ML17180B308
| ML17180B308 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/20/1995 |
| From: | Stang J NRC (Affiliation Not Assigned) |
| To: | Farrar D COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 9507280043 | |
| Download: ML17180B308 (13) | |
Text
....
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 July 20. 1995 Mr. D. L. Farrar Manager, Nuclear Regulatory Services Commonwealth Edison Company Executive Towers West III 1400 Opus Place, Suite 500 Downers Grove, IL 60515
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF PRELIMINARY ASP ANALYSIS OF OPERATIONAL CONDITION AT DRESDEN, UNIT 2
Dear Mr. Farrar:
Enclosed for your review and comment is a copy of the preliminary Accident Sequence Precursor (ASP) analysis of an operational condition which occurred at Dresden, Unit 2, on June 8, 1994 (Enclosure 1), and was reported in Licensee Event Report (LER) No. 237/94-018, Revision 1.
This analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL).
The results of this preliminary analysis indicate that this condition may be a precursor in the 1994 Annual Precursor Report.
In assessing operational events, an effort was made to make the ASP models as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators.
We realize that licensees may have additional systems and emergency procedures, or other features at their plants that might affect the analysis.
Therefore, we are providing you an opportunity to review and comment on the technical adequacy of the preliminary ASP analysis, including.the-depiction.of plant equipment and equipment capabilities.
Upon
- receipt and evaluation of your comments, we will revise the conditional core damage probability calculations where necessary to consider the specific information you have provided.
The object of the review process is to provide as realistic an analysis of the significance of the event as possible.
In order to incorporate your comments and meet our schedule for issuance of the 1994 Precursor Report, you are requested to complete your review and to provide any comments within 30 days of receipt of this letter.
We have also enclosed several items to facilitate your review.
contains specific guidance for performing the requested review, identifies the criteria which we will apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or specific actions in recovering from the event, and describes the specific information that you should provide to support such a claim. is a copy of LER No. 237/94-018, Revision I, which documented the event.
~
9507280043 950720.. :
- PDR ADOCK 05000237 S
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\\-, ______ --- -. ------------~-
OFFICE NAME DATE D. The final resolution of each licensee's comment-on the preliminary ASP analyses will be documented in a separate appendix of the 1994 Precursor Report, NUREG/CR-4674.. Dresden, Unit 2, is on ~~e distribution list for NUREG/CR-4674.
This request is covered by the existing OMB clearance number (3150-0104) for NRC staff followup review of eveots documented in LERs.
Your response to this request is voluntary and does not constitute a licensing requirement.
If you have any questions regarding thi$.reque~t, please contact me at (301) 415-1345.
Docket No. 50-237 Sincerely, Original signed by:
John F. Stang, Senior Project Manager Project Directorate 111-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation
Enclosures:
- 1.
ASP Analysis
- 2.
Guidance
- 3.
LER No. 237/94-018, Rev. 1 cc w/encls:
see next page DISTRIBUTION:
Docket File PUBLIC POI 11-2 R/F J. Roe R. Capra C. Moore J. *Stang OGC ACRS (4)
P. HilanG
. "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy
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07 /20 /95 OFFICIAL RECORD COPY
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D. L. Farrar Commonwealth Edison Company cc:
Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60603 Mr. Thomas P. Joyce Site Vice President Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450-9765 Mr. J. Heffley Station Manager Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450-9765 U.S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station*
6500 North Dresden Road Morr~s, Illinois 60450-9766 Regional Administrator U.S. NRC, Region III 801 Warrenville Road Lisle, Illinois 60532-4351 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Chairman Grundy County Board Administration Building 1320 Union Street Morris, Illinois 60450 Dresden Nuclear Power Station Unit Nos. 2 and 3
A.1-1 A.1 LER No. 237/94-018 Rev. 1 0
Event
Description:
Motor Control Center Trips Due to Improper Breaker Settings Date of Event:
June 8, 1994 Plant:
Dresden 2 A.1.1 Summary Following an unexpected trip of a motor control center (MCC) at Dresden 3 during surveillance testing, three MeCs were identified at Dresden 2 and Dresden 3 with improperly set feeder breakers.
A review of MCC loading indicated that load additions since the original settings were determined had created an overload situation. For two of the MCCs, the overload condition would only have existed if an emergency diesel generator (EDG) was running following a reactor trip with offsite power available. Load shedding following a loss of offsite power (LOOP) would have precluded an overload condition for this initiating event. For on-:._,f the MCCs, the overload condition would also
- have existed following a LOOP.
The conditional core damage probability estimated for the event is 6.1 x 10~.
The relative significance of this event compared to other postulated events at Dresden is shown below in Figure A.1.1 (to be provided in the final report).
. A.1.2 Event Description On June 8, 1994, Dresden Unit 2 was operating at 99% power, and Unit 3 was in refueling. The
...Unit 2/3 standby_gas treatme.nt (SBOT) system was in operation, and a 24 h endurance run for the EDG 3 was in progress, as was a Unit 2 high-pressure coolant injection (HPCI) surveillance.
Shortly after starting the Unit 2 HPCI auxiliary oil pump, MCC 39-2 tripped. As a result of the loss of power at MCC 39-2 (1) EDG 3 tripped on high temperature following loss of power to its cooling
. water pump and ventilation fan, (2) the 125-V de and 250-V de battery systems had to be realigned to alternate chargers, (3) a half-scram was generated as a result of loss of power to the RPS motor-generator, and (4) SBGT train A auto-started following loss of power to train B components.
MeC 39-2 loads were.stripped, and the MCC feeder breaker was reclosed. MCC 39-2 loads were reenergized within 30 min of the breaker trip.
The trip of MCC 39-2 was caused by an incorrectly set feeder breaker. The feeder breaker for the Mee had a General Electric dashpot type EC-2A overcurrent trip device which was original equipmenL The setting for this breaker was 400 A. A review of the original loading on the MCC indicated that the 400 A setting was adequate, but load additions made to the MCe over time had increased the available running load current above the 400 A setting.
Two other breakers were subsequently identified with similar problems-Mee 28-3 and 38-3. The EC2A trip devices for both of these MCCs had been replaced with newer General Electric solid state type RMS-9 trip devices. Both of these MCCs were also set to trip at 400 A The licensee noted in the LER that the.setting for MCC 38-3 was chosen to be identical with the original breaker ENCLOSURE 1
. -~
A.1-2 setting b~
on the assumption that MCC loading had not changed over time. However, since the loading had changed, the total connected load was greater than the protective device setting. At the time of the MCC 28-3 trip device replacement, it was recognized that the overcurrent setting was lower than the total connected load. However, it was assumed that the running load during accident conditions would be within the setting of the protective device.
Based on the loads associated with each MCC, the licensee concluded that MCCs 38-3 and 39-2 could be overloaded and trip during a safety actuation in which the associated EOG was running (e.g., for testing or following a spurious start) while offsite power was still available. For these MCCs, loads shed following a LOOP would preclude an overload condition. For MCC 28-3, however, the overload condition could exist for both LOOPs and other events in which the associated EOG was running.
A.1.3 Additional Event-Related Information Three EOGs provide emergency power to the two Dresden units: EOG 2 provides power to Unit 2 bus 24-1, EOG 3 provides power to Unit 3 bus 34-1, and swing EOG 2/3 provides power to either Unit 2 ":>us 23-1 or Unit 3 bus 33-1 in the ev~n* of a LOOP on Unit 2 or Unit 3, respectively. In the event of a dual-unit LOOP with a loss of coolant accident (LOCA) on one unit, EOG 2/3 provides power to the unit with the LOCA. In the event of a dual-unit LOOP without a LOCA, EOG 2/3 powers the unit that suffers the LOOP first. Unit 2 bus 24-1 and Unit 3 bus 34-1 can be cross-tied by closing two normally open breakers.
Two 250 V-dc and two 125 V-dc batteries are shared between both units. The 250 V-dc batteries primarily power large loads, such as de-powered pumps and valves, while the 125 V-dc batteries provide control power to components such as circuit breakers. Battery chargers that normally supply de power and provide battery charging can be powered from buses associated with EOG 2 (Unit 2) or EOG 3 (Unit 3) or the swing EOG. Each battery is sized to power its respective loads for 4h.
A.1.4 Modeling Assumptions*
Four possible situations were addressed in the analysis of this event. All three MCCs could have tripped following an initiating event in which emergency core cooling system (ECCS) actuation was
- required, offsite power was available, and the EOG associated with the MCC was running (e.g., for testing or following a spurious start). Analysis Case la addresses the situation in which one EOG was running.
Analysis Case lb addresses the situation in which two EOGs were running.
In addition, MCC 28-3 could have tripped following a LOOP. Analysis Cases 2a and 2b consider a plant-centered LOOP at Unit 2 and dual-unit LOOPs at Units 2 and 3. In all cases, the MCCs were assumed to trip if they could have tripped. This assumption may be conservative.
Case la. Postulated initiating event with offsite power available and one EOG running. This situation could exist if a transient or small-break LOCA occurred and one of the two EOGs associated with a unit was undergoing monthly surveillance testing. The greatest potential impact is associated with MCQ 39-2 and 38-3 at Unit 3. These MCCs, in addition to supplying power to EOG components (and turning gear components for MCC 38-3), also supply power to containment cooling service water (CCSW) cubide fans. CCSW provides decay heat removal for the containment cooling mode of low-pressure coolant injection.
The analysis assumed the two CCSW trains associated with the running EOG would be unavailable after the MCC tripped. The probability of a running EDG was estimated to be 2.8 x 10*3, based on an assumed 1-h surveillance run-time for
-~
A.1-3
. each EDG per month.
The significance for this case was estimated by setting basic events associated with the two impacted CCSW trains to true (failed) and calculating the increase in core damage probability for non-LOOP (transient and small-break LOCA) initiating events over a 1-year period using the IRRAS-based ASP
.model for Dresden. Long-term unavailabilities such as this event have typically been modeled in the ASP program for a 1-year period. assuming the plant was at power 70% of the time; this is equal to 6132 h (365 d x 24 hid x 0.7). The increase in core damage probability was multiplied by the probability that an EDG would be running to estimate the conditional probability for Case la. This conditional probability is less than 1.0 x 10..a.
Since this is substantially below the 1.0 x 10-6 documentation limit used in the ASP program, the calculational results are not included herein.
Case 1 b. Postulated initiating event with offsite power available and two EDGs running. This situation could exist if a transient or a small-break LOCA occurred and both EDGs associated with a unit were spuriously started. The analysis for this case is similar to Case la, except all trains of CCSW were assumed to be unavailable. The probability of spurious EDG start was estimated using a Sequence Coding and Search System search of BWR automatic or manual reactor trips with spurious EDG starts. Three such events were identified in 573 trips from power, resulting in an estimated probability of spurious EDG actuation of 5.2 x 10*3*
The resulting conditional core damage probability is estimated to be 4.3 x 10-s, also well below 1.0 x 10-6. As for ca.Se la, the calculational results are not included herein.
Case 2a. Postulated plant-centered LOOP at Unit 2. For a postulated plant-centered LOOP at Unit 2 only, offsite power remains available at Unit 3. Trip of MCC 28-3 will result in inoperability of swing EDG 2/3 and unavailability of power to 4-kV bus 23-1. Power can be recovered to bus 24-1 if EDG 2 fails by recovering offsite power or by closing the cross-tie from Unit 3 bus 34-1. Because of the shared de system at Dresden, de power will remain available for instrumentation even if Unit 2 batteries are depleted. Therefore, a sequence involving SRV reseat and isolation condenser or HPCI success following a postulated station blackout will not proceed to core damage (essentially all of sequence 44).
The probability of failing to recover power to bus 24-1 through closure of the cross-tie breakers from Unit 3 was assumed to be 0.12 (ASP nonrecovery class R3, see Appendix A, Sect. A.1 to the 1992 precursor report, NUREG/CR-4674, Vol. 17). This value was chosen because recovery appeared possible in the required time from the control room, but was not considered routine (the value*
chosen for this failure probability for this case is considered a bounding probability and does not substantially impact the overall analysis results). This value is used in lieu of the failure probability for EDG 3 in the IRRAS-based ASP models to reflect the failure to provide power from bus 34-1.
The probability of EDG common-cause failure was set to false to reflect the unavailability of EDG 2/3 and the availability of power on bus 34-1.
After elimination. of sequence 44 (since it does not proceed to core damage for a single-unit plant-centered LOOP) a conditional core damage probability of 1.6 x 10..a is estimated. As for Cases la and 1 b, the calculational results are not included herein.
Case 2b. Dual-unit LOOP at Units 2 and 3. For a postulated dual-unit LOOP (primarily grid-and weather-related LOOPs), offsite power is unavailable to both units. If the LOOP occurs at Unit 2 first, trip of MCC 28-3 will result in unavailability of swing EDG 2/3. EDG 3 will be required to power Unit 3 loads, leaving only EDG 2 to supply power to Unit 2 loads.
- *::..O*
A.1-4 The frequency of a dual-unit LOOP and the probability of failing to recover offsite power in the short-term and before battery depletion were estimated to be 1.7 x 10*2,tyear, 0.66, and 0.21, respectively, based on models described in Revised LOOP Recovery and PU'R Seal LOCA Models, ORNUNRC/LTR-89/11, August 1989. These models are based on the results of data distributions contained in Evaluation of Station Blackout at Nuclear Power Plants, NUREG-1032. The probability of the dual-unit LOOP occurring first at Unit 2 was assumed to be 0.5. The failure probability for EOG '213 was set to true to reflect its unavailability following trip of MCC 28-3. The common-cause failure probability for the EDGs was revised to 4.4 x 10*3 to reflect the unavailability of EOG 2/3.
A.1.5 Analysis Results The conditional core damage probability estimated for this event is 6.1 x 10~. The dominant core damage sequence, highlighted on the event tree in Figure A.1.3 involves a postulated dual-unit LOOP (primarily grid-or weather-related) with subsequent failure of all three Dresden EDGs and failure to recover offsite power prior to battery depletion. In the dominant sequence, EOG 2/3 fails due to MCC 28-3 trip following its alignment to Unit 2 (the postulated dual-unit LOOP affects Unit 2 first), and EOG 2 and 3 fail for unspecified reasons (random or common-cause failures).
The ca1culational results for Cases la, lb, and 2a were not included since they do not provide a significant contribution to the conditional core damage probability for the event. The calculational results for Case 2b are shown in Tables A.1.1 through A.1.5. Definitions and probabilities for basic events are shown in Table A.1.1. The conditional probabilities associated with the highest probability sequences are shown in Table A.1.2.
Table A.1.3 lists the sequence logic associated with the sequences listed in Table A.1.2.
Table A.1.4 describes the system names associated with the dominant sequences. Cutsets associated with each sequence are shown in Table A.1.5.
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I Dominant core damage sequence for LER 237/94-018.
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Al-6 Table Al.I. Dcfillitiam md probebili1ioa for IOloctod basic: eWllltl for LER 237194...()18 E'Wlll name EPS-OON~F-OONS EPS-OON-FC-002 EPS-OON-FC-003 EPS-OON-FC-0023 EPS-XHE-XE-NOREC IE-LOOP IE-SLOCA IE-TRAN OEP-XHE-XE-NOREC E VCII1 tree name LOOP Bue Cumm Delcriptioa probability probllbility COMMON CAUSE FAD..URE OF l.2E-003 4.4E-003 DIESEL GENERA TORS UNIT 2 GENERATOR FAll.S 4.4E-002 4.4E-002 UNIT 3 DIESEL GENERA TOR 4.4E-002 4.4E-002 FAULRE SWING DIESEL GENERATOR 4.4E-002 l.OE-+-000 FAll.S OPERA TOR F All.S TO RECOVER 8.0E-001 8.0E-001 EMERGENCY POWER LOSS OF OFFSITE POWER 9.IE-007 S.6E-003 INITIATOR SMALL LOCA INITIATOR l.7E-006 O.OE+-000 TRANSIENT INITIATOR 3.4E-004 O.OE-+-000 OPERA TOR F AD..S TO RECOVER 2.IE-001 2.IE-001 OFFSITE POWER Table A 1.2. Sequeuco cooditiooal probabilitiea for LER 237194-018 Sequeuco name 44 Conditional core damage probability
. (CCDP)
S.9E-006 Core damage probability (CDP) 3.SE-006 Importance (CCDP~DP) 2.JE-006 Total (All Scqueocea) 6.lE-006 Table Al.3. Scqucocc logic for dominant sequcocea for LER 237194...()18 EYCat tree name SoqilCllCO name Logic LOOP 44
/RPI, EPS, OEP Table Al.4. System namea for LER 237194...()18 System name Description EPS EMERGENCY POWER SYSTEM FAll.S OEP OFFSITE POWER RECOVERY RPI REACTOR SHUTDOWN FAll.S Modified Type for thiJ event?
y N
N TRUE y
N y
IGNORE y
IGNORE y
N
'le Caatributioa 96.7
.~
Al-7 Table Al.S. Cooditiooal cut aeta for higher proba:>ility sequeocea for LER 237~18 Cutsct No.
Ccmtn"butioo LOOP Sequence: 44 69.S 2
30.6 Frequency Cut llCtl 4.lE-006 EPS-DGN~F-OONS, EPS-XHE-XE-NOREC, OEP-XHE-XE-NOREC l.SE-006 EPS-XHE-XE-NOREC, OEP-XHE-XE-NOREC, EPS-OON-FC-002, EPS-DGN-FC-003
Background
GUIDANCE FOR LICENSEE REVIEW OF PRELIMINARY ASP ANALYSIS The preliminary precursor analysis of an operational eYent that occurred at your plant has been proYided for your review. This analysis was perfom1ed as a part of the NRC s Accident Sequence Precursor (ASP)
Program. The ASP Program uses probabilistic risk assessment teclmiques to proYide estimates of operating e\\*ent significance in terms of the potential for core damage. The types of events evaluated include actual initiating events such as a loss of off-site power (LOOP) or Loss-of-Coolant Accident (LOCA), degradation of plJnt conditions, and safety equipment failures or unanilabilities that could increase the probability of core damage from postulated accident sequences. This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), indi\\'idual plant examination (IPE), and the licensee e\\*ent report (LER) for this e\\'ent.
Modeling Techniques The models used for the analysis of 1994 eYents were dc\\'eloped by the Idaho National Engineering Laboratory (fNEL).
The models \\Yere developed using the Systems Analysis Programs for Hands-on Integrated Reliability haluations (SAPHIRE) software. The models are based on linked fault trees. Four initiating eYents are considered: ( 1) transients, (2) loss-of-coolant accidents (LOCAs), (3) loss of offsite power (LOOPs), and (4) Steam Generator Tube Ruptures (PWR only). Fault trees were de\\'eloped for each top eYent on the e,*ent trees to a supcrcomponent lc,*cl of detail. The only support system currently modeled is the electric power system The models may be modified to include additional detail for the systems/components of interest for a particular e,*ent. This ma~* include additional equipment or mitigation strategies as outlined in the FSAR or IPE. Probabilities are modified to reflect the particular circumstances of the e\\'ent being analyzed.
- Guidance.for Peer Review Comments regarding the analysis should address:
Docs the "E,*ent Description*' section accurately describe the event as it occurred')
Does the "*Additional EYent-Related lnfom1ation" section pro,*ide accurate additional information
>e.oncerning the configuration of the plant and the operation of and procedures associated with relennt systems')
Docs the ""'fv1odeling Assumptions" section accurately describe the modeling done for the event? Is the
- modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions.? This also includes assumptions regarding the likelihood of equipment reco,*ery.
Appendix E of Reference I provides examples of comments and responses for previous ASP analyses.
ENCLOSURE 2
Criteria for Evaluating Comments Modifications to the ~*ent analysis may be made based on the comments that you pro\\*ide.
Specific documentation will be required to consider modifications to the e\\*ent analysis. References should be made to portions of the LE~ AIT, or other event documentation concerning the sequence of events. System and component capabilities should be supported by references to the FSAR, IPE, plant procedures, or analyses.
Comments related to operator response times and capabilities should reference plant procedures, the FSAR.,
the IPE, or applicable t0perator response models. Assumptions used in determining failure probabilities should be clearly stated.
Criteria for Evaluating Additional Recovery Measures Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis. HoweYer, to assess the viability and effecti,*eness of the components and methods, the appropriate documentation must be included in your response. This includes:
non11al or emergency operating procedures,*
piping and instrumentation diagrams (P&!Ds),"
electrical one-line diagrams,"
results of therm::nl-hydraulic analyses, and operator training (both procedures and simulator)," etc.
Systems, equipment or 'Specific recO\\ery actions that \\Yere not in place at the time of the eYent will not be considered. Also, the documentation *should address the impact (both positive and negati\\*e) of the use of the specific recO\\*ery measure.. on:
the sequence of eYents, the timing of en~'llll*s, the probability of operator error in using the system or equipment, and other systems/processes already n1od.eled in the analysis (including operator actions).
For e:--;ample, Plant.A (a P\\VR) experiences a reactor trip, and, during the subsequent recoYery, it is discoYered that one train of the auxiliary feed\\\\*ater (AFW) system is una,*ailable. Absent any further
. infom1ation regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW unaYailable. The AF\\V modeling would be patterned after infom1ation gathered either from the plant FSAR or the IPE. Howe,*er, if infom1ation is received about the use of an additional system (such as a standby steam 1generator feedwater system) in reco,*ering from this eyent, the transient would be modeled as a reactor itrip with one train of AFW unavailable, but this unavailability would be mitigated by the use of the standby feedwater system. The mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:
standby feedwater system characteristics are documented in the FSAR or accounted for in the IPE, procedures for l.ISing the system during reco,*ery existed at the time of the event, the plant operat01rs had been trained in the use of the system prior to the event, a clear diagram offcthe system is available (either in the FSAR, IPE, or supplied by the licensee),
previous analyses have indicated that there would be sufficient time available to implement the procedure success[ ully under the circumstances of the event under analysis,
- Revision or practices at th.: time the event occuned.
the effects of using the standby feed,,*atcr syste111 haYe on the operation and recoYery of systems or procedures that are already included in the e,*ent 111odeling. In this case, use of the standby feedwater system may reduce the likelihood of recO\\*ering foiled AF\\\\' equip111ent or initiating feed-and-bleed due to time and personnel constraints.
l\\laterials Provided for Review The fol lowing materials haYe been proYided 111 the pL!ckage to facilitLJte your re\\*iew of the preliminLJry anLJlysis of the operational e,*cnt The specific LER, augmented inspection team (AIT) report, or other pertinent reports.
A summary of the calculationLJI results. An eYent tree with the dominant sequence(s) highlighted. Four tables in the LJnLJlysis indicLJte (I) LJ summLJry of the releYLJnt basic events including modifications to the probabilities reflect the circumstLJnces of the e\\*enL (2) the dominant core damage sequences, (3) the s~*stem nLJmes for the systems cited in the dominLJnt core dLJmLJge sequences, LJnd ( 4) cut sets for the dominLJnt core dLJmage sequences.
Schedule PlcLJse refer to the trLJnsmittLJl lettcr for schedules LJnd procedures for sub111itting your comments.
References I.
L. N. VLJndcn Hcu,*cl ct LJI., Prec11rsors ro Por11nrial Severe Core Damage Accidenrs: 1993. A Slarus Repon USNRC Report NUREG/CR-4674 (OR.J"IL/NOAC-232, Volumes 19 and 20), Martin i\\fariettLJ Energy Systems, Inc, OLJk Ridge NLJtionLJI LLJborLJtory and Science ApplicLJtions IntemLJtional Corp.,
September 1994.
Guidance revised March 9, 1995