Text

Technical Evaluation Rept on 'Submittal-Only' Review of IPEEE at Callaway Plant
	ML20212D292
Person / Time
Site:	Callaway
Issue date:	02/28/1999
From:	Kazarians M, Khatibrahbar, Sewel R; ENERGY RESEARCH GROUP, INC.
To:	; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20212D069	List: ML20212D071; ML20212D292;
References
	CON-NRC-04-94-050 ERI-NRC-95-510, NUDOCS 9909230062
	Download: ML20212D292 (62)
	v • d • e

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ERl/NRC 95-510

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i TECHNICAL EVALUATION REPORT ON THE

" SUBMITTAL-ONLY" REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS -

AT CAllRWAY PLANT 1 I;

I FINAL REPORT Completed: December 1996 Final: February 1998 Energy Research, Inc.

P.O. Box 2034 Rockville, Maryland 20847-2034 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

[2gg g 83 ENCLOSURE 3 -

- PDR _

ERI/h1C 95-5I0 TECHNICAL EVALUATION REPORT ON THE "SUBMITTAIrONLY" REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS AT CALLAWAY PLANT Completed: December 1996 Final: February 1998 M. Khatib-Rahbar Principal Investigator Authors:

R.T. Sewell, M. Kazarians , and M. Modarres2-Energy Research, Inc.

P.O. Box 2034 Rockville, Maryland 20847 i

Work Performed Under the Auspices of the United States Nuclear Regulatory Commission Office of Nuclear Regulatory Research 3 Washington, D.C. 20555 Contract No. 04-94-050 I

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' Kazarians and Associates. 425 East Colorado Street, Suite 545, Glendale CA 91205 2

University of Maryland. Department of Materials and Nuclear Engineering. College Park. MD 20742 J

TABLE OF CONTENTS EXECUTIVE SUMMAlt ................................. ..........v PREFACE .......................................... .. . . . . . . . . xi

ABBREVIATIONS ............. ........... .......... ....... ..... xii

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ........I 1.1 Plant Characterization ........... ............. ...... ......I 1.2 Overview of the Licensee's IPEEE Process and Important Insights . . . . . . . . . . . 2 1.2.1 Seismic . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...........2 {

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l.2.2 Fire..............................................3

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1.2.3 HFO Events . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

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l.3 ' Overview of Review Process and Activities . . .......................4 !

1.3.1 Seismic . . . . . . . . . . . . . . . . ......................... .5 .l 1.3.2 Fire . . . . . . . . . . . . . . . . . . . ..........................5 1.3.3 HFO Events .... .................. .. ............6

.2 ' CONTRACTOR REVIEW FINDINGS ........................ ........7 2.1 S eismic . . . . . . . . . . . . . . . . . . . . . . . .. . . .................... .7 2.1.1 Overview and Relevance of the Seismic IPEEE Process . . . . . . ......7 2.1.2 Success Paths and Component List .................... ...7 3 2.1.3 Non-Seismic Failures and Human Actions . . . . . . . . . . ..........8 l 2.1.4 Seismic Input . . . . . . . . . . . . . .........................9 !

2.1.5 Structural Responses and Component Demands . . . . . . . . . . . . . . . . . . 9 2.1.6 Screening Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ 10 2.1.7 Plant Walkdown Process .......... ..................... 11 l

2.1.8 Evaluation of Outliers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.1.9 Relay Chatter Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.1.10 Soil Failure Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.1.11 Containment Performance Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 14 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 ........ ......... .... .... .... 16 2.1.15 Other Safety Issues . . . . . ............................. 16

2.1.16 Peer Review Process . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 17 2.1.17 Summary Evaluation of Key Insights . . . . . . . . . . . . . . . . . . . . . . . . 17 2.2 Fire . ................................................ . 18 2.2.1 ' Overview and Relevance of the Fire IPEEE Process . . . . . . . . . . . . . . 18 2.2.2 - Review of Plant Information and Walkdown . . . . . . . . . . . . . ..... 19 l 2.2.3 Fire-Induced Initiating Events . . . . . . . . . . . . . . . . . . . . . . . .... 19 2.2.4 Screening of Fire Zones . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0 ;

2.2.5 Fire Hazard Analysis .. . . . . . . .......................... 20- {

2.2.6 . Fire Growth and Propagation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 ]

2.2.7 ' Evaluation of Component Fragilities and Failure Modes . . . . . . . . . . 22 l 2.2.8 Fire Detection and Suppression . . . . . . . . . .. . . . . . . . . . . . .. . . . . . 22 Energy Research, Inc. ii ERI/NRC 95-510 I i

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2.2.9 Analysis of Plant Systems and Sequences . . . . . . . . . . . .. .. . . 22 2.2.10 Fire Scenarios and Core Damage Frequency Evaluation . ....... 23 2.2.11 Analysis of Containment Performance . . . . . .. . .. .. ...... .. 23 2.2.12 Treatment of Fire Risk Scoping Study Issues . ................ 24 2.2.13 USI A-45 1ssue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2.3 HFO Events . . . . . . . . . . . . . . . . . . . . . . . . . ................... 25 2.3.1 High Winds and Tornadoes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.3.1.1 General Methodology . . . . . . . . . . . . . . . . . . . . . . . 26 2.3.1.2 Plant-Specific Hazard Data and Licensing Basis . . . .. 26 2.3.1.3 Significant Changes Since Issuance of the Operating License . . ... . . . . ... . . ... . ........ 26 2.3.1.4 Significant Findings and Plant-Unique Features ..... . 26 2.3.1.5 Hazard Frequency . . . . ............ . ..... 27 2.3.2 External Flooding . . . . . . . . . . . . ...................... . 27 2.3.2.1 General Methodology . . . . .............. ... 27 2.3.2.2 Plant-Specific Hazard Data and Licensing Basis ...... 27 2.3.2.3 . Significant Changes Since Issuance of the Operating License . . . . . . . . . ...................... 27 2.3.2.4 Significant Findings and Plant-Unique Features .. .. 27 2.3.2.5 Hazard Frequency ................ ....... 28 2.3.3 Transportation and Nearby Facility Accidents . .. . ... . . 28 2.3.3.1 General Methodology . . . . . . . . . ... . . . .. 28 2.3.3.2 Plant-Specific Hazard Data and Licensing Basis .... 28 2.3.3.3 Significant Changes Since Issuance of the Operating License ... ............................ 29 2.3.3.4 Significant Findings and Plant-Unique Features ...... 29 2.3.3.5 Hazard Frequency ........................ 29 2.4 Generic Safety Issues (GSI-147, GSI-148, and GSI-172) ................ 29 2.4.1 GSI-147, " Fire-Induced Alternate Shutdown / Control Panel Interaction" . 29 2.4.2 GSI-148, " Smoke Control and Manual Fire Fighting Effectiveness" . . . . 30 2.4.3 GSI-172, " Multiple System Responses Program (MSRP)" . . . . . . . . . . 30 3 OVERALL EVALUATION AND CONCLUSIONS . . . . . . ... ........ . 35 3.1 Seismic . . . . . . . . . . . . . . . . ............ ..... ....... . . 35 3.2 Fire . . . . . . ..... ..................... ... ........ . 36 3.3 HFO Events . . . . . . . . . . . . . . . . . . . . . . . .. .................. 37 4 IPEEE INSIGHTS, IMPROVEMENTS, AND COMMITMENTS . . . . . . . . . . . . . . . 38 4.1 S eismic . . . . . . . . . . . . . . . . . . . . . . . . . . ..................... 38 4.2 Fire . ..................... .... ............. .. .... 39 4.3 H FO Events . . , . . . . . . . . . . . . . . . . . . . . . . . . . .... ........... 39 5 IPEEE EVALUATION AND DATA

SUMMARY

SHEETS . . . . . .. ...... ... 43 6 REFERENCES............... . ... ............. ...... ... . 47 Energy Research, Inc. iii ERl/NRC 95-510

LIST OF TABLES Table 4.1 Open issues, and Their Resolutions, as Identified in the Callaway Seismic IPEEE . 40 Table 5.1 External Events Results . . . . . . . . . . . . . . . . . ................... 44 Table 5.2 SMM Seismic Fragility ................... ................. 45 Table 5.3 PWR Seismic Success Paths .... ............ ............. .. 46 Energy Research, Inc. iv ERI/NRC 95-510

EXECUTIVE SUMMAR This technical evaluation report (TER) documents a " submittal-only" review of the individual plant examination of external events (IPEEE) conducted for the Callaway Plant. This technical evaluation review was performed by Energy Research, Inc. (ERI) on behalf of the U.S. Nuclear Regulatory Commission (NRC). The submittal-only review process consists of the following tasks:

4 Examine and evaluate the licensee's IPEEE submittal and directly relevant available documentation.

Develop requests for additional informaticu (RAls) to supplement or clarify the licensee's IPEEE submittal, as necessary.

= Examine and evaluate the licensee's responses to RAIs.

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Conduct a final assessment of the strengths and weaknesses of the IPEEE submittal, and develop review conclusions.

This TER documents ERI's qualitative assessment of the Callaway IPEEE submittal, particularly with respect to the objectives described in Generic Letter (GL) 88-20, Supplement No. 4, and the guidance presented in NUREG-1407.

The Callaway IPEEE was conducted by Union Electric Company (Union Electric [UE]), with extensive contractor assistance. The CallawayIPEEE submittal considers seismic, fire, and high winds, floods and other (HFO) initiators for the external events analysis. The seismic IPEEE process was based on a focused-scope seismic margin assessment (SMA); the fire IPEEE was based on the Electric Power Research Institute's (EPRI's) fire-induced vulnerability evaluation (FIVE) methodology, with modifications and enh=aramank derived from findings and insights developed during EPRI-sponsored collaboration on fire analyses with UE and eight other utilities; and HFO events were evaluated using the screening approach from NUREG-1407 and GL 88-20, Supplement 4. The Callaway IFEEE was performed in accordance with quality assurance procedures. In addition, the submittal notes that all portions of the IPEEE received several levels of review.

The Callaway IPEEE submittal does not specify a freeze date for the modeling of plant configuration and operation. Documentation of the study was apparently completed in June 1995. Plant walkdowns were performed during the latter part of 1993.

Licensee's IPEEE Process With respect to seismic evaluation, NUREG-1407 assigns Callaway to the focused-scope review category.

UE elected to implement a focused-scope evaluation for the seismic IPEEE, following the EPRI seismic

_ margin methodology (SMM). In general, the seismic IPEEE addresses all major elements of concern for a focused-scope analysis, as identified by NUREG-1407. The study included an assessment of containment performance, examining Callaway's seismic capability to resist early containment failure. In addition, a relay chatter evaluation was performed. Soil-related evaluation of buried piping between the power-block and site-specific structures was also performed. A pre-screening at the design-basis level was conducted to elimmate most components from detailed screening and capacity assessment. High-confidence of low-Energy Research, Inc. v ERI/NRC 95-510 i

I probability of failure (HCLPF) capacities were calculated for only a small number of components. The seismic IPEEE process relied primarily on plant walkdowns that focused on evaluating component anchorage capability and the potential for elverse seismic-caused spatial interactions. Special area walkdown sheets (AWSs) were developed for the walkdown of pre-screened components. For the comparatively small number of components that could not be pre-screened, more detailed seismic j walkdowns were conducted using EPRI SMA procedures. Components required for successful integrity l of containment agamst early seismic-related failures were evaluated in a manner similar to components on i the safe shutdown equipment list (SSEL).

With regard to fire-related initiators, the licensee has conducted an extensive and detailed analysis of potential fire events at Callaway Plant. Several data bases and related documentation have been produced

to establish, and keep track of, fire-related plant features, including fire zones and areas. Logic models developed for the individual plant examination (IPE) were used in the fire analysis. Several extensive walkdowns of the plant have been conducted to support the analysis. The licensee has used the FIVE methodology for fire evaluation. The fire frequency and fire protection system data provided by EPRI were used for fire scenario quantification. The formulations in EPRI's FIVE model have been used to evaluate fire propagation, detection and suppression, and cable and equipment damage. Special attention has been given to human actions. For evaluation of redundant-train failure frequency, the IPE models of '

the plant have been used.

The Callaway HFO events IPEEE submittal is based primarily on the screening approach described in Supplement No. 4 to Generic Letter 88-20. The HFO portion of the overall IPEEE effort was performed (and reviewed) entirely by UE personnel, without the assistance of outside contractors. The examination process involved: (1) a review of plant-specific hazard data and plant licensing-basis information, considering the Callaway final safety analysis report (FSAR); and (2) implementation of a qualitative screening process. None of the HFO events was considered to be an important contributor to the potential ;

for induced severe-accidents, prunarily because the plant meets 1975 Standard Review Plan (SRP) criteria, and because no relevant changes have occurred since the time the plant operating license (OL) was issued. j High winds / tornadoes, external flooding, aircraft crashes, land transportation accidents, and nearby hazardous facility events have been discussed in the submittal, but these events were screened out in the IPEEE, since the plant conforms with 1975 SRP criteria.

Key IPEEE Hndings l In the Callaway seismic IPEEE, most SSEL components, and components required to prevent early containment failure, were pre-screened based on robust seismic design. Components that could neither be pre-screened, based on design consideratiom, nor screened out based on more detailed SMA walkdown procedures, were evaluated to have HCLPF peak ground acceleration (PGA) capacities of 0.3g. Hence, all components have been determined to be capable of_ withstanding the review-level earthquake (RLE),

. and the IPEEE submittal explicitly states that the overall HCLPF capacity for Callaway Plant is at least <

0.3g. (HCLPF assessments have been made with respect to the NUREG/CR-0098 median,5%-damped spectral shape, in accordance wii N guidelines of NUREG-1407). . Nonetheless, the focused-scope seismic evaluation for Callaway Plant has identified 21 open issues which required implementation of resolution approaches. A few minor fixes and four more noteworthy plant improvements have resulted ;

from such resolution. These safety enhancements have been made primarily with respect to a review of safe shutdown equipment; no improvements were found specifically or exclusively as a result of the containment performance evaluation.

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For the fire IPEEE, the licensee has concluded that there are no significant fire vulnerabilities at Callaway Plant. The total fire core damage frequency (CDF) from " unscreened" scenarios is estimated to be 8.9 x 104per reactor year (ry). The main contributors to the fire CDF are the control room and the two !

. emergency safety features (ESF) switchgear rooms.' The remainder of the compartments have been found i to be oflittle signi6cance to the overall fire risk. The IPEEE addressed fire risk for the control room and the cable spreadmg room. The possibility of forming a hot gas layer, and the possibility of fire and hot-

. gas propagation among adjacent compartments, have also been addressed.

- For HFO imtiators, the subnuttal notes that the plant meets 1975 SRP criteria and that, because no changes have occurred since the time of issuance of the plant OL, no HFO event is considered to have an important j potential to cause a severe reactor accident.

Generic Issues and Unresolved Safety issues For seismic events, the Callaway IPEEE specifically addressed the following additional issues: Unresolved i Safety Issue (USI) A45, " Shutdown Decay Heat Removal Reauirements"; Generic Issue (GI)-131, !

"Potantial Seismic Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Plants"; seismic-fire interaction aspects of the fire risk scoping study issues; USI A-17, " System

. Interactions in Nuclear Power Plants"; and USI A40, " Seismic Design Criteria". l For treatment of USI A45, the seismic IPEEE considered success paths that depend on seismic capability of the auxiliary feedwater (AFW) system and of bleed-and-feed cooling, as well as their support systems.

For treatment of GI-131, UE had earlier (in 1987) upgraded the hold-down assembly of the flux mapping system by increasing the size and strength of bolts and plates that comprise the assembly. With respect to seismic-fire interactions, the submittal concludes that it is unlikely that an inadvertent actuation of the fire protection system will occur as a result of the RLE, but if it were to occur, the result would be locahzed spraying of components not required to respond to the earthquake.- For USI A-17, the submittal states that closure is provided by the seismic capability walkdown, which addresses systems interactions, iWia: spatial, fire, and floodmg interactions. And, for USI A40, the submittal notes that the refueling water storage tank (RWST) was re-analyzed in 1990, and that an additional analysis of the RWST (for buckling concerns) was performed as part of the seismic IPEEE, for a RLE of 0.3g PGA. No vulnerabilities or concerns were noted with respect to any of the seismic-related GIs/USIs.

For the fire IPEEE, all of the Sandia fire risk scoping study issues, and USI A-45 issues, have been

addressed and resolved. For both cases, the licensee has dealt with the relevant concerns in detail, and has concluded that there are no outstanding problem areas. The licensee has presented a detailed discussion for each issue. Plant walkdowns have been undertaken to assist in resolution of these issues.

Seismic-fire interaction issues were addressed by examination of the potentials for an earthquake-induced

. fire event, for inadvertent seismic actuation of the fire suppression system and resu' ting adverse effects on safety equipment, and for seismically induced failure of the fire protection system. There are no areas in the plant where inadvertent acmarmn of fire suppression systems could lead to safety equipment damage that was not already considered in the fire analysis. Specific inspection and testing procedures have been instituted to verify the integrity of penetration seals, fire barriers, and fire dampers. Regarding USI A45, heat removal capabilities have been addressed in detail in the fire analysis via the IPE models used for

. CDF evaluation. The model gives credit to the possibility of bleed-and-feed cooling. It should be noted Energy Research, Inc.- vii ERl/NRC 95-510

that Thermo-Lag is employed at Callaway; however, with the exception of one fire area, the IPEEE fire analysis has not given any credit to the effectiveness of this material.

The submittal discusses the effects of the probable maximum precipitation (PMP) on the plant (GI-103).

The HFO IPEEE submittal does not describe the formal analysis of any other safety issues. It does,

. however, state that some generic and unresolved safety issues were addressed and are considered closed.

The submittal considers the following issues to be closed:

USI A-45, " Shutdown Decay Heat Removal Requirements"

USI A-17, " System Interactions in Nuclear Plants" GI-103, " Design for Probable Maximum Precipitation (PMP)"

Some information is also provided in the Callaway IPEEE submittal which pertains to generic safety issue (GSI)-147, GSI-148, and GSI-172. ]

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Vulnerabilities and Plant Improvements f

1 In the cover letter to the IPEEE submittal, the licensee draws the general conclusion that no plant-specific f

vulnerabilities (to potential severe accidents from external events) at Callaway Plant have been identified, i but opportunities for improvement have been considered. As noted below, such opportunities for !

improvement have related primarily to some enhancements and minor fixes to improve seismic safety, and {

to accident management considerations for fire events; no opportunities for improvement have been identified specifically in relation to HFO initiators.

Table 4.1 of this TER (derived from IPEEE Table 3.1.4-1) identifies the 21 open seismic issues for which the licensee is proposing resolution. In two cases, issues were resolved by demonstrating a HCLPF capacity in excess of the RLE. Other issues were resolved by further walkdowns, observations, judgments, and/or analyses. In addition, a few minor fixes and four more noteworthy enhancements were implemented to resolve other issues. From among the foregoing items, the seismic IPEEE submittal has drawn out the following improvements / findings as being most significant:

HCLPFResults The HCLPF capacity of component cooling water (CCW) heat exchangers anchorage was determined to be 0.41g PGA.

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The HCLPF capacity of diesel gererator (DG) control panels anchorage was determined to be i 0.49g PGA.

l Minor Fixes i Hand-held fire extinguishers have been remounted at heights not exceeding UL Standard drop heights. Hand-held fire extinguishers have been removed from the containment during normal operation; access it Cre suppression facilities is located near the containment hatch, for use by the fire brigade in responding to fires in the containment.

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. A floor grate, which crosses a building expansion gap behind the "A" DG starting air compressor skid, was clipped to the floor to prevent potential damage to DG sensing lines that pass through the grate.

Enhancements

- Several motor control centers (MCCs), whose top spray shields were found to be in contact with the adjacent walls, were bolted to the walls, to eliminate a spatial-interaction concern.

Missing shear pins in the foundations / supports of AFW pumps were installed.

Loose equipment found behind the control room was removed and signs were posted prohibiting further storage. ,

Chain hoists located in the emergency core cooling system (ECCS) pump rooms, the CCW pump rooms, and the AFW pump rooms, found to be in close proximity to their associated pumps, were moved to safe storage locations at the end of their rooms, and procedural guidance and training was instituted to prevent recurrence of this issue.

The submittal indicates that the noted fixes and plant enhancements have either been implemented or are planned for implementation.

No fire-related vulnerabilities have been identified from the IPEEE. No improvements and commitments were deemed necessary to further re6 ace the fire risk at Callaway Plant. The licensee has used the Nuclear Energy Institute's (NEI's) severe accident closure guidelines (NEI-91-04), to evaluate the need for plant improvements. Based on consideration of the existing plant features and operational practices (i.e.,

procedures and policies), the licensee has concluded that " Severe Accident Management Guidance" (SAMG) need not be put in place for those fire scenarios that exceed the guideline threshold value.

However, the licensee has elected to undertake analysis of the responses to spurious actuations which may result from a fire in the control room or either of the two ESF switchgear rooms.

The IPEEE for Ca'laway Plant has identified no vulnerabilities to any severe accident risk from HFO initiators. Since the HFO events contribution to the plant's CDF is believed to be small, the licensee has not considered any plant-specific improvements. Hence, no fixes or commitments related to HFO events are planned.

Observations j The seismic IPEEE of Callaway Plant has addressed the major elements specified in NUREG-1407 as recommended items that should be considered for a focused-scope evaluation. The submittal itself is considered to be well written and structured, and the identification and implementation of safety enhancements is judged to have produced a meaningful response to GL 88-20, for a focused-scope plant.

Only a few minor weaknesses of the licensee's seismic IPEEE submittal have been noted in this review.

. For the evaluation of fire initiators, the licensee has expenosi considerable effort on its IPEEE. The licensee has employed proper methodology, i.e., the EPRI FIVB methodology, and has employed proper data bases for fire occurrence and suppression system failure rates. The fire IPEEE submittal is well Energy Research, Inc. ix ERI/NRC 95-510

' written, with a clear and well orgamzed presentation. Tables and figures provide considerable supporting l

information for the analysis and conclusions. The final conclusions are reasonable, and are within the range of results expected for a pressurized water reactor (PWR). Clearly, the licensee has gained useful ,

insights from the task of inspecting every part of the plarit for potential fire vulnerabilities.

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i The HFO events portion of the IPEEE of Callaway Plant shows that the licensee has gained a qualitative j understandmg of the effects of high winds, external floods, and transportation events on potential severe-accident behavior of the plant. Since the HFO portion of the IPEEE submittal has been carried out by the licensee, without significant reliance on outside contracted assistance, the licensee itself has gained insights )

with respect to potential vulnerabilities due to HFO events affecting the Callaway site.

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PREFACE The Energy Research, Inc., team members responsible for the present IPEEE review documented herein, include:

Seismic R. Sewell Bre M. Kazarians Hieh Winds. Floods and Other External Events M. Modarres Review Oversieht. Coordination and Intevration M. Khatib-Rahbar, Principal Investigator, Report Review A. Kuritzky, IPEEE Review Coordination and Integration R. Sewell, Report Integration Dr. John Lambright of Larnbright Technical Associates, contributed to the preparation of Section 2.4 following the completion of the draft 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 NRC staff is acknowledged.

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ABBREVIATIONS AFW Auxiliary Feed Water AWS. Area Walkdown Sheets CCDP Conditional Core Damage Probability

-CCP Centrifugal Charging Pump CCW Component Cooling Water CDF Core Damage Frequency CDFM Conservative Determuustic Failure Margin CFR Code of Federal Regulations CRDM Control Rod Drive Mechanism CST Condensate Storage Tank DG Diesel Generator l

.ECCS Emergency Core Cooling System EOP. Emergency Operating Procedure EPRI Electric Power Research Institute ERI Energy Research, Inc. 1 ESF Emergency Safety Features ESFAS Engineered Safety Features Actuation System i ESW Essential Service Water i FCIA Fire Compartment Interaction Analysis I I

- FIVE Fire Induced Vulnerability Evaluation Method FSAR Final Safety Analysis Report GI Generic Issue

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GL Generic Letter GSI Generic Safety Issue HCLPF High Confidence of Low Probability of Failure (capacity)

HFO High Winds, Floods and Other (external initiators)

IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IRS In-Structure Response Spectrum LLNL Lawrence Livermore National Laboratory LOCA Loss of Coolant Accident

-LOSP Loss of Offsite Power MCA. Multi-Compartment Interaction Analysis MCC Motor Control Center MSFIS - Main Safety Features Instrumentation System

-MSL Mean Sea Level NEI - Nuclear Energy Institute NRC Nuclear Regulatory Commission

,c NSAC - Nuclear Safety Analysis Center

.NSSS- Nuclear Steam Supply System OBE Operating Basis Earthquake OL Operating License

'PGA Peak Ground Acceleration P&ID Piping and Instrument Diagrams

- PME . Probable Maximum Flood

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I PMP Probable Maximum Precipitation PORV . Power Operated Relief Valve PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor -

RAI Request for AdditionalInformation RCP Reactor Coolant Plunp RCS Reactor Coolant System RHR Residual Heat Removal RLE Review Level Earthquake RWST Refueling Water Storage Tank SAMG Severe Accident Management Guideline SCBA Self-Contained Breathing Aparatus SEWS Seismic Evaluation Work Sheet SG Steam Generator SI Safety Injection SIP Safety Injection Pump SLOCA Small-break Loss of Coolant Accident SMA Seismic Margin Assessment SME Seismic Margin Earthquake SMM Seismic Margin Methodology SNUPPS Standardized Nuclear Unit Power Plant Sites SPLD Success Path Logic Diagram SQUG Seismic Qualification Utility Group SRP Standard Review Plan SRT Seismic Review Team SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List SSI Soil-Structure Interaction TER Technical Evaluation Report UE- Union Electric Company UHS Ultimate Heat Sink USI Unresolved Safety Issue ZPA Zero Period Acceleration Energy Research, Inc. xiii ERI/NRC 95-510

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1 INTRODUCTION i- This technical evaluation report (TER) documents the results of the " submittal-only" review of the individual plant examination of external events (IPEEE) for the Callaway Plant [1]. This technical evaluation review, conducted by Energy Research, Inc. (ERI), has considered various external initiators, including seisme events; fires; and high winds, floods, and other (HFO) external events.

The U.S. Nuclear Regulatory Commission (NRC) objective for this review is to determine the extent to 1

which the IPEEE process used by the licensee, Union Electric Company (Union Electric [UE]), meets the '

intent of Generic Letter (GL) 88-20, Supplement No. 4 [2]. Insights gained from the ERI review of the i

IPEEE submittal are intended to provide a reliable perspective that assists in making such a determination.

This review involves a qualitative evaluation of the licensee's IPEEE submittal, development of requests ,

for additional information (RAls), evaluation of the licensee responses to these RAIs, and finalization of J

the TER.

The emphasis of this review is on describing the strengths and weaknesses of the IPEEE submittal, particularly 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, fire, and HFO events sections of the Callaway Plant IPEEE. Sections 2.1 to 2.3 of this report 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, ano HFO events reviews, respectively. Section 4. summarizes the IPEEE insights, improvements, and licensee

- commitments. Section 5 includes completed IPEEE data summary and entry sheets. Fin Aly, Section 6 provides a list of the references cited in the TER.

1.1 Plant Charactarization i

Callaway is a single-unit nuclear power generatmg facility consisting of a 4-loop Westinghouse pressurized !

water reactor (PWR). The licensed core power level of Callaway is 3,565 MWt. The containment

' building at Callaway is a large, dry containment structure designed by Bechtel Power Corporation. The

' containment walls and dome are steel-lined, reinforced, post-tensioned concrete founded on a containment

- base slab of reinforced, steel-lined concrete. The plant is located in Callaway County, Missouri, approximately 80 miles west of the St. Louis Metropolitan Area. The plant site consists of 2,767 acres located on a plateau approximately 300 feet above the Missouri River (which is about 5 miles to the south of the plant). The plant was licensed for full-power operation in October 1984, and the plant began

. commercial operation in April 1985.

Callaway is a sister unit to Wolf Creek Generating Station; both are among four units (Callaway, Sterling, Tyrone and Wolf Creek) that were planned for development under the Standardized Nuclear O' nit Power Plant Sites (SNUPPS) concept, where power-block components are designed to' site-independent demands determined from envelopes of individual demands for the four sites. Non-power-block structures and components are designed based on site-dependent demands. Power-block structures (i.e., the standard Energy Research, Inc. 1 ERI/NRC 95-510

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portions of the plant) at Callaway include the following: reactor containment building, auxiliary building, turbine building, fuel building, rad-waae building, control building, storage tanks (condensate storage tank

[ CST), refueling water storage tank [RWST], make-up water storage tank, demineralized water storage tank, and emergency fuel oil storage tank), diesel generator building, and transformer vaults. Non-power-block structures (i.e., the site-specific portions of the plant) at Callaway consist of: circulating wster cooling tower and pumphouse, essential service water (ESW) pumphouse, ultimate heat sink (UHS) raandan pond and cooling tower, the general yard area (including administrative buildings and technical support center, fire water storage tanks and pumphouse, etc.), the switchyard and offsite power sources, and the river intake structure.

The safe shutdown earthquake (SSE) for Callaway is different for seismic Category I power-block structures and equipment (designed based on SNUPPS considerations) and for seismic Category I non-power-block structures and equipment (designed based on site-specific considerations). The SSE peak f

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ground acceleration (PGA) for both horizontal and vertical motion is 0.25g for power-block structures and 0.20g for non-power-block structures; operating basis earthquake (OBE) design accelerations are one-half the corresponding SSE accelerations. 'Ihe design spectral shape in all cases is the Regulatory Guide (R.G.)

1.60 spectrum. Some of the Callaway seismic Category I structures are founded directly on bedrock (Pennsylvanian Graydon chert conglomerate), whereas some are founded on compacted structural fill and {

stabilized backfill. The depth of fill material over bedrock ranges from 19 feet to 54 feet, depending upon

' the particular structure. l j

1.2 Overview of the IJeansee's IPEEE Process and Important Insights I 1.2.1 Seismic l l

NUREG-1407 assigns Callaway plant to the 0.3g focused-scope seismic review category. The Callaway l seismic IPEEE is a focused-scope evaluation that implements the Electric Power Research Institute's (EPRI's) seismic margins assessment (SMA) methodology. The evaluation approach, however, departs somewhat from an EPRI SMA due to a unique pre-screening activity that was implemented. The purpose of the pre-screemng, as noted in the submittal, has been to take advantage of the robust seismic design of Callaway, to justify reducing the number of plant components for which a more detailed, conventional seismic margins evaluation is applied. Of the 592 components considered on the safe shutdown equipment list (SSEL) and the containment systems equipment list addressed in the IPEEE, only 22 were not pre-screened. The pre-screened components were physically examined by means of area walkdowns that focused on inspecting anchorage and interaction hazards, as well as on verifying the appropriateness of pre-screening.

A substantially unique feature of the design of Callaway Plant, which has impacts on the seismic IPEEE

_ (i.e., relative to the justification for pre-screening), relates to the SNUPPS concept, where power-block campanents are designed to site-independent demands determined from envelopes of demands developed' for the four SNUPPS sites. The SSE for power-block structures is defined by a Regulatory Guide (R.G.)

> 1.60 [4] spectral shape anchored to a PGA value of 0.25g. For non-power-block structures, seismic demands are based on individual plant responses (not envelopes of responses); the SSE for non-power-

' block structures at Callaway Plant is a R.G.1.60 spectrum anchored to a PGA value of 0.20g. Of significance is the fact that the design-basis spectral shape for Callaway is more severe than the spectral shape defining the seismic margin earthquake (SME), and yet, the PGA for the seismic margin earthquake

. (0.3g) is only somewhat greater than the design-basis PGA of 0.25g (for power-block structures). Hence, Energy Research, Inc.- 2 ERI/NRC 95-510

for power-block components, the design-basis seismic demands in many cases exceed the demands imposed by the SME. The licensee's pre-screening approach was designed to take advantage of this situation, which affects the evaluation of many components, since most components required for safe shutdown lie within the power block.

In addition to pre-screening, the IPEEE submittal indicates that the following steps were undertaken for the focused-scope seismic evaluation of Callaway Plant:

Project Planning Selection of SME/ Review Level Earthquake (RLE)

Selection of the Seismic Review Team (SRT)

Preparatory Work Prior to Walkdown-

Systems and Element Selection Walkdown

Seismic Capability Walkdown

Subsequent Walkdowns (As Needed)

SMA Work (HCLPF Calculations and Relay Chatter Evaluation)

Containment Performance Analysis Consideration of Unresolved Safety Issue (USI) A-45, Generic Issue (GI)-131, USI A-17 and USI A-40

Analysis of Seismic-Fire Interactions Resolution of Outliers

Peer Review

Documentation The seismic IPEEE submittal concludes that there are no vulnerabilities at Callaway Plant. Twenty-one (21) open issues have been identified and resolved through the seismic IPEEE. These resolutions have involved some plant hardware changes, minor fixes, and procedural implementations.

The submittal notes that Callaway was found to be a very seismically resistant plant, with most structures

- and equipment being over-qualified as a result of the SNUPPS enveloping of demands for power-block components. The study concludes that the plant has a high-confidence of low-probability of failure (HCLPF) capacity of at least 0.3g, and thus, meets the RLE requirements.

1.2.2 Fire The licensee has conducted an extensive and detailed analysis of potential fire events at Callaway Plant.

Several data bases and related documentation have been produced to establish, and keep track of, fire-related plant features, including fire zones and areas. Logic models developed for the individual plant examination (IPE) were used in the fire analysis Several extensive walkdowns of the plant have been .

conducted to support the analysis. The licensee has used the fire-induced vulnerability evaluation (FIVE)

- methodology for the fire assessment. The fire frequency and fire protection system data provided by EPRI were used for fire scenario quantification. The formulations in EPRI's FIVE model have been used to evaluate fire propagation, detection and suppression, and cable and equipment damage. Special attention has been given to human actions. For evaluation of redundant-train failure frequency, the IPE models of the plant have been used.

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' Multi-stage screening and detailed analyses have been employed to identify the dominant contributors to core damage frequency (CDF). The screening is based on deterministic factors and on a CDF threshold 4

of 10 per reactor year (ry). The licensee has assessed the overall fire CDF for " unscreened" scenarios 4

to be 8.9 x 10 /ry. The main contributors to this fire CDF are the control room and the two emergency safety features (ESP) switchgear rooms. The licensee has addressed the potential for fires in the control room and cable spreadmg room of the plant. The possibility of forming a hot gas layer, and the possibility of fire and hot-gas propagation among adjacent compartments, have been addressed.

The licensee has addressed Sandia's fire risk scoping study issues and USI A-45 issues. For both cases, the licensee's study has dealt with the relevant concerns, and has not revealed any outstanding problem areas.

The licensee has concluded that there are no significant fire vulnerabilities at Callaway Plant. The licensee has used the Nuclear Energy Institute's (NEI's) severe accident closure guidelines (NEI-91-04) to evaluate the need for plant improvements. Based on the existing plant features and operational practices (i.e.,

procedures and policies), the licensee has concluded that " Severe Accident Management Guidance" (SAMG) need not be put in place for those fire scenarios that exceed the guideline threshold value.

However, the licensee has elected to undertake analysis of the responses to spurious actuations which may result from a fire in the control room or either of the two ESF switchgear rooms.

1.2.3 HFO Events-The licensee has conducted a rather minimal analysis for HFO events. Since the Callaway plant was designed and constructed in compliance with the 1975 Standard Review Plan (SRP) criteria, the HFO l events were all screened out. Accordingly, the Callaway Plant IPEEE submittal finds no unduly significant sequences (i.e., vulnerabilities) with respect to HFO events.

1.3 cw: of Review Prams and Aedvities In its quahtative review of the Callaway IPEEE, ERI focused on the study's completeness in reference to NUREG-1407 guidance; its ability to achieve the intent and objectives of GL 88-20, Supplement No. 4; its strengths and wannann with respect to the state-of-the-art; and the robustness ofits conclusions. This review did not emphasize confirmanon of numerical accuracy of submittal results; however, any numerical errors that were obvious to the reviewers are noted in the review findings. The review process included the following major activities:

Completely examine the IPEEE and related documents j

Develop a preliminary TER and RAls

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 imponant 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 Callaway Plant.

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, e' 1.3.1 Seismic In conducting the seismic review, ERI generally followed the emphasis and guidelines described in the report, Individual Plant Examination ofExternal Events: Review Guidance [5], for review of a seismic margin assessment, and the guidance provided in the NRC report, IPEEE Step 1 Review Guidance Document [6]. In addition, on the basis of the Callaway IPEEE submittal, ERI completed data entry tables developed in the Lawrence Livermore National Laboratory (LLNL) document entitled "IPEEE Database Data Entry Sheet Package" [7].

In its seismic review of the Callaway IPEEE, ERI examined the following documents:

Sections 1, 2, 3, 4.3.5.1, 6, 7, and 8 of the IPEEE submittal for Callaway Plant [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 IPEEE?

Was the development of success paths performed in a manner consistent with prescribed practices?

Were random and human failures properly considered in such development?

e 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?

Were capacity calculations performed for a meanmgful set of components, and are the capacity results reasonable?

Does the submittal's discussion of qualitative assessments (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 risk outliers been addressed?

1.3.'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, 5, 6, 7, atx! 8 of Reference

[1], and on the licensee responses to fire < elated RAIs [8] The guidance provided in References [5,6] 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].

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The process implemented for ERI's review of the fire IPEEE included an examination of the licensee's methodology, relevant data, and results. ERI reviewed the methodology for consistency with currently accepted and state-of-the-art methods, paying special attention to the screening methodology and to the procedure used for estimating the frequency of occurrence of a fire scenario, to ensure that no fire scenarios were prematurely eliminated. The data element of a fire IPEEE includes, among others, such items as:

a cable routing fire zone / area partitioning a

fire occurrence frequencies e event sequences e

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 probabilistic risk assessments (PRAs). For a few fire zones / areas that were deemed important, ERI also verified the logical development of the screening justifications / arguments (especially in the case of fire-zone screening) and the computations for fire occurrence and CDF.

1.3.3 HFO Events The review process for HFO events closely followed the guidance provided in the report entitled IPEEE Step 1 ReWew 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 SRP conformance was appropriately executed. In addition, the validity of the licensee's conclusions, in consideration of the results reponed in the IPEEE submittal, was assessed.

Also, in some instances, computations of frequencies of occurrence of hazards, fragility values, and failure probabilities were spot checked. Review team experience was relied upon to evaluate the reasonableness l

of the licensee's evaluation. ;

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2 CONTRACTOR REVIEW FINDINGS 2.1 Seismic A summary of the licensee'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

a. Seismic Review Category and RLE Callaway Plant is assigned, in NUREG-1407, to the focused-scope seismic review category. The review level earthquake (RLE) is described by the NUREG/CR-0098 [9] median spectral shape anchored to a PGA value of 0.3g.

b. Seismic IPEEE Process UE has elected to implement a focused-scope evaluation, following the EPRI seismic margins methodology, for conducting the seismic IPEEE of Callaway Plant.

Callaway Plant is a single-unit PWR. Thus, no issues pertaining to multiple units have needed to be addressed in the seismic IPEEE.

c. Renew Findings In general, the seismic IPEEE addresses all major elements of concern for a focused-scope analysis, as identified by NUREG-1407. The study has included an assessment of containment performance, examining Callaway's seismic capability to resist early containment failure. In addition, a relay chatter evaluation was performed. Other than evaluation of buried piping between the power-block and site-specific structures, no soil failure analysis was performed. A pre-screening comparing RLE demands against design-basis demands was conducted to eliminate most components from detailed screening and capacity assessment. HCLPF capacities were calculated for only a few components, primarily because the plant seismic design basis (of 0.25g for power-block components, and 0.20g for non-power-block components) is not drastically different from the SME/RLE of 0.3g.

The approach implemented for the Callaway seismic IPEEE is thus relevant to the evaluation of seismic severe accident resistance, and is consistent with the format requested by NUREG-1407, with the modifications indicated in NRC's Supplement No. 5 to GL 88-20.

2.1.2 Success Paths and Component List For the seismic IPEEE of Callaway, UE has 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The SPLD identifies both preferred and alternate success paths. Reactivity control, reactor coolant system (RCS) pressure control, RCS inventory control, and decay heat removal are the essential " core protection safety functions" that have been addressed in the development of success paths. The preferred success path Energy Research, Inc. 7 ERI/NRC 95-510 l

. a relies on high-head safety injection (SI), via the charging or safety injection pumps, with secondary side heat removal via the auxiliary feedwater (AFW) system and steam generator power-operated relief valves (PORVs). The alternate success path also relies on high-head safety injection, via the safety injection pumps or charging pumps, with bleed and feed cooling for decay heat removal. In both success paths, reactor trip is the means relied upon for reactivity control, and residual heat removal (RHR) recirculation is required for coolant recirculation.

The success paths chosen for safe shutdown of Callaway, following a seismic margin earthquake, assume both loss of offsite power (LOSP) conditions and small break loss of coolant accident (LOCA) conditions (including reactor coolant pump [RCP] seal leakage). The IPE event trees for small break LOCA (SLOCA) events were used in developing the seismic success paths.

Once the success paths were determmed, the frontline and support systems that comprise these paths were identified. The submittal presents a large list of frontline systems, support systems, and primary containment safeguard systems that are needed to fully ensure integrity of the various elements of the success paths. System piping and instrumentation diagrams (P& ids) and emergency operating procedures (EOPs)~ were consulted to identify specific components and instrumentation that comprise the safe shutdown equipment list. The " rule-of-the-box" approach [10] was used in delineating separate components. A list of 592 SSEL components was developed from the systems analysis.

The submittal mentions that the Callaway SMA has involved a systems and element selection (success paths) walkdown, yet a general discussion of that walkdown is not provided. The submittal notes that the SRT added steam generator (SG) blowdown isolation valves and diesel generator (DG) exhaust silencers to the initial SSEL.

The submatal further mentions that components pertaining to contamment safeguard systems (contamment spray, cooling and isolation systems) were also included in the SSEL. However, the submittal does not provide a complete discussion of the success criteria for developing the containment systems equipment list (includmg an explicitly stated definition of successful containment performance).

Overall, the approach described in the Callaway seismic IPEEE for defining success paths and developing a safe shutdown equipment list and containment systems equipment list appears to be reasonable and consistent with the guidelines presented in NUREG-1407.

2.1.3 Non-Seismic Failures and Human Actions The Callaway seismic IPEEE implemented the approach of NUREG/CR-5679 [11J for screening out low availability mmponents and operator actions. The following screemng thresholds on failure rate have been

- used:

. 0.01 if the failure leads to loss of only one train in one system 0.001 if the failure leads to loss of all trains in one system 0.001 if the failure leads to loss of one train in multiple systems An item having a failure rate greater than the threshold value was not screened out, unless that item had a sufficiently high probability of recovery so as to make the resulting failure rate less than the threshold.

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The existing Callaway PRA, with modifications to the small LOCA event tree for seismic effects, was used in identifying non-seismic failures and human actions, their importance to plant failure cutsets, and their failure rates. Twenty (20) hems, involving single failures of components or human actions, were identified as being important to the integrity of success paths. All of these items, except one, were screened out based on having a failure rate below the appropriate threshold value. . With a rate of 0.00164, operator failure to iniriara flow to the RHR heat exchangers (i.e., failure to align component cooling water (CCW) to the RHR heat exchangers), did not screen out when conra ed to the 0.001 threshold. However, the licensee has considered this failure rate acceptable, citing the following three reasons in the submittal:

1. This operator requirement is clearly documented in EOPs and is a part of the normal shutdown process. This action is thoroughly understood and routinely practiced by the operators.

2. This action would not be required until several hours after an earthquake.

-3. The failure rate for this action is very close to the screening value.

The submittal concludes that tlicre are no non-seismic failures or human errors that result in low availability of systems or components relied upon in the SMA success paths.

Overall, the licensee's analysis of non-seismic failures and human actions is generally considered to be well executed. However, of the 20 identified potential failures, only four are operator actions. The submittal does not comment on how the earthquake itself might affect operator actions and failure rates.

Nonetheless, the licensee's treatment of non-seismic failures and human actions is judged to be generally reasonable and consistent with the requested guidelines of NUREG-1407. ,

2.1.4 ' Seismic Input The Callaway seismic IPEEE has used the NUREG/CR-0098 median 5%-damped spectrum, anchored to the RLE PGA value of 0.3g, to define seismic margin canhquake demandc. For pre-screening pmposes, these RLE demands have been compared to SSE demands. For normal screening, EPRI screening tables that are consistent with the RLE demands have been used. Anchorage evaluations of non-pre-screened items have been based on RLE demands. For HCLPF calculations, the RLE demands have also been used.

Even though some Callaway structures are founded on rock, the more severe soil spectrum of NUREG/CR-0098 has been used for all assessments.

Thus, the seismic input selected for use in the Callaway IPEEE is consistent with the relevant guidelines presented ni NUREG-1407.

2.1.5 Structural Responses and Component Demands

a. Owrall Approach In the Callaway seismic IPEEE, structural respons~es and component demands associated with the seismic margin earthquake were obtamed simply, by linearly scaling existing design-basis responses and demands.

. For example, design-basis in-stmeture response spectra (IRS) were scaled by the ratio of the RLE spectral acceleradon to the SSE at a building's dominant natural frequency, in order to obtain the IRS for seismic margin evaluation (see EPRI NP-6041-SL [10]).

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b. StructuralModels

.The submittal does not provide a specific description of the structural models and parameters used to

generate the original (design-basis) structural responses and component demands.

c. Soil Structure Interaction Analysis and Results

The submittal. states that the SNUPPS seismic analysis (for power-block structures) was originally

_ performed using the program FLUSH for soil-structure interaction (SSI) modeling based on the finite element method. No other details of tk SSI analysis are provided,

d. Review Findings Existing Final Safety Analysis Report (FSAR) structural responses were scaled to define structural demands, and existing FSAR in-structure spectra were scaled to define component demands. No new stmetural models were developed, nor new in-structure response spectra generated. This development of the structural responses and component demands used in the Callaway IPEEE is consistent with the relevant guidelines presented in NUREG-1407, 2.1.6 Screening Criteria The seismic IPEEE of Callaway Plant has involved both a unique pre-screening process and a conventional screening process.

a. Pre-Screening Pre-sc.dag has been based on comparing RLE demands (i.e., IRS values derived from the RLE input motion) to corresponding SSE demands (i.e., IRS values derived from the SSE input motion). In many cases for the power-block components, the SSE dernands were found to exceed the corresponding RLE damands. In such cases, components were pre-screened from a detailed seismic margins evaluation, yet still received a meaningful walkdown. It is noted that some power-block components could not be pre-screened. For non-power-block components, the RLE demands were found to always exceed correspondmg SSE demands, and hence, none of the non-power-block components could be pre-screened.

(Note that this process for pre-screemng does not necessarily ensure that the HCLPF will exceed the 0.3g RLE for pre-screened components),

b. ConwnrionalScreening For the 22 components which could not be pre-screened (out of 592 total SSEL components), the screening

. criteria and guidelines contained in EPRI NP-6041-SL were used as the basis for conducting additional i screening of components and stmetures for the Callaway seismic IPEEE. Components and structures were checked against the EPRI SMA (seismic margin assessment) screening caveats, in addition to being evaluated for anchorage adequacy and for potential seismic interaction concerns. The screening column applicable to' spectral accelerations between the limits of 0.8g and 1.2g was apparently used (Tables 2-3 ar.d 2-4 of EPRI NP-604;, Rev.1).

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c. Review Findings A set of pre-screening criteria were also applied to SSEL itans, resulting in the vast majority of components being screened out prior to the conventional SMA scraning. This pre-screening was based on a comparison of RLE in-structure spectra (using SMA parameters and criteria) with SSE in-structure i spectra (using FSAR criteria). With one marginal exception, the RLE versus SSE comparisom were performed in a manner so as to ensure a conservative approach to pre-screening. The comparisons can not absolutely ensure that the pre-screened components will have HCLPF capacities in excess of the RLE; however, it can be said that the approach gives confident assurance that such will be the case.

For items that were not pre-screened, the Callaway seismic IPEEE made use of the appropriate screening tables, criteria and procedures discussed in EPRI NP4041, Rev.1. Caveats of the seismic margin screening tables were used in the screenmg process for structures, as well as for electrical and mechanical

. equipment.' Anchorage and spatial interaction concerns were addressed in the walkdown(s).

The screemng criteria and procedures, and pre-screening approach, used in the Callaway seismic IPEEE are consistent with NUREG-1407 guidelines, and are judged to be appropriate for screening SSEL components.

2.1.7 Plant Walkdown Process

a. Preparatory Work Pre-walkdown preparatory work included gathering and reviewing plant design, qualification, and operational data, and performing preliminary demand and capacity calculatiom.

. b. Systems and Element Selection Walkdown A systems and element selection walkdown was performed, however, details of this aspect of the walkdown are not described in the submittal.

c. Seismic Capability Walkdoun Pre-screened components were walked down using a simplified procedure. Area walkdown sheets (AWSs) I were developed for the pre-screened components. For these components, the seismic review team (SRT) was aware of the significant seismic margin, and hence, the focus of the walkdown was on investigation of obvious anchorage and/or interaction problems. Development of the AWSs ensured that the walkdown i

was conducted in a highly efficient manner. ;

For the comparatively small number of components that were not pre-screened, a more detailed walkdown was executed, and seismic evaluation work sheets (SEWSs) were completed by the SRT. The SRT physically reviewed important attributes (include those pertaining to SMA screening caveats) of these ,

comy aents, according to procedures described in EPRI NP-6041-SL, again with particular emphasis on anchorage of equipment and systems interaction.

J The seismic walkdown process included the following elements: evaluation of SSEL components, !

investigation of seismic-fire interaction concerns, evaluation of potential seismically induced flooding / spray Energy Research, Inc. 1I ERl/NRC 95-510 !

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problems, information gathering and evaluation for resolution of generic issues and unresolved safety issues, and evaluation of components needed to prevent early failure of containment functions. For SSEL components, the SRT inspected the interiors of motor control centers, instrument and control panels, load centers, switchgear, etc., to verify adequate internal mounting of equipment, especially relays. Spot-checks were performed of piping, cable trays, manual valves, small filters, electrical penetration i assemblies, and self-actuated check valves. General seismic housekeeping issues were also addressed.

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d. Subsequent Walkdons Information was collected during the walkdown for cases where subsequent evaluation was deemed necessary. In some instances, post-walkdown evaluations were conducted before the SRT signed-off on worksheets. A second, confirmatory seismic capability walkdown was judged to be unnecessary. ;

e. Heatment ofinaccessible Components Only a few components listed on AWSs and SEWSs were not accessible for inspection as part of the i seismic walkdown. For these components, the evaluation was based on a review of existing drawings, I qualification packages, and/or photographs. l 1

JU Walkdon Duration and kaining ofSeismic Review Team (SRT)

' The seismic walkdowns were performed by a single SRT consisting of three trained members: a seismic capability engineer (EQE International); a systems engineer (NUS Corporation); and a third participant !

(a -Union Electric engineer) having the dual role of seismic capability and systems engineer. The I walkdown duration was approximately two weeks.

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The seismic capacity walkdown focused on evaluation of potential anchorage problems and spatial interaction problems for all components. Non-pre-screened components were visually inspected in detail, and were walked down using EPRI NP-6041 procedures (using SEWSs). Pre-screened components were walked down with the knowledge that high assurance of seismic margin existed; less-detailed AWSs were completed for these components. The submittal also makes note of specific areas and/or components that i were inaccessible due to high radiation / contamination, and indicates that documentation reviews were !

conducted in place of physical walkdowns.

The walkdown process conducted for the seismic IPEEE of Callaway is judged to be reasonable and appropriate for a focused-scope evaluation, and is capable of identifying outliers with respect to safe ;

shutdown and containment integrity following a seismic margin eanhquake. !

2.1.8 Evaluation of Outliers I

a. Owrall Approach l

e A number of outliers (termed "open issues" in the submittal) were identified as a result of the seismic capability walkdown. These items were resolved by one or more of the following methods: judgmental i evaluapon based on review of existing analyses and data, HCLPF calculation, implementation of a minor l l

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r fix, implementation of a significant hardware change, implementation of a maintenance action, or implementation of a new procedure.

b. HCLPF Calculations Two items were evaluated by means of HCLPF calculations, using the conservative de*erministic failure I

margins (CDFM) approach: (1) the CCW heat exchanger pedestal supports (i.e., with respect to identified l low design margin of anchor bolts); and (2) the diesel generator control panel anchorages (i.e., for j overhanging of panels with respect to the concrete base pad and embedment channels). HCLPF capacities I in excess of the RLE (0.3g PGA) were evaluated for these noted outliers. I

c. Other Calculations i 1

A deterministic evaluation of the buckling capacity of the refueling water storage tank (RWST) was performed, showing that the design-basis capacity exceeds the RLE demand. The submittal does not indicate that evaluations have been performed for other flat-bottomed tanks. Similarly, for concrete masonry unit walls, a screening analysis was conducted to show that the RLE demands were less than the design accelerations for such walls located in the vicinity of SSEL components. No HCLPF calculations ;

were performed for flat-bottomed tanks or masonry walls.

d. Rehw Findings l The seismic IPEEE submittal for Callaway Plant generally provides a very meaningful evaluation of identified outliers, i 2.1.9 Relay Chatter Evaluation Through a review of available documentation and data, the licensee has identified low-ruggedness relays in use at Callaway Plant in both Class IE and non-Class IE circuits. The licensee has identified the applications of these relays and the expected significance of relay chatter. The submittal does not indicate that all bad actor relays have been physically located by in.specting all relevant circuits. In other words, the possibility may exist that the use of low-ruggedness relays is not adequately or accurately captured in available schematics and documentation. The SRT performed spot-checks of various panels and cabinets, however, to verify relay types and ruggedness of relay mountings.

The documentation based relay evaluation identified instances of potentially anomalous plant behavior caused by relay chatter. In general, the evaluation concludes that relay chatter is acceptable, and does not represent a vulnerability with respect to the ability to safely shut down the plant. The submittal also makes note of vulnerable relays that may lead to spurious actuation of fire pumps in an eanhquake, which could lead to localized flooding / spraying of equipment.

No actions have been proposed to replace low-ruggedness relays.

In general, it can be concluded that a substantially comprehensive and well-structured relay chatter evaluation was conducted for the Callaway seismic IPEEE. UE personnel performed the relay evaluation, with waltdown assistance provided by contractors. Low seismic ruggedness relays (i.e., " bad-actor" relays) were evaluated for both Ciass IE and non-Class lE electrical systems, and a determination was Energy Research, Inc. 13 ERl/NRC 95-510

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made as to whether or not chatter of the relays is acceptable considering the application / function, normal snu'#Ekm, and SSEL impact of the relays. A walkdown was performed to spot-check relay types and mountings. Consequently, it is judged that the licensee has thus implemented an evaluation of relay chatter that is consistent with the guidelines of NUREG-1407. ' The resolution of findings with restsect to low-ruggedness relays, however, does not appear to be very meaningful.

. 2.1.10 Soil Failure Analysis The seismic IPEEE addressed potential failure of buried piping between the power block and site-specific structures, and notes that buried pipe design capability exceeds RLE demands. The submittal also notes that soil liquefaction analyses were performed as part of the plant design process, and that fill materials have been found to not be susceptible to liquefaction. No other mention of potential soil failures and their effects is made in the submittal.

This treatment of soil failures in the Callaway seismic IPEEE falls short of the guidelines described in NUREG-1407 for a focused-scope plant; however, the treatment is consistent with the modifications to focused-scope evaluation procedures described in Supplement No. 5 to GL 88-20.

2.1.11 Containment Performance Analysis The seismic IPEEE for Callaway Plant has involved an analysis of containment performance. A

' containment safeguards equipment list was developed for this purpose. The specific basis for developing this list is not described in detail in the submittal report. However, the list of selected components pertains to items needed to prevent early failure of containment funcuons. The evaluation focused on the following elements: containment integrity (including stmetural integrity and ruggedness of contamment cooling and mnrainmant spray systems); containment isolation functions (including ruggedness of penetrations and of actunnan and control systems that generate containment isolation signals); and prevention of seismically induced containment bypass (including interfacing systems LOCA and sieam generator tube rupture events).1Walkdown ofitems identified in the mnrainmant safeguards equipment list focused on identifying any potential problems asmciated with equipment anchorage or seismic spatial interactions. Most components were pre-screened.

Effectively, components important to successful containment performance were added to the SSEL and l embedded ra part of the seismic margin assessment, receiving a similar walkdown evaluation as for the i i

other SSEL components. The containment performance assessment in the Callaway seismic IPEEE is thus judged to be a meaningful examination of potential vulnerabilities that may prevent the success of early containment integrity.

2.1.12 - Seismic-Fire Interaction and Seismically Induced Flood Evaluations

a. Seismic-Fire Evaluation and Walkdown The Callaway 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 l fires, (2) seismic actuation of fire suppression systems, and (3) seismically induced failure of fire

< fsuppression systems. The submittal documents the findings of these primary aspects of seismic-fire

' interacuon evalw. ion, but minhnal treatment was given to seismically induced failure of fire suppression Energy Research, Inc. 14. ERl/NRC 95-510

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l systans. A seismic-fire walkdown has been conducted as part of the overall seismic-fire evaluation. The i

. walkdown included an investigation of hydrogen piping that runs through the auxiliary building, and of the RCP lube oil drain tanks in the reactor building. These items were found to be seismically rugged.

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b. ^ SeismicallyInduced Cres The assessment for seismically induced firec looked at potential breakage of liquid or gas flammnNes/ combustibles, hydrogen piping that runs through the auxiliary building, reactor coolant pump

. lube oil drain tanks in the reactor building, as well as other potmtial seismic-caused ignition hazards. L

c. Seismic Actuation ofFire Frotection System Equipment The evaluation for seismic actuation of fire suppression systems considered various elements of the fire j suppression syste:n (e.g., deluge valves, sprinkler system) and the potential for inadvertent actuation of !

fire pun ps due to relay chatter. The submittal concludes that it is unlikely that an inadvertent actuation I of the fire protection system will occur as a result of the RLE, but if it were to occur, the result would be {

localized spraying of components not required to respond to the earthquake.

d. Seismically Induced Failure ofFire Protection System Equipment

. The review of seismically induced failure of fire suppression systems looked at the seismic capability of limited aspects of the fire suppression system, and included an examination of the potential for adverse spatial intera::tions. Potential impacts of ground motion were not addressed for .everal aspects of fire protection capability.

c. Seismically Induced Flood Evaluation The submntal indicses that a walkdown for seismically indu,:ed flood concerns was performed as part of the IPEEE. The submittal documents a concern associatd with potential flooding of the 1974' level of

, the control building, due to possible sprinkler head breakage in an earthquake. However, the SRT confirmed that the most likely area of Noding would not impact SSEL equipment operation. The submittal also notes that inadvertent actuation of fire pumps can be expected due to relay chatter effects, but that this will only result in localized flooding if sprinkler head breakage happens to occur.

f ReviewIIndings The Callaway IPEEE has implemented a seismic-fire interactions evaluation which is essential complete, with exception of a '==aingful consideration of the potential and effects of seismically induced failure of fire protection systems.

2.1.13 Treatment of USl A-45 The success paths developed for the Callaway seismic IPEEE address decay heat removal requirements, following a seismic margm earthquake, via AFW capability and bleed and feed cooling. The elements of these decay heat removal functions have been evaluated as part of the consideration of SSEL components.

The SSEL was developed assuming decay heat removal requirements, under conditions of loss of offsite

' power and small LOCA, for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following the seismic margin earthquake. The submittal Energy Research, Ir.m 15 ERI/NRC 95-510 c !

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notes that the systems and components that fulfill the decay heat removal function, including support systans for the heat exchangers, pump seal cooling, and pump room cooling (CCW and ESW systems),

were included in the SSEL and their seismic adequacy assessed in both the pre-screening and walkdown L l

phases of th, seismic margin assessment. Piping associated with these systems was also sampled and evaluated for seismic adequacy during the walkdowns. Also, pre-screening / screening was performed to demonstrate that water storage tanks (RWST, reactor make-up storage tank, demineralized water storage tank, and CST) have adequate seismic capacity to insure reliability of water sources for decay heat removal. An analysis of elephant's foot buckling and diamond buckling was performed for the RWST.

In these evaluations, no vulnerabilities were noted with respect to seismic capability of decay heat removal

. systems.

Overall, the Callaway seismic IPEEE included a mamaingthi evaluation of potential vulnerabilities in decay heat removal systems, which is judged to address the relevant concerns of USI A-45. It is noted that the list of open issues presented in the IPEEE submittal suggests that two outliers have been identified that '

affect AFW capability: (1) missing shear pins in AFW pump foundations; and (2) potential interaction of a chain hoist with the AFW pump. Corrective actions were implemented for both of these outliers.

2.1.14 Treatment of GI-131 <

I GI-131 is'rclevant to Callaway Plant since the plant is a Westinghouse PWR, and it employs a movable flux mapping system. The seismic IPEEE submittal notes that, in response to NRC IE Information Notice 85-45 (June 1985), UE re-evaluated the seismic II/I analysis performed on the flux mapping system. As a result of this re-analysis, the hold-down assembly of the flux mapping system was upgraded by increasing the size and strength of bolts and plates that comprise the assembly. This upgrade was completed in 1987. For the IPEEE, the SRT attempted to check the modified hold down assembly, in order to verify the seismic adequacy of the earlier upgrade. However, the flux mapping system was not accanaihle to the SRT due to radioactive exposure concerns. UE has concluded that the earlier upgrade, addressing the NRC Information Notice 85-45, satisfies the requirements of GI-131.

In general, the licensee's approach to GI-131 is considered to be appropriate.

2.1.15 Other Safety Issues

a. Eastern U.S. Seismicity Issue Probabilistic seismic hazard calculations were performed for the Callaway site, as part of the resolution {

program for the Charleston Eanhquake Issue. Callaway was not identified as an outlier plant. The IPEEE submmal itself does not specifically comment on resolution of the eastern U.S. seismicity issue (Charleston j Earthquake Issue).

b. USl A-17 Regarding USI A-17, ' System Interactions in Nuclear Power Plants," the submittal states that closure for external events is provided by the seismic capability walkdown, which addresses systems intera:tions,

-including spatial, fire, and floodmg interactions. 'Ihe submittal concludes that such evaluation for external 1 events, together with separate evaluations performed for internal events, provide closure for USl A;17.

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c. USI A-40 ^

Concerning USI A 40, " Seismic Design Criteria," the submittal states that the RWST was re-analyzed in 1990, and that an additional analysis of the RWST (for buckling concerns) was performed as part of the seismic IPEEE, for a RLE of 0.3g. The submittal concludes that these analyses provide closure for USI A-40 concerns for safety-related above-ground tanks.

' d. Generic Safetyissues Some seismic-related information having relevance to Generic Safety Issue (GSI)-172 is provided in the submittal, as discussed in Section 2.4.3 of this TER.

e. Review Findings The seismic IPEEE includes a discussion of USI A-17 and USI A-40 which concludes that these issues are considered closed. The specific treatments of these issues were not evaluated as pan of this TER.

2.1.16 Peer Review Process The Callaway seismic IPEEE received independent peer reviews from UE personnel and from outside consultants. Each aspect of the seismic IPEEE was subject to various levels of review. . Various plant and ,

licensee departments have been involved in the review. The seismic SSEL was reviewed by UE's Training i Dept.rtment, and the relay evaluation was reviewed by UE's Engineering Department and by NUS Corporation. Stevenson & Associates, with assistance from consultant Dr. Robert J. Budnitz, reviewed '

all aspects of the Callaway SMA. Dr. John D. Stevenson accompanied the SRT during a portion of the seismic capability walkdowns, as peer reviewer.

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All review comments were addressed and resolved to the apparent satisfaction of the reviewers, who signedeff on the relevant ponions of the final IPEEE.

UE persomel had a meaningful (although not dominant) role in conducting the seismic IPEEE. The UE engineer who participated in the seismic evaluation received seismic qualification utility group (SQUG) training and EPRI Seismic IPE follow-on training. The submittal identifies all IPEEE participants, their i roles in conducting / reviewing the IPEEE, and their relevant qualifications. The IPEEE was executed in accordance with quality assurance procedures. Independent internal and external reviews were also pesformed of various aspects of the seismic analysis. The submittal notes that all portions of the IPEEE received several levels of review. ~ It is thus judged that a meaningful peer review was conducted of the

. Callaway seismic IPEEE, and that the major comments / concerns of the peer review team were substantially j addressed.

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2.1.17 Summary Evaluation of Key Insights The plant HCLPF capacity has been determmed to be at least equal to the RLE (NUREG/CR-0098 median, ;

5%-damped, soil spectrum. anchored to a PGA-value of 0.3g). The submittal provides sufficient s ;

justification to support this major finding. This finding is not surprising, given the high design basis and l recent vintage of Callaway Plant. i Energy Research, Inc.

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l Although the plant is seismically rugged, a number of comparatively minor outliers were noted, which the licensee has addressed through relevant safety enhancements or other resolution approaches. The submittal concludes that there are no seismic vulnerabilities at Caliaway Plant. Although opportunities for safety enhanmneras with respect to mitigation of relay chatter effects are evident from the submittal, the licensee has not proposed such enhancements.

2.2 Bre A sunwnary 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.

i 2.2.1 Overview and Relevance of the Fire IPEEE Process

a. Methodology Selectedfor the Fire IPEEE The IPEEE submittal includes extensive discussions regarding the methods and data used in conducting the fire analysis. The fire assessment is conducted using the FIVE methodology, and employs the approaches and data provided in Reference [12]. UE has participated in an EPRI-sponsored collaboration project, auned at testmg the FIVE methodology. The phased screening approach, as prescribed by FIVE, has been used. The unscreened areas and compartments have been subjected to fire modeling using the enn==*rized version of FIVE. The licensee has given special attention to control room fires. Possible occurrence of a hot-gas layer is assumed in modeling multi-compartmental fire propagation. A FIVE j Phase-III verification walkdown has been conducted as the final stop of the fire analysis.

b. ' Key Assumprions Used in Performing the Fire JPEEE \

Section 4.3.1.4 of the IPEEE submittal provides a list of assumptions and modeling conditions. Following are listed those assumptions and conditions that may have some notable effect on the final results of the i fire analysis:

1. Reactor sub-criticality is assumed to be successful in all cases.

2.- All cables are IEEE 383 qualified. ,

3. Fire barriers / boundaries are taken to be good as rated. '

4. The fire protection system is designed and installed per proper codes and standards.

5. A full fire-zone suppression system is capable of preventing a hot-gas layer from forming.

c. Status ofAppendix R Modifcations Callaway is a non-Appendix R plant. The licensee states that there are several exemptions regarding the fire protection system, but does not elaborate further on these. The licensee uses Appendix R definitions in the IPEEE fire analysis. For example, the definitions of fire areas and compartments are used per Appendix R guidelines. To define the safety component list, the IPE model has been used. There is no !

Appendix R safe shutdown list.

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.d. New or Existing PRA The IPEEE is a new fire study employing FIVE methodology.

2.2.2 Review of Plant Information and Walkdown

a. Walkdown Team Composition Two sets of walkdowns have been conducted to obtain and verify plant information. The walkdown teams i

- have used standardized forms to record their observations. , In Section 6 of the IPEEE submittal, the !

licensee describes in some detail the qualifications of the personnel involved in the analysis and the ,

participation of licensee's in-house personnel in the preparation of the fire IPEEE. The team was .

composed of expenenced engmeers and analysts. In addition, the fire analysis portion of the IPEEE was I "a part of an EPRI tailored collaborative project."

b. Signifcoat Walkdown Rndings l The scope of the walkdowns has included a wide variety of considerations, including fixed and transient

- Ignition sources, fire barrier qualifications, and potential for cross-fire compartment propagation. The '

- submittal does not indicate that the walkdown team discovered any fire vulnerabilities as a result of the plant walkdowns.

c.t Signi) cant Plant Features '

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' The following list describes some key features of the Callaway plant which are relevant to the fire analysis:

1. The auxiliary building houses engineered safety features equipment and nuclear steam supply l

. system (NSSS) avriliary equipment. 1

2. _ The control building includes the control room, cable spreading rooms, and other control and !

power distribution systems. I

' 2.2.3 - Fire-Induced Initiating Events l

a. Were Initiating Events Other than Reactor hip Considered?

Components and initiators modeled in the IPE have been used in the fire IPEEE. The possibility of a LOCA is not treated as a separate subject in the submittal. Other initiating events are also not addressed )

as a separate subject. From the discussions in Sections 4.3.2.2 of the IPEEE submittal, it can be inferred that the licensee has addressed initiators other than reactor trip in estimating the fire-induced CDF.

However, two issues are not explicitly discussed in the IPEEE submittal. First, it is not clear whether or j not the list ofIPE s =-ms has been expanded to include cables and components that could be involved i in initiating an accident. Second, in almost all of the discussions, there is no mention of cables as a component explicitly considered in the analysis.

is put of the screenmg protocol, the licensee states that areas in which a source for reactor trip could not be found were screened-out from further analysis. This approach, if applied without any reprd to other i i

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safety components, may lead to an optimistic screening conclusion. From a review of the submittal, however, it can be inferred that there are no fire areas or compartments for which this protocol has been applied in a non-conservative manner.

b. Were the Initiating Events Analyzed Properly 7 The initiating events (IPE initiators) have been addressed in the IPEEE. Section 4.3.3.2.1 of the submittal provides a list of event trees (and therefore, initiating events) that have been used in modeling the impact of fires. Later in the same section, from the discussions provided on pages 4-65 and 4-66, it can be concluded that the licensee has addressed all possible initiating events and has traced the cables relevant to these events.

2.2.4 Screening of Fire Zones

a. Was a Proper Screening Methodology Dnployed?

Screening of fire zones appears to have been performed properly. It has been conducted in multiple stages, using the protocol prescribed as part of the FIVE methodology. The licensee has used conservative measures to complete this task. In the first qualitative stage, if a compartment contained at least one safe shutdown component, it was retained for further analysis. In the second stage, the CDF threshold of 10'

'/ry was applied. Using the fire occurrence frequency for each compartment, together with the conditional CDF assuming loss of all cables and components in the compartment, the CDF contribution for the compartment was computed,

b. Have the Cable Spreading Room and the Control Room Been Screened Out?

The cable spreading rooms were included in the screening process. One of the two cable spreading rooms was screened out based on the application of the first quantitative screening.

The control room was treated as a special case, subjected to a detailed analysis. In this analysis, various panels were explored for potential fire occurrence, and the responses of the operators to the fire event was considered.

c. Were lhere Any Fire Zones / Areas that Have Been Improperly Screened Out?

No improperly screened fire areas / zones could be identified. This review finding was based on the information provided in the submittal. A random check of the frequencies and information on the type of equipment / cables present also revealed no errors or inconsistencies,. The final results seem to be reasonable, and are within the expected range of results for PWR plants.

2.2.5 Fire Hazard Analysis The data and methodology provided in Reference [12] has been used to estimate fire frequencies for individual compartments and fire areas. Weighting factors have been applied to apportion the overall fire frequency to the specific fire zones. Plant-specific fire occurrence data has not been used. The licensee states that the frequency of fire experiences at Callaway has been below the industry average.

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2.2.6 Fire Growth and Propagation -

a. heatment of Cross-2'one Fire Spread and Associated Major Assumprions Cross-zone fire propagation has been considered in two ways: (1) through multi-compartment interaction analysis (MCA), and (2) via fire compartment interaction analysis (FCIA). A hot gas layer is considered as the primary mechanism for propagation of the effects of a fire to an adjacent zone. As part of the qualitative screening, areas have been investigated for multi-compartmental fire effects, using such

. deterministic factors as: presence of fire-rated barriers, low combustible loading, and fire detection and

- suppression systems. However, the licensee has not taken credit for fire suppression systems in limiting the effects of a fire event. Also, it should be added that there are no normally open fire doors at this plant, and that the fire dampers are tested periodically for drop test and electrical circuit functionality. So far, no failures have been noted in these tests. The fire barriers are also subjected to a periodic surveillance 1 program. Only a few areas were not screened out based on this effort. This screening process was based on the criteria provided in Reference [12], which includes assumptions that all fire barriers are effective as rated, and that automatic suppression systems are effective in eliainating a hot gas layer. The licensee has also concluded that high-hazard fire areas (e.g., turbine building, diesel generator rooms and lube oil l storage rooms) are bounded by fire barriers that will contain a potentially severe fire in those areas.

After the qualitative and quantitative screenmg, those compartments that did not screen out were subjected to MCA. In the final analysis, only two areas were identified as being capable of having hot gas layers affecting adjacent areas.' For both of these areas, the final outcome was found to be risk insignificant.

b. - Assumprions Associated uith Detection and Suppression

- The Nuclear Saferj kaalysis Center's (NSAC's) approach to fire suppression modeling was used. The overall probability of aon-suppression was multiplied by the CDF value from the quantitative screening analysis, and aMmanal coirip.n- were screened based on this reduction. This approach does not take into account the possioility of bunched-together cables and equipment within a compartment.

The time to detect a fire for areas without an automatic suppression system was estimated based on fire source, target, and ionization detector positions.

c. heatment ofSuppression4nduced Damage to Equipment, ifAvailable There is no discussion of suopression-induced damage in the fire analysis. . Potential effects would include !

damage to cables and equipment as a result of the activation of the fire suppression system in extinguishing j a small fire in a given area. ' A somewhat related issue was discussed as part of the Sandia fire risk scoping study issues. The " Total Environmental Equipment Survivability" issue addresses the survivability of safe shutdown equipment from spurious actuation of a suppression system or from a mis-directed water stream during manual fire fighting.

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J 2.2.7 Evaluation of Component Fragilities and Failure Modes

a. Defnition ofFire-Induced Failures It is inferred that fire-induced failures have been properly considered. Reference [1] addresses hot shorts as a specific topic. From the discussions provided in the submittal, it can be inferred that the licensee has expended considerable effort in identifying valves susceptible to " hot short" and in determining their effects on the core damage pathways.

b. Method Used to Determine Component Capacities ,

The cable dusge temperature threshold criterion was taken to be 700 'F. This threshold is somewhat nonenservative, given that Reference [12] suggests 662 'F for IEEE 383 cables. However, the licensee claims that it has completed some sensitivity analyses where failure temperatures lower than that used in l the main analysis have been used. The overall impact on the final IPEEE results was consider 6d to be I minimal. No other damage criteria were presented in the submittal.

c. Generic Fragilities No specific discussion was provided in the submittal regarding equipment fragilities. Cable fragility was expressed in terms of a threshold temperature,

d. Plant-Specifc Fragilities No plant-specific fragilities have been mentioned in the submittal.

e. Technique Used to Treat Operator Recovery Actions Operator recovery actions were addressed conservatively. All basic events pertaining to human reliability in the core damage model were set to have failure probabilities of either 0.5 or 1.0. This approach is certainly conservative, and the treatment covers all possible adverse effects that a fire may impose on the proper execution of the required set of manual actions.

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 examined with respect to potential equipment and cable damage. Manual fire fighting, and the effect of fixed fire suppression systems on the formation of a hot gas layer, have been considered.

2.2.9 Analysis of Plant Systems and Sequences

a. Key Assumptions Including Success Criteria and Associated Bases The success criteria are directly taken from the IPE analysis and have not been modified for the fire analysis.

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b. Ewnt Trees (Functionalor Systemic)

Functional event trees have been displayed for the initiating events that were deemed to be possible as a result of a fire. No support systems were included in the models,

c. Dependency Matrix, ifit is Diferentfrom thatfor Seismic Events i

No dependency matrix has been provided in the submittal.

d. Plant-Unique System Dependencies The submittal does not present any unique system dependencies. The plant consists of only a single l reactor unit. '

e. Most Signijcant Human Actions Human actions, as discussed above, have been considered as an integral part of the fire scenario quantification. The submittal does not summarize the most significant human actions. Since a very conservative approach has been taken, it is not possible to distinguish the most significant human actions.

2.2.10 Fire Scenarios and Core Damage Frequency Evaluation I Using the FIVE methodology, several fire scenarios have been identified and analyzed in varying levels of detail. The analyses have taken into account numerous relevant issues, including the list of equipment and components failed by a fire, the conditional core damage probability (CCDP) given damage caused !

by the fire, the conditional probability of failure to suppress the fire, etc. A large number of fire areas :

' and compartments have been considered, and a summary of the analysis is presented in the submittal.

Many of the scenarios were found to have a CCDP ofless than 1 x 10-3 Three areas - the control room, -

and the two ESF switchgear rooms - were evaluated to have fire scenarios with a CCDP greater than this threshold value. The control room was analyzed in great detail, where the possibility of operators abandoning the control room was explicitly considered and modeled. l CDF was the principal parameter used for quantitative screening in the fire analysis. Sophisticated plant damage models (i.e., IPE models) have been used in the analysis. Fire ignition, propagation, detection, and damage have been quantified. The IPEEE submittal provides sufficient detail for the reviewer to verify a chain of computations. The IPEEE submittal presents numerous tables and figures that assist in this regard.

2.2.11 Analysis of Containment Performance

a. Signifcant Containment Performance Insights

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The submittal includes discussions of both inside-containment fires and outside-containment fires which ;

may affect containment isolation and heat removal. Containment fires were found to be of little '

significance, because of the RCP oil collection design and the separation of redundant cables. None of

- the significant fire scenarios outside the contamment were found to affect the containment-related functions in an adverse fashion.

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. b. Plant-Unique Phenomenology Considered No containment-related event trees have been used in any of the screening phases, nor in evaluating the

. unscreened fire zones.

2.2.12 Treatment of Fire Risk Scoping Study Issues

. a. Assumptions Used to Address Fire Risk Scoping Study Issues All of the Sandia fire risk scoping study issues have been addressed and resolved. The licensee has presented a detailed discussion pertaining to each issue, as summarized below.

1. Seismic-fire interaction has been addressed by examinations of: (a) the potential for a fire event resulting from an earthquake; (b) the potential for inadvertent seismic actuation of the fire suppression system and resulting effects on safety equipment; and (c) the potential for seismically induced failure of the fire protection system. A walkdown of the plant has been undertaken for these exammations. Hydrogen lines and the RCP oil system have been identified as potential fire sources. For both cases, the relevant components were found to be seismically rugged.

Inadvertent actuanon of suppression systems was addressed in some detail. There are no areas in ;

the plant where such an event could lead to safety equipraent damage not already considered in the !

fire analysis. Halon is used in those areas of the plant where the potential for equipment failure from water spray may exist.

2. A detailed discussion has been provided related to fire barriers. Specific inspection procedures are inarantad to verify the integrity of penetration seals and fire barriers. Fire dampers have been ;

ia==*ad at the time ofinstallation. ' Fire fightmg procedures address the potential for fire dampers l to fail to close properly.

3. Several procedures are implemented for fire detection, fire fighting, general personnel training, and for fire brigade training and drills. The general employee training procedures emphasize the ;

proper handling of accidents The fire bngade undergoes extensive drills on a regular basis. The drills are reviewed and critiqued by observers. !

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4. Regarding fire impacts on operator effectiveness in carrying out required tasks (from an !

environmental standpoint), a special procedure is implemented at Callaway for potential control room fires, and equipment are color coded for immediate identification under smoke or other stressful conditions. Also, self-contained breathing apparatus (SCBA) gear has been provided for the operators.

5. Control system interaction is addressed via the use of a remote shutdown panel and the capability to isolate the panels from one another.

b. Signifcant Findings i

1. The fire brigade undergoes sufficient training.

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2. The suppression systems do not have significant potential to adversely affect safety systems.

3. Procedures are available that address fire-related issues.

4. Existence of a remote shutdown panel, and the provision for isolating the panel from the control room, minimize the potential for control systems interaction.

2.2.13 USI A-45 Issue

a. . Methods ofRemoving Decay Heat The IPEEE fire analysis uses logic models developed for the IPE, which include the entire array of heat ;

removal capabilities of the plant.

b. . Ability of the Plant to Feed and Bleed The IPE model used in the IPEEE includes the provision for feed and bleed cooling. The licensee has used the dominant core damage scenarios to demonstrate how decay heat removal is modeled in the fire i analysis,

c. Credit Takenfor Feed and Bleed ,

Credit has been taken for feed and bleed capability. !

d. Presence of Thenno-Lag Reference [1] does not mention the presence of Thermo-Lag as part of the USI A-45 issue. However, in l Serian 4.3.1.4 of the submittal, it is mentioned that no credit was generally given to Thermo-Lag, with

- the exception of fire area A-6.' For this fire area, further fire modeling would have led to the same conclusion with or without crediting the effectiveness of the Thermo-Lag. Therefore, the effect of Thermo-Lag is concluded to be minimal with respect to fire risk.

2.3 ' HFO Events The Callaway IPEEE submittal reports no unduly significant sequences (i.e., vulnerabilities) with respect to HFO events. The submittal indicates that implementation of the screening approach described in Supplement 4 to Generic Letter 88-20, and in the guEance of NUREG-1407, has formed the basis for the conclusion that the plant is not vulnerable to HFC events.

The general methodology that has been implemented for the HFO events analysis makes use of the following screening steps: .

L High winds, external floods, and tr.=gnhiion and nearby facility accidents have been considered for analysis.' Other plant external hazards have not been considered.

2. For each of the hazards identifed in Step -(1), any changes that have taken place since the time the !

l operating license (OL) was issued have been identified. ;

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3. For each HFO hazard addressed, a quick screening has been performed to identify whether or not the plant meets the 1975 SRP criteria.

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4. The resulting HFO events analysis has been documented in the IPEEE submittal report.

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The review of changes since the time of OL issuance has revealed that there are no new conditions that would significantly affect the plant design basis. The submittal states that the following types of changes were reviewed:

Military and industrial facilities located within 5 miles of the site

- On-site storage or other activities involving hazardous materials Transportation characteristics in the vicinity of the plant site

Developments that could affect the original design conditions i

Because Callaway Plant was designed and constructed in compliance with 1975 SRP criteria, the screening approach was limited to these preceding areas.

I 2.3.1 High Winds and Tornadoes i 2.3.1.1 General Methodology The tornado analysis for Callaway Plant has been based primarily on the plant's FSAR assessment. The design wind velocity for all Category-I structures is 100 mph, at 30-feet above ground level, for a 100-year .

recurrence interval. The design-basis tornado is defined as a tornado that exceeds the design wind velocity with a frequency no greater than 10 per year. The submittal simply states that, because there have been no significant developments and changes since the time the plant operating licensee was issued, which would affect the original design conditions for high winds and tornadoes, no further analysis of high wind and tornauoes has been performed, i

'2.3.1.2 Plant-Specific Hazard Data and Licensing Basis I i

Since Callaway conforms with the 1975 SRP criteria, no plant-specific hazard analysis is performed.

2.3.1.3 Significant Changes Since Issuance of the Operating License The submittal states that no significant changes have occurred since the time of issuance of the OL.

Callaway Plant conforms to the criteria set forth in the 1975 SRP.

2.3.1.4 Significant Findings and Plant-Unique Features The submittal does not report findmgs of any plant design review nor of any plant walkdown. The design-basis tornado, with wind speeds of up to 360 mph, is simply considered to be harmless. Since Callaway Plant conforms to the 1975 SRP criteria, the submittal screens out bigh-wind and tornado events. As such,

no related vulnerabilities have been identified. No other related findings of significance have been cited in the submittal.

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2.3.1.5 Hazard Frequency

- Since Callaway Plant conforms to the 1975 SRP criteria, no hazard frequencies for high winds / tornadoes were presented.

2.3.2 External Flooding -

2.3.2.1 -

General Methodology Callaway plant is located at 840 feet mean sea level (MSL); all safety-related equipment are located at elevations of no less than 840.5 feet. The highest flood on record from the Missouri River was about 300 feet below this elevation. The plant's high elevation thus excludes the possibility of any major flooding.

- As such, floodmg hazard has been screened out in the HFO events IPEEE. Also, since no changes in the plant which affect flooding have occurred since the time of issuance of the OL, the submittal excludes external flooding from further consideration. ,

k 2.3.2.2 Plant-Specific Hazard Data and Licensing Basis l 1

I The probable maximum flood (PMF) is defined as the flood event that may be expected from the most severe combination of all possible weather and hydrological conditions affecting the Callaway Plant. One component of the PMF is related to the maximum flood level along the Missouri River, which is estimated j to reach 548 feet MSL. This flood level will not reach Callaway Plant, which is located at 840 feet MSL. 1 The PMF serves as an element of the design basis reported in Section 2.4.2.3 of the FSAR. The plant site dramage system is designed to remove water runoff from a 100-year storm. Even if the drainage system does not function properly, the PMF would overflow onto roadways, away from safety-related equipment, ;

' and into the retention pond. A maximum pond elevation of 839.87 feet MSL is expected. This level is j still less than the 840 feet plant level. The submittal states that accumulation of water precipitation on !

building roofs is not expected to cause any damage. The roofs have been designed to accommodate the maxunum load resulting from the 48-hour snow load. The roofs of all safety-related buildings have also been designed to prevent ponding or reliance on drains, with the exception of the diesel generator building and the^ control building (for these weaker roofs, the risk significance of ponding was not formally ,

2 calculated). Safety-related structures have been designed to withstand : snow load of 125 lb/ft (slightly less for the diesel generator and control buildings). On this basis, the submittal states that "... this provides a closure of GI-103 for Callaway." '

2.3.2.3 Significant Changes Since Issuance of the Operating License The submittal states that there have not been any significant developments affecting the original design condition, as regards to floods, since the time ofissuance of the operating license.

2.3.2.4 . Significant Findings and Plant-Unique Features No significant findmgs related to flood events were reported. Callaway Plant is located such that a PMF event is not a credible contributor to plant severe-accident risk.

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e 2.3.2.5 Hazard Frequency

- External flooding was screened out'due to the high elevation of Callaway Plant. As such, the frequency of external flooding that adversely affects the plant is expected to be extremely small. The submittal discusses the possibility of drainage system failures, but the licensee expects that the resulting flow into the retention pond (considering its elevation) will still produce flood levels below the locations of safety-related equipment. No discussion is provided in the submittal as to whether or not such an event would adversely affect non-safety-related equipment. For example, the submittal does not address when a PMF,

_with subsequent failure of the drainage system, may adversely affect the switchyard, transmission lines, or transformers, thus potentially causing loss of offsite power.

2.3.3 Transponation and Nearby Facility Accidents 2.3.3.1 _ General Methodology The Callaway IPEEE submittal has addressed aircraft crashes, as well as water, rail, and highway transponation accidents. Also, the submittal considers potential impacts of on-site hazardous material inventories.

In the analysis of aircraft crashes, the submittal states that there are eight federal airways and three transition routes within a 10-mile radius of the plant. No military routes were found to pass within 10 miles of the plant. The submittal states: "... there is no aircraft hazard whose probability of occurrence is greater than 1.01 x104per year as described in Section 2.2.3 of the FSAR ... *,

No specific methodology is cited in the submittal for the evaluation of nearby land and water transportation accidents. The submittal only states that the Missouri River runs approximately 5 miles southeast of the

. plant site and is a major transportation artery for barge traffic. Because of the significant distance from the plant, no related hazards are expected. This conclusion is consistent with the Callaway FSAR, Section ;

2.2.3.

The nearest main land route is Highway 94, which is located approximately 3.7 miles from the plant at its nearest point. According to FSAR Section 2.2.3, there is no risk to the plant from the transportation j of hazardous cargo along this route.

No military or industrial facility exists within a 5-mile radius of the plant. Thus, any related potential hazards have been screened out. Similarly, a survey of on-site toxic materials has resulted in the conclusion that any related hazards are mmimal. A survey of on-site explosive materials has also produced the conclusion that plant exposure for hazards from these sources is not significant.

2.3.3.2 - Plant-Specific Hazard Data and Licensing Basis For aircraft crashes, the SRP licensing basis and the Callaway FSAR were used to determine the frequency of a crash. 'Ihe subnuttal is organized and systematic in this analysis. Detailed data on the frequency of flights from nearby airpons, however, are not reported. Since most of the major transportation routes are far from the plant, the submittal screens out the respective transponation hazards.

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) i For the analysis ofland and water transportation events, no licensing basis information, details of specific events, or presentation of hazard data have been explicitly provided. Transportation events were i considered to be insignificant based on qualitative arguments.

The submittal states that potential design-basis accidents from nearby transportation events have been evaluated in FSAR Section 2.2.3, which reveals no on-site or offsite hazards that would have a significant I potential to adversely affect the plant.

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2.3.3.3- Significant Changes Since Issuance of the Operating License

. The submnal states that there have not been any significant developments, that affect the original design condition, with regard to transponation and nearby facilities accidents, since the time of issuance of the OL. '

2.3.3.4 Significant Findmgs and Plant-Unique Features .

No significant findmgs are dicme=A in the submittal for transportation and nearby facility accidents. No description of the walkdown conducted by the licensee, and no detailed description of the relevant

- qualifications of the walkdown team members, have been discussed.

2.3.3.5 Hazard Frequency

'No hazard frequency estimates were made for transportation and nearby facility accidents. Qualitative judgments, and conformance to 1975 SRP criteria, have been cited as the principal reasons for excluding these hazards.

2.4 rw s rwy i-- -- (cgr.147 car 148. =ad csr.172) 2.4.1 GSI-147, " Fire-Induced Alternate 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:

a' Electrical independence of remote shutdown control systems Loss of control power before transfer

. Total loss of system function

. Spurious actuation of components The licensee evaluated spurious actuation leading to LOCAs and interfacing system LOCAs. Since the submittal has followed the guidance provided in FIVE concerning control system interactions, all circuitry associated with remote shutdown is assumed to have been found to be electrically independent of the control room.

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., a 2.4.2 GSI-148, " Smoke Control and Manual Fire Fighting 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 efforts l By damaging or degrading electronic equipment By hampering the operator's Lbility to safely shutdown the plant 1

C . By initiating automatic fire protection systems in areas away from the fire

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Reference [13] identifies possible reduction of manual fire-fighting effectiveness and causing misdirected suppression efforts as the centralissue in GSI-148. Manual fire-fighting was credited in the analysis (page

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4-% and 4-104) within the confines of the FIVE methodology. However, no information was provided I

concerning the potential for smoke to reduce manual fire-fighting effectiveness or misdirected suppression 1 efforts.

2.4.3 GSI-172, " Multiple System Responses Program (MSRP)"

Reference [13] 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

subsection is only to identify the location in the IPEEE submittal where information having potential relevance to GSI-172 may be found.

Common Cause Failures (CCFs) Related to Human Errors Descripnon of the Issue [13): 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 imtiate CCFs include: manufacturing errors in components that

- affect redundant trains; and installation, maintenance 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.

Sections 3.1.1,3.1.2.1,3.1.2.4 (page 3-8), and 3.1.2.5 (detailed discussion of human actions screening) of the Caliaway IPEEE submittal provide information on operator recovery actions, as a consideration for !

success path selection in the seismic analysis. In regard to the fire analysis, the submittal provides !

Information on operator recovery actions in Sections 4.3.3.2 (with the most detailed discussion being provided in Section 4.3.3.2.2.F on pages 4-70 and 4-71, and Table 4.3.3.2-1), 4.3.5.4.C, 4.3.6.3, ,

4.3.6.5.D, and 4.3.6.6.

Non-Safety-Related Control System / Safety-Related Protection System Dependencies

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Desennnan of the h- [13]: 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 of interdependence between !

! ' safety-related and non-safety-related systems, and should identify potential sources of vulnerabilities. The Energy Research,'Inc. 30 ERI/NRC 95-510 I l

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l 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.

Information provided in the Callaway IFEEE submittal pertaining to seismically induced spatial and

- functional interactions is identified below (under the heading Seismically Induced Spatial and Functional Imeractions), whereas information pertaining to fire-induced alternate shutdown and control panel I interactions has already been identified in Section 2.4.1 of this TER.

Heat / Smoke / Water Propagation Efectsfrom Fires l

Descrintinn of the luna [13J: Fire can damage one train of equipment in one fire zone, while a redundant !

train could potentially be damaged in one of following ways: )

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 equipmem A random failure, not related to the fire, could damag" e +adant train. i l

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 caa cause unintended operation of equipment due to hot shons, 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 run, and electrical breakers could fail open or closed. The concern of water propagation effects resulting from fire is partially 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. The concern of alternate shutdown / control room interactions (i.e., hot shons and other items just mentioned) is addressed in GSI-147.

Information provided in the Callaway IPEEE submittal penaining to GSI-147 and GSI-148 has already been identified in Sections 2.4.1 and 2.4.2 of this TER. Section 4.3.5.4 of the submittal presents information pertaining to this issue.

Efects ofFire Suppression System Actuation on Non-Safety-Related and Safety-Related Equipment Description of the Issue [13]: Fire suppression system actuation events can have an adverse effect on safety-related components, either through direct contact with suppression agents or through indirect interaction with non-safety related components. This concern is addressed in GI-57.

Information pertaining to suppression-induced damage to equipment, as well as seismically induced inadvertent actuation of fire suppression systems, can be found, respectively, in Section 4.3.5.4, and Section 3.2.3, of the submittal. Discussion on GI-57 is provided in Section 4.8, which refers to the information in Sections 4.3.5.1.B and 4.3.5.4.B. It must be noted that Section 4.3.5.1.B is missing from the submittal. However, other referenced sections address the issue.

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p Efects ofFlooding and/or Moisture intrusion on Non-Safety-Related and Safety-Related Equipment Daenintion of the Issue [13]: Flooding and water intrusion events can affect safety-related equipment either directly or indirectly through flooding or moisture intrusion of multiple trains of non-safety-related equipment. 'Ihis type of event can result from external flooding events, tank and pipe ruptures, actuations of fire suppression systems, or backflow through parts of the plant drainage system. The IPE process j addresses the concerns of moisture intrusion and internal floodmg (i.e., tank and pipe ruptures or backflow through part of the plant drainage system). The guidance for addressing the concern of external flooding is provided in Chapter 5 of NUREG-1407, and the concern of actuations of fire suppression systems is provided in Chapter 4 of NUREG-1407.

The following infonnation is provided relevant to this issue: the Callaway IPEEE submittal discusses external iloods in Section 5.2; discussion is provided in Section 4.3.5.4 regarding actuations of fire suppression systems; discussion of seismically induced inadvenent actuation of fire suppression systems is provided in Section 3.2.3; and seismically induced flooding is discussed in Sections 3.1.4.2.1 (page 3-24), 3.1.4.2.2 (page 3-27), and 3.2.3.

Seismically Induced Spatial and Functional Interactions j Daenintinn of the kcna [13): 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 l in small piping that may disable essential plant shutdown systems; direct impact of non-seismically l 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' l failures; and indirect impacts, such as dust generation, disabling essential plant shutdown systems. As part l of the IPEEE, it was specifically requested that seismically induced spatial interactions be addressed during plant walkdowns. The guidance for performing such walkdowns can be found in EPRI NP-6041.

' The Callaway IPEEE bas included a seismic walkdown which investigated the potential for adverse physical interactions. m submittal states that EPRI NP-6041-SL guidelines were follt wed in the seismic walkdowns. Relevant information can be found in Sections 3.0, 3.1.1, 3.1.2 (page 3-3), 3.1.4 (with detailed discussion on the seismic capability walkdown presented in Section 3.1.4.2), 3.1.5.2.2, 3.1.5.3.2, 3.1.5.4,3.2.3, and 4.3.5.1 of the submittal.

Seismically induced Fires Description of the Issue [13]; 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 multiple safety-related systems. Seismically induced fir s is one aspect of seismic-fire interaction concerns, 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 4.3.5.1 of the Callaway IPEEE submittal provides a discussion of seismically induced fires.

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, Seismically induced Rre Suppression System Actuation Daa~intian of the I=aa [13): Seismic events can potentially cause multiple fire suppression system actuations which, in turn, may cause failures of redundant trains of safety-related systems. Analyses currently required by fire protection regulations generally only examine inadvertent actuations of fire suppression systems as single, i=iana-iaar events, whereas a seismic event could cause multiple actuations of fire suppression systems in various areas.

Section 3.2.3 of the Callaway IPEEE subnunal provides discussion of seismically induced fire suppression system actuation.

SeismicallyInduced Flooding Dancrgtion of the Issue [13]: 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.

The Callaway IPEEE submittal includes information on seismically induced flooding in Sections 3.1.4.2.1 (page 3-24), 3.1.4.2.2 (page 3-27), and 3.2.3.

SeismicallyInduced Relay Chatter Dancqption of theIssue [13J: Essential relays must operate during and after an earthquake, and must meet one of the following conditions:

rammin functional (i.e., without occurrence of contact chattering);

be seismically qualified; or

be chatter acceptable.

It is possible that enntact 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 guida :e specifically requested licensees to address the issue of relay chatter.

A detailed discussion of relay chatter is provided in Section 3.1.4.4 of the submittal. Some additional information on relay chatter is mentioned in Sections 3.2.3. Regarding containment isolation, Section 3.1.5.2.2 (page 3-46) notes that " actuation and control systems that generate containment isolation signals were included in the SSEL and evaluated as part of the Seismic Capability Walkdown," but the discussion

- does not explicitly discuss relay chatter effects.

Evaluation ofEanhquake Magnitudes Greater than the Safe Shutdoun Earthquake D=mintinn of the I=== [13]: The concern of this issue is that adequate margin may not have been

~ included in the design of some safety-related equipment. As part of the IPEEE, all licensees are expected l- ! to identify potential seismic vulnerabilities or assess the seismic capacities of their plants either by L

par. ug seismic PRAs 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.

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The Callaway IPEEE has included a focused-scope seismic margin assessment, as document:1 in Section 3 of the submittal. The seismic input for the analysis is described in Sections 3.1.3 of the submittal.

I Efects ofHydrogen Line Ruptures Description of the Issue [13]: Hydrogen is used in electrical generators 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 cover 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 portion of the IPEEE submittal.

A detailed discussion of potential hydrogen line and component ruptures (including nonseismic and seismic causes) is provided in Section 4.6 of the submittal. Other information on hydrogen line ruptures is mh on page 3-27 and in Section 4.3.5.1. Some information on hydrogen storage is also provided on page 4-34.

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3 OVERALL EVALUATION AND CONCLUSIONS 3.1 Seismic The approach chosen by the licensee for conducting the seismic IPEEE of Callaway Plant essentially addresses all relevant issues and concerns recommended in NUREG-1407 for evaluation of a focused-scope plant. The submittal itself is considered to be well written and structured, and the identification and implementation of safety enhancements is judged to have produced a meaningful response to GL 88-20, for a focused-scope plant.

The study has implemented a unique pre-screening approach which does not explicitly verify that component HCLPF capacities exceed the RLE; only adequacy with respect to design basis demands is verified by the pre-screening.

The focused-scope seismic evaluation for Callaway Plant has identified 21 open issues which required implementation of resolution approaches. A few minor fixes and four more noteworthy plant improvements have resulted from such resolution. These safety enhancements have been made primarily with respect to a review of safe shutdown equipment; no improvements were found specifically or exclusively as a result of the containment performance evaluation.

Based on this submittalmnly review, the following items have been identified as the primary strengths and weaknesses of the seismic IPEEE submittal for Callaway Plant:

Strencths (1) The study has implemented a relevant, meaningful approach for conducting the seismic IPEEE, and considers all major issues of concern.

(2) The study has made use of plant personnel (who received training in seismic evaluation methods),

yet utilized outside expertise where needed; thus ensuring that the study was completed (and reviewed) by qualified personnel and that the licensee was familiar with the evaluation.

(3) The submittal is clear and well-structured.

(4) The submittal's treatment of non-seismic failures and human actions is generally well executed.

(5) Even though many components were pre-screened, they still received a walkdown inspection which included an assessment of general condition, anchorage condition, and the potential for interactions. The basis and implementation approach for pre-screening, including the development and completion of Area Walkdown Sheets, is considered to be well conceived and supported.

Weaknesses (1) The submittal makes brief reference to evaluation of seismically induced failure of fire suppression systems, but provides no meaningful discussion of this aspect of seismic-fire interactions.

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o s 1 (2) The submittal does not repon HCLPF results for masonry / block walls and flat-bottomed tanks.

(This point is judged to be only a minor weakness, because the submittal does, in fact, demonstrate that the RLE input is exceeded by the design-basis input for these items.)

Despite these weaknesses, it is clear that the licensee has benefitted substantially from the seismic IPEEE effort, that meaningful plant improvements are being made, and that the licensee has improved its understanding of plant behavior in response to potential severe accidents caused by eanhquakes.

With respect to findings of the Callaway seismic IPEEE, it is important to highlight that existing anomalous conditions, as noted in the submittal, can become more significant concerns as time proceeds, and new ancmalies may also be encountered, due to plant degradation (e.g., cracking or rusting of anchoragehapports) with the passage of time. It is thus considered important that such anomalous conditions '>e monitored.

3.2 Bre The licensee has expended considerable effon in the preparation of the fire analysis portion of the IPEEE.

The licensee has employed proper methodology, i.e., the FIVE methodology, and has used proper fire occurrence and suppression system failure data bases for conducting the fire analysis.

Strennths The following items are considered to be the primary strengths of the submittal:

(1) The fire analysis portion of the IPEEE is well written. The overall presentation is clear and well organized. Tables and figures provide considerable supporting information for the analysis and conclusions.

, (2) An acceptable methodology and data have been used.

(3) The final conclusions are reasonable, and are within the range of results expected for a PWR.

Weaknesses Regarding the weaknesses of the licensee's effort, only a few items can be noted. In general, these weaknesses are deemed to be of comparatively minor significance. The following items are considered to be wonh mentioning:

(1) It is not clear how the initiating events considered in the IPE were used in the fire analysis. The principal concern of relevance is the need to ensure proper treatment of equipment failure caused by a fire that may lead to such initiating events as a LOCA and/or loss of offsite power.

(2) Sensitivities of the final results to specific assumptions were not determined In the screening analysis, there are several cases where the core damage frequency is only slightly smaller than the threshold value. Clearly, for these cases, the resulting conclusion to screen fire scenarios is highly sensitive to underlying assumptions.

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3.3 HFO Events The Callaway HFO events IPEEE generally follows the submittal guidelines of NUREG-1407. The following are considered to be the major strengths and weal nesses of the submittal:

Strenoths

(1) The submittal is well written and clear.

(2) The analysis of HFO initiators was performed entirely by licensee staff, using in-house expertise.

This approach has marimimi he t licensee's appreciation of HFO-based severe accident behavior.

Weaknesses (1) The submittal does not report findings from walkdowns, for HFO events.

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I 4 IPEEEINSIGHTS, IMPROVEMENTS, AND COMMITMENTS 4.1 Seismic l

Of 592 components considered in the seismic IPEEE, all but 22 components were eliminated from evaluation durmg the design-basis pre-screemng and area-walkdown process. After subsequent application of conventional seismic margin assessment (SMA) screening and analysis, it was determined that HCLPF calculations needed to be performed for only a small number of components. The results of the l

calculations for these components are HCLPF capacities in excess of the review-level earthquake (RLE).

Hence, all safe shutdown equipmeat list (SSEL) components, as well as components required for success of early containment integrity, have been determined to be capable of withstanding the RLE.

l Consequently, the IPEEE submittal explicitly states that the overall HCLPF capacity for Callaway Plant is at least 0.3g. (HCLPF assessments have been made with respect to the NUREG/CR-0098 median, 5%

damped spectral shape, in accordance with the guidelines of NUREG-1407.)

The submittal states that the IPEEE has identified no plant-specific vulnerabilities at Callaway Plant.

Twenty-one (21) open issues requiring resolution, however, were revealed and addressed in the study.

Table 4.1 (derived from IPEEE Table 3.1.4-1) summarizes these issues and the approaches implemented for their resolution. In two cases, issues were resolved by demonstrating a HCLPF capacity in excess of the RLE. Other issues were resolved by additional walkdowns, observations, judgments, and/or analyses.

In addition, a few minor fixes and four more noteworthy enhancements were implemented to resolve other issues. From among the foregoing items, the seismic IPEEE submittal has drawn out the following l improvements /findinFs as being most significant:

HCLPFResults The HCLPF capacity of component cooling water (CCW) heat exchangers anchorage was determined to be 0.41g PGA.

The HCLPF capacity of diesel generator control panels anchorage was determined to be 0.49g PGA.

Minor Fixes Hand-held fire extinguishers have been remounted at heights not exceeding UL Standard drop heights. Hand-held fire extinguishers have been removed from the containments during normal operation; access to fire suppression facilities is located near the containment hatch, for use by the fire brigade in responding to fires in the containment.

A floor grate, which crosses a building expansion gap behind the "A" diesel generator (DG)

I startmg air compressor skid, was clipped to the floor to prevent potential damage to DG sensing lines that pass through the grate.

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. < e Enhancements Several motor comrol centers (MCCs), whose top spray shields were found to be in contact with the adjacent walls, were bolted to the walls, to eliminate a spatial interaction concern.

Missing shear pins in the foundations / supports of AFW pumps were installed.

Loose equipment found behad the control room was removed and signs posted prohibiting further storage.

Chain hoists located in the emergency core cooling system (ECCS) pump rooms, the CCW pump rooms, and the AFW pump rooms, found to be in close proximity to their associated pumps, were moved to safe storage locations at the end of their rooms, and acedural guidance and training was instituted to prevent recurrence of this issue.

The submittal indic.ates that the noted fixes and plant enhancements have either been implemented or are planned for implementation.

4.2 Bre Overall, the licensee has concluded that there are no significant fire vulnerabilities at Callaway Plant. The total fire core damage frequency (CDF) from " unscreened" scenarios is estimated at 8.9 x 104/ry. This frequency is within the range of fire-induced CDFs obtamed for other nuclear power plants. The dominant contributors to the fire CDF are the control room and the ESF switchgear rooms. The control room fire scenarios are dominated by operator failure to control the plant from the remote shutdown panel.

No improvements or commitments were identified as being necessary to further reduce the fire risk at i Callaway Plant. The licensee has used the Nuclear Energy Institute's (NEI's) severe accident closure guidehnes (NEI-9104) to evaluate the need for plant improvements. Based on the existirig plant features and opentional practices (i.e., procedures and policies), the licensee has concluded that " Severe Accident Managernent Guidance" (SAMG) need not be put in place for those fire scenarios that exceed the guideline threshold value. However, the licensee has elected to undertake analysis of operator responses to spurious !

actuations which may result from a fire in the control room or either of the two ESF switchgear rooms. l The entire fire analysi., effort has provided an excellent opportunity for licensee engineers to improve their knowledge of the characteristics of the plant, of how the plant would behave under fire conditions, and I of what human actions will be necessary to protect the core from any adverse initiators. !

4.3 HFO Events l~ The IPEEE submittal coreludes that, since Union Electric's original analysis of the Callaway Plant y conforms to 1975 SRP crite sa, all HFO events can be screened out. The submittal follows the screening l approach suggested by Supplement 4 to Generic Letter 88-20. Accordingly, because no changes have been !

identified since the time the operating licensee (OL) was issued, all HFO events have been screened out.

Correspondingly, no vulnerabilities or safety enhancements have been identified, and no commitments for plant improvements related to HFO initiators have been made.

Energy Research, Inc. 39 ERI/NRC 95-510

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l Table 4.1 Open Issues, and Their Resolutions, as Identified in the Callaway Seismic IPEEE l

WALKDOWN ISSUE RESOLUTION

, Spray shields on MCCs NG03C, NG04C, The MCCs in question were modified by l' NG05E, NG06E, NG07F and NG08F are in attaching them to the adjacent wall by angle l contact with adjacent walls. This will cause braces attached to the spray shields.

impact with the wall during an earthquake.

The centrifugal charging pumps (CCPs), safety Westinghouse letter SCP-93-169 documented injection pumps (SIPS), and AFW pumps have that shear pins are not required for the SI/CCPs.

openings for shear pins and the seismic The AFW pumps have had shear pins installed.

qualification packages for these pumps appear to It has been determined that the AFW pump take credit for these pins. Shear pins are not drivers do not require shear pins; high strength installed in the AFW pumps and drivers, nor in bolts are sufficient.

the B train SI/CCP drivers.

The fire sprinkler piping and risers in the DG The fire piping was assessed during the II/I rooms are rod-hung. The configuration could hazards analysis as being able to withstand the result in the piping falling on the DGs during an SSE, based on engineering judgment.

earthquake. Additional review by the SRT indicates that there is enough interference to prevent the pipmg from falling on the DGs.

Hand-held fire extinguishers throughout the Fire extinguishers in the containment have been plant are mounted on relatively shallow hooks. removed to eliminate the concern. Also, During an earthquake, they could rock off the extinguishers have been drop-tested, and are hooks and fall to the floor. This may result in a able to withstand a fall from current mounting missile hazard due to the bottle or neck heights. Still, several extinguishers are being fracturing. remounted to meet the required drop height.

Carts, file cabinets, and other unsecured items A sign has been posted on the wall in *.he l behind the control room may shift during an vicinity of NF039A-C and SA036A-E stating l earthquake and impact safety-related cabinets. that no loose equipment should be stored along this wall. All loose equipment was verified to be removed on 2/1/94.

An argon cylinder has bee'n loosely taped for The gas cylinders noted during the walkdown support near EFV0352. have been verified to be removed as of 12/14/93.

The chain ho. ts over safety related equipment A memo stated that RFR 8191 addressed the may travel during an eanhquake, resulting in the issue of chain hoists and specified that they be chains impacting soft targets, such as cables, parked away from equipment, with the hook instrumentation lines, relief valves, and oil fully raised. A walkdown on 2/1/94 found the bubblers. chain hoists parked away from safety-related i

equipment.

Energy Research, Inc. 40 ERI/NRC 95-510 L

i e WALKDOWN ISSUE RESOLUTION Some of the fluorescent light fixtures could fall The IPEEE walkdown did not identify any during an eanhquake, because some "S" hooks fixtures that could fall on safety-related are open instead of being crimped closed. equipment. A follow-up walkdown on 12/14/93 found no light fixtures that could damage safety-related equipment if they fell. l The fire hose header in the lower cable A review of the hanger drawing for this piping spreading room is missing its vertical suppon shows that a clamp is not required at this clamp. location.

Sensing line for BBPT0455 is not in its support. The sensing line was put back in its support under a work request.

The floor grate behind the DG A starting air The floor grate was clipped to the floor by compressor is loose and could damage the means of a work request.

sensing lines passing through it in an earthquake.

The power cables to the DG room ventilation It has been determined that flexible conduit for fans are exposed. these fans is not required. l There is a missing bolt on cabinet RP147A and The missing bolt and screws were installed by l there are missing retaining screws on some of means of a work request.

the Foxboro modules inside the cabinet.

There is an unsupported fire sprinkler test The piping has been evaluated as II/I qualified.

header in the lower cable spreading room. No action required.

In the plant southeast quadrant of the ESW pipe Pipe schedule and wall thickness provide chase, there is a rod hung sprinkler riser 6 strength against failure, and even if failure inches from an adjacent rod-hung pipe. These occurred, the resulting flooding and/or spray pipes could hit during an earthquake, resulting would not damage SSEL equipment, in an inadvenent actuation of the fire system.

ESW return from DG B piping has rigid An initial review of Stress Run P-283 by restraints that do not appear to allow for Engineering indicated that SAM was accounted differential building motion. for and no action was required. Funher review by the SRT shows that restraint location was input to code wrong, resulting in wrong anchor movement.

Issue resolved for the IPEEE because the SRT judges that the piping will survive an earthquake without damage. Engineering judges that there is sufficient margin in existing design to accommodate the error with no further action.

l l Energy Research, Inc. 41 ERI/NRC 95-510 1

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WALKDOWN ISSUE RESOLUTION Main safety features instrumentation system Bechtel calculation J-200H found the cabinet (MSFIS) cabinet SA075A and engineered safety motion to be less than the measured gap.

features actuation system (ESFAS) cabinet Further review found the as-tested frequencies SA066A are very close together, but are not are less than the calculated frequencies, connected. Clearance between the two cabinets resulting in larger motion than calculated.

is approximately %' at the top, but 1/8" to 5/32" where bolt heads project. This could This will result in cabinet impact during the result in impact during an eanhquake. RLE. However, Engineering has determined that cabinet impact is acceptable. No further action is required.

CCW Heat Exchanger anchor bolts do not have A HCLPF calculation determined the seismic 10 bolt diameter edge distance, nor are they 10 capacity of the CCWHXs to be 0.41g.

bolt diameters apan.

Diesel Generator Control Panels KJ121 and A HCLPF calculation determined the seismic KJ122 are mounted on grout pads which are capacity of the panels to be 0.49g.

smaller than the panel depth, resulting in the panels being cantilevered over the grout pads.

There are block walls around the battery rooms. A review of the design calculations found ample This is a seismic interaction concern. margin for the IPEEE.

( The seismic capacity of the control rod drive The reactor internals were designed for a 1.0g

! mechanisms (CRDM) and the Reactor Internals ZPA (zero period acceleration) in both l needs to be verified, horizontal directions, and 0.67g in the vertical l direction. The CRDM housings have external lateral supports. This meets the requirements of I

the IPEEE.

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1 Energy Research, Inc. 42 ERI/NRC 95-510

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5 IPEEE EVALUATION AND DATA

SUMMARY

SHEETS Completed data entry sheets for the Callaway Plant IPEEE are provided in Tables 5.1 to 5.3. These tables have been completed in accordance with the descriptions in Reference [7]. Table 5.1 lists the overall extxnal events results. Table :i.2 summanzes general seismic data pertaining to the focused-scope seismic evaluation. Table 5.3 provides the PWR Seismic Success Paths table, which gives a description of the success paths developed for the focused-scope seismic evaluation. The IPEEE submittal does not provide sufficient information regarding core damage sequences and system failures pertaining to the fire events analysis; hence, no data mmmary tables are provided pertaining to the fire evaluation. Also, no PRA or bounding analysis was performed as part of the Callaway HFO events IPEEE; hence, no data summary tables are provided pertaining to evaluation for these external initiators, i

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. Energy Research, Inc. 43 ERI/NRC 95-510

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6 REFERENCES

1. "Callaway Plant IPEEE," Union Electric Company, June 30,1995.

2. " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -

10 CFR 50.54(f)," U. S. Nuclear Regulatory Commission, Generic Letter 88-20, Supplement 4, June 28,1991.

3. J. T. Chen, et al., " Procedure and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," U. S. Nuclear Regulatory Commission, NUREG-1407, May 1991.

4. " Design Response Spectra for Seismic Design of Nuclear Power Plants," U.S. Atomic Energy Commission, Regulatory Guide 1.60, Revision 1, December 1973.

5. R.T. Sewell, et al., " Individual Plant Examination for External Events: Review Guidance,"

Energy Research, Inc., 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 Shett Package," Lawrence Livermore National Laboratory, December 14, 1993.

8. " Docket No. 50-483, Callaway Plant: Responses to NRC RAI Letter on Callaway Plant," letter from Donald F. Schnell, Union Electric Company, to U. S. Nuclear Regulatory Commission, February 1996.

p.

9. "Developmera of Criteria for Seismic Review of Selected Nuclear Power Plants," NUREG/CR-0098, May 1978.

10. "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Electric Power Research Institute, EPRI NP-6041-SL, Revision 1, August 1991.

I1. " Enhancing the NRC and EPRI Seismic Margin Review Methodologies to Analyze the Importance i of Non-Seismic Failure, Human Errors, Opportunities for Recovery, and Large Radiological I Releases," U. S. Nuclear Regulatory Commission, NUREG/CR-5679, August 1990. l l

12. " Fire Induced Vulnerability Evaluation (FIVE)." Electric Power Research Institute, TR-100370, I Revision 1, September 29,1993.

13. " Staff Guidance ofIPEEE Submittal Review on Resolution of Generic or Unresolved Safety Issues (GSI/USI)," U.S. Nuclear Regulatory Commission, August 21,1997.

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l Energy Research, Inc. 47 ERl/NRC 95-510 i

y G. Randolph -

If you have any questions regarding the attached SER, please contact me at 301-415-1307 or at ind8nrc cov on the internet.

Sincerely, SNProject Manager, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosures:

1. Attendance List and Handout from 4/28/93 Meeting

2. Staff Evaluation Report

3. Technica! Evaluation Report cc w/encis: See next page Q)STRIBUTION:

Docket File PUBLIC PDIV-2 Reading l SRichards(clo)

OGC ACRS BHardin, RES ARubin, RES CWoods, RES DCoe RHeman RScholl(e-mail SE)

DGraves, RGN-IV - FES Meyno<3 daIeek 7/1Gyn ad4/cygcg TO feoelve a copy of mis document, Indicate v in me Dox OFFICE PDIV-2M C PDIV-2/LA C PDIV-2/SC I NAME JDp[Mrb E@ SDembek [

DATE 9 [ 9 /99 9 / 8 /99 9 / f /69 DOCUMENT NAME: G:\PDIV-2\Callaway\LTR83602.wpd ;

OFFICIAL RECORD COPY i l

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