ML16162A292
| ML16162A292 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 12/19/1979 |
| From: | Thatcher D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8001040344 | |
| Download: ML16162A292 (16) | |
Text
MEETING
SUMMARY
DISTRIBUTION SENT TO FOLLOWING LICENSEES, Duke Power Company (Oconee 1,2and3)
Metropolitan Edison Company (TMI-1)
Florida Power Corporation (Crystal River 3)
Sacramento Municipal Utility District (Rancho Seco)
Arkansas Power & Light Company (ANO-1)
Toledo Edison Company (Davis-Besse 1)
Docket File R. Reid NRC PDR V. Noonan L PDR P. Check ORB#4 Rdg G. Lainas NRR Rdg G. Knighton H. Denton Project Manager E. G. Case OELD OI&E (3)
D. Eisenhut R. Ingram R. Vollmer R. Fraley, ACRS (16)
W. Russell Program Support Branch B. Grimes TERA T. J. Carter J. R. Buchanan A. Schwencer Meeting Summary File D. Ziemann NRC Participants - R. Satterfield T. Ippolito D. Ross D. Thatcher W. Gammill T. Novak L. Shao W. Kane P. Matthews J. Miller R. Capra B. Wilson AEOD S. Israel Z. Rosztoczy G. Mazetis P. -Norian F. Ashe B. Sheron J. Joyce W. Jensen G. Kelly S. Lewis M. Rubin M. Mulkey H. Silver R. Hoefling s q4O
DEC 19 1979 Docket Nos.:
54=2jr, 270, 287 50-289-, 50-302 R ULA1DRY gOcKETREC?
50-312, 50-313 50-346 FACILITIES:
Oconee Nuclear Station, Unit Nos. 1, 2 -3 (Oconee)
Three Mile Island Nuclear Station, Unit No. 1 (TMI-1)
Crystal River Nuclear Generating Station, Unit No. 3 (CR-3)
Rancho Seco Nuclear Generating Station (Rancho Seco)
Arkansas Nuclear One, Unit No. 1 (ANO-1)
Davis-Besse Nuclear Power Station, Unit No. 1 (DB-1)
LICENSEES:
Duke Power Company (DPCo)
Metropolitan Edison Company (Met-Ed)
Florida Power Corporation (FPC)
Sacramento Municipal Utility District (SMUD)
Arkansas Power &*Light Company (AP&L)
Toledo Edison Company (TECo)
SUBJECT:
SUMMARY
OF MEETING HELD ON OCTOBER 23, 1979 WITH REPRESENTATIVES OF THE BABCOCK & WILCOX (B&W) OWNERS'GROUP, B&W AND OAK RIDGE NATIONAL LABORATORY TO DISCUSS THE "INTEGRATED CONTROL SYSTEM RELIABILITY ANALYSIS,," (BAW-1564)
On October 23, 1979, members of the NRC staff (staff) and Oak Ridge National Laboratory (ORNL) and ORNL's consultants from Scientific Applications Incor porated (SAI), met with representatives of the B&W Owners' Group and B&W in Lynchburg, Virginia, to discuss the integrated control system (ICS).
The reference document for the discussions was the B&W report BAW-1564 entitled "Integrated Control System Reliability Analysis," dated August 1979. During the ongoing review of-this document, a set of questions was generated by ORNL/SAI which served as the meeting agenda. A copy of these questi6ns is included as Enclosure 1. A list of attendees is provided as Enclosure 2.
BACKGROUND As one of the long-term items of the Commission Orders of May 1979, the B&W licensees were required to submit a failure mode and effects analysis (FMEA) of the ICS. In response to this requirement, B&W submitted a generic report entitled "Integrated Control System Reliability Analysis," (BAW-1564). This report was endorsed by each of the B&W:licensees as applicable to their facil ities. The staff has obtained assistance in the review of this report from ORNL which,. in turn, is utilizing members of SAI as subcontractors. This meeting was held to discuss the questions (Enclosure 1) prepared by ORNL and SAI based on their preliminary review of the report.
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- 9929391 1EC191 979 DISCUSSION.
After introductory remarks by R. Finnin of B&W, R. Satterfield of NRC, and J. Anderson of ORNL, the meeting centered around a discussion of each of the questions listed in Enclosure 1. Part of the introduction was a description of the approach by ORNL/SAI to the review of the report. They stated the plan being used was to investigate what questions are answered by the report and defi.ne what questions may still.have to be answered.
Specifically, ORNL/SAI's.objectives will be to see if:
- 1. the.concerns expressed by.the NRC staff and ACRS have been'addressed;
- 2. the technical-content of the report is valid;
- 3. the conclusions are valid and useful;
- 4. design changes are required; and,
- 5. additional work should be done.
Question 1. "There may be a significant difference between failure modes or conditions with a FMEA based on functional block diagrams rather than equipment block diagrams. Have the functional failure assumptions been compared with actual equipment failure modes.to assure that they are realistic and meaningful?"
Response
B&W expressed the opinion that the functional block approach was the most practical for the time frame and would indicate problem areas which could be developed on an-equipment block basis, if necessary. They pointed out some examples where there was some follow through to the equipment failures, par ticularly with respect to the operating history section of the report. B&W believes it.has found the "hard spots" in the ICS.
These are presented-in the report's conclusions and recommenda tions. Some discussion evolved concerning other areas which were being worked on by B&W, such as operator guidelines, par ticularly the Abnormal Transient Operating Guidelines (ATOG) program.
Question 2.
"The ICS signal input failure assumptions appear-to be all either "high" or "low" with some attempt to identify the "worst case."
Some of the operable plants under review have the potential for mid-scale failures.
There is reason to believe that some mid-scale failures may be worse than high or low failures, as experienced by the plant selected as typical, Rancho Seco.
Are there'plans for including mid b.CtI U -d la th T I
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Response
B&W emphasized that from a single input failure point of view the "high" or "low" were the worst cases.
The failures referred to in the question were the result of-multiple failures of inputs.
However, as pointed out by ORNL, the multiple failure could be the result of a single power supply failure.
The power supply area was highlighted as a particular concern in the report, however, this.concern was based upon operating experience and not the FMEA section of the report. The FMEA as performed by B&W would not highlight these type failures because of the definition of the ICS.boundary. There are presently no plans to-include "mid-scale" failures.
Question 3.
"Virtually all of the events/failures considered in the analysis appear to be based on "normal" conditions wherein all plant
- equipment is functioning at nominal design points.. Our limited information regarding operating experience suggests that many of the abnormal occurrences were the direct result of some plant equipment not functioning. For example:
three primary pumps instead of four running; one instead of two feedwater pumps running; one or more hand/automatic stations in manual; etc.
Since these seem to be'the more significant initial conditions for unsatisfactory ICS performance, how is their omission just ified? Are any of these "interesting" events analyzed but unreported?"
Response
B&W-is of the opinion that study-ing of off-normal alignments would not lead to anything significant beyond what is-in the report.
B&W did point out that in many of the recent trips.(CR-3 in particular) that the operators could have been more aware of new, post-TMI-2, reactor trip setpoints. It is &W's opinion that these trips are operability problems not safety problems.
ORNL agreed to some extent-however, while a reactor trip may be safe, it does represent a challenge to the control systems.
If a control system cannot do a specific job, then there is a problem. B&W pointed to the experience that showed ICS pre vented more trips than it caused.
Question 4.
"What process was.used to determine the "effect on the NSSS"?
Neither the technique nor the justification is included in the analysis.
What verification techniques were employed for the "effects" analysis?"
Response
Effects were evaluated by knowledgeable people with plant experience.
OFFICE......
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GOVERNMENT PRINTING OFFICE: 1979-289-369
OEC 19 1979
-4 Question 5. "The POWER TRAIN code obviously has limitations to its ability to simulate the NSSS and BOP response. How significant is this limitation on the analysis? -In particular:
(a) Describe the extent to which the simulation was used to predict results.
(b) Describe errors and uncertainties which might have resulted from the limited dynamic range and functional detail of the simulation.
(c)- Describe to what extent the simulation results were verified with plant data.
(d) Describe the extent to which the simulaiton is valid or invalid for each of the indifidual plants and their differ ences, especially feedwater systems.
(e) Does the simulation have capability for dealing with off normal operation such as three primary pumps or partial manual operation?"
Response
B&W pointed out that Power Train was used in about 75% of the cases; however, for post-trip cases, most of the information was developed from an engineering analysis. A general discussion of B&W's simulation followed.
Power Train IV has the following features:
2 Steam Generators modelled in continuous space, discrete time; steam lines; Feed water pumps; Feedwater heaters; Condenser; Pressurizer; Turbine dynamics; Valves.
The primary system includes pump character istics programmed from other codes as a-table and appropriate transport lags (-10 seconds). Pressurizer modelling includes the effects of surge flows, spray flows, internal flows with condensation and flashing, heaters, safety and power operated relief valves.
The ICS model uses a dedicated digital computer (EAI-640) and is a digital model of an analog system utilizing functional blocks.
One feedwater valve model is used to represent all FW valves.
The limiting ranges of PT-IV are reported to be:
Primary Pressure 1500 -
3000 psi Secondary Pressure 500 - 1500 psi Temperature (Pri. & Sec) 400 -
700 0F Feedwater Temperature 350 - 700 0F DATEit
_M38_-____i_0_ I._ _______
.F...c..R 369.......
MOCsFORM. 318.(9476) N$QCM.0244, GOVQCE9NMENT P$RMTING dFVICt:* 1979-289-369
DEC 1 9 1979 The hybrid model uses two EAI-680 analog and one CDC-1700 digital computers.
Due to computer limitations, there is not much detail of the feedwater system. A more complete model (not PT-IV). would include pump drains, flash tank levels and condensate pumps as well as main feed pumps. The condensate pumps have suction pressure trips that sometimes actuate when the interceptor yalves close. This is not modelled. Turbine trip is the transient used to check the code with plant data.
The validity of the comparison is judgemental.
The model is not valid at low powers.
Question 6. "The ability of the ICS'to respond properly to its design basis and other probable conditions is not addressed. That is, design problems associated with normal operation or maneuvering are not included unless a failure is supposed. This may.be outside the scope of the NRC request, but the ICS feedwater systems inter actions evidenced in operating plants indicate this may be of valid concern. Have the design problems and component limita tions associated with expected normal operation-been analyzed and documented?. Are these analyses available?"
Response
B&W explained that.there is little.motivation to spent a signifi cant-amount of time and money on the ICS because from plant avail ability there are other areas which offer more improvement. Their utility customers have no significant unresolved complaints about the ICS. There are internal B&W reports, "quick-look reports,"
on this subject; In addition, there is more forial work being done in the area of looking at customers' reactor trips.
Question 7.
"Is there any connectfon, physical or phenomenological, between RPS sensors and ICS inputs: Which common signals, if any, initiate trip and what is the potential for common signal con ditioning failures -initiating a plant transient through the ICS requiring KPS response derived from that signal."
Resonse:
This discussion was the result of general concern with inter action between protection systems (RPS) and control systems (ICS).
Based on the discussion and knowledge-of B&W design, it was agreed that there is no apparent problem in.this area.
Although the report does not address this concern at all, RPS signals are used by the ICS with suitable buffering.
Adequate redundancy is provided in the RPS to satisfy the requirements of IEEE-279.
M FORM 318 (976) -ACM 0240*
U.S. GOVERNMENT PRINTms OFFFCE:' 1979-289369
~DEC 19 1979 Question 8.
"FMEA categories for "causes", "detection", "propagation potential" would yield helpful information.
Has this type information been generated and is it available?"
Response
Identification of component causes was not considered necessary.
The categories used are based on IEEE Standard 352 The area of detection was not addressed and it was agreed that it may require a further look because of the fact that undetectable failures could exist for long periods of time and then single failures could lead to multiple failures.
Question 9. "The impact of power supply failures appears to be inadequately addressed, especially considering that events of much more significance than those analyzed have occurred at operating plants. How is the.omission of these considerations justified and is more comprehensive power supply failure analysis available?"
Response
Power supply failures were again discussed. According to B&W, further work is this area with the licensees is presently being considered. The FMEA scope did not extend to the power supplies such as those for the input instrumentaiton. Power supplied and theirreliability is a problem for the customer which needs to be resolved on a plant by plant basis.
Question 10. "A significant number of trips appear to have occurred when portions of the system were in manual. What fraction of time is it estimated.that control stations are in manual, and what are the problems associated with this mode of operation of the ICS?"
Response
B&W stated that they could not analyze all.combination of modes in this amount of time. We pointed out that we were asking for information to help determine which modes were more significant.
Manual modes are judged to be used mostly for startup and. testing.
Question II.
"How does historical failure data on ICS 721 and 820 compare with.
predictions based on nominal behavior?
Is there are evidence of accelerated failure?"
Response
B&W pointed out that Oconee and TMI-1 were the plants with the 721 system.
They.discussed the burn-in failure rate and then the relatively constant failure rate subsequent to it.
NRC FORM 318 (9-76) NRCM.9240
- U.S:
GOVERNMENT PRINTING OFFICE: l979-289-369
t)Ec 1 9 1979
-7 Question 12.
"Multiple failures are not treated although it is acknowledged by B&W that many failures are not annunciated and therefore may exist until other failures occur, resulting in effective multiple failures.
It appears that multiple failure situations may have significant probability of occurrence.
How is the omission of multiple failure considerations justified in the analysis?
Might Fault Tree Analysis have been a better technique for addressing the concerns and producing the results requested?"
Response
8&W has identified transients that have occurred, in.the Operating History Section. Therefore, with respect to multiple failures the report has identified critical areas.
Although this is true, an event tree of ICS may highlight other important multiple failures. This type analysis was considered to be too extensive for the time available.
Question 13:
"The analysis does not include information to substantiate the recommendation that improvement is needed-in power supplies, signal selection and signal reliability. Please supply the anal ysis or information which led to this recommendation.
In particular, does B&W have specific recommendations to improve the failure tolerance of the ICS?"
Response
8&4 has noted that a number of events have been caused by power supply failures and other-interface areas. The recommendations are based on these events and are in areas around the ICS boundary.
Question 14.
"Operating experience reports and oral. information not included in the-analysis suggest the ICS and/or the BOP system including the OTSG is sensitive to "tuning" and component problems such as feedwater valve'speed.and leakage.
Describe the extent.to which these problems are significant, how they have led to misoperation and RPS challenges, and how they might be avoided.
Are "tuning" problems inherent to this type of plant or do they represent design deficiencies which-can be corrected?"
Response
B&W explained that it is more than just ICS tuning,-it is more a matter of plant-tuning. If licensee complaints are experienced, more tuning may be a necessity. The area of pressurizer spray was, also talked about as a way to maintain pressurizer level.
OFFICE SU R N A M E DATE O 318
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GOVERNMENT PRINTING OFFICE: 1979-289-369 1
Question 15.
"Many-Licensee Event Reports, as well as this analysis, indicate that the operator is implicated in alarge number.
of occurrences of poor ICS operation. Many of these events also involve slightly off-normal conditions, such as non standard pump and valve.alignment. Do these events repre sent design deficiency, operator training deficiency.or a com bination of-these?, Does B&W have recommendations to correct these deficiencies and on what schedule can they be implemented?
Response
Most problems occur due to maintenance, testing, or equipment problems which require manual conditions. B&W pointed out that this information should not be the sole basis for performing
-operator retraining. We really do not know the success rate since all we see are the unsuccessful ones. B&W is not presently investigating for these.
In addition to the questions discussed above, the following request for specific information was discussed:
- 1. Rancho Seco equipment block diagrams and logic diagrams. The staff should request these directly from SMUD.
- 2. Identification.of differences in ICS design between Rancho Seco ande eah of the other 177FA plants. These design differences should be in the endorsement letters.
- 3. Identification In the FMEA of which "effects" resulted fron POWER TRAIN (PT) simulation and which resulted from engineering judgement. -.Not r'Upsued.
Real eyent date is particularly helpful.
Real event data should come from the,Icensees.
- 5. In Table 4.4 of the report,.what are specific indications to the operat r display failures?
The operator indications were discussed.. B&I provdd d ICS outputs for annunciator and or computer indication.
- 6. What is the ICS design basis?
& quoted the SA9 Section 7.2.3.1.
- 7. PT-11 manual and QA File.
This may be requested if the staff desiret U pursue system simulation.
- 8. Identify the power levels at which trips occurred.
A number of feedwater trips occurred around 40-50% power.
This informat ion should come froM.
the licensees.
- 9. Event information on operator/technician induced trips. Need event re ts id. e ttft1 EISO
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FORM318 (9-76)NCMSGOVER NT PiTNOFIE 1979 289-369
DEC 19 1979 CONCLUSION The discussion of the report questions lead to the general conclusion that although the report appears to be accurate, it does not appear to go far enough., The B&W definition of the ICS-and its boundaries tends-to limit the analysis portion to relatively straight forward conclusions. However, the operating history section does indicate some of the more.significant problems. It is expected that a draft report on the review of BAW-1564 will be forwarded to the NRC by ORNL in early December 1979 with a final report scheduled by-the end.of the year.
D. Thatcher, Reactor Engineer Systems Group B&O Task Force Office of Nuclear Reactor Regulation cc: See attached sheets OFFICE~1 B&O B&O TF I
B4&T.,.B SURNAM+
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.... RSA'? l W DATEp 12/13/79 12//7/79 12/&1/79 12/jf /79 NRC FORM 318 (9-76) NRCM 0240
- U.S.
GOVERNMENT PRINTING OFFICE: 1979-289-369
ENCLOUIRE 1 OAK RIDGE NATIONAL LABORATORY OPETATED BY UNION CARCIDE CORPORATION NUCLEAR DIV!SION POST OFFICE BOX X OAK RIDGE, TENN4ESSEE 37830 October 15, 1979 R. M. Satterfield, Chief Instrumentation and Controls Systems Branch Division of Systems Safety Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
Request for Supplemental Information:
Review of Integrated Control System Analysis Report BAW 1564, August 1979 Review of the subject document is continuing and in the-course of this review a number of questions have arisen concerning the techniques used and the results presented in the report, which we would like to have answered by B&W.
In addition we have identified some subject areas in which we have not posed specific questions but would like to discuss with B&W so that we might better understand the bases and background of the information and recommendations presented.
The questions and subject areas are attached.
We would like to propose that some of the ORNL and SAI staff members involved in the review visit the B&W facilities to discuss the infor::ation we seek. We believe that this would provide a more efficient exchange of information than correspondence, considering the short time allocated for this review. We request that you arrange such a meeting with B&W at your convenience.
->sincerdy, John L.
Anderson Reactor Systems Group Instrumental ion and Controls Division JLA:cwl cc:
Director, Division of Systems Safety, NRC F. R. Nvnatt Attn:
B. L. Grenier J. R. Penland, SAI L. Beltracchi, NRC R. S Stone R. Brodsky, DOE
. Thatcher, NRC S. J. Ditto D. B. Trauger H. N. Hill
Questions Regarding Report BAW-1564, Integrated Control System Reliability Analysis
- 1. There may be a significant difference between failure modes or conditions with an FMEA based on functional block diagrams rather than equipment block diagrams. Have the functional failure assumptions been compared with actual equipment failure modes to assure that they are realistic and meaningful?
- 2. The ICS signal input failure assumptions appear to be all either "high" or "low" with some attempt to identify the "worst case."
Some of the operable plants under review have the potential for mid-scale failures.
There is reason to believe that some mid-scale failures may be worse than high or low failures, as experienced by the plant selected as typical, Rancho Seco. Are there plans for including.mid-scale failures in the analysis and how is the validity of the analysis compromised by not including mid-scale failures?
- 3.
Virtually all of the events/failures considered in the analysis appear to be based on "normal" conditions wherein all plant equipment is functioning at nominal design points.
Our limited information regarding operating experience suggests that many of the abnormal occurrences were the direct result of s:'Ie plant equipment not functioning. For example:
Three primary pumps instead of four running; one instead of two feedwater pumps running; one or more hand/automatic stations in manual; etc.
Since these seem to be the more significant initial conditions for unsatisfactory ICS performance, how is their omission justified? Are any of these "interesting" events analyzed but unreported?
- 4. What process was used to'determine the "effect on the NSS"? Neither the technique nor the justification is included in the analysis.
What verifiration techniques were employed for the "effects" analysis?
- 5. The POWER TRAIN code obviously has limitations to its ability to simulate the NSS and BOP responses.
How significant is this limitation on the analysis?
In particular:
a) Describe the extent to which the simulation was used to predict results.
b) Describe errors and uncertainties which might have resulted from the liqi;ted dynamic range and functional detail of the simulation.
c) Describe to what extent the simulation results were verified with plant data.
d) Describe the extent to which the simulation is valid or invalid for each of the individual plants and their differences, especially feedwater systems.
e) Does the simulation have capability for dealing with off-normal operatflm such as three primary pumps or partial manual operation?
71
- 6. The ability of the ICS to respond properly to its design basis and other probable conditions is not addressed. That is, design problems associated with normal operation or maneuveri-g are not included unless a failure is supposed. This may be outside the scope of the NRC request, but the ICS feedwater systoms interactions evidenced in operating plants indicate this may be of valid concern. Have the design problems and component limitations associated with expected normal operation been 'analyzed and documented? Are these analyses available?
- 7. Is there any connection, physical or phenomenological, between RPS sensors and ICS inputs? Which common signals, if any, initiate trip and what is the potential for common signal or signal conditioning failures initiating a plant transient through the ICS requiring RPS response derived from that signal.
- 8.
FMEA categories for "causes", "detection", "propagation potential" would yield helpful information.
Has this type information been generated and is it available?
- 9. The impact of pcwer supply failures appears to be inadequately addressed, especially considering that events of much more significance than those analyzed have occurred at operating plants. How is the omission of these considerations justified and is more comprehensive power supply failure analysis available?
- 10. A significant number of trips appear to have occurred when portions of the system were in manual. What fraction of time is it estimated that control stations.are in manual, and what are the problems associated with this mode of operation of the ICS.
ll.* How does historical failure data on. CS 721) and 320 compare with predictions based on nominal behavior?
Is there any evidence or accelerated failure?
- 12. Multiple failures are not treated although it is acknowledged by B&W that many failures.are not annunciated and therefore may exist until other
'failures occur, resulting in effective multiple failures.
It appears that multiple failure situations may have significant probability of occurrence.
How is the omission of multiple failure considerations justified in the analysis? Might Fault Tree Analysis have been a better technique for addressing the concerns and producing the results requested?
- 13.
The analysis does not include information to substantiate the recommendation that improvement is needed in power supplies, signal selection and signal reliabirity. Please supply the analysis or information which lead to this recommendation. In particular, does 7&W have specific reco=mendations to improve the failure tolerance of the ICS?
- 14.
Operating experience reports and oral information not included in the analysis suggest the ICS and/or the BOP system including the OTSG is sensitive to "tuning" and component problems such as feedwater valve speed and leakagE.
Describe the extent to which these problems are significant, how they have led to misoperation and RPS challenges, and how they might be avoided. Ace "tuning" problems inherent to this type of plant or do they represent desi'.n deficiencies which can be corrected?
- 15. Many Licensee Event Reports as.well as this analysis indicate that the operator is implicated in a large number of occurrences of poor ICS operation. Many of these events also involve slightly off-normal conditions such as non-standard pump and valve alignment. Do these events represent eQr ipr r
- ifciency, onerator trainir: deficiency or a combination of these?
Does B&W have recommendations-to correct tEese deficiencies and on what schedule can they be implemented?
SUBJECT AREAS AND SPECIFIC INFORMATION REQUESTED Rancho Seco equipment block diagrams and logic diagrams.
Identification of difference in ICS between Rancho Seco and each of the other 177 FA plants.
Identification in the FMEA of which "effects" resulted from POWER TRAIN (PT) simulation and which resulted from engineering judgement.
Any information on how ICS interacts with feedwater system oscillati-nt Any real event data would be.particularly helpful.
In Table 4.4, what are specific indications to operator for display failures?
ICS design basis.
PT-11 manual and QA file.
Tentification of power levels at which trips occurred.
-=rct information on operatcr/technician induced trips.
Confidence levels on ICS MTBFs.
-V.
,1*
IKS ELiIALILITY ANALYSIS MIEETING OCiOBER 23, 1979 Rc'n Finnin B&
Art 2 c~ride SAI T J. R. Penland SAI John Cole Duke Ron Ir own Duke L. L.
Byr cob :aiw Consuxers Power Company Larry Stalter Toledo Edison Gary nnett Stephen Ditto ORNL 2ohi.Indorson ORNL od Sterf d
NRC Dale Tha-her I RC B.
- W C.
BABCOCK & WILCOX OPERATING PLAN 9
Mr. William 0. Parker Jr.
Vice President -
Steam Production Duke Power Company P.O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. William Cavanaugh, III Vice President, Generation, and Construction Arkansas Power & Light Company Little Rock, Arkansas 72203 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S.Street P.O. Box15830 Sacramento, California 95813 Mr. Lowell E. Roe Vice President, Facilities Development Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, Ohio 43652 Mr. W. P. Stewart Manager, Nuclear Operations Florida Power Corporation P.O.
Box 14042, Mail Stop C-4 St. Petersburg,Florida 3 3733 Mr. R. C. Arnold Senior Vice President Metropolitan Edison Company Parsippany, New Jersey 07054 Mr.
James H. Taylor Manager, Licensing Babcock & Wilcox Company rower Generation Group P.C. Box 1260 L3nchburg, Virginia 4505