ML20203C266
| ML20203C266 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/05/1997 |
| From: | Kugler A NRC (Affiliation Not Assigned) |
| To: | Gipson D DETROIT EDISON CO. |
| References | |
| GL-88-20, TAC-M83621, NUDOCS 9712150268 | |
| Download: ML20203C266 (6) | |
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v December 5, 1997 Mr. Douglas R. Gipson Senior Vice President Nuclear Generation Detroit Edison Company 6400 North Dixie Highway Newport, Michigan 48166
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (RAl) RELATED TO THE FERMI-2 INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS SUBMITTAL (TAC NO. M83621)
Dear Mr. Gipson:
On March 29,1996 (NRC-96-0037), Detroit Edison submitted the required response to Generic Letter 88 20, Supplement 4, " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," dated June 28,1991. Based on the staff's ongoing review of that submittal, additional information, as discussed in the enclosure, is requested in order for the staff to complete its review. The RAI is related to the fire and seismic areas of the IPEEE submittal. Currently, there are no questions in the high winds, flood, and other external events (HFO) area of the submittal. NRC requests that Detroit Edison respond within 60 days of the date of this letter, if you have any questions concerning this request, please contact me at (301) 415-2828 Sincerely, Original signed by:
Andrew J Kugler, Project Manager Project Directorate lil 1 Division of Reactor Projects ill/IV
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Office of Nuclear Reactor Regulation Docket No. 50-341
Enclosure:
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Mr. Douglas R. Gipson Fermi 2 Detroit Edison Company cc:
Jo;.n Flynn, Esquire Senior Attorney Detroit Edison Company 2000 Second Avnnue Detroit, Michigan 48226 Drinking Water and Radiological Protection Division Michigan Department of Environmental Quality 3423 N. Martin Luther King Jr Blyw P. O. Box 30630 CPH Mailroom Lansing, Michigan 48g0g-8130 U.S. Nuclear Regulatory Commission Resident inspector's Office 6450 W. Dixie Highway Newport. Michigan 48166 Monroe County Emergency Management Division g63 South Raisinville Monroe, Michigan 48161 Regional Administrator, Region lli
- U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Norman K. Peterson g
Director, Nuclear Licensing Detroit Edison Company Fermi 2 - 280 TAC 6400 North Dixie Highway Newport, Michigan 48166 w tw
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REQUEST FOR ADDITIONAL INFORMATION (RAl) RELATED TO =
- THE FERMI 2 INDMDUAL PLANT EXAMINATION FOR
' EXTERNAL EVENTS SUBMITTAL (TAC NO. M83621)
Fire -
c 1.
It is I,nportant that the human error probabilities (HEPs) used in the screening phase of the analysis properly reflect the potential effects of fire (e.g., smoke, heat, loss of -
lighting, and poor communication), even if these effects do not directly cause equipment damage in the scenarios being analyzed. - if these effects are not treated, the HEPs may be optimistic and result in the improper screening of scer arios Note that HEPs that are l
realistic with respect to an infomal events analysis could be optimistic with respect to a fire risk analysisc j
l-Please identify: (a) the scenarios screened out from further analysis whose
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quantification involved one or more HEPs, (b) the HEPs (descriptions and numerical values) for each of these scenarios, and (c) how the effects (e.g., smoke, heat, loss of l
lighting, and poor communication) of the postulated fires on HEPs were treated.
i 2.
NUREG-1407 (" Procedural and Submittal Guidance for the Individual Plant Examination of Extemal Events (IPEEE) for Severe Accident Vulnerabilities, Final Report *),
i Section 4.2 and Appendix C, and GL 88-20 (" Individual Plant Examination of Extemal Events (IPEEE) for Severe Accident Vulnerabilities"), Supplement 4, request that documentation be submitted with the IPEEE submittal with regard to the fire risk scoping t-study (FRSS) issues, including the basis and assumptions used to address these L
issues, and a discussion of the findings and conclusions. NUREG-1407 also requests l
that evaluation results and potentialimprovements be specifically highlighted. Control system interactions involving a combination of fire-induced failures and high probability random equipment failures were identified in the FRSS as potential contributors to fire risk.
i The issue of control systems interactions is associated primarily with the potential that a fire in the plant (e.g., the main control room (MCR)) might lead to potential control systems vulnerabilities. Given a fire in the plant, the likely sources of control systems interactions could happen between the MCR, the remote shutdown panel (RSP), and i
shutdown systems. Specific areas that have been identified as requiring attention in the resolution of this issue include:
j (a) Electricalindependence of the remote shutdown control systems: The primary concem of control systems interactione occurs at plants that do not provide independent remote shutdown control systems. The electricalindependence of ttie RSP and the evaluation of the level of indication and control of remote shutdown control and monitoring circuits need to be assessed ENCLOSURE i
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2 (b) Loss of control equipment or power before transfer. The potential for loss of I
control power for certain control circuits as a result of hot shorts and/or blown fuses before transferriag control from the MCR to remote shutdown locations needs to be assessed.
(c) Spurious actuation of components leading to componer:t damage, loss-of-coolant accident (LOCA), or interfacing systems LOCA The spurious actuation of one or more safety related to safe-shutdown-related components as a result of fire-induced cable faults, hot shorts, or component failures leading to component damage, LOCA, or interfacing systems LOCA, prior to taking control from the RSP, needs to be assessed. This assessment also needs to include the spurious starting and running of pumps as well as the spurious repositioning of valves.
(d) Totalloss of system function: The potentialfor totalloss of system function as a result of fire-induced redundant component failures or electrical distribution system (power source) failure nee 6 'o be addressed.
Please provide an evaluation of whether loss of control power due to hot shorts and/or blown fuses could occur prior to transferr ng control to the remote shutdown location and j
identify the risk contribution of these types of failures (if these failures are screened, please proside the basis for the screening). Finally, provide an evaluation of whether spurious actuation of cumponents as a result of fire-induced cable faults, hot shorts, or component failures could lead to component damage, a LOCA, or an interfacing systems LOCA prior to taking control from the RSP (considering both spurious starting and running of pumps as well as the spurious repositioning of valves).
3.
The previous question addresses the potential for hot shorts due to fires in the MCR.
Fires in other areas of the plant can also result in hot shorts that result in adverse conditions. Hot shorts in control cables can simulate the closing of control switches leading, for example, to the repositioning of valves, cpurious operation of motors and pumps, or the shutdown of operating equipment. These types of faults might, for example, lead to a LOCA, diversion of flow within various plant systems, deadheading and failure of important pumps, premature or undesirable switching of pump suction sources, or undesirable equipment operations. In instrumentation circuits, hot shorts may cause misleading plant readings potentially leading to inappropriate control actions or generation of actuation signals for emergency safeguard features. From the submittal, it cannot be determined to what extent the licensee has considered hot shorts as a failure mode for control or instrumentation cables. In particular, hot short considerations should include the treatment of conductor-to-conductor shorts within a given cable.
Discuss to what extent these issues have been considered in the IPEEE. If they have not been considered, please provide an assessment of how inclusion of poter.tial hot shorts would impact the quantification of fire risk scer,3rios in the IPEEE.
^ 3 4.
In the Electric Power Research Institute (EPRI) fire probabilistic risk analysis (PRA) implementation guide, test results for the control cabinet heat release rate have been misinterpreted and have been inappropnately extrapolated. Cabinet heat release rates as low as 65 Btu /sec are used in the guide.' In contrast, experimental work has -
developed heat release rates ranging from 23 to 1171 Btu /sec.
Considering the range of heat release rates that could be applicable to different control cabinet fires, and to ensure that cabinet fire areas are not prematurely screened out of the anal / sis, a heat release rate in the mid-range of the currently available expenmental L
data (e.g., 550 Btu /sec) should be used for the analysis.
Discuss the heat release rates used in your assessment of control cabinet fires. Please provide a discussion of changes in the IPEEE fire assessment results if it is assumed that the heat release from a cabinet fire is increased to 550 Btu /sec.
' 5.
On page 4 78 of the submittal, it is indicated that the quantitative screening approach used in the Fermi 2 fire assessment deviates from the fire-induced vulnerability evaluation (FIVE) methodology in that all equipment in a fire compartment is not
- assumed to be failed. - The submittal states that " obvious non-failures" were eliminated.
Three areas were screened in this phase of the analysis. Please indicate what equipment was assumed not to fail in any areas screened using this argument.
6.
Section 4.1.2 of the submittal indicates *. hat some minor discrepancies in the assessment were uncovered durino the final confirmatory walkdowns (e.g., equipment being located in wrong comnartments) but concludes that these errors would have no significant impact on the final results. Please provide a description of these discrepancies and a basis for the submittal's conclusion.
7.
The submittalindicates that some fire dampers have been declared inoperable. Please indicate if the inoperability of these dampers was included in the fire compartment interaction analysis (FCIA) portion of the fire assessment if credit was taken for these dampers, provide the basis for their assumed operation.
L Seismic 1.
Section 3.1.2.3.2 of the Fermi Unit 2 IFEEE submittal addresses control relays. Detroit Ediwn identified 13 low-seismic-ruggedness relays which perform a ' control interface
- function (Category 4). Based on a review of the affected circuits, Detroit Edison concluded that all 13 of these relays were ' acceptable for the seismic margins assessment (SMA) review.'
Please provide the applicable sections of Reference 3.42 and/or other documentation, as applicable, which provides the technical basis for the acceptability of the 13 control
- interface relays.
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2.
Sections 3.1.3 through 3.1.5 of the Fermi Unit 2 IPEEE submittal document Detioit -
Edisen's seismic evaluation in accordance with EPRI's SMA methodology for the IPEEE review level earthquake (RLE) (0.3g peak ground acceleration, NUREG/CR-0098
(* Development of Criteria for Seismic Review of Selected Nuclear Power Plants')
median-centered rock site spectrum). Considerable description of the approach, the results of the assessment, and the resolution of outliers is provided. However, the information provided is primarily of a qualitative nature. For example, while RLE in-structure response spectra were specifically developed for the IPEEE, these are not provided in the submittal. In addition, since evaluation to the RLE was apparently bad on extrapolation from design basis calculations, a comparison of the R.E in-structure response spectra to the design basis in-structure response spectra is necessary to assess Detroit Edison's approach. Lastly, the process of extrapolation from design calculations is only generally described. The conservative deterministic failure margin (CDFM) method of EPRI NP-6041-SL ("A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1") is not referenced as a basis for this extrapolation; consequently, the high-confidence low-probability of failure (HCLPF) capacities provided in the submittal have no defined technical basis.
To provide a more quan*.itative basis for assessment of Detroit Edison's approach, please glentify the five most limiting components or structural elements with respect to plant HCLPF capacity, and provide the following information:
a) the applicable RLE in-structuie response spectra b) the applicable design basis in-structure response spectra c) the calculations performed to determine the HCLPF capacity.
3.
For the following items, please provide the same (or comparable) information as requested in Seismic question #2 above:
a) Motor control center (limiting case) b) Control and instrumentation cabinet (limiting case) c) Masonry wall No. 212 d) Residual heat removal (RHR) heat exchanger supports e) Emergency diesel generator (EDG) fuel oil tank supports f) 4160 voit emergency switchgear/ bus High Wind, Flood and Other External Events (HFO) l 1.
There are 'lo questions in the HFO area.
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