ML20248C573
| ML20248C573 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/19/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248C400 | List: |
| References | |
| GL-88-20, NUDOCS 9806020191 | |
| Download: ML20248C573 (7) | |
Text
,-a.ee Enclosure
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PLANT SYSTEMS BRANCH STEP 1 REVIEW 0F INDIVIDUAL PLANT EXAMINATION OF EXTERNAL FIRE EVENTS MILLSTONE UNIT 3 DOCKET NO. 50-423
1.0 INTRODUCTION
Northeast Nuclear Energy Company (the licensee) submitted on December 23, 1991, their response to Generic Letter 88-20, Supplement 4 for.
Millstone Unit 3. On October 13, 1993, the Plant Systems Branch completed its step 1 review of their August 1990, individual plant examination of external events (IPEEE) of severe accident vulnerabilities for Millstone Unit 3.
As result of this review on April 3, 1994, additional information regarding their IPE was requested.
By letter dated June 4, 1994, the licensee provided this additional information.
2.0 DISCUSSION
The licensee used the results of the Millstone Unit No. 3 Probabilistic Safety Study (PSS) as the basis for their IPEEE review. The PSS is based on a Level III Probabilistic Risk Assessment (PRA). Therefore, the licensee used this PRA approach (PSS Sections 1.2.2 and 2.5.2) to determine the affect an external fire event in a critical plant area may have on the core melt frequency.
The major IPEEE fire analysis steps performed were: (1) selection of critical fire areas; (2) estimation of fire frequency in critical areas; and (3) quantification of fire related core-melt frequencies.
SPLB found the licensee's methodology for determining the results associated with these major steps of the fire analysis to be adequate.
3.0 EVALUATION
The Plant Systems Branch performed an initial review of the Millstone Unit 3 August 31, 1990 submittal, " Individual Plant Examination for Severe Accident Vulnerabilities," for fire events. This was an overview review that evaluated the licensees IPEEE methodology and approach to determine if the analysis yielded reasonable results. This review did not evaluate the accuracy of the data or the completeness of the assumptions and engineering judgements used to support the results of this analysis.
The following areas were reviewed: methodology selection; fire hazard analysis; review of plant information end walkdown; fire growth and propagation; evaluation of component
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fragilities and failure modes; fire detection and suppression; and anclysis of plant systems, sequences and plant response.
The licensee used a PRA approach to assess external fire events. The
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methodology used, was detailed in Sections 1.2.5 and 2.5.2 of the i
Millstone Unit 3 PSS. This methodology and approach was found to be 9906020191 990 6
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Millstone, Unit 3 consistent with the methods outlined in Electric Power Research Institute (EPRI) Final Report, TR-100370, Fire-Induced Vulnerability l
Evaluation (FIVE). The licensee analysis takes into consideration the j
results of their post-fire safe shutdown analysis and that the required fire protection features and fire barrier separation between redundant safe shutdown trains is fully implemented. At Millstone Unit 3, the separation between redundant shutdown trains generally follows the guidance of NUREG-0800, Standard Review: Plan (SRP) Section 9.5.1, " Fire Protection," except in the area of the charging and component cooling 1
pumps.
The licensee used the American Nuclear Insurers (ANI) data base on safety losses for determining the frequency of fire per reactor year for the following plant areas: diesel generator area; containment; cable spreading room; control room; turbine building; and auxiliary building.
This is a 1978 data base, which contains 37 fire events over a period of approximately 300 reactor years of operation. The licensee performed a
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comparison of the Electric Power hsearch Institute (EPRI) fire database and the ANI database, for example the ANI database indicates th t the theEPRIdatabaseindicatesthatthisfrequencyis2.84x10',4x fire initiation frequency for a diesel generator room is 3.
while
'In reviewing the licensees comparison of these databases, we did not find a significant difference between these fire frequencies and therefore, the frequencies as used by the licensee should not affect the results of the analysis.
The licensee performed an analysis which evaluated the mitigating actions and loss of safety functions for each fire area. This analysis used two quantification models that credit the detection and suppression capabilities in the plant. The first model uses an event tree with stages of ignition, detection, suppression and propagation. The second model is a critical path analysis which models stages in fire discovery, incipient stage, meaningful fire and the full room involvement. We' found this methodology for performing this portion of the analysis to be consistent with past PRA approaches.
The key assumptions used by the licensee to support the fire analysis for detection and suppression where (1) Unless an area, such as the control room, is occupied at all times no detection is initially credited to humans; (2) Smoke detectors are assumed to detect a fire before heat detectors and are credited with detecting a fire in the earliest discovery stage; (3) Human detection could occur because of a fire induced spurious signal if the smoke detectors have not responded; (4) If the smoke detector detects the fire and the fire is suppressed with a portable fire extinguisher, it is assumed that this is an early incipient stage fire and no safety loss has occurred; (5) If the portable extinguisher does not suppress the fire and the smoke detectors have not detected the fire, but either the heat detectors or humans have responded, then suppression may depend on either the automatic system or a hose station.
If the fire is. suppressed at this stage, it is assumed that a partial loss of safety function has occurred, based on the proportion of the safe shutdown components within the fire area; and (6)
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Millstone, Unit 3 t If the fire is detected but not suppressed, a total loss of all safety functions is postulated.
In addition, the key assumptions used by the licensee concerning the frequency of fire were: (1) all fires are assumed to increase in size, none are assumed to self extinguish; and (2) specific fire scenarios were not modeled, instead, it is assumed that any fire in an area has the potential of causing a loss of eith~er electrical equipment or cables, regardless of the source of ignition.
An oil fire is assumed to cause loss of a pump, and if it propagates, it is assumed to propagate to cables or electrical components.
1 We found the licensees assumptions associated with detection and suppression to be reasonable and the assumptions made concerning the frequency of fire and its growth within a fire area conservative.
In addition, the licensee's fire analysis made certain assumptions on the loss of safe shutdowa functions or the initiation of a plant transient based on partial or total fire loss of the zone. These assumptions are summarized above.
SPLB found these assumptions to be reasonable.
i The licensee assessment of fire suppression induced damage (Generic Issue 57 the licensing rev)ew for this facility.to safe shutdown equipment was addressedi i
The licensee addressed these potential system interactions in a submittal entitled " Millstone Nuclear power Station, Unit No. 3, Responses to SER Open Items," dated April 2, 1985.
In this submittal the licensee indicated that the carbon dioxide j
suppression systems used at Millstone are actuated by a cross zone l
detection system and that the actuation logic reduces the possibility of inadvertent actuation of these type of systems. The use of fixed water fire suppression systems has been generally limited to non-safety related areas (e.g., transformers, lube oil hazards in the turbine building).
Fixed fire water suppression systems have been used in a limited application to protect the charging pump cubicles and inside the containment in the area of the electrical penetration. The fira suppressions systems are fusible closed head type systems. The licensees inadvertent actuation assessment of these systems coiicluded that if a fusible sprinkler head were to actuate the water spray and potential damage would be limited to the area covered by the individual sprinkler head.
In addition, the licensees assessment concludes that even if all sprinkler heads were to actuate in these areas the discharge of water is not expected to have an impact on the plants ability to safely shutdown since these systems are used primarily to protect cables and cable trays.
The cables and cable trays at Millstone have been designed to be wetted down by water fire suppression systems. In addition, in the area of the charging pumps the plant layout (e.g.,
walls, distance between equipment) lends itself to ensuring that redundant components are not affected by inadvertent operation of the l
firesuppression4 system.
The licensee assessed the effects of manual fire fighting and water affecting the safe operation of the plant. At Millstone, there are o;;1y l
two areas where redundant trains of cable or equipment are located in
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the same plant area. These areas are the charging pump cubicles and the electrical penetrations inside the containment. The licensees Mi j
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Millstone, Unit 3 '(
assessment concluded that adequate separation is provided between redundant trains by using fire rated fire barriers which assures that one method of achieving and maintaining safe shutdown would be free from effects of fire or fire suppression activities.
In addition, the licensee assures that their fire brigade training program provides instruction on proper fire fighting techniques and the effective use of water. The licensee claims that this provides an additional level of assurance that redundant components will not be affected by manual water fire fighting operations.
The licensee modeled the failure of their fire suppression and detection capabilities.
In these models the licensee used fire suppression and detection failure data from estimates developed by Sandia National Lab, American Nuclear Insurers, and the National Fire Protection Association.
The failure probabilities derived from this data for fire suppression systems and the success values for the detection capability are considered reasonable predictions.
The licensee established a screening criteria for the various plant Four general classification of fire areas by type of fire areas.
consequence were identified.
These classifications are: (1) Fire areas where no initiating event would occur and no safe shutdown equipment would be lost as a result of the fire; (2) Fire areas where a transient event would occur as the result of the fire, but no safe shutdown components are affected; (3) Fire areas where no transient event occurs, but safe shutdown components would be lost as the result of the fire; and (4) Fire areas where an transient event occurs and safe shutdown components would be lost as the result of a fire.
From the licensee's analysis,13 fire zones had to be further quantified to determine their fire related core-melt frequency.
These fire zones control room (CB-9); Instrument rack room (CB-llA and CB-IIB);
are:
cable spreading room (CB-8); switchgear rooms (CB-1 and CB-2); motor control and rod control areas (SB-1 and SB-2); charging pumps and reactor plant component cooling pumps (AB-1); circulating and service water building (CSW-3 and CSW-4); and the diesel generator enclosures (EG-3 and EG-4).
Each of these critical fire zones were analyzed for plant damage states that could result after a safety loss due to fire.
This analysis considered the loss of safety function due to fire, the unavailability of redundant safety functions, and potential plant transient events.
Fromthisanalysis,thelicenseehaspredictedthecoredamagefrequenpy 2.44x10'7 these areas. The CDF for the control room (CB-9) is 7.28x (CDF) fo for.the cable spreading room (CB-8); 8.03x10'4 and CB-IIB); 9.8 for the Instrument rack room (C8-11 for the switchgear rooms (CB-1 and CB-2); 8.42x10.a for the motor control and rod control areas 4
(SB-1 and SB-2); 1.07x10 forthechargjngpumpsandreactorplant component cooling pumps (AB-1); 4.27x10 for the circ service water building (CSW-3 and CSW-4); and 1.45x10'ylating and for the diesel F
generator enclosures (EG-3 and EG-4).
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Millstone, Unit'1 SPLB found the licensee's analysis for these areas to be a reasonable approach for assessing the fire related CDF.
4.0 CONCLUSION
The licensee used the results of the Millstone Unit No. 3 Probabilistic Safety Study (PSS) as the basis for the IPEEE review. The PSS is based i
on a Level III Probabilistic Risk Assessment (PRA). The major IPEEE J
fire analysis steps performed were: (1) selection of critical fire 1
areas; (2) estimation of fire frequency in critical areas; and (3) quantification of fire related core-melt frequencies. The Plant Systems Branch performed an initial review of the Millstone Unit 3 IPEEE for fire events. This was an overview type review that evaluated the licensees IPE methodology and approach to determine if the analysis yielded reasonable results. This review did not evaluate the accuracy of the data or the completeness of the assumptions and engineering judgements used to support the results of this analysis.
The following
. areas were reviewed as part of this evaluation included: methodology selection; fire hazard analysis; review of plant information and walkdown; fire growth and propagation; evaluation of component fragilities and failure modes; fire detection and suppression; and analysis of plant systems, sequences and plant response.
Based on the current level of fire protection provided at Millstone Unit 3 and the results of this IPEEE review, we conclude that the licensee has taken a reasonable approach toward assessing their plant for fire related vulnerabilities and has adequately addressed Generic Letter 88-20, Supplement 4.
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Millstone Nuclear Power Station Unit 3 cc:
Lillian M. Cuoco, Esquire Joseph R. Egan, Esquire Senior Nuclear Counsel Egan & Associates, P.C.
Northeast Utilities Service Company 2300 N Street, NW P. O. Box 270 Washington, DC 20037 Hartford, CT 06141-0270 Mr. F. C. Rothen Mr. Kevin T. A. McCarthy, Director Vice President - Work Services Monitoring and Radiation Division Northeast Utilities Service Company l
Department of Environmental Protection P. O. Box 128 i
79 Elm Street Waterford, CT 06385 Hartford, CT 06106-5127 Emest C. Hadley, Esquire Regional Administrator, Region I 1040 B Main Street l
U.S. Nuclear Regulatory Commission P.O. Box 549 475 Allendale Road West Wareham, MA 02576 King of Prussia, PA 19406 Mr. John Buckingham First Selectmen Department of Public Utility Control Town of Waterford Electric Unit l
- Hall of Records 10 Liberty Square 200 Boston Post Road New Britain. CT 06051 Waterford, CT 06385 Mr. James S. Robinson, Manager Mr. Wayne D. Lanning NuclearInvestments and Administration Deputy Director of Inspections New England Power Company Special Projects Office 25 Research Drive 475 Allendale Road Westborough, MA 01582 King of Prussia, PA 19406-1415 Mr. John Streeter Mr. M. H. Brothers Recovery Officer - Nuclear Oversight Vice President - Millstone Unit 3 Northeast Utilities Service Company Northeast Nuclear Energy Company P. O. Box 128 P.O. Box 128 Waterford, CT 06385 Waterford, CT 06385 Deborah Katz, President Mr. M. R. Scully, Executive Director Citizens Awareness Network Connecticut Municipal Electric P.O. Box 83 Energy Cooperative Shelbume Falls, MA 03170 30 Stott Avenue Norwich, CT 06360 Mr. Allan Johanson, Assistant Director Office of Policy and Management Mr. David Amerine Policy Development and Planning Vice President-Human Services Division Northeast Utilities Service Company 450 Capitol Avenue - MS# 52ERN P. O. Box 128 P. O. Box 341441 Waterford, CT 06385 Hartford, CT 06134-1441 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road Waterford, CT 06385 l
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Millstone Nucle:r Power St: tion Unit 3 cc:
The Honorable Terry Concannon Mr. William D. Meinert Nuclear Energy Advisory Council Nuclear Engineer l
Room 4035 Massachusetts Municipal Wholesale l
Legislative Office Building Electric Company Capitol Avenue P.O. Box 426 Hartford, CT 06106 Ludlow, MA 01056 Mr. Evan W. Woollacott Attomey Nicholas J. Scobbo, Jr.
Co-Chair Ferriter, Scobbo, Caruso, Rodophele, PC Nuclear Energy Advisory Council 1 Beacon Street,11th Floor 1
128 Terry's Plain Road Boston, MA 02108 i
Simsbury, CT 06070 i
I Mr. John W. Beck, President
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Little Harbor Consultants, Inc.
Millstone -ITPOP Project Office P.O. Box 0630
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Niantic, CT 06357-0630 Mr. B. D. Kenyon (Acting)
Chief Nuclear Officer-Millstone Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Mr. Daniel L. Curry Project Director Parsons Power Group Inc.
2675 Morgantown Road Reading, PA 19607 Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street Chicago,IL 60603 Mr. G. D. Hicks Director-Unit 3 Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Senior Resident inspector Millstone Nuclear Power Station clo U.S. Nuclear Regulatory Commission l
P. O. Box 513 l
Niantic, CT 06357