ML20209G001

Staff Evaluation of Individual Plant Exam of External Events Submittal on Plant,Unit 1

ML20209G001
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Site:	Three Mile Island
Issue date:	07/09/1999
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9907160207
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4 UNITED STATES s

j NUCLEAR REGULATORY COMMISSION 4

WASHINGTON, D.C. 20086-0001

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STAFF EVALUATION REPORT QE THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)

SUBMITTAL ON '

THREE MILE ISLAND NUCLEAR STATION. UNIT 1

1.0 INTRODUCTION

On June 28,1991, the NRC issued Generic Letter (GL) 88-20, Supplement 4 (Ref.1) along with NUREG-1407, Procedural and Submittal Guidance, requesting all licensees to perform individual plant examinations of external events (IPEEE) to identify plant-specific vulnerabilities to severe accidents ~and to report the results to the Commission together with any licensee-determin_ed improvements and corrective actions. In a letter dated December 29,1994, the licensee, General Public Utilities Nuclear (GPUN) Corporation (now known as GPU Nuclear, Inc., or GPUN), submitted its response to the NRC (Ref. 2) for Three Mile Island Nuclear Station, Unit 1 (TMI 1).

The staff contracted with Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL) to conduct screening reviews in the seismic and fire areas, respectively, of the licensee's IPEEE submittal. The NRC staff conducted a screening review of the high winds, floods and other external events (HFO) areas of the submittal. The staff sent a request for additional information (RAI) to the licensee on January 8,1998 (Ref. 3). The licensee responded to the RAI on April 24,1998 (Ref. 4). Based on the results of the review of the j

licensee's submittal and response to the RA1, the staff concluded that the aspects of seismic events; fires; high winds; floods; and other external events were adequately addressed. The -

review findings are summarized in the evaluation section below. Details of the staff's and contractor's findings are presented in three technical evaluation reports attached to this staff evaluation report.

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In accordance with Supplement 4 to GL 88-20, the licensee provided information to address the resolution of Unresolved Safety issue (USI) A-45," Shutdown Decay Heat Removal Requirements," Generic Safety Issue (GSI)-103, " Design for Probable Maximum Precipitation i

(PMP)," GSI 57," Effects of Fire Protection System Actuation on Safety-Related Equipment,"

and the Sandia Fire Risk Scoping Study (FRSS) issues. These issues were explicitly requested in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407. The licensee did not propose to resolve any additional USIs or GS:s as part of the TMI-1 iPEEE.

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2.0 EVALUATION I

Three Mile Island ' Nuclear Station / Unit 1 is a Babcock and Wilcox (B&W)-designed pressurized l

water reactor (PWR) located about ten miles southeast of Harrisburg, Pennsylvania. The plant l

was designed to a seismic horizontal acceleration level of 0.06g operating basis earthquake l

and 0.12g safe shutdown earthquake. The TMl-1 facility was characterized in NUREG-1407 to be a 0,3g focused-scope plant. For the seismic analysis, the licensee used detailed plant l

walkdowns'and performed a seismic probabilistic risk analysis (PRA) to estimate the capacity of plant structures and equipment to withstand beyond-design basis eadhquakes and to estimate the frequency of seismic sequences. High confidence of low probability of failure (HCLPF)

L values of critical equipment groups were also estimated for TMI 1. For the fire analysis, the l

licensee performed plant walkdowns, fire area screening, and quantification of fire sequences due to fires in unscreened fire areas using the fire-induced vulnerability evaluation (FIVE)/PRA methods. For the analyses of other external events, the licensee used the progressive screening approach as described in NUREG-1407.

2.1 Core Damace Freauency Estimates

_The licensee performed a detailed seismic PRA of major plant structures and equipment and screened them based on Electric Power Research Institute's (EPRI) seismic margin assessment method (Ref. 5). The licensee estimated the total seismically induced core damage frequency (CDF) using the Lawrence Livermore National Laboratory (LLNL) seismic hazard curves to be 8.4E-5 per year. The CDF using the EPRI seismic hazard curves was also i

' estimated, and this estimate was found to be 3.2E-5 per year. As part of their seismic l

assessment the licensee also estimated HCLPF values of critical equipment and found that HCLPF values for the majority of shutdown-related equipment were higher than the NUREG-1407 criterion of 0.3g (see Section 3.1.4 of the IPEEE submittal). The licensee found the HCLPF values for some equipment to be in the range of 0.09g to 0.25g. These low HCLPF components were evaluated for their contribution to seismic CDF.

The total fire-induced CDF from eight unscreened fire areas was estimated to be 2.4E-5 per L

year. The licensee estimated the CDF values due to high winds and external floods to be 7.4E-7 per year and 8.1E-5 per ye_ar, respectively. The licensee also estimated the CDF values due to chemical release events and air transportation events to be 1.6E-7 per year and 4.0E-7 per year, respectively. The licensee estimated the CDF due to internal events, including l

' internal floods, to be about 4E-5 per year.-

l 2.2 Dominant Contributors l

The licensee performed Fussell-Vesely importance measure calculations based on the TMI-1 seismic PRA results. Importance measure calculation results were obtained for individual sequence importance, equipment importance, and recovery action importance. The top ten sequences account for over 50 p,ercent of the total seismic CDF. These sequences were

' initiated by seismic events ranging from 0.15g to 0.4g. The leading dominant sequence, which contributes about 13 percent of the total seismic CDF, is a sequence with seismic failure of J

_ Class 1E AC power with offsite power available. The most significant seismic failures of I

equipment are: 480V Class IE load centers (1 P,1S,1 R,1T), switchyard insulators, control

- room ceiling, emergency diesel generator (EDG) air start receivers, decay heat closed cooling L

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water heat exchangers, and EDG ground resistors. Relay chatter recovery actions were also identified as an important, but smaller, contributor to the overall seismic CDF.

The licensee estimated the fire-induced CDF to be 2.4E-5 per year. Fires in the West and East Inverter Rooms were found to contribute significantly (about 1.1 E-5 per year) to the overall fire CDF (47%). The CDF due to fires from two switchgear rooms (Room 1D and Room 1E) was about 8.9E-6 per year (39%). The CDF due to control room fires was about 3.1 E-6 per year (14%). Control room fires included consoles (designated as consoles CC and CR) and panel fires.

l The total CDF estimate due to high winds, external floods, chemical release events, and air transportation events was found to be 8.2E-5 per year. This CDF estimate was dominated by external floods (about 98%).

The staff finds that the licensee's IPEEE assessment has examined the significant initiating events and associated dominant accident sequences for TMI-1.

2.3 Containment Performance The licensee evaluated critical containment failure modes, such as containment bypass, containment isolation, and performance of containment barriers, using applicable seismic analysis models and evaluation criteria. The licensee's seismic evaluations included seismic fragility evaluations of the primary containment (Reactor Building), the 125V AC instrumentation buses which are needed for containment isolation, and relay chatter of the containment isolation valves. The licensee judged that the HCLPF values of the primary containment and the 125V AC instrumentation buses are greater than 0.3g. A containment penetration walk-down was also performed by the licensee to identify seismic II/l interactions affecting containment isolation failures. The containment walkdowns consisted of inspecting and evaluating unusual conditions or configurations (e.g., unrestrained components, potential spatial interactions, and unique penetration:,). One potential seismic ll/l interaction affecting one containment purge valve was identified and was evaluated for a need for plant improvements through a probability-based seismic sensitivity analysis. This evaluation concluded that failure of the valve was an insignificant contributor to risk, and no plant improvement was identified. No other unusual conditions were identified during the walkdown.

In addition, the licensee evaluated quantitatively (using the seismic PRA event tree approach) failures of eleven containment isolation valves. Based on the seismic PRA sequence evaluation, the licensee estimated a total CDF of 5.5E-7 per year for a plant damage state (PDS) that involves a small LOCA with injection failure, steam generator available, no cooling available for the containment, and a large containment isolation failure.

The licensee evaluated Mtential damage to the control circuits of the containment isolation valves (discussed aTWC lue to fires in the unscreened fire areas in the Auxiliary Building and the Control Building hat vuld result in containment isolation failure or bypass. Specifically, eleven containment isolation valves were identified for evaluation of fire-induced containment isolation failure using applicable fire-indu';ed faults such as hot shorts of cables and damage to isolation valves. Based on the fire sequence evaluations, the licensee estimated a CDF of 6.9E-8 per year for three plant damage states (PDS7L, PDS4L, and PDS8G). This frequency estimate is dominated by fires in the 1 E switchgear room which affected containment isolation valves. Fires in other buildings were also evaluated, and the evaluation indicated no impact on

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containment isolation failures. Loss of containment cooling systems such as the Reactor Building (RB) Spray system and the RB Emergency Cooling system, were also included in the q

evaluation of the failure of containment due to internal fires. The licensee concluded that the rnachanisms for containment failures due to loss of containment cooling systems are the same for internal fire scenarios as they are for other internal initiating events evaluated in the TMI-1 individual plant examination (IPE). The IPEEE fire analysis did not identify any additional containment failure modes that were not evaluated in the TMI-1 IPE.

The staff finds that the licensee's containment performance evaluations for seismic and internal fires have considered important containment performance issues and are consistent with the l

Intent of Supplement 4 to GL 88-20.

2.4 Generic Safety issues i

As a part of the IPEEE, a set of generic and unresolved safety issues (USI A-45, GSI-131, GSI-103, GSI-57, and the Fire Risk Scoping Study (FRSS] issues) were identified in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407 as needing to be addressed in the IPEEE. The staff's evaluation of these issues is provided below.

2.4.1 USl A-45," Shutdown Decay Heat Removal Requirements" The licensee's seismic evaluation of shutdown decay heat removal (DHR) is documented in Section 3.2.1 of the IPEEE submittal. This evaluation includes the development of a detailed i

seismic PRA, including USI A-46 relay chatter evaluations and plant walkdowns, to identify seismically induced equipment failures and seismic II/I interactions. The licensee made use of the equipment fragility values developed by EOE International, system models, and sequence models and estimated frequencies of seismic sequences due to all seismic events of intensity ranging from 0.05g to 0.6g. The IPEEE also evaluated the impact of fire-induced faults of DHR related equipment on the fire-induced CDF (e.g., hot shorts in the reactor primary side PORVs). The licensee's fire evaluation of shutdown DHR is documented in Section 4.9.1 of the IPEEE submittal. The licensee's evaluation of the HFO portion of shutdown DHR is documented in Sections 5.1,5.2, and 5.3 of the IPEEE submittal. The licensee made use of the system and sequence models developed for their IPE and calculated frequency estimates of applicable sequences due to HFO events (e.g., floods, airplane crashes, chemical releases from nearby raillines). The staff finds that the licensee's USl A-45 evaluation is consistent with the guidance provided in Section 6.3.3.1 of NUREG-1407, and therefore, the staff considers this issue resolved for TMI-1.

2.4.2 GSI-131," Potential Seismic interaction Irvolving the Movable in Core Flux Mapping System Used in Westinghouse Plants" TMI-1 is not a Westinghouse design. Therefore, GSI-131 is not applicable.

2.'4.3 GSI-103," Design for Probable Maximum Precipitation (PMP)"

Results of the licensee's evaluation in response to GL 89-22, " Potential for increased Roof Loads and Plant Area Flood Runoff Depth at Nuclear Power Plants due to Change in PMP Criteria Developed by the National Weather Service," are documented in the TMI-1 engineering calculations file system (

Reference:

TMI-1-90-9108). This evaluation - addressed the effect of

5 new roof ponding loads and new site drainage problems due to postulated new PMP data (i.e.,

HMR 52 published in August 1982). The licensee concluded that the new PMP is not a concem for TMI-1. The staff finds that the licensee's GSI-103 evaluation is consistent with the guidance provided in Section 6.2.2.3 of NUREG-1407, and therefore, the staff considers this issue resolved.

2.4.4 GSI-57," Effects of Fire Protection System Actuation on Safety-Related Equipment" The licensee addreased the applicable FRSS issues in Sections 4.8.1 and 4.8.4 of the IPEEE submittal. One of the FRSS issues addresses safety problems (e.g., the effects of inadvertent actuation of fire protection systems on safety systems) documented in GSI-57. The concern of this GSI also includes seismically induced fires, seismically induced suppressant diversion, and seismically induced actuation of the fire protection system. The staff finds that the licensee's overall GSI-57 evaluation is consistent with the guidance provided in Section 6.2 of NUREG-1407, and therefore, the staff considers this issue resolved for TMI-1.

2.4.5 Fire Risk Scoping Study issues The licensee addressed FRSS issues in Section 4.8 of the IPEEE submittal. These FRSS issues include: (1) seismic / fire interactions, (2) adequacy of fire barriers, (3) smoke control and manual fire-fighting effectiveness, (4) equipment survival in a fire-induced environment, and (5) fire-induced alternate shutdown / control room panel interaction. The staff finds that the licensee's evaluation is consistent with the guidance provided in NUREG-1407, and therefore, the staff considers this issue resolved for TMI-1, 2.5 Other Generic Safety Issues in addition to those USIs and GSis discussed above that were explicitly requested in Suppl 3 ment 4 to GL 88-20, four GSis were not specifically identified as issues to be resolved under the IPEEE program; thus, they were not explicitly discussed in Supplement 4 to GL 88-20 and NUREG-1407. However, subsequent to the issuance of the GL, the NRC evaluated the scope and the specific information requested in the GL and concluded that the plant-specific analyses being requested in the IPEEE program could also be used, through a satisfactory IPEEE submittal review, to resolve the external event aspects of these four GSis. The following

- discussions summarize the staff's evaluations of these GSis for TMl 1.

2.5.1 GSI-147," Fire-Induced Alternate Shutdown / Control Room PanelInteractions" The licensee addressed applicable FRSS issues in Section 4.8 of the TMI-1 IPEEE submittal.

One of the FRSS issues addresses safety problems documented in GSI-147. The licensee's FRSS evaluation addresses applicable issues of GSI 147 through the fire scenario analysis of the main control room fires (Sections 4.6.5 and 4.7.2 of the TMI-1 IPEEE submittal). Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no potential vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers this issue resolved for TMI-1.

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- 2.5.2 GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness" The licensee addressed applicable FRSS issues in Section 4.8 of the TMI 1 IPEEE submittal.

One of the FRSS issues addresses safety concerns documented in GSI 148. The licensee's FRSS evaluation addresses the applicable issues of GSI-148 in Section 4.8.3 and Section 4.8.4

'of the TMI-1 IPEEE submittal Based on the results of the IPEEE submittal review the staff considers.that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no potential vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers this issue resolved for TMI-1.

2.5.3 GSI-156," Systematic Evaluation Program (SEP)"

The licensee's IPEEE submittal contains information addressing the following SEP issues: (1) settlement of foundations and buried equipment (Section 3.1.1.6), (2) dam integrity and site flooding (Section 5.2), (3) seismic design of structures, systems, and components (Sections 3.1.1.5 and 3.1.1.6), (4) site hydrology and ability to withstand floods (Section 5.2.3), (5) industrial hazards (Section 5.3.1), (6) tornado missiles (Section 5.1.5), (7) severe weather effects on structures (Section 5.1), and (8) shutdown systems and electrical Instrumentation and control (Section 4.8.5). Based on the results of the iPEEE submittal review, the staff

- considers that the licensee's process is capable of identifying potential external events-related vulnerabilities associated with this issue. On the basis that no potential vulnerabilities associated with these issues were identified in the IPEEE submittal, the staff considers the IPEEE-related aspects of this issue resolved for TMI-1.

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

The licensee's IPEEE submittal contains information directly addressing the following external

- events-related MSRP issues: (1) effects of fire protection system actuation on non-safety-related and safety related equipment (Section 4.8.1.2), (2) seismically-induced fire suppression system actuations (Sections 3.2.2 and 4.8.1), (3) seismically-induced fires (Section 4.8.1), (4) effects of hydrogen line rupture (Section 3.1.2), (5) the IPEEE-related aspects of common cause failures related to human errors (Section 3.1.5),

(6) non-safety-related control system / safety-related system dependencies (Section 3.1.2),

(7) effects of flooding and/or moisture intrusion on non-safety-related and safety-related equipment (Sections 5.2.3 and 5.2.4), (8) seismically induced spatial / functional interactions (Section 3.1.2.1 ), (9) seismically induced flooding (licensee's response to RAI in Ref. 4 ), (10)

- seismically !nduced relay chatter (Section 3.1.4), and (11) evaluation of earthquake magnitude greater than safe shutdown earthquake (Section 3.1). Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no potential vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers the IPEEE-related aspects of this issue resolved for TMI-1.

2.6 Uniaue Plant Features. Potential Vulnerabilities. and imorovements TMl 1 is located on an island in the Susquehanna River, and the plant is protected by surrounding dykes (flood-walls) to protect against flooding.

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The licensee defined a potential seismic vulnerability for TMI-1 as any seismically induced core damage sequence greater than 1E-4 per year or any containment bypass or large early containment failure greater than 1E-6 per year. Using this criterion, the licensee did not identify any seismic vulnerabilities. The licensee did not define a potential severe accident vulnerability In the fire or HFO areas, and no vulnerabilities were identified.

The licensee identified six potential seismic-related plant improvements as part of their IPEEE review. Three of the six plant improvements were implemented (Attachment 1 of Ref. 4).

These three plant improvements are: (1) modification to the control room ceiling, (2) modification of an existing seismic restraint to penetration pressurization air tank, PP-T-1 A (if this tank were dislodged during a seismic event, it could possibly damage one of the reactor building purge inlet valves), and (3) modification of a seismic restraint to EDG air receivers.

The licensee determined that the other three seismic improvements were not cost effective. No fire-related plant improvements were identified. In the flood area, the licensee planned a plant improvement to mitigate the consequences of higher river flood levels (i.e., beyond the design basis flood level of 310 feet). However, the IPEEE submittal did not specify what equipment would be installed. The licensee also planned to develop mitigating guidelines in the event of severe flooding of the Susquehanna River.

The staff finds that the licensee's discussion on unique plant features, potential vulnerabilities and potential improvements is reasonable given the TMI-1 design and that the licensee is i

capable of identifying potential vulnerabilities and improvements.

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3.0 CONCLUSION

t On the basis of the above findings, the staff notes that (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 88-20 (and associated guidance in NUREG 1407), and (2) the IPEEE results are reasonable given the TMl-1 design, operation and history. Therefore, the staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the TMI-1 IPEEE has met the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues as discussed in this SER.

It should be noted that the staff focused its review primarily on the licensee's ability to examine TMI-1 for severe accident vulnerabilities. Although certain aspects of the IPEEE were explored in more detail than others, the review was not intended to validate the accuracy of the llcensee's detailed findings (or quantitative estimates) that underlie or stem from the examination. Therefore, tMs SER does not constitute NRC approval or endorsement of any I

IPEEE material for purposes other than thete associated with meeting the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues discussed in this SER.

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4.0 REFERENCES

I 1.

NRC Generic Letter 88-20, Supplement 4, " Individual Plant Examination of External i

Events (IPEEE) for Severe Accident Vulnerabilities - Title 10 CFR 50.54(f),"

I June 28,1991.

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2.

- Letter from R. W. Keaten of GPUN to NRC, subject: "Thiee MHe Island Nuclear Generating Station, Unit 1 - Response to Generic Letter 88-20, supplement 4 -

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Individual Plant Examination of External Events (IPEEE) for Severe Accidents,"

December 29,1994.

3.

Letter from NRC to J. W. Langenb6ch of GPUN, subject: " Request for Additional Information for the Three Mile Island Nuclear Station, Unit 1, related to Individual Plant Examination for External Events (IPEEE) Response to Generic Letter 88-20, Supplement 4," January 8,1998.

4.

. Letter from J. W. Langenbach of GPUN to NRC, subject:"Three Mile Island Nuclear Station, Unit 1, Individual Plant Examination for External Events - Response to Request for Additional Information" April 24,1998.

5.

"A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

EPRI-NP-6041-SL, Revision 1, August 1991.

Attachments: Appendix A Appendix B Appendix C Principal Contributor: E. Chelliah Date: July 9,1999 m

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Appendix A Submittal-Only Screening Review of the Three Mile Island Nuclear Station, Unit 1, Individual Plant Examination for External Events (Seismic Portion), Dated March 1999 t.

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