ML20032A681
| ML20032A681 | |
| Person / Time | |
|---|---|
| Site: | Armed Forces Radiobiology Research Institute |
| Issue date: | 09/30/1981 |
| From: | DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY |
| To: | |
| Shared Package | |
| ML20032A680 | List: |
| References | |
| NUDOCS 8111020178 | |
| Download: ML20032A681 (48) | |
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- Armed Forces Radiobiolcsry Research Institute AFRRI R-84 Annual Report October 1981 (Report Period: 1 October 1980 - 30 September 1981)
Part A - Changes in the Facilities and Procedures This section specifies actions taken during the report period that reflect changes to the existing Final Safeguards Report (FSR) not previously reported to USNRC. AFRRI is currently involved in a contested relicensing procedure. Under the relicensing effort, AFRRI has submitted an updated Safety Analysis Report (SAR) to USNRC. This SAR has been reviewed by the NRC staff; questions have been directed to and answered by AFRRI concerning the SAR. Also in connection with the AFRRI-TRIG A reactor relicensing, Technical Specifications, an Environmental Report, Environmental Impact Appraisal Data, an Emergency Plan, Reactor Operator Requalification Program, and Physical #ecurity Plan documents have been submitted to USNRC. Of these reports, only the AFRRI Physical Security Plan and the Reactor Operator Requalification Program have been reviewed and approved by USNRC and implemented by AFRRI. The remaining documents are presently under review as part of the AFRRI-TRIGA reactor relicensing effort. Since the SAR submitted by AFRRI has not been approved and, therefore, cannot be implemented, the actions taken during this reporting period are categorized on a chapter basis as listed in the existing FSR.
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Chapter I - Site A new animal facility, designated Building #47, was completed and accepted by AFRRI during this reporting period. This animal facility adjoins the AFRRI complex on its eastern side and is comprised of three main floors. Occupancy of this new animal facility recently began and is expected to be completed during 1982 after the cornpletion of on-going work on some environmental support systems. As a result, no permanent changes have, as yet, been finalized and implemented for the Emergency Evacuation Plan or the Environmental Monitoring Plan. These plans are currently being updated for subsequent implementation. In additior,, during this reporting period the new National Naval Medical Center (NNMC) Hospital, designated as Building #9, and a new Nursing Tower, designated as Building #10, were completed and placed in service on the grounds of NNMC. The figure in Attachment 1 illustrates the site plan with compass sectors and d
distances from the AFRRI stack shown.
Chapter II - Facility Floor plans for all levels of the AFRRI complex including the new animal facility addition (Building #47), identified under Chapter I - Site above, are provided as.
Chapter til - Reactor During the annual reactor shutdown maintenance
- period, performed in January / February 1981, the transient control rod was replaced with a new transient 2
control rod of the same design except for the too connecting fixture. The transient connecting rod, as a result, had to be modified slightly at its bottom end to securely accept and hold the new transient control rod. With the present core fuel loading, this new transient control rod has ar. integral worth of $3.35 (2. 345%k/k) at the mid-pool position (i.e. for infinite water reflection).
A back-up reactor make-up water system has been placed in service as an addition to the system.
The addition is a demineralizer which utilizes in-line mixed-bed resin canisters. Make-up water can now be supplied to the n actor tank via gravity flow by either a water distillation unit or the mixed-bed demineralizer unit just installed. The mixed-bed demineralizer unit utilizes city water as its feed supply with a manual gate valve and flow orifice installed upstream of the unit. Downstream of the unit M a conductivity cell and coupling which is kept "open" except when the unit is to be used.
The output from the mixed-bed demineralizer and distillation make-up water units join at a " Tee" with gate valve isolation provided. Piping from the two make-up water units to the reactor pool has been replaced and the piping terminus at the reactor tank location is now physically above the surface of the reactor pool to isolate the make-up system from the reactor pool and preclude any potential for backflow or syphoning.
No other changes to the reactor or reactor systems were made which affected the use or intent of any system and no changes to the Technical Specifications were made. A new set of Technical Specifications were submitted to USNRC as part of the AFRRI reactor relicensing effort; these have not yet been approved cr implemented.
Following is a listing of malfunctions that occurred during the reporting peciod together with the action (s) taken.
3
Malfunction: During a normal reactor startup, the steady-state timer sequencing and elapsed tirne scram signal for steady-state operational scram was determined to be malfunctioning.
The timer skipped nuinbers on the tenth-of-a-second digit during counting and often failed to give a scram signal at the set elapsed time.
Action Taken: Operators were notified to switch to manual timing for operations.
He steady-state timer was scheduled for replacement. The steady-state timer is not required for operations.
Malfunction:
During a weekly nuclear instrumentation check, high flux safety channel #2 was inadvertently saturated during NVT zero power pulse checkout (electronic testing of channel) resulting in a disabled amplifier. During the NV (Peak) checkout performed just prior to the NVT test, the channe! was operational and correct.
Action Taken: The linear amplifier was replaced and the channel was calibrated.
Malfunction: The %ermistor for bulk water temperature was found to be inoperable at start of day.
Checkout of thermistor with multi-meter indicated no electrical continuity.
Action Taken: Thermistor probe was replaced and the system calibrated.
Malfunction: During a 1.0 Mw(t) steady-state power run with both high flux safety channels #1 and #2 reading a constant 100%, a percent power scram occurred on high flux safety channel #1 due to an overly conservative setting on the bistable trip (electronic scram) for high flux safety channel #1.
4
Action Taken: Bistable trip setting for high flux safety channel #1 adjusted to yield scram signal at 110% of authorized power.
Malfunction:
An incorrect readout was displaying on remote water conductivity monitor unit in reactor control room.
t Action Taken: Water conductivity values were read from master unit in Equipment Room #3152. Slave unit in reactor control room repaired.
Malfunction: Unplanned manual scram due to stack exhaust fan power failure.
Action Taken: Power was restorea.
Malfunction: Overload in core dolly drive motor caused wiring connections in control rod drive housing for core dolly drive to open, resulting in loss of power to core dolly drive.
Action Taken:
The core dolly drive motor was removed, cleaned and the bearings were replaced. The wiring was repaired and the motor reinstalled.
Malfunction:
A core insoection, conducted as a result of a suspect k-excess measurement, revealed that an F-ring element was cocked.
Action Taken: USNRC, Region I was notified. USNRC, Region I advised AFRRI that this item was not reportable under either 10 CFR or AFRRI's Technical Specifications.
The cocked f;el element was removed from the core, visually inspected for damage, and 5
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measured for length and bow.
No discernable damage was detected by the visual inspection and both the length and bow measurements were well within Technical Specification limits. The subject fuel element was reloaded into the core and checked for proper seating; k-excess was measured (giving the same value as previously measured),
and operations were resumed. An informational letter was sent to USNRC, Region I.
Malfunction: During the zero power,ulse checkout portion of the normal reactor startup checklist procedure, the transient rod " air" light remained "on" forN 3 seconds af ter the transient rod air solenoid vented (i.e, af ter having received c. scram signal from the pulse timer). Transient rod drop action appeared to be retarded based upon visual observation; transient rod was fully inserting but "down" microswitch was not actuating.
Action Taken: Transient rod anvil was removed, inspected and cleaned - no problems identified. Transient rod drive was removed and the cylinder was inspected and cleaned -
slight amount of dirt and oil had accumulated. Piston action was bench checked manually and transient rod drive was reinstalled. USNRC, Region I was notified; advised AFRRI that no report was required. Several successful zero power pulses were subsequently performed.
Malfunction: K-excess was inadvertently not performed prior to first reactor power operation of the day.
Action Taken: USNRC, Region I was notified. K-excess was measured and agreed within $.01 from k-excess on previous day. An administrative procedure was instituted internally requiring that the SRO who reviews, approves, and signs the start-up checklist 6
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for a particular day must also initial the k-excess for that day. An informational letter was sent to USNRC, Region 1.
4 Malfunction:
Stack gas monitor was found to be inoperable at start of day due to seizure of pump supplying stack gas flow to proportional counter.
I Action Taken: Peactor start-up was postponed pending pump repair or replacement.
Pump was subsequently repaired.
b Malfunction: Early manual scram by operator was required during power run due to stack exhaust fan failure alarm caused by building fire alarm testing.
l' Action Taken: Event was logged as an unplanned scram. AFRRI Logistics staff was notified that alarm testing should be performed only aScr receiving approval of reactor operator / staff.
For the above listed events, all safety systems performed their intended functions.
i Chaoter IV - Exoerimental Facilities No changes have been made to the experimental facilities.
Chaoter V - Nuclear Analysis No changes have been made that affect the nuclear analysis.
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1 Chapter VI-Hazard Analysis Although no changes have been made that would affect the Hazard Analysis, a new i
Hazard Analysis, retitled Safety Analysis, has been prepared and submitted to USNRC within the updated SAR as part of the AFRRI-TRIGA reactor relicensing effort.
1 i
Chapter VII - Organization i
The administrative structure within AFRRI has changed slightly. In particular, the i
Scientific Stpport Department (SSD) has been reclassified as the Radiological Sciences j
Department (RSD). A list of current key AFRRI personnel follows:
i i
Director - CAPT Paul E. Tyler, MC, USN I
Deputy Director - COL Bobby R. Adcock, MSC, USA i
Chairman, Radiological Sciences Department - Lt Col James 3. Conklin, USAF, MC Chief, Radiation Sources Division - M A3 Ronald R. Smoker, EN, USA 1
The current Reactor Staff is:
4 1
Physicist-In-Charge - M A3 Ponald R. Smoker, SRO Chief Supervisory Operator - Mr. Marcus L. Moore, SRO Reactor Operator - Capt Joseph A. Sholtis, Jr., SRO Reactor Operate - CPT Leonard A. Alt, SRO l
Reactor Operator - SFC Harry H. Spence, SRO i
Reactor Operator - SFC Lorne H. Vernon, RO Reactor Operator Trainee - Ms. Angela Melcher g
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The current membership of the Reactor and Radiation Facility Safety Committee is:
COL Bobby R. Adcock, Chairman - AFRRI Lt Col James 3. Conklin - AFRRI CDR L.3. Ihuchler - NNMC M A3 Ronald R. Smoker - AFRRI M A3 Z.N. Church - ?.FRRI Capt Joseph A. Sholtis - AFRRI LT Kenneth P. Ferlic - AFRRI Mr. W.L. Gieseler - Consultant Mr. T.G. Hobbs - National Bureau of Standards Dr. Frank 3. Munno - University of Maryland Dr. N.K. Chawla - AFRRI Vr. 3.N. Stone - Naval Research Laboratory Mrs. B.S. Markovich - Recorder Chapter VIII-Procedures i
Minor procedural changes have been made to AFRRI's internal Radiation Sources Division Instructions, RSD 5-1 through RSD 5-8, which pertain to the AFRRI reactor.
Copies of these current RSD Instructions are provided as Attachment 3.
.i
Part B - Tests and Experiments 1.
No tests or experiments were performed during this reporting period that exceeded the limits stated in the Technical Specifications.
2.
All experiments performed during the reporting period were in accordance with Technical Specifications and authorized as either " Routine" or "Special."
3.
The reactor was utilized at the various experimental facilities such that the total power production was 30995.2 kW-hr, broken down as follows:
1 a.
In Steady-State: 30625.8 kW-br.
F b.
In Pulse: 369.4 kW-hr.
4.
Experimental workload for FY 82 is anticipated to be in the range of 25 MW-br to 50 MW-hr of total burn-up.
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RADIATION SOURCES DIVISION 27 March 1981 g
INSTRUCTION NUMBER 5-1 FUNCTIONS AND ORGANIZATION OF THE REACTOR BRANCH 1.
Purpose. To set forth the functions and organization of the Reactor Branch IAW AFRRI Instruction 5100.1, "AFRRI Organization," and to define the duties of the personnel.
2.
Applicability. The provisions of this instruction are applicable to the Reactor Branch staff.
3.
Cancellation.
RSD Instruction 5-1, " Functions and Organization of the Reactor Branch," 2 April 1980, is hereby cancelled.
4.
Functions.
Operates, calibrates and maintains the AFRRI-TRIGA Reactor and associated a.
systems in compliance with appropriate regulations.
b.
Provides radiation exposures in the various experimental facilities in the direct support of the AFRRI research mission.
Supplies instruction and technical assistance to the AFRRI staff, relating to c.
reactor operations, utilization, instrumentation and modification.
i d.
Conducts research to analyze reactor performance, develop and advance its capabilities, and evolve new applications.
5.
Organization. The block diagram presented in inclosure 1 defines the organizational relationships among the positions. The duties and qualifications for each position are defined in the following paragraphs:
a.
Physicist-In-Charge (PIC) of the Reactor Branch.
(1) The PIC will possess an NRC Senior Operator License for the AFRRI-TRIGA Reactor under the provisions of 10 CFR-50 and 10 CFR-55.
(2) The PIC is responsible to the Head, Radiation Sources Division, for operational, technical, administrative, and safety matters pertaining to the utilization of the reactor in support of AFRRI approved research programs.
(3) The PIC is directly responsible to the Head, Radiation Sources Division, for insuring compliance with the NRC Facility License (R-84) including all amendments, other NRC regulations, and AFRRI instructions.
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kADIATION SOURCES DIVISION INSTRUCTION NUMBER 5-1
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(4) The PIC will resolve confhcts between the various AFRRI user requirements concerning scheduling of the reactor in the event that the Chief Supervisory Operator (CSO) who schedules the routine use of the reactor, is unable to resolve the conflict. The PIC will contact the Head, Radiation Sources Division, in the event of a conflict involving a question of research priorities.
(5) The PIC will sign all Reactor Use Requests (RURs) thereby giving his official approval for operation of the reactor as indicated.
(6) The PIC, as the Head of the Reactor Branch, will participate in special development projects as authorized or directed by the Head, Radiation Sources Division.
(7) The PIC is responsible for conducting the training of Reactor Operators (RO)in preparation for NRC certification.
(8) The PIC is also the principal administrative officer of the Branch and as such will maintain the files and perieiically review / update operating procedures.
(9) The PIC is responsible for the completion of all required reports.
(10) During the absence of the PIC for more than one day, the Director will designate, orally or in writing, a qualified individual to act as the P!C.
In all other absences, the CSO is authorized to act on behalf, and instead of the PIC.
(11) The PIC may deviate from the provisions of the RSD instructions if such a
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deviation does not violate the AFRRI-TRIGA Reactor License, NRC regulations, applicable AFRRI safety directives, and he has received prior approval from the Head, Radiation Sources Division of each and all such deviations.
b.
Chief Supervisory Operator (CSO).
(1) The CSO will possess an NRC Senior Reactor Operator License under the i
provisions of 10 CFR-50 and 10 CFR-55.
(2) The CSO is responsible to the PIC for the efficient, safe operation of the reactor on a daily routine, and for its maintenance.
(3) After tentatively scheduling reactor time the CSO will forward to the PIC RURs submitted by investigators.
l (4) The CSO will schedule reactor time as requested by approved users for i
approved research. In the event of an unresolved scheduling conflict he will refer the matter to the PIC.
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(5) The CSO will schedule the activities of the assigned ROs.
(6) The CSO will conduct training on the operations / maintenance of the reactor system, and will conduct the recurring training of licensed operators IAW NRC regulations.
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RADIATION SOURCES DIVISION INSTRUCTION NUMBER 5-1 (7) The CSO will direct the supply activities of the Reactor Branch. The supply activities include requisitioning equipment and expendable supplies, maintaining supply records, and equipment inventories.
(3) The CSO is primarily responsible for execution of the Reactor Maintenance Program as prescribed in RSD Instruction 5-4.
c.
Reactor Operator (RO).
(1) The RO will be directly responsible for the safe and legal operation of the reactor in compliance with the Reactor License (R-84),10 CFR-20,10 CFR-50,10 CFR-55, and RSD instructions.
(2) The RO will perform duties as assigned by the CSO.
(3) The RO will possess either an NRC Senior Reactor Operator (SRO) or an NRC Reactor Operator License for the AFRRI-TRIGA Reactor, under the provisions of 10 CFR-50 and 20 CFR-55.
d.-
Reactor _Orerator Trainee.
(1) High School or equivalent.
/
(2) Be in training for NRC operator examination.
L-6.
Performance Ratings.
a.
The military performance ratings will be accomplished IAW AFRRI Instruction 1600.1," Performance Ratings and Schemes."
b.
The civilian performance ratings will be accomplished IAW DNA Instruction 1434.1, "Perf ormance Evaluation."
7.
Radiation Sources Division Instructions, a.
The Reactor Branch will function and operate the AFRRI-TRIGA Reactor and
- support systems IAW all applicable RSD instructions; i.e., RSD 5-series.
om.t /
RONALD R. SMOKER Major, EN, USA Head, Radiation Sources Division and Physicist-In-Charge s
3
e.
r-REACTOR BR ANCH ORG/NIZATION ADMINISTRATION Head (Physicist-In-Charge)
PIC OPERATIONS
/
Chief Supervisory i
s, Operator CSO i
Operator / Reactor Staff l
l l
l inclosure I to RSD Instruction 5-1.
l L
RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-2 f.
ADMINISTRATIVE LIMITS AND CONTROLS FOR REACTOR OPERATIONS 1.
Purpose. To establish the administrative limits and controls for reactor operations that will insure compliance with the Reactor License (R-84) and the Technical Specifications.
2.
Applicability. The provisions of this instruction are applicable to the AFRRI-TRIGA Reactor and aU operations conducted using the reactor, the reactor support equipment, and the irradiation facilities.
3.
Cancellation. RSD Instruction 5-2, " Administrative Limits and Controls for Reactor Operations," dated 8 April 1980.
4.
Operating Limits. The operation of the AFRRI-TRIGA Reactor and associated facilities will be performed IAW the AFRRI-TRIGA Reactor License (R-84), Technical Specifications, and with the following administrative operating limits and controls.
a.
Administrative Controls b.
Mode Operation Limits (See Table 1)
Miscellaneous Operation Limits (See Table 2) c.
r 5.
Administrative Controls.
k.
The minimum operable instrumentation for reactor operation shall consist of the a.
following:
(1) Two independent reactor power level monitoring channels which indicate the power level of the reactor during the steady state mode of operation.
(2) An NVT channel which indicates the integrated power produced during a pulse.
Scrams as indicated in Tableg n$,n.h )-
(3)
- m f
7,
~
(4) A minimum oi t vo area radiation monitors and a continuous air monitor in the reactor room. The area monitor directly over the reactor pool surface and the continuous air monitor shall produce audible and visible alarm signals in the control room.
The second area monitor in the reactor room shall have a visible alarm in the control room.
(5) Two area radiation monitors in the preparation area on the wall directly opposite both Exposure Room #1 and Exposure Room #2 plug doors to serve as radiation streaming detectors.
(6) A radiation detector system continuously sampling the effluent from the reactor building ventilating system exhaust stack and capable of producing a visible alarm at the control console.
b.
A daily "Startup Checklist" will be accomplished prior to performing any operations (other than those required for the startup). A Senior Reactor Operator (SRO) p will approve,and sign the completed "Startup Checklist." The licensed SRO that approves the checklist cannot be the same licensed operator that accomplished the checklist.
c.
Excess reactivity 1 be measured on each day any operation is planned. This measurement will be the firs eactor operation of that day.
d.
The reactor will be operated in Modes II and IIIIAW the " Mode II Checklist" and
" Mode III Checklist," respectively.
A " Nuclear Instrumentation Checklist" will be accomplished each week. The e.
completed checklist will be reviewed and signed by the Chief Supervisory Operator (CSO);
or the Physicist-In-Charge (PIC).
f.
An " Annual Shutdown Maintenance Checklist" will be accomplished by the CSO during each annual shutdown maintenance period. Tne checklist will be reviewed and signed by the PIC.
g.
Before any Mode II or Mode III operation, the reactor will be at a steady state power level for at least 30 seconds, since the last control rod movement, and with the power level at cold critical.
h.
When the reactor is operating, all doors and hatches into the Reactor Room will be closed with the exception that Door 3161 may be opened to permit personnel passage from the hallway into the Reactor Room or return.
r i.
During reactor operations, no person will be allowed inside the chain surrounding
\\.
the Reactor Pool except as authorized by the PIC or CSO.
j.
Reactor operator will be notified before opening, closing or use of any exposure facility.
k.
When an unprogrammed scram occurs, the reactor operator will notify the senior operator on duty and make appropriate logbook entries. Reactor operations will cease until the scram condition has been corrected.
1.
The CLOSE switch for the lead shield doors will not be depressed during normal operations while the core dolly is in Position 2.
The CLOSE switch may be depressed, with the core dolly in POSITION 2, to check the operability of the reactor core position safety interlock. This will be accomplished under the direct supervision of the CSO.
A " Shutdown Checklist" will be acco 'plished each day any "Startup Checklist" is m.
perform ed.
The reactor operator will maintain in the " Reactor Operation Logbook" a clean n.
and concise account of all operations.
The entries will be made IAW the " Reactor Operations Logbook Entry Checklist" such that all operations being performed could be reproduced at any future date. The future reproduction of any reactor operations would i
include using the "M alfunction Logbook," the " Maintenance Log," the "Reac tor use Requests," etc.
i l
+ -
For any reactor operation, excluding reactor maintenance, a minimum of two o.
NRC licensed operators will be present in the AFRRI complex. One operator will be an e
NRC licensed senior reactor operator. The second operator may be an SRO or RO. When the key is in the reactor console, a licensed operator will be present in the control room.
If the operator at the reactor console anticipates the need for assistance, a second staff member or operator will be available such that voice communication can be established between the two.
p.
" Reactor Operating Instruction Books" will be in the Reactor Control Room and be accessible to the operator at the reactor console. These notebooks will contain the following information pertinent to reactor operations:
(1) AFRRI-TRIGA Reactor License R-84 (2) AFRRI-TRIGA Technical Specifications (3) RSD Instructions (4) Pertinent AFRRIInstructions (5) RAD-SAF SOPS (HPPs)
(6) Completed Checklists (7) Current core data (8) Other data as desired
'\\.
TABLE 1 MODE OPERATION Lla11TS AND ASSOCIATED SCRAM SETTINGS MODE MAXIMUM POWER LIMITS AUTOMATIC SCRAM SETTINGS I & IA 1 MW Steady State 110% of Maximum Power (1.1 MW)
Fuel temperature of 500 C (2 TC channels).
II 1 MW Steady State 110% of Maximum Power (1.1 MW)
Fuel temperature of 500 C i
(2 TC channels)
III 2.3% 4 k/k Pulse Peak Power not to exceed 2750
($3.28)
MW. Fuel temperature of 500 C (2 TC channels).
TABLE 2 6
MISCELLANEGUS OPERATION LIMITS
.Old.
TSId PARAMETER Maximum Excess Reactivity 3.3%
3.5%
With infinite Water Reflector
($4.72)
($5.00) 6 k/k)
Maximum Absolute Reactivity 2.0%
2.1%
Change for Experiments (a k/k)
($2.86)
($3.00)
Bulk Water Temperature 50 C 60 C Maximum Fuel Temperature 500 C 600 C 4
Primary Coolant Conductivity 2.0 umho 2.0 umho em cm 1 - Operating Limit 2 - Technical Specification Limit 3 - Maximum Measured 4 - Averaged over one month 9
Reactor Experiments.
(1) Before any experiment can be conducted using the reactor and/or crperi-l mental facilities, the experiment will be reviewed by the " Reactor and Radiation Facility Safety Committee." If approved, the " Reactor and Radiation Facility Safety Committee" will issue either a "Special Reactor Authorization" or a " Routine Reactor Authorization."
(2) The scheduling of reactor experiments will be accomplished IAW AFRRI Instruction 3000.4 " Reactor Operations," and RSD Instruction 5-1.
(3) Before an experiment can be performed, the reactor operator will insure that the following conditions are met:
l (a) Except as noted in paragraph 4 below, a Reactor Use Request (RUR),
AFRRI Form 2, has been completed, approved, and signed by the PIC.
(b) No unsafe conditions exist with either the reactor systems or the i
experim ental f acilities.
(4) The following operations do not require an RUR:
(a) Reactor operator training (b) Instrumentation checks and calibrations (c) K-excess measurements s
l l
l (d) Reactor and reactor facility p,rameter measurements r.
(e) Maintenance (f) Tours (5) If it is anticipated that an experiment will cause a reactW.ty change in excess of + 0.5% a k/k, k-excess measurements will be made, at the core position where the experiInent will be performed, both with and without the experiment inserted. In addition, k-excess measurements will be made on all new experiments regardless of its anticipated worth.
Of]&/NAL-D/Gd6;b RONALD R. SMOKER, M
, USA Physicist-In-Charge, Reactor Chief, Radition Sources Division
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t TABLET-7.':!CMU't 29CTOR SAFETY SYSHM SCRAMS Mode in Which Effectivt Oricinatin: Chr.nel Set Poir.:
SS Pulea i
1.
Percent Pcwer
<1.1 Mwt".
X i
2.
Scram Butten on Conscie X
X 3.
Preset Timer Less than or equal X*
1 to 15 seconds 4.
Pcol Water Level 1 14 f'eet above X
X j
top of core 5.
Scram Button in Exposure X
X Room 6.
Fuel Temperature (2 1 6000C X
X independent channels) 7.
Loss of Ion Chamber High X
X Voltage l
S.
Loss of Facility Electrical X
X j
Power
(
Coes not scram reac:or - only drops transient rod following a puise.
t
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l
RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-3 REACTOR BRANCH EMERGENCY PROCEDURES 1.
Purpose. To set forth the emergency procedures to be followed in the event either an auxiliary monitoring system alarmed as a result of a radiation hazard, or a fire developed in the Reactor Building.
2.
Applicability. The provisions of this Instruction are applicable to the AFRRI-TRIGA Reactor, the Reacto-Branch staff, and the Safety Department staff.
3.
Cancellation. RSD Instruction 5-3 " Reactor Branch Emergency Procedures," 14 December 1976, is hereby cancelled.
4.
Radiation Hazard. TA suxiliary detection systems for the AFRRI-TRIGA Reactor are presented in Table I.
In addition the sensor, range alarm setting, and alarm notification are included for each individual system. Any deviation from these established setpoints will require the written authority of the PIC, coordinated through the Head, Radiation Sources Division, and the Head, Radiation Safety Department.
a.
Immediate action. The licensed reactor operator will respond to the activated alarms as follows:
(1) If any alarm listed in Table I occurs, the reactor will be scrammed within 15 seconds unless the operator is certain of the cause of the alarm, and that no hazard exists.
If the alarm occurred as a result of a malfunction on any instrument required by the f
Technical Specifications for the AFRRI Reactor (indicated by # in Table I) the reactor L
will be scrammed immediately. The PIC or his designee will be immediately notified, and the alarm will be recorded in the Reactor Operations Log.
(Z) In the event of an alarm from the Stack Particulate Monitor, the operator will ascertain if the LINAC is operating. This information will be considered before determining which action to take in (3) below.
(3) If any alarm listed in Table I occurs, the PIC or his designee will decide which one of the following actions to take:*
- (In the absence of the PIC or his designee, the licensed reactor operator at the reactor console will make the decision himself).
(a) Action #1 - Sound the fire alarm to evacuate the AFRRI complex.
l l
(b) Action #2 - Give oral orders to evacuate the immediate area in the vicinity of the radiation hazard.
(c) Action #3 - No radiation hazard exists and evacuation is not required.
l (4) At no time will any reactor operator take actions in response to instrumentation which would be in violation of the Technical Specifications of the AFRRI Facility License R-84.
[
b.
Action #1. The Reactor Branch staff will respond to this action as follows:
,F (1) AFRRI Instruction 3020.2, "AFRRI Emergency Evacuation and Fire Plan," will be implomented, and the Reactor Branch staff will respond in accordance.with that Instruction.
(2) If the reactor is operating, the operator will scram the reactor, insure that all rods are seated, and lock the reactor console. The operator will pickup all the keys in the Reactor Console Room, the Reactor Operations Log and the Emergency Checklist, close the doors to the Reactor Control Room, and then report to the PIC. The operator will turn over to the PIC the keys and the Reactor Operations Log.
(3) The reactor operator will insure that the experimental facilities are secured.
(4) A detailed investigation of the radiation hazard will be conducted, and a written report, summarizing the results of the investigation, will be completed by either the PIC or his designee, and submitted through the Head, Radiatian Sources Division to appropriate management levels within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
c.
Action #2. The Reactor Branch staff will respond to this action as follows:
(1) Same as subparagraph 4b(2).
(2) The Health Physics Division personnel will be immediately notified of the ardous condition by either the PIC or the reactor operator.
(3) The reactor operator will insure that the experimental facilities are c
secure <l.
t.
(4) A Command Post will be established at the nearest safe point to the evacuated area.
(5) If time allows, a two man re-entry team will consist of one member of the Reacto-Branch staff, and one member of Health Physics Division staff. The team will suit up accordingly and will enter the evacuated area to determine the type, size, and extent of the radiation hazard. The team will remove injured personnel if required. The corrective action taken will be determined by the information received from the re-entry team. The amount and type of protective equipment carried or worn by the re-entry team will be determined after consideration of the type of hazard existing.
(6) Injured and/or contaminated personnel will be handled in accordance with AFRRIInstruction 6310.1, " Management of Injured and/or Contaminated Personnel."
(7) Same as subparagraph 4b(4).
d.
A,ction #3. The Reactor Branch staff will respond to this action as follows:
(1) A thorough investigation will be conducted by the PIC or senior operator on duty to determine the cause of the alarm.
(2) Corrective action will be taken to insure chat the alarm will not sound unnecessarily in the future.
s
5.
Fire Hazard. The Reactor Branch staff will respond to the fire hazard emergency as e
follows:
a.
When the fire alarm is sounded (continuous tone bell), this indicates that AFRRI Instruction 3020.2, " AFRRI Emergency Evacuation and Fire Plan," has been implemented. The Reactor Branch will respond in accordance with the Instruction.
b.
Same as subparagraph 4b(2).
c.
The reactor operator will insure that the experimental facilities are secured.
d.
If the reactor operator on the console discovers the fire he will follow the procedures in subparagraph 4b(2). In addition, he will pull the fire alarm at Door 3162 aM 3/5+
when exiting the AFRRI complex.
e.
Personnel with contaminated clothing will not attempt to change clothes prior to evacuation, but will segregate themselves when outside the building.
f.
No attempt will be made to remove fuel elements from the reactor pool during a fire.
g.
If the fire occurred in the Reactor facility then an investigation will be initiated as stated in subparagraph 4b(4).
6.
Natural Disaster. The Reactor Branch staff will respond to a natural disaster such as flood, tornado, hurricane, earthquake, etc., as follows:
r I-If either a forecast is received that a natural disaster is imminent ur a natural a.
disaster strikes the AFRRI complex, the reactor operator will immediately scram the reactor, if operating, and then place the reactor in a secured condition.
b.
The reactor operator willinsure that the experimental facilities are secured.
7.
Bomb Threat. The Reactor Branch staff will respond to a bomb threat as follows:
Upon receipt of a bomb threat, Administrative Services Division will be notified a.
imm ediately.
b.
The Reactor Division staff will take appropriate action in accordance with AFRRI Instruction 5200.4, " Physical Security Plan for AFRRI", and AFRRI Instruction 5220.4, " Physical Security Plan for AFRRI-TRIGA Reactor Facility."
8.
Civil Disorder.
The Reactor Branch staff will respond to a civil disorder in accordance with AFRRI Instruction 5220.4, " Physical Security Plan for AFRRI-TRIGA Reactor Facility."
C//GM/L Sdf@ A, RONALD R. SMOKER, MAJpSA Physicist-In-Charge, Reactor IIead, Radiation Sources Division
TABLE I TECH SPEC SYSTEM REQUIRED RANGE ALARM SETPOINTS ALARM NOTIFICATION 6
1.
R-1" 1 to 10 mr/hr a)f00 mr/hr whe,n reactor personnel a)
Local audible buzzer and are on duty red light.
b) 20 mr/hr during non-duty hours b)5 Instrument module - audible buzzer and red light.
c)
Annunciator Panel - audible horn and red light.
5 5
2.
R-Z" 1 to 10 mr/hr 10 mr/hr*
Instrument module - red light.
5 3.
R-3" 1 to 10 mr/hr 10 mr/hr*
Instrument module - red light.
8 5
4.
E-3 and E-6" 1 to 10 mr/hr 10 mr/hr*
a)
Local red light.
b)
Instrument module - red light.
5 5
5.
Stack RAM" 1 to 10 mr/hr 100 mr/hr*
Instrument module - red light.
6 5
6.
Stack Gas 10 to 10 cpm 10 cp m *
- Instrument module - red light.
7.
Reactor Room Cam 50 to 5x10 cpm 10 cp m * *
- a)f Local audible bell and red light.
e 4
4 b)5 Annunciator panel-red light.
c)
Air damper closed lights on instrument module.
d)
Instrument module - red light.
e)3 Annunciator panel-audible horn and red light.
8.
NMC Critical-5 it y" 1 to 10 mR/hr a) 6~4 mr/hr - when reactor personnel Audible bell and red light.
are on duty * * *
- b) 29 mr/hr - during non-duty hours 9
Water Box 5
Gamma Monitor" 0 to 1.0 ma 0.5 ma Red light on instrument module.
S 10.
Exhaust Fan n/a 0CFM Bell and red light on wall.
W
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l c
Scintillation detector 1
Reactor Room b
Proportional detector 2
Room 3155 c
GM detector 3
IIallway 3101 d
Float activated switch 4
Prep Area 5
Control Room Loss of signal alarm: Instrument module - blue light Loss of signal: a)5 Local audible horn and red light; Loss of air pump alarm: Audible horn b) White light on instrument module Loss of power alarm:
Local yellow light; Loss of signal alarm:
White light on instrument module 5
Loss of signal alarm:
White light ADD TO TABLE I:
SYSTEM llANGE Al,AllM SliTPOINTS ALA141 NOTIFICATION Stack Particulatec 10 to 105 cpm 2 x 103 5
cpm Instrument module-red light l
9 I
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RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-4 REACTOR MAINTENANCE PROGRAM FOR THE AFRRI-TRIGA REACTOR 1
Puraose. To set forth an organized and systematic reactor maintenance program for the AFR3.I-TRIGA Reactor and to establish maintenance records that will provide a perrnanent maintenance history of the reactor systems and auxiliary support equipment.
2.
Applicability. The provisions of this Instruction are applicable to the AFRRI-TRIGA Reactor and the Reactor Branch Staff.
3.
Cancellation. RSD Instruction 5-4 " Reactor Maintenance Program for the AFRRI-TRIGA Reactor," dated 28 June 1979 4.
Definitions.
a.
Malfunction. The failure of a component of a reactor system that will prevent the system from either operating in its normal manner or fr.om performing its intended function.
b.
Modification. Any physical change (other than direct replacement), in either the components or the design of the reactor system.
c.
Major Modification.
Any physical change, in either the components or the design of the reactor systern, that will require an amendment to the existing technical specifications of the Reactor License (R-84) and therefore approval by the Nuclear j-Regulatory Commission (NRC).
L d.
Minor Modifications. Any physical change, in either the components or the design of the reactor system, that will not require an amendment to the e;tisting technical specifications of the Reactor License (R-84).
5.
Responsibilities. Responsibilities for the reactor maintenance program are delin-eated as follows:
Head, Reactor Branch (Physicist-in-Charge) (PIC):
a.
(1)
Has authority to approve any change in either the components or the design of a reactor system that does not constitute a major modification.
(2)
Performs periodic inspections to insure that the maintenance schedules and procedures are followed, and that all maintenance records are being properly maintained.
(3)
Reviews the Malfunction Log quarterly to insure that proper corrective measures have been taken.
(4)
Prepares documentation for proposed major and minor modifications, and maintains the major and minor modifications records.
b.
Chief Supervisory Operator (CSO):
(1)
Has primary responsibility for implementation of the reactor maintenance g-program in accordance with the Reactor License (R-84).
(2)
Supervises the reactor operators in the performance of all reactor maintenance.
(3)
Accomplishes the training of reactor operator trainees in the reactor maintenance program.
(4)
Supervises, when necessary, maintenance personnel from outside the Reactor Branch when working on reactor systems and auxiliary support equipment.
(5)
Maintains current and accurate reactor maintenance records enept for minor and major modifications.
(6)
Insures that all changes in the reactor systems are reflected in the as-built drawings.
(7)
Schedules preventative maintenance to include the annual shutdown maintenance period.
(8)
Recommends to PIC changes that will enhance the reactor maintenance program and procedures.
c.
Reactor Operator (RO):
r (1)
Performs reactor maintenance as directed by the Chief Supervisory L
Operator.
(2)
Immediately notifies the Chief Supervisory Operator whenever a mal-function occurs.
(3)
Informs the Chief Supervisory Operator whenever maintenance on any component of either the reactor or auxiliary systems is required.
(4)
Recommends to Chief Supervisory Operator changes that will enhance the reactor maintenance pregams and procedures.
6.
Maintenance Program.
The AFRRI-TRIGA Reactor Maintenance Program is conducted in ccmpliance with the technical specifications of the NRC Reactor License (R-84), and to insure the safe and reliable operation of the reactor in support of the AFRRI research program. In order to facilitate the administration of the maintenance program, the reactor system is divided into component systems. The maintenance program is divided into three general categories:
a.
Preventative Maintenance.
Consists of maintenance that is scheduled in advance and is accomplished in accordance with the schedule indicated on the Maintenance Chart.
s k
b.
Corrective Maintenance. Consists of maintenance that is non-scheduled and that requires either repairs or calibrations as a result of a malfunction or detection of a marginally operating component.
c.
Modifications.
Consists of major and minor modifications as defined in subparagraphs 4.c. and 4.d., respectively. Modifications are made to the existing reactor systems to increase the safety and reliability of reactor operations.
7.
Maintenance Records. The AFRRI-TRIGA Reactor maintenance records are main-tained in compliance with the technical specifications of the NRC Reactor License (R-84), and to provide a permanent history of all maintenance performed on each reactor and auxiliary system component.
a.
Maintenance Log.
(1)
The Maintenance Log will consist of a file system delineating each major component or system of the reactor.
The first section will be a list of current components in the system. Each section will refer to the frequency of inspection, or preventative maintenance, or calibration of the cornponent specified. Each section will describe the action required or refer to manuals or files giving such description.
Modifications to any component, change in any procedure covering calibration or inspection, or frequency of inspection will be referenced in each systems section.
(2)
There will be a space provided in the system for a written statement of action taken on any component along with a date, signature of operator completing the action and a block for CSO or PIC initials upon review of action taken.
b.
Malfunction Log. Consists of a bound notebook with all information pertinent to the reactor systems malfunctions in it. The entries in the Malfunction Log will be
(~~
made by the individual discovering the malfunction and will consist of the date the malfuncton occurred, the component that malfunctioned with a description of the malfunction, the corrective action taken, and the initials of the individual. The CSO or PIC will review the malfunction statements and initial the Malfunction Log.
c.
Maintenance Chart. A chart may be used to cover a calendar year depicting the date on which preventative maintenance will be performed for componenes in the system. The dates for preventative maintenance are depicted on tha chart with an (X).
The Maintenance Chart is to be used for informal reference only, actual records will be contained within the Maintenance Log file. The Chart is to be color ( 3ded for frequency of maintenance.
d.
Modification Records.
Consists of all proposed Modifica. ion Memorandums pertinent to reactor modifications. These records will be maintained in a current status i
by the PIC.
e.
As-Built Drawing Records. Consists of all as-built prints of the components of the reactor systems as originally designed and includes the modifications made to these reactor systems. These records will be maintained in a current status by the CSO.
f.
Calibration Data Records. Consists of final data taken during the calibration of the various reactor parameters specified in the technical specifications of the Reactor License (R-84). These records will be maintained in a current status by the CSO.
s
8.
Manuals. The maintenance program will be accomplished in accordance with the procedures set forth in the Reactor Maintenance Manual and as recommended in the
(
manufacturers' manuals except where updated or modified based on operational exper-ience.
9 Modification Procedures. A modification will be documented, approved and reviewed prior to being accomplished.
The proposed modification will be documented in a
" Modification Memorandum." The Memorandum will contain the following information:
a.
The purpose of the modification.
b.
A detailed description of the modification to include figures and as-built drawings as deemed ne::essarv.
c.
An analysis of the modification with emphasis placed on the reactor technical specifications, the safety of operation of the reactor, and the safety of personnel.
10.
A Minor Modification Memorandum will be prepared by the CSO, approved by the PIC and reviewed by the Head, RSD. A Major Modification Memorandum will be prepared by the PIC, reviewed by the Head, Radiation Sources Division, reviewed by the Reactor and Radiation Facility Safety Committee, approved by the Director and approved by the NRC.
11.
Reactor Core Maintenance.
The maintenance on the reactor core will be accomplished in accordance with the following conditions:
a.
Maintenance on the control rods will be performed only when the reactor is in a shutdown condition.
(
b.
The PIC will insure that the number of non-essential personnel in the Reactor Room will be kept to a minimum when core maintenance is performed.
c.
The core maintenance will be recorded in either the Reactor Operations Log, Maintenance Log, or Malfunction Log, or a combination of these, as required.
d.
A control rod will only be manually removed from the reactor core under the following conditions.
(1)
The reactor will be in a shutdown condition.
(2)
A minimum of three individuals will be present, and at least two of these individuals will possess an NRC license. One individual will be either the PIC or CSO and l
possess an NRC Senior Reactor Operator lice ~m. The second individual will possess either an NRC Senior Reactor Operator license cr an MRC Reactor Operator license.
(3)
One licensed operator will oh orve the core nuclear instrumentadon.
(4)
The minimum shutdown margin provided by the remaining control rods with the most reactive control rod fully removed shall be $1.00 (0.7% k/ ).
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The loading and unloading of the fuel elements in the reactor core will be e.
accomplished in accordance with RSD Instruction 5-8 " Reactor Core Loading and I.
Unloading Procedures."
Gk6/WAl- 0/6Vd2 RONALD R. SMOKER, J, USA Physicist-In-Charge, Reactor Head, Radiation Sources Division t
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o RADIATION SOURCES DIVISION 31 March 1981 INSTRUCTION NUMBER 5-5 REACTOR ROOM POWER HOIST OPERATIONS PROCEDURES 1.
Purpose. To set forth the procedures to be followed by the Reactor Branch staff in the operation of the Reactor Room Power Holst.
2.
Applicability. The provisions of this Instruction are applicable to the Reactor Branch sta ff.
3.
Cancellation.
RSD Instruction 5-5,
" Reactor Room Power Hoi-Operations Procedures", dated 14 December 1976 is hereby cancelled.
1 4.
General.
a.
The power hoist will not be operated without the permission of the Physicist-In-Charge (PIC) of the Reactor Branch or his designee.
b.
A minimum of two people will be present during the operation of the power hoist. One of the two people must be from the Reactor Branch.
c.
The key to the power hoist circuit breaker lock will remain secured in the Reactor Branch key box when the power hoist is not in use.
5.
Operation.
b The reactor will be in a shutdown condition before electrical power is applied to a.
the hoist.
b.
Restraining lines will be attached to all loads, prior to being lif ted by the power hoist, to control the swinging of the load, and to assist in guiding the load during horizontal movement.
d.
Any equipment to be used with the hoist such as hooks, slings, cables, chains, etc; must be visually inspected for wear and damage, and must be load tested prior to their being used with the hoist for any operation. The results of the visualinspection and the load testing will be documented in the Reactor Operations Logbook.
While electrical power is applied to the hoist and overhead rails, the following e.
additional administrative controls will apply:
(1) The personnel in the Reactor Room will remain clear of the hoist, load, and overhead power rails. Personnel will not stand directly under the load for any reason. All personnel should remain as far from the load as the length of the longest load-bearing cable, if possible.
i (2) Tools and poles, longer than 10 feet, will not be raised vertically in the Reactor Room to include the storage area over the Control Room.
(3) The control box will be continuously in the possession of the operator.
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Before any load is lifted in the vicinity of the reactor pool, simulated operations will be conducted in another location by the operator and the procedures approved b-the f.
PIC or his designee.
Before any load is lifted in the vicinity of the aluminum floor hatch in the g.
Reactor Room, the areas on the first and second levels beneath the hatch will be secured to ensure all personnel are kept out of areas directly below the load.
h.
The power hoist will be load tested and certified annually be licensed inspectors.
Results of the load test will be filed in the maintenance files and the maximum load rating will be stenciled on the hoist.
CAldvA/AL S/d//63 RONALD R. SMOKER, M J, USA Physicist-In-Charge, Reactor Head, Radiation Sources Division a
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RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-6 r
REACTOR OPERATOR TRAINING AND REQUALIFICATION PROGRAM J
1.
Purpose. To set forth a reactor operator training program to insure that candidates for USNRC Senior Reactor Operator and Reactor Operator licenses are properly trained in facility design, operation, maintenance, nuclear theory, and emergency procedures applicable to the AFRRI-TRIGA IWark-F Reactor.
To set forth a reactor operator requalification program, in accordance with 10 CFR-55, appendix A.
2.
Applicability. The provisions of this instruction are applicable to tl' Reactor Branch staff and to all operators licensed by USNRC on the AFRRI-TRIGA reactor.
3.
Cancellation. RSD Instruction 5-6, " Reactor Operator Training and Requalification Program" dated 27 June 1979 4.
Responsibilities.
The authority to implement both the reactor opere cor training program and the reactor operator requalification program is delegated ta the Head, Reactor Branch through the Head, Radiation Sources Division.
l 5.
Training Programs. The reactor operator training program is designed to provide the -
training for both Senior Operator and Reactor Operator candidates for. USNRC operator licenses. Personnel assigned to the Reactor Branch will be presumed to possess a general familiarity with reactor operations by virtue of previous training and experiment. All such personnel will enter the training program to achieve the desired level of competency-as a nuclear reactor operator. Certain individuals assigned to the Reactor Branch will be
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subsequently provided additional training to enable them to become a candidate for a USNRC Senior Operator's License. The reactor operator trainees will follow an organized program. The extent of the training program will be developed in view of the trainees'-
previous training and experience.
The program may include classroom lectures and examinations, on-the-job training and demonstration, console operation, and operation of the experimental facilities.
Principally, the reactor nuclear theory and radiological safety portions of the program will consist of classroom lectures while the remaining portions will be conducted by on-the-job techniques and supplemented by lectures as necessary.
a.
Reactor Operator Training. % eactor operator training will be conducted in accordance with the " Reactor Operatot
- qualification Program". This training program is designed to provide personnel previously trained in the basic principles of nuclear reactor operations with an acceptable familiarity and operational competency with the AFRRI-TRIGA Mark-F Reactor.
Completion of this - course demonstrates ability to
~ ll operator assigned tasks, without supervision, operate this reactor and accomplish a
during all normal reactor startups, operations, and shutdowns.
Must be capable of.
establishing and insuring a safe configuration during = any abnormal condition; must.
demor, strate familiarity with the design and. control of the reactor necessary for knowledgeable and competent operation; must possess a thorough knowledge of nuclear safety, the possible hazards associated with this facility, and response to the separate levels of alarm; must demonstrate a thorough knowledge of the directives applicable and the reports required of all operations.
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Senior Reactor Operator Training. The senior reactor operator training will accomplished in accordance with the " Reactor Operator Requalification Program". This g
training program is designed to provide personnel with that depth of knowledge ~iequired for supervisory direction, analysis cf routine reactor data and solution of. problems associated with reactor operation or experiment handling. Completion of this course,
demonstrates competence to direct the activities of licensed reactor operators in all emergency reactor operations to include accomplishment or, routine, non-routine, or required data forms, observance and interpretation of reactor parameters, and timely response to and solution of abnormal parametric conditions; must be capable of directing
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and training other operators and effecting an accurate and timely analysis of emergency indications.
f 6.
Requalification Program.
The requalification program will be conducted in -
accordance with the NRC approved AFRRI " Reactor Operator Requalification Program".
required to.',
All NRC licensed reactor operators and senior reactor operators are participate in the requslification program.
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j 7.
Recommendation for Licensing. Upon completion of the required training, the Head, Reactor Branch will determine the level of competency of the license candidate and forward to the Head, Radiation Sources Division, a recommendation for submission of the candidate to the USNRC for licensing. The Head, Radiation Sources Division will review the recommendation submitted and make a determination whether or not the candidate will be submitted to the USNRC for licensing. If approved the candidate will prepare an application either for licensing in accordance with 10 CFR-55.10 or for renewal of reactor license in accordance with 10 CFR-55.33. The application will be coordinated with the Head, Reactor Branch; Head, Radiation Sources Division; and Director, AFRRI prior to being forwarded to the USNRC.
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RONALD R. SMOKER, M J, USA Physicist-In-Charge, Reactor He'adr Radiation Sources Division
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RADIATION SOURCES DIVISION 18 March 1981 INSTRU, CTION NUMBER 5-7 r
PNEUMATIC TRANSFER SYSTEMS AND CORE EXPERIMENT TUBE OPERATION PROCEDURES 1.
Purpose.- To set forth the procedures to be followed by the Reactor Branch staff in the operation of the Pneumatic Transfer Systems (PTS), and the Core Experiment Tube (CET).
2.
Applica ility. The provisions of this Instruction are applicable to the Reactor Branch staff and the Radiation Safety Department staff.
3.'
Cancellation.' RSD Instruction 5-7 dated 8 April 1980 is hereby cancelled.
9 l 4.
_ General.
All use of the PTS and CET will be under the supervision of a Senior
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Reactor Operator. When not in use, the control key for the PTS will remain secured in j
the, Reactor Branch key box.
/t 5.'r / rr/dfation. No irraddtion will be carried out without an approved RUR and, when I
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requised, an SWP.'
6.
Operators and Mor tors.
a.'
A member of the Reactor Branch will operate the PTS and the CET. His duties include:
Physi' al operation of the PTS control panel.
(1) c f
(2) Communication with the reactor operator.
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(3) Inspection of rabbits to be used to preclude use of cracked or otherwise r damaged rabbits.
, b.
A fronitor from the Radiation Safety Department (SAHP) will be present when a retrieval is in progress, and will insure complirnce with applicable Health Physics
,4 Stardazd Operating Procedures of 'the SAHP.
In all cases the reactor operator is responsible for making the appropriate
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operatibna, lo'g ; book entries and for coordinating the insertion and retrieval of
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experiments.j -
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NOTE: Items 7 and 8 apply only to the Pneumatic Transfer Systems.
i 7.
Aluminum rabbits will be diverted to the hot cells, and therefore can only be R
irradiated in the "A" hy. stem.
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8.
If the anticipated radiation level of any returned rabbit is greater th LD /hr at 1 meter from the receiving station, the following precautions will be taken:
Unless experiment requirements dictate otherwise, the PTS operator will use the a.
remote control unit.
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b.
All non-essential personnel will exit from the Radiochemistry Laboratory prior to returning the rabbit.
r A radiation survey meter will be set up near the receiving station so that it will c.
be visible from the remote control unit.-
d.
The rabbit will be irradiated in the "A" system and then diverted to the Hot Cell or returned to the termir.us.
NOTE: Items 9 through 12 apply only to the CET.
9 Only CET type rabbits will be _used with the CET. CET rabbits will be raised or lowered into the CET through the use of a fishing pole assernbly which allows the operator to maintain up to approximately six feet between himself and the retrieved experiment.
At the direction of an SRO a plastic rabbit may be simply dropped into the CET without using the fishing pole assembly.
- 10. Irradiated CET rabbits will be raised near the pool surface, within the CET, for dose rate measurements by the SAMP monitor. Experiments too active for safe transfer will be returned to a lower portion of the CET and allowed to decay.
- 11. To protect the fishing pole assembly and the extractor from undue activation, insertions and retrievals are best performed when the reactor is shut down. Normally the extractor will not be lowered into the active core region when the power level is greater than 100 watts.
- 12. Open air transfer of a sufficiently " cool" rabbit will be accomplished only after non-essential personnel have cleared the immediate area.
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- 13. Radiological safety considerations will be coordinated through SAHP.
Cr0&MfL.StT/Cb RONALD R. SMOKER, h J, USA Physicist-In-Charge, Reactor Chief, Radiation Sources Division
T RADIATION SOUP.CES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 REACTOR CORE LOADING AND UNLOADING PROCEDURES 1.
Purpose. To set f~th the procedures to be followed by the R, actor Branch staff in the complete loading at.
.nloading of the AFRRI-TRIGA reactor core.
2.
. Applicability. The provisions of this instruction are applicable to the Reactor Branch staff, and the Health Physics Division staff.
3.
Cancellation. RSD Instruction 5-8 dated 14 December 1976 is hereby cancelled.
4.
General.
a.
The reactor core loading and unloading procedures contained herein apply to the preparation phase as well as the actuni loading and unloading phases. These procedures are based on two (2) separate core loadings and one core unloading of the AFRRI-TRIGA rt 'etor.
Emphasis is placed on following the procedures specified herein to insure continuity of operation and retention of experience within the Reactor Branch.
b.
All actidties associated with either the loading or unloading of the reactor core will be recorded n le Reactor Operations Logbook.
c.
The minimum number of personnel that will be required is: (1) Physicist-in-
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Charge (PIC) or his designee, (2) Chief Supervisory Operator (CSO), (3) One NRC licensed reactor operator, a.id (4) Health Physics Division representative.
d.
A daily Startup Checklist will be completed prior to the movement of any fuel elem ents.
c.
An approved Special Work Permit will be initiated prior to the movement of any fuel elements.
f.
If any new fuel elements are to be used, each element must be inspected when received at AFRRI. Each element will be removed from its shipping container, cleaned, and inspected for visual defects. Length and bow measurements must also be made and recorded. Smears of the element cladding for aloha contamination must be performed by I
the Health Physics Division representative prior to being handled by Reactor Branch l
personnel.
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g.
If any new thermocouple elements are to be used, a thermocouple calibration will be performed. The fuel element will be placed in a water bath, and Emf readings will be recorded over the range 20-100 degrees Centrigrade.
h.
At no time will more than six (6) new fuel elements be out of thc% shipping containers and on the reactor room floor level.
i.
The Physicist-in-Charge or his designee must directly supervise all sequences of loading and unloading the reactor core.
Page 1 of 7
..o RADIATION SOURCES DIVISION 27 Alarch 1981 INSTRUCTION NU51BER 5-8 j.
An NRC licensed reactor operator will continuously observe the nuclear instrumentation at the control console during all movements of control rods and fuel elements.
k.
No fuel element which has experienced burnup in the core shall be removed from the reactor pool unless at least two (2) weeks have transpired since its use in the core.
5.
Nuclear Instrumentation.
a.
The following nuclear instrumentation is the minimum required for a reactor ccre loading:
(1) Two ionization chambers will be located outside the core shroud, along the core centerline, and adjacent to core positions F-4 and F-12, respectiwly. The readouts for these chambers will be picoammeters, or equivalent.
(2) One BF e fission chamber will be located outside the core shroud, along 3
the core centerline, and adjacent to core position F-8. The readout for this chamber will be a scaler unit.
b.
The minimum nuclear instrumentation required for the unloading of a reactor core is:
(1) One compensated ionization chamber will be locate (' outside the core
( '
shroud, along the core centerline, and adjacent to core position F-12. The readout for this chamber will be a picoammeter, or equivalent.
(2) One BF or fission chamber will be located outside the core shroud, along 3
the core centerline, and adjacent to core position F-8. The readout for this chamber will be a scaler unit.
c.
An operational check of the channels will be made as follows prior to the movement of any fuel element.
(1) A neutron source (3-5 cuAs) will be placed in the neutron source holder, and an increase in the readings will be observed on all channels.
(2) The neutron source will be removed from the neutron source holder and the readings will be taken and recorded.
(3) Replace the neutyngou in the neutron source holder, and then generate a bias curve for the startup cliahnelli,pe'(2) above. Record all channel readings with the M.fi.
source.
These measuremerits will be performed several times in order to obtain reasonable reproducibility. These readings will be the basis for future calculations of source multiplication only in the loading of a reactor core. The neutron source reading will be the difference between the readings with and without the neutron source in place in the reactor core.
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r RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 d.
The nuclear instrumentation will be turne ; on and allowed to stabilize prior to the movement of any fuel elements, or making measurements of source effect.
6.
Core Loading.
a.
A 1/M curve is obtained by plotting the inverse multiplication vs the amount of fuel added (total amount in the core). The inverse multiplication is the ratio of the source reading to the reading with the fuel added. The loading curve will seldom be a straight line but may be either concave or convex dependent upon the geometry (source-detector distance).
Hence, a number of different channels will yield different predictions of criticality. Since not all channels will agree, a conservative approach will be taken and the smallest number of estimated fuel elements required for criticality will be used to dictate future steps.
b.
The fuel elements will be loaded in accordance with Table 1.
TABLE 1 FUEL LOADING SCHEDULE STEP #
- ELEMENTS REMARKS ADDED TOTAL
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1 4
4 Load four thermocouple elements,2 in the B ring and 2 in the C ring.
2
-14 18 Complete loading of B and C rings.
3 15 33 Load D ring.
4 15 48 Load E ring positions 1,2,4,6,8,9,10,12, 14,16,17,18,20, 22 and 24. This loading is designed to complete a compact array around the control rods as well as to fill water gaps.
5 9
57 Complete loading of E ring.
6 9
66 Load F ring in posittens 1, 5, 9, 13, 17, 21, 22, 23, and 27.
c.
After each step of the fuelloading, perform the following:
(1) Record readings.
(2) Withdraw control rods 50%
(3) Record readings.
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RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 (4) Withdraw control rods 100%
(5) Record readings.
(6) Calculate M,1/M for the step.
(7) Plot 1/M vs # fuel elements.
(8) Plot 1/M vs weight of uranium-235.
(9) Plot 1/M vs control rod position (50% and 100%).
(10) Predict criticalloadings.
(11) Estimate worth of the control rods.
(12) INSERT CONTROL RODS TO Ft'LL "IN" POSITION.
d.
AFRRI-TRIGA Core 1(aluminum clad elements) attained criticality with 72 fuel elements, 2811.33 grams uranium-235.
AFRRI-TRIGA Core II (stainless steel clad elements) attained criticality with 69 fuel elements,2630 grams isranium-235.
e.
Continue the loading sequence as detailed below until criticality is obtained, and
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until the excess reactivity is 40-50 cents:
TABLE I (Continued)
FUEL LOADING SCIIEDULE STEP #
- ELEMENTS REMARKS ADDED TOTAL 7
2 68 Load F ring positions 19 and 25.
8 2
70 Load F ring positions 3 and 11.
f.
Prior to loading the core to an operational configuration, the following measure-ments will be made:
(1) Control rod calibrations using the rod drop techniques.
(2) The worth of fuel elements in the remaining vacancies (E and F ring) vs water, taken one at a time.
(3) Estimate the core configuration for an excess reactivity of approximately
$3.20.
g.
The loading sequence will continue in order to attain a critical configuration with the transient rod in the DOWN position. This is the basis for the excess reactivity estimate of approximately $3.20.
Page 4 of 7
f RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 TABLE I (Continued)
FUEL LOADING SCIIEDULE STEP #
- ELEMENTS REMARKS ADDED TOTAL 9
2 72 Load F ring positions 7 and 15.
10 4
76 Load F ring positions 2,14,18 and 29.
Record critical rod Bank position; Calibrate the upper portion of the transient rod (0-25%)
via the positive-period technique.
11 4
80 Load F ring positions 8,10,24 and 30.
Calibrate the middle portion of the transient rod (25-75%) via the positive period method.
12 2
82 Load F Ring positions 16 and 20, and this should complete the operational configuration as stated above.
h.
Calibrate the four control rods via the positive period method, and then compute
(
the excess reactivity in the reactor core (K-excess must not exceed $5,00).
i.
Complete the core loading, insuring that the K-excess does not exceed $5.00.
TABLE I (Continued)
FUEL LOADING SCIIEDULE STEP #.
- ELEMENTS REM._RKS ADDED TOTAL 13 5
87 Load F ring positions 4, 6,12, 26 and 28.
J.
Recalibrate the four control rods via the positive period niethod, and then compute the K-excess reactivity in the reactor core.
7.
Core unloading.
a.
The reactor core will be unloaded starting with the F ring and ending with the B ring.
b.
The fuel elements will be individually removed from the reactor core, identified by serial number, and placed either in the fuel storage racks or a shipping cask.
c.
If the fuel elements are to be loaded into a shipping cask, the following actions will be taken in preparing the shipping casks for loading:
Page 5 of 7
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RADIATION SOURCES DIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 (1) A radiological survef will be made of the shipping cask upon arrival and before it will be removed from the truck.
(2) The cask will be moved from the truck to the Prep Area.
(3) The hatches, which provide access from the Prep Area to the Reactor Room, will be opened and the lifting hook to the power hoist lowered to the Prep Area.
(4) The power hoist will be operated in accordance with RSD Instruction 5-5.
(5) The lifting yoke will be attached to the cask and the cask lifted to the Reactor Room.
(6) The lid to the cask will be removed. The cask will be monitored by the llealth Physics Division representative while the lid is being removed, to insure that no radioactive materialis inside the cask.
l (7) The inside of the cask will be smeared for gross alpha and beta contamination.
(8) The inside of the cask will be vacuumed. The inside and outside of the cask will be washed down. The water drain line on the cask will be checked to insure that it is not blocked. Also verify the operability of the pressure relief valve and the temperature
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sensing thermocouple.
(9) If more than seven elements are to be loaded into the cask, it will be necessary to verify that a thermal neutron poison is present in the cask to prevent the loading of a critical mass.
(10) Move the cask by crane from the reactor deck and position the cask in the reactor pool.
d.
Load the cask with up to as many fuel elements as allowed by the license for tne cask. If grid index markings are present in the cask, record which fuel element is placed in which grid position.
e.
Lower the lid to the cask into the pool, place the lid on the cask, and secure the lid.
f.
Raise the cask from the pool, drain the water from the cask into the pool, and then cry the cask off. The cask will be monitored while being removed from the pool to insure that no radiation hazard exists as a result of a weakness in the shielding in the cask. The cask will be smeared for gross alpha and beta contamination.
g.
An air sample will be taken from the cask to measure the activity of the air.
The data from all radiological surveys will be recorded.
h.
After the air sample has been taken, observe the temperature ano e ssure inside a
the cask until the temperature and pressure reach an equilibrium.
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RADIATION SOURCES iIVISION 27 March 1981 INSTRUCTION NUMBER 5-8 1.
Label the cask accordingly and complete the appropriate pape work either for-
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temporary storage or for transporting.
J.
Move the cask to either a temporary storage area or to the truck for transporting. If the cask is to be placed in temporary storage, a criticality monitor must be available in accordance with 10 CFR 70.
J.
- l RONALD. SMOKER MAJ, EN, USA Physicist-in-Charge, Reactor Head, Radiation Sources Division l
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