ML19322C796
| ML19322C796 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/02/1979 |
| From: | Chipman G NRC - NRC THREE MILE ISLAND TASK FORCE |
| To: | Frampton G NRC - NRC THREE MILE ISLAND TASK FORCE |
| Shared Package | |
| ML19322C797 | List: |
| References | |
| RTR-REGGD-01.047, RTR-REGGD-01.078, RTR-REGGD-01.095, RTR-REGGD-01.097, RTR-REGGD-1.047, TASK-TF, TASK-TMR NUDOCS 8001240561 | |
| Download: ML19322C796 (42) | |
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'o UNITED STATES
$\\3_cqf('g NUCLEAR REGULATORY COMMISSION
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WASHINGTO N, D. C. 20555 oA !
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- August 2,1979
- +
MEMORANDUM FOR:
George Frampton, Deputy Director NRC/TMI Special Inquiry Group FROM:
Gordon L. Chipman Human Factors Task Force
SUBJECT:
HUMAN FACTORS WORK PLAN AND RESPONSE TO QUESTIONS We have completed the items you requested in your memorandum of July 19, 1979. The detailed work plan has been coordinated with all members of the task force, and it appears we have a working understanding of this Division of Labor. While there may be minor shifts in the future, the attached assignment and work scope should be considered as a working document.
The enclosed individual work plans have been developed in light of the overall work plan. The enclosed responses to your questions were prepared with input from all members of the Task Force.
We received the responses to the RFP, and expect to complete the evaluation processes and identification of the most qualified contractor today.
___._. N.early every_one wh.o. attend _ed_the_ B&W simulator _ pre.sen_tation was impressed with how tame the transient was after the first 20 minutes.
In light of this the scope of the task to construct a detailed time line and video tape a walk-through of the accident will be reevaluated after we have selected a contractor.
M(d4 /;fpme/a,.
G6rdon L. thipman Human Factors Task Force NRC/TMI Special Inquiry Group
Enclosures:
1.
Assignment and work scope 2.
Individual work plans 3.
Responses to questions cc:
R. Budnitz R. Haynes W. Johnston W. Parler J. Snell D. Allison 1r soo1 *o S Kt g
i J
Human Factors Task Forct ENCLOSURE 1 8/1/79 TASK OUTLINE HUMAN FACTORS EVALUATION OF CONTROL ROOM DESIGN AND OPERATOR PERFORMANCE AT TMI-2 I.
Backaround Following the accident at Three Mile Island Unit No. 2, the Commission established a Special Inquiry to assure that the NRC will have the fullest possible understanding of the events at Three Mile Island.
The purpose of that evaluation is to take whatever further steps may be necessary to prevent any similar accident in the future. A major area of investigation by the Special Inquiry is the response of the operating personnel to the events.
Specifically, the Inquiry must determine to what extent the control room design, operator training and selection, operator performance, and other factors, significantly influenced the sequence of events.
The work scope described below is essential to the completion of this objective.
II.
Soecial Inouiry Group Tasks l
Group Assigned A.
Operators /
1 o Examine background and experience prior to Met Ed employment n-Detennine educational background _. - _ _.. _ _ _ _ _ _ _ _.... _ _ _. _ _ _
1 o
1 o
Identify NRC requirements for selection 1
o Identify Met Ed requirements for selection 1
o Evaluate application of selection requirements to operators 1
o Identify NRC training requirements 1
o Identify Met Ed training program / requirements I
o Determine fonnal training of TMI-2 operators Curriculum
-- Lecture training Simulator training - how accurate simulation Instructor background Perfonnance of trainee Recurring training Training of Supervisors 1/ perators include:
CR operators, CR supervisors, plant auxiliary O
operators (Maint. Personnel) and appropriate management personnel.
s
' 1 o
Determine on job training for operators Duration Formality of program Capability of instructor (s)
Performarce of trainee Drills on watch 1-4
- o Evaluate training program Other utilities' programs comparison with Met Ed 1-4 0
Evaluate performance of operators in meeting NRC requirements (testresults,etc.)
o Actions / Inactions before the accident 4-2 Identify critical system malfunctions / misalignments 4-2
-- Identify human factors involvement in system malfunctions 4-2 Determine when and how recognition of malfunction was achieved
- 5..__ _ _4-2_._ _..*.__.. Determine when-and4cw correctien-of the malfunctions i
was achieved 2-3,4 o Actions / Inactions during accident Determine detailed sequence of events Identify significant operator actions / inactions B.
Precursors related to Human Factors 1 -2,3,4,5 o Identify significant precursors that could have impacted TMI-2 Determine the response by NRC, industry, Met. Ed. to each of these 1
o significant precursors.
1 o-For each precursor, determine what information was gained /should have been gained 1-2,4 o Determine what information feedback was utilized by NRC, Met. Ed.
l
-- To update training
-- To update procedures
-- To update plant To update control room l
Other
- These evaluations are to be performed in conjunction with the contractor.
9
. Compare emphasis on precursor events by Met Ed and other 1
o utilities C.
Control Room Design Identify NRC regulations and regulatory guides 1
o Identify published standards and recommended practices of
'I o
other organizations I
Identify the criteria utilized in TMI-2 CR design o
1 -2,4 o
Identify the CR design philosophy NRC philosophy Met Ed philosophy Vendor philosophy Architect / Engineer philosophy 1-2
- o Determine the dominant influence on TMI-2 CR design 2
- o Evaluate the conformance of CR design to human engineering principles L. -
_ _. _ _.... Compare design process to that utilized in other CR's of 2
- o the same vintage 2-
- o Evaluate the human factors considerations utilized in design of critical systems, controls and procedures 2
- o Compare CR design (from a human factors viewpoint) with designs of other complex man / machine systems NASA D0D Chemical Industry Nuclear Navy D.
Plant Design & Control (outside CR)
Identify NRC requirements for plant design and control I
o related to human factors Reactor and secondary system
~'
o Identify Met Ed influence on design I
- These evaluations are to be performed in conjunction with the contracto r, t
4-For plant control other ;han CR, determine 4-2,1 o
Miller, Doyle Human factors application Communication Signals in CR E.
Procedures Evaluate effectiveness of administrative (shift turnover) procedures.
4-l' o
Determine process for development of emergency procedures.
4-1 o
By whom Review and/or approval Update in light of precursors Evaluate effectiveness in time of emergency 4-2,1 o
Evaluate effectiveness in TMI-2 accident 4-2 o
Evaluate operators' use of procedures 4-2 o
4-2,1
- o. Examine need for simplification of procedures F.
- Evaluate,in conjunction with contractor, the adequacy of the All - _ - _.- fo-11owing,- -par-ticularly_as_they_reJa_ted to the accident
- C o NRC requirements Met Ed & their contractors and vendors in applying human o
factors principles Operator selection and training o
o Control room design o Feedback of information from precursors o Plant design 0 Emergency procedures
- These evaluations are to be performed in conjunction with the contractor.
~
. III.
Outside Contractor Tasks - Group 2, Contract Management Task A - Control Room Design at TMI-2 The Contractor shall:
1.
Identify the criteria which directly influenced the CR design as specified by the NRC and standards organizations.
a.
Review Title 10 of the Code of Federal Regulations, NRC Regulatory Guides and Standard Review Plans as provided by the NRC which dictated the design of the control room and point out those criteria which require the application of human engineering principals to such designs.
b.
Identify relevant standards and recommended practices published by organizations other than NRC which deal with nuclear power plant control room design.
Identify which of the criteria identified in 1.b. were c.
utilized in the design of the TMI-2 control room.
2.
Identify the actual design basis and operating logic which led to the s. - - - -- - - as-built -design-of-the -control-room..-_ Review.the design studies and analyses of Metropolitan Edison Co. and its associates as provided by the NRC leading to as-built design of the control room and deter-mine what human engineering principles were applied.
i i
=
s 3.
Determine if the CR was designed in accordance with the design basis and criteria identified in 1 and 2 above.
a.
Review control room contractual documents, Final Safety Analysis Report, Construction specifications and as-built drawing as pro-vided by the NRC to determine the human factors aspects of the con-trol room design.
b.
Visit Three Mile Island site for familiarization and to complete accurate descriptien of the control room in its pre-accident con-figuration.
In conjunction with the NRC Special Inquiry contract project manager c*
("NRC project manager") compare the human factors aspects of the actual design of the control room (as determined under 3 a and b above) with the criteria and bases that led to the design (as determined under 1 and 2 above).
d.
In conjunction with the NRC project manager, and using the results of 3.c., identify those implicit philosophical or broad based design concepts which had a significant impact on the human factors design of the control room (i.e. single failure concept),
c.._ _
_ _ _ __ e.
Determine _if the _quan.tity_and prominance of information presented in the control room are consistent with the design bases and criteria.
4.
Compare the design process for TMI-2 CR with that used in other nuclear plant control rooms of the same vintage.
a.
In conjunction with the NRC project manager, identify a limited number of plants (at least 2) of the same generation as TMI-2.
b.
Obtain reconnaissance level information (documents and discussions) on human factors criteria and design bases used.
Visit the control rooms identified above, and assess the degree to c.
which these designs were 6onstructed in accordance with their respec-tive criteria.
NRC will assist contractor in obtaining access to such control rooms.
- d. On a broad basis and in conjunction with the NRC project manager, compare the process that resulted in the application of human engineering principles to the design of the control room of TMI-2 with that of the plants identified in 4.a above.
Task B - Control Room A 2 h12y The' Contractor shall:
1.
Construct a full scale mock-up of the TMI-2 control room panels utilizing photographs for the panels identified in the table below.
The mock-up must be transportable in sections.
Drawings of the panels will be provided by NRC in conjunction with Task A.3.b.
Visit the TMI-2 CR to provide familiarity with the actual CR layout.
2.
Prepare a timeline diagram of the control room activities during the first 15D minutes of the a: ident.
a.
Using event chronologies and operatorinterviews provided by NRC, define operator activities, b.
NPC will identify the critical timeline actions / inactions within the control room which significantly influenced the outcome of the accident.
3.
Video tape an enactment of the timeline sequence of events for use in the analysis of operator performance.
4.
Based on the emergency procedures and other fomal guidance available to the operators, develop an idealized timeline.
5.
In conjunction with the NRC project manager, i'entify the control room design facters which influenced critical actions / inactions (2.b. above).
Emphasis should be placed on the most significant human engineering issues.
Table of Control Room Panels to be Modeled L._.
Control Rtiom Desk - CONS-1 Computer Console - CONS-2 Aux. Systems Control Console - CONS-3 Plant Control Console - CONS-4 Turbine Control Console - CONS-5' Electric Control Consoles - CONS-6A, 6B, 6C Fire Detection Panel - PNL-7 Coolan: Systems Monitoring Panel - PNL-8 Reactor Coolant Drain Tank Panel - PNL-8A Push Pull Control Panel - PNL-9 Plant Equipment Temp. Recording Panel - PNL-10 Radiation Monitoring Panel - PNL-12 SFAS Panel - PNL-13 Control Rod Drive Panel - PNL-14 Containment Isolation Panel - PNL-15 Turbine Supervisory Panel - PNL-16 l
Turbine Auxiliary Monitoring Panel - PNL-17 l
Station Electric Aux. Monitoring Panel - PNL-18 l
Vital Power Panel - PNL-19 Nuclear Instrumentatinn - CAB-2'], 21 HVAC Panel - PNL-25 Diesel Generator No.1 & 2 Panels - PNL-25, 29 Computer Programmers Console - CAB-188A
i Task C - Ooerator perfomance h'e Contractor shall:
1.
Determine the adequacy of the training program to assure the operators' capability to diagnose problems and take appropriate actions during nomal and emergency conditions.
a.
NRC will provide documents describing operator training program, TMI-2 emergency operating procedures and will make available NRC operator l
licensing personnel to describe the regulatory program.
i b.
In conjunction with the NRC project manager, determine if the operator training was adequate in particular with respect to the s9nificant actions / inactions taken by the operators on the TMI-2 accidec.t.
2.
Identify the basis for each significant action / inaction resulting from operator performance that cannot be attributed to inadequate training.
Where additional interviews with operators are required, this will be arranged through the NRC project manager.
Types of results that might be obtained include:
mismatched operator aptitude, poorly defined lines of authority or task assignments within the CR, etc.
3.
Evaluate the adequacy of the transfer of information between shifts, and between operators and maintenance personnel at TMI.
Review NRC and Met Ed requirements for infomation transfer and the implementation of these require-ments at TMI-2.
Compare the information transfer procedures of TMI-2 with those used in the plants identified in Task A.4.a. above.
k -- -Task D -- Acolication of Human Factors--Princiales to Control Room Desian Tne Contractor shall:
1.
Identify the systems components and procedures in the control room which played a critical role during the first 150 minutes of the accident.
NRC will identify the critical timeline actions / inactions (critical points) a.
I within the control room which significantly influenced the outcome of the accident (B.2.b. ) and provide applicable emergency procedures (C.l.a).
b.
For each critical ooint, identify the systems, components and procedures in the control room which did or should have played a role in the decision process.
i NRC will provide documentation of the chronology of events and existing c.
i operator interviews as necessary.
Requirements for additional interviews will be coordinated through the NRC contract manager.
2.
For each critical system, component and procedures identified in 1 above, identify the relevant human factors considerations.
This will include the factors in the relevant Human Factors engineering standards.
3.
Detemine the degree of compliance of the critical system and component l
designs and procedures to the applicable human factors principles (standards).
4.
Where areas of non-compliance are identified in 3 above, in conjunction with the NRC project manager, cetermine the impact on operator performance at critical points.
5.
Utilizing the information obtained in Task A and B and in 1-4 above, i,
and in conjunction with the NRC project manager, evaluate the inte-gration of the control room design with the reactor system design in the context of human factor program development.
This should include the utilization of task analyses of:
the role of the CR operator; generating CR staff selection and training requirements; development and testing of operational procedures (including emergency actions);
and the effectiveness of Licensee Event Reports (LERs) feedback.
6.
Identify the approach taken by other agencies and organizations in the design of comparable complex man / machine systems with respect to the application of human factor principles and one example of advanced CR design concept being offered by a U.S. nuclear plant supplier.
The agencies and organizations investigated should include comparable industries (chemical, etc.), the amed services and NASA.
The procedural and decision-making process employed by each selected crganization will be compared to the process utili:ed in the design of TMI-2.
Significant variations should be identified and their impact on the performance of the operation estimated in conjunction with the NRC managers.
IV.
Deliverables
- 1. For all tasks A-D a.
Letter status reports every 2 weeks.
b.
Preliminary final letter report of all findings by September 28, 1979.
c'.
NRC will provice coments Tf October T,19797nd Final Letter report of all findings incorporating NRC comments by October 10, 1979.
- 2. Contractor may be required to deliver the CR mock-up to the Washington, DC, metropolitan area before the termination of the contract.
If the mock-up is not requested by the NRC by contract temination, the contractor may dispose of the mock-up.
- 3. The contractor shall be available to brief NRC Comissioners and other groups (not to exceed 10 briefings or hearings) regarding their work and findings on an as needed basis.
- 4. Except as specifically authorized by this contract, or as otherwise approved by the Contracting Officer, records or other information, documents and material furnishing by the Commission to the Contractor in the performance of this contract shall be used only in connection with the work performed under this contract.
The Contractor shall, upon comple. tion or termination of this contract, transmit to the Comission all records or other infomation, documents and material, and any copies thereof, furnished by the Comission to the Contractor or data developed by the Contractor in the performance of this contract.
L
4 EliCLOSURE 2 WORK PLANS g.
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'M-RESPONSE TO QUESTIONS
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' ~ * *
- 1.
Describe briefly what kinds of licenses NRC grants and NRC's requirements for shift manning.
[
~
The NRC presently grants two kinds of operator licenses:
Reactor Operator (RO)
Senior Reactor Operator (SR0)
The educational and training requirements for the operators are described in Section 2 below.
The primary NRC requirement for staffing the control room with licensed, reactor operators is stated in 10 CFR 50.54(k):
"An operator or senior operator licensed pursuant to Part 55 of this chapter shall be present at the controls at all times during operation of the f acility."
Additional requirements are described in Standard Review Plan 13.1.2-2, the most pertinent of which are:
"(a) A licensed senior operator who is also a member of the station supervisory staff should be onsite at i l time s
-when at least one unit is loadec with fuel.'
"(b) For each control room from which one or more reactors are in operation, an additional operator should be onsite and available to serve as relief operator for that control room."
In addition to the aforementioned licensed operator requirements the manage-rent of _each station having one or more units containing fuel should:
- 1) either " qualify and designate at least one member of each shif t operating crew to implement radiation protection procedures...
or assign a health physics technician to each shift..."
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g
. 2)
Include in their shift crew assignmnts a Licensed Senior Reactor Operator, a Senior Reactor Operator limited to fuel 3
handling with no other concurrent operational duties, hpecial license " fuel handling foreman," to directly supervise the i
core alterations after initial fuel loading; 3) a fire brigade should be assigned f or each shift the minimum size of which is 5 persons - unless a specific site evaluation has been completed and some other number justified;
- 4) assignments of onsite shif t operating crews should have the following minimum (which includes licensed operators):
station with one unit: 3 people at all tims + 2 additional people when unit is operating multi-unit station: 3 people per unit
+ 1 additional at all times person per operating unit fiRC's minimum shift crew requirements for licensed reactor operators during power operation are shown on Table 1.
(Page 4)
Presently, there are no f;RC requirements for having a shift supervisor with greater technical skills than that of the Licensed Senior Reactor Operator.
Licensing exams for Reactor Operators and Senior Reactor Operations (initial exams vs. requalification or cross licensing exams) are made up by f4RC personne l.
These licensing examinations are administered and graded by NRC emplopes, as well as people from flational Laboratories, college professors
. who wcrk with research reactors; e.g., Oak Ridge National Laboratory Hanford Savannah River MIT U of Illinois Georgia Tech U of California Requalification exams and cross licensing exams are made up and administered by the utilities' training staff - with NRC audit review of the examinations and the results.
2.
Describe NRC's educational and training requirements for operators, senior operators, engineers.
The following list summarizes URC's guidelines for educational and training requirements for licensing nuclear reactor operators.
Reactor Operator selection High School Diploma or equivalent criteria Two years of power plant experience Minimum of one year at a nuclear power plant (meaningf ul operations, construction, or design work)
Training Requirements Cold training - for potential reactor operators who have no previous experience -
This program usually starts two years before fuel loading.
Phase I Basic Courses 12 weeks usually at nuclear training ceters or universities; includes 10 veeks of nuclear physics, health physics, chemistry, and plant technology and 2 weeks of practical operational training on a nuclear training or research reactor where the applicants participate in fuel loading experiments, coefficient measurement experinents, perform reactivity calculations, and manipulate controla during 10 reactor startups.
TABLE 1 MIf11 MUM SHIFT CREW RE0VIREMEf;TS FOR LICEf! SED REACTOR OPERATORS DURING POWER OPERATI0fi Total Number of Control Room Arrangement Type SR0 R0 Licensed Operators Single Unit A
1 2
3 Dual Unit Single Control Room B
1 2
6 Common Control Room C
2 3
5 Triple Unit Single Control Room D 1
2 9
Dual Single 1
2 3
+
Control Rm.
E Single Dual 2
3 5
Control Rm.
Common Control Room F 3
4 7
3.., _ _
EXCEPTIONS TO CURREllT STAFF POSITI0ft:
TECHil! CAL SPECIFICiT10f1S - MINIMUM ALLOWABLE LICEfJSED REACTOR OPERATORS DURIfiG POWER OPERATI0f4 f;uclear Plant Type SR0 R0 Comments Browns Ferry 1, 2, & 3 E
3 4
Short i R0 La Crosse A
l 1
Short 1 R0 Oconee 1, 2, & 3 E
3 4
Short 1 R0 Point Beach 1 & 2 C
1 3
Short 1 SR0 L
'(,
. Phase II Design Lecture Series.
6 veeks - f amiliarizes trainee with the design features of his plant.
Phase III Observation Training and Simulator Training Observes day to day operation of a nuclear power plant including surveillance testing and radiation protection programs. The trainee receives two to three months of training on a power plant simulator in which he observes and participates in all phases of plant operations, startup, load and power changes, normal, abnormal, and emergency conditions.
The observation training is from 1 to 3 months, the simulator q.._..____.
_1 raining is.from.2.to 3 months _The min.icum.. time for the combination is 4 months.
Phase IV Onsite testing (6 month minimum) applicants for a cold examina-tion must complete a site training program at the reactor site for which the reactor operator will receive his license.
In addition to classroom work the applicant engages in the day to day plant activities including procedure writing, construction checkout, and preoperational testing.
-Hot -Training Program - Training for applicants who will be taking their NRC
. license exam following reactor criticality.
(Usually candidates are selected
-from plant auxiliary operators who normally have l!s - 2 years of operating experience at that f acility (training conducted onsite).
The technical training for the hot training program is the same as that which is required for applicants in the cold training program. However, the hot
- training applicants participate in on-the-job training which involves manipula-tion of controls during 5 reactivity changes and at least 2 training startups of the reactor.
The two training startups of the reactor can be replaced by training at an appropriate nuclear power plant simulator.
Sore trainees participate in reactor and plant operations during the commissioning phase of the f acility up to' the point where the reactor operates at a power level of 20% - this provides the applicants with hands on experience. The hot training program generally proceeds in a manner similar to the one described for the cold licensing' program, and covers a period of 6 to 8 months (minimum of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> lectures and 3 months of observation and " hands on" manipulation of power plant controls).
License Examinations - for Reactor Operator - same for Hot and Cold training progr ams.
i hYitten examinations:
six to eight hours essay type questions covering:
a) principles of reactor operation b) features of f acility design c) general operating characteristics d) instrumentation and control i
e) safety and emergency systems f) standard and erergency operating-procedures g) radiation control and safety l'
-l
. ~
Oral examinations:
four to six hours to verify applicant's capability of:
a) reading.and interpreting the control instrumentation of the facility b) manipulating the control equipment in a safe and competent manner c) operating the f acility during normal and energency conditions d) knowing the radiological safety practices and the purpose and function of radiation monitoring equipment.
The oral examination is conducted at the plant - mostly in the control room -
where the applicant points out and explains the function and way of using plant instrumentation and controls. The applicant is tested on hypothetical p___ _ _
accident scenarios. - including _ mock _ manipulation of_c_o,ntrols.
The applicant must also tour the plant and demonstrate his knowledge of the plant, emergency systems, administrative procedures, radiation monitoring equipment, etc.
Senior Reactor Operator - Selection Criteria High School Diploma or equivalent Four years of responsible power plant experience (minimum of one year at a nuclear plant, and a maximum of two years of the remaining three years of power plant experience niay be fulfilled by academic or related technical training on a "one-for-one basis")
L I
. Examination reouirements hYitten examination comprised of questions in the same seven
. categories as the Reactor Operator examination, plus five other areas; i.e.,
h) reactor theory 1) radioactive material handling, disposal, and hazards
-j) specific operating characteristics k) fuel handling and core parameters
- 1) administrative procedures, conditions, and limitations.
Criteria for selection and training of nuclear power plant personnel are generally within the general guidelines of ANSI N18.1 - 1971 - (See Tabis 2) per Reg. Guide 1.E "In some cases, plant design features or unusual operating conditions may indicate that additional cr specialized expertise beyond qualifications presented in ANSI U18.1 - 1971 is needed.
This determination will be made on a case-by-case basis."
It is the utility's responsibility to obtain competent personnel to man the power plant.
In their safety analyses the utility is required to submit their plant's organizational plans including job requirements and descr:gtions and personnel
' qualific at ions. The utilities are also required to submit details of their training plans for all employees.
Details of the subject matter for each course, on site, as well as at other training institutions, simulators, etc., are provided to NRC.for review and approval per Standard Review Plan 13.2.
The utility's safety analysis reports: "should demonstrate that the training provided, or to be provided,
.-for each position on the plant staff will be adequate to provide assurance that all plant staff' personnel qualification requirements will be met as of the time needed; l
l
$f' 9-prior to operator license exams, prior _ to fuel loading, or prior to appointment. or reappointmnt to the position."
The decision for a particular individual to pursue a particular license is
~
his decision. influenced-by salary and status -- however it would appear. that the utility has veto power over corporate expenditures on the individual's tr aining.
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it far es."ainali""
pl...t r ure irm r..% 'cre n.n in paragrapt. 2 an.13 for nuricar M
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g Additional NRC requirement (Radiation Protection Manager -- Bachelor's Degree fiS equivalent to one year of
=ma-a.
professional experience; Ph.D = two years professional experience), at least three years of professional
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experience should be in applied radiation protection work in a nuclear facility dealing with radiological M
problems similar to those in a nuclear power station, preferably in a nuclear power station.
E""'"'
3.
What training did the operators on duty at TMI get? Describe training course including classroom work, hands-on work, types of scenarios followed, use of multiple failure training.
Where does typical operator
' om from? Describe apprentice, on-the-job, refresher, and simulator training.
Who does the training? Does f;RC monitor or review the training and how?
The control room licensed operator (R0 and SR0) on duty at the time of the accident were all experienced operators with a background of several years operating Navy nuclear power plants.
They had all spent approximately 18 months as " auxiliary operato'rs" within the Met Ed system during their training and qualifications to become Reactor Operators at Ti1I.
The training program of Met Ed was typical of many utilities and followed the requirements and guidelines of the NRC (NUREG-0094).
Subsequent to the accident the TMI-2 operators have been critical of the training they received in general.
In particular, one of the R0s noted the following:
l'~-'
~1;'~Trainin'g is based on the assumption that the design encompasses enough enargencies and anticipated casualties, thus the outcome of any event could be assumad before hand.
2.
Although some instrument errors are used durine training, basic philosophy is everything is going to work. Never failed 2-3 safety systems on a simulator. Also simulator can 't nodel a solid pressurizer.
3.
Training is in response to the types of questions NRC is going to ask during exams. Train to pass a test rather than operate a plant.
4.
NRC looks for responses to predetermined erergencies.
5.
Because of above, training philosophy is inadequate.
1 versight Hearings
-e Weaver Committee ' Serial No.
S.
B 0
art 1, May 11, 1979, beginning page 151.
PODIFORIGINAL
e 2
The specific Met Ed Training Program has been described in detail.2 This program will be reviewed in detail to evaluate its adequacy.
The Met Ed Training Staff has provided connents on the adequacy of the training program as it directly relates to the accident at TMI.
The plant procedures (2103-1.3) and the operator training specify that the plant shall not be allowed to go solid at any time except for hydrostatic tes ti ng.
The training emphasizes this prohibition, stressing the
~
possibility of exceeding the high-pressure safety limit of 2750 psig because the pump discharge head is 2900 psig.
For high pressurizer w.
_. _. _. _ _ _ _ _ _ _. _ _ _ _. _. require securing makeup and increasing level, the procedures (2103-1.3) letdown.
For low pressurizer pressure, the procedures require the opposite:
isolate letdown, increase nakeup, and, in addition, turn on the heaters.
2Ernst L. Blake Jr. of Shaw, Pittman, Pitts and Trowbridge letter to Mr. George Frarpton, NRC/TMI Special Inquiry Group, July 20,1979, docu-ments G/712-9.a-1 and G/712-9.a-2.
I
3 The training staff was asked what the operators would be expected to do, based on their training and experience, if a high pressurizer level indi-cation called for one set of actions and a low RCS pressure called for another. Members of the training staff stated that the operators would definitely have reacted to the high level to avoid going solid.
This is based on the TMI training and the B&W operating procedures.
The necessity of maintaining pressure is stressed in connection with the avoidance of departure from nucleate boiling, but this is of no concern at the lower power levels existing after the reactor trip.
The training staff was asked if the opeators were taught the significance of saturation pressure.
They stated that they were taught this, in the w_.
. basic thermdynamics. that is_ taught._They were. asked i.f. the saturated condition would cause the operators to suspect steam voids in the primary system. They stated under the conditions that existed at the beginning of the event, the training staff would not expect the operators to check for the saturation condition im:mdiately. The operators would not expect voiding with the pressurizer full.
The injection of cold auxili'ary feed-water and HPCI would cause the operators to expect a pressure reduction.
The training staff was asked if the possibility of a level rise in the pressurizer caused by steam flashing in another part of the prirrary,, system had ever been recognized and brought to the operators " attention. The answer was "never." The only training in this area is the discussion of the possibility of flashing in the hot legs if the pressurizer level is not raintained.
4 The training staff was asked if the operators were trained to verify the closure of the electromatic relief valve during events that can be expected to result in its operation.
They stated that they were trained to check if it was open, but considering the other events that were occurring during the accident, they would not be expected to check this right away.
The reans available to check this were the console demand signal, which indi-cated closed, and the discharge line temperatures.
The training staff stated that high discharge line terperatures were not very meaningful because the EMOV had been leaking prior to the incident, which resulted in terperatures that were not much lower than those existing with the valve open.
These temperatures and their status are printed out by the alarm printer but these alarms might not get printed out for 20 to 30 A
-- minutes under the-conditions-of -the-accident-because-of the -large nuitber
.~
of alarms.
The training staff was asked if the operator training included actions to be.taken if there was a pressure rise in the reactor coolant drain tank (RCDT).
They stated that the training on the " Response to High RCDT Alarm" procedure (No. 2204-301B, C5) covered this. However, the alarm and indicators for this system are behind the panels. Also,-determining the soruce of the leakage requires a process of elimination because the RCDT is' connected to other_ leakage such as the RC pump seals and..yalve packing leakoffs.
?'.
5 It was noted that the safety features actuation system was bypassed by the operators very soon after actuation, even though the coolant injection might not be be throttled back until later.
The training staff was asked 4
if the operators were trained to do this. They stated that they were trained to reset as soon as possible.
This is done to prevent injection of sodium hydroxide into the reactor.
In addition, the operators were trained to be prepared to raintain a 220-inch level in the pressurizer by throttling the HPCI valves. Also, they had to be prepared to throttle the flow to the-makeup pumps to prevent exceeding the 550-gpm flow limitas tion, as the flow would increase if the RCS pressure decreased.
The training staff was asked under what conditions the operators were trained to shut down the RCS pumps. They stated that the procedures (2203-1.4, Revision 3) and training required punp shutdown for high vibration, low anperage, or low reactor coolant flow, all conditions that existed during the event.
In the training staff t view, the conditions that existed during the March 28 event did require shutdown of the punp.
They stated that the operator is trained that failure to trip the pump under these conditions could lead to pump seal failure or loss of the inpeller.
e
~
6 A general description of typical training programs and NRC requirements follows.
Typical Operators Applicants for reactor operator and senior operator licenses hired by the utility come from (1) conventional plants throughout the utility, (2) government operated nuclear reactors, and (3) new hires off-the-street.
Both the operator and senior operator cust be high school graduates or equivalent. Many utilities employ preselection screening using tests designed to determine an individual t suitability for nuclear training.
Over one half of the operators have little, if any, nuclear experience at the time they are selected for training.
While conplete statistics have not been evaluated, a sanple size of 303 was found to have the following characteristics.
The median age was 36 years for the SRO and 33 years for the R0s in the sample.
With all over 25 yers of age, about 80 percent of the currently licensed SR0s and about 50 percent of the reactor operators have fornal education beyond high school.
Initial Trainino of Personnel The t'RC requires two types of training prograns.
The " cold" program pro-vides the necessary tiraining for personnel who will sit for NRC license
-examinations prior to initial fuel loading.
The " hot" program is for applicants who will sit for license examinations following criticality of the reactor.
~
~
The NRC required training for cold applicants usually starts 2 years before fuel loading.
Applicants who have previous nuclear experience are phased in proper times in accordance with their experience.
Applicants with no experience are phased in at proper times in accordance with their experience. Applicants with no experience co@lete the following program:
B Phase 1 - Basic courses which normally last 2 weeks are conducted at the nuclear training centers or universities.
Phase 2 - Design lecture series takes 6 weeks and familiarizes the trainees with the features of his plant.
Phase 3 - Observation training is conducted at both the sinulator and on an operating nuclear power plant.
The training requires 4 months and consists of observation of day-to-day operation at an operating plant and " hands on" training on a simulator.
Phase 4 - is onsite training and takes approximately 1 year.
The hot licensing program follows the same material outlined under the cold licensing program.
The training is conducted onsite and requires 6 months to co@lete.
the training program requires a minimum of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of lectures and 3 months of observation and " hands-on" manipu-lation of power plant controls, on a day-to-day basis.
Training includes 1 week of simulator operation which involves observing reactor transients
'and coping with accident conditions.
On the comTrcial plant, only the control operators and supervisor are licensed. The balance of plant
~~
personnel complete a training program, but are not licensed.
8 Oualification and Recualification Procram Many utilities employ a preselection screening process using tests designed to determine an individual t suitability for nuclear training. B ecause many of those selected for training have been out of school for a number of years, sone conpanies have found it advantageous to first conduct a review of basis mathematics and physics for the candidate.
Generally these reviews last 4 weeks.
Training programs, together with the training schedule prior to fuel loading, are submitted to the NRC for approval.
Usually the training program for applicants with no previous nuclear experience starts 2 years before fuel loading.
Applicants who have previous nuclear experience are phased in at the proper times in accordance with their experience.
Applicants with no previous experience are required to conplete the entire training program outlined below. The programs outlines below are minical programs.
Applicants must be highly motivated and dedicated to success-fully conplete these prograns.
Many applicants will require additional tutoring and time to becone competent operators.
At the corpletion of the training program, operators and senior operators are certified by utility management and are then examined by NRC licensing examiners.
If they successfully pass the written and oral examinations, they are issued a license to operate the plant. The written examination for the operator consists of seven categories and generally takes 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to couplete.
The senior operator written examination consists of the i
l
p.
9 sane seven categories plus an additional five.
Approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> are required to complete the examination.
The written examinations are followed by an operating test conducted by an NRC examiner.
A typical operating test takes from 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and consists of a one-on-one oral examination, which tests the applicant t knowledge of the plant operations and procedures.
The NRC approved requalification program requires that each licensed indi-vidual demonstrate his competency over a 2-year period to renew his license.
The program requirements include preplanned lectures and on-the-job taining, including control reactivity manipulations, understanding of systens, procedures, design changes, changes to facility license and the emergency procedures. The NRC licenses both the operator and supervision to operate all systens in the nuclear power plant.
Practice of Casualty Drills and Plant Evaluations The utility requalification program provides training and evaluation of the performance of abnornal and emergency procedures. This is acconplished by a training supervisor, reviewing step-by-step, the procedure with the licensed operator or supervisor at the operating control board.
Casual ty training and evaluation on a reactor simulator is an integral part,,of many of the plant requalifications programs.
On the simulator, the student i
observes the symptons and performs the immediate actions required tu cope with the accident condition.
t i
i l
10 i
The NRC requires the licensed personnel to discass step-by-step the emergency procedures on the control boards. Many of the utility training programs include training on simulators when the licensee analyzes and copes with the accident condition.
Continued Review of Personnel Performance and Removal From the Program of Those Wno Do Not Meet Standards The utility requalification program provides an ongoing review of personnel performance through preplanned lectures, control manipulations, review of abnormal and emergency procedures and annual written and oral examina-tions. A grade of less than 80 percent on any category of the written examination requires attendance of the lecture on that category sterial.
A grade of less than 70 percent overall on the annual written examination requires mandatory participation in an accelerated requalification program.
The individual is removed from his licensed duties until he has successfully conpleted the accelerated program and scored not less than 70 percent on a reexamination.
Further, if a licensee has not been actively performing the functions of an operator or senior operator for a period of 4 months, he nust demonstrate to the NRC his understanding of facility operation before he is permitted to resume his licensed duties.
1.
Schedule.
The requalification program shall be conducted for a continuous period not to exceed 2 years, and upon conclusion shall be prorptly followed, pursuant to a continuous schedule, by successive requalification programs.
7fnp -
11
~2.
Lectures.
The requalification program shall include preplanned lectures on a regular and continuing basis throughout the license period in those areas where annual operator and senior operator written examina-tions indicate that.enphasis in scope and depth of coverage is needed in the following subjects:
a.
Theory and principles of operation.
b.
General and specific plant operating characteristics.
c.
Plant instrumntation and control systems.
d.
Plant protection systems.
e.
Engineered safety systems.
f.
Norm 1, abnorm 1, and emergency operating procedures.
g.
Radiation control and safety.
h.
Technical specifications.
i.
Applicable portions of Title 10, Chapter I Code of Federal Regulations.
Other training techniques including films, videotapes and other effec-tive training aids my also be used.
Individual study on.the part of each operator shall be encouraged.
However, a requalification program based solely upon the use of films, videotapes and/or individual study is not an acceptable substitute for a lecture series.
6 12 3.
On-the-job training.
The requalification program shall include on-the-job training so that:
a.
Each licensed operator of a production or utilization facility manipulates the plant controls and each licensed senior operator either manipulates the controls or directs the activities of in-dividuals during plant control manipulations during the term of their licenses.
For reactor operators and senior operators, these manipulations shall consist of at least 10 reactivity control manipul'ations in any conbi-ation of reactor startups, reactor shutdowns or other control manipulations which demonstrate skill and/or familiarity with reactivity control systens.
b.
Each licensed operator and senior operator has demonstrated satisfactory understanaing of the operation of all apparatus and rechanisms and knows the operating procedures in each area for which he is licensed.
c.
Each licensed operator and senior operator is cognizant of facility design changes, procedure changes, and facility license changes.
d.
Each licensed operator and senior operator reviews the contents of all abnornal and emergency procedures on a regularly sck~eduled basis.
I
13 e.
A simulator may be used in meeting the requirements of paragraphs 3a and 3b if the simlator reproduces the general operating charac-teristics of the facility involved, and the arrangement of the instrumentation and controls of the simlator is similar to that of the facility involved.
4.
Evaluation. The requalification program shall include:
a.
Annuai written examinations which determine areas in which retraining is needed to upgrade licensed oeprator and senior operator knowledge.
b.
Written examinations which determine licensed operators " and senior operators " knowledge of subjects covered in the requali.fication pro-gram and' provide a-trasis for evaluating their knowledge of abnormal and emergency procedures.
c.
Systematic observation and evaluation of the performance and competency of licensed operators and senior operators by super-visors and/or training staff members including evaluation of actions taken or to be taken during actual or simlated abnormal and emer-gency conditions.
- d. Simulation of emergency or abnormal conditions that my be accomplished by using the control panel of the facility in olved or by using a simulator.
Where the control panel of the facility is used for simulation, the actions taken or to be taken for the
x.
14 energency or abnornal condition shall be discussed; actual manipula-tion of the plant controls is not required.
If a simulator is used in meeting the requirements of paragraph 4c, the simulator shall accurately reproduce the operating characteristics of the facility involved and the arrangement of the inst'rumentation and controls of the sinulator shall closely parallel that of the facility involved.
e.
Provisions for each licensed operator and senior operator to parti-cipate in an accelerated requalification program where performance evaluations conducted pursuant to paragraphs 4a through 4d clearly indicate the need.
S
\\
f
m.
4.
Control Room and Instrumentation The principal design criteria for TMI-2 were developed in consideration w e-e of the AEC's General Design Criteria that 11s55E6 PROPOSED in 1967.
(Approved in
).
The following is a listing of criteria applicable to the control room and the plant instrumentation -
as described in the PSAR (through Amendment 12,10/71).
CUTIN 0N 11 - CONTROL ROOM (Category 3)
The f ac111:7 shan be previded wi:h a c:=:rol ree f ce which actices :o mais-tais safe opera:1cual s:atus of :he plan: can be con ro n ed.
Adequate radia-tics protec:ica shan be provided even under acciden: condi:10:s, to equip ent in the c ::rol co: or c:her areas as necessarf :o shu: den and maintain saf e ec :rol of the facility wi:hou: radia:1ca exposures of personnel 1: excess of 10 CyR 20 11=its.
I: shan be possible :o shut :he reae::: dcun and =ain:at: 1:
in a safe conditics if access :c the ec ::o1 ce= is los: due :o fire or c:her cause.
P00R ORIGINAL
_CRITI?l_Q 12 - INS Rt; MIN"ATION AND CONTROL SYStIMS (Categcry 3)
......w
- 10:::.:===:::10 and cen::cis-shan-be provided as = quired to =cci:or and
=ais:ais variables withis prescribed operating ranges.
CRITIRION 13 - FISSICN ?ROCESS MONITORS AND CONTROLS (Ca:egerf 3)
Means shan be provided for monitoring and aintai ing control over the fission p:ccess th: ughou: core life and for all ec di:ic=s that can reasonably be anticipated :o cause variaticas 1: reactivity of the core, such as indication of positien of cc=::c1 rods and conces::a:1c of soluble reactivity con:rol poiscas.
l Cn'"I n 0N 16 - MON!!O? LNG ?IAC~0R COCLANT ??2SSi;?2 3CCCARY l
(Category 3)
Means shall be p cvided for =cci:: ing~:he reactor coolan: pressure boundary 1
-to detec leakage.
1 CnTIn05 17 - MON!!0 RING RADICACTIVIIT FILIASI (Ca:egorf 3)
Means shan be p cvided for :: i:oring :he contain=en: at=csphere, :he facili:y efflues discharge paths, and :he facili:y envircus for radioactivi:y that c=uld be released f c nor=al opera:ic s, f c: anticipated ::ansients, and f c= accident condi:1cas.
m l
C?l!I?l0N 18 - MONITORING TUIL AND ~4ASTI STORAGE (Ca:egory B)
M:=1:::1=g and alar = instrumenta:ics shall be provided for fuel and waste s: Or-ago and handling areas for ccedi:ic=s : hat =igh: ::::ribute to loss of ec=:inui:yl ir decay hes: re=cval and :o radia:1c= exposures.
I CRI"'IRION 22. SI? ARA'"!ON OF PROTI:" ION AND CCICOL INS 3U!Ci".ATION '
SYS"'J.S (Category 3)
Protection syste=s shall be separated frc= :: trol instr =entation systems to the extent that failure or re=cral frc= serrice of any ecstrol 1:str=enta-l tien syste: component or channel, er of those ec==c to ec= trol instr =enta-ties and ;;ctectics circuitry, leaves intact a syste= satisfying all require-
=:sts for the protectics channels.
Beyond the proposed General Design Criteria, the PSAR does not reference any specific requirements for control room design, data display equipment, or instrumentation.
The closest that the PSAR comes to outlining instrumentati'on require-
' ~ ~
ments is in paragraph 7.l.'l.5 Pr'i$ciples of Design (Instrumentation and Control-Protection systems) - where it states that "The PROTECTION SYSTEMS are designed to meet the requirements of the proposed " Criteria for Nuclear Power Plant Protection Systems," (IEEE 279 dated August 30,19681"An annotated copy of IEEE 279-1968 indicating relevant portions is attached Nevertheless, in Section 7.4 of the PSAR, Met Ed outlines their control room design philosophy - see attached.
Apparently, the control room described e
in Section 7.4 of the PSAR melts the AEC's proposed (1967) General Design Criteria.
P00R~0RIGINAL l
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In 1969, the flRC was in the process of-proposing guidelines with respect to minimum requirements for control room design considerationr..
The enclosed memorandum. S. Levine dated 6/11/69, exemplifies the status of AEC control room design requirements prevalent at the time TMI was licensed.
Group 1 will be tracking down the evolutionary process associated with control room design, instrumentation, and monitoring equipment requirements.
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