ML19257C429
| ML19257C429 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/17/1980 |
| From: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | CALIFORNIA, STATE OF |
| Shared Package | |
| ML19257C422 | List: |
| References | |
| NUDOCS 8001290106 | |
| Download: ML19257C429 (27) | |
Text
. _.. -..
. ~
January 17, 1980 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312
)
(Rancho Seco Nuclear Generating Station) )
LICENSEE'S ANSWERS (SET NO. 2) TO CALIFORNIA ENERGY COMMISSION'S SECOND SET OF INTERROGATORIES TO THE SACRAMENTO MUNICIPAL UTILITY DISTRICT. _ _,
4.
INTERROGATORY:
Specify each type of feedwater_ transient which may result in a reactor trip.
For each transient, are operator procedures the same?
Identify those procedures.
(Are Emergency Procedure D-14 " Loss of Steam Generator Feed" and Procedure A-51
" Auxiliary Feedwater System" the only procedures available for operators in the event of a feedwater transient?
(Reference SMUD Response to CEC Interrogatory #11)).
If different procedures apply for different feedwater transients, how does an operator recognize the type of event involved in order to choose the proper response procedure?
Explain.
ANSWER:
Two basic types of feedwater transients may result in reactor trip.
These transients are loss of feedwater flow or excessive feedwater flow.
Emergency Procedure D-14 " Loss of Steam Generator Feid" is written for the loss of feedwater transient, the more significant hazard of the two feadwater transients.
Annunciat'or procedures H2PSB for annunciator panel windows 13 and 14, indicating steam gerarator levels high or low, direct the operator to take action as necessary to restore levels to normal plant conditions. - The most likely cause of an overfill condition is loss of normal automatic level control.
In this case the operator's training provides him wit' the capability to shift from 1836l238 8001290 i
t
automatic to hand control and reduce steam generator levels to the appropriate level for the existing power.
The operator is able to recognize whether it is a loss of feed transient or an overfeed transieric by monitoring of the steam generator level indication.
9.
INTERROGATORY:
Based upon the occurrence of a worst-case feed-water transient (loss of of f-site power with single failures in available safety systems), identify the most severe over-cooling events which could occur.
Specify those procedures which ensure adequate core cooling and all related operator actions.
ANSWER:
A loss of off-site power is not expe.gted to initiate a safety features trip at a reactor coolant system pressure of less than 4600 pounds.
Upon loss of all reactor coolant pumps, the auxiliary feedwater system would initiate and control steam gener-ator level at 50 percent, putting the reactor coolant system in the natural circulation mode.
An over-cooling event would only occur if some failure in the auxiliary feedwater system caused the level control valve to malfunction and raise level in tne steam generator excessively.
In the extreme instance this excessive level may generate sufficient driving head to cause natural circulation at such a rate to provide a cooldown in excess of 100-degree F. per hour.
To avoid complete filling of the steam generator, the operator would take manual control of the level control valves and reduce feedwater flow.
Those procedures which deal with providing adequate core cooling are Procedures D-14
" bass of S team Generator Feed", Procedure A-51 " Auxiliary Feedwater System", and Procedure B-4 " Plant Shutdown and Cooldown".
I8ffe 239.
- 19. INTERROGATORY:
Ideritify each employment position _or___clase of personnel at Rancho Seco.
Provide a description of the A-.tagement and reporting of these employees, as well as descriptions of the job duties Include the following:
(a)
Numbers of employees in each position.
(b)
Responsibilities and qualifications of each position.
(c)
Identify specific individuals who assist or maintain the operation of the plant equipment.
(d)
Identify all plant personnel (other than licensed operators) who have lef t employment at Rancho Seco in the last two years, including the reason they left.
ANSWER:
The overall on-site responsibility for. operations, maintenance and administration of the Rancho Seco Nuclear Generating Station rests with the Manager of Nuclear Operations.
Reporting directly to the Manager of Nuclear Operations are the Engineering Supervisor and the Plant Superintendent.
There are three separate divisions under the Plant Superintendent's responsibility.
The Operations Supervisor Leads the division which includes all personnel directly responsibla for operating the unit. This includes six shif t supervisors, fou senior control room operators, four control room operators, seven auxiliary operators, ten equipment attendants and seven power plant helpers.
The Maintenance Superintendent is responsible for the administration of the Maintenance Division.
Reporting directly to him is one Mechanical Supervisor and one Instrument Control and Electrical Supervisor.
The Mechanical Supervisor exercises management control over six meet.raical foremen and twenty-three mechanics and mechanic apprentices.
The I&C and Electrical Super-visor exercises control over three instrument control foremen, one electrical technician foreman and four electrical foremen.
At the 18fk6 240
-~.--.-... ~
working level that organization also includes sixteen instrument control technicians or apprentice technicians, eleven electricians or electrician apprentices, and five electrical technicians or technician apprentices.
Also reporting to the Maintenance Superintendent are three electrical engineers and two mechanical engineers who are used for support of the technicians and the mechanics in carrying out maintenance.
The third division under the Plant Superintendent is the Health Physics and Chemistry Control Division, managed by one Radiation and Chemistry Supervisor.
Reporting to this supervisor are three Senior Chemistry and Radiation Assistants, one Health Physicist and one Nuclear Chemist.
The Senior Chem-Radiation Assistants exercise supervision over nine Chem-Rad Assistants who carry out the daily health physics and chemistry duties in support of both maintenance and operation.
Under the Engineering Supervisor, who reports directly to the Manager of Nuclear Operations, there are three Nuclear Engi-neers, two Electrical Engineers, and three Mechanical Engineers.
In addition, the Engineering Supervisor exercises management con-trol of three in.1pectors through the Quality Control Coordinator.
There are also three engineering aides and/or technicians who assist the engineers in carrying out their functions under the Engineer-ing Superviso r, i859 241.
~ _... -
Under the management structure described above the following numbers of employees are in each of the positions described:
Mechanical Engineers 5
Electrical Engineers 5
Nuclear Engineers 3
Quality Control Coordinator 1
Inspectors 3
Engineering Aides / Technicians 3
Shif t Supervisors 6
Senior Control Room Operators 4
Control Room Operators 4
Auxiliary Operators 7
Equipment Attendants 10 Power Plant Helpers 7
Instrument and Control Foremen 3
Instrument and Control Technicians and Apprentices 16 Electrical Foreman 4
Electricians and Apprentices 11 Electrical Technician Foreman 1
Electrical Technicians and Apprentices 5
Mechanical Foremen 6
Mechanics and Apprentices 23 Senior Chemical and Radiation Assistants 3
Chemical and Radiation Assistants 9
Health Physicist 1
Nuclear Chemist 1 2
/
The desirable qualifications and types of duties for which these individuals are responsible are described in the employment announcements which are provided pursuant to Licensee's response to " California Energy Commission's Second Request _ for Production of Documents to the Sacramento Municipal Utility District."
The individuals holding the positions described assist or maintain the operation of plant equipment.
The SMUD Employee Status Notification forms for all per-sonnel who have left employment of the Rancho Seco Nuclear Opera-tions Department in 1978 and 1979 will be provided pursuant to the Licensee's response to " California Energy Commission's Second Request for Production of Documents to the Sacramento Municipal Utility District."
The Employee Status Notification forms iden-tified above provide, in some cases, a brief description of the individual's reason for leaving.
- 21. INTERROGATORY:
What personnel are assigned t'o operational crews that have diagnostic capability to identify multiple failure accident events?
What is the basis for demonstrating diagnostic skills?
ANSWER:
Each operating crew has assigned a shif_t supervisor, a senior control room operator, and a control room operator.
All three of these individuals have been licensed by the Nuclear Regu-latory Commission as either senior reactor operators or reactor operators.
In the course of their training, these licensed opera-tors have received simulator training in a basic three week program and annual simulator training to retain their proficiency.
The simulator provides the capability of instituting failures in i 8didL 243
.~.._..
various systems of the nuclear plant.
These failures are insti-tuted singularly or as multiple f ailures to test-the operators' ability to accurately diagnose and take action.to mitigate the consequences of these failures.
- 22. INTERROGATORY:
What are the responsibilities _o_f. supervisory and higher plant management for decision-making under accident conditions?
What training has been provided for these personnel?
Give details.
ANSWER:
The Operations Supervisor, the Plant _S_uperintendent, and the Manager of Nuclear Operations form the direct chain of command above the shif t supervisor level.
Inherent in the responsibility of this chain of command for the safe operation of the nuclear plant lies the authority to provide to the shift supervisor what-ever guidance and direction is needed to cope with plant conditions under an accident scenario.
Specifically, the Operations Supervisor's responsibility deals directly with the operation of the reactor plant in placing it in a safe shutdown condition.
The Plant Superintendent is responsible for the overall direction of operations, maintenance and health physics support to mitigate the consequences of the accident.
The Manager of Nuclear Operations has responsibility to keep higher level management within the District informed of the status of plant conditions.
The Manager of Nuclear Operations would also be the focal point for determining whatever additional support is needed and assuring that management provides the additional assistance.
Specific training given to these individuals to equip 1839L 244 them with the general knowledge needed to cope with accident con-ditions has included qualification as licensed senior reactor operators and extensive training in the actual operation and limitations of the reactor plant.
In addition, the Plant Superin-tendent and the Manager of Nuclear Operations have been active in industrial organizations dealing with plant activities from f acil-ities across the country.
This participation provides these two individuals with knowledge of activities, experience with and improvements in plant management at other similar units.
Most recently all three of the f acility management per-sonnel have begun participation in a command and control training program being presented by an outside contractor.
The purpose of this program, which is currently in the first of three phases, is to provide additional training in the command and control aspects of mitigating the circumstances of various accident scenarios.
- 23. INTERROGATORY:
What simulator training has been provided for
" multiple failure" accidents like TMI (or other similar tran-sients) ?
Give details of training and capabilities of simulator to provide experience (s) of this t;rpe.
ANSWER:
Between April 20, 1979, and June 22, 1979, all licensed senior reactor operators and reactor operators were provided simu-lator training at the Babcock and Wilcox simulator in Lynchburg, Virginia.
The purpose of this training was to thoroughly acquaint them with the indications expected during an accident similar to the multiple failure accident that occurred at Three Mile Island.
This training included classroom discussions of the basic underlying causes of the accident, a description of how the plant's parameters changed during the course of the accident, and, finally, 1831.245.
how the accident was terminated.
In the simulator the accident initially was demonstrated under conditions in which the operators observed the course of the various plant parameters.
Then the accident was again simulated allowing the, operators to exercise control to mitigate and stop the accident before it reached conditions in which core damage occurred.
Subsequently additional simulator courses were given as part of the annual retraining.
During these courses multiple failure accidents have been imposed and the operator has been given the opportunity to exercise his diagnostic skills and training in mitigating the consequences of those multiple failure accidents.
The simulator has the capability of introducing over sixty individual casualties in the various reactor plant systems.
The specific systems which are covered in these casualties include the coolant makeup system, the reactor and its instrumentation, the reactor coolant system, the steam and turbine system, the condensate and feedwater system and various auxiliary systems.
The individual casualties can be combined to create multiple failure scenarios and to present the operator with a complex problem in which to practice his training and diagnostic skills.
The programming available at the simulator also permits the instructor to f ail equipment sequentially and thereby allow full exercise of the operator's training.
This tests the operator's skill and abilities to make initial diagnosis of a f ailure, begin corrective action, discover another failure, and then exercise some alternate corrective method to keep the unit in a safe condition.
18fA 246
- ~..
- 24. INTERROGATORY:
What instruments in the control _ room hav_e_ scale readings which are adequate for normal operations, but would be of f-scale in the event of a TMI-type accident?
ANSWER:
Those instruments available in the control _ room which may be driven off scale for an accident of the magnitude of the accident at Three Mile Island are as follows:
a.
The pressurizer level may go off scale high or low, b.
The reactor coolant system outlet temperature may go off scale high or low.
c.
The reactor containment building radiation monitors would go off scale.
- 27. INTERROGATORY:
What efforts are made to obtain_ s_t.udent_ev.aluation of training programs?
Is there a feedback mechanism for incorpor-ating student comments on training into future training programs?
Has feedback become incorporated into changes in operational pro-cedures?
Control room hardware?
Describe mechanisms for feedback and give specific examples of how it has affected training pro-grams, operational procedures, and/or control room design.
ANSWER:
In the course of conducting operator.. training programs the training staff solicits and welcomes comments from the students that will provide improvements in the quality, technique or clarity of the training.
Similarly, in the course of carrying out the simulator training program student comments to the instructors are welcomed and these comments are considered in upgrading of the training program.
The feedback mechanism for incorporating comments into future training programs is direct student communication, since the instructors conducting the program have a very major responsi-bility in control of the course content as well as its presentation.
The simulator program in particular enjoys a very strong positive feedback in that simulator instructors on occasion come to the site i edifL 247 for discussions with site training personnel and operating person-nel to solicit topics for discussion in the lecture portion of the simulator program.
In addition, the types of transients which operators are interested in having programmed into the simulator are acted on positively.
One of the major sources of changes to operating procedures actually come from operating personnel or students during their control room training.
Any individual operator who finds a better technique for carrying out a specific operation or finds an error in a currently approved procedure has the opportun-ity to make his comments known informally to the shif t supervisor or the operations supervisor.
These comments are considered, discussed with the operator and, if found to be a more efficient and precise technique for carrying out a particular procedure, will be incorporated as an approved change into a standing operating procedure.
Engineering personnel who are contemplating changes in control room hardware solicit the opinion of operators for place-ment of that hardware in existing panels.
The location of switches or meters is typically determined only after some discussion with operating personnel so that placement will allow them to operate the equipment with as much ease as possible.
Specific examples of feedback effects include the following:
18SS.248
.~
a.
Because of the interest of individuals who, during the simulator training reviewed television tapes of casualties, including loss of feedwater, five tapes were purchased from B&W so that throughout the year operators would have an opportunity at Rancho Seco to review these tapes.
b.
As a result of direct input from students requesting additional specific instruction in the health physics area, the health physics program has evolved from a three and one-half hour program to an eight hour program.
c.
Students have made comments regarding the difficulty of learning how to read pocket dosimeters.
As a result of this input a two and one-half foot long, scaled-up model of a dosimeter was produced and has significantly enhtnced the ability of the training staff to teach students how to properly read radiation dosimeters.
d.
In the placement of the recently installed auxiliary f eedwater flow detectors within the control room, operator feedback was requested and acted upon to place these meters in the most advantageous position available.
e.
As a result of input from operating personnel, a complete rewrite of the auxiliary boiler procedure was undertaken to improve its organization and content.
l bh;h-249.
~ -.....-
- 28. INTERROGATORY:
On a sketch of the Rancho Seco_ control. room, show the locations of instruments and related controls which must be used by operators to maintain control of the facility in the event of feedwater and loss-of-turbine transients.
(a)
Identify locations of steam generator level indicators, feedwater flow indicators, reactor coolant system temperature and pressure indicators, pressurizer indicators, Engineered Safety Feature of the indicators, and corresponding controls and valve position indicators related to all elements of the feedwater system, reactor coolant system, pressurizer elements, Engineered Safety Feature controls, especially high and low pressure injtetion system controls reactor control rod positions, reactivity in core, and Integrated Control System.
(b)
Identify locations of reactor coolant drain tank status indicators and instruments showing responses of radiation monitoring equipment.
(c)
Provide physical descriptions of instruments and controls giving type of dimensions, methods for indicating operational limits on gauges, heights of instruments from floor level, and parallax problems (if any).
(d)
Discuss any line of sight problems which may exist when operators are working at control panels and trying to read associated instrumentation.
(e)
Discuss mechanism for " tagging out" inoperative equipment and instruments and discuss potential for
" tags" blocking operator views of critical data.
(f)
Provide photographs or drawings showing important sections of control panels and providing legible indications of functional system. layouts of controls, status indicators of functional system layouts of controls, status indicator signals, and instrumentation associated with the above listed systems and others that are critical to plant operation during transient recovery and/or shutdown operations.
(g)
Identify location of computer input / output equipment and associated data displays using CRTS.
Describe displays for trend recording information handling as provided by computer output.
ANSWER:
Figure 7.4-1 of the Rancho Seco Final Safety Analysis Report ("FSAR") provides a general layout of the Rancho Seco control room.
Engineering drawings E402, Sheet 1, E402, Sheet 2, E406, E410, N25.01-8, N25.01-9, N25.01-10, N25.01-ll, N25.01-12, N25.01-13, N25.01-14, N25.01-17, N25.01-18, N25.01-34 and 18% 250
_13_
N25.01-36 provide the arrangement of instruments and controls on each individual panel.
These drawings are available for inspection.
(a)
The location of all indicators and controls,.with the exception of the indication for reactivity in the core, identified in Interrogatory 28(a) are shown on the engineering drawings described above.
There is no core reactivity meter.
(b)
Reactor coolant drain tank status indicators are not in the control room.
The location of radiation monitoring equipment indications-is shown on Figure 7.4-1 of the FSAR.
(c)
The physical description of instruments and controls giving dimensions, methods for indication, heights from floor, and parallax problems (if any) may be determined by the California Energy Commission under the terms of Licensee's response to
" California Energy Commission Request for Inspection of Rancho Seco Facility", which is being filed simultaneously herewith.
(d)
Licensee is not aware of any line of sight problems which exist when operators are working at control panels and trying to read associated instrumentation.
(e)
Administrative Procedure AP-4 provides the procedure used at Rancho Seco for " tagging out" inoperative equipment and instruments.
Since magnetic markers are used instead of large tags, there is no potential for tags blocking operator views of critical data.
(f)
The information requested is contain'ed in the engi-neering drawings listed in response to Interrogatory 28(a)above, and/or can be ascertained under the terms of Licensee's response to the " California Energy Commission Request for Inspection of 18 8 25i
Rancho Seco Facility".
(g)
Figure 7.4.1 of the FSAR shows the location of computer input and output equipment.
CRTs are not used at Rancho Seco.
Trend recording, as described in Licensee's answer to Interrogatory 25 (in Set No. 1), is displayed on the computer console.
- 29. INTERROGATORY:
NUREG-0585 has emphasized the need for better licensee verification of correct performance of operating activities. (pp.2-7).
(a)
Please describe SMUD's current procedures for verification of correct performance of operating activities.
(b)
What changes, if any, are contemplated in these procedures?
(c)
Describe all accuments which contain the results ofi verification of correct performance of operating activities over the last three years.
ANSWER:
(a)
The current procedures for verification of correct operating activities encompass two general programs.
The first program consists of shif t operating logs.
These include the shif t supervisor's log, control room operator log, and various logs under the responsibility of the auxiliary operator, equipment attendant, or power plant helpers.
The shif t supervisor and control room operator logs essentially record the major overall plant activities of each shif t, with a review provided by the next higher level individual in the operating organization.
For example, the control room operator's log would be reviewed by the shif t supervisor and the senior control room operator, and the shif t supervisor's log would be reviewed by the operating division supervisor.
18jF2 252.
In addition, the administrative procedure for conduct of the watch requires that in the course of his watch, both the shift supervisor and the senior control room operator review those copies of shift supervisor's logs and control room operator.'s logs which have been generated since their last watch routine.
The senior control room operator will also review the logs completed by other members of the operating crew.
The primary purpose of these logs is to insure that the other operators within the operating crew do make the rounds and monitor the status of equipment and tank levels in their areas of responsibility.
The second portion of the verification of correct operating performance and activities'is covered by a very detailed surveillance program.
This program encompasses a large number of procedures which require specific activities to be carried out by members of the operating or maintenance staff at specific
' intervals.
In some cases, these intervals may be as short as once every one hour, and as long a; once each eighteen month refueling period.
The actual operating crew has procedures which require that data on the status of the plant must be recorded on a shif t basis, a daily basis and a waekly basis.
These procedures cover details as prescribed in the operating license.
The shift supervisor is required to sign off at the completion of these procedures, verifying that the information contained in the appropriate sections of the procedure are in fact accurate and have been reviewed by a senior licensed operator.
Procedures dealing with the testing of e%djdits
~
operability, upon completion, are forwarded to the appropriate engineer or technician for review to insure that the acceptance criteria have been adequately addressed and met in the course of conducting the test.
(b)
The changes to these procedures are not contemplated to be extensive in the future.
It is contemplated that upon full qualification of the proposed on-shif t technical advisors these individuals will also participate in monitoring and verification of on-shif t operating activities.
It is currently anticipated that the individuals filling this role will be hired in the near f uture and their training will commence so as to have them fully qualified by January 1, 1981.
(c)
The documents which contain the results of verifi-cation of correct performance of operating activities over the last three years include completed copies of the surveillance test procedures, completed copies of the various operating logs, com-pleted copies of the plant computer log, and copies of the surveillance test program schedule.
The surveillance procedures identify the specific information which is required to be checked and recorded.
The operating logs identify equipment to be moni-tored and the parameter to be recorded.
The written operating shif t supervisor and control room operating logs contain the narratives by the appropriate member of the crew describing the activities on his particular watch.
The surveillance program schedule documents the specific surveillance to be conducted, the day on which it is to be conducted, and whetner or not it was conducted'on that prescribed day.
181Bgs 254 i
- 30. INTERROGATORY:
What in-plant drills, if any, does.SMUD conduct to train operating personnel to deal with unusual situations?
(Reference NUREG-0585 at 2-7)
ANSWER:
SMUD conducts in-plant drills covering _ radiation casualties, fire, personnel injury and evacuation.
Specific drills related to operating casualties affecting the actual operation of the reactor plant and auxiliary systems are carried out at the annual one-week simulator program for each licensed operator.
- 31. INTERROGATORY:-
NUREG-0585 states as follows: "IO]ne of the most important lessons learned from the Three Mile Island accident is there there is a neeed to rapidly improve the human factors engineering in the design and layout of existing and future control rooms..." (pp.2-8).
What steps has SMUD taken, either as a response to TMI or independently, to assess and improve human factors engineering in the Rancho Seco control room?
ANSWER:
The District has requested proposals from contractors to perf orm human f actors engineering in the control room and the anticipated operations control center to upgrade the design and layout of the current control room and the expected operations control center.
- 32. INTERROGATORY:
in NUREG-0585 it is stated with respect to operator training:
" At present, a basic fundamentals course of approximately twelve weeks is required as part of the operator training program.
A prerequisite to satisfactory performance of nuclear power operation is the fundamental understanding of nuclear technology.
The Task Force believes twelve weeks to be insufficient time to provide a broad and comprehensive level of understanding in the fundamentals of nuclear technology.
It is recommended that the NRC, perhaps in consultation with INPO, examine the content of the casic fundamentals course and establish definitive instructional requirements for the cours."
(pp. A-7 to A-8).
4 i Bjlfr 255.
(a)
Do SMUD operators receive the twelve-week course described above?
(b)
Does SMUD agree that additional training is necessary to provide "a broad and comprehensive level of understanding in the fundamentals of nuclear technology?"
If not, please explain.- If so, what steps has SMUD taken and/or is SMUD taking to provide such additional training?
ANSWER 3 2 (a) :
SMUD operator licensing candidates receive a fundamentals course of approximately sixteen weeks in length.
This course includes mathematics, nuclear reactor physics, nuclear technology in general, health physics, chemistry and related technologies.
ANSWER 32(b) :
The answer to 32(a) sufficiently. describes SMUD's approa'ch to the nuclear technology fundamentals course and its duration.
- 33. INTERROGATORY:
Describe the training and tests given to and procedures available to non-licensed Rancho Seco personnel (i.e.,
managers, engineers, auxiliary operators, maintenance personnel, technicians) relating to the operation of the facility.
ANSWER:
Non-licensed operating parsonnel are trained in basic power plant fundamental training programs.
This program includes approximately three to four weeks of orientation aimed at training in piping and instrument diagram reading, plant system understand-ing, safety and first aid, emergency plan, and radiation protection This course normally follows an on-shif t assignment of one to two months to provide an adjustment period to shift work for the new non-licensed operators.
Non-licensed operating personnel are also issued a power plant fundamentals course which is to be completed during their on-shif t working periods or at home.
This course consists of a r. umber of books covering various aspects of power 18152L 256 operation.
At the end of each book a quiz is given and graded to measure the non-licensed operators' level of understanding and-retention of the material presented.
During the three to four week orientation course quizzes are usually administered to.
provide a measure of the trainee's attention and to assist 'the instructor as a feedback mechanism for any subsequent training sessions.
Supervisory, technical and maintenance personnel receive their training in on-the-job programs related directly to their individual tasks.
Typically the training responsibility lies specifically with the individual's supervisor.
The training staff provides assistance to the individual's supervisor when requested to do so.
All personnel -- operating, maintenance, management and technical -- are required at the outset of their employment at Rancho Seco to undertake training in the emergency plan and. plant safety, as well as in health physics and radiation protection.
The plant operating procedures, surveillance procedures, maintenance procedures and emergency plan are all available to non-licensed Rancho Seco personnel at various locations around the site.
Copies are located in the auxiliary building, in the administration building library and in various maintenance areas.
- 34. INTERROGATORY:
In response to CEC Interrogatory.15, SMUD provided test and training results for three groups of licensed personnel.
With reference to those test results:
(a)
For each phase identified on Tables 15-1, 15-2, 15-2, and 15-4, what result represented a satisfactory score?
Pl' ease explain.
18$FE 257.
(b)
What criteria were used to determine which licensed personnel would receive particular training?
(c)
On Table 15-4, it appears that Operators 1 and 25 consistently received low results.
Did these operators receive any special training in view of these results?
Please explain.
ANSWER 34(a) :
Those training phases identified _i.n_ Tables _15-1, 15-2 and 15-3 represented test scores which had a 70 percent passing score.
However, f ailure to achieve a 70 percent did not mean that the individual would be removed from the training pro-gram.
An individual's success or f ailure in the overall licensing program rests with his or her ability to achieve a satisfactory level of knowledge to pass the final written and oril examination prior.to being recommended for licensing to the Nuclear Regulatory Commission.
The scores identified in Table 15-4 as requalification training scores have as a minimum satisfactory grade 70 percent.
Failure to achieve the 70 percent mark would require the individual to undergo an accelerated retraining program during which time he would not be able to conduct licensed operator activities.
Throughout the training course instructors are available for, additional assistance to trainees who are having difficulties.
In addition, an independent training consultant has been retained to evaluate trainees and to identify weak areas that require addi-tional attention on the part of the training staff in the course of carrying out the overall training program.
ANSWER 34(b) :
All licensed personnel participate in the same basic training programs in accordance with established operating training programs at Rancho Seco.
ANSWER 34(c):
Operators identified as 1 and 25 on TaLle 15-4 did 1812 258 receive special training.
This training consisted either of one-on-one instructor to traineee tutoring or assignment of a shift supervisor to the trainee to assist in upgrading skills.
- 35. INTERROGATORY:
If SMUD denies in whole or in part any of the requested admissions filed by CEC to SMUD, dated December 21, 1979, please provide the basis for each such denial.
ANSWER:
The following answers provide the basi.s. for some of the denials asserted in " Licensee's Answers to California Energy Com-mission Requests for Admissions to Sacramento ' Municipal Utility District".
The second number identifying each answer below corresponds to the number of the admission requests denied.
ANSWER 3 5-28:
The f ailure of operators to recognize that coolant was being discharged to the reactor coolant drain tank was not a significant contributing f actor to the accident at TMI-2.
The significant factor that operators should have recognized was decreased reactor coolant system pressure which would indicate a loss of inventory.
The location of the loss is of secondary importance.
A ISWER 35-29:
The f ailure of the operators to_ achieve core cooling by using the decay heat removal system did not exacerbate the TMI-2 accident.
The decay heat removal system is a low pressure system which should not have been used until temperatures and pressures dropped to values within the design of the system.
ANSWER 35-32:
The power operated relief valve at TMI-2 failed in the open position.
A second valve would not have aided operators 18f5259
_22_
in responding to the accident, but indeed could have caused a second path for leakage from the primary system.
ANSWER 35-35 :
The feedwater transient would not_ be exaected to activate the high pressure injection system; however, if high pressure injection were initiated on low pressure, procedures instituted since TMI-2 direct operators to stop the main coolant pumps and therefore rely upon natural circulation.
There is no twenty-minute criteria in these procedures.
ANSWER 35-54 :
Manually raising inventory on the feedwater side of the steam generator is not normally required in the event of a feedwater transient.
The only procedures now in effect at Rancho Seco that direct operators to manually raise inventory on the feedwater side of the steam generator are in preparation for initiating natural circulation from a condition of forced circulation with the reactor tripped.
ANSWER 3 5-57:
Procedures now in effect at Rancho Seco direct operators to continue operation of the high pressure injection system upon its automatic initiation until 50 degrees F.
subcool-ing is achieved.
It is not expected that a feedwater transient will initiate automatic activation of high pressure injection.
Fu r the r, it is noted that the requested admission erroneously inplies operation of high pressure injection until 50 degrees F.
superheat is attained.
i83P7 260 ANSWER 35-64:
Licensee is not aware of any control _ room _
indicatior_s not located sufficiently close to the control location to permit adequate control of the facility by one operator.
Only recently was the requirement instituted for a second licensed operator in the control room.
This requirement was necessitated by a special duty assigned to the second operator and not because the design of the control room in any way inhibited adequate control of the f acility.
- 36. INTEROGATORY:
The questions below relate to.the_following sequence or events which may occur in connection with a feedwater transient.
Scenario
~
A feedwater transient occurs, concurrent with a loss of main feedwater to the steam generators.
The feedwater transient results in an immediate reactor trip.
The AFW is activated and the operators align the level of the AFW in the steam generator. '
Primary reactor pressure drops below 1600 psi.'
When primary reactor pressure reaches 1600 psi, the HPI is activated and primary reactor coolant pumps are turned off.
The HPI will remain on for at least 20 minutes.
(a)
Does the foregoing scenario accurately describe an expected sequence of events in the event of a feedwater transient?
If not, please describe such a normal response.'
(b)
At each step in the foregoing scenario (either as described above or as revised per response to the previous question), please describe:
(i) Each action which must be taken by operators (including the number of operators who must be involved and the procedure, if any, which calls for such action);
(ii)
The reason for each operator action; (iii) The training which each operator has received to ensure that the foregoing actions will be correctly implemented; and (iv)
Provide a facility diagram (including a detailed control room diagram) showing where each action must be implemented.
ANSWER 36(a):
The foregoing scenario does not accurately descri.be 18551 261
~
an expected sequence of events in the event of a feedwater tran-sient.
The expected sequence of events would include the initia-tion of a feedwater transient by loss of all main feedwater to the steam generators; a reactor trip as a result of this tran-sient; and activation of the auxiliary feedwater system.
At that point the auxiliary feedwater system would automatically control steam generator level at approximately the 30-inch level, and no operator action would be necessary to align the level in the steam generator.
The reactor coolant system would not drop below 1600 psi, and therefore the reactor coolant pumps would not be turned off.
Additional makeup water would be added as necessary to assure that propec operating pressure and pressurizer level are maintained.
Actions which must be taken by the operators in this scenario are described in Emergency Procedure D-2 " Turbine / Reactor Trip", and Emergency Procedure D-14 " Loss of Steam Generator Feed".
These actions can be catried out by three operating personnel from the control room.
ANSWER 36 (b) (i) and (ii) :
The steps prescribed _for operatnr action and the reasons therefore in Procedures D-2 and D-14 are self-descriptive and generally assure that the reactoc is placed in a safe snutdown condition, that auxiliary electrical loads are properly transferred as necessary, and that the reactor coolant system is maintained in such a configuration that subcooling can be assured and that auxiliary feedwater is provided to the steam generators to provide an adequate heat sink.
ANSWER 36(b) (iii) :
Each of the three operators involved in the scenario above have been certified as licensed operators by the Nuclear Regulatory Commission.
The extensive training that both 18'J4 262
~
the simulator and on-site training provide include familiarity with the control room and Operating Procedures D-2 and D-14 to assure that these operators can correctly implement the steps of.
those procedures.
ANSWER 36(b) (iv) :
See the answer to Interrogatory 28. _____
4 18 % 263.
g UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD _ _ _ _ _ _ _ _
In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312
\\
(Rancho Seco Nuclear Generat3 ng S tation) )
AFFIDAVIT OF R.
J.
RODRIGUEZ County of Sacramento)
SS State of California )
R.
J.
Rodriguez, being duly sworn according to law, deposes and says that he is Manager, Nuclear Operations Depart-ment of the Sacramento Municipal Utility District; and that the answers contained in " Licensee's Answers (Set No. 2) to California Energy Commission's Second Set of Interrogatories to the Sacramento Municipal Utility District" are true and correct to the best of his knowledge and belief.
m Y\\M R.
J.
Rodrig ez '\\
Sworn to and subscribed before me this '
day of January, 1980.
Notary Public
)h$h 2h4 e...............................e ometAL SEAL PATRICIA K. GEISLER NOTARY POSUC CAUFQWNBA SACR ENT C TV My Commason Empires November 22.1933 9
-