ML18082A945
| ML18082A945 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/08/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0694, RTR-NUREG-694 NUDOCS 8008200279 | |
| Download: ML18082A945 (32) | |
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Public Service Electric an(\\ Gas Company 80 Park Place Newark, i\\! J. 07101 Phone 201 /430-7000 August 3, 1980 Director of Nu~lear Reactor Regulatior.
United States Nuclear Regulatory Comm-~*3sion Washington, D. C.
20555 Attention:
Mre A. Schwencer,.Acting Chief Licensing Branch 3 Division of Licensing Gentlemen:
NUREG-0694 DATED REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR 3ENERATING STATION DOCKET NO. 50-3ll Public Service Electric & Gc..s hereby s11bmits, in the enclosure to this letter, additional information concerning the Dated Requirements" identified in NUREG-0694, "TMI-Related *Requirements for New Operati;1c-Licenses 11
- It should be r.oted that impl2Illentation of certain specific modifications described in the enclos.1re cannot be completed by the dates Epecified in NUREG-0694.
The principal reason for these del.:~ys is that the limited number of vendors who can supply eq11ip:;:ncnt to meet the requirements are attempting to provide ti:i.1ely delivery to the en ti re industry, but have limited produt:tion capabilities.
We have provided our proposed sche(1.ule for implementation of these specific it.ems.
Should you ha7e any questions, do not hesitate to contact us.
15~u1J.,~~
0~~~-
R. {. i.1i ttl General Manager -
Licens ~.ng and Environment Engine0ring and Construction Enclosure CC:
Mr. Leif Norrholm
'BO"). I Solem Resident Inspector I I, 9!) 2001 l40llM) D* 77
I.A.1.1
. ".:' *.... *.* ~, ~--'-'-.*-* _. -* -*----*-~ -----.
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DATED REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION SHIFT TECHNICAL ADVISOR The Shift Technical Advisor shall have a technical education, which is taught at the college level and is equivalent to about 60 semester hours in basic *subjects of engineering and science, and specific training in the design, function, arrangement and. operation of plant systems and in the expected response of the plant and instruments to normal operation, transients and accidents including multiple failures of equipment and operator errors.
This requirement shall be met by January 1, 1981.
(See NUREG-0578, Section 2.2.lB, and letters of September 27 and November 9, 1979).
Response
One Shift Technical Advisor (STA) is presently assigned to the shift to cover both Salem Units 1 and 2.
A class of STA's are presently undergoing training which will meet or-exceed the above requirements.
This training consists of a 36-week course which was developed by Westinghouse and is being taught by a Westinghouse college graduate instructor *. This class of STA's will complete the course and be assigned to shift duties by January 1, 1981.
A copy of the course of instruction is provided as Attachment*!.
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- ~--------*--*. -------- --~-------- -- -
I.A.2.1 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING AND QUALIFICATION Applicants for SRO license shall have 4 years of responsible power plant experience, of which at least 2 years shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shal_l be academic or related technical training.
Certifications t~at operator license applicants have learned to operate the controls shall be signed by the highest level of corporate management for plant operation.
These requirements shall be met on or after May 1, 1980.
(See March 28, 1980 letter).
Revise training programs to include training in heat transfer, fluid flow, thermodynamics, and plant transients.
This requirement shall be met by August 1, 1980.
(See March 28, 1980 letter).
Response
Requests for waivers from the requirements for an SRO license have been submitted in PSE&G letter of July 16, 1980 from F. W. Schneider to Stephen s. Hanauer.
Additional correspondence on our program to license both RO's and SRO's will be forwarded within the next few weeks.
Certification that operator license applicants have learned to operate the controls will be signed by F. w. Schneider, Vice President -
Production.
M P80 83/1 2
The licensed cperator (RO & SRO) training programs have been revised to in~lude heat transfer, flui~ flow, thermodynamics and plant transients.
I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS FOR LICENSED OPERATORS Training instcuctors who teach systems, integrated responses, transient and simulator cour.;es shall successfully C'omplete a SRO examinatiorc.
Applicants shetll be submitted by, August 1, 1980.
(See Instructors shall attend appropriate retraining programs that address, as a minimum, current operating history, problems and 1:hanges to procedures and administrative limitations.
In the event an instructor is a licensed SRO, his retraininq shall be the SRO requalification program.
Programs shall be initiated by May 1, *1980.
Response
(See All training instructors that teach systems, integrated reponses and transient courses presently hold valid SRO licenses.
Instructors participate in
~he applicable portions of the requalif ication prograrn.
I.A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMS Applicants for operator licenses will be required to grant permission tc the NRC to inform their facility management regarding t~~ results of examinations, M P80 83/l 3
Contents of the licensed operator requalif ication program shall be modified to include instruction in heat transfer fluid. flow, thermodynamics, and mitigation of accidents involving a degraded core.
These requirements shall be met by May 1, 1980.
(See March 28, 1980 letter).
The criteria for requiring a licensed individual to participate in accelerated requalification shall be modified to be consistent with the new passing grade for issuance of a license.
This requirement shall apply to all annual requalif ication examinations conducted after March 28, 1980.
(See March 28, 1980 letter).
Requalification programs shall be modified to require specific reactivity control manipulations.
Normal control manipulations, such as plant or reactor startups, must be performed.
Control manipulations during abnormal or emergency operations s_hall be walked through and evaluated by a member of the training staff.
An appropriate simulator may be used to satisfy the requirements for control manipulations.
This requirement shall be met by August 1, 1980.
(See March 28, 1980 letter).
Response
Applicants for operator licenses will grant permission to the NRC to inform the Company regarding the results of examinations.
The contents of the licensed operator requalification pro-gram have been modified to include instruction in heat transfer, fluid flow, thermodynamics and mitigation of accidents involving a degraded core.
M P80 83/l 4
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The accelerated requalif ication program has been modified to reflect the new passing grade for issuance of licenses.
Simulator Training is scheduled for each licensed operator on the basis of one week per year and is normally set up many months in advance.
Under most circumstances, simulator availability precludes scheduling training on short notice and therefore if the NRC wishes to observe or interact with operators during this training, arrangements should be made through the Manager -
Salem well in advance.
The requalification program has been modified to require specific reactivity control manipulations.
During the two year period of the RO/SRO license, each individual is required to perform *the approved manipulations listed below.
The items marked with an asterisk (*) must be performed annually.
Each individual is required to perform or participate in these manipulations, based on the availability of plant equipment and systems.
As many of these manipulations as practical will be performed at the facility; those manipulations not performed at the facility will be performed on a simulator.
Reactivity Control Manipulations
- l.
Plant or reactor startups; to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established.
- 2.
Plant shutdown.
M P80 83/l 5
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~3.
Manual control of steam generators and/or feedwater during startup and shutdown.
- 4.
B9ration and/or dilution power operation.
~s.
Any-significant (~ 10%) power changes in manual rod control.
- 6.
Any reactor power change of 10% or greater where load change is performed with EHC in manual.
~7.
Loss of coolant including:
- a.
significant steam generator leaks
- b.
inside and outside primary containment
- c.
large and small, including leak-rate determination
- d.
saturated reactor coolant response
- 8.
Loss of instrument air.
- 9.
Loss of electrical power (and/or degraded power sources).
~10. Loss of core coolant flow/nat_ural circulation.
- 12. Loss of service water.
- 13. Loss of shutdown cooling.
- 14. Loss of Component Cooling System or cooling to an individual component.
- 16. Loss of all feedwater (normal and emergency).
- 17. Loss of protective system channel.
- 18. Mispositioned control rod or rods (or rod drops).
- 19. Inability to drive control rods.
- 20. Fuel cladding failure or high activity in reactor coolant.
- 21. Turbine or generator trip.
M P80 83/l 6
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- 22. Malfunction of automatic control system(s) which affect reactivity.
- 23. Malfunction of reactor coolant pressure/volume control systems.
- 24. Reactor trip.
- 25. Main steam line break (inside or outside containment).
- 26. Nuclear instrumentation failure(s).
Reactor Operators must perform the manipulation to take credit.
Senior Reactor Operators may take credit for the manipulation if they direct or evaluate the control manipu-lations as they are performed.
The use of Technical Speci-fications is maximized during the manipulations.
Manipula-tlons are documented and become a part of the individual's record.
I.C.l SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION Analyze the design basis transients and accidents including single active failures and considering additional equipment failures and operator errors to identify appropriate and inappropriate operator actions.
Based on these analyses, revise, as necessary, emergency procedures and training.
This* requirement was intended to be completed in early 1980, however, some difficulty in completing this requirement has been experienced.
Clarification of the scope and revision of the schedule are being developed and will be issued by July 1980.
It is expected that this requirement will be coupled with Task I.C.9., Long-Term Upgrading of Procedures.
(See NUREG-0578, Sections 2.l.3b and 2.1.9, and letters of September 27 and November 9, 1979.
M P80 83/l 7
Response
Information add~essing this item was provided in PSE&G letters, R. L. Mittl to o. D. Parr, dated January 4, 1980 and March 28, 1980.
Additional information was provided for Salem 1 (Docket No. 50-272) in PSE&G letter, F. P. Librizzi to A. Schwencer, dated April 11, 1980~ which is also applicable tu Salem 2.
II.B.l REACTOR COOLANT SYSTEM VENTS Install reactor coolant system and reactor vessel head high-point vents that are remotely operator from the control room.
This requirement shall be met before January 1, 1981.
enclosure 4 to letters of September 27 and November 9, 1979).
Response
(See This requirement will be met by January 1, 1981, as des-cribed in PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980.
II.B.2 PLJ'l~NT SHIELDING Complete modifications to assure adequ~te access to vital areas and protection of safety equipmertt following an accident res\\11 ting in a degraded core.
This requirement shall be met by January 1, 1981.
(See NUREG-0578, Section 2.l.6b, and letter3 of September 27 and November 9, i979).
Response
This requirement will be met by January 1, 1981, as des-cribed in PSB&G letter, R. L. Mittl to A.
Schwenc~r, dated July l,*1960.
M P80 83/1 8 r :* 1
II.B.3 POST-ACCIDENT SAMPLING Complete corrective actions needed to provide the capability to promptly obtain and perform radioisotopic and chemical analysis of reactor coolant and containment atmosphere samples under degraded-core conditions without excessive exposure.
This requirement shall be met by January 1, 1981.
(See NUREG-0578, Section 2.l.8a, and letters of September 27 and November 9, 1979).
Response
An improved system for obtaining post-accident reactor l
coolant and containment atmosphere samples has been purchased.
This system was described in PSE&G lett~rs dated January 4, 1980 (Ro L. Mittl to O. D. Parr) and July 1, 1980 (R. L. Mittl to A. Schwencer).
The vendor (Sentry) is backlogged and cannot deliver the system until late December, 1980.
It is estimated that three months will be required to install the new system.
It is therefore expected that the system will be operational by April 1, 1981.
Interim methods for accomplishing this function were also described 'in the above referenced letters.
These methods will continue to be utilized until such time that the new system is operational.
II.D.l RELIEF *AND SAFETY VALVE TEST REQUIREMENTS Complete' tests to qualify the reactor coolant system relief and safety valves under expecte~ operating conditions for design basis transients and accidents.
M P80 83/l *9
This requirement shall be met by July 1, 1981.
(See NUREG-0578, Section 2.1.2, and letters of September 27 and November 9, 1979).
Response
This requirement will be met by July 1, 1981, as described in PSE&G letter, R. L. Mittl to o. D. Parr, dated March 28, 1980.
II.E.1.2 AUXILIARY FEEDWATER INITIATION AND INDICATION Upgrade, as necessary, automatic initiation of the auxiliary feedwater system and indication of auxil,iary feedwater flow to each steam generator to safety-grade quality.
This requirement shall be met by January 1, 1981.
(See NUREG-0578, Sections 2.l.7a and b, and letters of September 27 and November 9, 1979.
Response
Automatic initiation of the AFW System and indication of auxiliary feedwater flow to each steam generator meets safety grade requirements.
Supporting documentation was provided in the following PSE&G letters:
- 1.
November 1, 1979 (F. P. Librizzi to D. G. Eisenhut)w
- 2.
December 14, 1979 (F. P. Librizzi to A. Schwencer)w
- 3.
December 27, 1979 (F. P. Librizzi to A. Schwencer)w
- 4.
January 4, 1980 (R. L. Mittl to o. D. Parr)
- 5.
February 4, 1980 (F. P. Librizzi to A. Schwencer)*
- 6.
March 28, 1980 (R. L. Mittl to o. D. Parr)
- 7.
May 5, 1980 (F. P. Librizzi to A. Schwencer)*
- 8.
June 27, 1980 (R. L. Mittl too. D. Parr)
- 9.
July 1, 1980 (R. L. Mittl to o. D. Parr)
M P80 83/l 10
The letters marked with an asterisk (*) were submitted on the Salem 1 docket (50-272), but are also applicable to Salem 2.
11.E.4.l CONTAINMENT DEDICATED PENETRATION Install a containment isolation system for external recombiners or purge systems for* post-accident combustible gas control, if used, that is dedicated to that service only and meets the single-failure criterion.
This requirement shall be met before January 1, 1981.
(See NUREG-0578, Section 2.1.Sa and 2.1.Sc, and letters of September 27 and November 9, 1979.
Response
This requirement does not apply to Salem 2 (reference PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980).
II.F.l ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION Install continuous indication in the control room of th~
following parameters:
- a.
Containment pressure from minus 5 psig to three times the design pressure of concrete containments and four times the design pressure of steel containments;
- b.
Containment water level in PWR's from (1) the bottom to the top of the containment sump, and (2) the bottom of the containment to a level equivalent to 600,000 gallons of water; Containment water level iri BWR's from the bottom to 5 feet above the normal water of the suppression pool;
- c.
Containment atmosphere hydrogen concentration from 0 to 10 volume percent.
- d.
Containment radiation up to 108 Rad/hr;
- e.
Noble gas effluent from each potential release point from normal concentrations to 105 uCi/cc (Xe-133).
M P80 83/l 11
Provide capability to continuously sar,tple and perform onsi te analysis of the radionuclide and particulate effluent samples.
This instrwilentation shall meet the qualification,
, redundancy, testability and other design requirements of the proposed revision to Regulatory Guide 1.97.
This requirement shall be met by January 1, 1981.
{See NUREG-0578, Section 2.l.8b, and lettecs of September 27 and November 9, 1979).
Response
Containment pressure indication will be modified to meet the above requirements by January 1, 1981.
A system to measure containment water level has been purchased.
The vendor (DeLaval) has indicated a delivery date of Feb~uary, 1981, although expediting efforts may improve this date sornewh9t.
It is estimated that one month will be required to install the system, including a one week unit outage for final hookup.
It is therefore expected that the system will be operational following the first available outage of a2propriate duration after April 1, 1981.
The installed containment hydrogen analyzers have been recalibratPd for a range of 0 to 10 volume percent.
Two high range (107 Rad/Hr. gamma) radiation monitors* have been purctased for installation inside containment.
The vendor {Victoreen) has indicated a delivery date of M P80 83/l 12
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December, 1980.
It is estimated that installation of these monitors wi.1.1 be completed by April J_, 1981.
Final hookup will be made during the first available outage of appro-
- priate duration following April 1, 1981.
In the inte~im, the unit is equipped with a singl~ un-shielded detector with a range of 1C7 Rad/Hr. (gamma).
The noble gas effluent monitor has a range of 102 uCi/cc.
This unit if". to be expanded by the addition of an over-lapping instrument for provide a 105 uCi/cc range.
This instrument has been purchased from Vir:::toreen, the manu-facturer of the presently installed equipment~
At this time the vendor has been unable to provide a firm delivery date and completion of this item may extend beyond January 1, 1981.
Interim monitoring methods are addr~~sed in PSE&G letters dated Jana~ry 4, March 28 (R. L. Mittl to o. D. Parr) and July 1, 1980 (R. L. Mittl to A. Schwencer).
The capability to continuously sample and pe_rform onsi te analysis of effluent samples is presently available.
M P80 83/l 13
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II.F.2 INADEQUATE CORE COOLING INSTRUMENTS Install, if required, additional instruments or controls needed to supplement installed equipment in order to provide unambigrious, easy-to-interpret indication of inadequate core cooling.
This requirement shall be met by January 1, 1981.
(See NUREG-0578, Section 2.l.3b, and letters of September 27 and November 9, 1979).
Response
It is our intent to install a system to indicate reactor vessel water level, as described in PSE&G letter, R. L.
Mittl to O. D. Parr, dated March 28, 1980.
The system is under development by Westinghouse and the Westinghouse Owners Group.
At this point, a firm delivery date _has not been indicated by Westinghouse, although it is estimated to be in the March-April, 1981 time frame.
This level device requires removal of the reactor vessel head for installation and, as such, it will be installed during the first refueling outage.
III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES Provide radiation monitoring and ventilation systems, including particulate and charcoal filters, and otherwise increase the radiation protection to the onsite technical support center to assure that personnel in the center will not receive doses in excess of 5 rem to the whole body or 30 rem to the thyroid for the duration of the accident.
Provide direct display of plant safety system parameters and call up display of radiological parameters.
For the near-site emergency operations facility, provide shielding against direct radiation, ventilation isolation capability, dedicated communications with the onsite technical support center and direct display of radiological and meteorological parameters.
M P80 83/l 14
This requireme:nt shall be met by Janua1~y 1, 1981, although the safety parameter information requirements will be staged over a longer period of time.
(See NUREG-0578, Section 2.2.2b and 2.~~.2c and letters of Sept(~mber 27 and November 9, 1979 and April 25, 1980).
Response.
The Technical Support Center (TSC) is described in PSE&G letter, R. L. Mittl too. D. Parr, dated January 4, 1980, along with plans for upgrading ~he TSC to meet the above requirements.
It is expected at this time that the upgrading effort will be complete by January 1, 1981 with the exception of a common data base di.splay of plant safety system parameters.
In this regard, it is our understanding that the require-ments stateo above have been revised in NUREG-0696, "Functional Cciteria for Emergency Reeponse Facilities",
which is to r.e issued for comment in August, 1980.
It is also our understanding that a revised schedule for imple-mentation of a total requirements package is also und~r development.
The near-site Emergency Operations Facility (EOF) is described in PSE&G letter, F. W. Schneider to S. A. Varga, dated July 3, 1980.
The EOF is being activated at our Quinton Training Facility, approximately 8 miles from the Salem Station, and will be equipped as required to comply M P80 83/l 15
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'.I with the requirements set forth in NUREG-0696, in accordance with the implementation schedules under development by the NRG.
III.D.3.3 IN-PLANT RADIATION MONITORING Provide the equipment, training, and procedures to accurately measure the radioiodine concentration in areas within the plant where plant personnel may be present during an accident.
This requirement shall be met before January 1, 1981.
(See NUREG-0578, Section 2.l.8c, and letters of September 27 and November 9, 1979).
Response
This requirement will be met by January 1, 1981, as des-cribed in PSE&G letter, R. L. Mitt! to o. D. Parr, dated January 4, 1980.
M P80 83/l 16
A
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Attachment One I
' I Shift Technical Advisqr Training Program
. I The Public Service Electric and Gas Company Shift Technical Advisor Training Course is intended to provide engineers with sufficient knowledge to prepare themselves to assume the duties of a Shift Technical Advisor.
The level of knowledge and degree of competence that this course provides should be sufficient to permit th~ shift technical advisor to fil_l the following roles:
- An Accident Assessment Function to evaluate various accidents and abnonnal situations to verify that they are progressing according to the expected course of events.* If the sequence is deviating from the anticipated, the Shift Technical Advisor must be able to recognize this fact and recommend the. appropriate operator action to cor.rect the situation.
An Operating Experience Assessment Function to evaluate applicable operating experiences from similar plants and to apply these evaluations to the upgrading of the plant safety.
This program is not-intended to qualify a person to license or be equivalent to a Senior Reactor Operator. This training will be accomplished by fonnal cl ass room/simulator training by experienced Westinghouse training engineers.
This fonnal training will encompass the following areas:
- Nuclear Reactor Theory
- NT'R Training Large PWR Core Physics
- Health Physics and Chemistry
- Detailed Plant Lecture Series o
Reactor Engineer Training
- Advanced Training on Simulator A detailed listing of the complete traini~g program is provided in Addendum A.
The enclosed course will be conducted at the Public Service Gas and E1ectr1c Company facilities with the exception of the Advanced Simulator and NTR training which will be conducted at the w,estii:ighouse Nuclear Training Center at Zion, Illinois and the Reactor Engineer Training which will be conducted at NSD headquarters in Pittsourgh, PA.
i I
I Weeks 1 - 6 Weeks 7 & 8 Weeks 9 & 10 Weeks 11 - 25 Weeks 26 & 27 Weeks 28 - 34 Weeks 35 & 36
'D
- 1.
TENTATIVE RECOMMENDED COURSE SCHEDULE March 10 Ap ri 1.l 8, 19 80 April 21
- ~ May 2, 1980 May 19 May 30, 1980 June 2 Sept. 12, 1980 September 14 - September.20, 1980 1st Group Septenber 21 - September 27. 1980 2nd Group
- September 29 - Noveniler 14, 1980 November 17
~ November 21, 1980 1st Group November 24
- November 28, 1980 2nd Group
Weeks 1 - 6 Weeks 7 & 8 Weeks 9 & 10 Weeks 11 - 25
- Weeks 26 & 27 Weeks 28 - 34
. Weeks 35 & 36 2
SHIFT TECHNICAL ADVISOR Reactor Th~ory Reactor ~inetics PWR.Core Physics Heat Transfer Fluid Flow Thennodynami cs. *.
Health Physics Pl ant Chemistry Detailed Plant Lecture Series NTR Operations (WNTC, Zion,Illinois)
- I Reactor Engineer Training (Pittsburgh)
Simulator, Advanced Training (WNTC, Zion,Illinois)
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- -*---~---------**- -----------~..... ---*-*--*--*.::... -.:. '
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- (
TYPICAL ON SITE INSTRUCTION SCHEDULE
-~
.\\I:,.
8:00 -
10:00
. self Study 10:00 -
12:00
- Site Instructor Administer W Exams Monday 12:00 -
12:45 Lunch 12:45 -
1:45 Exam Review l:45 -
4:45 Instruction
- s*:oo -
12:00 Instructfon
.* 12:00 -
12:45 Lunch Tuesday 12 :45 -
2:45 Instruction 2:45 -
4: 45 Self study/Tutoring 8:00 -
12:00 Instruction 12:00 -
12:45 Lunch Wednesday 12:45 -
2:45 Instruction 2:45 -
4:45 Self Study/Tutoring 8:00 -
12:00 Instruction 12:00 -
12:45 Lunch Thursday l2:45 -
2:45 Self Study Fri day 8:00 -
12:00 Self Study 12:,00 -
12:45 Lunch 12: 45 -. 4:45 Self Study I
~
Weeks l & Z 4
- I
- Detailed Course Outline
~
Addendun A Basic PWR Orientation 1.'. Introduction to Fl u1 d Systems
- 1. Primary Plant Systems
- 2. Secondary Pl ant Systems
- b. Introduction to Control and Protection Sys,tems.
- Nuclear* Reactor Theory Atom1 c Physics Review
- a. Molecular and atomic structure b., Nucleus, elemental particles -- electrons, protons and neutrons.
- c. *Atomic number and the periodic table
- d. Energy levels
- e. Ionization and coulonb forces
- Neutron Physics
- a. Nuclear Reactions b *. Neutron interaction with matter (abso1'1ltion, capture, fission. scattering).
- c. Microscopic and macroscopic neutron cross-sections.
- d. Neutron energy dependence on cross-sections (1/v and resonance.)
- e.
- The fission process
- f. Neutron flux
- g. Neutron spatial and energy distribution
- h. Reaction rate
- 1. Heutron diffusion and slowing down Reactor Physics
- a. Neutron chain reaction
- b. Neutron multiplication factor and criticality concepts c:. Infinite multiplication factor (four* factor for111ula)
- d. Effective multiplication factor (six factor fonnula)
- e. Neutron balance problems
- f. Buckling.-- georetric and material
- g. Neutron leakage and reflector savings h..
- Critical size
- 1. Neutron balance flux shape in a critical reflected rea*
e s
Weeks 3 & 4
~eactar K1net1cs
- I
.r::
Weeks 5 & 6
- -* ---------~-------- - --
- a.
- b.
c:.
d..
- e.
- f.
- g.
- h.
- f.
j..
- k.
Characteristic time parameters for neutrons Time dependent neutron flux considering only prompt neutrons.
Delayed neutrons Relationship of delayed and prompt neutrons.to reactor central.
Neutron flux transients Defin1 tion and LD"li ts of reactivity Relationship between reactivity and reactor perio~
(Inhour Equation).
Transients and stable reactor period Exa~le problems involving neutron flux and reactor period.
Effective ~layed neutron fraction Problem session Subcriti cal Reactor Theory
- a. Neutron sourc~s
- b. Subcrltical multiplication and Keff relations~ip
- c. Approach to critical techniques Introduction to Reactor Control
- a. General control. concepts
- b. Reactivity control in a reactor -- rods, rroderator, reflector and fixed poi sons
- c. Reserve reactivity requirements and inherent -
reactivity changes -- buniup, varying poisons,
- tenrtJeratures *
- . cf..
Control rad rr..;:.terials, configuration and classificatior
- e. Control rod differential and integral reactivity worth fe Neutron flux spa ti al distributions
. g. Control rod worth curves and example calculations brge PWR Core Physics
- Control Rod Reactivity Effects
- a. Introduction to core design and excess reactivity loading.
- b.
- Poi son bui 1 dup
- c~ Ft.el bumup
- d. Control rod flux distribution affects
- e.
Rod worth variations
- f. Plant curves
- g.
P~bl~m s7ssions
---~ -** ---** -*-
- ---*-*-*- -*---~ ----*--=:...:::.
- - ---*---- _t ___.::.,. ____.. _......:..... _____
--~----*
*--*~-*
Weeks 5 & 6 (cont.)
6
. L.
Soluble Boron Control
- a.
Purpose
- b. Boron effects on rmderator temperature *coefficient.
- c. Boron worth curves
. d.. Soron concentration changes with core 11 fe
- e. Problem session Moderator Temperature Effects *
- a. Temperature coefficient dianges with core bumup
- b. Factors affecting temperature coefficient
- c. Safety consi derat~ ons
- d. Mo*oorator coefficient curves
- e. Problem session
' F1.el Temperature Affects a.;.
Fuel compos 1 ti on
- b. Doppler coefficient c:.
Fuel heat transfer
- d. *p1ant curves
- e.
- Problem sessions Total PO'tier Defect
- a. Coefficients wh i ch make up the power defect
- b. Changes in power de'fect due to:
ll Fuel burnup Z)
- c. *Use of pl ant curves
- d. Problem session Poison Effects
- a. Xenon
- b. Samarium
- c. Poison rod effects d.. Power hi story effects on Xenon
- e. Plant curves Estimated critical con di tfons and shutdown margin cal c:ulat1o utilizing curves for problem session.
Weeks 7 & 8 7
Core Operational Concepts
- a. Spat1 al flux cons1 de ration
- b. *eoL to EOL core characteristic change PWR Thenrodynami cs - Fundamentals
- a. Introduction to Power*Plant Funct1on
- b...Def1 nf tions
- c. The First Law of Thenrodynami cs
- 1. lhe General Energy Equation
- 2. Enthalpy
- 3. Examples of Steady State, Steady Flow Systems
- 4. Bernoulli's Equation
- 5. Iceal Gas Behavior
- d. Heat Engines
- l.
- Thennal Efficiency
- 2. Reversibility and the Second Law of ihenoodynamics 3.,, Entropy
- 4. The Ca mot Cycle PWR Thenrodynami cs - Appl 1ed
- a. Steam Power Engines 19! Properties of Water
- 2. Steam Tables 3". The Rankine Cycle
- 4. Components of the PWR Cycle tr.
PWR Power Pl ant
- l. The Real Rankine Cycle
- 2. The Pressurizer
- 3. Calorirretric Heat Transfer
- a. Types of Heat Transfer
- b. Conduction Through a Pure Substance
- 1. Conduction Through Composite Bodies
- 2. Conduction in Cylindrical Georretry
- 3. Conduction with Internal Heat Generation
~~*==~===========:;;.= __________________ _:__ _ __:____:*_.. ---...... :
Weeks 9 & 10 8
- c. Convection 1~ Forced Convection
- z. The Overall Coefficient
- d. Heat Exchangers
- e. Heat Transfer with Change of Phase
- l. The Boiling Curve
- 2. DNB
- 3. The Con~nser
- 4. Fouling
- f. Natural Circulation
. g. Plant Transient Response
- 1. Steam Generator Shrank and Swell 2.. Primary System Temperature Changes Fluid Flow
- a. Fl ui d Fl ow
- 1. Vf s cos i ty
- 2. Types of Flow
- b. Fluid Flow Through Nozzles and Orifices
- c. Nozzle Design
- 1. Nozzle Fl ow Rate
- d. Venturi' s
- e. Orifices
- f. Punp and Fl ow Characteristics 1.. Centrif!Jgal Pumps
- 2. Positive Displacerrent Pumps
- 3. Pt.mp Combinations
- 4. Friction and Head Losses in a System
- 5. Cav1 tat ion Heal th Physics and Chemistry
- I Introduction to Sample Radiation Safety Calculations
- 1. Types of radiation and interaction with matter
.b. Biological damage from radiation
- c. Radiation quantities and units
- d. Radiation safety tenns and definitions
- e. Natural radiation background and its sources
---~---*.:. ---* ---- '. - _.,.:. __. __ ---*--****-**-:----~----.:. _________.._ _____
- __.
"*'....... \\............ -.. *:...__,.
Weeks 9 & 10 (cont.)
9 B1o1og1cal Effects
- a. Bf ology of the human body and terminology
- I b.. Radiation effects on cells and human a1'"9ans and systems c:.
~ntemal and external radiation exposure and effects
- d. Chronic and acute exposures e.. Scmati c and genetic effects
- f. Acute overexposure syndrome Principles of Radiation Protection
- a. General nethods of limiting radiation exposure
- b. Standards for radiation protection (10CFR20)
- c. Maximt.m pennissible dose an!! established working limits (external and internal exposures)
- d. Radiation rreasurerrents and rooni t.Oring
- e. Contamination and its control f".
~adioactive material handling and disposal
. g. Special radiation safety problems in a reactor facility
- h. Radiation shielding
{ *. Procedures utilized in radiation safety practices
- j. Protective clothing and errergency equiprrent I
Principles of Radiation' Detection
- a. General radiation detection principles
~
. b. Cl assi fi cation of radi atien detectors
- c. The gas filled radiation *detector
- d. The four conm:m datectors
- e. Semi-conductor detectors Radiation Monitoring and Dosirretry
- a. Radiation monitoring and dosimetry systems b'.. Personnel and area m:mi taring c:.... Environrrental rronitoring (Radiation surveys; air, water surface contamination)
- d. Typical radiation safety rooni tori ng instrtm"ents
. ~... -
. -.. --~*---* *--.
Weeks 9 & 10 (cont.)
Week 11 Week 12 Week 13 Wee!( *14
~eek 15 0
10.
Rad1 at1on Detector Usage 1 n P1 ant
._a. *.Portable rooni tors
.b. Radiation rrcnitoring system
..... _t} __ Area roni tors
- 2).. Process rooni tors Chemistry --
Primary and Secondary System5
- a. Primary chemistry
--*. l) Types of corrosion and control Z) pH concept
- 3) Primary syste*m* chemistry specifications and controls
- 4) Radiation chemistry
- b. Secondary chemistry
- 1) Steam generator all vol a tile treatnent specification Z) Condensate. and feedwater specifications Detailed Pl ant Lecture Seri es
- b. Reactor Vessel, Internals and Fuel
. c-. Reactor Cool ant Pumps and Steam Gen'=rator
- a. Chemical and Vo1l.IITe Control System
- b. Makeup Water System
- c. Containment Structure and Systems
- a... Residual Heat Rerroval System
- b. Safety Injection System
- c. Containment Spray System a-.
Iodine Reirova1 System
- b. Containrrent Isolation System
- c. Component Cooling Water System
.d. Auxiliary Feed.iiater System
- a. Exc0re Nuclear Instrurrentation System
- b. Incore Instrurrentation System
- c. Ful 1 Length Rod Control Sys tern
Week 16 111:-!..,;,
- Week 1.7 Week 18 e
Week 19 Week ZO Week 21 Week*zz
- -~---'-----*----------- _*.....,: __ _
11 t* Rod Position Indicating System b*;- Rod Insertion Limit
- c. Pressurizer Pressure ana Level Control Systems cl.:
Steam Dimtp System
- a. Steam Generator Level Control System tt;;:*= Reactor Protection System
- a. Process Control Sys tern Logic Diagrams
- b. Protection/Safeguards Logic Diagrams c..
I&C Systems Integration
- a. Hain Steam System
- b. Auxiliary Steam System
- c. Condensate System
<4._ Feed System
- a. Electrical *Distribution b.. Electrical Controls c:. Diesel Generator
- d. Safeguards Sequence a.. Fuel Handling
- b. Spent Fuel Pool
- c. Spent Fuel Pool Cooling d.. Service Water System
- e. Ventilation System System
- a. Instrunent and Service Air System
- b. Fire Protection System
- c. Steam Generator Slowdown System
- d. Sam;Jling System
-~ ------,----*---~-----.,--------- ------*------- *--*-*-* ******-----**---~-- ---------*....
Week 23 Week
- 24.
Weeks 25 Weeks 26 & 27 Week 28.,
- e
- .*.--c... --"-*
12
- I
- a. Turbine Generator
~
- b. Turbine Generator Support Systems
- c. Electro-Hydraulic Control System d *. Voltage Regulator
- a. Liquid Waste System
- b. Gaseous Waste Syst~m
- c. Solid Waste System
- a..Plant Computer
- b. Technical Specifications NTR Operations Control & Protection Systems Introduction
- a.
Core 1 imi ts
- b.
NI trips C.
LOF Trips
- d.
Setpoints Solid State Protection System IEEE Standards Design Cri te ri a Safeguards Panel.
Technical Specifications PLS Document Accident Analysis
- a. Increase in Secondary Heat Removal
- b. Decrease in Secondary heat Removal Accident Analysis
- a. Mass/Energy Releases (Secondary)
.b. Loss of Flow/Locked rotor.
., -; ~.......
Week 29 Week 30 Week 31 Week 32
. -- *-**-~-- ***------*---*1*"3**~:.. _____,_.
-*--~-*-** **---*--*-***--*-'-***-
e Accident Analysis
- a. Reactivity Excursions
- b. Overpressure Protection/AM Accident Analysis ao Loss of Coolant bo Radiological Assessments Recent Technological Development Gross Failed Fuel System Rx Vessel and Internals Fuel Assemb1y*Mech. Design Core Component Mech. Design Nuclear Fuel Design -
Core Design Characteristics Nuclear*oesign Report Core Control and Load Follow Hybrid Computer Tour
'I Initial Rx Startup Program thru Hot Fun.ctionals Reactivity Computer Fuel Shipping and Accountability Fuel Handling Initial Fuel Operations and Equipment Refueling Initial Rx Startup Test Program Phase III Detailed Discussion of Plant Startup Procedures Physics Reduction and Data Measurement Incore Codes Plant Computer and Load Follow Programs Plant Physics/Computer Programs
- ----~--- - -
Week 33 Week 34 Weeks 35 & 36.
I
_....:.__--"--'-'---'-~-----...;_-- -- **-----C..*,_.. ____,_: __ _
14 e*
Inser~1ce Inspection NRC Regulations Plant Licensing Process Relo~d Licensing
- Licensing Issues Plant Startup and Power Operati~ns Fuel Cycle Economics Fuel Management Design Fuel Management
. Fuel Cycle Flexibility Fuel Perfonnance Public Acceptance Nuclear.Industry Discussion
- I Simulator Advanced Training (WNTC)
- a. Familiarization and Basic Operatic~~
be Reactor Startup
- c. Pl ant Heatup and Cool down
- d.
Power Operations
- e. Selected Accident Recognition