ML19340E117
| ML19340E117 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/30/1980 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-TM P-80444, NUDOCS 8101060448 | |
| Download: ML19340E117 (43) | |
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i public SeMee Company of Oderdo 4
Dece-ber 30, 1930 i
Fort St. Vrain j
Unit No. 1 i
P-30444 1
i Mr. Darrel Eisenhut, Director j
l Division of Reactor Licensing
[
Office of Nuclear Reactor Regulation i
U. S. Nuclear Regulatory Commission
]
Washington, D.C.
20555
SUBJECT:
Fort St. Vrain Unit No. 1 l
Action Plan Requirements i
REFERENCE:
P-80438
Dear Mr. Eisenhut:
In the above referenced letter we indicated in our response to Item j
I.A.1.1 that we would submit our STA program by January 1, 1981.
Please find attached the following documents which serve as our response: : Compares the Fort St. Vrain training program with that proposed by INFO.
There. are some minor differences in responsibilities which were dictated by our specific organi:ation and some differences in educational courses which were 4
dictated by differences in reactor technology of an HTGR versus an LWR.
.0ne of the major differences involves simulator training. As you are aware, we have no simulator facilities on site and have no general simulator facilities wnich we can use.
7 Sets forth' the Technical Advisor's duties, resconsibilities and accountacility.
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_, : Sets forth the Technical Advisor's qualification and training requirements as well as
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requalification requirements.
. : Sets forth.the detailI for Tecnnical Advisor training.
With the
. exception of management /sucervi;ory training the Technical
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' Advisors at' Fort St.
Vrain have received the training outlined in-this attachment.
The Technical Advisors are enrolled in a
1 management / supervisory course which is scheduled for comcletion in June, 1931.
S101060 Yg
. The above listed attachments represents our program for the Technical Advisors and, of course, any new Technical Acvisor candidates would be reautred to fulfill the program set forth.
Concerning the long term use of Technical Advisors (TA's) and tne eventual phase out of TA's our position is as follows:
1.
Retain the TA's as cermanent Fort St. Vrain staff for the duration of operating license in lieu of upgrading SRO's to a college level engineering degree.
2.
Vograde the operator and SRO training and requalification reauirements to include courses in heat transfer, fluid flow, and thermodynamics as well as other upgrading as outlined in Attachment 5, but retain the Technical Advisors as the college level expertise rather than providing this expertise to the operator.
In our opinion and for our particular set of circumstances the above position has many distinct advantages.
Upgrading operating shift personnel to a college level degree in engineering can only result in excessive turnover rates in the operating staff and an overall recuction of experience in the long term.
In our opinion, no amount of training, college level or otherwise, can substitute for hands-on experience.
Since our Technical Advisors are on call, rather than on shift, we can avoid the problem of having degreed personnel on shift while at the same time provide the required college level expertise for accident situations, thus avoiding excessive turnover in the Technical Advisor staff while maintaining a high experience level in the operating staff.
While this type of program may represent a departure from your l
present thinking we believe it is in keeping witn the objectives of l
the TMI Action Requirements and is deserving of your censideration.
l With reference to Item III.D.3.3 we have included as Attachment 6 some further clarification.
Very truly yours, WM-c Don W. Waremoourg Manager, Nuclear Production For St. Vrain Nuclear Generating Station l
CWW/alk Attachments
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s ATTACRMENT 1 COMPARISON OF INFO WITH FSV PLUT l
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a TASK I.A.l.1 - Compare INPO with FSV Plan for Technical Advisors I.
Function A.
" Provide assistance during normal and abnormal operating conditions."
Fort St. Vrain
" Provide assistance during abnormal conditions."
II.
General Qualifications l
A.
None listed for Fort St. Vrain III. General Duties i
A.
"~ognizant of plant and equipment..."
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Fort St. Vrain
" Cognizant of plant..."
IV.
Tvpical Resoonsibilities l
A.
INPO's statements are based on an on-shift Technical Advisor; therefore, they state "during transients...etc."
l Fort St. Vrain's statements are based on an on-call Technical l
Advisor; therefore, they state "after transients and during l
accidents..."
B.
INFO has Technical Advisor report' abnormalities to Shift Super-l visor.
I Fort St. Vrain has Technical Advisor report abnormalities to Operations Manager.
C.
INPO has Technical Advisor recommend procedure changes to Shift i
Supervisor.
Fort St. Vrain has Technical Advisor recommend procedure changes to Superintendent of Operations.
D.
INPO has Technical Advisor interpreting and applying Technical l
l Specifications.
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I Fort St. Vrain has Technical Advisor applying Technical Specifications.
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_2-E.
INPO has Technical Advisor " review planned shift activities..."
Fort St. Vrain does not (since Technical Advisor is not on shift).
F.
INPO has Technical Advisor evaluate effectiveness of plant in-structions.
Fort St. Vrain does not.
G.
INPO has Technical Advisor evaluate core power distribution during and following load changes and perform hot channel factor and rod programming.
Fort St. Vrain has Technical Advisor evaluate region peaking factors and/or core reactivity status.
H.
INPO has Technical Advisor prepare special reports as requested by Superintendent of Operations.
Fort St. Vrain has Technical Advisor prepare special reports as requested by Technical Services Supervisor.
V.
Accountability A.
INFO has Technical Advisor " observing plant status..."
Fort St. Vrain has Technical Advisor " reviewing plant status..."
B.
INPO has Technical Advisor " maximizing plant safety during transient or accident situations..."
Fort St. Vrain has Technical Advisor " maximizing plant safety during accident situations..."
VI.
Experience A.
INP0 requires 12 months nuclear power plant experience at same plant.
Fort St. Vrain requires 6 months.
B.
INPO talks of experience gained prior to initial fuel loading.
Fort St. Vrain - N/A
_3 VII. Absences from Shift Technical Advisor Duties A.
INPO requires technical Advisors not actively performing Technical Advisor functions for 30 days to receive training.
Fort St. Vrain states 60 days before training is required.
VIII.
Education and Training Requirements A.
Fort St. Vrain requires an official college transcript.
INPO does not.
B.
INFO requires curriculum and instructor of courses not administered by colleges or universities certified by INFO.
Fort St. Vrain requires instructor to have appropriate academic credentials and retains outline of lectures for 4 years.
C.
INPO lists number of contact hours for a given course.
Fort St. Vrain does not.
D.
INPO requires " Laplace transforms to interpret control response" (Mathematics).
Fort St. Vrain does not.
E.
INPO requires Statics.
Fort St. Vrain does not.
F.
INPO requires " fuel densification" training for reactor materials.
Fort St. Vrain requires " binary alloys" training for reactor materials (as recommended by Dr. Olson).
VIII. Applied Fundamentals - Plant Specific A.
INPO requires training at the college level on plant specifics.
Fort St. Vrain plant specific training is under the direction of the Training Supervisor.
. B.
INPO separates the subject matter for plant specifics under various categories.
Fort St. Vrain combines most of the subject matter under one category.
All of the required subje'ets are covered. Management / supervisory skills is still a separate category.
X.
Simulator Training A.
INPO requires simulator training.
Fort St. Vrain has no simulator.
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ATTACHMENT 2 TECHNICAL ADVISOR DUTIES I
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12/8/80 s
TECFNICAL ADVISOR DUTIES Function Provide advanced technical assistance to the operating shift during abnormal operating conditions.
1.0 General Duties (1) During assigned tour of duty be cognizant of plant status.
(2) Maintain independence from nor=al plant operations as necessary to make obj ee:1ve evaluations of plant operations and to advise or assist plant management in correcting conditicus that may compromise the safety of operatices.
(3) 3e readily available to provide appropriate assistance to the operating shift.
2.0 Tveical Reseensibilities (1) After transients and during accidents, compare actual critical parameters.
(i.e., neutron power level; reactor coolant system, pressure and tempera-ture; reactor building pressure, temperature acd radiation level; and plant radiation levels) vi:h those predicted in the Plan Transient and Acciden: Analysis, to ascertain whether the plant responded to the in-cident as predicted.
Report any abnormalities to the Operations Manager and provide assistance in for=ula:ing a plan for appropriate action.
(2) Make a qualitative assessment of plant parameters during and following an acciden: in order to ascertain whether core damage has occurred.
(3) During e=ergencies be observant of critical parameters, ascertain that there is adequate core cooling including availab111:7 of a heat sink for the coolan: system, and in the event that cri:ical parameters become unavailable due to instrument failure, perform calcula: ions or through 7
l c:her means determine approximate values f or the para =eters in question.
(4)
Investigate the causs(s) of abnor=al or unusual events occurring and l
assess any adverse affects. Recommend changes to procedures or equip-men as necessary to prevent recurrence.
(5)
Evalua:e the effectiveness of plant procedures in terns of terminating or mitigating accidents and =ake recon =endations to the Superintendent of Opera:1ons when changes are needed.
l (6) Assist the opera: ions staff in applying the require =en:s of Technical Specifications.
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_2-(7) Evaluate region peaking factors and/or core reactivity analysis.
(8) Review abnormal and emergency procedures.
(9) Prepare special reports when requested by the Technical Services Supe.r-visor.
(10) Provide an engineering evaluation of Licensee Even: Reports from other plants as assigned.
3.0 Accountability The TA is accountable for the following:
(1) Contributes to aximizing safety of operations by independently reviewing plant status and advising operations supervision of conditions that could compromise plant safety.
(2)
Contributes to maximizing plant saf ety during accident situations by indeoendently assessing plant condi:Lons and by providing the technical assistance necessary to mitigate the inciden and mini =1:e the eff ect on personnel, the environment and plant equipment.
1 ATTAC1 DENT 3 TECHNICAL ADVISOR OUALIFICATIONS AND TRAINING t
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12/8/80 TECHNICAL ADVISOR OUALITICATIONS AND TRAINING The Public Service Cenpany of Colorado Technical Advisor course is to provide qualified personnel with suffician: knowledge to prepare them-selves to assume the duties of Technical Advisors.
1.0 Exeerience 1)
The Shif: Technical Advisor shall have a minimum of 18 months of nuclear power plan experience, at least two months of which shall he a: an operating nuclear plant.
- 2) A maxi =um of six months of this experience =ay be obtained in the military or at a p;cduction nuclear plant and should be evaluated on a case-by-case basis.
- 3) A maximum of three months of systems and opera:1ons training may be applied toward these experience requirements.
At least 6 months of this experience shall be at the station at which the posi: ion is to be filled.
This may be vaived in part when two essentially identical plants are involved.
2.0 Education and Training Recuirements A valver for any of the required education or training shall be granted only by the Manager of Nuclear Production and should be evaluated on a case-by-case basis.
For courses completed at an accredited college, a semester credit hour shall be considered equivalen: :o approximately 15 contact hours in a full-time training program. An official college transcript will be re-i quired for each insti:ution that he attended.
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'Jhen courses prescribed in Section 2.1.2 are not aaninistered by an accredited college or university, the instrue:c shall have appropriate academic credentials. An outline of the lec:ures will be retained in l
the files for four years.
2.1 Education 2.1.1 prerecuisites 3evond Eish School Dieloma - I: is assumed that many candidates may have received previous ::aining and are qualified to begin the coursework prescribed in 2.1.2.
Pre-requisite education considered necessary for successful com-pletion of the advanced coursework is iden:1fied below.
This coursework may be waived without formal documentation of specific course completion.
Mathematies Trigonometry, Analytical Geometry, College Algebra
3 "
Chemistrv i
Inorganic Chemistry Physics Engineering Physics (Haat, Mechanics, Light Sound, Electrici:7 and Magnetism) l 2.1.2 College Level Fundamental Education Mathematics Engineering Mathematics through the introduction to ordinary differential equations Reacter Theory Atemic and Nuclear Physics through 2-group Diffusion Theory Kinetics, Reactivity Changes Reactor Chemistry Inorganic Chemistry (as related to reactor systens)
Corrosion - Reaction Rates Nuclear Materials Strength of Materials Reae:c: Material Properties (phase diagrams, binery alloys)
Thermal Sciences (for nuclear systems)
Thermodynamics Laws of Thermodyrm ies Properties of k'ater and Steam Steam Cycles and Ifficiency Fluid Dynamics Bernoulli's Equatica Fluid Friction and Head Loss Elevation Head Pump and System Characteristics Two Phase Flow Heat Transfer Methods of Heat Transf er Boiling Hea: Iransfer Heat Exchangers Electrical Sciences Electronics (Circuit Theory, Digital Electronics)
Motors, Generators, Transformers, Switchgear Instrumenta: ion and Centrol Theory
3-Nuclear Instrumentation and Control Reactor Instrumentation Reactivity Control and Feedback Nuclear Radiation Protection and Health Physics Radiation Detectors Biological Effects Radiation Survey Instrumentation Shielding 2.2 Plant Specific Training The Technical Advisor training program will be accompanied by formal classroom training. The Technical Advisor training will be under the direction of the Training Supervisor. The formal training will con-sist of the following subjects:
- 1) Detailed Plant Systems Series
- 2) Health Physics
- 3) Plant Controls
- 4) Plant Protective System
- 5) Technical Specifications 6)
Emergency Procedures
- 7) Administrative Procedures
- 8) Refueling
- 9) Reactor Physics and Fort St. Vrain Core Physics
- 10) Heat Transfer, Thermodynamics, Fluid Dynamics and Engineering Materials
- 11) Accident Analysis 12)
OPOP's Technical Advisors who have expertise in any of the above subjects will be exempt from taking those course.
Example: Holding an SRO License 2.3 Management /Suoervisor Skills l
Subject Leadership Interpersonal Communication Motivation of Personnel Problem and Decisional Analysis Command Responsibilities and Limits Stress Human Behavior NOTE:
For initial Technical Advisors, to be completed by June, 1981.
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2.4 Annual Recualification Training Subject Material _
Review of transient and accident analyses of FSAR events e=phasizing the individual's role in accident assessnent. Review selected in-dustry events and LER's that could have led to more serious indidents.
2.5 Absences from STA Duties Persons not actively perfor=ing the STA functions for a period of sixty (60) days or longer shall, prior to assuming responsibill:ies of the position, as a mininem receive training sufficien: to ensure he is cognizant of f acility/ procedure changes that occurred during his absence.
Persons no: perfor=ing the STA function for a period of six (6) months or longer shall, prior to assuming the responsibilities of the position, receive the annual requalification training described in this docu=ent.
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ATTACH'ENT 4 DETAILED TECHNICAL ADVISOR TRAINING 1
(T FORT ST. UHAlH TF '4ItH: DEPARTHENT O
Teclinical Advisor Training Date 8/04/80 - 8/_28/80 Hon.l.Ty Tuesday _
Wednenday
.Tliu r uday Friday 8/04/80 8/05/80 8/06/80 53/07/80 8/08/80 Site and PCRV PCRV Piping and Reactor Fuel & Reflectors Control Rod Drives Review and Quiz overall Plant System Core Components llTGR Tecli. Course FSAR, Section 11 and SD-17 Section D and E Appendix C SD 11-2 P. Moore P. Moore S. Willford S. W111 ford P. Moore Write and give quiz 8/11/80 8/12/80 8/13/80 8/14/80 8/15/80 RSD and Orifices System 23 Systems 41 and 42 Systems 46 and 47 Review and Quiz Primary Coolant l
S. Willford
)
i P. Moore P. Moore P. Moore P. Moore Write and give quiz 3
8/18/80 8/19/80 8/20/80 8/21/80 8/22/80 Systems 62, 61 and 61 System 82 System 91 System 92 System 92 I
3 T. Clarlsler P. Moore T. Clirialer S. Willford S. W111 ford 8/25/80 8/26/80 8/27/80 8/28/80 8/29/80 l
l Quiz System 21 System 21 Steam generators Review and Quiz Aux 111 aries Basic Design Overall System System 22 T. Clirisler T. Christer Write and give quiz R. Wadas P. Moore P. Moore Write and give quiz e
I FORT ST. VitAt ti T!
Vit tlG IlEl'ARTilEllr i
Teclinical Advisor Training 11a t e 9/01/80_ _9/_26/80
]Ionday Tuesday Wednesday Tliurnday Friday _ _
9/01/80 9/02/80 9/03/80 9/04/80 9/05/80 lloliday Systems 53 and 54 Systems 51 and 52 Systems 24 and 25 System 84 Systems 31, 32 and 33 EllC T. Chrisler/P. Moore S. Willford P. Moore P. Moore 9/08/80 9/09/80 9/10/80 9/11/80 9/12/80 General Atomic Seminar on Accident Analysis
}
9/15/80 9/16/80 9/17/8Ti 9/18/80 9/19/80 Thermodynamics Systems 44, 45 and 48 Ileat Transfer and Fluid Systems 13, 14, 15 and 1()
Materials Flow (Fuel llandling)
Dr. Olson S. Willford Dr. Olson P. Moore Dr. Olson
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9/22/80 9/23/80 9/24/80 9/25/80 9/26/80 llent Transfer /Thermo-Review Coytrol Systems Control Systems PPS dynacles Exam pp3 OPOP's T. Chrisler S. Willford/P. Moore S. Willford/P. Moore S. Willford i
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Teclinical Advisor Training Date 9/29/80 - 10/24/80 flonila y Tuesday
~ Wednesday I
Thur uilay Friday 9/29/80 9/30/80 10/01/80 10/02/80 10/03/80 Quis Review Ilealth I'hysics Nuclear Instruments Nuclear Instruments and Radiation Monitors S. Willford/
S. W111 ford P. Moore T. Chrisler T. Chrisler T. Chrisler 10/06/80 10/07/80 10/08/80 10/09/80 10/10/80 Emergency Procedures Emergency Procedures Emergency Procedures Emergency Procedures Emergency Procedures P. Moore P. Moore P. Moore P. Moore P. Moore l'
10/13/80 10/14/80 10/15/80 10/16/80 10/17/80 Quis No School Review Quiz and Technical Specifications Technical Technical Specifications Specifications P. Moore S. Willford S. H111 ford R. Wadas l!
10/20/80 10/21/80 10/22/80 10/23/80 10/24/80 j
Quiz Basic Nuclear Concepts Basic Nuclear Concepts Basic Nuclear Concepts Basic Nuclear Con-Tapes 5, 6 and 7 Tapes 8, 9 and 10 Tapes 11, 12 and 13 cepts Tapes 14, 15 and 16 P. Moore
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Technical Advisor TraininS Date 10/27/80 - 12/05/80 Honday Tuesday
_ _ _ _ _ _Wednemlay Thurenlay Friday 10/27/80 10/28/80 10/29/80 10/30/80 10/31/80 Reactor Operation Reactor Operation Reactor Operation Reactor Operation Reactor Operation Tapes 1, 2 and.1 Tapes 4, 5 and 6 Tapes 7, 8 and 9 Tapes 10, 11 and 12 Tapes 13 and 14 11/03/80 through 11/21/80 In Plant Training 11/24/80 11/25/80 11/26/80 11/27/80 11/28/80 Chemistry Physics Physics and floliday Holiday (in plant)
SR 5.1.5W (Shutdown margin)
V. Lucero P. Moore P. Moore /C. Fuller 12/01/80 12/02/80 12/03/80 12/04/80 12/os/80 Radiological Emergency Self Study and Review Review Review Final Exam Response Plan of the ADH's and Engineered Safeguards in the Ahstract Book T. Chrisler
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PP-80-1330 INTER DEPARTMENT MEMO - PUBLIC SERVICE COMPANY OF COLORADO October 20, 1980 DATE Don W. Warf t r W anager, Nuclear Production Fort St. Vrain TO OEPARTMENT oR C6VISloN R. E. Wada_,
raining Supervisor Fort St. Vrain FROM CEPARTMENT OR Q4Vl53ON ATTN.
TECHNICAL ADVISOR TRAINING SUBJ.
Letter G-80154 from the NRC lists the recommendulon:, from INFO as to the training requirements for Technical Advisors.
s ction 6.3 of INFO's recem-mendation refers to the " Management / Supervisory Skills" training. This includes the following subjects:
)'
Leadership 2)
Interpersonal Communication
- 3) Motivation of Personnel
- 4) Problem and Decisional Aulysis
- 5) Command Responsibilities and Limits 6)
Stress
- 7) Human Behavior 1._
This should involve 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of training.
Mr. Wilbur Weir of the downtown Education Department has advised me that all these areas are covered in the 18 (3-hour) sessions of " Basis of Supervision," which begins in November, 1980, and is completed in May, 1981.
I would recommend that the Technical Advisors - Roger Heller, Charles Fuller 2
and Frank Novachek attend the " Basis of Supervision" course to fulfill this portion of their training requirements.
The NRC states that Technical Advisor training should be completed by January 1, 1981. However, within the time frame allowed, it seems reasonable the NRC would accept this schedule as a reasonable time to complete this portion of their RLff'
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training.
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'R. E. Wadas Training Superviser
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FSV SEMINAR FOR SHIFT TECHNICAL ADVISORS Introduction The Shif t Technical Advisor (STA) position at Fort St. Vrain (FSV) has been created in response to the NRC's instructions following the Three Mile Island (TMI) incident. A training-workshop se=inar is described herein to supplenent the training progra=s that may be available under PSC's direction.
The intent of this seminar is to acquaint PSC personnel in those areas of Safety Expertise developed at GAC which are directly beneficial to the technical advisors in performing their intended function.
Three specific objectives of the se=inar are proposed:
1.
To provide a description of the analytical =edels used to provide insight into plant transient responses to upset and accident condi-tiens. Results frc= the analytical =odels will be related to actual plant behavior.
2.
To review accident cases involving interruption of forced cooling and various possible responses which are not included in the Final Safety Analysis Report (FSAR),
3.
To provide the STAS with sc=e background on accident progression analyses.
The e=phasis in this area vill be to provide the STAS with sc=e working knowledge of event trees and how they are used to evaluate possible consequences of accident conditions.
Descriotion of the Se=inar The seminar is planned as a 5-day workshop involving approxi=ately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classrec= time with adequate ti=e for direct participation by the attend-All =aterials presented will be asse= bled into a notebook for use ees.
by the attendees.
Further, an audio recording of the secinar will be =ade for PSC's use.
A descriptive cutline of the =aterial to be presented and discussed is as follows:
1.
General Overview of Safety Philosochv, Accroxi=atelv 1/2 Dav 1.
In this discussion the general safety philosophy governing the design of FSV will be reviewed.
Exa=ples of the tcpics are:
The Plant Protective System (?PS) and why it works the way it does.
Why shutdown of syste=s and interruptiens of forced circu-lation are not unanticipated events.
The i=portance of the operator in responding to accident situations.
Defense in-depth philosophy.
Actions in progress to reduce challenges to safety syste=
(PPS Setpoint Re-evaluation).
2.
A discussion of FSAR accidents and why they are conservative predictions of plant perfor=ance.
The following accidents which have been updated since the FSAR will be discussed:
Firewster Cooldown.
Hypothetical Design Basis Depressuri:ation Accident (b3A-2).
Prestressed Concrete Reactor Vessel (PCRV) depressuriza-tion following a per=anent loss of forced circulation (DBA-1).
3.
Consequences of TMI incident on FSV and responses to NRC questions will be discussed.
Specific topics are as follows:
Shielding study.
Adequacy of radiation instru=entation and pri=ary coolant sa=pling.
Evacuation planning :ene and ingestion pathway distances.
II.
Review Plant Transient Resconse - Accroxt=stelv 1 Dav 1.
Description of the Transient Analysis Progra: (IAP) Code -
The IAP code =odels the Nuclear Stea: Supply (NSS) co=ponents and the turbine stea= syste= to predict the plant transient response during upset conditions.
a)
Introduction A brief de'scription of the purpose and application of the code.
Background and status of the code.
b)
Description of the =ajor plant co=ponents and how they are modeled. Model description and i=portant assu=ptions or li=its.
The =ain ele =ents of the code to be discussed are:
The pri=ary coolant systc= co= prised of the core, circulators, and steam generators.
The turbine stea= syste= cc= prised of the turbine generator, extraction stea=, bypass stes=, and feed-water train.
2.
Description of the overall plant control system and how it is modeled.
This discussion will include an overview of the general design philosophy for the control syste=s and include such subj ects as:
a)
Steam te=perature droop.
b) Attemperation program.
c)
Interlock Sequence Switch (Manual / Auto =atic Control).
d)
Ti=e constants.
3.
Co=parison of predicted versus observed plant perfor:ance.
Two exacples will be discussed:
a)
Loop trip.
b)
The discussion will illustrate the sequence of events, plots of IAP results versus actual FSV results from the Model Verffi-cation Data. The discussion will include the effects of
=anual control intervention.
4.
Group Discussion on Transient Plant Perfor=ance a) Operation in the 12-22" power range (not recoc= ended for prolonged ti=e periods),
b)
Discuss the characteristics of a well behaved plant transient versus an ill-behaved transient.
The areas of concern are:
Rate of chcnge of steam te=perature at the steam generator cutlet.
Rate of change of gas te=perature at the core outlet.
c)
Influence of Rod Withdrawal Prohibit on transients.
III.
Review of Plant Resconse to Accidents, Accroximatelv 2 1/2 Davs 1.
Description of Reactor Energency Cooling Analysis (RECA) Code -
The RECA code codels the core and pri=ary coolant system com-ponents to predict the detailed fuel and gas te=peratures during accident conditions.
a)
Introduction A brief description of the purpose and applications of the code.
I b)
Model description and i=portant assu=ptions or li=1ts.
Core Model - a three di=ersional ther=al si=ulation with axial flow channels representing the coolant flow.
Natural convection flow analysi capability.
Reactor kinetics are not included.
2.
Co=parison of Predicted versus Measured Results.
a)
Four reactor scra=s between 30% and 50% power were analyzed and co= pared to observed results.
The co=parisons verify the analytical codels for flow distribution and heat transfer in the core.
b)
An analysis of shutoff of forced circulation is co= pared to observed results.
The results verify the capability of analyzing natural circulation flows in the core.
3.
Results for Accidents Involving Interruption of Forced Circu-lation (10FC).
a)
PCRV Pressure Intact Discussion of natural circulation effects in the core (reverse flow) and in the pri=ary flow circuit.
Discussion of'eritical co=ponents and their bases.
The co=ponents to be discussed are fuel, orifice valves and the ther=al barrier.
Results for restart of forced circulation.
Co=parison of results for 1-1/2 hour IOFC versus no IOFC.
b)
Depressuricing the PCRV Discussion of the li=iting case of 5 hrs. 10FC.
Discuss the li=iting co=ponents in the flow circuit when forced cooling is rest =ed.
Discussion of case involving depressurizing the PCRV co==encing after 2 hrs. IOFC.
This-discussion will include the characteristics of co=ponents in the heliu=
purification syste=.
4.
Accidents Involving Depressuri:stion of the PCRV.
This discussion will generally cover only those cases which have undergone =ajor revision fro = what is presented in the FSAR.
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Discuss the characteristics of critical co ponents and their bases.
The co=ponents to be discussed are the fuel and the ther=al barrier.
Discuss why reverse flow is not important when the PCRV is depressurized.
Discuss sources for driving circulator pelton wheels, i.e.
feedwater, condensate and firewater.
Illustrate cooling capability and transient results.
Discuas the effects of delay ti=e in restarting forced circulation.
Discuss the effects of different cooling flow rates when i
forced circulation is re-established.
5.
Group Discussions Group discussion will be encouraged as a for= of =utual benefit to the participants. Possible topics for group discussion are:
a)
Variations of Accidents.
Co==on ele =ents a=eng IOFC accidents.
Effects of initial power at part load.
b)
Potential Operator Actions.
Orifice valve readjust =ents.
Restart of forced circulation after =aking co==it=ent to depressurize.
c)
Evaluation of Plant Status During an Accident.
I What can one =enitor to deter =ine that core cooling is progressing as expected.
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When can core cooling be ter=inated.-
IV.
Accident Precressien Analvses. Accroxi=ately 1/2 Dav 6
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1.
Introduction.
Description of the general topic of risk assess =ent.
2.
Event Trees and Fault Trees.
a)
Discussions of the evolut4cn of event tree and fault tree
=ethods.
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b)
Discussion on the approaches to event tree construction.
Sinple event tree construction exacple.
Sinple event tree,onstruction problen.
3.
Applications of Event Trees.
a)
Based upon the experiences at FSV, the developnent of the,
event tree for in:erruption of forced circulation will be discussed as an exa=ple.
b)
Discussions of event trees in accident progression analyses and how they provide an aid to assessing precursor events.
c)
Use of event trees in incident report writing.
d)
Uses of event trees as a training tool.
Recognizing i=portant risk contributors.
Identification of short tern and long tern strategies to cope with a given event.
Developing operations procedures.
7.
Su--nrv. Accroxinatelv 1/2 Dav 1.
Review highlights of the nacerial presented by GAC.
2.
Review the nain points covered in the group discussions, l
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t l-I ATTACHMENT 5 REOUALIFICATION PROGRAM, LICENSED OPERATORS
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.4J REQUALIFICATION PROGRAM 1.0 GENERAI, DESCRIPTION The Licensed Operator Requalification Program at Fort St. Vrain is designed to maintain a high degree of knowledge and job proficiency as required by 10CFR50 and 10CFR55, Appendix A.
The program will apply to NRC Licensed Opers.cors. Sr.nior Licensed Operators, and Special Senior Licensed Oparators (fuel handlers).
Special Senior Licensed Operators shall participate only in chose pcreions of the program that are applicable to their specific licensed activicias.
Tha R. qualification' Program vill be conducted on a two-year cycle, divided into raqualificaticn year 1 and 2.
Each requalification year runs fron Memorial Day ?.o Memorial Day.
The Tcaining Supervisor has the responsibility e.nd authority for implementing the Licen. tad Operator Requalification Training Program.
Licer. sad Personnel sre exempt trom those portions of the Requalification Progau for which they have primary responsibility as part of their normal du t:.e s. 7cr exampla:
a)
Indiriduals who administer and grada examinarions need not take the examination.
b)
Individuals who prepare, review, and/or approve significant Facility and procedure changes or Reportable Occurrences need not review them as part of requalification.
Individuala who are not normally engaged in actual control manipulations or directly supervise the actual control manipulation (back-up" licenses) shall participate in all portions of the Requalification-Program not included in their particular job duties.
2.0 REQUALIFICATION SCHEDULE 2.1 Licensed and Senior Licensed Operator's lec:ures are presented each requalification year between Labor Day and Memorial Day. Lectures may be postponed during periods of intense plant activity such as re-fueling or major shutdown. Lectures shall be conducted in the following areas and should be tailored to include equipment and systems that have exhibited unusual operating problems and/or subjects that required fort-ification as determined by the previous annual requalification examinations.
a) Theory and Principles of Operation b) General and Specific Plant Operating Characteristics including expected, response to equipment failure c) Plant Instrumentation and Control Systems d) Plant Protection Systems e)
Engineered Safety Systems f) Normal, Abnormal and Emergency Procedures g) Radiation Control and Safety and Plant Chemistry I
h) Technical Specifications f
i) Applicable portions of Title 10, Chapter 1, Code of Federal Regulations j) Heat Transfer, Fluid Flow and Thermodynamics k) Mitigation of Accidents involving a Degraded Core 2.2 Special Senior License Operator's lectures will be conducted within the three months prior to the annual refueling outage. Lectures will be conducted in the following areas and will be tailored to include equipment and systems that have exhibited unusual operating problems and/or subjects I
l that require fortification as determined by the previous annual requalifi-cation examinations.
a) Reactor and Fuel Characteristics b) Equipment, Instrumentation and Design c) Procedures and Limitations d) Emerhricy Systems and Safety Devices e) Health Physics and Radiation Protection The requalification exam shall be administered at the conclusion of the lee-ture series. The criteria for adequate performance for Special Senior Licensed Operators will be the same as described in Section 4 Prior to in-core fuel handling, each Special Senior Licensed Operator shall manipulate the controls of the fuel handling machine in moving new fuel to' storage.
3.0 REQUALIFICATION PROGRAM STRUCTURE Licensed Operator Requalification Program shall be structured in two main sections which shall normally run concurrently. The implementation of any section or aspect thereof, may be delayed or rescheduled when the results of examinations indicate that some other aspect required priority. The two main sections of the program are:
3.1 LECTURE SERIES The lecture series will be of an estimated length of 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, but in no case less than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> each requalification year.
3.1.1 Lecture Series Attendance b'hile it is intended that all licensed individuals shall attend all lecture series presentations, an individual may be excused frem attendance at specific lectures when subject matter is in the area of the individual's expertise or when the individual has demonstrated sufficient knowledge by attaining a minimum of 80*
in the applicable category (s) of the previous year's annual written requalification examination.
t 3.1.2 Quizzes and/or Written Study Assignments Quizzes and/or written study assignments will be administered periodically for evaluation of individual knowledge and progress.
A minimum grade of 80% is acceptable. Grades less than 80% will require additional training in the area (s) of weakness. Training Instructors will be excused from taking those quizzes in subjects they have presented, quizzes they have prepared, or quizzes they have graded.
3.2 ON THE JOS/ ADDITIONAL TRAINING a) Performance of Reactivity Manipulation b) Significant Facility and Procedure Changes c) Abnormal and Emergency Procedure Review d) Plant operation drills 3.2.1 Plant Controls and E=ergency Manipulations Over the two-year requalification period, all licensed individuals shall manipulate the plant controls for all manipulations listed below. All control or emergency operation manipulations, if not actually performed, must be simulated by a walk-through with, and l
evaluation by, a member of the Training Staff. Normal control man-l l
1pulations such as reactor startups and shutdowns must be performed.
l Control Maniculations
- l)
Reactor startup with observable temperature feedback
- 2)
Reactor shutdown with observable temperature feedback l
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- 3)
Manual manipulation of steam generator and/or feedwater l
controls during startup and shutdown l
- 4)
Reactor power change of 10% or greater where load change is performed with load limit control or where flux, temp-erature, or circulator speed control is on manual
I Emergency Operation Manipulations
- 5) Moisture inleakage
- 6) Reactor scram
- 7) Loop shutdown
- 8) Circulator trip
- 9) Abnormal power change
- 10) Main turbine emergencies
- 11) Extended loss of sctive core cooling l?) Abnormal radioactivity
- 13) Fire 14)
Radiological alert or radiological emergency
- 15) Earthquake
- 16) Loss of instrument air header
- 17) Loss of hydraulic power
- 18) Loss of instrument bus
- 19) Loss of a DC bus
- 20) Steam leak or inadvertent opening of a main or hot reheat steam relief valve
- 21) Loss of access to Control Room
- 22) Loss of HVAC to 480 Volt switchgear room
- These manipulations must be performed on an annual basis; all other manipulations shall be performed at least once per two-year cycle.
3.2.2 Sirnificant Facility and Procedure Changes and Operating Experience Feedback a) All the licensed individuals are advised of significant l
changes to the Facility License, Facility Design and l
l Procedures by distribution of documents which state 1
change and the basis for change. Licensed individuals cognizance of the change is documented.
b)
It shall be the responsibility of the Training Super-visor to determine if changes to the facility and/or procedures are significant in nature to warrant review by Licensed Operators.
c) Operating experience feedback to licensed operators shall be conducted as specified in Plant Administrative Procedures.
d) Licensed individuals whose management function involves the prepration, review and/or approval of changes to the Facility License, the physical plant modification or operating procedures are not required to take part in the training on these topics by the Licensed Operator Requalification Program.
3.2.3 Technical Specifications, Abnormal and Emergenev Procedure Review Annual review by licensed individuals of all Technical Specifications,
, Emergency Procedures and Abnormal Operating configurations is re-quired. The method of reviewing shall include an appropriate com-b(nation of:
a) Walk-through of simulated emergency incidents corresponding to Reactivity and Emergency Manipulations of Section 3.2.1.
b) Plant operation drills c) Classroom review d)
Individual study 4.0 EVALUATION 4.1 The Review Panel a) An evaluation of the licensed individual's knowledge and skill shall be made by a Review Panel. A Review Panel shall consist
of the Training Supervisor and the Superintendent of Operations or their delegates.
b) The Review Panel uses the techniques described in the following sections to evaluate each licensed individual's knowledge and skill. The Review Panel may recommend one of more of the following courses of action:
- 1) Based on the annual nxamination grades, individuals may be excused from attending specific lectures.
- 2) A licensed individual receiving a grade of less than 70%
in any requalification examination category or an overall grade of less than 80% shall be placed in an accelerated review program. Within three weeks of review of the annual written requalification examination, licensed indi-viduals who receive less than 70% in any examination cate-gory or less than 80% overall shall undergo an oral exami-nation administered by a member of the training staff appointed by the Training Supervisor. The Training Super-visor shall notify the Administrative Services Manger and l
the Operations Manager of the oral and written examination
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results and provide a recommendation regarding the licensed individual's removal from licensed duties.
c) Records of all Review Panel activities, evaluations, and de-l cisions are maintained as part of the Licensed Operator Requal-ification Program documentation.
1 d) Report shall be submitted to the Administrative Services Manager.
5.0 EXAMINATIONS The licensed Operator Requalification Program includes:
a) Annual written examination b) Annual oral examination L
c) Periodic written quizzes on the current subjects of the lecture series d) Plant operation drills e) In addition to forming part of the basis for evaluating the per-formance of licensed individuals, the results of these examinations are used by the Training Supervisor in classroom and on-the-job training described above. Changes in content and/or schedule may be made to upgrade licensed individual knowledge and skill as deemed adviseable by the Training Supervisor.
6.0 PERFORV.ANCE EVALUATION 6.1 Routine Performance evaluation documentation shall be reviewed a minimum of once each calendar year.
6.2 Special Pollowing each review, a written evaluation of the licensed individual response shall be made by the Review Panel and retrained in the appro-priate training file. This evaluation shall be reported to the Admin-I istrative Services Manager.
If a licensed individual has been on duty during an incident involving use of an Emergency Procedure, his actual response shall be evaluated l
by the Shif t Supervisor and the Superintendent of Operations.
6.3 Inte rviews The Review Panel may conduct personal interviews with licensed indi-viduals to f acilitate reaching a decision regarding the course of I
action to be recommended.
Interviews are scheduled on an as-needed basis by the Review Panel.
Records of interview findings will be retained in the appropriate training file.
7.0 RECORDS Copies of the following requalification records shall be retained for two years following requalification togram completion:
a)
Lecture Attendance Records b)
Topic Quizzes and Quiz Answers c)
On-the-job Training Records d)
Annual Written Examination and Answer Key e)
Review Panel Evaluation f)
Acceler,ated Training Programs (if assigned) g)
License Personnel Graded Examinations and Quizzes 8.0 SIMULATOR TRAINING AND TESTING No suitable simulator is available for requalification training and testing.
As an alternate to this requirement, the following training will be conducted:
- 1) Increased emphasis will be placed on Plant Transients during the lecture series.
- 2) Annual requalification oral exams for all Reactor Operators and Senior Reactor Operators shall be given.
- 3) Training Instructors will simulate all reactivity and emergency conditions listed in Section 3.2 and walk through each of these manipulations with all Reactor Operators and Senior Reactor Operators over a two-year period.
If an operator has actually performed or supervised the performance of a manipulation satis-factorily, he will be given credit for performing that manipulation.
- 4) Plant operation drills will be periodically given on an individual or team basis and will be used to evaluate individual or team re-sponse to the drill conditions.
4 Each Reactor Operator and Senior Reactor Operator shall be involved in a mini =um of 5 drills / year.
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FURTHER CLARIFICATION, NUREG 0737, ITEM III.D.3.3 1
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III.D. 3. 3 Improved Inplant Iodine Instrumentation under Accident conditions PSC December 12,1979 (P-79299) REPLY to NUREG-0578 Section 2.1.8.C Improved In-Plant Iodine Instrumentation "PSC will identify the areas at Fort St. Vrain which require con-tinuous occupancy to mitigate the consequences of an accident.
Acceptable portable or cart mounted iodine samplers with attached single channel analyzers are not on the market today to the best of our knowledge.
To meet the requirenents of the Section, PSC proposes to take air samples, utilizing charcoal filter adsorbers, from those locations that require continuous habitability during accident conditions. Sampling will be analyzed utilizing a multi-channel analyzer, which will be located external to the reactor building to assure analytical capability in a timely f ashion under accident conditions.
These procedures will be in effect by January 1, 1980. We will continue to evaluate the development of portable iodine monitors and will purchase such equipment if and when reliable equipment becomes availabic.
By January 1, 1981, PSC will have the capability to remove an iodine sampling cartridge to a permanent, low background, low contamination area where accident condition iodine concentrations can be accurately measured."
PSC December 27,1979 (P-79312) SUBMITTAL:
Due to the co= plex geometry of the FSV reactor building, instal-lation of continuous on-line radioiodine monitors within the building does not appear f easible. As centioned in the P-79299 PSC reply (above) acceptable portable iodine samplers with attached single channel analyzers are not currently available on the market today to the best of our knowledge.
To meet the requirements of this section, PSC will use portable high volume air samplers utilizing charcoal filter adsorbers to obtain samples from those locations that require continuous or infrequent habitability during accident conditions.
It should be noted from Section 2.1.6.b of this letter that access to the reactor building should not normally be required past 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into the accident.
4 4
These portable high volume air samplers exist at FSV and pro-cedures are available for conducting air sample surveys through-out the postulated accident. The objective will be to assure personal exposures during sampling to as low as practicable.
The portable air samplers have been modified to incorporate a charcoal cartridge of the same size used in the radioiodine effluent monitor. Gamma spectral analysis of the cartridge will be I
perf ormed external to the reactor building to assure analytical capability in a timely fashion under accident conditions. Proc e-dures exist for gamma counting the cartridge and providing specific activities for all iodine radionuclides expected.
PSC December 24, 1980 (P-80 444 ) SUBMITTAL:
I As mentioned in the PSC December 27,1979, (P-79312) SUBMITTAL, an analysis of DBA#1, identified in FSAR Section 14.10 and Appendix D, was performed to evaluate rsJiological conditions in the reactor building as well as required operator actions involving reactor building access following the accident. DBA#1 involves i
a release of radioactivity equivalent to that described in Regulatory Guides 1.3,1.4, and 1.7.
Based on the required operator actions identified in the above analysis, areas requiring l
iodine sampling have been identified.
i Existing and budgeted equipment will be sufficient to assure that all needed airborne iodine concentration data can be collected under accident conditiona.
Existing iodine sampling equipment l
consists of 4 RADECO H809V portable low volume air samplers with particulate filters and capability to use charcoal or silver zeolite cartridges. Budgeted equipment includes two additional low-volume samplers of the above type and four continuous Air Monitors with iodine detection capabilities.
Sampling (Ref erence 1) and calibration (Reference 2) procedures 7
for the iodine sampling equipment have been developed. The continuous air monitors will be used to provide a continuous indication of reactor building iodine concentrations. Grab-samples will be taken following an accident using the portable low volume samplers equipped with particulate filters and charcoal cartridges. The cartridges will be taken to the radicchemistry laboratory for analysis. Silver zeolite cartridges are also available if necessary.
The radiochemistry laboratory is currently located in a room in the Security Search and Identification Building, which is expected to be a low background area during accident conditions, although not designed as such. The laboratory will be relocated to a i
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O permanent location in the Technical Support Building when construction is completed. This building is designed to be a low background and low contamination area during accident con-ditions. Equipment to be used for airborne iodine analysis includes a GeLi scintillation detector and multichannel analyzer.
The expected presence of noble gases will not interfere with the iodine analysis capability. Sample preparation (Reference 3) and operating (Reference 4) procedurer for the anlysis equipment are currently in use.
Due to the excellent resolution of the analysis system employed at Fort St. Vrain and the expected noble gas concentrations follow-ing the design basis accident, purging of the charcoal cartridges prior to counting should not be necessary.
If necessary, shorter samples can be taken to reduce radiation levels te acceptable va lues.
Health Physics technicians and radiochemistry lab assistants receive department training which includes the use of air sampling and analysis instrumentation. Periodic retraining is perf ormed to ensure competency in these areas.
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REFERENCES 1.
Fort St. Vrain Nuclear Generating Station, Health Physics Procedure HPP-12, " Sampling Procedure for High and Low Volume Air Particulate and Iodine Collection".
2.
Fort St. Vrain Nuclear Generating Station, Health Physics Procedure HPP-58, " Calibration Procedure for Airflow Measuring Devices".
3.
Fort St. Vrain Nuclear Generating Station, Radiochemistry Procedure RCP-9, "Sanple Preparation f or Gamma Spectral Analysis".
4.
Fort St. Vrain Nuclear Generating Station, Radiochemistry Procedure RCP-26, " Operating Procedure f or the Canberra "8100" Quanta System".
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