ML19296C489
| ML19296C489 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/11/1980 |
| From: | Bridenbaugh D, George Minor CALIFORNIA, STATE OF |
| To: | |
| References | |
| NUDOCS 8002260319 | |
| Download: ML19296C489 (21) | |
Text
Js.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of:
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)
SACRAMENTO MUNICIPAL UTILITY
)
DISTRICT
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Docket No. 50-312 (SP)
(Rancho Seco Nuclear Generating )
Station)
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Prepared Direct Testimony of Dale G.
Bridenbaugh and Gregory C. Minor Concerning Operator Training and Human Factors Engineering February 11, 1980 Sponsored by the California Energy Commission n
8002260 G
3 (C)
-1
p.
Prepared Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor Concerning Operator Training and Human Factors Engineering I.
Introduction Mr. Minor is a partner in MHB Technical Associates, a consulting firm located in San Jose, California.
His educational background is in electrical engineering (B.S.,
University of California at Berkeley, 1960; M.S.,
Stanford University, 1966).
In addition, he participated in General Elactric Company's 3-year Advanced Course in Engineering, graduating in 1963.
A full description of his experience and background has been provided in response to discovery requests.
During the period between 1960 and 1976, Mr. Minor was employed oy the General Electric Company in engineering and management positions involving the design of components and systems for use on nuclear reactors.
These systems included reactor monitoring, control, and safety systems.
Between 1972 and 1976, he was Manager of Advanced Control and Instrumentation Engineering, responsible for designs of new safety and control systems.
These included the design of new control room concepts involving new display and control techniques for use as a man-machine interface.
Specific emphasis was placed on human factors in the operator interface.
Mr. Minor is presently a consulting engineer with MHB Technical Associates, consulting on nuclear power issues for public and private organizations at a state, national and
.t s
international level.
Also, he was recently a participant on a Peer Review Group of the NRC/TMI Special Inquiry Group regarding both the accident sequence and tha human factors sections.
He is a member of the Nuclear Power Plant Standards Committee for the Instrument Society of America.
Mr. Bridenbaugh is also a partner and technical consultant of MdB Technical Associates.
He is a graduate engineer, familiar with the operation of nuclear generating plants, including operating difficulties that lead to reductions in nuclear power plant reliability and operability.
He received his Bachelor of Science in Mechanical Engineering from the South Dakota School of Mines & Technology in 1953.
From June, 1953 until February, 1976, he worked as an engineer and manager with the General Electric Company on a wide variety of most of the aspects of power generation equipment design, manufacture and operation.
During the last ten of those 22 years, he was in management positions in the General Electric Nuclear Energy Division where he had the responsibility to manage the monitoring of operation of nuclear power plants and implement solutions to operational problems.
In these positions, he monitored the performance of both boiling water reactors and pressurized water reactors and was cognizant of the performance record of large fossil generating stations.
For approximately five years during this assignment, he was also responsible for managing the corrective action programs required to resolve contractual complaints for the commercial nuclear power reactors supplied by General Electric, both domestic and oversees.
An additional duty held during this period of time was to develop 2.
a Nuclear Division Master Performance Improvement Plan, aimed at bringing about tne long-term improvement of boiling water reactor performance.
Prior to the management assignment in the Nuclear Energy Division, he spent several years as a field engineer at the first large scale commercial nuclear plant built by General Electric Company for Commonwealth Edison Company at Dresden, near Chicago, supervising the construction, start-up and modification, and repair of various portions of the plant.
He was also responsible during this time for acting as the General Electric Site Manager during the first major refueling and maintenance outage conducted at the Dresden plant.
For the past three years, Mr. Bridenbaugh has been a partner and technical consultant on energy with his consulting partnership, MHB Technical Associates.
In this capacity, he has provided technical advice to various governmental bodies and individual groups on subjects primarily related to the design and operation of commercial nuclear power plants.
As an example of this type of work, in 1978 he served as a consultant to the United States Nuclear Regulatory Commission to review the URC Plan for Research to Improve the Safety of Light-Water Nuclear Power Plants, sub-sequently documented in NUREG-0438, usued April 12, 1978.
He has also served in various consulting capacities to the General Accounting Office, the States of New Jersey and Illinois, and Suffolk County, New York, and to the governments of Norway, and of Sweden in the evaluation of nuclear programs.
Mr. Bridenbaugh is a registered professional nuclear engineer in tne State of California, helding Certificate No. 973, and is 3.
s also a member of the American Nuclear Society.
Additional detail of his experience is included in his resume produced in response to discovery requests.
II.
Discussion of Operator Training Issues The purpose of this testimony is to assess the adequacy of the Rancho Seco operators and the adequacy of their training to assure they can perform the actions necessary to deal with normal, abnormal and emergency operating conditions associated with the B&W nuclear system utilized at the Rancho Seco plant.
Specifically, this testimony will focus on the issues accepted by the Atomic Safety and Licensing Board'r (Board) Order Ruling on Scope and Contentions, dated October 5, 1979, as defined by the following questions:
a.
Board Question CEC 1-7.
Do the operator training actions responding to Subparagraph (d) of Subparagraphs a-e for Rancho Seco fail to give sufficient attention to providing appropriate analytical bases for operator actions?
(Subparagraph (d) states:
" Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.")
b.
CEC 3-1.
Whether personnel adequately understand the mechanics of the facility, basic reactor physics, and other fundamental aspects of its operation?
c.
CEC 3-2.
Whether personnel are properly apprised of new information pertinent to the facility's safe operation and ability to respond to transients, particularly information on operating experiences of other reactors?
4.
o d.
CEC 3-3.
Whether NRC and SMUD adequately ensure that emergency instructions are understood by and are available to plant personnel in a manner that allows quick and effective implementation during an emergency?
e.
Board Question Hursh & Castro No. 35 Rancho Seco, being a Babcock and Wilcox designed reactor, is operated by personnel and management whose competence has not been adequately tested and evaluated; namely, testing has not been conducted as to whet such employees can act responsibly and approp.' ;ely to make judgment decisions during a loss of feedwater transient, personnel interviews have not been conducted to properly evaluate the test results with such employees and some employees have never been tested because of grandfathering and, therefore, is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
f.
Board Question Hursh & Castro No. 34.
Rancho Seco, being a Babcock and Wilcox designed reactor, has not adequately trained unlicensed operators to respond to orders necessary for action which would be required in the event of loss of feedwater transient and, therefore, is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
These questions can be more simply stated as follows:
1.
Do the operators possess sufficient understanding of the analytical bases of operating procedures and do they possess an adequate knowledge of the fundamental (physics) aspects of plant operation?
2.
Is an effective _ procedure in place to ensure that new procedures and information are communicated to the operators?
3.
Are emergency instructions understood and effective?
4.
Has individual testing of the operators been 2dequate to ensure each has proper understanding?
5.
Are tha unlicensed operators properly trained to respond to emergency procedures?
5.
s Each of these fundamental areas are addressed in the following discussion.
A.
Analytic Basis of Operating Procedures and Fundamental Understanding of Plant Operation.
It is essential that personnel responsible for decisions regarding plant operation possess a high degree of understanding of the bases underlying the operating procedures.
The evident failure of TMI operators to have such an understanding was apparently a significant contributing factor to the severity of the TMI accident.
Much discussion and debate has taken place in past years over the question of extensive automation vs. human control.
In general, since it is virtually impossible to predict all possible sequences of nuclear plant accidents, the practice has been to automate for approximately ten minutes and to call for operator action shortly thereafter.
Since no procedure can ever address all possibilities, the ability to observe and analyze is essential.
This requires a thorough understanding of the bases for the procedures developed.
This is particularly true after TMI because the procedures adopted since that accident place heavy new responsibility on operators.
Various reports, procedures, training programs, interrogatory responses and depositions have been evaluted to determine if some general conclusion could be reached regarding the under-standing of operating procedures to be found among the Rancho Seco operators.
The preliminary finding after reviewing this material is that SMUD in general complied with the letter of existing requirements.
Numerous records have been produced 6.
s demonstrating the attempts to explain the underlying bases of procedures to the operators for the procedures adopted in the post-TMI era.
The essential question is, however, have such efforts been soccessful?
The ultimate answer to that question will only be revealed through time.
However, some indications are now available.
A caution against expecting too much is available in one study commissioned by the NRC's internal review headed by Mitchell Rogovin.
The Human Factors Evaluation conducted by the Essex Corporation 1/ finds, for example, that:
In general, the first (limited) definition of human error holds the operator responsible for most mistakes and has one pervasive remedy for errors -
more, and more effective, training.
The operator is expected to learn how to operate control panels regardless of the quality of panel design or procedure.
However, when errors occur where poor design or procedures are causal factors, improved or increased training will not of itself resolve the problem.2/
This seems particularly applicable to the issues considered in the first question (CEC 1-7).
The evidence indicates that all actions suggested or requested by the NRC Order have been followed, but this still may not be " sufficient" to assure that proper action can be taken in all emergency situations.
A substantial amount of uncertainty and lack of understanding was exhibited in the depositions of the three licensed operators 1.
NUREG/CR-1270, Human Factors Evaluation of Control Room Design and Operator Performance at Three Mile Island-2 (Final Report).
2.
Id., p.
110.
7.
s conducted on January 24-25, 1980.
For example:$!
- Lack of knowledge concerning length of time in which the OTSG will go dry.
(Tipton, p. 16.)
- Lack of knowledge concerning basis for concern re vessel weldments.
(Morisawa, p. 72.)
- Uncertainty regarding conflicts between procedures and between procedures and technical specifications.
(Morisawa, pp. 66-69 and Tipton, p. 56.)
- Uncertainty concerning need to take action at HPI pump runout.
(Tipton, pp. 43-45.)
- Uncertainty regarding action to be taken regarding RCP during an over-cooling transient and effectiveness of natural circulation vs. reflux boiling.
(Tipton, pp. 71-75 and Morisawa, pp. 37-38.)
Of additional concern is the attitude that there are no potential problems (the Kemeny Report called this "mindset").
An example of this is contained in Mr. Comstock's deposition wherein he asserta that the B&W system is far superior to the Westinghouse system with regard to feedwater transient response.4/
While each person is entitled to his own opinion, the majority of the opinion at this time seems to view B&W systems as significantly more susceptible to transients.
In fact, at a recent NRC Staff review of this issue with the ACRS, statements were made that the B&W design " places so much responsibility on the operators."5/
This implies a need to be doubly sure that B&W operators fully understand that fundamentals as well as the weaknesses of this design.
3.
Examples cited are from depositions of Daniel E. Comstock, Wayne S. Morisawa and Dennis E.
Tipton, conducted January 24 and 25, 1980.
4.
Deposition of Daniel E. Comstock, January 24, 1980, p.
9.
5.
ACRS Subcommittee meeting, January 8, 1980.
8.
o In conclusion, we find there is no assurance that SMUD operators have an analytical understanding significantly better than that of the TMI operators.
B.
Procedures for Communication of New Information.
SMUD's general procedures for the conveying of new information significant to safety are described in the December 4,
- 1979, 22.bI The weakness in Set No. 2 answers to Interrogatory No.
the process described is the apparent lack of a requirement that such information be passed on to the shift crews.
For example, the statement is made that:
Events which occur at other units and come to the attention of the facility management can also be promulgated (Emphasis added.)
Further, it is stated:
Significant events or potential problems can also be discussed in the routine retraining program.
(emphasis added) and:
The annual one-week simulator course at the B&W Training Simulator provides an opportunity for operators to see and practice transients which have occurred at other B&W units.
(Emphasis added.)
The fact that such things "can be" done does not ensure that they are done.
This is evidenced again by the operators' depositions where it is found that:
No transients other than TMI have been discussed.
(Tipton, p. 97.)
Procedure changes are not formally transmitted.
(Tipton, pp. 94-95.)
6.
Licensee's Answers (Set No. 2) to the California Energy Commission's First Set of Intarrogatories dated November 15, 1979, pp. 18-19.
9.
r
/
- No formal discussion or information concerning the September 21, 1979, North Anna event was apparently conducted.
(Morisawa, p.
73 and Tipton, p. 97.)
Additionally, it was asserted that:
- No formal procedure exists to ensure that operators actually read the Standing Order (changes).
(Tipton,
- p. 95.)
No system exists to make NRC (NUREG) reports readily available to the operators.
(Tipton, p. 139.)
The system for the communication of operating experience, procedure changes and other information helpful in developing a depth of understanding appears to be haphazard and in need of improvement.
At a minimum, there needs to be a means to ensure that new procedures and significant events are promptly communicated to operators in a manner designed to make certain that the events and procedures are thoroughly understood by operators.
C.
Effectiveness of Emergency Procedures.
Use of emergency procedures was considered at some length in the depositions of Tipton and Morisawa.2/
Both of these discussions highlighted the difficulty of dealing with complex emergency procedures while responding to a transient condition.
Not the least of the problem is determining which of several pro-cedures actually applies.8/
It is also indicated that SMUD has committed to the NRC that the operator will memorize the immediate action steps.EI It is not clear from the depositions whether the 7.
Ibid.
3, Tipton, pp. 54-59; Morisawa, pp. 66-67.
8.
Ibid.
3, Tipton, p. 56.
9.
Ibid.
3, Tipton, p.
142.
10.
operators accept that commitment as being a requirement, as heavy reliance on the written procedur? is described.
At a minimum, all operators should be required to memorize the steps of the main emergency procedures (such as turbine trip / reactor trip and loss of steam generator feed) and demonstrate ability, periodically, to use them and to understand the interrelationship of the various procedures.
D.
Effectiveness of Operator Testing.
Some questions must always remain regarding the effective-ness of the operator training testing program.
It is not possible to test all operators for all possible sequences under the real environment of time and stress.
Criticism has been levied by various review committees.
The Kemeny Report found that:
The agency should be directed to upgrade its operator and supervisor licensing functions.
These should include the accreditation of training institutions from which candidates for a license must graduate.10/
The study commissioned by the Rogovin review found that the TMI training was in full compliance with regulatory standards but was still deficient.11/
The implication is that the standards are inadequate or, at the least, inadequately followed.
SMUD's training program is not substantially different from that used at TMI.
The same simulator is used and the course content is basically the same.
Until new standards are adopted, a question as to its effectiveness must exist.
It has been reported 10.
Report of the President's Commission on the Accident at Three Mile Island, p.
63.
11.
Ibid. 1, p.
100.
11.
by the NRC that numerous studies and updates of training standards are underway.12/
There is also a study underway by General Physics to develop a performance measurement system for training simulators.13/
This is reported to be a computer based system to evaluate training performance.
All of these improvements are needed to make the assessment of training a more exact task.
Based upon the informa-tion be have reviewed, SMUD operators' training appears to be similar to that received by TMI operators and, accordingly, there is no basis to conclude that they have adequately been trained to respond to off-normal conditions.
E.
Training of Unlicensed Operators.
The ability of unlicensed operators to deal with emergency situations received substantial attention in the NRC's preliminary assessment of the TMI accident.1S/
As reported therein,15/ the nonlicensed operators may perform many essential and critical tasks such as the closing or opening of safety related valves, transfer of radioactive wastes, etc.
There is reason to be concerned regarding the general informality of the training of nonlicensed operators.
A good description of the "on-the-job" 12.
NRC Staff Responses to California Energy Commission's First Set of Interrogatories to the Nuclear Regulatory Commission, December 11, 1979, pp. 29-33.
13.
EPRI NP-783; Interim Report, " Performance Measurement System for Training Simulators," May, 1978.
14.
NUREG-Oo00, Investigation Into the March 28, 1979, Three Mile Island Accident by Office of Inspection and Enforcement.
15.
Id., pp. I-2-50-53.
12.
training program is found in the deposition of Dennis Tipton.bb!
This "on-the-job" training program means that unlicensed operators may not know how or where to perform certain actions the first time they are called upon to perform them.17/
If the first time is an emergency requiring unlicensed operator action, they may not be sufficiently trained to respond properly.
The issue of nonlicensed operators (as well as nonlicensed management) is continuing to receive much attention in the on-goins reviews.
It is recommended that these reviews be closely followed, that SMUD commit to improvements in such methods that may be recommended, and that a formal program be developed and documented as soon as practical.
F.
Conclusions on Operator Training.
There is substantial reason to judge the operator training and level of understanding at Rancho Secc as inadequate.
While SMUD has attempted to demonstrate that the training program meets all industry standards, there is no reason to believe that this produces an adequately trained operator.
The general agreement that industry standards in the past have been inadequate, coupled with the greater demands imposed on the operator by the greater sensitivity of the B&W system, point to a substantial need for improved training methods at this plant.
16.
Ibid.
3, Tipton, pp. 109-114.
17.
Ibid.
3, Tipton, pp. 113-114.
13.
III.
Discussion of Human Factors Engineering Issues This portion of our testimony will address the two issues designated by Board Question CEC 5-3a related to the adequacy of instrumentation at Rancho Seco, and Board Question Hursh & Castro No. 31 related to the adequacy of the Rancho Seco control room design.
A.
Board Question CEC 5-3a.
Are the special features and instruments installed at Rancho Seco adequate to aid in diagnosis and control after an off-normal condition engendered by a loss-of-feedwater transient?
The instrumentation in the Rancho Seco centrol room is adequate to meet the minimum requirements for operating the reactor but has several limitations during off-normal conditions.
In an effort to improve the ability to respond to a feedwater transient and/or loss of feedwater accident, several instrumentation changes were required to be implemented as a result of the Lessons Learned Task Force (short termkb1!
The effect of these changes is to add information for the operator to use in making his decision about the status of the Auxiliary Feedwater System, particularly during a transient.
However, this is not to say that all off-normal conditions are now adequately instrumented.
The basic weakness in the instrumentation systems identified by most of the major studies of the TMI accident was the inability to directly know the water level in the reactor vessel or more generally to know when the saturation conditions are reached (i.e.,
when the reactor coolant starts boiling and voiding).
17.
N'JREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations, USNRC, July, 1979.
14.
The present instrumentation system requires inference from two.or more indicators to determine if the reictor vessel is filled (i.e., pressurizer level and coolant parameters).
Even if the vessel is diagnosed to be underfilled, this method cannot tell the operator the amount of coolant lost and the actual level in the vessel.
Given the intensive focus on the Transient /AFW/PORV/LOCA accident sequence, it is unlikely that an operator will improperly diagnose this particular problem in the near future.
But in the long term, the operator's ability to diagnose an off-normal condition involving loss of coolant in the face of a yet-undiscovered series of obscure failures wou? d be enhanced by a direct indication of vessel level during saturation conditions.
The NRC Staff, following the TMI accident, recommended that PWRs be provided with a more direct reading of vessel water level.18/
However, because of the complexity of accurately measuring reactor water level in a PWR vessel, it would be necessary to research this problem carefully to assess the best method for obtaining such a measurement.
Even in the relatively unhurried period of the post accident analyses, researchers had difficulty accurately estimating the core water level history using the recommended measurement technique (i.e., using pressurizer level and reactor coolant parameters) and, therefore, relied on such indirect means as 18.
NUREG-0560, Staff Report on the Generic Assessment of Feedwater Transients in Pressurized Water Reactors uesignea Dy B&W Company, USNRC, May, 1979.
15.
abnormal behavior of in-core neutron detectors to establish level.19, 20/
The NRC has also required PWRs to install a "subcooling meter" and " additional instrumentation" to detect inadequate 22/
core cooling.
SMUD has committed to comply with these Lessons Learned Requirements, but the details of the changes to be made and the range of plant conditions the changes will cover are not clear at this time.
Another area of uncertainty in attempting to diagnose off-normal conditions is in attempting to detect with certitude the initiation of natural circulation.
Presently, the operators are required to read out several parameters and make a judgemnt as to whether the plant has achieved natural circulation.
Unfortunately, these readings are not always reliable or available (e. g., the thermal couples readings rely on computer availability),
which makes the operator's task more difficult.
The operator would be less likely to make errors in diagnosis if he were provided with a dedicated indication of natural circulation which was reliable under all off-normal conditions.
This problem is particularly important on B&W plants which have a lower driving head due to the lower position of the steam generators relative to the reactor vessel.
19.
NSAC-1, Analysis of Three Mile Island-Unit 2 Accident, NSAC, July, 1979, Appendix CI.
20.
NRC/TMI Special Inquiry Group, Draft Report, Dec., 1979, Accident Sequence Section.
21.
Ibid. 17, Section 2.1.3.b (2.1.3.b).
22.
Letter, October 30, 1979, Harold Denton (NRC) to All Operating Nuclear Power Plants, subject:
Discussion of Lessons Learned Short Term Requirements, pp. 9-14.
16.
It is our belief that the ability to quickly diagnose the Rancho Seco plant would be enhanced by the foregoing additional types of displays and that without them the present instrumentation and measurements place an undue burden on the operators.
B.
Board Question Hursh & Castro No. 31.
Rancho Seco, being a Babcock and Wilcox designed reactor, has a control room configuration which is poorly and inadequately designed for plant operators to avoid a loss of feedwater transient, and there-fore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
Compared to the TMI-2 control room, the Rancho Seco control room appears to have several significant advantages from a human factors point of view.23/
It also has some weaknesses.
Some of the major differences are as follows:
It is less conjested; it is sma.'.ler, it has fewer feet of inner consoles and front row vertical panels than TMI; and it has fewer displays; but it has over 100 feet of panels in the b :k room; it lacks physical diversity in control; it uses very few mimics; and it employs numerous vertical moving-pointer arc-scale meters mounted above eye level.
On the whole, it appears that the advantages outweigh the disadvantages, leading to a better design at Rancho Seco than that at TMI-2.
However, this does not mean that the Rancho Seco control room is optimally designed for handling feedwater transients or any other upset condition.
The design appears to be 23.
We have not had an opportunity to inspect the Rancho Seco control room before preparing this testimony.
Thus our views concerning the Rancho Seco control room may change following the scheduled inspection.
17.
optimized for normal operation but may be lacking the needed displays and reliable data to handle upset conditions.
The added fact that the B&W design has inherent sensitivity to feedwater transients may amplify the importance of human factors deficiencies in periods of high stress.
In general, essentially all nuclear control rooms are inadequate and poorly designed from a human factors engineering point of view.
This view is substantiated by various studies such as the Lockheed/EPRI Study which states:
The study [of five operational control rooms] revealed both major and minor problems in the design of con-trol rooms which increased the potential for operational errors and unnecessarily added to the training burden and rigor of selection criteria for operator candidates.
In short, the control boards reviewed had not been designed to promote error-free operation, especially during potentially stressful circumstances.24/
The Essex Study of TMI-2 was even more emphatic about inadequacies in control room development.
Their findings stated:
- Human engineering planning at TMI-2 was virtually nonexistent.
- NRC and the nuclear industry have virtually ignored concerns for human error.25/
Rancho Seco is not identified as being one of the plants evaluated in the above studies.
NRC review of control rooms during the Rancho Seco licensing was cursory or non-existent.
The Kemeny Commission called for a correction of this shortcoming in their findings:
24.
Human Factors Methods for Nuclear Control Room Design, EPRI NP-ll19-SY, June, 1979, p.
1-1.
25.
Human Factors Evaluation of Control Room Design and Operator Performance at Three Mile Island-2, NUREG/CR-1270, Essex Corporation, January, 1980, Vol.
1, p. 99.
18.
Other safety emphasis should include review and approval of control room design; the agency should consider the need for changes in the overall design to aid understanding of plant status, particularly in response to emergencies.26/
It is our opinion that Rancho Seco should be evaluated against consensus standards for human factors engineering.
Also, the on-going NRC funded studies of human-factors and the man-machine interface should be extended to existing control rooms, including Rancho Seco, to evaluate possible enhancement and improvement.
C.
Conclusions Regarding Human Factors Engineering.
It is our opinion that operation of the Rancho Seco reactor without direct indication of (1) reactor vessel coolant level, (2) the onset of saturation conditions, and (3) the initiation of natural circulation, and in a control room environment which is not designed with optinimum human-factors considerations, could contribute to errors in diagnosis and control of upset conditions.
26.
Report of the President's Commission on the Accident at Three Mile Island, October, 1979, p. 63.
27.
NRC Staff Response to First Set of CEC Interrogatories, dated December 12, 1979, Response No. 13.
19.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of:
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312(SP)
DISTRICT
)
)
(Rancho Seco Nuclear Generating)
Station)
)
)
AFFIDAVIT OF GREGORY MINOR Gregory Minor, being duly sworn according to law, deposes and says as follows:
I have prepared and am familiar with the attached document entitled " Prepared Direct Testimony of Gregory Minor and Dale Bridenbaugh".
The opinions set forth therein are my own and, to the best of knowledge, the facts set forth therein are true and correct.
I Dated:
Feb ruary 8,1980 eb o 651 NE GREGOR'l MINDR Sworn and subscribed before me this 8th day of February, 1980.
OT n t - \\N\\ % L h m u A Nothry Public OFl ICIAl. SEAL
,..Q MARY McDEARMID d NoT 'RY PUBUC CAUFOWNI A.
!Igg/ P.incipal0fficein SACRAMENTO Can:y d
v 4.,
J My Commiten Erpires Feb. 20.1930
' e-~ ; 'n ~; ;, ;-,. ; :-y-;, ;-- J
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of:
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312(SP)
DISTRICT
)
)
(Rancho Seco Ni';1 ear Generating)
Station)
)
)
AFFIDAVIT OF DALE BRIDENBAQQE Dale Bridenbaugh, being duly sworn according to law, deposes and says as follows:
I have prepared and am familiar with the attached document entitled " Prepared Direct Testimony of Gregory Minor and Dale Bridenbaugh".
The opinions set forth therein are my own and, to the best of knowledge, the facts set forth therein are true and correct.
/
d Dated:
February 8, 1980 kN [
be d- -
DALE BRI'DENBAUGH
/
Sworn and subscribed before me this 8th day of February, 1980, tom. - @%. M_ '
ci Nohary Public
..1..
OFFICIAL SEAL t
4 MARY McDEARMID
- l NoT ARY PUBLIC. CALIF fWNI A i,
d
~
Psincipal Office in SACPA'!Ulf 0 Caun'y d
My Contms:en Eroira Feb. 20.198'] ' >
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