ML20248A789
| ML20248A789 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/26/1998 |
| From: | Williams J NRC (Affiliation Not Assigned) |
| To: | James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| References | |
| GL-88-20, TAC-M83622, NUDOCS 9806010041 | |
| Download: ML20248A789 (10) | |
Text
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Mr. J:m:s Knubel May 26, 1998 Chi;f Nuclear Officer Power Authority of the Stata of New York j
123 Main Street
. White Plains, NY 10601
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT-REQUEST FOR ADDITIONAL INFORMATION REGARDING GENERIC LETTER 88-20, SUPPLEMENT 4, " INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS,"
Dear Mr. Knubel:
By letter dated June 26,1996, the New York Power Authority submitted results of the Individual Plant Examination of Extemal Events (IPEEE) for the James A. FitzPatrick Nuclear Power Plant in order to provide information requested by Generic Letter 88-20, Supplement 4. The NRC staff i
has determined that additional information will be required in order to complete its review. The l
additional information required is discussed in the enclosure to this letter.
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Your prompt response to this request will assist the staff in its effort to complete this review in a timely fashion. Please contact me at (301) 415-1470 if you have any questions on this topic.
Sincerely, Original Signed by:
Joseph F. Williams, Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Request for Additional Information cc w/ encl: See next page DISTRIBUTION:
LOooket File' S. Uttle W. Hardin, RES PUBLIC J. Williams PDI-1 R/F OGC i
J. Zwolinski ACRS S. Bajwa C. Hehl, Region I
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Mr. J:mes Knubel May 26, 1998 Chtf Nucle:r Officer Power Authority of tha Stato cf l
New York 123 Main Street White Plains, NY 10601 1
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT-REQUEST FOR l
ADDITIONAL INFORMATION REGARDING GENERIC LETTER 88-20, l
SUPPLEMENT 4, "lNDIVIDUAL Pl. ANT EXAMINATION OF EXTERNAL EVENTS,"
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Dear Mr. Knubel:
By letter dated June 26,1996, the New York Power Authority submitted results of the Individual Plant Examination of Extemal Events (IPEEE) for the James A. FitzPatrick Nuclear Power Plant in order to provide information requested by Generic Letter 88-20, Supplement 4. The NRC staff has determined that additional information will be required in order to complete its review. The additional information required is discussed in the enclosure to this letter.
Your prompt response to this request will assist the staff in its effort to complete this review in a timely fashion. Please contact me at (301) 415-1470 if you have any questions on this topic.
Sincerely, Original Signed by:
Joseph F. Williams, Project Manager Project Directorate I-1 Division of Reactor Projects - t/11 Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Request for Additional Information cc w/ encl: See next page DISTRIBUTION:
Docket File S. Little W. Hardin, RES PUBLIC J. Williams PDI-1 R/F OGC J.' Zwolinski ACRS S. Bajwa C. Hehl, Region 1 j
l DOCUMENT NAME: G:\\FITZWl83622A.RAI j
To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with I
ctt:chment/ enclosure "N" = No copy n
- l OFFICE PM:PDI-1 ff2 lE LAPDI-1s 1 !
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SBeiwe /' E N lDATE 05 @ /98 05N48 06/')l4 8 I
Official Record Copy
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UNITED STATES g
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2 WASHINGTON, D.C. 2006H001 8
%,*****,o May 26, 1998 t
Mr. James Knubel Chief Nuclear Officer Power Authority of the State of New York 123 Main Street White Plains, NY 10601
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT-REQUEST FOR ADDITIONAL INFORMATION REGARDING GENERIC LETTER 88-20, SUPPLEMENT 4, " INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS,"
Dear Mr. Knubel:
By letter dated June 26,1996, the New York Power Authority submitted results of the Individual Plant Examination of Extemal Events (IPEEE) for the James A. FitzPatrick Nuclear Power Plant in order to provide information requested by Generic Letter 88-20, Supplement 4. The NRC staff has determined that additional information will be required in order to complete its review. The additional li, formation required is discussed in the enclosure to this letter.
Your prompt response to this request will assist the staff in its effort to complete this review in a timely fashion. Please contact me at (301) 415-1470 if you have any questions on this topic.
Sincerely, Joseph F. Wi!Iiams, Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Request for Additional Information l
1 cc w/ encl: See next page I
James Knubel James A. FitzPatrick Nuclear Power Authority of the State Power Plant of New York cc:
Mr. Gerald C. Goldstein Regional Administrator, Region i Assistant General Counsel U.S. Nuclear Regulatory Commission Power Authority of the State 475 Allendale Road of New York King of Prussia, PA 19406 1633 Broadway New York, NY 10019 Mr. F. William Valentino, President New York State Energy, Research, Resident inspector's Office and Development Authority U. S. Nuclear Regulatory Commission Corporate Plaza West P.O. Box 136 286 Washington Avenue Extension Lycoming, NY 13093 Albany, NY 12203-6399 Mr. Harry P. Salmon, Jr.
Mr. Richard L. Patch, Director Vice President - Engineering Quality Assurance Power Authority of the State Power Authority of the State of New York of New York 123 Main Street 123 Main Street White Plains, NY 10601 White Plains, NY 10601 Ms. Charlene D. Faison Mr. Gerard Goering Director Nuclear Licensing 28112 Bayview Drive Power Authority of the State Red Wing, MN 55066 of New York 123 Main Street Mr. James Gagliardo White Plains, NY 10601 Safety Review Committee 708 Castlewood Avenue Supervisor Arlington, TX 76012 Town of Scriba Route 8, Box 382 Mr. ArthurZaremba, Licensing Manager Oswego, NY 13126 James A. FitzPatrick Nuclear
(
Power Plant Mr. Eugene W. Zeitmann P.O. Box 41 President and Chief Operating Lycoming, NY 13093
)
Officer
{
Power Authority of the State Mr. Paul Eddy
]
of New York New York State Dept. of 99 Washington Ave., Suite No. 2005 Public Service Albany, NY 12210-2820 3 Empire State Plaza,10th Floor Albany, NY 12223 Charles Donaldson, Esquire Assistant Attomey General New York Department of Law 120 Broahay New York, NY 10271
REQUEST FOR ADDITIONAL INFORMATION GENERIC LETTER 88-20. SUPPLEMENT 4 INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 A.
Seismic 1.
It is stated in the submittal that: "The existing unreinforced concrete block walls were previously evaluated in the IE Bulletin 80-11 program, in which response spectrum analysis was performed using the 1 percent DBE floor spectra and the allowable tensile strength of 35.4 psi. In the IPEEE analysis, the High Confidence Low Probability of Failure (HCLPF) capacities of the block walls were estimated by extrapolating the foregoing linear analysis results by raising the damping value to 7 percent and the allowable stress to 42.5 psi (1.7 times the design allowable stress of 25 psi)." In Section 8.3 of the submittal, it is stated that the block walls in the emergency diesel generator (EDG) building, (EGB-272-6,7,9 and 10) will be strengthened. The following specific questions relate to the capacity evaluation and strengthening of the unreinforced concrete block walls:
Provide the technical basis, such as past test results or detailed analytical a.
evaluation, for the assumed damping value of 7 percent and the allowable stress of 42.5 psi, and for the extrapolation from the Bulletin 80-11 linear analysis results to obtain HCLPF estimates.
b.
Please provide information on the current status of the effort to strengthen the block walls in the EDG building.
2.
Regarding the relay chatter evaluation on pp. 3-31 of the submittal, the submittal states that, " relays that are either bad-actor relays or are not covered by the generic equipment response spectra (GERS) were assumed to chatter during an earthquake"(i.e., a failure probability of 1.0 is implied). On the other hand, on pp. 3-81 of the submittal, a failure probability of 0.1 is implied for screening criteria.
Please explain this inconsistency. Also, please provide the schedule for resolving the issue of bad actor relays in the EDG building.
3.
Regarding the potential rupture of a hydrogen line during an earthquake event, the submittal states that the operating procedure has been changed to close the hydrogen supply line at supply points in the event of an earthquake.
Considering the swift onset of an earthquake and the potentially rapid progress of a hazardous hydrogen explosion, please describe the timing of operator actions needed to close the hydrogen supply lines and justify that this simple procedural change is sufficient to prevent a potential fire (i.e., explain why this procedural change alone is sufficient and and how it will be implemented in the hazardous environment likely to exist following a seismic event).
Enclosure
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Time (s) f the analysis using a more justifiable value and provide the resulting change in scenario contribution to core damage frequency.
Figure R.1 Sensitivity of the hot gas layer temperature predictions to the assumed heat loss factor 2.
The Fitzpatrick submittal assumes propagation time delays of 5 minutes for tray-to-tray fire spread and 15 minutes for fire spread from an electrical panel. The basis upon which these propagation time delays have been assumed has not been provided in particular, these assumptions have been applied in fire zones RR-1, CS-1, CT-3/RR-1, RR-1/AD-4, RR-1/TB-1, and RR-1/RB-1 A. These assumptions pre-set the fire growth and propagation time, which in tum impacts the time-dependent rate of heat release and the resulting critical damage times, without appropriate consideration of the specifics of a given fire scenario. In particular, the propagation behavior of a fire will be influenced by a number of case-specific factors including the intensity and duration of the initial source (or exposure) fire, the proximity of the secondary fuels to the initial source (or exposure) fire, the fire behavior properties of the cables or other secondary fuels (e.g., ignition temperature, heat of combustion, total fuel mass), the physical configuration of the exposed trays (e.g., vertical versus horizontal, exposed surface area, and the density of the cable loading), proximity of the fire source and secondary fuels to walls and/or the ceiling, and the room-specific ventilation conditions.
For each fire scenario in which fixed propagation delays were used to estimate the rate and extent of fire propagation, please: (a) indicate if FIVE (or similar) calculations were performed for the scenario and provide the results (equipment or cables damaged) of these calculations; and (b)if experimental data were used, please indicate, 1
. which experimental results were used, how they were utilized fn the analysis, justification of the use of these experimental results to the scenario being
=
analyzed.
The discussion of results applicability should compare the geometries, ignition sources, fuel type and loadings, ventilation characteristics, and compartment characteristics of the experimental setup (s) with those of the scenario of interest.
3.
Fires in the main control room (MCR) are potentially risk-significant because they can cause l&C failures (e.g., loss of signals or spurious signals) for multiple redundant divisions, and because they can force control room abandonment. Although data from two experiments conceming the timing of smoke-induced, forced control room abandonment is available (Reference 3) the data must be carefully interpreted, and the analysis must proper 1y consider the differences in configuration between the experiments and the actual control room being evaluated for fire risk. In particular, the experimental configuration included placement of smoke detectors inside the cabinet in which the fire originated, as well as an open cabinet door for that cabinet. In one case, failure to account for these configuration differences led to more than an order of magnitude underestimate in the conditional probability of forced control room fire abandonment (Reference 4). In addition, another study raises questions about control room habitability due to room air temperature concems (Reference 5).
Please provide the detailed assumptions (including the assumed fire frequency, any frequency reduction factors, and the probability of abandonment) used in analyzing the MCR and justifications for these assumptions. In particular, if the probability of abandonment is based on a probability distribution for the time required to suppress the fire, please justify the parametric form of the distribution and specify the data used to quantify the distribution parameters. Alternatively, estimate the sensitivity of the fire CDF to the probability of control room abandonment by assuming a value based on the FitzPatrick control room characteristics.
4.
The FitzPatrick submittal utilizes an approach to the analysis of electrical cabinet fires that is the same as the approach recommended by the EPR/ Fire PRA / implementation Guide.
Enclosed ignition sources cannot lead to fire propagation or other damage.
Fire spread to adjacent cabinets cannot occurif the cabinets are separated by a double wall with an air gap or if the cabinet in which the fire originates has an open top.
Oil-filled transformers and high-voltage components in cabinets, for example, are susceptible to energetic faults leading to cabinet breach. Switchgear fires at Yankee-Rowe in 1984 and Oconee Unit 1 in 1989 both resulted in fire damage outside the cubicles. Cabinets are also susceptible to warping underintense heat loads, which wc,uld invalidate any assumption of limited combustion air. Assumptions in the FitzPatrick submittal on fire propagation from enclosed ignition sources should be verified, especially in such typically important areas as the re!a/ room and cable spreading room.
.. Please provide the basis for the assumption and a discussion on how the specific enclosures were analyzed to ascertain that the assumption is applicable to them.
5.
The FitzPatrick submittal assumes cabinet heat release rates of 65 Blu/sec. In contrast, experimental work has developed heat release rates ranging from 23 to 1171 Btu /sec.
Considering the range of heat release rates that have been observed and cou'd be applicable to different control cabinet fires, appropriate values should be selected to ensure that cabinet fire areas are not prematurely screened out of 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 500 Btu /sec.
6.
The submittal notes that non-lEEE383-qualified cabling is present in the plant, but that j
installations since construction have used only lEEE383-qualified cable. The description of the cable parameters, 900* F for ignition, 732' F for damage, are more typical of qualified cable.
Please identify the fire zones and scenarios where non-lEEE383. qualified cable is present, the ignition and damage temperatures r peropriate to that 1
cable, and the impact on the results of the fire study if those parameters are I
assumed. Identify and estimate the fire CDF contribution from any scenarios screer.ad inappropriately by assigning qualified cable properties to non-qualified cable.
7.
Section 4.7.3.1 cites the EPRI Fire PRA Implementation Guide methodology for the treatment of automatic suppression system unavailability ir. the multi-zone analysis. This reference treats manual recovery of automatic suppression systems as being independent of subsequent manual efforts to suppress the fire. This assumption is optimistic, as the fire conditions (e.g., heat, smoke) that lead to the failure of recovery efforts can also influence the effectiveness of later suppression efforts. Such an approach, therefo;e, can overlook plant-specific vulnerabilities.
It is important that all relevant factors be considered in an evaluation of the effectiveness of fire suppression. These factors include: (a) the delay between ignition and rYector/ suppression system actuation (which is specific to the configuration being analyzed); (b) the time-to-damage for the critical component (s) (which is specific to the fuel type and loading as well as to the configuration being modeled); (c) the resper.se time of the firo brigade (which is plant-specific and fire-location-specific); (d) the time required by the fire brigade to diagnose that automatic suppression has failed and to take manual action to recover the automatic suppression system; r.nd, (e) performance shaping factors (PSFs) affecting fire bripade actions. These PSFs could include factors such as perseverance (persistent efforts made to recover a failed automatic suppression system), smoke obscuration, and impaired communications (Reference 4).
Finally, it should be noted that the Nuclear Regulatory Commission (NRC) staff's evaluation of the FIVE methodology specifically stated that licensees need to assess the effec'iveness of manual fire-fighting teams by using plant-specific data from fire brigade training to determine the response time of tbs fire fighters.
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5-Please identify those scenarios for which credit is taken for both manual recovery of automatic suppression systems and manual suppression of the fires (if manual recovery efforts are unsuccessful), and please indicate the plant equipment that l
may bs affected by the fires. In the analysis of these scenarios, how are dependencies between manual actions treated? Please justify the treatment, considering the expected fire environment, the recovery actions required, and the manual fire suppression actions required.
1 8.
The failure probability for automatic suppression applied values compatible with the FIVE methodology. This data is acceptable for systems that have been designed, installed, I
and maintained in accordance with appropriate industry standards, such as those i
published by the National Fire Protection Association (NFPA).
Please verify that automatic firs suppression systems at FitzPatrick meet NFPA standards.
C.
High winds, flood, and other external events i
1.
It is stated in the submittal that an analysis has indicated that, for the condition of probable maximum precipitation (PMP), the load capacity of the reactor building roof (50 2
lb/ft ) will only be exceeded if two of three roof drains on one side of the roof are blocked and, as a result, the water depth exceeds 9.6 inches.
Please provide information on any plant procedures that address the prevention of blockage of the reactor building roof drains, or if there are no such procedures, i
provide justification for their ommission.
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. References R1.
P.J. DiNenno, et al, eds., "SFPE Handbook of Fire Protection Engineering," 2nd Edition, National Fire Protection Association, p. 3-140,1995.
R2.
L Y. Cooper, M. Harkleroad, J. Quintiere, W. Rinkinen, "An Experimental Study of Upper Hot Layer Stratification in Full-Scale Multiroom Fire Scenarios," ASME Joumal of Heat Transfer,1Qi, 741-749, November 1982.
R3.
J. Chavez, et al., "An Experimental Investigation of Intemally Ignited Fires in Nuclear Power Plant Cabinets, Part Il-Room Effects Tests," NUREG/CR-4527N2, October 1988.
R4.
J. Lambright, et al., "A Review of Fire PRA Requantification Studies Reported in NSAC/181," prepared for the United States Nuclear Regulatory Commission, April 1994.
RS.
J. Usher and J. Boccio,
- Fire Environment Determination in the LaSalle Nuclear Power Plant Control Room," NUREGICR-5037, prepared for the United States Nuclear Regulatory Commission, October 1987.
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