ML20206M260
| ML20206M260 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/06/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20206M239 | List: |
| References | |
| GL-88-20, NUDOCS 9905140232 | |
| Download: ML20206M260 (8) | |
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STAFF EVALUATION REPORT INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL OMAHA PUELIC POWER DISTRICT FT. CALHOUN STATION. UNIT 1 DOCKET NO. 50-285
1.0 INTRODUCTION
On June 28,1991, the NRC issued Generic Letter (GL) 88-20, Supplement 4 (with NUREG-1407, Procedural and Submittal Guidance) requesting that all licensees perform individual plant examinations of external events (IPEEE) to identify plant-Apocific vulnerabilities to severe accidents and to report the results to the Commission together with any licensee-determined improvements and corrective actions. By letter dated June 30,1995, Omaha Public Power District (OPPD), submitted its response to the NRC.
The ataff contracted with Energy Research, Inc. to conduct a completeness and reasonableness " Step 1" review of the licensee's IPEEE submittal and its associated documentation in March 1995 and sent a request for additional information (RAI) to the licensee in June 1996. The licensee responded to the RAI in August 1996. On the basis of this review and further review by a senior review board, the staff concluded that the aspects of seismic; fires; and high winds, floods, and other extemal events were adequately addressed. The review findings are summarized in the evaluation section below. Details of the contractor's findings are presented in the enclosed technical evaluation report (TER).
In accordance with Supplement 4 to GL 88-20, the licensee also proposed to resolve in its IPEEE submittal generic safety issues (GSis) GSI-57, " Effects of Fire Protection System Actuation on Safety-Related Equipment," GSI-103, " Design for Probable Maximum Precipitation (PMP)," Unresolved Safety issue (USI) A-45, " Shutdown Decay Heat Removal Requirements" and the Sandia Fire Risk Scoping Study (FRSS) issues. Review findings regarding these issues are included in this SER.
2.0 EVALUATION The Fort Calhoun Station (FCS) consists of a single unit, Combustion Engineering two-loop pressurized-water reactor (PWR), with an electrical output of 501 MWe. The site is located on the west bank of the Missouri River, approximately 19 miles north of Omaha, Nebraska. The licensee performed a focused-scope NRC seismic margins analysis (SMA), consistent with NUREG/CR-4334 using a seismic margin earthquake (SME) of 0.3g peak ground acceleration (PGA). FCS is also a USI A-46 plant, and the IPEEE seismic walkdowns were performed in conjunction with the USl A-46 walkdown. Thus, significant information was shared between the i
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two programs. For fire events, the licensee performed an evaluation based on the Electric Power Research Institute's (EPRI)" Fire PRA implementation Guide." For the analyses of high winds, floods and other extemal events (HFO), the licensee generally followed the progressive screening procedure as described in NUREG-1407 and NUREGICR-2300. For the HFO assessment, the licensee compared the plant design basis to the 1975 Standard Review Plan (SRP) criteria although the plant was constructed too early to be a SRP plant. The licensee also used deterministic bounding analyses and simplified probabilistic bounding analyses to demonstrate that risks associated with HFO events are low.
Core Damage Freauency Estimates A PRA quantification for fire events that included Level 1, 2, and 3 PRA information was performed in which the contribution to plant core damage frequency (CDF) was estimated to be g.2E-5/RY before the implementation of certain plant procedural changes and 2.7E-5/RY after the changes. Since the licensee used the seismic margin approach for the IPEEE seismic analysis, there is no estimate for the seismic contribution to CDF at FCS. The licensee estimated that the contribution from other extemal events (i.e., extemal floods, high winds, transportation and industrial events) are insignificant at FCS except possibly for dam-break floods and periodic flooding. Initial results from probabilistic bounding analyses for dam-break floods provided the licensee incentive to modify procedures in AOP-1, Act of Nature. After procedural improvements, the CDF for dam-break-induced flood was estimated to be 6E-7/RY, and the CDF estimated for periodic flooding was 3E-6/RY. The licensee estimated that the CDF due to extemal events, after certain hardware and procedural improvements, is about 3.1E-5/RY excluding seismic events since they were not quantified. The licensee estimated that the CDF due to intemal events is about 1.4E-5/RY, including intemal flooding.
These CDF estimates compare reasonably with those of other plants.
Dominant Contnbutors Since the licensee used the seismic margins approach (SMA) for the seismic assessment, an identification of dominant seismic contributors to plant CDF was not made. The SMA was used to evaluate potential seismic weaknesses to provide a basis for assessing cost-effective plant modifications where weaknesses were identified. As a result of the SMA, the licensee focused its efforts on improving the seismic margin of systems and equipment associated with high-pressure injection, long-term cooling with the auxiliary feedwater system, and the motor control centers.
The fire CDF is dominated by fires in the control room and in the east basement of the auxiliary building. The important system / equipment contributors to the estimated fire CDF that appear in the top sequences are associated with the loss of the power operated relief valves (PORV) and i equipment failures leading to an interfacing system loss of coolant accident (ISLOCA). The dominant contributors for HFO are dam-break-induced flood and periodic flooding due to intense rainstorms.
The licensee's IPEEE assessment appears to have examined the significant initiating events and dominant accident sequences.
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.Y 3-e Containment Performance The licensee has assessed containment performance uader seismic conditions at FCS focusing on the likelihood of a large early release owing to containment bypass or isolation failure. It was estimated that the conditional probability of large early release, given a core damage sequence induced by earthquakes at the plant review level earthquake (RLE), or below, is about 1%.
The licensee has also assessed the containment failure modes that could be caused by fire. It was estimated that fire-induced containment bypass events (Event V) represented less than 2%
of the total CDF for internal fire events. However, considering the consequences of these events, they were a dominant contributor to plant risk from fire events (48% of the estimated risk from fire). Table 4.7 of the licensee's submittalillustrates these values.
The lice.nsee's containment performance analyses for seismic and intemal fire events appeared to have considered important severe phenomena and are consistent with the intent of Supplement 4 to Generic Letter 88-20.
Generic Safety lasues As a part of the IPEEE, a set of generic and unresolved safety issues (USl A 45, GSI-131, GI-103, GSI-57, and the Sandia Fire Risk Scoping Study [FRSS) issues) were specifically identified during the initial planning of the IPEEE program and explicitly discussed in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407 as needed to be addressed in the IPEEE. The staff's evaluations of these issues are provided below.
- 1. USl A-45, " Shutdown Decay Heat Removal Requirements" t
The licensee addressed USI A-45 in the seismic area qualitatively by comparing the l insights gained by performing the SMA with the generic issues as summarized in !
Appendix 5 of GL 88-20. A number of plant improvements were identified to address outliers for USI A-46 that are relevant to USl A-45, and credit was taken for the correction of these outliers. The IPE results were also cited in which it was determined that multiple paths were available for decay heat removal (DHR) at FCS. From its assessment, the licensee concluded that the plant DHR pathways are seisnlically robust with considerable expected margin above the safe shutdown design basis earthquake of 0.17g. It was further concluded that there are no unique DHR vulnerabilities for seismic events. For fire, the licensee used its fire PRA and compared those results with the A-45 insights from GL 88-20. Similar conclusions were reached as for the IPE. That is that feed and bleed provides for effective DHR under accident conditions and that feed and bleed failures are not dominant contributors to CDF due to the availability of redundant auxiliary feedwater in systems that do not tend to fail when feed and bleed is lost. The licensee concluded that there are no unique DHR vulnerabilities with respect to fire. The staff finds that the licensee's evaluation of USI A-45 is consistent with the guidance provided in Section 6.3.3.1 of NUREG-1407.
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- 2. GSI-131, " Potential Seismic Interaction involving the Movable In-Core Flux Mapping System used in Westinghouse Plants' Since FCS is not a Westinghouse design, GSI-131 is not applicable.
- 3. GI-103, " Design for Probable Maximum Precipitation" The licensee has explicitly assessed Gl-103 in its submittal. The licensee reports that FCS has a roof drain system, and that the roof designs are capable of withstanding ponding loads even with all of the drains plugged. Also, the reactor is sited on ground that is above the surrounding area so that water drains away from the site. The licensee concluded that the new Probable Maximum Precipitation (PMP) criteria will not havvany impact on Fort Calhoun due to the roof design and surrounding conditions. The ste.i finds that the licensee's GSI-103 evaluation is consistent with the guidance provided in Section 6.2.2.3 of NUREG-1407.
- 4. GSI-57," Effects of Fire Protection System Actuation on Safety-Related Equipment" The licensee has assessed the impact of inadvertent actuation of fire protection systems on safety systems which is also one of the issues identified in the FRSS. The submittal states that this issue was evaluated prior to GSI-57 in response to inspection and Enforcement Notice IE 83-41, and that the results from this evaluation coupled with the SMA walkJown indicate that there are no concems at FCS regarding this issue. The staff finds that the licensee's GSI-57 evaluation is consistent with the guidance provided in EPRI's Fire-induced Vulnerability Evaluation (FIVE) which was accepted by the NRC staff.
- 5. Fire Risk Scoping Study lasues The licensee has explicitly addressed the Fire Risk Scoping Study issues which are discussed in more detailin Section 2.2.12 of the TER. The staff finds that the licensee's evaluation is consistent with the guidance provided in NUREG-1407.
In addition to those safety issues discussed above that were explicitly requested in Supplement 4 to GL 88-20, four generic safety issues 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 generic letter, the NRC evaluated the scope and the specific information requested in the generic letter and the associated IPEEE guidance, 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 extemal event aspects of these four safety issues. The following discussions summarize the staff's evaluation of these safety issues at FCS.
- 1. GSI-147, " Fire-induced Altemate Shutdown / Control Room Panel Interactions' The licensee's IPEEE submittal contains a brief discussion addressing this issue in Section 4.8.5 of the submittal on Fire Risk Scoping Study lasues. In the discussion, the licensee states that the functional requirements for this issue are identical to those
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o covered by its treatment of Appendix R cable spreading room fire requirements. The licensee evaluated hot shorts leading to LOCAs and interfacing system LOCAs.. Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is reasonable and is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with thir 'ssue was identified in the IPEEE submittal, the staff considers this issue resolved.
- 2. GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness" l The licensee's IPEEE submittal contains information addressing this issue in Section .
4.8.3. The licensee performed a review of the FCS fire protection program against the I evaluation guidance provided in EPRI's FIVE on this issue. The licensee concluded that l the FCS fire protection program provides adequate assurance that fire will not significantly increase plant risk. Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is' reasonable and is capable of identify!ng potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers this I issue resolved for FCS.
- 3. GSI-156, " Systematic Evaluation Program (SEP)"
i The licensse's IPEEE submittal contains information to directly address the following external-event-related SEP issues: settlement of foundations and buried equipment !
(Sections 3.1.1.1, 3.1.4.1.2, 3.1.4.3, and 5.4); dam integrity and site flooding (Sections j 3.2.5.2, and 5.2.1 to 5.2.4); site hydrology and ability to withstand floods (Sections l 3.2.5.2 and 5.2); industrial hazards (Sections 5.3 and 5.4); tornado missiles (Section i 5.1); severe weather effects on structures (Sections 5.1 and 5.4); design codes, criteria and load combinations (Sections 3.1.3.1, 3.1.4.1.1, 3.1.4.1.2, 3.1.4.2.1, 5.1.3.1, 5.2.1, 5.2.3, and 5.3); seismic design of structures, systems and components (Sections 3.1.3.1,3.1.4.1.1,3.1.4.1.2, and 3.1.4.2); and shutdown systems and electrical instrumentation and control features (Sections 3.2.1 and 4.9.1). Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is reasonable and is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerabilities associated with this issue were identified in the licensee's l lPEEE submittal, the staff considers this issue resolved for FCS.
- 4. GSI-172, " Multiple System Responses Program (MSRP)"
The licensee's IPEEE submittal contains information directly addressing the following extemal-event-related MSRP lasues: effects of fire protection system actuation on safety-related and non safety-related equipment (Sections 4.8.1.2 and 4.8.1.3), smoke control and manual fire-fighting effectiveness (Section 4.8.3), heat / smoke / water propagation effects from fires (Sections 2.4.1, 2.4.2,4.8.4 and 4.9.3), effects of hydrogen line rupture (Sections 3.2.6 and 4.8.1.1), non safety-related control system / safety-related system dependencies (Sections 3.0,3.1.1.2,3.1.4.1,3.1.2.2, 3.1.4.2, 3.1.4.3, 3.2.5, 4.8.1, and 4.8.5); seismically induced spatial interactions (Sections 3.0, 3.1.1.2, 3.1.4.1, 3.1.2.2, 3.1.4.2, 3.1.4.3, 3.2.5, and 4.8.1 ), seismically induced fires (Sections 3.2.6 and 4.8.1.1), seismically induced flooding (Section 3.2.5),
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seismically induc.ed fire suppression system actuation (Sections 4.8.1.2 and 4.8.1.3),
seismically induced relay chatter Section 3.1.2.5), effects of flooding and/or moisture intrusion on non safety-related and safety-related equipment (Sections 3.2.5,4.8.1, 4.8.3,4.8.4,4.9.3 and 5.2), and evaluation of earthquake magnitude greater than safe shutdown earthquake (Section 3.1.3). Regarding IPEEE-related aspects of common cause failures associated with human errors, recovery actions for certain seismic scenarios are addressed in Sections 3.0, and 3.1.2. Similar information for fire scenarios may be found in Sections 4.0,4.6.7 and 4.8.4.
Based on the overall results of the IPEEE submittal review, the staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with GSI-172. Therefore, 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 the issue to be resolved for FCS.
The licensee also addressed severalissues that were not specifically required as a part of the IPEEE review as requested in Supplement 4 to GL 88-20 and NUREG-1407. These issues include USI A-17, " Systems Interactions in Nuclear Power Plants," USl A40, " Seismic Design Criteria," and GSI-106, " Piping and the Use of Highly Combustible Gases in Vital Areas."
USI A-17 and GSI-106 have been resolved in separate evaluations from the IPEEE program as indicated in NUREG-0933, "A Prioritization of Generic Safety lasues." Also, NUREG-0933 identifies four plants for which additional studies are needed to resolve USI-40, but since FCS is not one of the four plants identified, this issue does not apply to FCS. No other specific USIs or GSis were proposed by the licensee for resolution as part of the FCS IPEEE.
Uniaue Plant Features. Potential Vulnerabilities. and imorovements I
in Table 7.2 of its submittal, the licensee repor1ed three unique safety features at the plant that are cited as being design strengths. These are associated with extra protection from high winds (design for 300-400 mph), diesel generators independent of plant systems for cooling, and an early waming system for floods.
To identify potential vulnerabilities, the licensee used a set of closure criteria based on <
NEl 91-04, " Severe Accident lasue Closure Guidance (Rev.1)." These criteria are based on overall core damage frequency or containment failure frequency and on percent contributions to these overall frequencies. Based on these criteria, the licensee did not identify any potential vulnerabilities associated with seismic, fire, or other extemal events. However, a number of plant-specific improvements have been identified by the licensee as a result of the seismic, fire and HFO assessments. In the seismic area, certain tanks, electrical relays, and anchoring of electrical and mechanical components and heat exchangers were identified for modifications to increase the plant's capability to withstand earthquakes. As increased protection against flooding, provisions for portable pumps, new and more detailed flood procedures, upgrading of flood doors, sandbags, and sealing of conduits were initiated. For fire, actions wee initiated to add special procedures addressing spurious PORV openings due to fire and addressing interfacing LOCA in the shutdown cooling suction line, and to upgrade anchoring of cabinets containing flammable liquids and anchoring for a fire pump fuel oil tank, and to provide protection for sight glasses for the diesel generator day tanks. The submittal states that some of these improvements are completed and some are in progress. These procedural and design
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O improvements, listed in Table 7.1 of the licensee's submittal, are intended to improve plant safety and reduce the potential for seismic, fire and HFO vulnerabilities at FCS.
3.0 CONCLUSION
On the basis of the overall review findings, the staff concludes that: (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to Generic Letter 88-20 (and associated guidance in NUREG-1407), and (2) the IPEEE results are reasonable given the FCS 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 FCS IPEEE has met the intent of Supplement 4 to Generic Letter 88-20 and the resolution of specific generic safety issues discussed in this SER.
It should be noted that the staff focused its review primarily on the licensee's ability to examine FCS for severe accident vulnerabilities. A'Ithough certain aspects of the IPEEE were explored in more detail than others, the review was not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that underlie or stemmed from the examination. Therefore, this SER does not constitute NRC approval or endorsement of any IPEEE material for purposes other than those associated with meeting the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues discussed in this SER.
Principal Contributor: B. Hardin Date: May 6, 1999
4 TECHNICAL EVALUATION REPORT OF THE FORT CALHOUN STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL k.
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