ML20216J092
| ML20216J092 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/18/1998 |
| From: | Hansen A NRC (Affiliation Not Assigned) |
| To: | Jeffery Wood CENTERIOR ENERGY, TOLEDO EDISON CO. |
| References | |
| GL-96-06, GL-96-6, TAC-M83613, TAC-M96803, TAC-MA0414, TAC-MA414, NUDOCS 9803230350 | |
| Download: ML20216J092 (11) | |
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NUCLEAR REQULATORY COMMISSION WASHINGTON, D.C. 30selN1001 March 18, 1998 Mr. John K. Wood Vice President - Nuclear, Davis-Besse Centerior Service Company clo Toledo Edison Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, OH 43449-9760
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO: GENERIC LETTER 96-06,
" ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" (TAC NO. M96803); REQUEST FOR RELIEF FROM CERTAIN ASME CODE REQUIREMENTS (DAVIS-BESSE SERIAL NO. 2506, TAC NO. MA0414); AND INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS (IPEEE)
Dear Mr. Wood:
Generic Letter (GL) 96 06 The staff has your submittals dated January 28,1997, February 28,1997, July 28,1997, and September 30,1997, regarding GL 96-06, " Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions." Your January 28,1997, submittal provided the initial response to the GL for the Davis-Besse Nuclear Power Station.
The February 28,1997, submittal provided additional detail regarding actions taken to address the issues of water hammer, two-phase flow, and thermally-induced pressurization of piping runs penetrating the containment. The submittal of July 28,1997, provided a status update.
The September 30,1997, submittal provided a report on the resolution of the issues addressed by the GL.
In your submittals, you indicated that a total of 13 pipe lines penetrating the containment were p
susceptible to thermally-induced overpressurization. You indicated that bypass check valves would be installed on five lines and a relief valve on one other line. In addition, you initiated procedural changes to address the issue on three other lines. You determined that the
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- remaining four lines were acceptable based on further evaluation.
In order to complete its review of your responses to the GL, the staff needs the additional information described below regarding the evaluation of these four lines.
In the submittal of September 30,1997, you indicated that two pipe lines meet ASME Code pressure limitation under faulted-load combinations (penetrations 14 and 56). Please provide the following information for the piping runs associated with these lines:
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J. Wood.
Provide the applicable design criteria for the piping and the valves. Include the required load combinations.
Provide a drawing of the piping run between the isolation valves. Include the lengths and thicknesses of the piping segments and the type and thickness of the insulation.
Provide the maximum calculated temperature and pressure for the pipe run. Desenbe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model used in the analysis and the basis for the j
heat transfer coefficients used in the analysis.
In the submittal of September 30,1997, you indicated that the air-operated isolation valves associated with penetration 21 and 32 pipe lines provide inherent relief to prevent overpressure.
Please provide the following information for each pipe line:
Describe the applicable design criteria for the piping and valves. Include the required load combinations.
Provide a drawing of the valve. Provide the pressure at which the valve was determined to lift off its seat or leak and describe the method used to estimate this pressure. Discuss any sources of uncertainty associated with the estimated liftoff or leakage pressure.
Provide the maximum calculated stress in the piping run based on the estimated liftoff or leakage pressure.
- Relief Requests The staff is currently reviewing your December 22,1997, requests for relief from certain American Society for Mechanical Engineers (ASME) Code requirements for inservice inspection at Davis-Besse. Please provide the following additionalinformation:
RR-E2: In the basis for relief section, you state that the use of CP-189 will not improve the capability of examination personnel. Inclusion of these standard in the 1992 Edition and later Edition and Addenda of the Code suggests that it is an experience-based improvement over the SNT-TC-1 A. Also, it is a consensus (ANSI) standard. Please provide your plans to implement Subarticle IWA-2300 (1992 Edition and Addenda) for qualifying NDE personnel for all the ASME Section XI components, and containment in particular.
RR-E3 and E4: In the bases for relief section for these two relief requests, you state that the Davis-Besse coating maintenance program will provide an adequate level of safety and '
quality. With respect to the DBNPS coating program, please provide the following information related to the Service Level 1 coating (that is, protective coating (s) used inside the containment):
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Standards, procedures, or guidelines used for (1) surface preparation, application, surveillance, and maintenance activities for protective coatings; and (2) documenting the quality and condition of reapplied paint. Maintenance activities include reworking of degraded coating, removing of degraded coating to sound coating, preparing the surface, I
applying new coating, and verifying the coating quality.
RR-E6: You are requesting relief from the requirements of IWE-2420(b) and (c), related to successive inspections. The request does not provide any discussion of quality and safety, when the flaws and degradation areas are accepted by engineering analysis (lWE-3122.3).
l Discuss your plans for monitoring the progression of the flaws and degradation areas when they are accepted by engineering analysis.
RR-E7: Before processing this relief request, the staff needs to know how the preload and condition of the bolts of the pressure-unseating containment penetrations (which would tend to induce tension in the bolts under accident conditions) are monitored and maintained.
i Provide a summary of such penetrations in the Davis-Besse containment and procedures used to monitor and maintain the functionality of their bolts.
IPEEE The ongoing staff review of your IPEEE submittal has identified a need for additional l
information related to the fire and seismic areas. The request for additional information (RAI) j on fire issues was developed by Sandia National Laboratories, and the RAI on seismic issues was developed by Brookhaven National Laboratories, both under contract with NRC. All
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questions were reviewed by the Senior Review Board, which is comprised of NRC staff and Sandia consultants with probabilistic risk assessment expertise in extemal events.
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- 1. NUREG-1407, Section 4.2 and Appendix C, and Generic Letter (GL) 88-20, Supplement 4, request that documentation be submitted with the IPEEE submittal with regard to the fire risk scoping study (FRSS) issues, including the basis and assumptions used to address these issues and a discussion of the findings and conclusions. NUREG-1407 also requests that evaluation results and potential improvements 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.
The issue of control systems interactions is associated primarily with the potential that a fire in the plant (for example, the main control room (MCR)) might lead to control systems vulnerabilities. Given a fire in the plant, the likely sources of control systems interactions are between the control room, the remote shutdown panel, and shutdown systems. Specific areas that have been identified as requiring attention in the resolution of this issue include:
(a) Electricalindependence of the remote shutdown control systems: The primary concem of control systems interactions occurs at plants that do not provide independent remote l
J. Wood shutdown control systems. The electricalindependence of the remote shutdown panel (RSP) and the evaluation of the level of indication and control of remote shutdown control and monitoring circuits need to be assessed.
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(b) Loss of control equipment or power before transfer. The potential for loss of control I
power for certain control circuits as a result of hot shorts and/or blown fuses before transferring control from the MCR to remote shutdown locations needs to be assessed.
(c) Spurious actuation of components leading to component damage, a loss-of-coolant accident (LOCA), or an 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, a LOCA, i
or an interfacing systems LOCA before control is transferred to the remote shutdown panel 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) Total loss of system function: The potential for total loss of system function as a result l
of fire-induced redundant component failures or electrical distribution system (power source) failure r:seds to be addressed.
j Describe how your procedures provide for transfer of control to the remote station (s).
Provide an evaluation of whether loss of control power due to hot shorts and/or blown fuses could occur prior to transferring control to the remote shutdown location and identify the risk contribution of these types of failures (if these failures are screened, please provide the basis for the screening). Finally, provide an evaluation of whether spurious actuation of l
components as a result of fire-induced cable faults, hot shorts, or component failures could j
lead to component damage; a LOCA, or an interfacing systems LOCA prior to taking control from the RSP (considering spurious starting and running of pumps as well as the spurious repositioning of valves).
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rate have been misinterpreted and have been inappropriately 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.
l Considering the range of heat release rates that could be applicable to different control l_
cabinet fires, and to ensure that cabinet fire areas are not prematurely screened out of the l-analysis, a heat release rate in the mid-range of the currently available experimental data (for example, 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.
- 3. Fire severity factors were used for analyses of control room electrical cabinets, switchgear I
ioom electrical cabinets, indoor transformers, diesel generators (skid fires), motor generator l
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I J. Wood sets, pumps (other than reactor coolant pumps (RCPs) and mean feedwater (MFW)), and ventilation subsystems based on the Fire PRA Implementation Guide. The severity factors were used to adjust the basic ignition frequencies of the associated components for those areas surviving screening. These adjusted ignition frequencies were appartsntly used in l
scenarios where fire suppression was credited, including the detailed analyses for compartments G.02 and ll.01 (turbine building). Since the success of fire suppression would reduce the potential for a large fire, there appears to be a significant possibility that the use of a fire severity factor, when fire suppression is modeled, double counts for suppression efforts.
Please describe the fire scenarios in which automatic fire suppression was credited in conjunction with the fire severity factors used in the Davis-Besse fire assessment. For each case explain why such credit does not constitute double counting for suppression.
- 4. Hot shorts in control cables can initiate the closing of control switches, leading, for example, I
to the repositioning of valves, spurious 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. For MCR abandonment scenarios, such spurious operations and actions may not be indicated at the RSP(s), may not be directly recoverable from remote shutdown locations, or may lead to the loss of remote shutdown capability (for example, through loss of RSP power sources). In instrumentation circuits, hot shorts may cause misleading plant readings, potentially leading to inappropriate control actions or generation of actuation J
signals for emergency safeguard features.
The submittal indicates that hot shorts were considered in the assessment. However, this assessment included the possibility of a single wire shorting to ground for cases where this I
could cause a component to fail to its undesired state. This possibility is not consistent with the definition of a hot short, which refers to applying electric power to an unpowered component. In addition, a probability of 0.2 was assigned to this event, a value which is probably optimistic since a fire is very likely to lead to a short to ground if not suppressed in the early stages.
Discuss to what extent the shorts to ground affected the results of screening and detailed analyses in the IPEEE. Also, assess the effects of assuming that the probability of a short to ground is 1.0 rather than 0.2. For cases where hot shorts (other than shorts to ground) were considered, discuss their effect on the quantification (screening and detailed analyses) of fire risk scenarios in the IPEEE.
- 5. The turbine building (fire area 11) and the auxiliary feedwater (AFW) pump rooms (fire areas E and F) are treated as separate areas in the fire analysis, apparently on the basis that they are enclosed by adequate boundaries. However, the MFW pumps in the turbine building are separated from the AFW pumps by a water curtain. As a result, the possibility of fire j
propagation between the turbine building and AFW pump rooms should have been considered in the fire compartment interaction analysis (FCIA).
J. Wood.
Please include the above areas / compartments in the FCIA and describe the results of the revised analysis, including any impact on the subsequent quantitative screening and detailed analyses.
- 6. The analysis of compartment A.05 states that cables of both trains of AFW and emergency core cooling system (ECCS) pumps are located in this compartment. Due to the train separation and fire detection and suppression available in the room, it is assumed, based on the fire hazards analysis report (FHAR), that a fire would only cause the loss of one train of each of these systems.
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Please analyze the effect of the above assumption on the IPEEE results by assessing the probability that a fire could cause a failure of both trains of one or both of these systems, leading to core damage.
- 7. In the event of a control room fire, reentry is credited for the operation of long-term heat removal functions (provided associated controls are not damaged) which would not be required for several hours. It is assumed that operators could reoccupy the control room "within no more than a few hours (nominally 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) after the evacuation." It is also I
stated in Section 4.2.6.3 that "it was assumed that there was negligible probability that the control room could not be occupied before long-term actions to preserve core cooling (those that would.be relevant 12 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the control room was evacuated) would be needed." In the control room summary (Section 4.2.6.6), it is noted that even if the control room must be evacuated, options for core cooling remain available. The reoccupation of the control room is again assumed "such that additional options could be implemented for long-term cooling."
Please discuss the basis for assuming that the control room could be re-entered in the event of a fire. Include a description of the fire scenarios which are dependent on this assumption. For each scenario, describe the systems that would be available, consistent with the associated fire suppression time and assumed damage, for successfully maintaining long-term core cooling. In the analysis, consider available systems and controls inside or outside the control room, and discuss the potential for success of the options l
available for each scenario.
- 8. It is important that the human error probabilities (HEPs) used in the screening phase of the analysis properly reflect the potent lal effects of fire (for example, smoke, heat, loss of lighting), 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 scenarios. Note that HEPs which are conservative with respect to an intemal events analysis could be nonconservative with respect to a fire risk analysis.
Please identify: (a) the scenarios screened out from further analysis whose quantification involved one or more HEPs; (b) the HEPs (descriptions and numerical values) for each of these scenarios; and (c) how the effects of the postulated fires were treated.
J. Wood.
Seismic
- 1. Davis-Besse has been identified in NUREG-1407 as a plant belonging to the 0.3g focused-scope seismic margin assessment bin. Hence, the reduced-scope evaluation at 0.15g, as performed in the Davis-Besse seismic IPEEE, does not conform to the review guidance in NUREG-1407 and Supplement 4 to Generic Letter 88-20. Accordingly:
ac Provide a list of structures, systems, and components (including safe shutdown equipment list (SSEL) items and containment systems equipment) that did not screen at j
the 0.3g review level earthquake (RLE).
- b. Provide the basis for disposition of each item that did not screen at 0.3g RLE, including the results of new calculations for high confidence low probability of failure (HCLPF) capacity.
- c. Provide an evaluation to the 0.3g RLE of masonry / block walls that may influence the perfurmance of success path components.
- d. Provide an evaluation to the 0.3g RLE of flat-bottomed tanks, as requested in NUREG-1407 and GL 88-20 for focused-scope plants. Address tank failures themselves as well as flooding concems resulting from tank failures.
- e. Provide figures comparing the design basis ground spectrum and in-structure response spectra (IRS) to the IPEEE 0.3g peak ground acceleration (pga) RLE ground spectrum and in-structure response spectra (IRS). If scaling was used, describe the scaling method. If new IRS were generated, describe the analyses performed to generate all significant RLE IRS.
- 2. Additional information and clarification are needed for the success path selection described in the submittal:
- a. The SSEL provided as part of the submittal includes high-pressure injection (HPI) ptimp 1-1 and its peripheral equipment, but the second HPl train is apparently not on the list.
Please clarify if the second HPl train is qualified for the success paths, and if not, please justify its omission.
- b. Please provide a discussion of how the hot leg vents provide the reactor coolant system (RCS) pressure control function. Do these vents have enough capacity? Is the action automatic, or do the operators have to manually open such vents? If the latter, discusa the reliability of such action during a seismic event in view of the timing, stress, any alarms, etc.
' 3. Nonseismic failures and human actions are not specifically addressed in the Davis-Besse submittal. Regarding nonseismic failures and human actions, NUREG-1407 states that
" success paths are chosen based on a screening criterion applied to nonseismic failures and needed human actions. It is important that the failure modes and human actions are I
1 J. Wood q clearly identified and have low enough probabilities to not affect the seismic margins evaluation." Please describe how nonseismic failures and human actions were treated in j
the Davis-Besse analysis and address the concerns of the above quoted statement from NUREG-1407.
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Hiah Wind. Flood and Other Extemal Events (HFO)
There is no RAI in the HFO area.
Please provide your responses to the above requests within 120 days of receipt of this letter.
You may contact me at 301415-1390 if you have any questions.
Sincerely,
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Allen G. Hansen, Project Manager Project Directorate lll-3 Division of Reactor Projects lil/lV Office of Nuclear Reactor Regulation Docket No. 50-346 cc: See next page l
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.g' J. Wood i lt is important that the failure modes and human actions are clearly identified and have low enough probabilities to not affect the seismic margins evaluation." Please describe how j
nonseismic failures and human actions were treated in the Davis-Besse analysis and address the concems of the above quoted statement from NUREG-1407.
Hiah Wind. Flood and Other Extemal Events (HFO) i There is no RAIin the HFO area.
Please provide your responses to the above requests within 120 days cf receipt of this letter.
j You may contact me at 301-415-1390 if you have any questions.
Sincerely, Original signed by:
i Allen G. Hansen, Project Manager Project Directorate lll-3 Division of Reactor Projects Ill/IV i
l Office of Nuclear Reactor Regulation Docket No. 50-3/S l
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I J. Wood ~ !
l It is important that the failure modes and human actions are clearly identified and have low enough probabilities to not affect the seismic margins evaluation." Please describe how nonseismic failures and human actions were treated in the Davis-Besse analysis and address the concerns of the above quoted statement from NUREG-1407.
- Hiah Wind. Flood and Other External Events (HFO)
There is no RAI in the HFO area.
Please provide your responses to the above requests within 120 days of receipt of this letter.
You may contact me at 301-415-1390 if you have any questions.
Sincerely, l
Original signed by:
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Allen G. Hansen, Project Manager Project Directorate lll-3 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation I
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Docket No. 50-346 I
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.e John K. Wood Davis-Besse Nuclear Power Station, Dnit 1
- Toledo Edison Company cc:
Mary E. O'Reilly Robert E. Owen, Chief FirstEnergy.
Bureau of Radiological Health Davis-Besse Nuclear Power Station.
Service 5501 North State - Route 2.
Ohio Departenent of Health Oak Harbor, OH 43449-9760 P.O. Box 118 Columbus, OH 43266-0118 James L. Freels Manager-Regulatory Affairs Toledo Edison Company James R. Williams, Chief of Staff Davis-Besse Nuclear Power Station Ohio Emergency Management Agency 5501 North State - Route 2 2855 West Dublin Granville Road Oak Harbor, OH 43449-9760 Columbus, OH 43235-2206 Gerald Charnoff, Esq.
Donna Owens, Director Shaw, Pittman, Potts Ohio Department of Commerce and Trowbridge Division of Industrial Compliance 2300 N Street, NW.
Bureau of Operations & Maintenance Washington, DC 20037 6606 Tussing Road i
P.O. Box 4009 Regional Administrator Reynoldsburg, OH 43068-9009 U.S. Nuclear Regulatory Commission 801 Warrenville Road Ohio Environmental Protection Agency Lisle, IL 60523-4351 DERR-Compliance Unit ATTN: Zack A. Clayton l
Robert B. Borsum P.O. Box 1049 Babcock & Wilcox Columbus, OH 43266-0149 Nuclear Power Generation Division 1700 Rockville Pike, Suite 525 State of Ohio Rockville, MD 20852 Public Utilities Commission 180 East Broad Street Resident inspector Columbus, OH 43266-0573 U.S. Nuclear Regulatory Commission 5503 North State Route 2 Attomey General Oak Harbor, OH 43449 Department of Attorney 30 East Broad Street James H. Lash, Plant Manager Columbus, OH 43216 Toledo Edison Company Davis-Besse Nuclear Power Station President, Board of County 5501 North State Route 2 Commissioner of Ottawa County Oak Harbor, OH 43449-9760 Port Clinton, OH 43252 e