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{{#Wiki_filter:TEC H SPfcS 6  .- o                  DOCWET [d#lld DEFENSE NUCLEAR AGENCY G.
ARMED FORCES nADIOnlOLOGY RESE ARCH INSTITUTE DETHE SD A. F.e AnYLAND 20889-5145 RSDR
 
==SUBJECT:==
Submission of Annual Report U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
 
==Dear Sir:==
 
Attached please find the 1991 Annual. Report for the AFRRI TRICA reactor facility, submitted as required by license R-84,-facility docket 50-170.
Should you need any further information, please contact the undersigned at (301) 295-1290.
                                                                                    -l l)i]il Attachment                                          WARK WOORE as stated                                          Reactor Facility Director CY Furn:
U.S. Nuclear Regulatory Commission                                                            ~q
                                                                                                          ^
ATTN: Mr. Warvin Mendonca, Wall Stop 11H10 Washington,'DC 20555 U. S. Nuclear Regulatory Commission, Region I' ATTN: Mr. Thomas Dragoun 475 Allendale Road King of Prussia, PA 19406 pi) 9203300221 911231          2                                                        ' ' 8 .VV PDR- ADOCK 05000170 .
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1991 ANNUAL REPORT OF AFRRI TRIGA REACTOR
                                  )
Submitted by:
MARK MOORE Reactor Facility Director Docket 50 - 170 License R - 84
 
i Reviewed and Approved b                                        ,
Yhulbh t M ARK L. MOORE [j 23/lbAcq Date            '
Reactor Facility Direditu Approved for Release (mk3 '"&A 9 )
ROBERT L. BUM @\RN ER                                              Date Captain, MC, USN Director
 
l 1991 ANNUAL REPORT                                          ;
TABLE OF CONTENTS Intr 6 duction                                                                                  f f
Generrn Information                                                                              i becticn I Changes to tite f acility des!gn, performance cheracterletics and operational procedures. Results of survelllance tests and inspections                              !
Section II Energy generated by current reactor core and number of pulses 1 82.00 See,lon III                                                                                      '
Utscheduled shutdowns Section IV Safety-related corrective mainterance Section V Faci!Ity changes and changes to procedures as described in the Safety Analysis Report. New experiments or tests during the year.
Section VI Sumary of radioictive effluent released Section VII Environmental radiological surveys Soction VIII Exposures granter than 25% of 10 CFR 20 limits Attachment A 10 CFR 50.59 analysis and Refueling Plan for fuel-follower control rrta.                ,
Attachment B Current Reactor Administrative and Operational Procedures Attachment C Amendment No. 21 to Facility Operating Licensa Attachment D Routine Reactor Authorizations Attachment E 10 CFR 50.5g safety evaluations of modifications, changes, and enhancements to procedures or facilities (other than fuel-follower control rods).
Attachment F May 1991 sumary of changes to administrative and operational procedures                    -
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)    INTRODUCTION                                                                                                              ,
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                                                    .-                                                EXPOSURE ROOM 2
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f' PLUG 000R Cutaway View of AFRRI TRIGA Reactor
 
1991 ANNUAL REPORT INTRODUCTION:
In 1991, the AFRRI reactor staff accomplished two important milestones in continuing efforts to improve the operational capabilities and reliability of                                                          #
the reactor facility. Af ter Nuclear Regulatory Commission (NRC) approval of the installation through Amendment 21 to the f acility operating license (Attachment C), the reactor staf f installed three new f uel-follower standard control rods, s i
new longer air-follower transient control rod, and related equipment during an exter.ded maintenance shutdown period of November-December 1991. Significant post-installation testing and calibration was performed to verify the validity of the original safety analysis and refueling plan (Attachment A) and the reactor was returned to regular steady-state operation on 16 December 1991.
Approval for resumption of pulsing operations was withheld by the Reactor Facility Director pending completion of an testing program of incremental sized pulses in early 1992. That testing program will include pulses up to 82.S0, but the current administrative pulse size Iimit of $2,00 remains in offect for normal operations.
Also, the microprocessor-based instrumentation and control consolo installed in 1990 completed its first fu!I year of successful operation. Two modifications mndo to the console to improve its capabilities are discussed fully in Sections I and V and Attachment E.
The Reactor Facility was inspected by the Defense Nuclear Agency Inspector Osneral fren 25 to 27 September 1991. The inspection found that the AFRRI TRICA Rsactor facility received an everall grade of SATISFACTORY with no significant and discrepanciesnotedinorganization, sofeguards. There were no reportableoperations,materialhandlinglntenance deficiencies found in the ma records, operational log books, or training records. During the inspection, four minor deficiencies were documented. None of the cited deficiencies were considered significant and, neither singly nor in aggregate, impaired the performance or degraded the safety of the TRICA nuclear reactor operations.
Concurrently with that inspection, the reactor facility also received a Nuclear R actor Security Inspection by representatives of the Assistant to the Secretary of Defense, Atomic Energy. That inepection also gave the reactor facility an overall rating of SATISFACTORY.
The Reactor Facility was also inspected by NRC personnel from Region I on 23-2S April 1991. No vioistions were identified during this inspection.-
Changes were made to the procedures and facilities during 1991. These_ changes ccre sepported by an extensive safety review process in accordance with the provisions of 10 CFR 60.59. The changes will be discussed fully in Sections I and V.
No trainees war.,added to the reactor staff during 1991. Two former trainees cbtained Senior Reactor Operator licenses and one former trainee received his Reactor Operator license. One Senior Reactor Operator, Mr. Thomas Wright, d; parted curing the year. Also, one Senior Reactor Operetor, Mr. Stephen Holmes, chose license was terminated on 29 November 1990 was relicensed in 1991.
R: quests from non-AFRRI investigators continued to supplement the substantial inhouse experimental work load. These experimenters included representatives from the National Institutes of Health (NIH), Smithsonian Institution, National Institute of Standards and Technology-(NIST), Naval Medical Research Institute 2
_ - ~ - . _                  _.                      _ _ _ _ _ _ _ _                    . _ _              _ . . _ - _ _ _ _ _ _ _ , - _ .,          _
 
(NMRI), and the University of Waryland at Baltimoro. Tho reactor staff was also tasked with providing personnel to assist in conducting inspections of the Fast Burst Reactor facilities at Aberdeen Proving Grounds, Maryland and White Sands Wissile Range, New Wexico. At the request of Cornell University, an operations audit of their reactor facility was conducted.
The AFRRI documentation on financial assurance for decommissioning was accepted by the NRC on 11 February 1991.
No Licensee Event Reports were submitted during the year, however there were several malfunctions and unplanned shutdowns that are discussed in Sections III and IV.
The remainder of this report is written in a format to include notification items required by the AFRRI TRICA Reactor Technical Specifications. Items not specifically required but of general informational value are presented in the General Information section. Each section following the general information corresponds to the required section as listed in Section 6.6.1.b of the AFRRI TRICA Reactor lechnical Specifications.
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GENERAL INFORMATION Key Personnel Reactor and Radiation Facility Safety Committee.
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CENERAL INFORMATION:
All personnel held their poaltions as listed throughout the entire year unless otherwise specified.
: 1. Current key AFRRI personnel (as of 31 December 1991) are as follows:
Director - CAPT Rabert L. Bumgarner, USN (ef fective 30 August)
Chairmate, Radiation Sources Department - CAPT C. B. Calley, USN Chairman, Safety and Health Department - Mr. Thomas J. O'Brien and AFRRI Radiation Protection Officer      (ef fective 1 June)
: 2. Reactor Facility Director - Mr. Mark Moore (SRD)
: 3. Current key Reactor Operations Personnel:
Reactor Operations Supervisor - Capt Matthew Forsbacks, USAF (SRO)
(effective 31 May)
Training Coordinator - CPT Christopher Owens (SRO ef fective 7 May)
Walntenance/ Procurement - Wr. Robert George (SRO)                                '
Administration - MSO Harry Spence (SRO)
Other Senior Reactor Operators - Wr. John Nguyen (SRO ef fective 7 May)
Wr. Stephen Holmes (SRO ef fective 15 Oct)
Reactor Operator - SFC Wichael Laughery (R0 ef fective 7 May)
: 4. Senior Reactor Operator Candidates: None
: 5. Departures during CY 1991:
Mr. Thomas Wright (SR0 license terminated 31 May)
: 6. There were several changes to the RRFSC during the 19g1 calendar year. Mr.
Thomas O'Brien replaced Mr. Douglas Ashby as Chairman, Safety and Health D partment effective 1 June. CDR Joseph E. DeCicco of the Navy Dosimetry Center sorved as a regular member from 1 January until his retirement on 30 November.
Dr. Samuel Levine was appointed as a special member effective 9 December to assist in revlow of the fuel-follower concrol rod installation. Mr. John Dickson and Ms.-Leslie Moore of the Calvert Marine Museum were appointed as special members on 10 September to perform special projects related to reactor pool cater quality.
The 1991 RRFSC consisted of the following membership in accorcience with
~ AFRRI Reactor Technical Spe::lfications (as of 31- December):
Chairman: Col Nicholas Wanderfield,-USAF (Director's Representative) 4
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i Regular Wembers:
Mr. Thc. mas o'Brien (Chairman, Safety and Health Department, AFRRI)
Wr Mark Woore (Reactor Facility Director, AFRRI)
Dr. Marcus Voth (Director, Breazeale Reactor snd Professor of Nuclear    i Engineering, Pennsylvania State University)
Wr. Ron Luerson (Safety Directorate, Naval Research Laboratory)
Special Members:                                                        j CAPT C.B. Calley, USN (Chairman, Radiation Sources Dept)(Certifled HP)
Dr. Samuel Levine (SHL Nuclear Associates)(Nuclear Engineer-Reactor Specia list)
Mr. John Dickson (Calvert Marine Museum)(Water Quality Specialist)
Ws. Leslie Moore (Calvert Marine Museum)(Water Quality Specialist)      :
Non-voting Observer:
Mr. James Caldwell (EPA, Montgomery County, WD)
Recorder:
Ws. Carol King W:etings of the RRFSC were held:
20 February 1991 27 June 1991 (Subcomittee) 24 September 1991 17 December 1991 4
                                                                                                        \
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SECTION I Changes to the Facility Design Performance Characteristics and Operational Procedures.
Results of Surveillance Tests and Inspections.
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SECTION I Changes to the facility design, performance characteristics, operational procedures, and results from surveillance testing are contained in this section.
A. DESIGN CHANCES:
1,  As previously discussed in the Introduction section, three new fuel-follower sta dard control rods, a new air-follower transient control rod, and associated equipment were installed during the year. Also, the control rod drive support structure was modified to permit easier access to the transient rod drive mechanism for ma;at.enance. (Attachment E-1)
: 2. The pool water gamma activity monitoring system was redesigned to replace vacuum tube technology with a more reliable solid stato system (Attachment E-2). Also, the primary water conductivity readout modules in the control room were replaced by newer models. The sensors and overall systyn design remain the same and no changes to the flow of the primary coolant system were made. (Attachment E-3)
: 3. The air particulate monitor (CAM) automatic damper closure system was modified by adding an override circuit to allow an operator to open the reactor room air dampers while a CAM is alaraing. This temporary modification would allow controlled venting of the reactor room should a cladding failure occur when it would be necessary to pull air through the absolute filters, for example, during pulse testing of the new fuel-follower control rods. (Attachment E-4)
: 4. The steady-state timer originally installed on the new reactor console was a count-down timer. That timer was replaced with a count-up timer to facilitate determination of steady-state run lengths since runs are often terminated based on dosimetry measurements and not at predetermined elapsed times. (Attachment E-5)
: 5. The lead-acid oattery back-up system that previously provided standby power to the criticality monitor and radiation area monitors (RAM.) was replaced by a larger uninterruptible power supply (UPS) necessary for providing more stable current and decreasing maintenance requirements.
Replacement of thc lead-acid batteries also eliminates a hazardous chemical problem and hydrogen explosion hazard.
6.
Hardware and software modifications were made to the control console to install the new low pulso display option and new pulse mode scram timer (Attachment E-6). The pulse timer allows the operator to adjust the amount of time the transient rod remains up during a pulse as per earlier preapproved console installation plan. The new small pulse display option allows high sensitivity pulso data acquisition and a more readable display of pulse characteristics on the console CRT.
B. PERFORMANCE CHARACTERISTICS:
Installation of the fuel-follower control rods and readjustment of the reactor core loading resulted in an increase in the nominal control rod worths by an average of 80.40 (0.0028 Ak/k) and increased the k-excess measurement at infinite water by 10.67 (0.0047 Ak/k). The negative 6
 
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l temparate,o co3fficient of reactivity and the reactor tank constant were not affected. Spwific details of the installation and testing program will be presented in the Startup Report to be submitted as required by the reactor Technical Specifications o early 1992.
C. ADVINISTRATIVE PROCEDURES:
Several changes to the Reactor Administrative Procedures were approved and implementad during the year. A complete set of current administrative procedures is included at Attachment B.
Administrative Procedure Al was changed to clarify the recording requirements when an operator is using over-the-counter medications by requiring operators to check possible side effects in the Physician's Desk Reference, notify the RFD, and log the medication to a Drug Log.
Administrative Procedure A2 was changed to delete the escorted access roster for the exposure room prep area. All persons allowed in the area now receive trainirig in radiation safety to allow them unescorted access.
Administrative Procedure A3 was revised to simplify the 10 CFR 50.59 worksheets to implement recommendations from various NRC inspections.
Administrative Procedure A4 is a completely new procedure developed in conjunction with NRC inspection recomnsdations on inventory procedures for special nuclear matorial.
D. OPERATIONAL PROCEDlRES:
Numerous changes were made to the operational procedures to improve clarity and to account for the new fuel-follower control rods, new low pules display, and other facility modifications. The changes trade during the major revision in May 1991 at listed at Attachment F and subsequent 1991 changes are summarized below. A complete set of current operational procedures is at Attachment B.
: 1. Procedure 4, Personnel Radiation Protection, was revised to specifically indicate that AFRRI Instruction 6055.8, Occupational Radiation Protection Program, is the radiation protection program followed by the reactor.
: 2. Procedure 8, Reactor Operations, was extensively revised as follows:
: s. Basic: The basic procedure was revised to clarify that retpirator equipment is intended for use only during emergency conditicns, not on a routine basis. Changes were also made to record the SP.0 on-call at the beginning and end of each day or if a change occurs during the operational day instead of on each logbook page. Also, e new designation, physicist in ch-ege (PIC), is added to clarify the individual in charge physienliy present at the reactor.
: b. Tab A: The logbook entry chneklist was changed to show that SRO on-call and PIC entries would bs in black ink.
: c. Tab C: The nuclear instrumentation setpoints were adjusted for the new reactor control console and to delete references to the obsolete stack particulate monitoring system.
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: d. Tcb 0: The k-execes procedure was revised to change the standard power level for performance from 15 watts to 5 watts. This allows better readability of the power level on the reactor console linear power indicator.
E. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS:
All required maintenance and surveillar.et tems were accomplished as '
required. Malfunctions discovered ar e o. . a; lod in Section IV.
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SECTION                    II Energy Generated by Current Reactor Core and Number of Pulses
        $2.00 or Larger.
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SECTION II Energy generated by the reactor core:
MONTH                  KwHrs JAN                  1883.3 FEB                  3115.1 MAR                    4417.6-APR                  5993.6 VAY                  1747.0 JUN                  4715.9 JUL                  1575.3 AUG                  2303.6 SEP                  2078.8 OCT                  1343.0 NOV                  1509.4 DEC                  1370.9 TOTAL                32053.5 Total energy generated this year:          32053.5 KwHrs Total energy on fuel elements:          734168.6 KwHrs Total energy on FFCRs:                      1370.9 KwHrs Total Pulses this year 2 82.00:                None-Total Pulses on fuel elements 2 82.00:          4112 Total Pulses on FFCRs 2 82.00:                  None Total Pulses on fuel elements:                    9765 Total Pulses on FFCRs:                            None a
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SECTION III Unscheduled Shutdowns l
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i SECTION IfI Unscheduled Shutdowns:
There were four unscheduled shutdowns during this reporting period.
: 1. During a 1 MW steady-state run on 7 March, a reactor scram occurred when a high voltage drop in the NP-1000 channel exceeded the trip point of 10%.
Investigation showed that the trip point had been set more conservatively than the required 20% limit. The trip point was readjusted to 20% and operations continued.
: 2. An "NPP High Voltage Low" scram occurred during a 1 MW steady-state run on 1 August. The voltage circuit was tested and no problem-could be found.
The run was resumed and the reactor operated normally.
: 3. During a 1 MW steady-state run on 25 October, a reactor scram occurred because of a momentary loss of site electrical power affecting a broad local area. All required safety instrumentation continued to operate on back-up uninterruptible power supplies and normal power was restored in only a few seconds.
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: 4. During control rod calibrations on 4 December at a power level of 5                                i watts, a reactor scram occurred with a " Timer Scram" indication on the control console CRT screen. At the time, the timer scram was not enabled nor was the steady-state timer running. Testing of both the related console hardware and software did not indicate any cause for the scram and calibrations resumed with no further problems.
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SECTION IV Safety-Related Corrective Maintenance.
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SECTION IV Safety-related corrective maintenance:
The following are excerpts from the malfunction logbook during the reporting period. The reason for the corrective action taken, in all cases, was to return the failed (quipment to its proper operational status.
17 Jan 91    Problem: While performing k-excess, the console scrammed when the Auto mode button was pushed. A " Console Scram Button' message appeared on the CRT.
Solution: System was tested but the malfunction could not be made to reoccur. Reactor operated normally thereafter.
18 Jan 91    Problem: While moving the core from position 700 to position 250, a
                " Console Scram Button" message appeared on the CRT. The reactor was not at power at the time.
Solution: System was tested but the malfunction could not be made to repeat. Reactor operated normally thereaf ter.
05 Feb 91    Problem: While raising control rods to begin a steady-state run, the transient rod scrammed while being raised. No message appeared on the CRT , no other rods scrammed, and the transient rod drive did not drive down until the DOWN button was pressed.
Solution: System was tested but the malfunction could not be made to reoccur. Reactor operated normally thereaf ter.
04 Mar 91    Problem: While preparing for a steady-state run, an NM-1000 high power trip occurred and would not allow the rods to be raised. The CRT scram message could not be cleared.
Solution: NM-1000 scram clearing sequence was performed, but the scram would not clear until a complete console prestart check sequence was run. Reactor operated normally thereaf ter.
22 Apr 91    Problem: The console would not pass the automatic prestart check sequence due to a '0IS064 Timeout' message on the CRT.
Solution: A defective power supply in the CSC ccmputer expansion chassis was replaced, the system was tested, recalibrated, and returned to full service.
25 Apr 91    Problem: Criticality monitor R-5 alarmed at a steady-state power level of 70 KW, lower than expected.
Solution: The calibration was checked and appeared to have drifted slightly. The monitor was recalibrated with a known source and returned to service. The calibratire was checked periodically to ensure r.o further drift was observed.
06 May 91  Problem: During morning startup procedures, multiple scram messages, DOS error messages, and timecuts occurred.
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Solution: The problem was traced-to a possibility of excessively high humidity in the control room.-The ventilation system was adjusted and when the humidity was lowered below 804 all systems returned to normal.
15 May 91    Problem: The secondary air particulate monitor (CAM)-exhibited a
                      ~
very erratic trace while the primary CAM showed no fluctuation-in effluent radiation concentration.
Solution: The problem was traced to a faulty chart lamp. The lamp was replaced, the CAM was tested with a known source, and returned to service.
31 May 91    Problem: While raising control rods for k-excess measurements, the three standard control rods scrammed but the rod drives did not drive down. The transient rod did not scram and no message appeared on the CRT.
Solution: A loose relay contact.in the NP/NPP high voltage power supply was secured and the reactor operated normally.                  ,
10 Jun 91    Problem: A malfunction of the reactor area security system computer caused the computer to not accept keyboard comands and to not' respond to alarms. Reactor Facility Director notified as well as AFRRI security and logistics sections.
Solution: The reactor area backup system was turned on so that reactor-related alarms would appear on security monitors at the front desk which is mar.ned at all times. The priraary system was repaireo by a contractor and returned to full service on 12 July. No loss of security effectiveness occurred while the backup system was operational.
4 17 Jun 91    Problem: During the daily startup procedures, a large volume of air began escaping around the transient red anvil. The RFD was notified.
and an investigation begun.                                              1 2
Solution: The reactor steff disassembled the transient _ rod _ system and deteralnoa that a teflon seal on the piston 'had worked -loose due to the lack cf a required locking pin in the retaining collar.:This pin was left out by the-contractor during rebuilding.of the drive as part of the fuel-follower control rod installation program. A new taflon seal was machined and a.new locking pin inserted. Cleaned and lubricated piston and barrel. Removed, inspected, and replaced 0-ring seal:at bottom of barrel. Reassembled drive and performed. rod drop tasts and recalibrations as' required._ Notified General Atomics that locking pin had not' bean installed during rebuilding of drive _
assembly at General: Atomics factory.
02 Jul-91-- Problem:-When the console CRT-screen was 'reconfigured after.-_
displaying the graph of'a pulse, the digital linear power readout showed 2 mi'lliwatts while the CRT bargraph had a' scale of-0-1.2-milliwatts.
                  =_                                                        _
12
 
Solution: Trua rsector power determin2d to be 2 milliwatts based on independent monitoring of counts out of fission chamber. As soon as the power was raised enough to require a change of bargraph c,cale (to 10-120 milliwatts) the system operated normally. General Atomics notified of possible software problem in CRT reconfiguration program.
05 Jul 91  Problem: After completing automatic console prestert check sequence, a " Pulse Power High' scram message would not clear. While repeating prestart sequence in an attempt to clear the scram, the safe and shim rod drives started to drive up. There was no magnet power to the drives and all control rods remained in the core.
Solution: The console was tested and the malfunction could not be made to reoccur. The reactor operated normally thereafter.
09 Jul 91    Problem: None of the console keyboard function keys used for alternate CRT displays (pulse graph, accumulated operator time, etc) would work.
Solution: The console computer was reinitialized and all keys then worked normally.
16 Jul 91    Problem: While the reactor was secured, an 'NPP High Voltage Low" scram message appeared on the console CRT screen.
Solution: When the key was inserted into the console the scram message cleared. No cause for the scram could be found and the reactor operated normally.
03 Oct 91    Problem: While closing the ERI plug door, the door stopped moving and a loud grinding noise was apparent in the gear reduction box.
Solution: The ER2 gear box was moved to ER1 and the door operated with no more problems. The defective gear box was returned to the factory for repairs.
07 Oct 91    Problem: During prestart checks, a malfunction of'the DRIVE UP microswitch caused the safe rod drive to drive up instead of down when the reactor was scrammed. The magnet released and the rod itself dropped correctly. When the drive continued past the upper limit, the position indicator belt and flexible wire guide broke.
Solution: The drive was removed; belt, wire guide, and DRIVE UP.
microswitch replaced; drive assembly -inspected, tested, and-reinstalled. Rod travel was adjusted, rod drop time checked, and rod-worth curves checked.
NOTE: The reactor was out of service for the annual maintenance shutdown period 12-27 September. No malfunctions occurred during that time. The reactor was also taken out of service for installation of the fuel-follower control rods on 8 November. No malfunctions occurred af ter the reactor returned to steady-state cperation on 16 December.
\
13 I
 
SECTION V Facility Changes and Changes to Procedures as Described-in the Safety Analysis Report. New Experiments or Teets during the Year.
 
SECTION V Changes to the facility and procedures as described in the Safety Analysis Rsport and new experiments or tests performed during the year are contained in this section.
A. As previously discussed in the Introduction section, three new fuel-follower standard control rods, a new longer air-follower transient control rod, and related equipment were installed during the year, Also, the control rod drive support structure was modified to permit easier access to the transient rod drive mechanism for maintenance. P ktachment E-1)
B. The pool water gamma activity monitoring system was redesigned to replace vacuum tube technology with a more reliable solid state system (Attachment E-2).
Also, the primary water conductivity readout modules in the control room were replaced by newer models. The sensors and overall system design renain the same and no changes to the flow of the primary coolant system were made. (Attachment E-3)
C. The air particulate monitor (CAM) automatic damper closure system was modified by adding an override circuit to allow an ope:ator to open the reactor room air dampers while a CAM is alarming. This temporary mdification would cllow controlled venting of the reactor room should a cladding failure occur then it would be necessary to pull air through the absolute filters, for example, during pulse testing of the new fuel-follower control rods. (Attachment E-4)
D. The steady-state timer originally installed on the new reactor console was a count-down timer. That timer was replaced with a count-up timer to facilitate determination of steady-state run lengths since runs are often terminated based on dosimetry measurements and not at predetermined elapsed times. (Attachmer.t E-5)
E. The lead-acid battery back-up system that previously provided standby power to the criticality monitor and radiation area monitors (HAMS) was replaced by a larger uninterruptible power supply (UPS) necessary for providing more stable current and decreasing maintenance requirements. Replacement of the lead-acid batteries also eliminates a hazardous chemical problem and hydrogen explosion hazard. The installation of these items was approved along with two other UPSs that were installed in 1990. (See Attachment C-5,1990 Annual Report)
F. Hardware and software modifications were made to the control console to install the new small pulse display option and new pulse mode scram timer (Attachment E-6). The pulse timer allows the operator to adjust the amount of time the transient rod remains up during a pulse as per earlier preapproved console installation plan. The new small pulse display option allows high sensitivity pulse data acquisition and a more readable display of pulse characteristics on the console CRT.
C. There were no new experiments or tests performed during the reporting period that are not encompassed in the Safety Analysis Report. However, all previously cpproved Routine Reactor Authorizations were replaced by a revised set approved by the RRFSC on 24 September 1991. The revised Authorizations do not include any cxperiments that were not in the previous authorizations. The revision served primarily to update format and references and to obtain approval from the current RRFSC membership. A set of the Authorizations is at Attachment D.
14
 
l  - Attachments E-1 through E-6 are a summary of safety evaluations made for changes not submitted to the NRC pursuant to the provisions of 10 CFR 50.59. Each modification was described and qualified-using Administrative Procedure A3, Facility'Wodification. This procedure utilizes a step-by-step process to document the fact that there were no unreviewed safety questions and no changes required to the Technical Specifications.
15
 
SECTIONS VI through VIII Summary of Radioactive Effluent Released.
Summary of Radiological Surveys.
Exposures Greater Than 25% of 10 CFR Limits.
            >                                                    1
 
SECTION VI Summary of radioactive effluent released:
A. Liquid Waste - The reactor produced no liquid waste during CY 1991.
B. Caseous Waste - There were no particulate discharges in CY 1991. The total Ar-41 discharges in CY 1991 were 13.408 Curies.
On a quarterly basis: Jan - Mar 1991 49E9,9 mci Apr - Jun 1991 3139.5 mCl Jul - Sep 1991 4195.4 mci Oct - Dec 1991 11BS.4 mci C. Solid Waste -  All solid material was transferred to the AFRRI byproduct license; none was disposed of under the R-84 license.
SECTION VII Environmental radiological surveys:
A. The environmental sampling of soil, water, and plant growth reported radionuclide levels that were not above the normal range. The radionuclides that were detected were those normally expected f rom natural background and f rom long-term fallout.
B. The environmental monitoring (dosimetry) program reported the following results for CY 1991.
: 1. The average background of 19 thermoluminescent dosimeters (TLD) located outside a 15 mile radius from the AFRRI site was determined to be 84.40 2 1.69 millirem.
: 2. The average reading of approximately 30 environmental stations located on the AFRRI site was determined to be -0.45 2 0.31 millirem above background.
: 3. The single highest environ,wntal station reading was 14.39 6.07 millirem above background. This station is approximately 500 meters from the AFRRI reactor building.
: 4. The above results are expressed at a 95% confidence level.
C. The in-plant surveys, including analysis of effluent filters,-showed no measurable activity (except as reported in this section) in all areas outside the restricted-access areas.
D. Tnere were no special environmental studies cor. ducted during the year.
SECTION VIII Exposures greater than 25% of 10 CFR 20 limits:
There were no exposures to staff or visitors greater than 25% of 10 CFR 20 limits.
16
 
ATTACHMENT A 10 CFR 50.59 Analysis and Refueling Plan for Fuel-Follower Control Rods
 
I Facility Modification.Worksheet 1 10 CFR 50.59 Analysis Install Fuel Follower Control Rods Proposed Change Subinitt:d by:            Cape Forsbacka                                Date 23 OCT 91
                . Description of change:
Replace current standard control rods with fuel follower control rods (FFCRs).
A complete description of the FFCRs is included in the attached report " Max-imum Temperature Calculation and operational Characteristics of Fuel Follower Control Rods for the AFRRI TRICA Reactor Facility"(AFRRI TR91-1).
: 2. Reason for change:
FFCRs will be installed to overcome long werm burnup effects and increase the excess reactivity of the reactor core.
l
: 3. Verify that the proposed change does not involve a change to the Technical-Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.
This modification DoES INVOLVE A TECHNICAL SPECIFICATIONS CHANGE.                See l
attached approval from NRC. Full documentation is in Reactor File 605.06.
Analysis attached? Yes          X AFR21 TR91-1 is the safety analysis report regarding FFCR installation and use.
: 4. The proposed modification constitutes a :hanges in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).
!                Procedure X        Facility      'X            Experiment Revised: 15 May 91                                                                        Page 3
 
Facility Modification Worksheet 1
: 5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this description need not contain the final SAR wording.
The following sections in the SAR will require modifcation to reflect the installation of FFCRs:
                          - Section 4.9, Fuel Elements
                          - Section 4.10, Reactor Control Components
: 6. For facility modifications, specify what teating is to be performed to assure that the systems involved operate in accordance with their design intent.
See attached plan, "AFRRI TRIGA Refueling Plan".
Revised: 15 May 91                                                                Page 4
 
Facility Modification Worksheet 1
: 7. Specify sssociated information.
New drawings are:        Attached          x (Blueprints of FFcas are in che reactor blueprint file)
Does a drawing need to be sent to Logistics?        Yes      No x Are training matcrials effected?                  Yes _x_ No Will any Logs have to be changed?                  Yes      No  x Are other procedures effected?                    Yes        No s
List of items affected:
Training Materials:
                                  - New SRO Training Package will need modification to reflect presence of FFCRs in the core.
Maintenance Procedures:
                                  - Annual ehutdown checklist will have to be modified because FFCRs will not be brought out of the pool once they have been irradiated.
                                  - Elongatation and lateral bend will have to be measured on the FFCRs.
: 8.      Create an Action Sheet containing a list of associated work specified in item #
7, attach a copy, and submit another to the RFD.
Action Sheet:          Submitted      x                Not Required
                                                                                /
Reviewed and approved by RFD                /A                  -      Date N RRFSC Concurrence                                                        Date Revised: 15 May 91                                                                  Page 5 E_ __ .. _ . .    .
 
        '                                                                                                                                                                          l
              .        p* ~%
UNITED STATES I                            , ,. . c
                      !-                      ,I                      NUCLEAR REGULATORY COMMISSION o,                      c                                            WASHINGTON. D.C. 2005 i
s f                                                                      October 8, 1991 Docket No. 50-170 Colonel George W. Irving, III, BSC, USAF
!                        Director Armed Forces Radiobiology Research Institute Bethesda,- Maryland 20814-5415
 
==Dear Colonel Irving:==
 
==SUBJECT:==
ISSUANCE OF AMENDMENT NO.- 21 TO FACILITY OPERATING LICENSE NO. R-84' - ARMED FORCES RADI0 BIOLOGY RESEARCH INSTITUTE (AFRR The Cosmission has issued the enclosed Amendment No. 21Ito Facility Operating 4
i                        License No. R-84 for the AFRRI TRIGA Research Rear. tor.. The amendment :onsists of changes to the Technical Specifications in respnse to_your_ submitcal of April-30, 1990, as supplemented on December 17, 1990. March 5 -1991, May 17 1991, August 16, 1991, and September 10, 1991.-
The amendment (1) corrects errors in typography and grammar,_ (2) increases the maximum licensed steady state reactor power to 1100 kilowatts, (3) author-m izes installation of fuel follower control rods,-(4)-clarifies the transfer of
!                        Reactor and (5) allows        Facilityoperational Director (RFD)    flexibilityresponsibilities in performing  in-thesurveillance absence of testingthe RFD,f            o l
4                        the ventilation _ system for the reactor facility.
Enclosure 2 is a copy of. the related Safety Evaluation supporting Amendment
,                        No. 21.
!                                                                                                              Sincerely,
( ]Y-
                                                                                                                        ~
i fl                      .
jL                -
                                                                                                                                                  /-
Alexander Adams,- Jr., Project ; anager Non-Power Reactors, Decommissioning and:
Environmental Project Directorate
!                                                                                                              Division of Advanced. Reactors
                                                                                                                    -and Special Projects                  -.
Office of. Nuclear- Reactor Regulation
 
==Enclosures:==
 
1.- Amendment No. 21 2._      Safety Evaluation-cc w/ enclosures:-
See next page f                                                                                                                                                                                  I
}
  ., _ -n - .  + -n                "w,.-s,      4r-t+-  ,-e-m            e- w - r w-a---ew  . x y,a.-              -        <,c,-    w  r,    w-.y-    w  er , ,.- --
 
4 i
AFRR1 TRIGA Refuellag Plan Receiving and Storage of FFCRs Prior to Installation Prior to receiving the FFCRs, the Hot Cell will be prepared for fuel storage. The room will be reasonably free of dust and debris, a gamma detector (criticality monitor) will be in place, a high security lock will be installed, and cradles for holding the fuel will be available.
The FFCRs will enter AFRR1 through the shipping and receiving department (LOGS). A memorandum will be prepared (See attachment 1) to instruct LOGS personnel to not open the containers holding the FFCRs. LOGS personnel will be instructed to call RSDR and SHD immediately upon receipt of the FFCRs. SHD will perform a radiological survey in accordance with HPP-0 3 of external radiation and external contamination of the packaging in the loading dock area prior to moving the FFCR containers to the Hot Cell.
Once the FFCRs have been cleared by SHD to be moved to the Hot Cell, RSDR staff will conduct a series of tests within the Hot Cell to determine if any uranium is on the outer surface of the FFCR claddmg. The testing methodology is similw to series of measurements outlined in 10 CFR 70.39 which deals with the certification of caixation or reference radiation sources.
The following tests will I;c conducted:
: 1. Dry wipe test. 'lle entire surface of the FFCRs will be wiped witn fiher paper
  ) using moderate finger pressum. Any radioecovity on the fiker paper will be detenrdned by measunng the radaanon levels using SHD's counting lab.
: 2. Wet wipe test. The entire surface of the FFCRs will be wiped with filter paper, moistened with water, using moderate finger pressure. Any rachoactivity on the fiher paper will be determmed by measuring the radia:.lon levels using SHD's countag lab.
: 3. Water Boil Test. De FFCRs will be @y immersed in boihng water for one hour. The residue obtained by evaporating the water will then be moutored for the presence of uranium using the SHD en=*= lab.
If measurable quanuoes of radinactivity are present following any of the above tests, the FFCRs will be thoroughly elemaad and the tests will be iW. In the event that radioactivity is found after repeated tests and c62 g,            the shipment will be rW and returned to the manufacturer.
Following the cuumlatinn of mermful tests showing no contamination of the FFCR extenor cladding with uranium or other radioactive enatammanet, the FFCRs may be moved to storage in the Reactor Room.
I
 
l i
r                .                                                                                                                    !
l 4              Core Preparation 4
Prior to shutdown and FFCR installation, neutron activation foils / wires in will be used in ERI l            with the bare core configuranon at a fixed position one meter from the core. Determine neutron energy spectrw, save data for leer comparison with FFCR loaded core. Next, completely unload all fuel from the reactor core. Following the removal of fuel, the standard control rods I            and transient control rod will be rernoved. The rod drive support structure will then be modified as required to allow for the ease of transient rod drive maintenance. Next, the new transient rod .
will be installed.
!            The FFCR conrxctmg rods will be ret up to allow for the removal of FFCRs without bringing them out of the reactor pool. Once the FFCRs have been irrwliatad, they will become highly-radioactive due to their fission product inventories, so it will be important to keep them under wut.r for shielding purposes.
i installation of FFCRs into AFRRI TRIGA Reactor Core 4            Measuremaart of the requirej connecting rod lengths will be made by measuring the entire length of the ==aderd rods connecad to the barrel. Correcting for differences between the FFCRs and i            the enadard tods, the connacting rod lengths for the FFCRs will be determined. The Wag rods will then be asenchad a the FFCR and the entire nait will be insamilad into the reactor core, i
The connecting rods will then be attached to the banel assembly to complete the imenilmenon of                            i i            the FFCRs.
1
!            Following inenilmaina, the FFCRs will centered in their grid locations using the set screws for i            the berrel assemblies. The goal is to minimiae any rubbing of the FFCR as it travels up and down. Once the FFCRs have been centered, drop time tests will be j '.. ;4 to insure
!            compliance the technical specifications.
Refseling AFRRI TRIGA Ranctor Case A conservative approach to refaeling the teactor core will be taken. These instructions will suppnessent Reactor Oper=einmal Procedure VII. Once the critical loadmg has been achieved, l            excess reactivity will be n=elemaad using the transient rod until there is enough excess reactivity                        :
to na'wm rod worth curves. Excess reactivity will be detenmined aher each fuel loadmg step                                I until the optJetional coe83 erdion is achieved. Care will be takaa to ensure that the $5.00 mariamm allowed cIcces reactivity is not **caadad, Following Reactor Operational Proceduit VII, install the thermaramplad fuel alamamen and load                              i the B-ring. Place neutron sourse in source holder and BF, neutros detectors in F-7 and F-18, 1 hen load *lammaet for grid lacasinas C-1, C-5, C-6, C-9, C-10, C-11 D-14, D-15, E-18, E-                                l
;            19,F-22, and F-23. Loading these alanwmet will allow for the asutronic coupling of the FFCRs and the neutmn source, At this point the rods will be withdrawn as described in step 2.a. of the
 
i r
procedure, and the first subcritical multiplication measurements will be taken.
Complete loadmg the C-Ring, load D-ring elements 2, 6, 8,12, and 18, and repeat the suberitical multiplication measurements. Complete loading the D-ring, and load E-ring elements 5,6,14, 15; perform suberitical mukiplication measurements. Lead E-ring elements 1,2,8,9,10,16, 17, and 24; perform suberitical multiplication preearements. Load E-ring elements 2, 23,7, 11,13, and 20; perform subciiiical multiplication measurements. Lead the following sets of elements and perform suberitical multiplication maanirements after each step (note: this load
(
pattern may be modified to accommodate instrumentation or other items that may obstruct fuel loading): E-4 and E-12; E-21 and E-22; F-1 and F 2; F-3 and F-30; F-4 and F-29; F-5 and F-28; F-6 and F-27; F-26 and F-16; F-15 and F-17; F-14 and F-18; F-13 and F-19; F-12 and F-20; F-il and F 21; F-10; F-9; and F-8. This ined.ng pattern allows for the FFCRs to exercise a high infhwner over the neutron population while the core is still very subcritical.
Core Calibration Once the core has been loaded to the operational excess reactrvity, core calibrations will proceed in the same mannar as followag an annual shutdown. Differential and integral reactivity worth curves will be generated for each rod in core poetions 250,500, and 750. Up to this point all testing has been done at very low powers, so the fission product inventory in the FFCRs will be low. Since the prnhahility that the FFCR cladding may fail is highest shortly after innallahna, a senes of pulses will be perfanned to stress the FFCRs before the fission product inventory has much of a chance of building up. During this pulsing operation the water will be closely momtored for any fission fragments. In the event that fission fragments are found in the pool water, all activities will stop, the NRC and GA will be notified, and the leaking element (s) will be found and isolated.
A thermal power cahbration will be perfonned in the usual mammer. Next the power coefficient of reactivity curve will be generated followed by the reflection coefficient measurements in positions 250, 500, and 750. The neutron energy spectrum experiment in ERI and ER2 will be i@ to ensure that the characesr of the radiation field has not been snochfied.
I
 
'j 1
1 l                      Refacilag Checklist i
:                      Pre-shipment 4                                                                                                                              !
l
: 1. Clean Hot Cell.
r                      2. Inspect Hot Cell, ensure area is reasonably dust free.
l                    3. Inemil high security lock on Hot Cell.
;                    4. Ensure radiacon monstor in hot cell is functional.                                                    i
: 5. Prepare fuel cradles, install in Hot Cell.
.                    6. Send memoranda to SHD and LOG on receipt of FFCR shipment.
i 3
Receipt of shipment i
: 7. SHD performs radiological surveys of exterior of package in accordance with their procedmes, i                    8. FFCR packages are transferred to Hot Cell.
i                    9. Open FFCR shippie4 packages in Hot Cell.
l                      10. Place FFCRs in prepared cradles.
;                    Hot cell testing
: 11. Perform dry wipe test.
;                    12. Perform wet wipe test.
!                      13. Perform water boil test.
:                    14. Repest wet wipe test.
: 15. If all easts me ==ccm==41, FFCRs sany be moved to the Reactor Room for storage.
!                    Core preparation I
l                    16. Use neutron acavation wire set from Reactor Experiments to establish base line neutron j-                    energy spectrum using LAdasy's set up.
: 17. Uniond all feel from the core in accordance with Procedure Vll.
I
: 18. Modify the rod drive support structure.
FFCR instausdos
: 19. Instau new transiset rod.
;                    20. Fabricate commacong rods for FFCRs.
l                    22. Install FFCRs, insure that they are not " bottomed out" with they are fully down.
: 23. Measure FFCRs agaiset FFCR standard, record resuks.
I              -.        .        . .      - -        .-      .. _ . .      . ... - .. -        . _ . -            . . -
: 24. Reinstall FFCRs, install rod drive motors, and center FFCRs in their core grid locations.
: 25. Perform drop time tests to insure compliance with Technical Specifications.
Core refueling
: 26. Place neutron source in its holder, and place BF, or fission detectors in core grid locanons F-7 and F-18.
: 27. Load the thermocoupled elements irto core grid locations B-5 and C-6.
: 28. Complete loading the B-ring and load C-1, C-5, C-6, C-9, C-10, C-11, D-14, D-15, E-18, E-19, F-22, and F-23. Perform subcritical multiplication maniments.
: 29. Complete loading the C-ring. Im1 D-ring locations 2, 4, 6, 8,10,12,14,16, and 18.
Perform subcritical multipbcation inneenmemst.
: 30. Complete loading of D tmg, and lend E-ring elements 5,6,14, and 15. Perform subcritical muloplicadon --wements.
: 31. Load E-ring h 1, 2, 8, 9,10,16,17, and 24. Perform subcritical nmitiplication nietsurtNwnet.
: 32. Load E-ring elements 15,19, F-22, and F-23, place neutron source into its holder, and perform subetincal multiplication enessurements.
: 33. Load E-ring elan ==en 2, 23, 7,11,13, and 20. Perform subcritical muinplication mensmements.
                                                                                            '^
: 34. Load E rieg alaama*= 4 med 12, and perfona subcritical                                -- ir=ba enemrements.
: 35. Load E-ring ela===*= 21 and 22, and pwform subcritical andtiplication numenements.
                                          **** 140TE ****
When critical coanguration is achieved, estimase excess reactivity using the transient rod. Fa=*ia=* loading until there is enough excess reactivity to perform a control rod worth measurement
: 36. Load F-ring enemments 1 sad 2, perform subcritical - ti&r'wlexcess reactivity                  -
inensurements.
: 37. Imad F-ring .                3 and 30, perform subentical >==lriplicatinalexcess reactivity mensarements.
 
J l
1      38. Imd F ring elements 4 and 29, perform suberitical multiplicadon/ excess reactivity        l measurements.
: 39. Load F-ring eicments 5 and 28, perform excess reactivity measurements.
: 40. Load F-ring elements 6 and 27, perform excess reactivity measurements.
: 41. Load F ring elements 16 and 26, perforin excess reactivity measurements.
l l      42. Load F-ring elements 15 and 17, perform excess reactivity measurements.
.      43. Imd F-ring elements 14 and 18, paform excess reactivity measuremcots.
                                                  * * *
* NOTE * *"
Do not exceed $5.00 excess reactivity!
          . . . . . . Per.,xin thamal power calibration at 500 to insure proper placement of operational channel. Perform rod worth curves for all rods in position 500 when opermeinmal loading is achieved
: 44. Imf F-ring elemanet 13 and 19, perform excess reactivity measurements.
: 45. Imd F-ring ah 12 and 20, perform excess reactivity measurements.
: 46. Load F-ring elements 11 and 21, p L- excess nectivity inessurements.
i l      47. Load F-ring elements 10, perform excess reactivity measureements.
: 48. Load F-ring elements 9, perform excess reactivity measurements.
: 49. Load F-ring ch 8, perform excess rescarvity . .                , . . .  .
.      Core calibration
<      $0. Install all cora instrm=nammina into their permanent positions.
i 4
: 51. Perform rod worth curves for all rods in core poutions 250, 500, and 750.
: 52. Perform thennel power calibration and s2 safety chambers,
: 53. Fire $1.10 pense, take pool water sample, manknr for finnan fragsments.
: 54. Fire $1.20 pulse, take pool water u:.nple, =nnianr for fission fragments.
: 55. Fire $1.30 pulse, take pool water sample, mandar for fission fragmesus.
: 56. Fire $1.40 pulse, take pool water sample, mouitor for fission fragments,
: 57. Fire $1.50 pulse, take pool water sample, monitor for fission fragments.
: 58. Fire $1.60 pulse, take pool water sample, monitor for fission fragments.
: 59. Fire $1.70 pulse, take pool water sample, monitor for fission fragments.
: 60. Fire $1.80 pulse, take pool water sample, monitor for fission fragments.
: 61. Fire $1.90 pulse, take pool water sample, monitor for fission fragments.
: 62. Fire $2.00 pulse, take pool water sample, monitor for fission fragments.
: 63. Fire $2.00 pulse, take pool water sample, monitor for fission fragments.
: 64. Fire $2.00 pulse, take pool waar sample, maniw for fission fragments.
: 65. Fire $2.00 pulse, take pool wanar sample, maniw for fission fragments.
: 66. Fire $2,00 pulse, take pool water sample, maniw for fission fragments.
: 67. Fire $2.00 pulse, take pool water sample, maaiw fw fission fragmeans.
: 68. Fire $2.00 pulse, take pool water sample, monitor fw fission fragments.
: 69. Fire $2,00 pulse, take pool wahr sample, mnaiw for Assion fragments.
: 70. Genersee power coefficient of reactivity curve, take pool water sample and monitor fw fission fragments.
: 71. Fire $2.00 palme, take pool water sample, monitor for fission fragments.
: 72. Opersee reactor at 1.0 MW fw 30 minness, take water semple and amaitor fw fission fragments.
: 73. Fire $2.00 paise, take pool water sample, monitor for fission fragments.
: 74. Fire $2.00 puhe, take pool water maple, manieor fw fission fragsments.
: 75. Repeat neutron energy spectra measurements in ENand ER2.
: 76. Perform messarements of reflecaor coeffbeients in panisinsu 250,500, and 750.
I i
 
e ContrW NurnDW sunlect                                                                                                  "YtOlE""      8"*""" N / A Installation of FFCRs                                                                  T4Te' 23 OCT 91 Action neewired RFD APPROVAL AND CONCURRENCE
    -- wome, noum for neeere. iosecnee trosy the requeements eac=grouno ano acton tamen or recommenoeo uust ce su            e cen'ty cetaaea to
      'dentity tBe 4Chon w@out f9 Course to other sourtee )
: 1. Request that we plan to install FFCRs -in mid to late November,1991 in a:cordance to fuel reloading plan attached.
: 2. Training caterials will require an audit to ensure that they will reflect the reactor configuration with FFCRs installed.
: 3. Formal maintenance procedures involving FFCR surviellance should be developed in the next few months.
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TECHNICAL REPORT                                                                                                                                                          l I
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;                                                              and Operational Characteristics
,                                                              of Fuel Follower Control Rods for the AFRRiTRIGA Reactor Facility
            ;                                                                                                                                                  ^r . < '                  '                          :
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Contents IHtroduCtIon...............................................................................................................I General Deserlption of Fuel Follower Control Rods..............................................1 F F CR Muimum Fuel Temperat ure Calculat10n....................................................                                                            2 Power Density in F F CR F uel Elemen t ................................ ........ .... ......... ...... 3 Maximum Temperature in FFCR Fuel Element............................................. 4 Fuel Temperature in Pulse Mode Operation............................. ...................... 0 F F C R O per atlon al C h ar ac t erls t le s .. ..... .. . . . .... . .. .. . . ........ .... ... . . .. ... . .... .... . .. ... ... . . .. .... . . . . 8 Conclus10n..................................................................................................................9 NeferenCes..................................................................................................................9 Appendix A:        Determination of Free Convective Heat Transfer Coefncient.....11 Appendix D:        Reactor Core Loading and Unloading ..........................................15 111 1
 
i INTRODUCTION                                                                                                                          l Operational requirements of the Armed Forces Radlobiology Research                                                                    !
Institute (AFRRI) TRIGA reactor facility necessitate the implementation of
                                                                                  . Fuel follower control rods are like the                                    -
4 fuel standard      follower TRIGAcontrol    rods control rods    as(FFCR's) described in section 4.10.1 of the TRIGA Safety Analysis Report (SAR) except that they have a fuel filled                                                                ,
follower ratl.er than an alt or aluminum follower. The primary purpose of the FFCR's is to offset the long term effects of fuel burnup.
requires that The modificationsCode ofofFederal a portionRegulations (s) of a licensed        (CFR;  Titleas10, facility,            Part 50.50)in _the facility described SAR, be documented. with a written safety analysis. The SAR ensures that all safety issues associated with the implementation of FFCR's have been reviewed. This technical report will show that implementing FFCR's will allow the standard control rods to function in their intended purpose and will restore core reactivity economically. FFCR's have been implemented in approximately a dozen TRIGA reactors and have been used for over 20 years without reported failure.                                                                                                        -
This report has been submitted to the AFRRI Radiation Facility Safety Committee to ensure that all safety questions have been reviewed before submission to the U.S. Nuclear Regulatory Commis;lon (NRC), as required under 10 CFR 50.50.
GENERAL DESCRIPTION OF FUEL FOLLOWER CONTROL RODS The current AFRRI TRIGA standard control rods were installed in 1964.                                                                  -
The standard control rod consists of a ser. led aluminum tube (0.065 inch thick) approximately 1.25 Iriches in diameter and 31 inches long. . The upper 15.25 inches of the tube contain a compacted borated graphite rod D with 25 percent free ~ boron or other boron compounds), which functions                                                      a  (as C neutron absorber or poison. The lower end of the tube contains a 15,25 inch long and 1.125-inch diameter solid aluminum rod called the aluminum follower. The follower functions as a mechanical guide for the control rod as it is withdrawn from or inserted into the reactor core.
The proposed FFCR's differ from the current standard control rods in the following respects:
* The aluminum cladding is replaced by smooth stainless steel fSS304) cladding with a wall thickness of 0.020 inch. The inner anc                                            outer-diameters are 1.085 inches and 1.125 inches, respectively.-
* The length of the control rod is increased to 37.75 inches; the absorber
                                  . and fuel follower section are both nominally 15 inches long.
* The outer diameter of the absorber section and the fuel follower are both 1.085 inches.
* The fuel follower has a solid zirconium rod as its central core with an outer diameter of 0.225 inch.
l
    - _ _ _ . _ . , ,                            x_.._.._      _ _ . _ _ . . _ . , _                                .. _ _ _ ..        . _ . _ , _ _ . .
 
The absorber or poison material of the proposed FFCR's is, however, identical to the standard control rods presently installed.
) The fuel contained in the FFCR consists of a fuel moderator element in                              '
which zirconium hydride is homogeneously mixed with partially enriched
! uranium. The FFCR fuel element contains 12 percent uranium by weight and has a nominal entlehment of 20 percent in the 235U isotob>e. The FFCR fuel element contains about 30.0 grams of ::sU this is 79,o of the 335U loading of a standard AFERI TRIGA fuel element. The nominal hydrogen-to-zirconium ratio in the FFCR fuel element is 1.7 with a range between 1.0 The and 1.7. The FFCit fuel element contains no burnable poison.
stainless steel eladding on the FFCR fuel element has a hardness greater than the aluminum control rod guide tubes, so wearing will occur on the guide tubes rather thsn on the FFCR fuel elements.
FFCR MAXIMUM FUEL TEMPERATURE CALCULATION A thermal hydraulle analysis of the FFCR fuel element to determine the maximum fuel temperature uses the following model:
* The neutron mean free path for neutrons of all energies is smaller than the diameter of the TRIGA fuel rods, so the reactor must be treated as a heterogeneous reactor. Thus, the active volume of the core is taken to be the volume of fuel contained within the reactor core.
* The ratio of power in a fuel element with 12 wt % uranium versus 8.5 wt-% uranium is 1.21. This is determined by General Atomics design calculations.8                                                                                '
* The reactor is operating at a steady state power level of 1.0 hiW, and the heat flux across the fuel element is described by Fourier's law of thermal conduction:8 q''(r) = -kVT(r)                                                              (1) where q"      = heat flux at por! tion r k =(r)t hermal conductivity T(r) = temperature at position r For steady state heat transport, the heat production rate and the rate of energy loss due to heat transport are equal. This can be generally expressed as q' ' ' (r) = V* g' ' (r)                                                              (2) where q"'(r) = volumetric heat rate (heat production rate) at position r.
Substituting      equation equation of thermal                        (1) ion;into equation (2) yields the time independent conduct q' ' ' (r) = -Y* kVT(r)                                                                (3) 2 l
 
Equation          is, thus, the second-order ordinary differential equation that must be so(3) lve d to determine the maximum temperature attained in the fuel portion of the FFCR.
Using this model to determine the maximum fuel temperature divides the analysis into two separate tasks: determining the power density in the FFCR in a D ring grid position and solving equation (3) for the given power density.
Power Density in FFCR Fuel Element The anticipated fuel loading for the AFRRI TRIG A reactor core with FFCR's installed will consist of 77 standard Tn!GA fuel elements and the three FFCR fuel elements. Presuming that the control rods are fully withdrawn to achieve a power level of 1.1 MW, the total active fuel volume will be 30,507.9 cm 3. Thus, the average power density at 1.1 MW will be 36.0 W/cm3.
The maximum fuel temperature is the itnportant parameter, so only the radial variation of the core centerline power density is considered. To determine the maximum power density in the D ring location of the FFCR fuel element, the following calculations are made:
For the AFRRI TRIGA, the radial and axial peak-to average power ratios are 1,55 and 1.30, respectively.3        Thus the maximum power density (heat rate) will be q' ' ', = (1.55) (1.go) q ' ' ',y,                                                    (4)
                                      = 72.4 W/cm To determine q"'om relative to q'"                            it is useful to compute a scalin6 factor from the gro,s,'b variation of TI[e,rmal neutron flux in the radial normalized radial flux distribution for the AFRRI TRIGA core i)sdirection                  best (the represented by a Dessel function of the first kind of order zero:
ftherm " O (2.405r    R        }                                                      (}
e where R* = 21.78 cm, the extrapolated core radius r = 11.99 cm, radial position of D ring element and the Bessel function scaling factor is J, (1.3240) = 0.6074.
The power density for thn D ring is thus computed to be 3
q' ' D-ring = (0.6074) q' ' '        = 44,0 W/cm                                      (6)
Because the FFCR fuel element differs from the standard fuel element in concentration of uranium, the power density in an FFCR fuel element is 3
 
greater than the power density in a standard fuel eleinent by a factor of 1.21.8 Taking the above scaling factor into account, the power density of an FFCR fuel element is found to be q' ' 'FFCR = (1.21) q' ' D-ring                                                              (7)
                = 53.2 W/cir.3 Note that the calculation of q'"                                takes into account the most limitin condition for power peaking in a [Ov/o fuel element versus an 8.5 w o fue1 element.
As developed in equation
                                                                  , q"'r    is still considerab
* less as (7)determi'neS in equation 4.      less than the theoretical roachmaximum would have                  q"'lso a accounted for the reduce (d ) volume in conservative an FFCR fuelapp!cment.
e Maximurn Ternperature in FFCR Fuel Eleinent Equation 3 azimuthal (sy)mmetry (see Figure 1): takes the following form for cylindrical geometry with 1[dr.dr)+q'''=0 r
k                                                                                      (8)
The boundary conditions required that constrain equation (8) are as follows:
dT
      - = 0 at r - Rg and T = Tg at r = Rg                                                          (9)
J
                                                            ,      Te\ ,
                                                      '~
                                                              +Tj E        .Tg d,
Figure 1. Cross section of FFCR fuel element.
4                                          i
 
Integrating equation (S) twice yields a general solution of the form 2
T (r) = -q' ' '                  -
                                                                + Cg in(r) + C2                                                        (10)
Applying the boundary conditions to solve for the temperature difference between the outer edge of the rircoelum rod and the inner surface of the cladding, q'''RI 2 O g=                                                                                                          (11) 2k g C2=Ti-                                (In(Rg ) -              )
2 q'''R          R                        R I
Tg - T, s                              [( )2 - 21n ( ) - 1)                                                  (12) f            1                        i To account for the transfer of heat from the fuel through the cladding to                                                        i the coolant, we must consider the heat conduction between the inner and and the heat conduction from the outer outer surfacesurfaces of the cladding      of thetocladding, the coo qaTa,nt, qWe make the assumption that no heat is produced in the cladding or SN3e coolant, so the heat conduction and gnaa*
from  the conduction The heat            outer surface            of the leaving      the fuel  fuel,is qua,iven g                bymust be q,i,,
equal      to 2                                                2 ,,'
qfuel = r(R,2 -R g )g4,,, = r((R,/R g)2 - 1)LR g q                                                            (13) where L = length of fuel element dT                2rke L(T -T) 9 clad = -kch                                                                                                  (14) dr    clad                  R  +c In(                    )
R, qfluid = W T, - Tf ) = 2r(R, + c)Lh(T                                          c -T)f Note that the area, A, used in computing qa,, is the logarithmic mean area of the cladding.                    Recall that 9 fuel " 9 clad " 9 fluid                                                                                      (15)
So we can solve for the temperature differences in the above equations in terms of q"':
5 i
 
2
                                        +c                R Ri R
l          T    -T c = In (-                    ) ((J)2 , g)                            q,,,
R                    R                        2 I                                                      (16)
R E)2 , g) Ri
* q,,,
1 T    -
C f = (R,+c)h ((R              g 2
with equation (12 Adding expression  equations for the max (16)imum temperature                            in the FFCR. ,) and solving for T, gives us the I
l q,''R1 2. R                                R (E)2 - 21n( 2) - 1                                                          (17)
Tg=Tf+          4k g      ,R g                          R                ,
1 2                                                        +c q'''RI                R                            1        R                1 l                        +
[(R 2)2              -1)(k,          In(          )+              )
2                      g                                  R,        h(R, + c.) ,
where T i = maximum fuel temperature T, = bulk coolant te'nperature R' = 1.38 cm (radius of FFCR fuel element) eR=      0.280
                  = 0.051    cmcm        (radius (cladding                      of Zr rod))
thickness k    = 0.18 W/cm 'C k* = 0.138 W em.'C(thermal conductivity of UZril)'
g h    = 1.339 W cm .'C (thermal conc'uctivity of SC304)8 3
q"' = 53.2 W/ m 3 (from(free                  equation (7))          convective heat transfer                  coefficien Note that the free convective heat transfer coefficient, h, was an l      experimentally derived quantity. The method by which h was determined is presented in Appendix A, Solving equation (17) using a volumetric her.t rate of $3.2 W/cm3 and a bulk water temperature of 48.0'C (the conditions at which h was determined) yields a maximum fuel temperature of 210.2'C.
The maximum temperature achieved in the FFCR is nearly 180*C less than the normal temperature of 390'C in a standard fuel element in the B-ring during a 1.0 hiW steady-state power operation.
Fuel Temperature in Pulse Mode Operation The Nordhelm Fuchs model predicts the maximum fuel temperature achieved in a pulse mode operation 5
The fundamental assumptions of this model are as fohows:
* The neutron flux in the reactw la separated into a spatial component (shape factor) and a time-dependent component (amplitude factor), such that f(r,t) = vn(t)t(r)                                                                                (18) l l
6
 
  ,                                                                                                                                                                l where v = neutron velocity n(t{ = neutron density (amplitude factor), proportional to power f(r, = shape factor The shape factor is assumed to remitin constant during a pulse.                              This is called the point reactor model.
* The production of delayed neutrons and the effects of source neutrons                                                l are neglected,
* The pulse from a thermodynamic standpoint is adiabatic, so dT
                                                            = Kn(t)                                                                            (19) dt
                                                      $vhere T = fuel temperature K = reciprocal of heat capacity From the first and second assumptions we can write the time dependent neutron density as dn          p-/
                                              - a                      n (20) dt            E where R = mean lifetime of neutrons in the reactor p = delayed neutron fraction p = reactivity                                                                                                      ,
To account for a step insertion of reactivity, we can write -
p*p y      -
uaT (21) where a = negative of the temperature coefficient of reactivity pg= step-Insertion of reactivity Taking the derivative with respect to time of the above equation and
!                                      substituting the result from equation (18),
                                                    = -aKn                                                                                    (22) i dt Applying the chain rule to equation (20) yields d" . . (# ~ 4) dp              aK1                                                                                (23) 1                                                                                                  7 1
i e, -        g  .
                                                                      -9.+w. e. 9 --*w    ,              w-., r. , ,        --w w- .--. y-    .--- .y
 
i I
l l
Integrating equation .(23) and solving for the constant of integration gives us the result
                                                                                                                                                                                                      -i 1
n=            ((p, - /)2 , (p , p)2]                                                                                                        (24) i                                              2aK1 f                                  The pulse is terminated when n becomes negligibly small. This occurs when p=2p-p,                                                                                                                                    (25)
Equation (25 gives us the condition for the total energy release from the pulse, which ) manifests itself as a temperature rise in the fuel element w it is substituted into equation (21).
I 2(po - #)
core, ave
                                                            "                                                                                                                          (0) a Calculations by General Atomics show that a complete core of 12 w/o fuel erature coefficient of reactivity, a, that is- 75% of the would for value  have ana8.5 tempw /o fueled core.1 The value for a                                                                          is taken to be
                                    - 80.0118/' C.      (This  value        was          experimentally                      verified,,w,dh                              a series of 23 of pulses ranging C-within  from  $1.30 8.5%    of the    to $2.00-published  thatvalue. resultedThe                in an      effect averafe    o add  a'm,,g,Niree 80.0128/'FFCR's -would, however, have a -neg)ligible effect on the overall 12 w/o-temperature wellicient of reactivity for the entire core.1
;                                  Applying the Nordhelm Fuchs model for self limiting power exultsions, we                                                                                          ,
can determine the maximum average increase in temperature for the entire core using equation (26). For a maximum allowed reactor- pulse with a i
84.00 step insertion of reactivity, the maximum attained average temperature rise is calculated to be 333'C. Applying the power peaking factors from- the
!                                    previous section, the maximum calculated temperature rise in ' an FFCR
                                                                    ; the temperature rise for _ an FFCR for a $4.00 pulse would    be 1.48'AT is calculated    to IEc "83'C. - Assuming an initial temperature of 25'C, the maximum temperature value would be 518'C. Note that even in the.
limiting case, neither the technical specification safety limit of 1000'C nor j                                    the limiting safety systene setting of 600'C is violated.
FFCR OPERATIONAL CHARACTEPISTICS FFCR's are a standard design offered as a stock item by General Atomics and have been used in several TRIGA- reactors for over 20 years. FFCR's are currently implemented in approximately a dozen TRIGA reactors.- There                                                                                        '
has been no reported evidence of fuel failure as a result of FFCR use in the
                                    ' United States, . The operational issues- to be resolved are the effects of-l                                    burnup' on the - FFCR and- the influence of FFCR's on the temperature -
coefficients of reactivity, shut-down margin, and rod worth.
FFCR- control rod worth curves-.will be generated the same way that standard control rod worth curves are,                                        Since the poison section of the FFCR will be the same as that of the currently installed standard control 8
l _. . _ _ _ _ _                          .
 
        -    .                .-      -        .=      _                .        - -
rods, the worth of the poison section of the FFCR will be that of the currently installed control rods. Measurements made by AFRH1 reactor staff of control rod worth of the currently installed standard control rods yielded a nominal rod worth of $1.90. The fuel follower is expected to add at least (0.70 of reactivity' when the control rod is fully withdrawn, so the total rod worth for an FFCR is estimated to be $2.00. The transient control rod and its follower have been measured and have a total nominal worth of
    $ 4.01. The shutdown margin, as established by ANS/ ANSI 15.1, is computed as follows:
Total rod worth              $11.81 k , y , ,, (maximum)      -  $ 5.00
                                                        $ 0.S1 Worth of TRANS rod        -  $ 4,01 Shutdown margin              8 2.80 The shutdown margin with the most reactive control rod removed from the reactor is $2.80--well in excess of $0.50 minimum allowed value.
Once operational rod worth curves are established and power monitoring channels have been calibrated by the thermal power calibration method, power coefficient of reactivity curves will be generated. The issues regarding the measurement of shut-down margin and excess reactivity are addressed in Appendix D. Reactor Core Loading and Unloading.
Structural changes in the FFCR's will be monitored on an annual basis as part of the annual shutdown and maintenance. Specific effects to be monitored are the elongation and lateral bending of the fuel. FFCR fuel elements that have an elongation greater than 0.100 inch or a lateral bend greater than 0.0025 inch will be removed from service.
CONCLUSION The analysis in this report shows that installing FFCR's in the AFRRI TRIGA reactor core will not result in an unsafe condition or violation of technical specifications. - The primary parameter of interest, the maximum fuel temperature, was computed to be 210'C in the limiting case for steady-state operation and 518'C in the limiting case for pulse operation.
Operational issues regarding maximum excess reactivity, shutdown margin, and burnup have also been addressed, and it has been determined that sufficient surveillance capabilities exist to prevent any unsafe or illegal condition.
REFERENCES 1.
General Atomics, letter to M. Moore on fuel follower control rods, 28 October 1988.
: 2. Ei-Wakil, M. M., Nuclear Heat Transport, The American Nuclear Society, Lagrange Park, IL,1978.
9
 
t e
1 4                        3.        Defense Atom t e Support Agency, ATRRl/USAEC Facility License R 94, Complete teith Applications and Amendments, Dethesda, MD,1902.
;                        4.        Wallace. W. P., and Simnad, M.                            T.,  Afetallurgy of TRIGA Fuct Elements, G A 1949, General Atomics, San Diego, CA,1901.                                                                                                                  l i
;                        5.        Hetrick, D. L., Dynamics of Nuclear Reactors The University of Chicago Press, 1971.
!                          6.        DNA Contract DNA001-80 R 0030 to General Atomics for fuel follower j                          control rod construction.
)                          7.        Jaluria, Y., Natural Convection Neat and Afass Trans/c r, Pergamon Press, 1980.                                                                                                                                                              j i
                                                                                                                                                                                                    )
t i
l 2
i i
i s
F N
i
                                                                                .10
 
APPENDIX A: DETEllS11 NATION OF FRFE CONVECTIVE IIEAT TRANSFER COEFFICIENT Introduction We can measure the bulk water temperature within the AFRRI TRIGA core to determine the average free convective heat transfer coefficient of the cooling water. This experiment involves inserting a temperature measuring                                                                                        t probe between the D. and C-ring fuel elements while the reactor is operating at a steady state power level of 1.0 h!W and measuring the water temperature at various axial positions. Once the bulk water temperature has been determined, Newton's law of cooling can be used to calculate the average free convective heat transfer coefTicient.
Experimental Apparatus and Procedure The equipment used in this experiment consists of two approximately 18 foot lengths of chromal alumel thermocouple wires fused together at one end, encased in a 10-foot long, 0.375 inch diameter aluminum (Al) tube, and the thermocouple display readout on the AFRRI computerized reactor control
* console (Figure A 1).
The potential difference generated at the thermocouple junction as the water is heated by the reactor is amplified and- displayed by the thermocouple circuitry in the AFRRI computerized reactor control console.                                                                          The thermocouple is initially inserted into the core to correspond to position I.
The thermocouple resides in each region for several minutes to allow it to attain thermal equilibrium. Once thermal equilibrium is attained, ten temperature readings are taken at 10 second intervals.                                                              After each temperature measurement, the thermocouple is withdrawn to the next position, and the temperature measuring procedure is repeated.
Figure A 2 shows that the temperature is messured in five axial' positions:
(I 3 inches below midpoint-(14 inches of thermocouple wire inserted- into midpoint in axial dimension; (III halfway between midpoint th)e core); (II) f graphite slug; (IV) at top of ) fuel re and bottom o                                                                                                            .
above top of fuel region.
Thermocouple junction To new conde
                                          =            G  -.
Al tube s
Figure A 1. Experimental apparatusc 11
 
                                                              ~                        ~
4      s
                                                              -                    4      IV 4        lit
                                                                -                <;~        n 4        i H;0 Figure A 2.        Axial measuring points.
Safety Considerations There are two safety considerations associated with this experiment: radiation streaming and an unintentional positive change in reactivity if the thermocouple wires are rapidly withdrawn from the reactor core while it is at power.        Radiation streaming is avoided by flooding the aluminum tube with water and bending the tube so that it is at an angle not normal to the top of the core. The thermocouple wire displaces only 0.043 imh3 of water when it is fully inserted in the core, so using the void coefficient of reactivity, the thermocouple wire represents a negative reactivity insertion of only 0.001 cents.            If we were to estimate conservatively that the thermocouple wire had the same neutron-absorbing properties of a control rod, the maximum negative reactivity would be only 0.01 cents. Thus, there is no possibility of a reactivity accident associated with the apparatus used in this experiment.
Data Table A-1 summarizes the data gathered during a 1.0 MW steady-state run of the AFRRI TRIG A reactor.                                  The variation in the temperature measurements is most likely due to variance in the radial position of the temperature probe in the channel.
Table A 1. Bulk Water Temperature at Each Axlal Position in the AFRRI TRIGA Reactor Core Axlal                      inlet                              Measured core position              temp ('C)                            bulk water temp ('C)
I                      22                                      72.9 11                        24                                      65.0 111                        25                                      48.6-IV                          26                                      51.6 V                          27                                      59.7 l
l l
12
 
Analysis and Conclusion The purpose of this experiment is to determine the bulk water temperature within the core shroud; thus, it is the iowest measured value of the water temperature that is sought. Figure A 3 illustrates the temperature variation within a cooling channel.
Table A 1 shows that the measured value that most closely represents the bulk water temperature within the core shroud is 48.0'C.
The free convective heat transfer coefficient, h, is found by solving equation (8) for boundary conditions given by a standard TRIGA fuel element.
Equation (A-1) gives the solution in terms of h.
                                              ,,,      2  ,                    ,                  .-1 (Tg-T)-g                      (    )2 - 21n( )-1 1                      4k g        . rg        rt      1      r In( , + c,)
h=(            )                                                    -
(A-1) q'''r(
r, + c, E        T 2,[                      c        o 2      , rg            ,
where Ti    = measured fuel temperature at 1.0 MW T,    = mea.sured bulk coolant temperature in the core r, = fuel outer radius,1.810 cm ri    = fuel inner radius, 0.229 cm e, = cladding thickness, 0.051 cm k, = thermal conductivity of fuel, 0.18 W/cm.'C k' = thermal conductivity of clad, 0.138 W/cm *C qi " = volumetric heat rate.
l'he measured fuel temperature in the B ring at 1.0 MW steady state power level is 300'C, and the calculated volumetric heat rate is 65.9 W/cm 3.
Using the measured value  3 of the bulk coolant temperature of 48.0'C yields a value of 1.339 W/C'-cm for the free convective heat transfer coefUcient.
a
                                                        '\            r 5            W i
1 w
                                                                ,p    .-
Figure A 3. Ternperature variation within a cooling channel.
13
 
l
                                                                                                                                                                )
Newton's law of cooling expresses the linear relationship between the heat transfer rate, Q, and the temperature difference between the clad surface temperature, T,, and the bulk water temperature, T,, as Q = hA(Tf - T,)
where h is the overall convective heat transfer coefficient and A is the area of the fuel element.' The value of h determined for the AFRR1 TRIGA is unique in that it takes into account the flow configuration, fluid propertice, and the dimensions of the fuel elements. Assuming that the dependence of Q on the temperature difference T - T h, computed using data from B rin,g                              ents  elem,is      roughly with their  higherl'near,  heat transfer      then the value o rate and local temperature difference, will be close to the value for h for D-ring positions, t
t l                                                                                14 1
: r.    , ,
c    ,    - , . . , - -      ,... - , . , - - , - - - . ,                ..          -          n-,,.  ~ ,, - - - , - -  ~cn,- ,,,v-+, ~
 
                                      -.=          -.    .--    - -      --
APPENDIX D:                  REACTOR COltE LOADING AND UNLOADING General Loading and unloading of the reactor core shall be performed under the supervision of the Reactor Facility Director or the Reactor Operations Supervisor.
I Specific
: 1.        Setup
: a. Ensure that at least one nuclear instrumentation channel is operational,
: b. Ensure that the source is in the core.
: c. Ensure that an operator monitors the reactor console during all fuel movements.
: d. Check new FFCR's before insertion into the core; this includes cleaning, visual inspection, and length and bow measurements,
: e. Install all control rods.
: f. If irradiated fuel elements are to be removed unshielded from the pool, obtain a Special Work Permit Department SHD do not remove f(uel elements with aSWP)                  fromhistory the Safety and Hea (greater than (1 KW)in the previous 2 weeks from the reactorpower    pool.
: 2.        Core Loading
: a. After each step of fuel movement perform the following:
(1) Record detector readings.
(2) Withdraw control rods 50%; record readings.
(3) Withdraw control rods 100%; record readings.
(4) Calculate 1/M.
(5) Plot 1/M versus number of elements (and total mass of      U).
(6) Predict critical loading.
(7) Insert ALL rods; continue to next step.
: b. Load elements in the following order:
(1) Load the B ring and C-ring thermocouple elements.
15
 
(                                                                                                                                                                                              l 1
1 i
(2) Connect thermocouple outputs to reactor control console display.
]                                            (3) Install any other thermocouple elements.                                                                                                      1 an d      C ring elements (total of 18 standardelements  (4) Complete plus 3 loading FFCR's).of D                                                                                                            ,
!                                              (5) Load D ring (total of 33 standard elements plus 3 FFCR's).
)                                              (0) Load the following E ring elements in order:
10,17,18, 20, 6, 8, 9,10 (total of 41 standard elements plus 3
  ~
FFCR's).
)                                              (7) Complete the E-ring by loading the following elements in order:
;                                                    15, 21, 11, 5, 14, 22, 4, 12, 13, 1 (total of .51 standard
:                              elements plus 3 FFCR's).                                                                                                                                      1
.                                              (8) Load the following F-ring elements in two elements per step until criticality is achieved, using the following loading order:
!                                                  22, 23, 24, 21, 20, 25, 26,-27, 28, 29, 30, 1, 2, 3,_4, 5, 19, 18,                              ~
4                              17, 16, 15, 14, 13, 6, 12, 7, 11, 8, 10, 9.                                                                                                                  i I                              Once criticelity has been achieved, perform control rod worth measurements                                                                                    *
!                              at core position 500 by rod drop technique. Calculate shutdown margin
{                              (SDM):
l                                                    SDM = total control rod worth - K,y,,, - TRANS rod worth d
(9) Load core to $2.00 excess reactivity by loading two elements per step using the loading order in instruction 8.
1
;                                            (10) Verify control rod worth using rod drop techniques; calculate '
i                              SDM, i
(11) Load the core to achieve' a K                                                    that- will allow calibration of-f                              the TR.ANS rod based on the last available*Torth- curve of the- TRANS rod
;-                              (approximately $4.00). Calculate the reactivity value of each element as it '
is added.
(12) Calibrate all control rods.
;                                            (13) Calculate SDM.
(14). Estimate K,,,,,,- with a fully loade'd core (must not _ exceed
[
15 instructi(on)8, and recalibrate all control reds.                                                          Load    core to fully operational Calculate SDM.
16
_ _ _ _ _ . . _ , . _ -                      _ . , ,      - - _ . _ - . _ . _ . . _ . . ~ _ . ~ . _ . _ ,                                          , _ - , _ _ . _ - . _ . . _ . . .
 
  - - - -          , -                -    . - .        - - . - - . -              . - . ~      . .                    -  . -.      . . _    - . . . -  _ . -  -
                              !f necess(ary.16)          Recalibrate    Adjustallthe    corerods.
control    loadingCalculate          patternSDM.
to meet operations requirements
: 3.        Core Unloading
: a.      Unload the reactor core starting with the F. ring and ending with the D ring,
: b. Remove the fuel elen.;nts individually from the reactor core, identify them by serial nutaber, and place them in the fuel storage racks or a shipping cask.                                                                                                                          '
: c. If elements are to be loaded into a shipping cask, clean the cask completely, and check for rad'ological contatnination before placing the cask in or near the pool. Load cask in accordance with procedures specific to the cask,
: d. Once the cask is loaded, perform an air sample and survey; check temperature and pressure inside cask, if necessary.
: e.      If elementa are placed in temporary storage away from core monitoring, ensure that criticality monitoring in accordance with 10 CFR 70 is in place.
l 17
 
l l
l l
1 1
ATTACHMENT B                                                                                                    t 1
Current Reactor Administrative and Operational Procedures i
i 1
l
:                                                          . . e.. , . _ y . . _, . , . ,,_ ,    .
I i
i i
i PROCEDURES For The AFRRI Reactor Facility O                        C' 4< <              >
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TABLE of CONTENTS r
ADMINISTRATIVE PROCEDURE                                                                                Revised Date Procedure A1 Fitness For Duty ........ ............................. ............................. 15 May 91 Procedure A2 Personnel Passage nrough The Prep Arca.......................... 15 May 91 Procedure A3 Facility Modification ........................................ .................... 15 M ay 91 Procedure A4 Special Nuclear Material Accountability...............................                  15 May 91 OPERATIONAL PROCEDURE Proce dute 0 Procedute Changes .................................................................... 15 M ay 91 Procedure 1 - Conduct of Experiments ........................................................... 15 M ay 91 Procedure 1, TAB A Reactor Exposure Room Entry .................................... 15 May 91 Procedure 1, TAB B - Core Experiment Tube (CET) ..................................... 15 May 91 Procedute 1, TAB C - Extractor System........................................................... 15 M ay 91
,                      Procedure 1, TAB D Pneumatic Transfer System (PTS) ............................... 15 May 91 Procedure 1, TAB E - In. Pool /In Core Experiments ........................................ 15 May 91 Procedute 2 - Reactor Staff Training ............................................................... 15 M ay 91 Procedute 3 Maintenance Procedures ........................................................... 15 May 91 Procedure 4 Perscanel Radiation Protection ............... . ............................... 15 Nov 91 Procedure 5 Physical Security ....................................................................... 15 M ay 91 Procedure 6 - Eme rgency Procedure s ............................................................. 15 M ay 91 Procedute 7 Core Loading and Unloading .......... ........................................ 15 May 91
 
1 TA
      -.BLE      of CONTENTS                                                                                              Revised Date.            ;
1 Proce d u re 8 R eactor C pe r ations ......... ............. ............ ...... . ...... ................ 15 Nov 91 Procedure 8, TAB A - legbook Entry Checklist .... ........... . ..... . ..................... 22 Jul 91 Procedure 8, TAB B Daily Operational Startup Checklist ..... .... ................ . 15 hiay 91 i                Procedure 8, TAB B1 Daily Safety Checklist .............................. .. .............. 15 May 91 Procedure 8, TAB C Nuclear instrumentation Set Points ........ .....................15 hiay 91 Proce d u r e 8, TAB D K.Exce ss ..... .............. . .................... ....................... ... . . 15 M a y 91 Procedure 8, TAB E Steady State Operation ....................... .......................... 15 May 91 Procedure 8, TAB F1 Square Wave Operation (Subcritical) ......................... 15 May 91 Procedure 8, TAB F2 Square Wave Operation (Critical) ............................... 15 May 91                                    i Procedute 8, TAB G1 - Pulse Operation (Critica1) ........................................... 15 May 91 -
Procedure 8. TAB O2 Pulse Operation (Subcritical) ..................................... 15 May 91 Procedure 8, TAB H Weekly Operational Instrument Checklist ................... 15 May 91 Procedure 8, Tab I Daily Operational Shutdown Checklist ........................... 15 May 91 Procedure 9 Reactor Roorn S afe ty ................................................................. 15 M ay 91 Procedure 10 - Staek Gas Monitor Procedure ................................................. 15 May 91 Procedure 11 - Air Particulate Monitor Procedure .......................................... 15 May 91 f
I
 
                                                                                          ''(
Lhocedure A1 ADMINISTil
    , ..    . - T.iVE![i m _ __ PROC _SDUREE I                                                                                                            i
;                                                                                                            t This procedure has been approved by the Reactor Facility Director                                            '
                                              /            iDo    FVlfP f)u Q '
Reactor Facility D fctor FT          KaGG u
                                                                                            ' DatI Reviewed by the Reactor Staff fd/S h              2-9 New4/
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Spence                              Date
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FITNESS FOR DUTY                                              l OENERAL The AFRRI Reactor Facility is a drug free work place. The use of illicit drugs by any RSDR staff member is prohibited. Personnel using over the counter or prescription drugs which cause drowsiness or otherwise alter one's state of con-sciousness will not be permitted to operate the AFRRI TRIGA reactor. In addi-tion, reactor operators, operators in training, and management will be monitored for attitude and behavioral changes that may impact an individual's reliability.
SPECIFIC
: 1. RSDR staff members shall participate in drug free awareness programs spon-sored by AFRRI. Military and civilian staff members shall submit to drug screen-ing programs conducted by their respective services. If a staff member's drug screening test yields a positive result, that staff member shall not be permitted to operate the reactor pending verification of the test. The Reactor Facility Director (RFD) is required to ensure that the cutoff levels for alcohol or controlled sub-stances as established in 10 CFR 26 are not exceeded by NRC licensed personnel.
Any staff member determined to be a drug user will be terminated.
: 2. Personnel are instructed to inform their physician of their job description and requirements prior to being issued a prescript!on medication. They are instructed to inquire about any medication side effects expected and the physician's opinion regarding interference with safe job performance. This information shall be re-layed to RFD as soon as possible.
Personnel are encouraged to minimize their use of non prescription over-the-counter drugs for self medication purposes. Specifically, sedatives, cough and cold preparations, appetite suppressants, and pain relievers have central nervous system side effects. If these medications are used in any quantity, an operator must in.
form the RFD or ROS and~be relieved from operating on that day or until any side effects have resolved once the medication has been discontinued.
Personnel are instructed to read the information in the Physician's Desk Refer-ence (PDR) concerning medication they are taking. If the PDR indicates that the medication will adversely affect an operator's abil!ty to safely perform his/her du.
ties, he/she must inform the RFD or Reactor Operations Supervisor (ROS) that he/she must be relieved from operating on that day.
Revised: 15 May 91                                                                      Page 1
 
i 1
: 3. The RFD shall continuously monitor the reliability of individuals under his/her                                  i command by the following criteria-l
* Any court martialor civilconviction of a serious nature. Minor traffic                    l violations are not a consideration.
* Negligence or delinquency in duty performance.
* Significant mental or character traits, or aberrant behavior, sustained by medical authority, that might affect the reliable performance of duties.
* Behavior patterns that s'..ow or suggest a contemptuous attitude toward the law or regulations                                                                  _
                                                          +
Drug abuse or alcohol misuse.
* Poor attitude, lack of motivation toward assigned duties, or financial it.
responsibility.
The RFD will be observed by his superiors to ensure his/her adherence to reliability criteria. Individuals who exhibit any of the listed behavions or actions will be removed from licensed activities, t
Revised: 15 May 91.                                                                                        Page 2.
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r This procedure 'has been ap roved .y the Reactor Facility Director
                                              )41b/to L                                                        MJu#t/  '
Rea6tn't Facil'ity DVgftor                                                          Dale                            I h i A /l                                                            .2 7 t% v t/                                    l Chairnfan, Safety and Health Department                                            D(ite                            l Reviewed by the Reactor Staff                                                                                          i
                                                                    /djd/                t/zvAl Wrighf                            Date 4L/.Ad... v                    erh1/1/
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                    > y.a ,ggy                                                  y;ppy i      PERSONNEL PASSAGE THROUGH THE PREP AREA GENERAL Access to the Reactor Prep Area is lirnited to personnel who are granted access in accordance with the reactor physic 6 security plan and Operational Procedure 1.
The Reactor Facility Director is responsible for mairaining an unescorted access roster for the Reactor Prep Area and for providing a Prep Area briefing to all persons listed on that roster. This administrative procedure does not recapitulate the operational procedure. Rather,it presents specific guidelines for Reactor Prep Area pascage for individuals who are authorized access.
SPECIFIC There are three specific conditions when Reactor Prep Area passage is considered appropriate.
: 1. ROUTINE PASSAOE:(EXPOSURE ROOM DOORS CLOSED)
: a. Perronnel who are authorized unescorted access to the Reactor Prep Area may pass through the Prep Area but are required to radiologically frisk themselves when exiting the Prep Area,
: b. Personnel who are being escorted through the Reactor Prep Area may pass through the Prep Area with their escort. Both individuals must radiologically frisk themselves when exiting the Prep Area.
: c. Only appropriate personal dosimetry that has been issued by SHD or AFRRI Security is requi;ed for routine passage through the Reactor Prep Area. There is no requirement to wear a pocket chamber in addition to the AFRRI issued TLD for routine passage.
: 2. CONTROLLED PASSAGE:(EXPOSURE ROOM DOOR OPEN)
: a. Only personnel associated with the experiment / operation being performed are normally authorized access to the Reactor Prep Area during an exposure room operdng. These personnel will be required to wear the AFRRI TLD dosimeter (issued with their AFRRI badge), pocket chamber (if dose rate at face of door is 2 mr/hr or more), and the following special dosimeters if they enter the exposure room:
AFRR1 wrist dosimeter or finger ring dosimeter.
Revised: 15 May 91 Page i
 
4 All personnel who enter an exposure room will log their pocket chamber reading ia the pocket chamber log prior to :ntering the room for the first time that day and will enter the final pocke chamber reading following their exit from the exposure room at the end of the day. Each individual who enters an exposure room is responsible for monitoring his accumulated dose throughout the day to ensure the he/she doc not exceed the AFRRI daily exposure limits of 50 mR/ day or 100 mR/ week. Extremity dosimetry is required only if work is to be performed on an experimental array or within 1 meter of the core projection.
: b. Personnel authorized unescorted access to the Reactor Prep Area or personnel being escorted through the Reactor Prep Area may pass through the Prep Area when an exposure room is open with permission from the Reactor Staff person in charge during the opening if the following guidelines are met:
(1) The person desiring passage must stop just inside the Prep Area door upon seeing that an exposure room door is open and request permission from the Reactor Staff member in charge before proceeding. At that time, the Reactor Staff and Safety Sta" members monitoring the Exposure Room opening will determine if the radiation level at the uutside entrance to the exposure room in direct line of sight with the core projection is less than or equal to 2 mR/hr. If this reading is less than or equal to 2 mR/hr, the Reactor Staff Member may grant passage permission. If the reading is greater than 2 mR/hr, passage will be denied.
(M3 .rsonnel who pass through the Reactor Prep Area must tadiobgically frisk. themselves before exiting the Prep Area. There is no requhement for these individuals to wear a pocket chamber just to pass through the Prep Area.
: 3. OPEN PASSAGE:(NON-ROUTINE)
The prep area may be opened for passage by personnel traveling between buildings at AFRRI when maintenance is being performed on the normal connecting                        a hallw;y. This is not a routine occurrence and warrants written approval from the Reactor Facility Director with concurrence from the Chairman, Safety and Health Department. In addition, the Prep Area must be monitored at all times by appropriately train:d personnel. Prior to the opening of the Reactor Prep Area for open passage, the Safety and Health Department shall conduct a radiological survey of the area and cenify that no radiation areas exist within the Prep Area and that the non painted areas of the Prep Area floor are free of contamination.
There is no requirement for personnel who pass through the Prep Area to wear pocket chambers or frisk themselves during periods of open passage. Finally, open passage will be suspended during exposure room openings.
    ~
Revised: 15 hfay 91                                                                Page 2    )
 
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                                                                                          ,I    _
                                                                  /  -
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[.                                                                  Reviewed by the Reactor Staff
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FACILITY          MODIFICATION GENERAL                                                                                              i Changes to the Reactor Facility and operational procedures must comply with requirements specified in the Reactor License, and 10 CFR 50.59. It is required '
that modifications to the facility or procedures as described in the Safety Analysis Report (SAR) be documented with a written safety analysis. Under 10 CFR 50.59, a licensee may make changes to the facility provided there are no changes made to the Technical Specifications, there are no unreviewed safety questions, and that a proper safety analysis is carried out, documented, and reviewed.
Applicability:
The Facility Modification Procedure applies to proposed facility changes or changes in the operating procedures.
The referenced procedure will not cover routine replacement of parts or components with equivalent parts or components.
DESCRIPTION This administrative procedure consists of these instructions, the Facility Modifi-cation Worksheet Guide, and two worksheets to facilitate.a 10 CFR 50.59 review of modifications and to determine if a detailed safety analysis is necessary. The in-structions in the Facility Modification Worksheet Guide are used determine which worksheet must be completed for the modification. One of three conclusions re-garding the proposed facilMy modification will be reached:
: 1. The modification requires prior approval or a license amendment from the USNRC,
: 2. The modification may be made according to the provisions of 10 CFR 50.59(a)(1) (Facility Modification Worksheet # 1), or
: 3. The modification does not require a 10 CFR 50.59 safety analysis (Facility Modification _ Worksheet # 2).
l Revised: 15 May 91 -                                                                  Page1 E          _ _ .                  .        _.          ._        ~      ._        _    -.- _ _ .-        .  - .
 
Facility Modification -Worksheet Guide.
: 1. Technical Specification Change: If the proposed modification requires a changes in the Technical Specifications, a license amendment is required prior to-making the change. NRC approvalis required; do not implement the change with-out this approval.
: 2. Unreviewed Safety Question: If an unreviewed safety question is created by the proposed change as defined in 10 CFR 50.59(a)(2) such that the change increases the probability of occurrence or severity of an accident described in the SAR, can malfunction in a manner that can cause an accident of a different type than de-scribed in the SAR or can decrease safety margins as d: fined in Technical Specifica-tions, then NRC approvalis required. Do not implement the change without this approval.
: 2. If the proposed modification makes a change in the facility ss described in the SAR or changes a' procedure as described in the SAR, the change can be per-formed under a 10 CFR-50.59 analysis with a safety review, if there are no un-reviewed safety issues (10 CFR 50.59(a)(2)). The change may be made following a review by the RRFSC. _Go to Facility Modification Worksheet # 1.
                                                                                                            ~
: 3. If the proposed modification does not make a change to the facility 'as described in the SAR or to a procedure as described in the SAR and does not pose an un-reviewed safety issue, a 10 CFR 50.59 analysis is not required. Go to Facility Mod-ification Worksheet # 2.
i Revised: 15 May 91                                                              Page 2 l
 
  . - . .  ~ ~ _ . . _        .  -_ _ . _ .        -  _    _  ...        . _        _ . . _
d Facility Modification Worksheet 1 10 CFR 50.59 Analysis Proposed Change                                            ,
Subrnitted by:                                                        Date
: 1. Description of change:
: 2. Reason for change:
: 3. Verify that the proposed change does not involve a change to the Technical Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.
Analysis attached? Yes            _
: 4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).
Procedure                Facility        Experiment Revised: 15 May 91                                                                              Page 3 -
l
 
Facility Modification Worksheet 1
: 5. Specify what sections of the SAR are applicable, In general terms describe the.
necessary updates to the SAR. Note that this description need not contain the final SAR wording.
i
                                                                                          -l
: 6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.
1 i
Revised: 15 May 91                                                                Page 4
 
Facility Modification Worksheet 1
: 7. Specify associated information.
New drawings are:                    Attached Not required Does a drawing need to be sent to Logistics?                                  Yes            No Are training materials effected?                                              Yes          No Will any_ Logs have to be changed?                                            Yes            No Are othe; procedures effected?                                                Yes            No List of items affected:
: 8.        Create an Action Sheet containing a list of associated work specified in item #-
7, attach a copy, and submit another to the RFD.
Action Sheet:                      Submitted                                            Not Required-1 Reviewed and approved by RFD.                                                                                    D ate -
RRFSC Concurrence                                                                                            - Date Revised: 15 May 91                                                                                                                    Page 5-
          . . -                    .      -      . _ . . _ . _ . . - - _ . c.            . - . _ . . . - . . _ . _ . _ . -    1._ . . . _ . .    . . . _ _ . _ _ . .      _ u .. ..
 
1 1
Facility Modification Worksheet 2 No 10 CFR 50.59 Analysis Required
:      Proposed Change Modification to:        Procedure            Facility            Experiment Submitted by:                                                      Date
: 1. Description of change:
f i
: 2. Verify that tre proposed change does not involve a change to the Technical Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved safety question as defined in 10
:      CFR 50.59(a)(2).
: 3. If change involves a facility modification, attach a drawing if appropriate. If-structural facility drawings need updating, forward a copy of changes necessary to
;      Logistics.
: 4. Determine what other procedures, logs, or training material may be affected 4
and record below.
;      5. List of associated drawings, procedures, logs, or other materials to be changed:
!      6. Create an Action Sheet containing the list of associated work specified above, attach a copy, and submit it to the RFD.-
Action Sheet:            Submitted                Not Required Reviewed and approved by RFD                                            Date RRFSC Notified                                                        Date i
Revised: 15 May 91                                                              Page 6
 
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wa                cws stig g }} g g g ] g'g4g SPECIAL      NUCLEAR          MATERIAL ACCOUNTABILITY
: 1. PURPOSE This procedure prescribes the responsibility for all work involving the receipt, use, and disposal of Special Nuclear Material (SNM) at AFRRI.
: 2. REFERENCES a.10 CFR 70 Domestic Licensing of Special Nuclear Material b.10 CFR 73 Physical Protection of Plants and Materials c.10 CFR 74 Material Control and Accounting of Special Nuclear Material
: d. NUREG/BR-0006/7 Instructions for Material Balance Reports and Transac-tion Reports
: e. AFRRI Health Physics Procedures
: 3. RESPONSIBILITY
: a. Overall responsibility for all SNM at AFRRI rests with the Reactor Facility Director (RFD). The RFD designates individuals to be in charge of supervising the use of and accountability for specific types of SNM.
: b. Members of the reactor staff will use SNM procured under license R-84 in ac-cordance with the reactor Technical Specifications and Rerctor Operating Proce-dures. Users of SNM listed under the byproduct material license will be
!    governed by Health Physics Procedures (HPP 5-X series). The RFD will desig-l    nate a member of the reactor staff (Inventory Officer) to conduct inventories and maintain accountability records for all SNM as required by Reference d,-
i    this procedure, and HPPs.
: c. The inventory officer will receive training in fuel records,-accountability pro-l    cedures, and inventory procedures prior to assignment as inventory officer. This will be documented on the appointment memo. (Appendix A)
: 4. RECEIPT AND TRANSFER OF SNM Receipt of SNM at AFRRI and transfer af SNM from AFRRI are regulated by 10 CFR and 49 CFR and performed in accordance with HPPs 0-3 and 0-5. When any SNM is received or transferred, the responsible inventory officer will ensure that DOE /NRC Form 741 is submitted as required by Reference d and 10 CFR 74.15.
Revised: 15 May 91                                                                Page1
 
The RFD must authorize any receipt or transfer of SNM Approval is also re-quired from the RXSC for any SNM on the byproduct material license.
: 5. ACCOUNTABILITY AND INVENTORY PROCEDURES
: a. Safety and Health Department (SHD) personnel will conduct leak tests as re-quired by HPP 5-2. A physical inventory of all sealed SNM sources on the by-product license will be conducted as of March 31 and September 30 of each year. This semiannual inventory will be conducted to ensure annual compliance with 10 CFR 70. The specific inventory method will follow ALARA considera-tions. Reports required by References c and d (DOE /NRC Form 742/742C) will be prepared and submitted within 30 days (10 CFR 74.13).
: b. A current reactor fuelinventory sheet (Appendix B) will be maintained. It will give the core loading (core position) for all in-core fuel elements and FFCRs. During each inventory , the inventory officer will sign the inventory sheet verifying loading. This document will be updated if changes occur and at least 10% of the in core elements will be verified by serial number as to location annually.
;    c. Inventory procedures for reactor fuel elements and fuel-follower control rods I
will be as follows (conducted as of March 31 and September 30 each year by the l    inventory officer):
(1) Perform K-excess to verify the core is loaded as per the fuel invea-tory sheet.
(2) Sign the fuel inventory sheet to certify correct loading and attach to summary accountability sheet. (Appendix C)
(3) Inventory by serial numbers fuel elements in stcrage except for en-tombed damaged elements (previously recorded)
(4) Account for any fuel elements or FFCRs received or shipped out during the reporting period (5) Ensure number of elements and FFCRs present matches the num-ber of fuel element record sheets and that Fuel Element Record (AFRRI Form 255) sheets are current (6) Transfer data from fuel inventory sheet to summary accountability shect and verify correct total number of fuel elements and FFCRs (7) Calculate fuel burnup in grams based on kilowatt-hours generated during the reporting period and subtract from the previous material balance Revised: 15 May 91                                                                  Page 2
 
(8) Complete and submit DOE /NRC Forms 741, 742, and 742C using procedures in Reference d.
(9) File the fuelinventory sh:et, the summary accountability sheet, and copies of all DOE /NRC forms in reactor files,
: d. Fission chambers and other SNM on the R-84 license will be inventoried by piece count in conjunction with the fuel inventory, recorded on the summary ac-countability sheet, and reports submitted as in (8) above. All smear testing /han-dling will be in accordance with Reference e.
: 6. RECORDS All records and reports pertaining to any SNM held at AFRRI will be maintained in reactor facility files by the inventory officer. Duplicate receipt and transfer files, radiological survey documents and leak test results will be maintained in SHD files.
: 7. REPORTS OF SNM INCIDENTS Reports of loss, theft, attempted theft, or diversion of SNM will be submitted as re-quired by 10 CFR 74.11. Other incidents involving SNM may be reportable under reactor Tec'mical Specifications or 10 CFR 20.402-20.405
: 8. SIGNATURE AUTHORITY All reports and other outgoing correspondence dealing with SNM will be signed only by the Reactor Facility Director or his designee (see paragraph 3a).
f Revised: 15 May 91                                                                  Page 3
 
Appendix A INVENTORY                    OFFICER is appointed Inventory Officer to perform duties as per Ad-ministrative Procedure A4. He has received proper training in accountability proce-dures, inventory procedures, and fuel records.
Reactor Facility Director                  D ate Revised: 15 May 91                                                                                    Page 4
 
APPENDIX B - FUEL INVENTORY WITHDRAWN FROM PUBLIC DISCLOSURE I
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  . .. .                        ~ . .. .. ... .-            . - ..                      --
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Appendix C                                                                                                                                  \
l Summary Accountability Sheet Number                                        Weight Fuel Elements In Core Fuel Elements In Tank Storage Total Fuel Elements Fuel Follower Control Rod (Location)
Chambers Foils Sources l
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i Revised: 15 May 91                                                                                                          Page 11'
 
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PROCEDURE CHANGES GENERAL This establishes procedures for permanently or temporarily changing reactor oper-ating procedures.
SPECIFIC
: 1. Permanent changes are made by revising the entire procedure. The revised pro-cedures will be approved by the Reactor Facility Director (RFD) and reviewed by-the Reactor and Radiation Facility Safe _ty Committee (RRFSC).
: 2. Temporary changes may be made in pen and ink on the current procedure when initialed by the RFD or Reactor Operations Supervisor (ROS). These changes must be documented, approved by the RFD, and reviewed by the RRFSC at the next scheduled meeting.
: 3. Temporary procedures may be established by the RFD for a specific situation.
: 4. All procedures (temporary or permanent) will have an signature block for all op-erators and reactor staff members. Operators will review new or modified proce-dures and sign the signature block prior to operating the reactor console When the block is completed,' the procedure will be placed in the Reactor Procedures Binder and kept available for operator review.
: 5. All changes will be accomplished under the following guidelines:
: a. The change will result in no decrease in the safety of the actions being addressed.
: b. The change will result in no decrease in the efficiency of procedure performance,
: c. The change will not affect the ability of the procedure to perform its intended function.
l l
Revised: 15 May 91                                                                                    Page 1
 
    .- .    . -      - = - -              .
                                              ---        .=    ..  . . _  . .      . . - . ._          _ - .
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: 7. All changes will be staffed to the following:
: a. Chairman, Safety and Health Department (SHD)                                                ,
: b. Reactor and Radiation Facility Safety Committee (RRFSC)
: c. AFRRI TRIGA Reactor Facility staff Procedures that may affect other areas such as building changes, security, etc.. will  .
be staffed to the appropriate office (s) prior to routing to Chairman, SHD
* NOTE: Procedural changes that _do not deal specifically with health physics pro-
',        cedures or radiation safety issues need not be staffed through Chairman, SHD.
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CONDUCT OF              EXPERIMENTS l
GENERAL
: 1. All experiments will be observed during irradiation with the exception of CET experiments or those in which no movement is possible. The closed-circuit televi-sions (CCTVs) in the exposure rooms and over the r actor pool can be used to meet this requirement.
: 2. All experiments will be set up so as to preclude movement unless the experiment apparatus is designed for movement (such as rotators, etc.).
: 3. All animal experimental arrays (shielding)in the exposure rooms that are set up on wooden tables or on styrofoam will have an absorbent pad placed over the wood or styrofoam surface to prevent sanitation problems from the animal waste.
; 4. The Reactor Staff will conduct a thorough inspection of all experiments to de-termine that no unauthorized materials are irradiated.
: 5. ALAR A will be practiced during all experiments.
SPECIFIC
: 1. A Reactor Use Request (RUR)is required for any experiments included under authorizations outlined in the Technical Specifications, section 6.4.2.a. and section 6.4.2.b.. RURs are not required for reactor parameter measurements as outlined in the Technical Specifications, section 6.4.2.c. Any experiment performed by the reac-tor staff (except T.S. 6.4.2.a) for the purpose of determining information to be used to enhance, define, ascertain, or develop methods to expand the performance of the reactor will not require an RUR. Facility tours will not require an RUR but will require verbal approval of either the Reactor Facility Director (RFD) or the Reactor Operations Supervisor (ROS).
: 2. Expr.timent Review (Processing of RURs):
: a. Check the RUR for completeness (Section I should be filled out).
: b. Forward the RUR to the Radiation Biophysics Department, Opera-tional Dosimetry Division (BRPD) if dosimetry support is required.
: c. Forward it to the Safety & Health Department (SHD) for radiologi-cal safety coordination.
Revised: 15 May 91                                                              Page 1
 
l
: d. Check experiment protocol against reactor cuthorizations.
: e. Fill in Section II of RUR with special instructions, as appropriate, Assign an RUR sequence number. Write in estimated or measured ex.
periment worth and the core position of the experiment facility to be                                                                  t utilized in the appropriate block (lower left hand corner of form),
: f. Have the RFD, acting RFD, ROS or acting ROS review and sign the form.
: g. Ensure the RUR form is placed in the reactor control room prior to the irradiation.
: 3. Conduct of Experiments. Perform setup and irradiation of experiments in ac-cordance with the following procedures:
: a. Exposure Room Entry - TAB A.
: b. Core Experiment Tube (CET) - TAB B.
: c. Extractor System - TAB C.
: d. Pneumatic Transfer System (PTS) - TAB D.
: e. In pool /In core Experiments - TAB E.
: 4. Complete the RUR by filling out Section IV with the appropriate information.
: 5. Attach form to clipboard in the control room.
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0(W M Q f & g &.nyewyg:4QQMQantTfDyg*3 REACTOR EXPOSURE ROOM ENTRY
: 1. REFERENCES a.10 CFR 20, " Standards for Protection Against Radiation"
: b. USNRC licenses: R-84,19-08330 02
: c. AFRRI Instruction 6055.8B
: 2. GENERAL
: a. PURPOSE: This procedure specifies all safety and security procedures for ac-tivities involving entry into the AFRRI TRIGA Reactor exposure rooms, currently designated exposure rooms 1 and 2 (rooms 1123 and 1122).
: b. AUTHORIZED ENTRY: Both AFRRI picture badge and U-badge person-nel, may enter a reactor exposure room under the supervisica of the Reactor Facil-ity Director (RFD) or his representative. Visiting personnel (V badge) require special authorization by both the Chairman, Safety and Health Department (SHD) and RFD to enter either exposure room. In general, permission to enter the expo-sure rooms will be granted personnel whose duties require such entry; however, per-mission may be denied to personnel for serious or repeated safety or security violations, or for safety reasons emanating from conditions in the exposure rooms.
All personnel who are granted unescorted access to the prep area or warm storage will receive a special prep area safety briefing prior to being granted access. Only personnel who have been granted unescorted access will be given the combination to the prep area . The RFD is responsible for maintaining a roster in the prep area tw personnel who have been granted unescorted access. Other personnel re-quiring unescorted access to the prep area or warm storage for a specific purpose or time period may be granted special access in writing by the RFD with concur-rence of SHD. However, these personnel who are granted special access from the RFD will not be given the combination to the prep area,
: c. ER ENTRY INSTRUCTIONS - All personnel will:
(1) Know the Reactor staff representative is in charge of all operations in the prep area. Obtain permission to enter either exposure room from the Reactor staff representative.
(2) Wear AFRRI TLD whole body badge and pocket dosimeter.
Revised: 15 May 91                                                              Page 1
 
l l
l (3) Wear wrist or finger dosimeter if work is to be performed on an experi-mental array or within one meter of the core projection.
(4) Wear booties, eye protection, gloves and coat.
(5) Check and log pocket dosimeter reading on log in prep area prior to entry.
(6) Familiarize themselves with approximate radiation levels in the room, based on radiological surveys performed and data obtained by SHD.
(7) Ensure that all materials removed from the exposure room are properly la.
beled and entered on the exposure room entry log (AFRRI FORM 130) and the ac-tivated materials control log.
(8) Glove and coat requirements may be waived by the Reactor Representa-tive on an individual basis for personnel who will not be touching anything in the exposure room. There must be a specific reason for waiving;such requirements.
: d. DEPARTURE FROM REACTOR EXPOSURE ROOM ENTRY PROCE-DURES: Any departure from the following procedures will require a special work permit (SWP). Exceeding any radiation dose limits will require a written justi-fication from the supervisor of the research project which must be approved by the Head, SHD.
: 3. SHD EXPOSURE ROOM SURVEY
: a. EXPOSURE ROOM CAM: Prior to opening either exposure room, the re-spective CAM must read 2000 cpm or less, above background. If the CAM reads 2000 cpm or greater above background, change the filter of the C. M. If 10 min-utes or more have elapsed since the end of the reactor run, the door may be opened to the first step to facilitate radioeffluent clearance in the room._ Then check the CAM after 1 minute and if the reading is below 2000 cpm above back-ground, proceed with the exposure room opening. Ifit is above, change the filter and wait another minute. If the CAM alarms during;or immediately after a run,--
change- the filter and reset the CAM.
: b. DOSE RATE AT FACE OF DOOR: If the dose rate at the face of the plug-door in the direct line of sight of the reactor tank bulge reads greater than 100          ,
mr/hr, the door will be closed sufficiently to preclude access. The plug door will be reopened upon agreement of the- SHD and RFD representatives for reevaluation of radiation levels.
: c. DOSE LEVELS IN ROOM: Exposure rates will be measured at specific sites-in the rooms. These measurements will be given to both the reactor representative-and the personnel entering the room. Additionally the readings will be entered in -
Revised: 15 May 91                                                                Page 2:
 
I the room entrance log (AFRRI FORM 130) and kept in the prep area. The levels will be measured at:
(1) The reactor door face in the direct line of sight of the reactor tank bulge (2) At the contamination line in the entrance of the room (3) The middle of the room (4) One meter from the tank wall or shield (5) Contact with the tank wall or shield (6) The area (s) where individual (s) will be working for an extended pe-riod of time and any other place deemed necessary by the SHD or reac-tor representatives,
: d. ROUTINE ENTRY: Entry is routinely permitted only when the maximum reading in any occupiable area is 1 R/h or less. Entry may be permitted iflevels are 1-5 R/h, but no work will be permitted in fields over 1 R/h. If personel are working in a specific area for an extended period of time , the dose rate in that area will be measured.
(1) Readings over 100 mR/hr (closed window) will be reported to the Reactor representative by the SHD monitor. These areas of the exposure room will be iden-tified to the Reactor representative and entry personnel. When appropriate, after consultation with the SHD and Reactor representatives, stay times will be assigned for entry personnel. All personnel entering will be assigned a stay time if they will be working in the high radiation area. AFRRI limits of 100 mR/ week and 50 mR/ day are to be used as the basis of stay time determinations.
(2) All exposure room entries will be checked by the SHD monitor for compli-ance with radiation safety aspects of applicable Reactor Use Requests (RURs). If not, non compliance will b reported to the RFD and to SHD.
: e. FILLING OUT THE SURVEY OF EXPOSURE ROOM OPENING LOG:
The exposure room opening log sheet must be filled out completely for each open-ing of an exposure room. - Care must be taken to fill out each blank on the entry log sheet. If a section is not applicable to the particular opening, N/A should be filled in the blank.
: 4. NON MONITORED OPENING:
: a. Personel may enter the exposure rooms without a SHD monitor present if ALL the following conditions hold:
i (1) The reactor has not been to power in that ER since the last survey.
Revised: 15 May 91                                                                    Page 3
 
I (2) Survey meter readings at the door indicate safe entry conditions (should be less than 1 mR/hr).
(3) The ER CAM should be observed, and its reading (net) should be                                          -
less than 200 cpm above background.
: b. An entry will be made in the exposure room log by a reactor staff member, with a note that the survey has been waived.
: c. SHD must be notified if any radioactive materials or equipment are to be re-moved from the prep area.
: 5. PERSONNEL PROTECTION PROCEDURES
: a. Dosimetry and protective clothing requirements are given in paragraph 2.c, ER Entry instructions.
: b. Entry is permitted only after the SHD monitor has completed the survey and reported results to those about to enter (excluding non monitored openings - Refer-ence Paragraph 4, above).
: c. All personnel shall record initial dosimeter reading in the prep. area dosimeter log prior to entering the exposure room for the first time each day.' Personnel shall read dosimeters when leaving the exposure room and record a final dosimeter read-ing in the prep area log at completion of daily operations. Net doses over 10 mrem -
must be reported to the SHD Monitor,
: d. Protective clothing will be removed in such a way as not to contaminate
            " clean" areas by items from " dirty" areas,
: e. All personnel will " frisk" themselves before leaving the prep area.
: 6. SPECIFIC ACTIONS TO OPEN EXPOSURE ROOM DOORS-
: a. Turn up exposure room lights (this can be waived for experiment needs).
: b. - Check plug door tracks for obstructions; ensure all obstacles are clear of the door (including ropes).
: c. Ensure that only authorized personnel (see 2.b.) are present in the reactor prep area during: exposure room openings.
: d. When facility safety interlocks and opening procedures have been satisfied, in-sert key into exposure room door key panel and open door. DO NOT LEAVE KEY IN LOCK UNATTENDED.
t Revised: 15 May 91                                                                                          - Page 4 '.
: e. Open door in accordance with entry procedures. Ensure all required data is logged in entry log.
: f. Ensure that individuals who will be moving lead, bismuth, or other heavy ma-terials are wearing steel toed shoes,
: g. Limit exposure times of all personnel entering the exposure rooms based on the results of the radiation survey.
: 7. ACTIVATED MATERIALS
: a. PLACING MATERIAL IN EXPOSURE ROOM: Before placing any equip-ment or material in an exposure room for irradiation the following will be ob-served:
(1) Equipment tagged as AFRRI property: a DF must be sent to both the RFD and the AFRRI property officer. The DF must state that the equipment is knowingly being irradiated and therefore request that it be removed from the prop-erty books. It must also state that should the material remain byproduct material after a reasonable amount of time it will be disposed of as radioactive waste. The-DF must contain all nomenclature as well as an . adequate description of the equip-ment in order for it to be identified.on the property book.
(2) Non-tagged AFRRI equipment or material (to be returned): a DF. or state-ment on the reactor RUR must be sent to the RFD giving the~ kinds and. amounts of byproduct material expected to be produced (that is the material that the experi-menter wishes to be returned) and a copy or number of their radionuclide authori-zation number. The DF or RUR statement must be specific and contain an -
accurate description of the material being exposed (converted to. byproduct).
Other information will be required from personnel before any material is allowed                          ,
;                  to be removed from the prep or warm storage areas-(see next section of this proce-dure 7.b. and .7.c.)
(3) Non-tagged equipment or material (not to be returneil): A DF or state i                  ment on the-RUR that the experimenter understands that byproduct material pro-                            .
duced as a result of their irradiations will be disposed of as radioactive waste, and additionally any' material not specifically requested to be held, will be disposed of-as radioactive waste in.the next shipment, i                              (4) Non-AFRRI owned equipment / material: A signed memorandum from the -
responsible property. owner that they understand- that byproduct materials gener-l                  ated in excess of their license will be disposed.of as radioactive waste unless prior-arrangements have been made with the reactor /SHD staffs for storage. Any mate-Revised: 15 May 91                                                                    Page5
                ._            _.        _ _ . _ _ _ _ _ _ . . . _    _.--. _ . _          _  m. __ .        _ _ . . . _ _
 
                                                                                  +
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rial not removed within a reasonable amount of time will automatically be dis-posed of as radioactive waste.
: b. SURVEY OF MATERIALS COMING OUT OF EXPOSURE ROOM (1) All materialleaving the exposure rooms :..ust be surveyed for activation or .
contaminaticn Survey meter readings will be used to determine dose levels. Smear surveys may be used, if the SHD representative deems them necessary. All materi-            ,i
,        als will be labeled appropriately in accordance with HPP 0-2.                                I (2) All special equipment that has been activated such as chambers,;otators, motors, meters, etc., will be stored under the control of the reactor license or the AFRRI byproduct license in warm storage-or the prep area. Removal'of items                  '
from the prep area will only be allowed in accordance with HPP 31                          ;
: c. DISPOSITION OF ACTIVATED MATERIALS All activated or contaminated materials will be under the control of the reac -
tor license while such materials remain in the reactor controlled area. Removal of          j any radioactive materials from the reactor controlled area will be done in accor-            i dance with HPP 31.
: 8. COMPLETION OF ENTRY                                                                        '
i
: a. The Reactor Staff Representative will check to see that all personnel have left    -l the exposure room before the plug door is closed. In the event that the warning horn in either exposure room is disconnected, for testing-or experiment require-ments, the exposure room plug door shall not be closed wil at least two (2) li-              l censed reactor operators visually inspect the room to ensure .that no personnel l
remain in the room. To ensure compliance with the reactor Technical Specifica-tions, the names of these licensed operators present at the exposure room closing shall be entered into the reactor operations logbook and on AFRRI FORM 130.-
At the completion of the test or experiment, the warning horn shall be reconnected and tested. All actions regarding the warning horn shallin entered in GREEN ink-in the reactor: operations- logbook.
: b. The SHD monitor will not leave the area while the plug door is open without            !
notifying the Reactor -Staff Representative.
: c. Lock the exposure room door control panel; reset lights, if appropriate.
: d. Resecure the prep area on departure.-
Revised: 15 May 91 -                                                              ' Page 6
 
i OPF.RATjoNAL PROCEDURil                                                                '
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OPERAT10NAL      PROCEDURL                                                                Procedure 1, TAB D CORE EXPERIMENT TUBE (CET)                                                            }
GENERAL ALARA principles will be practiced during CET operations.
SPECIFIC
: 1. CET Insertion into the core:
: a. Ensure a reactor operator is monitoring the reactor console,
: b. Ensure a reactor staff member is present in the reactor room,
: c. Establish comrnunications between the reactor room and the control room.
: d. Test fuel handling tool for operability,
: c. Lower the fuel-handling tool into the core and attach to the desired element. Notify operator on th; console that you are prepared to lift fuel element. When acknowledged, lift fuel element from the core.
: f. Transfer ehment to a storage rack location and secure fuel handling tool cable,
: g. Loosen CET bracket bolts and remove CET bracket.
h While the CET is held down, cut cable tie from around the CET.
: i. Lift CET from the storage rack location and transfer to the reactor carriap.
: j. Notify the ;onsole operator that you are prepared to lower the CET into the core; when acknowledged, lowe: 'he CET into the core ensur-ing that it is properly seated in the lower grid plate,
: k. With a downward pressure on the CET to keep it seated, secure the CET bracket with the two bolts.
: 1. Ensure appropriate entries are made in the operations logbook and the fuel book, and that the reactor core pegboard is updated.
: 2. Irradiation:
: a. Clean the rabbit (s) using alcohol and water.
Revised: 15 May 91                                                                                        Page1
: b. Once clean, do NOT handle the rabbit except with gloves, Kimwipes, or handling tools.
: c. Ensure that the rabbit cap is secured tightly,
: d. The rabbit may,be inserted into the CET either before the run begins or while the reactor is at power .
: c. If the rabbit is to be inserted while the reactor is at power, then after notifying the reactor operator on console, drop or lower the rabbit into the core WITH THE CAP UP. Ensure that you spend a minimum amount of time in the vicinity of the carriage. Do NOT lower the rab.
bit with the extractor tool while at power.
: f. Complete irradiation and shut down reactor.
: g. Ensure appropriate entries are made in the operations logbook r.nd the CET logbook.
: 3. Rabbit Retrievals:
: a. Ensure that a reactor staff member and a Safety & Health Depart-ment (SHD) monitor are present in the reactor room. Any staff mem-ber who will be handling the sample following the irradiation may be required to wear a pocket chamber and appropriate extremity dosime-try depending on the radiation levels of the irradiated sample. If the CET is in the core, a reactor operator must monitor the console during the retrieval,
: b. Test the rabbit extractor (" fishing pole") for operability,
: c. Insert the extractor head mechanism into the CET and reel out cable until you reach the low end indicator painted on the cable.-
: d. Drop the extractor head firmly on the rabbit,
: c. Ensure the SHD monitor has a teletector positioned near the CET top to monitor the rabbit.
: f. If the CET is in the core, notify the reactor operator that the rabbit is being pulled and continue when acknowledged.
: g. Ree) in the cable at a rate commensurate with radiation levels; lower the rabbit back into the CET if the rabbit is excessively hot.
: h. Stop when the rabbit is visible at top of CET; have SHD take an ac-curate radiction reading,
: i. If radiation ievels are acceptable, swing rabbit away from carriage i
and have another individual grab it with a handling tool 'If the radia-tion levels are not acceptable, lower the rabbit back into the CET. The rabbit will again be withdrawn for teevaluation of radiation levels Revised: 15 May 91                                                                                  Page 2
 
I when the SHD and RFD representatives concur on w aceptable radi.                                                              i
)                                                    ation level in accordance with ALARA and mission nquirements.
                                                                                                                                                                                    >~
j, Release extractor head and detach rabbit from head.
: k. Unless working with the rabbit. er radiation levels are low, store rab.
[
1                                                    bit or irradiated material in a lead pig or storage cask, i                                                    1. .%Iake appropriate entries in the operations and CET logbooks.                                                            l
: 4. CET Removal from Core:
.1
: a. Complete steps la c above.
: b. Loosen the CET bracket bolts while holding the CET downt remove the CET bracket.                                                                                                              i
: c. Notify the console operator that you are prepared to remove the CET from the reactor core.
: o. When acknowledged, transfer the CET to the storage rack, ensuring                                                          1 that it is kept as low in the water as possible.                                                                              ,
: e. Secure the CET with cable ties.
4
: f. Secure the CET bracket with the two bolts.
: g. Remove the fuel element from the stork;. rack and transfer to core.                                                          '
Notify the console operator and receive acknowledgment prior to inser-tion of the element .
: h. Ensure the element is properly seated in the lower grid plate by lis-tening for the " double clicks".
: i. Make appropriate entries in the operations and fuel logbooks and up.                                                        ;
date the reactor core pegboard.                                                                                                '
t
(
Revised: 15 May 91                                                                                                                      PageS
  - _ , , . . , . . , . . . _ . , - . . . _ , . - - , . - -.,_..,,,,_-,,,,..__...--,._..,,.,m..-_
_ r.,.. , _ , _ . _ . , _ . , . . _ _ . . < - . = . . . . . . - .      .m.
 
                                    '                            ~
OPERATIONA1. PROCEDtJRE':
s                          Procedure 1C r
This procedure has been approved by the Reactor Facility Director AA b hdhmIY-              . _ _ ? hdif Re-
                      .or Faiility Dirpoff            D            j!      [ D' ate '
4 A /YA
                /Chaiffnan,      ~
                                              ~
At% 9/
Safety ahttrealth Departrnent' Date Reviewed by the Reactor Staff 1R4+          S*. 2 9 t/
Wrisht                Date I!.4<. . , s    c3-7n=il Date
                              &Ceort e / [- '24A9/
For; acka              Date A ch    M/k,fff Spence                Date
                                %%M              J9 a%91 Laugher'y 0              D'' ate
                                  /Atm %s % , h mui91 uyen ' "          D ate' L. htt+w 34ny+
Dwens                    Date Date Date Reviewed by RRFSC24 SEP1991 Date 4
g
 
OPERATIONAL FROCEDURE                                                  Procedure 1. T AB C EXTRACTOR SYSTEM GENERAL The extractor s, ya VM s iried for operability prior to the initial experiment for the day.
SPECIFIC
: 1. Assembly of the extractor system:
: a. Inside the exposure room:
(1) Move the inside receiver section into position in front of the core; screw tube supports to the floor .
(2) While holding the appropriate connecting tube in position, tie the strings in the tube to the two ends coming out of the exposure room wall and to the two ends in the receiver section.
(3) Align the ends of the tubes and slide the clamp over each joint.
(4) Place the alignment tools into the appropriate holes to check the tube alignment; tighten down the clamps.
(5) Remove the alignment tools,
: b. Outside the exposure room:
(1) Remove tube plug.
(2) Move the receiver section close to the tube projecting from the wall.
(3) While someone else is pushing the table toward the wall, insert two screws into the holes on the securing bracket (beneath the table).
(4) Tie the string from the end of the small tube to the end of the wire cable.
(5) Pull the string in the large tube slowly while having someone inside the room guide the string.
(6) When the cable is all the way through both tubes, thread the cable through the receiver tube while moving the receiver table into final po-sition against the wall (if necessary, add an additionallength of cable to the take up reel).
Revised: 15 May 91                                                                    Page 1
 
i i
!                                        (7) Position and tighten clamp over the joint; position carrier in tube j                                        and connect cable to each end; remove the tape on the take up teel.
i                                        (S) Pull back on the drive motor assembly untilthere is no slack in the
!                                        cables; tighten the adjustment bolts on the drive assembly.                                    ,
f                                        (9) Connect the electrical cables to the motor, control unit, and limit switches.                                                                                      ,
: 2. Disassembly:
i                              a. Reverse the order of the above with the following changes:
)                                        (1) Defore loosening the motor assembly, place tape on the cable drum
;                                        to keep the cable from moving (ensure the carrier is in the receiver sec.
tion).
i (2) Before pulling the cable through the tubes, attach a new string to it.
(3) Leave enough slack for disassembly inside the exposure room.
(4) Cut the string at thejointsin the room and tape the ends to the tubes or tie the ends together .
: b. Ensure the tube plug is in place, and the control unit is secured.
e                      3. Operations:
: a. On the motor control, initially set controls as follows:
(1) Power switch:"OFF".
(2) Torque control:"OFF".
(3) In/out switch: " BRAKE".
l                                        (4) Speed control:"0%".
: b. Plug motor controlinto AC outlet; switch the power switch to "ON".
: c. Switch in/out switch to appropriate position.
: d. Slowly increase speed to an appropriate level; as the carriage approaches its-i                            full in/out position, decrease the speed slowly to "0%"..
: c. Turn the in/out switch to " BRAKE".
: f. During retrieval operations, the Safety and' Health' Department (SHD) moni.                            -
tor will be present.
: Revised: 15 May 91'                                                                            Page 2
    . . - . . . - - . . - . - . . - . = - . - .                  . _ . - - . . .  . . . -. -    ._-- . .- - - . - ..- .. - . - ,- :
 
OPERATIONAL)      i PROCEDtJRE" 1 m, ,a w . .
u.-                                                  ,                        N - Procedure 1D .
r This procedure has              en approved by the Reactor Facility Director
                                /bf([j ;~ ~ - nnm04hw19r Re  ,t o r _ Facility pec]                                U
                                                                                                  'j              Dale
                        /L A&,                                                              MNcey 9(
                    "Chait' man, Safety and f(alth Department                                      ' D ate Reviewed by the Reactor Staff T /e/,rdl/                                929    *p / -
Wright'                                          Date L C-<. q                                    C-79Al Cyeotje                                          Date 4'Y%/[jjAL 79 A,4; F sba'cka                                        Date L(                              29 +7 % / I Spence                                          Datc h to.A~ o                                    14 %ff Laughcly0                                        date
                                      /h4w Nww 'h m% 9 /
                                    $gyed'                                            D'at e.
( l u 8) w % #f9/
                                                    ~
Owens                                            Date D ate .
Date Reviewed by RRFSc 24 SEP 1991 D ate -
  ?!                                                                                  k                                        $h?"
 
OPER AT10NAt.' PROCEDtJRE                                                                ,
W ' ' Ptocedure l, TAD D PNEUMATIC TRANSFER SYSTEM (PTS)                                                                                                            .
GENERAL
: 1. This PTS procedure is inactive. If the PTS Facility is reactivated, then this pro-cedure must be reviewed and approved by the RRFSC and the Reactor Facility Di-rector.
: 2. ALAR A principles will be prscticed during PTS operations.
: 3. All PTS operations will be directly supervised by a reactor operator present in the Hot Lab.
SPECIFIC
: 1. PTS Setup:
: a. Position core at 700 (inside region 111).
: b. Ensure communications are established between the hot lab and the control                                                                                      '
room.
: c. Inspect rabbits to be used in the PTS for cracks or other damage,
: d. Aluminum rabbits must be diverted to the Hot Cell and therefore may only be used on the "A" system.
: e. If the anticipated radiation level of an." returned rabbit is greater than 1.0 R/hr at 1 meter, take the following precautions:
(1) Use the remote control unit, unless experiment requirements dic-tate otherwise.
(2) Place a radiation survey meter next to the receiver / sender station so that it can be monitored from the remote control unit.
(3) The rabbit will be irradiated in the "A" system and then diverted to the Hot Cell or returned to the irradiation location.
: 2. Manual Operations:
: a. Ensure all switches on both the local and remote control units are in the "OFF" position; place the local / remote switch in the desired position,
: b. Place blower switch in the "ON" position.
l l                        Revised: 15 May 91                                                                                                                                      Page 1 i
            ._. , . , _ . _ . . _ , _ - . _ _ . _                                          . . - _ . . ~ _ , . _ . _ _ _ . - _ _ _ . _ _ . . . _ _ . . ~
: c. Insert key into local control unitt turn key to "0N" position.
: d. Ensure tubes are empty,
: c. Set mode switch (man /off/ auto) to "hf AN" position. Blower will start,
: f. Set in/out switch to the "OUT" position and the tube on/off switches to "ON"; allow the system to run for a short time,
: g. Set tube on/off switches to "OFF" and turn in/out switch to "lN".
: h. Load samples into tubes.
: i. Check communications with reactor operator at the reactor console.
J. When the reactor is at the designated power level, set the tube on/off-switches to "ON" one at a time, to send rabbits into the irradiation location.
: k. Begin stopwatch or timer.
: 1. Tur". sube on/off switches to "OFF" and turn in/out switch to "OUT".
: m. Ensure a Safety & Health Department (SHD) monitor is present during re-trievals.
: n. Set on/off switch to "ON" one at a time; rabbits will return to sender / receiver station,
: o. Set all switches to "OFF", and remove key from control unit.
: 3. Automatic hiode:
: a. Complete steps 2a d above,
: b. Set mode switch to " AUTO" position. Blower will start.
: c. Complete steps 2f i above.                                                                  '
: d. Set timer (0 to 5 minutes) by turning the red and black arrows to the desired irradiation time,
: c. When the reactor is at the desired power level, briefly push the timer push button and release. The rabbits willleave the receiver / sender station and will au.
tomatically return at the end of the preset irradiation period. The timer will au-tomatically reset.
: f. Turn all switches to "OFF" and remove kr.y from control unit.
Revised: 15 May 91                                                                      Page 2
 
i i                                                                                                                              -
l 4
i
: 4. Diverting Samples:                                                                                            l r
: a. Diversion of sarnples to the Hot Cell may only be made using the "A" systern.                              ;
i                b. After the rabbit has returned to the receiver / sender station, set the di.                                !
vert / send switch to " DIVERT" and hold it until the loading port handle trips to                            l the rear position,                                                                                            i i
;                c. Send the divert / send switch to " SEND" and hold for a few seconds. The rab.
bit will leave the receiver / sender station and travel to the Hot Cell.                                      l t
* i 3
1 l
1 l
l Revisedi 15 May 91                                                                                    Page 3 1
l............._......-._._...-,.-...--.-.-,_,-,,_.,.~, , , , , ,
 
i OPERAT10NAC JPROCEDURB'.                                                  Procedure 1E i
This procedure ha been approved by the Reactor Facility Director 1A($ta):                  ----_ b' A u C-            f Reacfor Facifity Dj/pt          D  Of    ly (l        D/te LA                ",l
* Au9/
Chairn6n, Safety and H61rif' Department              'D ate Reviewed by the Reactor Staff itM.4              T- M ' V/
Wrigtft                  Date
                                  /LW.,#                9 -1 9-41 l    I.iL 21/              /(
F 9,rjbacka              Date
                                @hww                &#AAI 3p~ence                  tfate
                                % %k -              JS w 1/
Laugheyf G                DWtc
                                !)rdsu 90.,.su 79 M A)
Date Sj{u)'eVi b ta J9MWff
                                                        $1 Owens "                    Date Date Date Reviewed by RRFSC24 SEP 1991 Date
            .g 't          '.- ,        .
 
'                                                                                                                    ^ ~
                                                                            'F OPERATION 5LIPROCEDURC@,
ar mm am m.w an T  ds~ Procedure
                                                                                                                - ~-a 1, TAB E IN-POOL /IN-CO}iE EXPERIMENTS GENERAL ALARA principles will be followed during these experiments. These procedures apply to all in pool or in core experiments except CET operations (See Procedure 1 Tab B).
SPECIFIC
: 1. All operations will be supervised by a licensed operator
: 2. Actions willbe taken to prevent damage to the reactor core or aluminum Mnk.
: 3. Ensure that a member of the reactor staff and a SHD representative are present in the reactor room during the removal of samples from in pool or in core loca.
tions.
: 4. The removal of experimental materials from the pool or core will be monitored with a radiation survey meter; additionally, a reactor operator will monitor the re-actor console during insertion and removal of in core /in pool experiments.
1 1                                                                                                                                              '
i I
l Revised: 15 May 91                                                                                                  Page'l
_. . _ _ _ _ . . _ _ . . _ . . . _ _ . _ . . - . - ~ . _ _                          _-__._.._.__.._u_..,.___,                _ _ _ -
 
1 1'
a.J
                                                                                                                                                          ~
                                                                                                            ^                            ^
                                                                                                                                                                                      . Procedute 2
;                                OPERA _TIONALS
                                    . - . .          -a ~.                  iPROCEDUiLEf mn                                                                              m-      4 1
i                                                                                                                                                                                                                      '
j                                                                                                                                                                                                                              l i
;                                This procedure has been approved by the Reactor Facility Director i                                                                                                                                          0                                                                                  ,
C11 Uf@ifi..
i i                                                                                                                            /          L 0          59 i            l            O/                                  M A<J              D ale W                                          '
j                                                                      Reactor Facility Directg l                                                                                        Reviewed by the Reactor Staff I
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                                                                                                                                                                                                                            ~
Reviewed by RRFSC 24 SEP 199I Date-
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x.
OPER ATION Ali PROCEDURE 3..                                    ,c Procedure 2 REACTOR STAFF TRAINING
: 1. The reacter staff training is delineated in the current "AFRR1 Reactor Opera-tor Requalification Program".
: 2. The Reactor Facility Director (RFD) determines who is allowed into the train-ing program. As part of the training /requalification program, the following will be performed:
: a. A training file will be maintained for each trainee / operator,
: b. When a section of training is completed, it will be annotated on the training checklist in each file.
Revised: 15 May 91                                                                Page1 I
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Reviewed by the Reactor Staff 1 town /            T. t 9 ?/
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                                              .                      - Reviewed by RRFsc 24 SEP1991 Date
                      - '  *f      '
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                                                      . ' .<:,.. .sp .. r gir. -
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                                                                                                                    ..    . Procedure 3 MAINTENANCE PROCEDURES GENERAL Maintenance procedures are provided in other references.
SPECIFIC
: 1. Preventive maintenance procedures for each item of the reactor systems are pro-vided in the maintenance logbook and console systems manuals.
: 2. Annual shutdown procedures are given in the Annual Shutdown Checklist which is revised each year by the Reactor Operations Supervisor (ROS) and ap-proved by the Reactor Facility Director (RFD).
: 3. Malfunctions are annotated in the Malfunction Logbook. Entries are normally made by the operator who discovered the deficiency. When corrective actions have been made and annotated in the m:lfunction logbook, the RFD or ROS shall re-view and initial the entry.
: 4. Procedures for maintenance of specific equipment are provided in the manufacturers' literature.
Revised: 15 May 91 Page 1 i
 
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  'This procedure has been approved fy the Reactor Facility Director
                    -    'i            ot,:n  wn      n=  m (6 #N9 [
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                      - A            m                            11-20-9]
Chalt in fafety and Health Department Date i
Reviewed by the Reactor Staff
                            )Wh                ).A N %'
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8t'd          Date Reviewed by RRFSC 17 DEC 1991 Date b'
 
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* 7    7"                  Pn,ahn 4 PERSONNEL          RADIATION          PROTECTION GENERAL All activities performed in areas of potential personnel radiation exposure will be done in accordance with ALARA principles. These areas are the reactor room, upper equipment room (3152), lower equip: at room (2158), ivarm storage, prep area, exposure room 1, exposure room 2, and the hot lab / cell. AF R R1 Instruction 6055.8B, Occupational Radiation Protection Program, is the radiation protection program followed by RSDR.
SPECIFIC
: 1. Reactor Room:
: a. CET Operations: See Procedure 1 Tab B.
: b. When working inside chained area around pool: The reactor operator on the console shall be respor.sible for controlling entry into the chained area during operations
: 2. Warm Storage: See HPP 31.
: 3. Prep Area: See Prep Area Briefing.
: 4. Exposure Rooms: See HPP 31 and Procedure 1 Tab A.
: 5. Hot Lab / Cell: See H PP 3-5 and Procedure 1-Tab D.
: 6. Upper and Lower Equipment Rooms:
: a. No written radiation protection procedures are required for entry into these rooms,
: b. Access to these areas is controlled by the AFRRI Reactor Physical Security Plan.
: 7. Personnel Dosimetry and Monitoring: See HPP 3-1,3 2, and the Prel' Area Briefing.
l -
1 Revised: 15 Nov 91                                                                Pace 1
 
OFERATIONAtMPPhcEDURENWA
                . a.    .m . 4            maa :          '      m, 7' As.e r. -              *      -
Procedure 5
[This procedure has been approved by.,,,.,.j..
                                                                          ,          the )q}eact,or, Facility Director
                                            "                                            !      h.h id (w Reactor Facility Dirq#or nn 's.      date Reviewed by the Reactor Staff 7%N                  7-19-9/
Wrigh't                        Date
_ L 6, ,,n                  9-WI
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                                                  ]9 hhs., Jfw 4f
_Owens                            Date Date Date Reviewed by RRFSC 24 SEP Igl .
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g            , ' . '- '' .          ,
                                                                    ...o    .
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  .g.,        -
yggg;g. gig.g                                    ;sjr; PHYSICAL SECURITY GENERAL Physical Security requirements are given in the AFRRI Reactor Physical Security Plan.
SPECIFIC
: 1. The reactor control room and the reactor room will be secured if no reactor -
staff member is present for a prolonged period of time during duty hours.
: 2. Control of keys is delegated to the Reactor Operations Supervisor. Key inven-tories will be performed annually, not to exceed 15 momhs..
: 3. The Physical Security Plan will be reviewed annually.
l l
l Revised: 15 May 91                                                          _Page1
 
3 OPERAT10NAlc.,,,_m,_-,,
: PROCEDURE                  .                                                                                              - Procedure 6 i                          .w                                                                          -
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i This procedure has been approved by the Reactor Facility Director                                                                                                i j\    Q w r V u..
                                                                                                                                          ~x,
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                                                                            $)            .
jb                                    Tbc.a 9/                                      ;
Reactor Facility Direep                                                                              D' ate I                                                                                                                                                                                          1 Reviewed by the Reactor Staff                                                                                      :
stL}/                        T 2 '/ 1/
Wrighi                                Date                                                                '
AL ,#                            v- r 9- 9l
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                                                                                ,Fppbadka                                Date
                                                                              %m                              Wrfif/
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                                                                              $guyen#                                    Ddte (IoLA VH7)xet                                                                                                -
Owens                                    Date.
Date Date i
Reviewed by.RRFSC i
Date
                                                                                                                                                                                      ~
4'                                                      . '
k 4                                .
s                        ,
  . _ _ . _ - . _ _ _ . _ . . .                                        .m__ . _ m. .. _                                  _ _ . . _ . . . _ . , . _ _ _ _ _ . _ . - . _ _ . . _ . _
 
l OPERAyONAL]ROCEDURE,7 hocedet 6
                                              =_                                    -.
EMERGENCY PROCEDURES GENERAL The reactor ernergency organization, emergency classes, and emergency action lev-els are set forth in the AFRRI Reactor Facility Emergency Plan and its implement-ing Procedures.
SPECIFIC Perform the following, as appropriate (need not be done in order).
: 1. Reactor Emergency:
: a. SCR AM reactor,
: b. Check radiation monitors; use portable survey instruments to assess situa-tion, if necessary.
: c. Notify ERT Commander of situation,
: d. Activate emergency organization.
: 2. AFRR1 Complex Emergency Evacuation:
: a. SCR AM reactor,
: b. Secure any exposure facilities which are in use so that personnel access to that facility is not possible,
: c. Remove logbook, emergency guide, radios, teletector, tool kit, and keys; re-port to ERT.
: d. Ensure reactor area doors are secured upon departure.
: 3. Proper classification of emergency situation: All SROs must review the vefer-enced Emergency Plan documents and be able to properly classify the events as they occur. Below is a tabulation of emergency classification to be used as guid-ance.
Revised: 15 May. 91                                                            Page 1
 
4 i
EMERGENCY Radiation              Activate AFRRI Complex      Activate Emergency
.          CLASS      Alarms        Emergency Evacuation          Response Team (Unanticipated)                                                  l l    Class O Fire Alarm (non reactor)                  Yes                Yes Class 1 R 1 > 1 min.                                      Yes.
R2> 1 min.
* Yes R3                              No                No R5 > 1 min.
* Yes R6                            No                No E3 > 1 min.
* Yes E6 >1 min.
* Yes SOM> 1 min.
* Yes Reactor Stack Fan Monitor                        No                No Fire Alarm (reactor)                      Yes                Yes Class 2 CAM > 'l min.
concurrent with R1, R2,-
;                      RS, and/or SGM
* Yes NOTE:
* A decision to evacuate the Institute will be made by the ECP Commander based on input from the ERT Commander.
l
  --Revised: 15 May 91-                                                        Page 2
 
i l
i OPERATIONAL ' PROCEDURE ?                                                                                Procedure 7 d
r I
il'his procedure has been approved by the Reactor Facility Director                                                    j I                                                                      -    .-      . ..
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Reviewed by the Reactor Staff 17sCO                r. 29-?/
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Uyen' ' ~                            Diite d 9Aiut 1Mheet Owens                                  Date Date Date Reviewed by RRFSC24 SEP 1991 Date
          $ f { . Nb $ l ;'' ? f ^.* { .. ? f l f ? O D f ]..f & h i? [ & ( ,
 
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.                                    REACTOR CORE LOADING AND UNLOADING GENERAL Loading and unloading of the reactor core shall be perforrned under the supervi-sion of the Reactor Facility Director or the Reactor Operations Supervisor.
SPECIFIC                                                                                          :
: 1. Setup
: a. Ensure that at least one nuclear instrumentation channelis operational.
: b. Ensure that the source is in core.
: c. Ensure that an operator monitors the reactor console during all fuel move.
ments.
: d. Check new FFCRs before insertion into the core; this includes cleaning, vi.                .
sual inspection, and length and bow measurements,
: c. Install all control rods,
: f. If irradiated fuel elements are to be removed unshielded from the pool, obtain a Special Work Permit (SWP) from the Safety and Health Department (SHD);
do not remove fuel elements with a power history (greater than 1 KW) in the previous 2 weeks from the reactor pool.
: 2. Core loading
: a. After each step of fuel movement perform the following:
(1) Record detector readings.
;                                  (2) Withdraw control rods 50%; record readings.
(3) Withdraw control rods 100%; record readings.
(4) Calculate 1/M.
(5) Plot 1/M versus number of elements (and total ma.es of U-235).
,                                  (6) Predict critical loading, i
(7) Insert ALL control rods; continue to next step.
: b. Load fuel elements in the following order:
(1) Load the B ring and C ring thermocouple elements.
Revised: 15 May 91                                                                  Page 1 L.                        - . .            = a. -        - - - -            - . - - - -        ;-------"
 
1 (2) Connect thermocouple outputs to reactor control console display.
(3) Install any other thermocouple elements.
(4) Complete loading of Ib and C ting elements (total of 18 standard el-ements plus 3 FFCRs).
(5) Load D ring (total of 33 standard elements plus 3 FFCRs)
(6) Load the following E ring elements in order:
16,17,18,20,6,8, 9,10 (total of 41 star.dard elements plus 3 FFCRs).
(7) Complete the E ring by loading the following elements in order:
15,21,11,5,14,22,4,12,13,1 (total of 51 standard clernents plus 3 FFCRs)
(8) Load the following F ring elements in two elemems per step until criticality is achieved using the following loading order:
22,23,24,21,20,25,26,27,28,29,30,1,2,3,4,5,19,18,17,16,15, 14,13,6,12,7,11,8,10,9.
Once criticality has been achier,:d, perform control rod worth measure-ments at core position 500 by rod drop technique. Calculate shutdown margin:
SDM = Total Control Rod Worth          K excess  TRANS Rod Worth (9) Load core to $2,00 excess reactivity by loading two elements per step using the loading order in instruction 8.
(10) Wrify control rod worth using rod drop techniques, calculate SDM (11' aac the core to achieve a K excess that will allow calibratiori of the TRANS rod based on the last available worth curve of the TRANS rod (approxirnately $4.00). Calculate the reactivity value of each ele-ment as it is added.
(12) Calibrate all control rods.
(13) Calculate the shutdown margin.
(14) Estimate K excess with a fully loaded core (must not exceed $5,00).
(15) Load core to fully operational load using loading order in instruc-tion 8, and recalibrate all control rods. Calculate the shutdown margin.
(16) Adjust the core loading pattern to meet operational requirements if necessary. Recalibrate all control rods. Calculate the shutdown mat-gin.
Revised: 15 May 91                                                                Page 2
 
8.
At the end of each dry in which a Daily Operational Startup Checklist or Daily Safety Checklist has been completed, perform a Daily Operational Shutdown Checklist (Tab I),
: 9. Complete the monthly summary .
: 10. Respirator equipment will not be used on a routine basis. Respirator equipment is provided for use during emergency conditions only.
l l
l Revised: 15 Nov 91 Page 2 4
 
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This procedure has been approved by the Reactor Facility Director
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                                                                                                                      ~
Reactor Facility Dirgr                                                                                Date P
Reviewed by the Reactor Staff A & W p snwij                                                        ~
G orge s                                                                Date
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                                'Fo sbhb '                                                                  Date usa 2 Fee 1/
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Lauhhei9                                                                Date WT.Nw 492 4/
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0 M d.ff t 6
                      ,--                REACTOR      OPERATIONS GENERAL Logbook entries will be made in accordance with the Logbook Entry Checklist (Tab A).
SPECIFIC
: 1. The names of the individuals who supervised and performed the daily and weekly checklists will be shown at the top of the checklist. Checkmarks or num-bers, as appropriate, will then be entered on each checklist line as that item is formed.
: 2. Perform reactor Daily Operational Startup Checklist (Tab B), utilizing appro ate nuclear instrumentation set points (Tab C). In the case of no planned opera-tions, a Daily Safety Checklist (Tab B1) may be performed.
: 3. Record at the beginning of each day in the reactor operations logbook the SR on-call for that date. At the end of the "zeding hy,                                                    WRW1
    =ce!! fer +he gec : ; - gh+, "feeend, er he!!&y _(Med 'he m:ee                                          /J W 1 2.
of 'S
: 4. At the begining of each working day, also record
* ame of the physicist in charge (PIC)present at the reactor facility. If th " C vised entry will be made in the logbook.                  9        changes during the day, a re-either the PIC or SRo on-call
: 5. Perform K-excess measurement (Tab D).
: 6. Perform operations in accordance with the following:
: a. Steady state operation (Tab E).
: b. Square wave operation (Tab F).
: c. Pulse operation (Tab G).
: d. CET operations (Procedure 1, Tab B).
: e. Pneumatic Transfer System (r'rocedure 1, Tab D).
: f. In-pool /in-core experiment (Procedure 1, Tab E)
: 7. Perform Weekly. Operational Instrument Checklist once during each calen week (Tab H).
i Revised: 15 Nov 91
:                                                                                                      Page 1
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:                              Reviewed by the Reactor Staff 4W                        Ilusj Get!e                        Date
                                    /$Yo 1~l N                    tisJint Pors ek'a                      Date
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Date Date Reviewed by RRFSc 2 4 SEP 1991 Date l
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LOCBOOK ENTRY CHECKLIST
: 1. The reactor operations logbook is a before the-fact record, that is, entries will be logged before the operator actually performs the planned function. Any late en-tries will be so noted.
: 2. The operations logbook will have a hardbound cover and will be sequentially numbered by volume. The pages will be dated at the top of each page and each page will be sequentially numbered.
: 3. The Reactor Facility Director (RFD) will review each logbook upon its comple-tion; he will make an appropriate entry in the back of the logbook and sign the entry. The operator who makes the final entry at the end of a logbook is responsi-ble for ensuring that the ROS is notified that the logbook is ready for RFD review.
: 4. Allitems in GREEN (see below) that are not closed out during the working day will be carried in GREEN at the end of the day and again at the beginning of the next operational day.
: 5. The entries will be made in ink and in accordance with the following designated color code:
: a. BLACK and BLUE-BLACK:
(1) Console locked and unlocked. The individual at the console will enter his/her name and the supervisory licensed operator's name, if necessary.
(2) Checklist number and completion time.
(3) Power level at criticality and subsequent power level changes.
(4) Reactor SCRAM.
(5) Mode of operations. Use appropriate stamp or entry to designate the op-eration:
(a) Steady State (b) Square Wave (c) Pulse Revised: 22 JUL 91 Page1                ;
                                                                                                      \
                                                                                      .---_m__._._-.a
 
(6) Operation of reactor associated facilities such as lead shield doors, pneu-matic tube systems, etc., unless such operations cause a change of reactivity (see 5.b.(2) below).
(7) Change of personnel at the console. Name of personnel v.ill be entered along with the licensed operator present in the control room, if the person at the console is not a licensed operator.
(8) The operator in charge will be designated in the logbook whenever multi-ple operators are signed on the console.
(9) Completion of the daily startup and shutdown checklists and weekly checklist.
(10) Signature of reactor operator to close out '.ne log for the day.
(11) Designation of the SRO on call and physicist in charge (PIC).
(12) Reactor calibrations and data.
(13) Allline outs, entry errors, changes in mode of operation stamp lines, and end of page line outs will be initialed or signed by the operator,
: b. RED:
(1) K-excess measurements, to include experiment worth determinations.
(2) Actions which affect reactivity:
(a) Core movement.
(b) Fuel movement.
(c) Control rod physical removal for maintenance.
(d) Experiment loading and removal from the CET, PTS, pool, or core.
: c. GREEN:
(1) Any reactor malfunctions noted upon discovery / occurrence with a second entry noting corrective action has been completed.
(2) Additionalitems entered at the discretion of the operator such as addi-tion of makeup water to the reactor pool, etc.
(3) Any Technical Specification required equipment taken out of service for any reason. A second entry is made when the unit is returned to service.
, 6. When an operation requiring entry into the logbook falls under more than one
; color code, the color to be used will be determined via the following order of prece-dence: RED - GREEN - BLACK / BLUE-BLACK.
Revised: 22 JUL 91                                                                  Page 2
 
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                                                                                                .              .c,..
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                                                                                . Reviewed by RRFS$4 SEP 1991 Date bhb5?$$l$?ll ll%$?AblNQQ jl}l$f[
 
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                                                                    ~ m,                                  '
Procedure 8. TAB B a
DAILY O_PERATIONAL STARTUP CHECKLIST Checklist number                                                                                            Date SRO On Call                                                                                                  Supervised by Assisted by Operators                                                                                                    Time completed I. EQUIPMENT ROOM (Room 3152)
: 1. Air comp ressor pre ssu re (psi) . ... ... .... . ..... ..... .. ..... .. .. . .. .. . .. . . ..                                                                            ;
: 2. Air compressor water trap drained .......... .... ....................... ..
3 . Ai r d ry e r Op e ra t in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 4. Doors 231,231 A, 3152, and roof hatch SECURED.................
II. LOBBY AREA Lobby andio ala rm tu m e d off . . .. .. .. . . .... . . . . . . . .. . . .. . ...... . ... . .. .. . . ... . .
111. EQUIPMENT ROOM (Room 2158)
: 1. Pre filte r diffe re ntial pressure ......... ... .. .. ..... . ... .........................
: 2. Primary discharge pressure (psi) ..............................................
: 3. Deminetalizer flow rates set to 6 gpm ........................... .........
: 4. Staek roughing filter (inches of wate r) .....................................__.
: 5. Stack absolute filter (i.m e s o f wat e r) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 6. Visu al inspeetion of are a .. . ........ . .. . . . . . ... ... . ... . . .. . . ... ... . .. . .. . ..
: 7. D oo r 215 8 S E C U RE D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
IV. PREPARATION AREA Visu al i n spection of are a .. .. .. . . . . .. . . .. .. . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . .. . . . . . .
V. REACTOR ROOM (Room 3161)
: 1. Tt ansient rod air pressure (psi) .............................. .................
: 2. Shielding doors bearing air pressure (psi) ................................
: 3. Visual inspection of core and tank ...........................................
: 4. Number of fuel elements and                                                      fu el elem en t s . .... .. .......... ..
control rods in tank storage                                                  cont rol r o d s .... ........ ...... ....
: 5. Air particulate monitor (CAM)
(a ) Op era tin g a n d Tra cin g . .. .. .. .. .. .. . . . . .. . .. . .. . . . . . . . . .. .. .. .. . . . . . . .. . . .. ..
(b) Alarm test completed, damper closure verified...................
: 6. D oo r 316_2 SECU RED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 7. Stack gas monitor quality assurance checked ........ .................
_J AFRR1 FORM 61a (R)                                                    Revised: 15 May 91                                                                                  Page 1 1
l
 
I 4                                                                                                                                                                                                                                        i VI. REACTOR CONTROL ROOM (Room 3160)
: 1. Emergency ait dampers teset                                      ......................................................
: 2. Console recorders dated -                              ..........................................................                                                                                !
: 3. Staek flow and fuci temperature recorders dated                                                                              .........................
l
: 4. Logbook dated and reviewed                                          ........................................................                                                                      j
: 5. Water monitor box (resistivity must be > 0.5 Mohm cm)                                                                                                                                              1 (a) Background aetivity(epm) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
(b) Water monitor box resistivity [Mohm cm) ....................................
(c) D M 1 r esistivit y [M o h m cm ) ....,,. .......... .. .... . ... ........ . .. . ...... ... ... .. . ..
(d) D M 2 resisti ity (M o h m cm] .... ...... ... .. .......... .. .. .... .. .... .... .. ... . ..... .
: 6. S t a ck g a s flo w r a t e [K cfm ) . . . . . . . . . . . . . . .. . . . . . . . . . . .. .. . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . .
: 7. S t a ek lin ca r flow ra t e (K ft/ min) ........... .. .. .. .. .... .. ... ... ... .. ... . . ..... . .. ... . . .
: 8. Gas stack monitor (a) Background (cpm) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
(b) Alarm check . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
(c) H igh ala rm set t o 800 M P C Ar 41 .................................................
: 9. Radiation monitors Monitor                      Alarm Point                                            Reading                                                              Alarm Setting Functional                                          (mR/hr)                                                                          (mR/hr)
(a) R 1                                                                                                                                                                              500 (b) R 2                                                                                                                                                                                10 (c) R 3                                                                                                                                                                                10 (d) R 5                                                                                                                                                                                50 (c) E-3                                                                                                                                                                                10 (f) E 6                                                                                                                                                                                  10
: 10. TV monitors on                      .............................................................................
: 11. CAM high level audible alarm check . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                                              __
: 12. Wa t er t em p e r a t u re (in let) . .. . . . . . . .. .. .. .. .. .... .. .. .. .. .. . . .. .. .. .. .. .. .. .. .. .... .. .. .. . .
: 13. Water levellog completed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 14. Time delay operative                            ....................................................................
: 15. Source level power grcater/ equal to 0.5 cps. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 16. Prestatt operability checks performed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
!                    17. Interlock Tests (a) Rod raising, SS mode                                                                    (c) I kW/ Pulse mode (b) Rod raising, Pulse mode                                                                  (f) NM-1000 HV (c) Source- RWP                                                                              (g) Inlet Temp -
_ (d) Period RWP
                    '18. SCRAM checks (at least one per rod)
(a) % Power 1                                                                              (h) Reactor key
                      - (b) % Power 2 (i) Manual                                                                                                          '
                        -(c) Fuel temp 1                                                                              (J) Emergency Stop
,                        (d) Fuel temp 2                                                                              (k) Timer l
(e) HV loss 1                                                                                (1) CSC Watchdog (f) HV loss 2                                                                              (m) DAC Watchdog (g) Poollevel
: 19. Z ero p o wer p ulse . . . . . . . . . . .. . . . .. . . . .. . . .. . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . .
L l
l AFRRI FORM 61a (R)                                                    Revised: 15 May 91                                                                                                            Page 2
            . ._              _      _ . _ ~___.                                                  _ . _ - -                                              . . _ _ , _                                    _ . _ _      _ -    _ _ . _ -
 
OPERATIONA13      uuam.a w%-.                        IPROCEDU.REi_d w              .u :a          aew, m:  _
O .e--,".3  . m&x - ..
                                                                                                                                                                    ' Y'Piscidurf 8, TAB B1 l
This procedure has been approved by the Reactor Facility Director
                                                                                                                                                                                                          ^ ~
                                                                                                                                "!b=::ngg g%,'        %
                                                                                                                                        . ; 7 y';j' yli I                4 9/
Reactor Facility Dir(dtor                                                                        /Date Reviewed by the Reactor Staff TWalI                      f. 2M-9)
Wright                                Date it. c u m.,                      r- 2 Mi cor e                              Date
                                                                                                                  .. k 248,1/
Fo sbaMa                              Date n                n%4/
fpence                            ' Date                                                                          '
                                                                                                    #1 L s}m                    AMhr 9/
Laughet'y CT                          tate-O G k wyn 79 nau,9' Date fpuyen'o is b ut v dYr J W 7/
Owens                                    Date Date Date
                                                                                                                              ' Reviewed by RRFSc 2 4 SEP 1991 D ate 4
                                                                                                                                                    ,      ,            'c                    8 '
1
  ,        ..          m
                                                                                                                                                        -      %--  , , .    --..r.-,    -,ev,    -m-- , -w-      -
 
  . - . ... . - -                        . ~ . . ~ . - - ~                                _ - - .                        . - .          . . -                . - - -                .          . -
l
,                                                                                                                                                                                                        .1 1
l 1
OPERATIONAIF PROCEDU,REi                  , m a um w a d          %.*IE . m-iI                          ._                    s._
                                                                                                                                                                ~ Procedure 8. TAB B1 DAILY SAFETY CHECKLIST Checklist number                                                                                            Date                                                      _
SRO On Call                                                                                                  Supervised by Assisted by Operators                                                                                                    Time completed I. EQUIPMENT ROOM (Room 3152)
: 1. Air compressor pressure (psi) ............................................ .....
: 2. Air compressor water trap drained ..........................................
: 3. Ai r d ry e r Op e ra t in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .      ,
: 4. Doors 231,231 A, 3152, and roof hatch SECURED.................
II. EQUIPMENT ROOM (Room 2158)
: 1. Prefilter differential pressure ....................... ............................
: 2. Primary discharge pressure (psi) ..............................................
: 3. Demineralizer flow rates set to 6 gpm .....................................,
: 4. Stack roughing filter (inches of water) .....................................
: 5. Stack absolute filter (inches of water) ......................................
: 6. Visual ins pe etio n o f area ............... .... . ........ .. .. ... ... ... ............... ..
: 7. D oo r 215 8 S E C U R ED . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
III. PREPARATION AREA Visual ins pe etion of are a ....... ............................. ........................ ..
IV. REACTOR ROOM (Room 3161)
: 1. Transient rod air pressure (psi) ................................................
: 2. Shielding doors bearing air pressure (psi) ................................
: 3. Visual inspection _ of core and tank ........................................ ..
        -4. Number of fuel elements and                                                        fu el elements ...................
control rods in tank storage                                                cont ro l ro d s . . . . . . .... ... . .. .. .. ..
: 5. Air particulate monitor (CAM)
(a) Op e ra t in g a n d Tr a cin g ..... .. ........ . . .. .... .. . . .... .. .. .. . . . . .. .. .... .. .. ..
(b) Alarm test completed, damper closure verified...................
: 6. Doo r 3162 - SECURE D . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . .
        -7. Stack gas monitor quality assurance checked ..........................
AFRRI FORM 61b (R)                                                      Revised: 15 May 91                                                                                      Page 1-
                                                                                                                                                          ... .              ... .. ~ _ - .._ ._._ -..
 
i V. LOBBY- AREA I
Lo b b y a u d io a la r m t u r n e d o ff . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
I VI. REACTOR CONTROL ROOM (Room 3160)
: 1. Emergency ait dampers teset                                    ......................................................                                            !
: 2. Console recorders dated                                    ................................................................                                        1
: 3. Stack flow and fuel temperature recorders dated                                              ..........................
: 4. Logbook dated and reviewed
: 5. Water monitor box (resistivity must be > 0.5 Mohm cm)
(a) B a ckgro u n d a etivit y(cp m) ..... .. ... ....... ... .. .... .... . ...... .. ....... . ... .. ... . . .
(b) Water m onitor box resistivity (Mohm-cm) ....................................
(c) D M 1 r esist ivit y [M o h m -cm] . .............. ... ... . .. ... ...... ...... ........ . ... .. .. .. .
(d) D M 2 r esistivit y [M o h m cm ) . .... .. .... . ... .. .... .. ... .... .. ...... ....... ............
: 6. S t a ck ga s flo w ra t e [K cfm ] . . .. . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .
: 7. S t a ck lin ea r flow r a t e (K ft/ min) . ... .... .............. ....... .. ... . .. . .. ... . ..... .. . .... . .
: 8. Gas stack monitor (a) Baekground (cpm)
(b) Alarm check (c) High alarm set to 800 M PC Ar-41 ................................................
: 9. Radiation monitors Monitor                                  Alarm Point                    Reading                          Alarm Setting Functional                      (mR/hr)
(a) R 1                                                                                                        (mR/hr) 500 (b) R 2 10 (c) R-3
_                                10 (d) R-5 50 (c) E-3 (fj E-6                                                                                                                        10 10
: 10. TV monitors on
: 11. CAM high level audible alarm check
: 12. Wa t er t em p era t u re (inle t) ........ ... ... ...... .. .. .. .. . .. ... .. .. ...........................
: 13. Water level log completed
: 14. Source level power greater /equa1 to 0.5 cps.                                    .....................................
1
{
I AFRRI FORM 61b (R)                                              Revised: 15 May 91                                                          Page 2
 
                                                      ~
OPERATIONAL                                                                                      Procedure 8. TAB C
                'PROCEDURH[
i This procedure has been approved by the Reactor Facility Director
  \
                                                                                .;i -
                                                                  - .      . .0 V M, a , i i    I 2RAEm l/
Reactor Facility Directp                                                    ' D a t e' l                                    Reviewed by the Reactor Staff 7'. w u L t.      S*-'4- 9 I Wright '              Date IL G.,a              9 -71 9 b    l (L 79          1%
F rsbacka              Date
,                                          _ nce-          2144y'11 1 pence              ~ Date
                                            & l m /a 29 Mle,f/                                                            '
Laughdy(/              bate Adu , % ruin, ln mau 9(
uyen #              Dite bobran d7n7N4/
                                          '0 wens                  D ate Date Date Reviewed by RRFSC 24 SEP1991 Date Ik      b                              bhhhhf}bhhhhhh$$[hfhIbi,
 
w;.      , .
                            ~ . . 3. s. . . . .gggg;ggy                            ggg NUCLEAR            INSTRUMENTATION          SET    POINTS              l GENERAL 1
These set points may be adjusted for a specific operation by the RFD or ROS but in no case may they be set at a point non conservative to the technical specifications.
SPECIFIC The following are channel or monitor set points (alarm, scram, rod withdrawal prevent),
: 1. Scrams:
: a. Fuel Temperaturc 1 & 2:          575 degrees C
: b. High Flux 1 & 2:                  110% (1.1 M W) 1
: c. Safe Chambers 1 & 2 HV Loss:      Loss of 20%
: d. Pulse Timer:                      Less than 15 seconds
: e. Steady State Timer:                as necessary
: 2. Rod Withdrawal Prevents:
: a. Period:                            3 seconds b.1 KW (Pulse Mode):                  1KW
: c. Source:                            0.5 CPS
: d. Water Inlet Temperature:          60 degrees C
: e. Fission Chamber HV Loss:          20 %
: 3. Alarms:
: a. R AM S:                          As directed in procedures
: b. CAMS:                              10,000 CPM
: c. Stack Gas:                        800 MPC Ar-41
: d. Water Monitor Box Gamma:          7000 CPM
: e. Criticality Monitor (RS):          50 mR/hr day -
l                                                    20 mR/hr night Revised: 15 May 91                                                          Page 1
 
OPERATIONAISf-PROCEDUREiQ@:MW3CM'J'
      - :~..un  .a e    . x:a c:
                                                          @IE"
                                                              -~
                                                                    . IdfNPiEcEdhre58,TABD
                                                                    ''. . w w.,a ,
c l
l                                                                                                            i This procedure has been approved by the Reactor Facility Director 3        .,..,z  .77                                        l j          ,[            }    j'    g7 4 Reactot' Pa~cility Dis /ctor                              Date Reviewed by the Reactor Staff 744 {l-          >K l9-9/
Wright '                Date ILL<., n                T-h-9
                              //              !  b 79        s, o sbacka              Dafe v .      %            Eff)lsfif fpence                  Date hf_$~rla            09 'mm f/
Lau thir>C)              D' ate d 4e Lunt Wh >vu 4l gyen'f'                  Date L J J k , k & n d w h *7/
O'wens ~                  Date Date Date Reviewed by RRFSC24 SEP1991 D ate
    ,          y      , *          "* 4
                                    .        *,                                        4 F
 
1 1
ggq,,.      . \ ;L ::.g; -
                                                      ~ yp;ygg; ggg. nun wgw      aggwymrg ..;;;;..y                    y g,.
                    . w.
                                                                                .suzu                        .. . a.
2      . .
K-EXCESS                                                                            !
4 E FF ECTI V -
: 1. Withdraw SAF and SHIM rods 100% ar.d ..i:h& . :he TR AMS :od 25 u.                                            leTBu9
: 2. Use the REG rod to bring the reactor to cold critical at 5 watts. If criticality can not be reached with the REG rod full out, use the TRANS rod to bring to crit-ical.
: 3. When power is stabilized at 5 watts, record rod positions in reactor operations
:                    logbook, entering all information in red ink.
: 4. Using rod worth curves, compute K-excess for the core position used and re-cord in the reactor operations logbook .
Note: Use the curves for infinite water when doing K-excess between positions 300 and 700, 4
4 i
i t
l Revised: 15 May 91                                                                                          Page1; l
 
OPERATION      ^1hfPRDCEDURE" a -~ ,u,r s .. -.    :,ws ..
Procedure 8. TAB E This procedure has been approved by the Reactor Facility Director
                                                                                                  *:n-i llo a.m
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Wright'                  Date il kw              s-M -St
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U~ wens.                  Date Date D ate -
Reviewed by RRFSC24 SEP1991 D ate -
                                                                                                                                ^      ' '
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OPERA~. 110NAlf?                                ey y~    .'i      t Pr60edure 8, TAB E m . .~P.ROGDUREt
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STEADY STATE OPERATION GENERAL The reactor shall not be operated at a power greater than '.l.0 MW.
SPECIFIC
: 1. Set the mode switch to manual mode and clear allwarning messages and scrams.
: 2. For runs greater than 200 KW, adjest alarm points on R-1 and R-5 to full scale.
: 3. Raise control rods with the appropriate banking, taking into consideration the location in the pool, power level, and experimental requirements.
: 4. If final approach to criticalis to be made in Auto mode, perform the following:
: a. Set the the thumb wheel dials to the desired power.
: b. Raise the rods to the appropriste banking.
: c. Select the rods that are to servoed.
: d. Make sure that all rods that will be servoed have been raised at least 5%
: e. Enter Auto mode.
: 5. Scram the reactor at the end of the run using the manual or timer scram.
: 6. Ensure the appropriate entries have been made in the operations logbook.
: 7. If no further steady state runs, square waves or pulses are anticipated, adjust R-1 and R-5 alarm points to their normal settings.
Revised: 15 May 91                                                                    Page1
 
                                                                                ^~              ~#
OPERATIONAlfd_PROCEDUREP
                              . - . - - +              - . - - . _ . ,            ^                            "" _ M.,, j PrClure  S TAB F1
                                                                                                                                  - _ .~  -_
i This procediare has been approved by the Reactor Facility- Director
                                                                                              , .,y -            .,
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_      5 -Date 2, /,,
Reviewed by the Reactor Staff 7 Wx$4                  ., . m . ii Wright                            Date
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                          - ,                    -                  ._..=
 
I'ERANg]Rg@pyR                              j        ', jg,]pfmyM M SQUARE WAVE OPERATION (Suberitical)
GENERAL The square wave mode will not be used above a demand power of 250KW.
SPECIFIC
: 1. Set R1 and R5 to full scale
: 2. Determine the transient rod critical position using the core position, the final transient rod position, the rod curves and the equation below. Note that a square wave insertion can not exceed 75 cents.
CRITICAL POSITION (S) = FINAL POSITION (S)-INSERTION (S)
: 3. Apply air to the transient rod and raise the anvil to the citical position that was calculated above.
: 4. Bring the reactor cold critical using the three standard contral rods; use a rod configuration commensurate with the core position and experimental requirements.
If Auto Mode is used, select the rods to be used. Ensure thet 6ese rods have been raised at least 5% before entering Auto Mode. Set the cold critical power level on the Power Demand thumb wheels and enter Auto Mode.
: 5. Stabilize the reactor in Manual Mode.
: 6. Set power demand thumb wheels to desired power level.
: 7. Select the standard control rods to be servoed. Make sure that all control rods to be servoed have been raced at least 5%.
: 8. Scram the transient rod.
: 9. Raise the anvil to the desired final position.
: 10. Allow the power level to fall below 10 watts.
: 11. Switch into Square Wave mode.
: 12. Depress Fire button.
: 13. As the power level approaches the power demand level, the console will switch into Auto Mode. If power can not reach the demand power,it will automatically change to manual mode. At this time, either switch to Auto Mode or bring the re-actor to the desired power level manually.
1 Revised: 15 May 91                                                                  Page1
 
l
: 14. Scram the reactor at the end of the run using the manual or timer scram.
: 15. Ensure all pertinent information has been entered in the reactor operations log-book.
: 16. If no forther steady state runs, square waves or pulses are anticipated, adjust R 1 and R-5 alarm points to their normal settings.
Revised: 15 May 91                                                              Page 2
 
OPERATIONild:JPdOCEDORE$38
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8, TAB F2 This procedure has been approved by the Reactor Facility Director A                    c_ kh                                        ,
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PERgTig g [ fRgg p p { [ i n 6 y > ,                            Wfm 8, TAB F2 SOUARE WAVE OPERATION (Critical)
GENERAL The square wave mode will not be used above a demand power of 250KW.
SPECIFIC
: 1. Set R-1 and R-5 to full scale.
: 2. Bring the reactor cold critical using the three standard control rods; use a rod configuration commensurate with the core position and experimental requirements.
If Auto Mode is used, select the rods to be used, ensure that these rods have been raised at least 5% before entering Auto Mode, set the cold critical power level on the Power Demand thumb wheels, and enter Auto Mode.
: 3. Determine TRANS rod anvil setting for desired insertion. Insertion cannot ex-ceed 75 cents. Raise the anvil to that setting.
: 4. Stabilize the reactor in Manual Mode.
: 5. Set power demand thumb wheels to desired power level.
: 6. Select the standard control rods to be servoed. Make sure that all control rods to be servoed have been raised at least 5%.
: 7. Switch into Square Wave mode.
: 8. Depress Fire button.
: 9. As the power level approaches the power demand level, the console will switch into Auto Mode. If power can not reach the demand power, it will automatically change to Manual Mode. At this time, either switch to Auto Mode or bring the re-actor to the desired power level manually.
: 10. Scram the reactor at the end of the run using the manual or timer scram.
: 11. Ensure all pertinent information has been entered in the reactor operations log-book.
: 12. If no further steady state runs, square waves, or pulses are anticipated, adjust R-1 and R-5 alarm points to their normal settings.
Revised: 15 May 91                                                                Page 1
 
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This procedure has been approved by the Reactor Facility Director f                          !du u UL@ y            .J
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Reviewed by the Reactor Staff R/d O                                  S' 27 ti Wright                                              Date-LC.-                                        9-T %I e      c f                    Date -
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Date -
Reviewed by RRFSC 24 SEP1991
                                                                                                                      - D ate -
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OPERATIONAL    l PROCEDURE'                                    f DPlocedure 8,' TAB G1 -
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PULSE CyERATION (CRITICAL)
GENERAL Pulses above S2.00 must be approved by the RFD (prior to pulse initiation). Spec-ification on the RUR may be used to meet this requirement.
SPECIFIC
: 1. Set the alarm points on R-1 a:-ri R-$ (criticality monitor) to full scale.
: 2. Bring the reactor cold critical using the three standard control rods; use a rod configuration commensurate with core position and experimental requirements.
Note: A series of repetitive pulses may be fired using the same rod positions on the same day as long as the reactor power is not increasing and is less than 1 kW.
: 3. Stabilize in the manual mode.
: 4. Raise the transient rod anvil to the desired pulse position. (This position is ob-tained from the control rod worth curves for the appropriate core operating posi-tion)
: 5. Select the proper pulse detector according to the table below. If the Cerenkov detector is selected, turn off the reactor room and tank lights.
Detector 1 = Pulse Ion (Maximum insertion = S2.00)
Detector 2 = Cerenkov (Maximum insertion = S4.00)
: 6. Enter Pulse Mode and enter an identifying string at the prompt, The power level must be below 1 kW to enter Pulse Mode.
: 7. Fire the pulse by depressing the " Fire" button on the reactor console.
: 8. Record the appropriate data in the reactor operations logbook from the pulse display.
: 9. Reset R-1 and R-5 to their normal alarm points when pulsing, square wave, or steady state operations are complete.
l Revised: 15 May 91 Page 1
 
                                                                                                                                                            ?
OP~ERXTION
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                                                                                                          -Date Date Reviewed by RRFSC 24 SEP1991 D ate .
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G2 h PULSE    OPERATION            (SUBCRITICAL)
G2NERAL Pulses above S2.00 must be approved by the RFD (prior to pulse initiation). Spec-ification on the RUR may be used to meet this requirement.
SPECIFIC
: 1. Set the alarm points on R-1 and R 5 (criticality monitor) to full scale.
: 2. Given a core position, set the transient rod at a position corresponding to the dollar value determined by the following equation:
S Value = Total worth (S) Transient rod - Desired pulse (S) Wlue
: 3. Bring the reactor cold critical using the three standard control rods, use a rod configuration commensurate with core position and experimental requirements.
Note: A series of repetitive pulses may be fired using the same rod positions on the same day as long as the reactor power is not increasing and is less than 1 kW.
: 4. Stabilize in the manual mode.
: 5. Select the proper pulse detector according to the table below, If the Cerenkov detector is selected, turn off the reactor room and tank' lights, Detector 1 = Pulse Ion      (Maximum insertion = S2.00)
;                    Detector 2 = Cerenkov          (Maximum insertion = S4.00)
: 6. Scram the Transient rod.
: 7. Raise the Transient rod anvil to 100%.
: 8. Let the power decay to approximately 1 watt or less.
: 9. Enter Pulse Mode and enter an identifying string at the prompt.
: 10. Fire the pulse by depressing the " Fire" button on the reactor console.
  - 11. Record the appropriate data in the reactor operations logbook from the pulse .                  1 display.
: 12. Reset R-1 and R-5 to their normal alarm points when pulsing, square wave, or steady state operations are complete.
  . Revised: 15 May 91 Page1
                                                                                          .,w--t    -c
 
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OPERATIONAL ' PROCEDUR!! g.                                                    . Procedure Si TAB H This procedure has been approved by the Reactor Facility Director
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Reviewed by the Reactor Staff
                                          .                        l 170k4/              '/ 29-9/
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                                . st/tu._            2744/7/
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Date Reviewed by RRFSC 24 SEP1991 Date h              bb                      hkkkkhhhh[hhhhhh![U
 
                        ?!$ M *                                                                                                                                  $1h WEEKLY OPERATIONAL INSTRUMENT CHECKLIST CllECKLIST #                                                      DATE, SUPERVI5ED BY ASSISTED DY ,_                                                    REVIEWED DY                                                  _._
: l. WATER LEVEL INDICATOR I
l A, in pool, cast sidc. deptess float on watc' level indicator...................... .......
l B . O b ser v e ser a m o n co n so le . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .! . . . . . . . . .
: 11. WATER RESISTIVITY                                                                                                              <
List resistivity readings for previous week from daily startup checklists. Deterrnine the average at each point is > 0.5 Mohm cm.
MON            TUE                      WED          THU                  FRI                                  AVO Monitor Box DM1                        -                                  ~                                                              ~
DM2                                                                                                                                                      '
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111. RADIATION ALARMS A. Test alarm functions for high level and failure Monitor            Failure alarm functional                          HIGH Level alarm functional R1                                                                            __
R2 R 5 (criticality)
E3 E6 Reactor Room CAM Oas Stack Monitor
                                                                                                                        ~
B . R e set ala r m s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .
IV. OTH ER A. Top lock key seals at Security Desk and at LOG verified intact...............
B. Ch a n ge Filt er in t h e St aek O a s M onito r .................................................._                                    _
I T
AFRRI Form 66(R)                                      Revised: 15 May 91 i
i
 
OPERATIONAL  PRONDUREi                                          Procedure 8,TABI I
Trhis procedure has been approved by the Reactor Facility Director r
(    ,      blfh A            __
E
                                                                                                                                        /
Reactor Facility Directf/                        Date Reviewed by the Reactor Staff
                                                                                              ,weM          $~ 2.9. 't/
Wriglft                Date
                                                                                            /, G<, , a        T-74 -9 I Date
                                                                                          /idor I r/L w -s F gbacL4a                Date
                                                                                          ,          u.. 8%'lty6/
Spence                'Date 9M $nid4 197/or4/
Laugh 6fyQ              tate 0 V w k nr    "  1 % u Al Da%e
[q'guyenihtthdra ??nar1/
Owens                    Date Date Date Reviewed by RRFSC 24 SEP1991 Date
                                                                $3$$$$$[N((Mh5NE[hfh8NikfE$!D$2iE6ddi,                                            1
                                                                                    ..                    -                                        i
 
t Ot'EP ._.T10_2{AITT,- PROCEDURB.
                                                - -                                                                                        l-~--em.-
                                                                                                                                              .: 4 m 4. Procedure 8. TAD I D AILY OPF.7 ATIONAL SHUTDOWN CIIECKLIST Checklist No.                                                                                                              Date Time Completed ,_                                                                                                          Supervised by Assisted by          ,,
: 1. REACTOR ROOM (Room 3161)
: 1. A 11 r o d d r i v e s D O W N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 2. C a r r la ge ligh t s O F F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .
: 3. D 0 0 r 316 2 S E C U R E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: 4. D o o r 3161 lo c k ed wit h k e y .. .... .. .. . . . . .. . . . . . . . . . . .. . . . . .. .. . . .. . . . . .. . . . . .. . . .. .. . . .
FM
: 11. EQUIPMENT ROOM (Room 3152)
: 1. Distillat10n unit discharge valve CLOSED .....................................
: 2. A ir d rye r O P E R AT I O N A L . .. .. .. .. . . .... . . . . .. .. . . .. . . . . . . . . . . . . . . .. .. .. . . .. .. . . .. . . .
  ,    3. Doors 231, 231 A, 3152 and Roof hatch SECU R ED .....................
111. EQUIPMENT ROOM (Room 2158)
: 1. Prim a ry disch a rge p r essu re (P SI) ................ .............................. .....
L. D emineralizer flow ra tes set to 6 G PM ..........................................                                                                                ,,,,,,
: 3. Visu a l in sp eet to a fo r le a k s ... .. .. .. .. . . .. .. .. . . .. . . .. . . .. .. .. . ... .. . . .. . . .. .. . . . . .. .. ...
: 4. D o o r 215 8 S E C U R E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .              .
IV. PREPARATION AREA
: 1. ER 2 plug door CONTROL LOCKED; D oor closed; and h a ndwh eel PA D LOCK ED ................................
: 2. ER 2 ligh ts ON a nd rh co sta t a t 10% ..............................................
: 3. ER 1 plug door CONTROL LOCKED; D oor closed; an d handwheel PAD LOCKED .............................,.
: 4. E R 1 ligh t s O N a n d r h eo st a t a t 1(, 4 ......... .. .... .... ... . ................ .......
: 5. Visu a 1 in sp eet io n o f a r ea .............. ..... .. .... .... .. .. .... . . . . .. . . .. ... . .. . . .. .. .. . . ..
AFRR1 Form 62(R)                                          Revised: 15 May 91                                                                                              Page1
                                                                                                                                                                                      \
 
V. LOBOY ALARM LobbyaIarmaudtoON........................................................................
L_._...                                                                                                                                                    l 1
VI. REACTOR CONTROL ROOM (P am 3160)
I 1. R e a et o r t a n k ligh t s O F F .. .. ...... .. .. . . . .. . . .. .... .... .. ... . .. .. . . .. . . . . . . . . . . . .
: 2. Con sole ch a r t r ecor d er pen s r aised .................................................                                                        i 3.TVmonitorsOFF...........................................................................
: 4. Console LOCKED, and all required keys returned IoloCkboX......................................................................................
: 5. D iffu ser a n d seco n d a ry pu mp s O F F ................ .............................. .
: 6. Pu :ificat ion a n d p rim a ry p u m ps O N ............................... .. .. ...........
: 7. R eaet or mon t hly u sage su mm a ry complet ed ..................................
: 8. Exposure rcom camera power supply turncd OFF ........................
: 9. R a d l a t t o n m o n it o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
MONITOR                                                  READING          HIGH LEVEL ALARM SETTING (mr/hr)                                  'j
: a. R 1                                                                            20    _
: b. R.?                                                    _.                      N/A
: c. R 3                                                                            N/A
: d. R 5                                                                            20
: c. E 3                                                                            N/A
: f. E 6                                                          ,                N/A
: g. R 6                                                                            N/A s
AFRRI Form 62(R)                                                                            Revised: 15 May 91                                    Page 2
                                                                                                                                                                              )
l
 
            - _ . .. .- _..--_                -    _ ~ - _ _        = . _ . -      .-      -. . _ _ .      .      _.-
OPERATIONAL                  PROCEDtJRE                                                                    Procedure 9
  )
This procedure has been approved by the Reactor Facility Director                                                        I
                                                                                          'I!p Pjj;                        l f
  ;                                        [A                'bt                                    fL9 f9 r l                              Reactor Facility Director g                                            Date i
Reviewed by the Reactor Staff l
f%4vN                f LY* '//
Wrigfit                            Date ett.<,,w                      G-71^)
                                                                        ?tt ,st F sbacka                          Date
                                                  .:fl %              N/f%1l Sjience                      "D a t e M $nda, DG95% Q Lau h6yj                        4 ate
                                              & M1 hwsw            ''
Ynva, A) 14 u'en#~                          Dilte i s h AYC7i Owens                              Date Date Date Reviewed by RRFSC 24 SEP1991 Date M4        =        I          *                                                            *
                                                                                                                    .,j
 
                                                                                                                            *T55Ea '****
0?E?^?2 W. OWO9$ *iL ,                                                        .
REACTOR ROOM SAFETY GENERAL The following safety procedures will be observed while in the reactor room.
SPECIFIC
: 1. lioist Operations: Perform the following before/during any hoist operations:
: a. Inspect any lifting equipment (ropes, cables, etc.) for wear or dam-age prior to use,
: b. rinsure that the hoist has n current load testing (within last 12 months) ,
: c. Ensure areas beneath the hoist are clear of personnel when opera-tions are underway. This is particularly important when using the hatches between several floors.
: d. Each time a load approaching 10,000 pounds is handled, test the brakes by raising the load a few inches, applying the brakes, and check-ing for slippage.
: c. Ensure a load is not lowered below the point where two full wraps of cable remain on the drutn.
: f. Ensure no tools or poles longer than 10 feet are raised vertically in the reactor room while the power rails are energized.
: 2. Mercury or mercury compounds in any form are not allowed in the reactor room at any time.
Revised: 15 May 91                                                                                Page 1
 
                                  .. v+
LO - ckDifR ad., ~
OPERXTIWA,a,s..pk6.a.v.ws_d?jW                                          C,.en JE~IOV.w;FM:W%aL,c
                                                                                                                    ,-        aman s Pro,cedure e          10 j                                                                                                                          i
.                          iThis procedure has been approved by the Reactor Facility Director
                                                                    '[A                  Ia                          fMllidli61[L9
                                                                                                                    ;  aua u m Date          1/
Rea or Facility DirectoPa
                                                        -  A                    A 8A'                                      2 9' h e 9f Chair 6an, Sfafety'anii fferth Department ' Date Reviewed by the Reactor Staff 1:wL]f                            T. Ls9/
Wright'                                    Date Ilb,,,a                              C' 24 -91 d(                                        k?9          ,, a f sbackh                          - D at'e Ee                                21f%4f fpenJe                                'Date 9HJa d<                            MA f/
Laugher                                    Diite
[ Q1 ion" + MhtvAl yen V                              Da(e L kl.n                        NOhaf /
OWens                                        Date Date Date Reviewed by RRFSC 24 SEP1991 l
Date r, y .
                                                                                                                                                ,gggg 74:etg4g
_ . _ . .            _ . , . . _        _    ~ . .        _ . _ _ . - ,                    . _ _
 
                                                                              ..c. .a 3 STACK GAS MONITOR PROCEDURE GENERAL This procedure specifies all the requirements for operation of the Stack Oas Monitor (SOM)in the reactor room. This instrument is used to sample and measure the gaseous effluent in the building exhaust system.
SPECIFIC A quality assurance check (QA) is performed daily, prior to reactor operations, as part of the reactor start.up. This check is performed in the following manner:
: 1. The particulate filter is changed if necessary.
2 The front cover of the detector shield is removed and the check source is inserted all the way in to the face of the detector. The blue alert light should come on as the count rate rises above the alert setpoint.1 he red high level alert light and bell should come on as the count rate rises above the high level set point. The audible alarm can be silenced by pushing the rec button on the front of the SOM cabinet.
: 3. The detector voltage system set point is checked by pressing menu function 5 and then menu function 1. The high voltage set point is displayed under itern 7 on the display screen. Press 0 twice to return to the main menu.
: 4. The air sampling flow rate (should be greater than 3.5 cubic feet per minute).
: 5. The counts per minute reading shown on the display should be checked against the plot of counts per minute versus Julian date to determine if it falls within the plus or minus 5% deviation lines for the detector and check source. Ifit does, the cheek source should then be removed and the detector cover replaced.
if the counts per minute consistently fall outside the +/ 5% window, it is considered an abnormality and should be reported immediately to the Reactor Facility Director and to the Safety and Health Department.
: 6. The SGM alarms will be acknowledged by pushing the "ACK" button on the SGM keyboard.
t Revised: 15 May 91                                                                Page1
 
1                    ortxArio.  -o  NAG.o_m.enoc.._rp0d@_.!.";f T > , %+ w' 1^~T#'~"!.                  o w h . ~ 2M      !W x 2. M . ,s f 1i
_:rioc.a                ,
,                    This procedure has been approved by the Reactor Facility Director
[_ 17 .                  Jh in y ,$/'2.9 [77 Reac r / Facility Direct &                      . JUS    U J=U        Date ikA$
                                                      'hairian, Safet'y a'nd Trealth Department # Date
                                                                                                            **c7 h k u 9/'
Reviewed by the Reactor Staff f ubX4                T l *! 1)
Wright                      Date 4 c<,e                  s t4 9j co                        Date 4 dr e/ 1 / 4 w a ,,
F9 sbdeka                    Date 42>;+rra.-              e#7kA'/
                                                                          ~ 3l 5ence      ,            'D ate M 1.-ric< M//fu4/
L u hfry                    -Date i f.4  h.Mmw4f                                                      ~
6 uyefi ' '                  Date
                                                                          / h AL '26sv t
Gwens                        Date Date Date Reviewed by RRTSc 24 SEP199' Date
 
0(ERADON.:RPf5e M iE2 W HFiIIE 2 E?E N B E hr*F AIR    PARTICUL YE MONITOR                    PROCEDURE GENERAL This procedure specifies how to test the CAM to ensure proper operation of this monitoring device.
SPECIFIC This procedure uses a radioactive source to test the alarm set points of the CAM,
: 1. OPERATING and TRACING Observe to see that the CAM is operating and tracing.
: 2. ALARM TEST WITH SOURCE Open the detector chamber door and slowly bring a radioactive source near the detector. Observe the meter on the fron' of the CAM. The yellow light will oc come on at approximately 4,000 counts per minute. The red light will come on at approximately 10,000 counts per minute, the abrm will sound and the dampers will close. Reset the alarrr., close the chamber door and replace the source in the drawer.
Revised: 15 May 91                                                                Page1
 
ATTACHMENT C                                -
Amendment No. 21 to Facility Operating License
                              -- _ - - _ _  u
 
f * %,
                  /                        o,
              !                      / i                                                    UN*TED STATES
              !)                  ,.. ,4( [3                    NUCLEAR RELLLATORY COMMISSION
                *, 84                                                                  WA$HINGTON. D C. 2tF66 ARMEDFORCESRAD10B10LOGYRES{ARCHINSTITUTE DOCKET NO. 50-170 AMENDMENT TO FACILITY OPERATING LICENSE
* Amendment No. 21 License No. R-84
: 1.        The U.S. Nuclear Regulatory Comission (the Comission) has found that:
A.      The application for amendment to facility Operating License No. R-84 filed by the Armed Forces Radiobiology Research Institute (the licensee), of April 30, 1990, as supplemented on December 17, 1990, March 5, 1991, May 17, 1991,' August 16, 1991, and September 10, 1991, complies with the standards and re Energy Act of 1954, t.s amended (the Act)quirements                                              of the Atomic
                                                                                                                              , and the Consnission's        rules and regulations as set forth in 10 CFR Chapter I; B.      The facility will o)erate in conformity with the application, the provisions of t1e Act, and the rules and regulations of the Consnission; C.      There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comunission's regulations; D.      The issuance of this amendment will not be inimical to the coninon defense and security or to the health and safety of the public; E.      The issuance of this amendment is in accordance with 10 CFR Part 51 of the Consnission's regulations, and all applicable requirements have been satisfied; and F.      Prior notice of this amendment was not required by 10 CFR 2.105(a)(4),
and publication of notice for this amendment is not required by 10CFR2.106(a)(2).
: 2.            Accordingly, paragraph 2.C.(1) of Facility Operating License No. R-84 is hereby amended to read as follows:-
(1) NaximumPowerLevM AFRRImayoperatethereactoratsteady)statepowerlevelsup to a maximum of 1100 kilowatts-(thermal , and at pulse power levels not to exceed a pulse reactivity insertion of 4.00 dollars.
              ~                _        _              _ , _ . _ _ _ _ _ . . _ _ -
 
2-
: 3. Accordingly, the license is amended by changet. to the Technical Specifications as indicated in the enclosure to this license aniendment, and paragraph 2.C.(2) of License No. R-84 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications co.itained in Appendix A, as revised through Amendment No. 21. are he eby incorporated in the license.
The licensee shall operate the fa;ility in accordance with the Technical Specifications.
4    This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas  . Nur  , Director Office of Nuclear Reactor Regulation
 
==Enclosure:==
 
Appendix A Technical Specifications Changes Date of Issuance:    octot>er 8,1991
 
  ._._ _ __-                            ___r                      _ . . _ . . _ _ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ - _ _.
ENCLOSURE TO LICENSE AMENDMENT NO. 21 FAClll1Y OPERATING l! CENSE NO. R 84 DOCKET NO. 50-170 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
1 Remove                                                                                                    insert i                                                                                                          i 2                                                                                                        2 7                                                                                                        7 8                                                                                                        8 22                                                                                                        22                                                      ;
,                                              23                                                                                                        23                                                      ;
25                                                                                                        25 26                                                                                                        26 27                                                                                                        27 28                                                                                                        28 29                                                                                                        29 30                                                                                                        30                                                      ,
31                                                                                                        31 32                                                                                                        32-34                                                                                                          34 35                                                                                                          35-f 9
v e-e--= --r  --e-..--a,- r-, ww .a-v,v  *w,  --,re,--,v,v-          .-y,,,-    v.,my-  ,rw-,-w,>,v,,,w , y-w -    ,w  , , ~,v,,-,vwry~,,eiew    r-,o  w gv ,mvr -c' g me~- Sw=t- g c, , %yg,'s,--: w
 
TECllNIC AL SPECirtC.4TIONS Foil Tile AfitRI ItEACTOR FACILITY LICENSE NO R 84 DOCKET #50170 TABLE OF CONTENTS M
i0  DErlNITIONS 1.1    ALARA                                                                    1 1.2    Channel Calibration                                                      1 1.3    Channel Check                                                            i 1.4    Channel Test                                                            i 1.5    Cold Critical                                                            1                  '
l.6    Core Grid Position                                                      1 1.7    Experiment                                                              i 1.s    Experimental Facilities                                                  1-1.9    Fuel Element                                                            2 1.10  instrumented Element.                                                    2_
1.11  Limiting Safety System Setting                                          2 l          1.12  Measured Value                                                          2
!          1.13-  Measuring Channel                                                        2 l          1.14  On Call                                                                  2 Operable 1.15                                                                            2 1.16    Pulse Mode                                                              3 1.17    Reactor Operation                                                        3 1.18    Remeter Safety Systems                                                  3 1.19    Reactor Secuted                                                          3 1.20    Reactor Shutdown                                                      'S 1.21    Reportable Occurrence                                                    3 1.22    Safety Channel                                                            4 1.23  So.fety Limit                                                            4 1.24  Shutdown Margin                                                          4 1.25  Standard Control Rod                                                      4l 1.26  Steady State Mode                                                        4 1.27  Transient Rod                                                            4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS                                    3 i
2.1    Safety Limit . Fuel Element Temperature                                  5              i 2.2  . Limiting Safety System Setting for Fuel Temperature                      6        - - -
3.0 LIMITING CONDITIONS FOR OPERATIONS                                                    7 3.1    Reactor Core Parameters                                                  -- T .
3.1,1 Steady State Operation                                              7 3.1.2 Pulse Mode Operation                                                7
                  - 3.1.3 Reactivity Limitations                                              6 3.1.4 Scram Time                                                          8 1
Amendment NJ. 21
: e. Henetor Pool
: d. Core Experiment Tube
: e. Portable Detm Tubu
: f. Pneumatic Transfer System
      .. Incore Locations 15    117,L EL Ehf ENT A fuel element is a single TRIG A fuel rod, et the fuel portion of a fuel follower controi rod.          l t 10 INSTRUMENTED ELEMENT An instrumented element is a special fuel element in which sheathed chromal/alumel or equivalent thermocouples ue embedded in the fuel.
1.11 LIMITING S AITTY SYSTEM FPTTING limiting safety system settings are settings for automatic protective devices related to those variables having significans safety functiott 1.12 MEASURED VALUE A meuvred value is the magnitude of a variable u it appears on the output of a aneuuring channel.
1,13 MEASURING CllANNEL A measuring channel is that stombinatien of sensor, interconnecting cabla or linn, amplifiers, and output device that ue connected for ti.e purpoet of measu*ing the value of a variable.
1.14 E CALL A person is considered on call if
: a. The individual hu been specifically designated and the operator knows of the designationi
: b. The individual keeps the operator poeted as to his/La whernabouts and telephone number; and
: c. The individual is capable of getting to the reactor facility within 80 minuta under normal circumstantw.
1.15 OPERABLE A system channel. device, or component shall be considered operable when it is capable of performing-its intended function (s) in a normal manner.
3 Amendment No. 21 1      .
                                                                                                                    .1
 
30 LIMITING CONDITIONS FOR OPEft ATIONs 3: ItE ACTOH CORE PAR AMETEltS 3.t.1 fTEADY STATE APEH ATION A tohe abilit y This specification applies to the manimum reacter power attalned during steady state operation.
ON*etive To assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a set point for the high flus litniting safety systerns, so that automatic protective action will prevent the safety limit from being reached i
during steady state operations, Erecificatleris l                    The remetor steady state power level shell r.
* exceed 1.1 rnegawatte. The normal i
steady state operating power limit of the reactor should be 1.0 megawatt. For                l l                    purposes of testing and calibration, the rtactor may b, operated at power levels not to exteed 1.1 megawatts during the testing period.
buit Thermal and hydraulic calculationr and eterational experience indicate that TRIG A fuel may be safely operated up to power levels of at leut 1.5 megawatts with natural convective cooling.
3.1.2  PULSE MODE OPERATION Arrhes.kjjity.
This specification applies to the muimum thermal energy produced in the teactor u a result of a prompt critical insertinn of reactivity.
Obieetive The objective is to usure that the fuel temperature safety limit will not be exceeded.
Specification The maximum step insertion of reactivity shall be 2.8% Ak/k (84.00) in the pulse mode.
Dbtil Based upon the Fuchs-Nordheim rr athematical model (cited by C.E. ClifTord et al. in the April Reador"),          1961 GA an insertion    ReF.
of 2.8,  rt #2119, Ak/k          "Model results in      of theaverags a maximum    AFRRI TRIGAfuel temperature of less than 550'C, thereby stayin,g within the limiting safety settings that protect the safety limit. The 50 C margin to the Limiting Safety 7
Amendment No. 21
 
System Betting and the 450'C margin to the safety limit amply allow for uncertainties due to extrapolation of meputed data. accuracy of meuured data, and location of instrumented fuel elements in the core.                                                              l
:                                      3.1.3          REACTIVITY 1.lMITATIONS 1
A r+1ie nbilit y 4
These specifications apply to the reactivity condition of the reactor and the                                            i
]
reactivity worths of control rods and experiments. They apply for all modes of                                        l  l operati)n.                                                                                                              >
Obiective e                                                    The objective is to guarantee that the reactor can be shut down at all timo                                            '
and that the fuel temperature safety limit will not be exceeded.
St-iMettlug
                                                                                                                                                                            ?
: a.          The reactor shall not be operated with the mutmum available excus remetivity sbove cold critical with or without all experiments in place greater than 86.00 (s.s% Ah/k),
: b.          The minimum shutdown margin provided by the remaining control rode with the most reactive control rod fully withdrawn or removed shall be 80.60 (0.365 Ak/h) for any condition of operation.
l 11A111
: a.          The limit on available excess reactivity utablishn the maximum power if all control elements are removed.
: b.          The shutdown margin usurn that the reactor (an be shut down from any operating condition even if the highut worth control rod remains in the                                    i fully withdrawn position or is completely removed.
3.1.4        SCRAM TIME Aeollembility The specification applit to the time required _to fully inert any control rod to a-full down position from a full up position.                                                                              ,
Oblective                                                                                                                >
The objective is to achieve rapid shutdown of the reutor to prevent fuel damage.
Specification The tisne from scram initiation to the full insertion of any control. rod from a full up position shall be less than I second.
f                                                                                                                                                                            k 8
Amendment No. 21 L.          ,_.u__                        _        _ _ . _ . _ . _ , _ _ _ , .                      _ _ . _ _ . _ . _ _ . _ _
 
E,kffjfit ation Functional checks shall be made annually, but not to exceed 15 months, to insure the following:
: a. With the lead shield doors epen, neither exposure room plus door can be electrically opened,
: b. The core dolly cannot be enoved into position 2 with the lead shidd doors cloud e    The wuning horn $all sound in the exposure room before opening the lead shield door, which allows the core to move to that exposure room unless cleued by two licensed operators.
U.hNL These functional checks will verify operation of the interlock system. Experience at AFRRI indicatn that this is adequate to insure operability.
4.2.5  REACTOR FUEL ELEMENTS Acelleability This specincation applies to the surveillance requiremerits for the fuel elements.
Obiettive The objnlive is to verify the integrity of the fuel element cladding.
Soetifications All the fuel elements prennt in the reactor core, to include fuel follower control rods, shall be insputed for dart. age or deterioration, asi meuured for length and bow at intervals sepusted by not more than 500 pulsa of inurtion greater than
            $2.00 or annually (not to exceed 15 months), whichever occur .nret . Fuel elements in long term storage need not be measured until returned to core; however, fuel elements routinely moved to temporary storage shall be meuvred every 500 pulses of insertion greater than 82.00 or annually (not to exceed 15 months), whichever occurs fint.
Elst The frequency of inspection and meuvrement is bued on the parameters most likely to affect the fuel cladding of a pulse reactor, and the utilisation of fuel  l elements whose characteristics are well known.
The limit of transverse bend has been ebown to rwult in no dimculty in disusembling the core. Analysi4 of a woret case scenario in which two adjacent fuel elements suffer sufficiently severe tranevme bends to twult in the touching of the fuel elements has shown that no darnage to the fuel elements will roult via a hot spot or any other known mechanism, 22 Amendment No. 21
 
4.3 COOL ANT FYSTEMS Attlitatilitz This spec:ication applies to the surveillance requirements for monitoring the pool water and the water conditioning system.
Objective The objective is to usure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating possible radiation huards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other component, in the primary water system.
S t.*eific ations
: a.        The pool water temperature, u meuurid near the input to the water purification system, shall be meuured daily, whenever operations are planned.
: b.        The conductivity of the water at the output of the purification system shall be meuvred weekly, whenever operations are planned.
EAdi Dued on experience, observation at then intervals provides acceptable surveillance of      l limits that usure that fuel clad corrosion and neutron activation of dipolved materials will not occur.
4.4  VENTIL ATION SYSTEM Aeolienbility This specification applies to the facility ventilation system isolation.
Obleetive The objective is to usure the proper operation of the ventilation system in controlling the releue of radioactive material into the unrestricted environment.
Eggtification The operating mechanism of the positive naling dampees in the reactor room ventilation system shall be verified to be operable and visustly inspected at leut monthly (interval.
not to exceed six weeks).
Datig Experience accumulated over years of operation hu demonstrated that the tests of the ventilation system on a monthly basis are sufficient to assure proper operation of the system and control of the releue of radioactive material.
23 Amendment No. 21-l
 
50 DESIGN FEATUllCS 5i SITE AND FACILITY DEscitII' TION Acclieability This specification applies to the building that houses the reactor.
Obi etive The objective is to restrict the amount of radioactivity releued into the environment.
So,eifiestions
: a.      The reactor building, u a structurally independent building in the AFRill complex, shall have its own ventilation system branch. The effluent from the reactor ventilation system shall exhaust through absolute filters to a stack having a minimum elevation that is 18 fut above the roof level of the highest building in the AFRRI complex.
: b.      The reactor room shall contain a minimum free volume of 22,000 cubic feet,
: c.      The ventilating system air ducts to the reactor room shall be equipped with positive sealing dampers that are activated by fail safe controls, which will-automatically close off ventilation to the reactor room upon a signal from the reactor room air particulate monitor.
: d.      The reactor room shall be designed to rwtrict s.it leakage when the positive sealing dampers are closed.
EElill The facility is designed so that the ventilation system will normally maintain a negative pressure with respect to the atmosphere, so that there will be no uncontrolled leakage to the ervironment. The free air volume within the resctor building is confined when there
!        is an emergency shutdown of the ventilation system. Building construction and gukets around doorways help rwtrict leakage of air into or out of the reactor oom. The stack height insures an adequate dilution of effluents well above ground level. The separate ventilation system branch lasures a dedicated air flow system for reactor effluents.
* 5.2 BEACTOR CORE AND FUEL 5.2.1    REACTOR FUEL Arcticability These specifications apply to the fuel elements, to include fuel follower control rods, used in the reactor core.
Obiective The objectiva are to (1) usure that the fuel elements are dwisned and fabricated in such a manner u to permit their use with a high degree of reliability assure thatwith    twpect the fuel    to their elements usedphyelcal in the coreand  nuclear characteristics, are substantially those analyse and (2) d in the Safety Analysis Report.
Soeeifications The individual nonirradiated standard TRIG A fuel elements shall have the following characteristics:
25 Amendment No. 21
 
i n.
Uranium content: Muimum of 9.0 weight percent enriched to leu than 20% uranium.386. In the fuel follower, the maximum uranium content will be 12.0 welght percent enriched to less than 20% uranium.235, b,
llydrogen.to elttordum atom ratio (in the Zril ):
1.0 Zr atoms with a range between 1.6 and Ih. Nominal 1.7 li atoms to
: c.        Clatiding: 604 etainlese steel, nominal 0.020 inch thick.
d.
Any buenable poleon und for the specific purpose of compensating for fuel burnup or lont. term teactivity adjustmente shall be an integral put of the manufactured fuel elements.
Undt A muimum uranlurn content of 9 weight percent in a standard TRIGA element le greater than the design value of 8.6 weight percent, and encompuses the maximum probable variation in individual elemente. Such an increue in bding would result in an increase in power density of len than 6%. An increue in local power density of 6% in an-individual fuel element reduces the safety mugin by 10%, at most. The hydrogen.to elrconium ratio of 1.7 will produce a rnaalmum prenure within the cladding well below the rupture strength of the cladding.
The local powe, denalty of a 12 0 weight percent fuel follower is 21% steater -
than an 8.5 weight percent standud Tit!GA fuel element in the D Ring. The volumefuel TRIGA            of element.
fuel in a fuel followed rod is $6% of the volume of a standud Therefon the actual power produced-in the fuel followed rod D-ring.
the      is 83% less than the power , produced in a standed TRIGA fuel element in 6.2.2 REACTOR CORE Adelienbility These specifications apply to the configuration of fuel and in core experimente.
Obiective The objective le to restrict the arrangement of fuel elemente and experimente so as to provide moeurance that excessive power densities will not be produced.
Soecifications a.
The reactor core shall consist of standard TRIGA reactor fuel elemente in a close packed array _ and a minimum of two thermocouple instrumented TRIG A reactor fuel elemente.
              -b.
Ther, shall be four single core positions occupied by the three standud control rode and transient rod, a neutron start up source with holder, and positions for poselble in core experimente, e.
The core shall be cooled by natural convection water bw, s
d.
In core experimente shall not be placed in adjacent fuel positions of the B-ring and/or C ring.
26
: Amendment-No. 21.
I
 
4
: e.            Fuel elemente indicating e.n elonsstion great:r thin 0.100 inch, a lateral bending treater than 0.0625 inch, or significant visible damage shall be considered damaged, and shall not be und in the reactor core.
Datia Standard TRIG A cores have bun in un for years, and thelt safe operational characteristice are well documented. Experience with TRIGA reactore hu shown that fuel element bowing that could result in touching hu ouvrred without deleterious effects. The elongation limit hu been specified to (a) usure that the cladding matnial will not be subjected to strence that could caun a non of integrity in the fuel containment, and (b) nasure adequate coolant flow.
5.2.3 CONTROf RODS Arctic ability These specifications apply to the control rode und in the reactor core.
Obiettive The objective is to usure that the control rode are dulgned to pumit their un with a high degree of reliability with respect to their physical and nuclear characteristice.
Specific ations
: a.                  The standard control rode shall have scram capability, and shall contain borated graphite, B4C powder, or boron and its compounde in solid form as a poison in aluminum or stainless eteel cladding. Thue rode moy have an aluminum, sir, or fuel follower. If fuel followed, the fuel region will conform to the Specifications of 5.2.1.
: b.                  The transient control rod shall have uram capability, and shall contain borated graphite, B C powder, or boron and its compounde in polid form u a poleon in alumindm or stainlue stul cladding. This rod may incorporate an aluminum, poison, or air follower.
Eatig The poleon requirements for the control rods are estisfied by using neutron.
absorbing borated graphite, B C powder, or boron and its compounds. Thue materials muel be contained i,n a evitable cladding material, such u aluminum or                          i etainless steel, to insure mechanical etability during movement and to leolate the                        '
poison from the pool water environment. Scram capabilities are provided by the rapid insertion of the control rode, which is the primary operational omfety feature of the reactor. The transient control rod is designed for use in a puleing TRIGA reactor.
27                                                ,
Amendment No. 21 1
 
6.3 S'ECI AL Nt.' CLEAR MATERI AL STOR AGE Mdicat1hty This specification applies to the storage of reactor fuel at times when it is not in the remeter core.
Obieetive The objective is to usure that stored fuel will not become critical and will not reach an Unskfe temperature.
St et(fit tti9n All fuel elemente not in the teactor core shall be stored and handled in accordance with appliemble regulations. Inadiated fuel elements and fueled devices shall be stored in an array that will permit sufficient natural convective cooling by water or air, so that the fuel element or fueled device temperature will not exceed design values. Storage shall be such that groups of stored fuel elements will remain suberitical under all conditions of moderation.
        !!alig The limits imposed by this specification are conservative and usure safe storage and                          l handling. Experience shows that approximately 67 fuel ele:nents are required, of the design used at AFRRI, in a closely packed array to achieve criticality. Caicuistions show that in the event of a full storage rack failure with all 12 elements falling in the most reactive nucleonic configuration, the mus would be less than that required for criticality.
Therefore, under normal storage conditions, criticality es,nnot be reached, t
28 1
Amendment No. 21
 
eo M)MINISTR ATIVE CONTROLS e! QRG AN17 ATIOS 6 1.1      ST R t'CTt'R E The organisation of personnel for the management and operation of the AFRRI reactor facility is shown in Figure 1. Organtiation changes may occur, bued on Institute requireniente, and they will be depicted on internal documents. However, no changes may be made in the Operation, Safety, and Emergency Control Chain in which the Reactor f acility Director hu direct responsibility to the                              l Director, AFRRI.
thrector. AFRRI g, ,              4% ,,,        Ore'aumal. Safet). Reactor and Radiauen Factluy Safety Commutee Safety and Health Expt.
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* Ra&auon                                              i Sources Dept.                                        {
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Advm                                                      Abitv) l l                                                          l Rextor Factlary Dusetor        ,,,,,,,,,,,,,,j l
Reactor Opersoon Supervisod I
Reactor Operanons Sta!P rine. t. o ri..i..ii.. .: r er.....i v.e u...e....          ..a o,. . i.. .e 36. A rax: n..si., r..iiit y.
A.y e..et.t st.ff m.mb.e bu useos 4. ch. IMr.et.: f.e m .46et. .f e.foty.
29 Amendment No. 21 3
 
e.l.2            EIMPONSIBILITY The Director, facility.          The ReactorAFRHi,Facility shat ave Directorlicense resp)asibility (RFu              for the reactor shall be responsible  for administration and operation of the Reactor Facility and for determination of applicabihty of procedures, experiment authorientions, maintenance, and operations. The RfD may designate an individual who meets the requirements of Section 6.1.3.1.a to discharge the RFD's responsibilities in the RFD's absence. During brief absences (periods less than four hours) of the Reactor Facility Director and his designee, the Remetor Operations Supervisor shall discharge these responsibilities.
6.1.3              STAFFING 6.1.3.1 Selection of Personnel
: a.          Reactor Facility Director                                                                  l At the time of appointment to this position, the Beactor Facility Director shall have 6 or more years of nuclear experience liigher education in a ecientific                nuclear engineering field may fulfill up to 4 years of experience on a onedor one basis. The Facility Director must have held a USNRC Senior Reactor Oprator license on the AFRRI reactor for at least 1 year before appointment to this position.                                ;
: b.          Reactor Operations Supervisor (ROS)
At the time of appointment to this position, the ROS shall have 3                        l years nuclear experience. Illgher education in a science or nuclear engineering field may fulfill up to 2 years of experience on a one-for one bula. The ROS shall hold a USNRC Senior Remeter Operator license on the AFRRI reactor. In addition the ROS shall have 1 year of experience as a USNRC licensed Senior kenetor Operator at AFRRI or at a similar facility before the appointment to this position.
: c.          Reactor Operators / Senior Reactor Operators At the time of appointment tc this positlen, an individual shall have a high school diploma or equivalent, and                  . " possess the appropriate USNRC license.                                                                                ,
: d.          Additional staff u required for support and training. At the time of appointment to the reactor staff, an individual shall poseus a high school diploma or equivalent.
6.1.3.2 Ooerations
: a.          Minimum staff when the reactor is not secured shall include:                                    ,
: 1.          A licensed Senior Reactor Operator (SRO) on call but' not necessarily on site
: 2.          Radiation control technician on call
: 3.        - At leut one licensed Reactor Operator (RO) or Senior. Reactor Operator (SRO) present in the control room
: 4.          Another person within the AFRRI complex who la able to carry out written emergency procedures, instructions of the operator, or to summon h(lp in case the operator becomes incapacitated 30 Amendment No. 21
_ _ _ _ . _ _ _ _ _ _ . .                                                                            .    ..          _ _ _ . _                  _        _.s.. __ ..  ~. u _
: b.      Maintenance actisiues that could affect the reactivity of the reactor                                                      .
shall be accomplished under the supervision of an SRO.
: c.      A list of the names and telephone numben of the following personnel                                                        i shall be readily available to the operator on dutyi
: 1. Management personnel (Reactor Facility Director, AFRR1 Director)
: 2. Radiation safety personnel (Head, Safety and Health Department)                                                  l;
: 8. Other operatione personnel (Reactor Staff ROS) 6.1.4      TRAINING OF PERSONNEL A training ana retraining program will be maintained, to insure adequate levels of proficiency in persons involved in the reactor and reactor operations.                                                                  >
6,2 RgviEW AND AUDIT . Tile REACTOR AND R ADIATION FACILITY SAFETY                                                                                        '
COMMITTEE fREFSCI 6.2.1      COMPOSITION AND QUALIFICATIONS 6.2.1.1 Ccmoosition
: a.      Regular RRFSC Members (Permanent Memben)
(1) The following shall be members of the RRFSC.                                                                            i (a) Chairman, Safety and Health Department, AFRR1                                                                  l (b) Reactor Facility Director, AFRR1 (2) The following shall be appointed to the RRFSC by the Director, AFRRI:
(a) Chairman se appoln'ed by the AFRRI Directorat.,
i                                                                          (b) One to three non AFRRI members who we knowledgeable in fields related to reactor safety. At least one shall be a Reactor Operations Specialist, or a Health Physics Spetlalist,
: b.      Spelai RRFSC Memben (Temporary Memben)
(1) Other knowledgeable pereone to serve as alternates in item a(2)(b) above u appointed by the AFRRI Director.
l l                                                                  (2) Voting ad ja memben, invited by the Director of AFRRI, to anoist in review of a particular problem.
: c.      Nonvoting memben as invited by the Chairman, RRFSC.
31 Amendment No.=21
  - _ __                                    - . - . -        __ _                                                      --.m.-                        __ . _ _-          _ . . _ _ . _ . .
 
I 6.2.12 Quahfit11}Qat The minimum qualifications for a person on the RRFSC shall be 6 yeus of profeusonal experience in the discipline ce specific field represented. A baccalaureate degree may fulfill 4 years of experierce.
62.2      FUNCTION AND AUT110RITY 6 2 2 i Funetlen The Reactor and Radiation Facility Safety Committn is directly responsible to the Director, Afiliti, The committee shall review all radiological health and safety matters concerning the reactor and its usociated equipmtnt, the structural reactor facility, and thon items listed in Section 6.2.4.
6.2 2 2 Authority The RitFSC shall report to the Director, AFRill, and shall advise the Reactor Facility Director in those ueu of responsibility epecified in Section 6.2.4.
6.23      CIIARTER AND RULES 6 2.3.1 Alternates Alternate members may be appointed in writing by the RRFSC Chairman to serve on a temporary buis. No more than two alternatn shall participate on a vollng buis in itRFSC activities at any one time.
6 2.3.2 Meettnr Freauency, The RitFSC or a subcoramittu thereof shall meet at leut four timu a calender year. The full RRFSC shall mut at leut semi annually.
l 6.2,3.3 Quorum A quorum of the RRFSC for review shall consist of the Chairman (or designated alternate) and two other members (or alternate members), one of which must be a non AFRRI member. A majority of thon prennt shall be regular members.
6.2.8.4 Votine Rules Each regular RRFSC member shall have one vote. Each spe<lal appointed member shall have one vote. The majority is 61% or more of the regular and special members pruent a,nd voting.
6.2.3.6 hiinuttu Minutes of the previous meeting shall be available to regular members at leut I week before a regular scheduled meeting.
32 Amendment No. 21
: f. Any other area of Facility operations considered appropriate by the RRFSC or the Director /AFRRI.
: g. Reactor Facility AL ARA Program. This program may be a seet n of the total AFRRI program.
6s PNoCEDURES 6A 1  Written instructions for certain activities shall h approved by the Reactor Facility Director and reviewed tr the R:actnr and Radiation Facility Safety Committee (RRFSC) The pre:: lures shall be adequate to assure sJe operation of the reactor, but shall not peelude the use of independent judgment an I attion as deemed necessary. These activities are as follows:
l
: a. Conduct of irradiations and experiments that could affect the operation and safety of the r:.ctor.
: b. Reactor start. training program,
: c. Surveillance, testing, and calibration of instruments, components, and systems involving nucless safety.
: d. Personnel radiation protection consistent with 10 CFR 20.
: e. Implementatica of required plans such as the Se urity Plan and Emergency Plan.
: f. Resetor core loading and unloading.
: g. Checkout startup, standard operations, and securing facility.
6.3.2 Although substantive changes to the above procedures shall be .nade only with approval by the Reactor Facility Director, temporary changes to the procedures that do not change their original intent may be made by the ROS. All such temporary changes shall be documented and subsequently reviewed end approved by the Reactor Facility Director.
6.4 REVIEW AND APPROVAL OF EXPERIMENTS 6.4.1 Before issuance of a reactor authorisation, new experiments shall be reviewed for radiological safety and approved by the following:
: c. Reactor Facility Director
: b. Safety and Health Department
: c. Reaceor and Radiation Facility Safetv Committee (RRFSC) 6 4.2 Prior to its performance, an experiment shall be included under one of the following types of authorisations:
: a. Soecial Remeter Antheriatlan for new experiments or experimente not included in a Routine Acactor Authorisation. These experimerts shall be performed under the direct supervision of the Reactor Facility Director or designee.
: b. Routine Reactor Authorisation for experiments safely performed at least once. These experiments may be performed at the discretion of the Reactor Facility Director and coordinated with the Safety and Health Department    l when appropriate. These nothorisations do not require additional RRFSC review.
34                                                ,
1 i
Amendment No. 21 1
1 i
i
 
c.
hjp? Parameters Authorisation for routine measurements of reactor parameters, routine core measurements, instrumentation and calibration                    E cheth, maintenance, operator training, tours, testing to verify reactor outputs, b-other reactor testing procedures. This shall constitute a single authorisattoa. Then operations may be performed or der the authorisation of the Reactor Facility Director or the Resetor Operat . sne Supervisor.
6.4.$
Substantive (reactivity worth more than 280.25) change to previously approved experiments shall be made only after review by the RRFSC and after approval (in writing) by the Reactor Facility Director or designated alternate.- Minor changes thst do not significantly alter the experiment (reactivity worth of Ins than 180.25) may be approved by the ROS. Approved esperiments shall be carried out in accordance with utablished procedures.                                          l 6.5 J1EQUIREI; ACTIONS 6.5.1 ACTIONS TO BE TAKEN IN CASE OF SAFETY LIMIT VIOLATION a.
The reactor shall be shut down immediately, and reactor operation shall not be resumed without authorisation by the NRC, b.
The safety limit violation shall be reported to the Director of NRC Region-1, Office of Inspection and Erforcement (or designate); the Director, AFRRl; and the RRFSC not later than the next working day, e,
A Safety Limit Violation Report shall be prepared,. This report shall be reviewed by the MtFSC, and shall dweribe 11) appiitable circumstances l- i preceding the viola ion, (?) effects of the violation on facility components,            '
structures, or systems, and (3) corrective action taken to prevent or reduce the probability of recurrence.
d.
The Safety Limit Viobtion Report shall be submitted to-the NRC; the Director, AFRRl; and the RRFSC within 14 days of the violation.                        l 6.5.2  REPORTABLE OCCURRENCES Reportable occurrenen as defined in 1.21 consequences, corrective actions, and measu(res to prevent recurrence) shall b reported to the NRC. Supplemental reports may be required to fully deseribe the final resolution of the occurrence.
: a. Promot Notification With Written Followan. The types of events listed .
below shall be reported as soon as possible by telephone and confirmed by 35 Amendment No. 21
 
4 y
%M.      ATTACHMENT D 4 v-yp        soutine neactor Authorizations
 
i l
Routine Reactor Authorization #1 September 1991 latteduction: The Reactor and Radiation Facility Safety Committee has reviewed and approved t!,e operations described below. These operations have been performed many times in the past and are now considered part of the routine operations for use of the AFRRI reactor.
Authorintion: As permitted by NRC license R-84, the Reactor Facility Director, Reactor Operations Supervisor, and NRC licensed operators may perform measurements, conduct operations, effect maintenance, perform experiments, conduct tours, and conduct operatoc training within the scope of AFRRI procedures, Technical Specifications, and NRC regulations.
ALARA principles will be followcx! at all times during the conduct of operations.                                            ;
Approved:                        bl>-r2/
Reactor Facility'
                                                                                    /            .
                                                              $NMi        ,, SHD R              ~
l                                                                NM. Manderfield COL /iiSAF, MSC -
Chairman, Reactor and Radiation Facility Safety Committee -
[2cactor        Radiation F ility Safety Committee mw M. L        re                                CDR J.E. DeCicco, MSC, USN Y~                            0            et    A.u
[s. Ashby - /
g.
s CAPT C.B. Galley, fSC, USN A Luersen
    ##h1
: 23. VM
 
i Routine Rcoctor Authorization #2 September 1991 latI9 duction: The Reactor and Radiation Facility Safety Committee has reviewed and approved the dosimetry operations dewribed below. These operations have been safely perfctmed over many years and are classified as routine.
Authunzations: As permitted by NRC license R 84, dosimetry instrumentation and other measuring instrumentation may be used alone, attached to, or included with experiments in the reactor radiation facilities within the site boundary, subject to the limitations imposed by Technical Specifications. All dosimetry devices, (quipment or experiments will be used or performed under the supervision of the RFD or his designee. In particular, the dosimetry devices and the experiment shall:
o      not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table 11 Appendix B, of 10 CFR 20 are exceeded, a      not create inventories of I 131 through 1135 greater than 1.3 curies and Strontium-90 inventories greater than 5 millicuries, o      limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o      be doubly encapsulated if the release of the contained material can cause corrosion to the radiation facility, o      have an absolute worth less than $3.00, and o      have been inspected and approved by a reactor operator prior to performance of the experiment.
The term " dosimetry instrumentation" shall include, but not be limited to: fission chambers, ionization chambers including those with flowing gas, thermoluminesent devices, foils, tablets, and phantoms meeting the limitations listed above.
Such dosimetry instrumentation may also be irradiated separately for the purpose of dosimetric calibration or device evaluation and testing.
ALARA principles will be followed at all times during the desi        and conduct of experiments.
Approved;          h,.
* ODv-Reactor Facility Dtrdptx                    Chairman 'S14D h///$
N.M. Manderficid COL (USAF, MSC Chairman, Reactor and Radiation
:                Facility Safety Committee
 
Recctor and Radiation Facility Safety Committee M.I
___oore
__/_ M OA 33). Ashby            -
Ell $d R. Luttsgn l  Y6 M. Votti ~
CDR J.E. DeCicco, MSC, USN bb CAPT C.B. Galley, M      ,USN
 
Routine Reactor Authorization #3 September 1991 IntrodKuon: The Reactor and Radiation Facility Safety Committee has reviewed and approved the reactor expenment described below. This experiment has been performed on numerous occasions in past years and is considered routine.
Authorintuon5: As permitted by NRC license R-84, The Reactor Facility Director may permit a principal investigator to irradiate animals, animal tissue and other biological materials in the reactor irradiation facilities subject to the limitations imposed by Technical Specifications. This authorization includes shiciding and support materials, sensors, control devices and the use of phantoms and other dosimetry instrumentation authorized under Routine Reactor Authorization #2. In particular, the experiments shall:
o      not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o      not create inventories of I-131 through I-135 greater than 1.3 curies and Strontium-90 inventories greater than 5 millicuries, o      limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o      be doubly encapsulated if the release of the contained tnateriel can cause corrosion to the radiation facility, o      have an absolute worth less than $3.00, and o    either have movement precluded or be monitored by a Senior Reactor Operator.
ALARA principles will be followed at all times during the        ign and conduct of e periments.
Approved.          h/                      '
Reactor Facility Dir@
                                                            -yan,gHD
                                                                                  $N        J N.'WI. Manderfield COL, U$AF, MSC Chairman, Reactor and Radiation Facility Safety Commi* tee
                                                            ,                        e
 
Reactor and Radiation Facility Safety Committee M.
[    hQ    '
f kore pD. Ashby      ,
LA&J R. Luersp J
M. Voth _
CDR J.E. DeCicco, MSC, USN O      a      o CAN C.B. Galley, ApC, USN i
 
                                            =_ -
Routine Reactor Authorization #4 September 1991
 
==
Introduction:==
The Reactor and Radiation Facility Safety Committee has reviewed and approved the operations described below. These experiments have been performed on numerous occasions in the past and are considered routine.
Authorizations: As permitted by NRC license R-84, the Reactor Facility Director may irradiate biological or non biological materials with an atomic number less than 93 and any experiment stnictural suppon and experiment containers in the reactor irradiation facilities subject to the limitations imposed by Technical Specifications and applicable procedures. !n particular, these experiments shall:
o      not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o      not create inventories of I-131 through I-135 greater than 1.3 curies and Sr-90 inventories greater than 5 millicuries, o      limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o      be doubly encapsulated if the release of the contained materia _1 can cause corrosion to the radiation facility, o      have an absolute worth less than $3.00, and o      either have movement precluded or be monitored by a Senior Reactor Operator.
ALARA principles will be followed at all times during the design and conduct of ex          'ments.
Approved:
Reactor Facility Di(eplor pirman,            D                  .
                                                                                    /4 NW. Manderfield COli USAF, MSC Chairman, Reactor and Radiation Facility Safety Committee
 
l 2eactor and Radiation Facility Safety Committee I
r
    .      hx)re
_C1  %
Ashby[,
            .1 id[/w
    . Luerip i\      Al
: t. votti lDR J.E. DeCicco, MSC, USN
[b        e    U APT C.B. Galley, MS% USN g
(
l
 
RSDR                                                                                    Forsbacka/mjf/51221 Date: 19 SEP 91 MEMORANDUM FOR RRFSC
 
==SUBJECT:==
Request Concurrence to Classify Krypton Gas Irradiation Experiment Protocol Under Routine Reactor Authorization #4 1.
 
==Purpose:==
The Naval Medical Research Institute (NMRI) has a requirement for a radioactive inett gas for use in the study of tissue inert gas exchange kinetics and bubble dynamics. Krypton when irradiated produces a isotope, namely Kr-85m, which is useful in this research. Inert gasses under pressures up to 2 atm have been irradiated safely under Routine Reactor Authorization # 102 dated September 4,1990.
: 2. Experiment
 
== Description:==
See attached letter from LCDR Novotny dated 16 SEP 91.
: 3. Safety Analysis: A worst case scenario accident analysis involves the complete release of the irradiated Krypton gas after a 1 MW-hr irradiation. The gas irradiation vessel has a volume of 446 ml. At a gas pressure of 40 psig, the vessel will deliver 1214 ml of gas at STP (0.0542 moles). The Neutron Activation Tables (Ed. Gerhard Erdtmann,1976) allow us to determine the quantities of radioactive Krypton produced after an irradiation at 1 MW for 1 hour. The composition of the gas is as follows:
Isotope        Abundwe                      On==+ity in Veael          Product      Half Life "Kr            0.35 %                        1.498E-2 g                "Kr          34.9 hr "Kr            2.25 %                      9.872E-2 g -              "Kr            13.3 sec
              '2Kr              11.6 %                    0.5218 g                  "Kr            1.86 hr "Kr              57.0 %                      2.656 g                    "Kr          4.5 hr-
              "Kr                17.3 %                    0.8157 g                  "Kr          76 min        i The heutron Activation Tables tell us the decays /second per microgram of target element given a 1 hour irradiation in a thermal neutron flux of-10" n/cm'sec, so the total amount of radioactivity produced is as follows:
Radioisotope              Total Ona=*ity Produced                Concentration in Recctor Room of Released "Kr                                  9.6 nucroci              1.05E-8 microCi/ml -
              "Kr                                  1.98E4 microCi-          2.16E-5 microCi/mi "Kr                                  7.39E5 microCi          -8.05E4 microCi/ml "Kr                                  5.48E4 microci            5.97E-5 microCi/ml s
              "Kr                                  7.01E3 microci            7.64E-6 microCi/ml The free volume of the reactor room is 32,400 ft', so the concentranon of radioisotope in the reactor room is found by dividing the total quantity produced by the free volume. Adding the above quantities of radioactive matenals yields a total of 0.82 Ci of radioactive laypton.
 
l 10 CFR 20, Appendix B specifies the allowed concentrations of radioactive mHerials in air in restricted and unrestricted areas based on inhalation for 40 hours per week for a 13 week period.
These limitations are as follows:
Allowed Concentration                Allowed Concentration Radiniiolene                  in RestricleiArca                    ULUntestricted Area
          ''Kr                  3E-9 microCi/mi                      lE-10 microCi/mi
          'Kr                    IE-6 microCi/ml                    3E-8 microCi/ml "Kr                    IE 6 microCi/ml                      3E-8 microCi/ml "Kr                  6E-6 microCi/ml                      IE-7 microCi/mi "Kr                    !E-6 microCi/ml                      2E 8 microCi/ml Given the condition of the total release of the irradiated krypton gas, it would take 15 half lives (27.9 hours) for "Kr and 10 half lives (45 hours) for "Kr to decay to below the specified allowed concentration in an unrestricted area. After 48 hours, "Kr concentrations will still exceed the 10 CFR 20, Appendix B limit to an unrestricted area by a factor of 50. To resume normal operations, it will be necessary to slug discharge the remaining gas through the reactor room dampers and dilute it with the air passing through the reactor gas stack. AFRR1 TR83-1, " Safety Analysis of Modifications to Upgrade the Reactor Ventilation System at Armed Forces Radiobiology Research Institute" specifies that 3430 cfm of air is exhausted from the reactor room into the reactor gas stack which has a total flow rate of 35,000 cfm. Simply opening the dampers dilutes the gas 10 a factor of 10. Opening the dampers for one minute in ten minute intervals lets out approximately one tenth of the gas and further diltas the remaining gas in the room. Since the air system is designed to change the air in the reactor room 4.4 times per hour, the dampers will need to be opened for 14 one minute periods to completely clear the room of        t krypton gas.
A straight forward approach to ensure the safe irradiation of krypton gas would be to perform the irradiation with the reactor room air dampers closed to isolate the air in the reactor room.
In the event of a gas vessel failure, the room will remained isolated for 48 hours to allow most of the radioactive gas to decay to acceptable levels for release to an unrestricted area. Since the design basis accident radiation release referred to in the facility safety analysis report is approximately 7 Ci for a mptured fuel element (Safety Analysis Report for AFRRI TRIGA Mark-F Reactor, page 6-19), the reactor confinement should easily contain the 0.82 Ci that could be released in a gas vessel rupture. After a 48 hour decay period, the dampers will be opened for one minute at ten minute intervals for 2.5 hours to sufficiently dilute the remaining radioactive gas as it is released to an unrestricted area.
These precautions are extremely conservative as 10 CFR 20.103 specifies that the concentration levels are based on a 40 hour per week exposure for 13 weeks. Because the gas will be effectively completely decayed in 14 days, there is no chance of exceeding the quarterly limit.
The hazard is further mitigated due to the fact that radioactive krypton is reble gas and only presents a submersion dose hazard unhke fission products which are far more dangerous.
l
: 4. Experiment Procedure: The following steps shall be taken during the irradiation of krypton gas in the gas irradiation vessel described earlier:
: 1) Manually close reactor- room dampers, ensure good seal of dampers with a vane anemometer. (DELETED 24 JAN 92)
: 2) Load krypton gas into irradiation vessel.
: 3) Irradiate krypton gas at 1 Mw for I hour.
: 4) Monitor gas pressure in vessel to ensure gas has not leaked.
A) If gas has pressure has decreased, stop irradiation immediately.. Notify                        .
ROS/RFD. Continue to monitor gas pressure. Assume worst case scenario and                          I (SEE ATrACHED GWIGE) isolate the reactor room for 48 hours. Following the 48 hour waiting period, open the dampers for one minute at ten minute intervals for 2.5 hours. Leave the dampers open after the 2.5 hours of opening and closing the dampers.'.' At least one reactor operator will be present in the reactor facility until the gas has been released from the reactor room.
B) If gas has been irradiated without incident, the gas will be cryogenically transferred into transfer vessel. Once the gas is outside of the reactor facility boundaries, the dampers shall be reopened. (OfANGED 24 JAN 92)
Submitted by:
j af Matt Forsbacka Capt, USAF-'
Reactor Operations Supervisor i
Approved by Reactor Facility Director:              Approved- by Chairnaan, Safety and Health Cew a;:
M      Moore Reactor Facility Director                          pDotig;              ,
                                                                                            'y
                                                                                      , Safety and Health Dtpartment -
1
_ - _ - _ _ - _                          _ _ _ .      Y
 
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l Approved by Reactor and Rs fiation Facility Safetv Co ntnittee:
Oh (    #
N.W. Mandertield                                                                                                        -
Colonel. USAF, MSC Chairman, RRFSC C7 w 4        ugfshby Chairrhan, Safety and Health Department r
dhbw Mark Moore                                    ['
Reactor Facility Director kb        uh Mark Voth                                                            C  Jofieph DeCicco, MSC, USN mm                              aiv                            d h b ,,
Ron Luersen                                                          CAPT C.B' Galley,/4SC, USN 1
l l .: _ -      _-_                    . _ - - _ . _        _ _ .      ...~..-~....._._.-._--._;_.-            . - - --. - . . - ...-
 
Routine Reactor Authorization #5 September 1991 Initeduction: The Reactor and Radiation Facility Safety Committee has reviewed and approved the operations desenbed below. These experiments have been performed ot. numerous occasions in the past and are considered routine.
Authorizations: As permitted by NRC license R 84, the Reactor Facility Director may irradiate up to 1000 gm of active or passive electronic components in the reactor irradiation facilities subject to the limitadons imposed by Technical Specifications and applicable procedures, in particular, these experiments shall; o    not be able to cause the release of radioactive gases and acrosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o    not create inventories of I-131 through I-135 greater than 1.3 curies and Sr-90 inventories greater than 5 millicuries, o    limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o    be doubly encapsulated if the release of the contained matenal can cause corrosion to the radiation facility, o    have an absolute worth less than $3.00, and o    either have movement precluded or be monitored by a Senior Reactor Operator.
ALARA principles will be followed at all times during the design and conduct of experi      nts.
Approved:
Reactor Facility Direcief
                                                                                      ~
6dehairman S .D i
NW. Manderfield col., USAF, MSC Chairman, Reactor and Radiation Facility Safety Committee -
 
Reactor and Radiation Facility Safety Committee
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(24W R. Luerso,n I
cu hi, Voth CDR J.E. DeCicco, MSC, USN
      $1 alley, DAPT C.B.
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ATTACHMENT E 10 CFR 50.59 Safety Evaluations of Modifications, Changes, and Enhancements to Procedures or Facilities (Other than Fuel-Follower Control Rods)                                    1
                                                                )
 
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Facility Modification Worksheet 1 10 CFR 50.59 Analysis Proposed Change                                        Modify Transient Rod Drive Support Structure i
capt Forsbacka                    Date    U CI 9I Submitted by:
: 1. Description of change:
Raise transient rod drive to position approximately 60 inches above its current position. This will require a connecting rod which is approximately 60 inches longer than the current connecting rod.
: 2. Reason for change:
The purpose of this modification is to facilitate the maintenance of the transient red drive mechanism and to unclutter the space between the pool water and bottom of the carriage. This modification will allow for the complete change out of the transient rod drive mechanism without disrupting the other control rods.
: 3. Verify that the proposed change does not involve a change to the Technical-Specifications or produce an unresolved safety issue as specified in 10 CFR                                        '
50.59(a)(2). - Attach an analysis to show this.
This change does not involve a Technical Specifications change.
Analysis attached? Yes - X
: 4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).
Procedure                                        Facility X        Experiment
      ; Revised: 15 May 91                                                                                      Page 1      j A TTAcHMEwi~ 5-l  - - _ - _ _ _ _ _ _ -
 
Facility Modification Worksheet 1
: 5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this description need not contain the final SAR wording.
Section 4.10 Reactor Control Components will need to be modified to describe the new physical position of the transient control rod drive mechanism.
: 6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.
No modification to the actual rod drive mechanism will be made. This modification.
is simply to change the physical location of the control rod drive mechnaism.
The transient rod drive mechansim will operate in accordance with its design intent.
Revised: 1$ May 91                                                                  Page 4 l
 
Facility Modification Worksheet 1 l
l~                                                                                                                                                i
: 7. Specify associated information.
New drawings are:        Attached        ,jt.JPhoto is attached, drawing to be produced -
                                                                ** * ***** d***)
Nm giM Does a drawing need to be sent to Logistics?              Yes X    No Are training materials effected?                          Yes      No x Will any Logs have to be changed?                        Yes      No    x Are other procedures effected?                            Yes      No  -v List of items affected:
Q%&                                          M yab.
I
: 8. Create an Action Sheet containing a list of associated work speci6ed in item #-
7, attach a copy, and submit another to the RFD.
Action Sheet:          Submitted      X                      Not Required-l Reviewed and approved by' RFD                                              Date A                                          Ni[
RRFSC Concurrence -                  ,                                  - Date l
Revised: 15 May 91                                                                                                          Page 5 L
 
k L
Safety Analysis of Transient Rod Drive Support Modification The transient rod drive will be raised approximately 60 inches above its current location. The purpose of this modification is to facilitate maintenance of the transient rod drive mecimanism.
No modification of the function of the transient control rod drive will be made. All normal testing will be performed to insure that the systems will operate in accordance with its design intent.
The photographs below illustrate the modification made:
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Action Sheet                                                                                                                            !,
subiect omce symoon DCBD sv. pen..      ]
Modification of Transient Control Rod Drive Support                                                        23 OCT 91 ACtlen A44Wi red Concurrence and Approval uemer.noom for meewa,icescru oneay me reouvemems. escurouno ano accon tuen or recommenoso vust ce su ,c>ency          r      cernoea to
,conety me ecnon *imout recourse to emw sources i
: 1. The transient control rod drive mechanism needs to be raised by about 60 inches above its current location to allow for ease of maintenance.
LOGD needs to design and appropriate support structure to keep the rod drive mechanism securely in place above its current location.
: 2. Once built, we'll need to make a drawing of the finalized design for our as-built drswings. Autocad would probably be best.
: 3. No training materials should be effected by this modification, but we should brief the staff on why the change was made.
on              LA              OCttA4. /h                              O CACud '                    h                . Jh k4CkP-                                                .
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i                                                      1 c      ne  n.                                    e_:
Office                                peesne                      Phone                                  in nenD        w. 9 wmn .                                        s-179n                                      7/
o(ruf      6 h r- 4 Q fe r.                                  s i in o                                      V Disoetenes (Dtg Show Aadiuonal Coorenstkm on Reverse $de er Conenuetten Sheet Actmn Otticer (Name, grace, t,;amne ord signature)
Matt Forsbacka. Capt USAF. 5-1221 eD3"                  Hegroooo when Seoersted from caserned Deamnt
 
ATTACHMENT E-2
 
Attachment I hEFERENCE: ADMINI3TRATIVE FROCEDURE I. FACILITY MODIFICATIONS ANALYJ13 OF FROPOSED MODIFICATION Modification Nomenclature:              REPLACEMENT          OF  WATER          GAMMA ACTIVITY MONITORING SYSTEM Analysis: SFC LAUGHERY                Date:    12 February 1991 SECTION A
: 1. Document analysis          to determine if a change                      to the Technical Specification is required.                      Include 10 CFR and/or Technical Specification references as applicable.
NO CHANGE TO THE TECHNICAL SPECIFICATION IS REQUIRED
: 2. If    your  analysis        determines    that            a          Technical Specification change is required, go to SECTION B.
: 3. If a Technical Specification change is not required, document an analysis to determine if the prvposed modification would constitute a change to the facility as described in the SAR. Include 10 UFR ind r                                    Ab references as applicable.
NO CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR IS REQUIRED. The proposed modification is the replacement of the vacuum tube operated monitoring system with a solid state system.            This modification is necessary because replacement vacuum tubes are no longer available.
: 4. Document    an  analysis        to  determine          if  the          proposed modification would constitute a change in a procedure as described in the SAR. Include license and/or SAR references as applicable.
PROPOSED MODIFICATION DOES NOT CONSTITUTE A C11ANGE IN A PROCEDURE AS DESCRIBED IN THE SAR
: 5. Document an analysis to determine if the proposed modification would constitute a change in the tests or experiments as described in the SAR. Include tieens-and/or SAR references as applicable.
PROPOSED MODIFICATION DOES NOT CONSTITUTE A CHANGE IN THE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR
: 6. If the proposed modification does not constitute a change to the facility, procedures,            tests or experiments as B TTRC HMEn T    E- L
 
  . ECT10t1 C
: 1.      tio 10 CFR 50.59 is required. The analysis in SECTIO!15 A.3. A.4,                A.5 provide the bases for this determination that:
: a.                no change                in the Technical  Specifleation    is required.
: b.                  no change in the facility as described in the current SAR is proposed.
: c.                  no change in the procedures as described in the current SAR is proposed, and
: d.                  the proposed test or experiment:
(1)  coincides with those described in the current SAR; c                                    (2)  is permitted by the Technical Specifications, and (3)  has been previously reviewed and conducted.
: 2.      This modification has been reviewed in accordance with ALARA principles. Comments (if necessary) are as f ollo -
1 b                                    j          Mhh
: 4.      Tw . - _                    approved by RFD            i          _ _ D.,:  t.fg /fi/
: 5.      RRFSC Concurrence                                              /
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Attachment i 1*Ei EhEth;E : ADMINISThATI' E i-hOaDUhE I. FAtlLITY MvDIFICnTION6 ANALYbid OF EhulO5ED MODIF1'AT1uN Modirteattun ik.me nc l a tu re      REPLACEMENT            OF      WATER  CONDUCTIVITY MONITORING SYSTEM Ana ly s u : SFC LAUGHERY          Date:  12 February 1991
    .iECTION A
: 1. Document analysis to determine if a change to the Technicil
              .3pectfication is required.        Include 10          CFR and/or Technical Specification references as applicable.
NO CHANGE TO T!!E      HNICAL SPECIFICATION IS REQUIRED
: 2. II your analysis c . ermines that a Technical Specification change is required, go      SECTION B.
: 3. If a Technical Specification change is not required, document. an analysis    to determine    it  the prvposed                  modification would constitute a change to the tacility as described in '!r .4 h .
Include 10 CFR and/or SAh reIerences as applicable.
NO CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR IS REQUIRED.
The proposed modification is the replacement of the present monitoring system with a new system by. the same manufacture This modification is necessary because replacement parts for the old system are no longer available.
: 4. Document an analysis to determine if the proposed modificatien would constitute a change in a procedure as described in the SAR.
Include license and/or SAR references as applicable.
PROPOSKD HODIFICATION DOES NOT CONSTITUTE A CHANGE IN A PROCELURE AS DESCRIBED IN THE SAR
: 5. Document an analysis to determine if the proposed modification would constitute a change        in the        tests or experiments                    as described applicable.
in the SAR. Include license and/or SAR references as PROPOSED MODIFICATION DOES NOT CONSTITUTE A CHANGE IN THE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR
: 6. If the proposed modification does not constitute a change to the facility, procedures, tests or experiments as described in th-SAR TAnswer to SECTIONS A.3, A 4 and A,6 is 'Nv"                      in all casesi.
Go to SECTION C.
'7T/MHME]YT Ed                                                                                              1
_ _ _ _ _        _ - - - - -    .J
 
dECT M!i C 1    !J . 10 CFh 50.03 $s required. The analysis in IE' f l0!1.3 A.3. A.4,                        A.5 provide the bases for thi.e de t e rmi r.s t i . n thst:
: a. no change                            in    the  Technical        Specifteation        is required.
: b. no change in the facility as describ-;d                                            in the current SAR is proposed.
: c. no change in the procedures as described in the                                          I current SitR is proposed, and I
: d. the proposed test or experiment (1)                        coincides with those described in the current 4
                                                                  '6 AR ;
(2)                        is permitted by the Technical Specifications, and (3)                        has been previously reviewed and conducted.
: 2. This modification has been reviewed in accordance with ALARA principles Comments (if necessary) are as f ollows:
: 4. Reviewed and appr >ved by RFD                                      Ilb >    -
Date1[lfh/
: 5. RRFSC Concurrence .                                      d# ,([u __ _ Date.2fffl 9
 
E ATTAOiMENT E-4
 
_ _ _ _ _ _ _      _--_- - - - -                                                                      ~
Facility Modification Worksheet 2 No 10 CFR 50.59 Analysis Required Proposed Change                              " "*" " # 1 ' " '
* 8 Y " ' * * * * " ' d
* Modification to:                            Pr ocedure .              Facility        "      Experiment Submitted by: ceor2e Date s %    m                        -
: 1. Description of change:
that will allow operators to override the cam dampor closure system to                                                                  - -
the dampers to open when the came are alaming.                                          System will only be operated upon the orders of the RFD.
will be detected before any daily operations. Operation of the                                                    ems damper closure s THIS FFCRS.                      SYSTEM IS TEMPORARY AND WILL BE RDt0VED AFTER SUCES                                                  -
USE oF THIS SYSTEM VILL BE ALLOWED oNLY BY DIRECT AUTHR 2 Verify that the proposed change does not involve a change to the Technical
                                                                                                                                                                          =
Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved safety question as defined in 10 CFR 50.59(a)(2),
intended.
: 3. If change involves a facility modification,t attach a drawing if a structural Logistics,                          facility drawings need updating, forward a copy of changes necessary to n/A
: 4. Determine what other procedures, logs, or training material may be affected and record below.
Staff bziefi
: 5. List of associated drawmgs,ng                          procedures, logs,      will or beotreautred. her materials to be changed:
Sea attached drawing of system for reactor f.iles.
: 6. Create an Action Sheet containing the list of associated work specified above, attach a copy, and submit it to the RFD.
Action Sheet:                                Submitted      x                Not Required _
Reviewed and approved by RFD RRFSC Notified
                                                                                                                              /      Date[b I/
[        Date 17 DEC f991
                    ^
Revised: 15 May 91 Page 6 HTi?%HM6tvr E-9            _ _ _ - - _ - - - - - - - _ _                                      . _ _ _
 
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3                                *c                                        : {  ,,.                                              w 4
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  $                                                                                                                                  si e                                                                                                                                  x
    )
          .m.
 
                                                                                                                  ' ~ - --
l TRAINING POR DAMPER OVERRIDE BOX GENERAL The damper closure system located in the penthouse requires a closed loop to cause the dampers to close. When the loop is open the dampers can be opened 'nd will remain open. A two conductor wire is run from the penthouse to a relay Ln the motor control center. This relay is operated by AC voltage from the primary CAM. If the primary CAM alarms, AC power will be turned off to the relay in the motor control centnr.
If the backup CAM alarms it will break the AC power with a relay in the black box in the reactor room under R2.
Under normal operation power from the primary cam energizes the relay in the motor control center. This opens the loop to the penthouse damper closure system which will allow the dampers to be open.
DAMPER OVERRIDE:
The Cverride system consists of a two conductor wire with a plug together.
on the end of each wire so that they can be plugged The wire breaks into the two conductor wire which comes from the penthouse and simply acts as an extension of that wire. One end of the new wire connects into the penthouse wire at the relay in the motor control center. The l
other end of the wire, the end with the connector, is located in 3152 on the wall in a labeled box above the air compressor.
Under normal operations the connector will always be connected.
will operate This                  properly.will act like a normal wire and the system In the event of an emergency and if the RFD orders it, the connector can be disconnected.
Opening this connector will open the circuit to the penthouse thus allowing the dampers to be opened.
The orders                  connector from the should RFD. never be disconnected without direct
 
                                                                    ~
contro Numoer Action Sheet                                                                                                                  '
swo,0.,                                                                                    omee    ,mt.o.    .w.,en.e      ,
n em a                          !
                                                                                                                        >: : s Temporary Installation of CAM Damper Override                                    sete                              q System                                                                                      e DEC 91              j AC16on Sequired l
Training of Staff in use of override system, removal of system when no longer raquiri womorenowm eor necore. icesence oneay ene roouvemeets eackroun o ano action taaen or 'ecommencea vust te sec eavy cetnoea +o cents v ecton .cout recovise to omer sources i
: 1. A CAM damper override system is proposed for the purpose of allowing the air dampers to be opened if necessary during the pulse testing of the FFCRs.        THIS SYSTEM IS TEMPORARY, AND USE OF THIS SYSTEM MAY ONLY BE AUTHORIZED BY THE RFD.
: 2. Once the FFCRs are fully tested, this override system will be removed.
l l
I, tConenue on can bono) w__
Coormaoiseas                                  a,,rweie OWs                                      %                                  Pheae                  inattale          Date RSDR                Mark Moore. RFD                                    5-1290 Dispatched (Dtt)
  ^
Show Adeltional CoordinstNm on AWeree Ede er Continuetton Sheet Action Omcor (Narne. graoe. phone and segrature)
Matt Forsbacka Capt USAF. 5-1290 DNA Form 547 4 Jan as                  nega$ed when Seperstvo mun Clase#Aed Docurnent J JW      d
 
4
,                    ATTACHMENT E-5 pu iiiiin
 
4e        , 4    ,
                          . r ni Mi F        ,1 N I N I . o b /' ' 'F    t    ".'""1        -  I'.4it  W ' t ' I t' l ' % 1  '
N.
                          ..h              y i fe }    .:4 1, Mt t iicic;A T 10 N Mc 1 i t i c *s t t on N : mere; i si a re Au x Lil a ry_ Ti toe t a n_. H o tL C oli ti ol e An a !.v 3 i s Ofil 14nu11hu rY                Date 2Q 30 m 'et 1990 J,ECTION A
,                          1    Document analycis t,o determine if a change to the Te c hn a -:31 5pecification is required.                          Incjude 10 CFR and/or Technical
                                .3pucification references as spplicable.
NO CilANGE TO Tile TECllNICAh SPECIFICATION IS REQUIhED.
                          .;      if your analysis determiner., that a Tevhnim i Stua1 tie ,tlun change is reanir+d, go to CECTION B.
J.      If a Technical 6pecifluation ;hange 1 nut requate,s. 1.4 om-nt an snaLys)e to determine ti the proposed modifientiva would constitute a change to the facility as described in the f. A R .
Include 10 CFR and/or SAR references as applicable.
TIIERE IS NO CilANGE TO THE FACIh1TY AS DESCRIBED IN Tile GAR.
: 4. Document an analysis to determine if the proposed modification would constitute a change in a procedure as described in the SAR. Include license and/or SAR references as applicable.
T!fERE IS NO C11ANGE IN ANY PROCEDURES AS DESCRIBED IN Tile SAH.
: 5. Document an analysis to determine if the proposed modification would constitute a change in the tests or experiments as described in the SAR. Include lict;nse and/or SAR references as applicable.
THERE ARE WO CHANGES IN TIIE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR.
: 6.      If the proposed modification does not constitute a change to the facility, procedures, tests or experimants as described in the SAR (Answer to SECTIONS A.3, A.4 and A.S is 'N0"                                                        in all cases). Go to SECTION C.
e 77t]C HMsgr E~ ~l
 
              .Ec110N O 1              N.                              p; i: Fis 00.59 te re;uired. The an4lytis in 3ECT MN A A . 3.
is 4.                              A . '- prcvide the b a s e t.; f or this detertnirnt ien t hM
: s.      no                              change  in    the  Technical    Opeuitioiti n          1*
required,
: b.      no change                              in  the facility    is dv e c. r i be d in  ' h --
current SAh in proposed.
e        no change in the procedures as des 'ribed in the current SAR is proposed. and
: d.      the proposed test or capt ringmt (1)                            evincides with those described in the current 6AR;
( !:                            is permitted by the Technical ap"eification*,
t3)                            has been previously reviewed and conducted.
J.
This modification has been reviewed in accordance with ALAhA principles.                                                          Comments (if necessary) are as follows:
The ne.w reactor console has a COUt1T DOW11 timer.
The auxiliary timer is a COUt1T UP timer which facilitates record keeping during experiments by displaying the actual time the reacter was at power.
3.
W Reviewed and approved by, R03-_ .                                                                    _____,___ Date L M t._fi
: 4.              Reviewed and approved by RFD,.b                                                                    b g _ Date L U d 7/
: 5.            RRFSC Concurrence _                                                                        #ff
_    _          4        __ Date @ [f 7/
t
 
ATTACfWENT E-6 0
 
1 Facility Modification Worksheet i 10 CFR 50.59 Analysis Proposed Change                                  t'p g r a d e pulse dispiny capabilitie. and install variable pulse timer as specified in origin.?1 contract with General Atn~fc.                                                            __
Submitted by:                                    capt Forsbacka                      Date    13 SUV 91
: 1. Description of change:                                                                                              _
Modification of both NPP-1000s for small pulse operation. This modification allows a gain change for high sansitivity pulse data aquis4. tion. Addition of an Eagle dig timer and relay logic to scram the reactor after initiating a pulse.
: 2. Reason for change:
To fulfill original GA contract obligations.
: 3. Verify that the proposed change does not involve a change to the Technical Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.
Analysis attached? Yes Not Required
: 4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).
Procedure Facility          Experiment Revised: 15 May 91                                                                                          Page 3 B TT/9 CHM EN T                6-6
 
Facility Modification Worksheet 1
: 5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this det.cription need not contain the final SAR wording.
Section 4.14. Scram Logic Circuitry: Add description of pulse mode scram t ic:e r .
: 6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.
Botti modifications are for the purpose of ensuring that the console operates in accordance with its design intent.
f I              Revised: 15 May 91                                                                  Page 4
 
Facihty Modification Worksheet 1
: 7. Specify associated information.
New drawings are:              Attached                                        x                                                              -
Not required Does a drawing need to be sent to Logistics?                                                            Yes .            No
* Are training materials effected?                                                                        Yes X            No Will any Logs have to be changed?                                                                        Yes            No    x Are other procedures effected?                                                                          Yes          v  No List of items affected:
Training materials will have to reflect the ability to view posses in two modes.
Procedure 8. TAB G1 and C2. will have to be modified to instruct the operator whisn mode of pulse display should be uced.
: 8. Create an Action Sheet containing a list of associated work specified in item #
7, attach a copy, and submit another to the RFD.
Action Sheet:              Submitted X                                                                            Not Required Reviewed and approved by RFD                                                                                                    Date/3  /V [
h RRFSC Concurrence                                                                                                                Date 17 DEC 1991 s
  ~
Revised: 15 May 91                                                                                                                        Page 5
 
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t            :              o~              :            m                ?
g      ~v .                      .
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                        ,,                        UNITED STATES
        !-      'O ) i                NUCLEAR REGULATORY COMMISSION
        ',, Sg F 4                            W A$HINGTON. D C. h?A
            ^'' . ,' f*'f
* cctober 8, 1991 Docket No. 50-170 Colonel George W. Irving,111. BSC, USAF Director Armed Forces Radiobiology Research Institute Bethesda, Maryland ?0814-5415
 
==Dear Colonel Irving:==
 
==SUBJECT:==
ISSUANCE OF AMENOMENT NO. 21 TO FACILITY OPERATING LICENSE NO. R ARMED FORCES RAD 10 BIOLOGY RESEARCH INSTITUTE (AFRRI)
The Commission has issued the enclosed Amendment No. 21 to Facility Operating License No. R-84 for the AFRRI TRIGA Research Reactor. The arnendment consists of changes to the Technical Specifications in response to your submittal of April 30, 1990, as supplemented on December 17,1990, March 5,1991, May 17, 1991, August 16, 1991, and September 10, 1991.
The amendment (1) corrects errors in typography and grasunar, (2) increases        -- -
the maximum licensed steady state reactor power to 1100 kilowatts. (3) author-izes installation of fuel follower control rods, (4) clarifies the transfer of Reactor Facility Director (RFD) responsibilities in the absence of the RFD, and (5) allows operational flexibility in performing surveillance testing of the ventilation systese for the reactor facility.
Enclosure 2 is a copy of the related Safety Evaluation supporting Amendment No. 21.
Sincerely, c              r Alexander Adams, Jr., Project Ianager Non-Power Reactors, Decommissioning and Environmental Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation
 
==Enclosures:==
: 1. Amendment No. 21
: 2. Safety Evaluation cc w/ enclosures:
See next page C.-
 
                                                                                        **""""**~
Action Sheet s.e, t                                                                                om.e s,m-                  ....en.e D R B P'                  N/A        '
Installation ot upgraded pulse display and variable                          p.to 13 NOV 91          i pulse timer.
A tion Required Modify Procedure 8. TAB G1 und TAE CL womeranowm tw socore tooecnte trear tee reownements. escaroung saa octon teen or recommenoso vwat t4 *w"wey mareo to eenm tne acton .itnout recouru to othw sourcee 6
: 1. The new pulse display will allow us to select high or low resolution, depending on the size of the pulse. The low resolution will be used for large pulses and tha high resolution will be used for small pulses. We need to inodify procedure
: 8. TAB G1 and TAB G2 to reflect this option.
(Contrive on > tonc) l                                              I Coertiencuene                                Appreenis Otsce                                ifeme                            Phone                            Inty          Date AC        w._
RSDR        Mark Moore. RFD                                        5-1290                          '
                                                                                                              /?PD      fjM*f(
WQ hk u su :L Diepesched (Ots)
Show Additional Coeralinet6en on A* verse Side er Corttmusten Sheet Acuen ovner m grooe, onone ano egne,ures Matt Forsbacka, Capt USAF. 5-1290 P*NAfonn 53    547    Megrened when separet d from Closetw Document
 
FCOM                        12/06<B4 15: 46  P. 1 l
FAX TRANSMISSION GENERAL ATOMICS TRIGA REACTORS FROM:        a111 nood                              PAGE 1 OF 2 TO:          narry spence, ArnaI
 
==SUBJECT:==
Work Performed week of 11-18-91        DATE: 12-06-91 QLD._1URIMEEEI a) Installed Pulsa Mode Scram Timer. Added a Eagle digital timer and relay logic to scram the reactor a preset time after initiating a pulse. The timer is mounted behind the chart recorders in the CSC, and is intertacnd to the CSC computer and scram loop through a pair of relays.      The timer is only active in PULBE mode, and is started by firing the pulse rod.
There is no provision to disable the timer in pulse mode. The console software provides for a maximum scram delay of 15 seconds in pulse. If the operator wishes to scram the reactor prior to the 15 cocond cottware scram timeout, the desired time must be entered to the eagle timer prior to initiation of the pulse. Note that due to the software configuration the Csc will not post the Pulse Mode Scram Timer scram message until all pulse data acquisition is completed.
Rob and Mike were given a copy of the Pulse Mode Scram Timer and the updated CSC scram loop schematics, b) Replaced the serial data cable between the CSC computer and the Tektronix high resolution monitor. The new cable is a two piece cable, which allows the disconnection of the high resolution monitor at the rear of the computer chassia.      The AFRRI vire list was updated to reflect the wiring change (see    .0 Rob or Mike for a copy of the list).                            T
                                                                                .0 EHEREM't. ORDBRB t a) Fuel Follower Control Rods. You        should have all of the information on this subject already.
 
        *0M                                          12 esesa i5 as P. I b) small Pulse Data Acquisition.
                  }!ARDWARE: Modified both NPP-1000's for the small pulse option.
This modification allows a gain change for high sensitivity pulse data acquisition. Also modified DOH-32 relay output #18 (in the DAC) to enable the gain change in pulse mode.
Rob and Mike hava ocpien of the changes.
SOFTWAREl Installed now software to allow the acquisition of both normal and small pulsos. The high sensitivity pulso data acquisition will acquire up to 1000 MW pulses (full scale) .
5 r
t
__                                                                  TOTAL P. 2
 
ATTACHMENT F n
May 1991 Summary of Changes to Administrative and Operational Procedures i
t s
 
Summary of Changes to Administrative and Operational Procedures introduction in May,1991 RSDR stalf completely updated the Administrative and Operational Procedures to implement grammatical changes, a new format for the procedures, and corrections throughout including updating the core position references.
Administrative Procedures
{
A1. Add reporting requirements to (2).
A2 Delete reference to ' escorted access roster *.
A3. Complete revision / simplification of 10 CFR 50.59 worksheets.
A4. Completely new procedure developed in conjunction with NRC recommendations.
Operations] Procedures                                                                                                        E O.      Change ' initial block' to " signature block".
: 1. Add " acting ROS" to 2.f.
lA. Delete escorted access roster (2.b.).
Change "open* to " enter" (4.a.).
Delete material that duplicated HPP 3-1 (7.c.).
Update badge types for entry (2.b.).
: 18. Change " element F28' to ' desired element" (l.c. and 4.g.).
Delete reference to " upper end indicator
* painted on cable (3.b.).
IC. Delete requirement place lead bricks on tube supports (1.a.(1)).
Delete references m former ' limit switch" (l.a.(5)).
Delete reference to having prep area ' scaled off" during operations (3.f.).
1D. No changes.
IB. Change "SRO' to "!W operator' (1.).
Add "in the reactor room' to (3).
: 2.      Delete statement 'a record of operatious will be kept for each trainee / operator" (2.c.),
d
: 3.      Add *and conwle system manuals
* to (1).
: 4.      add 'during operations' to (1.b.).
: 5.      Add (3).
: 6.      No changes.
: 7.      Revise references to Shutdown Margin throughout.
: 8.      Add (5.f.).
8A.      Delete reference to " Mode 1, I A, etc' (5.a.(5)).
8D.      Delete measurement of pool level in 3161 (V.3.).
Change ' conductivity" to ' resistivity" (VI.5.).
Change ' pool" to " inlet" (VI.17.g.).
f,H l . Sarne changes as in 88.
8C.      No changes.
8 D. Change 15 Watts to 5 Watts.
Delete reference to monthly summary sheet (4).
8E.      Update title.
Change ' servo nxxle' to " auto irode' throughout.
Change 800 Kw to 200 kW in (2).
Add (7).
8Fl. Change ' servo' to " auto
* throughout.
Add general staternent.
Add specific (1).
Add second sentence in (13).
Add (16).
8F2. Change ' servo" to " auto
* throughout.
Add general statement.
Add specific (1) and (12).
Add second sentence in (9).
301. Add mnimum pulse sitcs to (5).
Add different types of operations to (9).
802. Add pulse sires to (5),
g            Add different types of operations to (12).
 
811.      Add (IV.b.). Remove referes                                  i 1.0 to automatic scram reset, change "conouctivity to
                                  ' resistivity".
: 81.        Update (VI.2.) and (VI.4.).
: 9.        Add 'while the power rails are energized
* to (l.f.).
Add 'or mercury compounds in any form
* to (2).
: 10.        Moved from Procedure 8. Tab K to new Procedure 10. Add additional instructions to step 3 to better illustrate how to check the high voltage setting of the SGM l
: 11.        Formerly was attachment to Procedure Vill Tabs B and Bl.
_ _ - _ _ - _ _ - _ _ - _}}

Latest revision as of 01:06, 13 May 2020

Afrri Triga Reactor 1991 Annual Rept. W/Undated Ltr
ML20091A862
Person / Time
Site: Armed Forces Radiobiology Research Institute
Issue date: 12/31/1991
From: Bumgarner R, Maria Moore
ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9203300221
Download: ML20091A862 (233)
v • d • e


Text

TEC H SPfcS 6 .- o DOCWET [d#lld DEFENSE NUCLEAR AGENCY G.

ARMED FORCES nADIOnlOLOGY RESE ARCH INSTITUTE DETHE SD A. F.e AnYLAND 20889-5145 RSDR

SUBJECT:

Submission of Annual Report U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Dear Sir:

Attached please find the 1991 Annual. Report for the AFRRI TRICA reactor facility, submitted as required by license R-84,-facility docket 50-170.

Should you need any further information, please contact the undersigned at (301) 295-1290.

-l l)i]il Attachment WARK WOORE as stated Reactor Facility Director CY Furn:

U.S. Nuclear Regulatory Commission ~q

^

ATTN: Mr. Warvin Mendonca, Wall Stop 11H10 Washington,'DC 20555 U. S. Nuclear Regulatory Commission, Region I' ATTN: Mr. Thomas Dragoun 475 Allendale Road King of Prussia, PA 19406 pi) 9203300221 911231 2 ' ' 8 .VV PDR- ADOCK 05000170 .

-R PDR.

~._.. _ . __ . , . _ _ - . .. _._.:.._....a._ . . - , -

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l 1991 I

i ANNLAL REPORT i

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1991 ANNUAL REPORT OF AFRRI TRIGA REACTOR

)

Submitted by:

MARK MOORE Reactor Facility Director Docket 50 - 170 License R - 84

i Reviewed and Approved b ,

Yhulbh t M ARK L. MOORE [j 23/lbAcq Date '

Reactor Facility Direditu Approved for Release (mk3 '"&A 9 )

ROBERT L. BUM @\RN ER Date Captain, MC, USN Director

l 1991 ANNUAL REPORT  ;

TABLE OF CONTENTS Intr 6 duction f f

Generrn Information i becticn I Changes to tite f acility des!gn, performance cheracterletics and operational procedures. Results of survelllance tests and inspections  !

Section II Energy generated by current reactor core and number of pulses 1 82.00 See,lon III '

Utscheduled shutdownsSection IV Safety-related corrective mainterance Section V Faci!Ity changes and changes to procedures as described in the Safety Analysis Report. New experiments or tests during the year.

Section VI Sumary of radioictive effluent released Section VII Environmental radiological surveys Soction VIII Exposures granter than 25% of 10 CFR 20 limits Attachment A 10 CFR 50.59 analysis and Refueling Plan for fuel-follower control rrta. ,

Attachment B Current Reactor Administrative and Operational Procedures Attachment C Amendment No. 21 to Facility Operating Licensa Attachment D Routine Reactor Authorizations Attachment E 10 CFR 50.5g safety evaluations of modifications, changes, and enhancements to procedures or facilities (other than fuel-follower control rods).

Attachment F May 1991 sumary of changes to administrative and operational procedures -

1

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) INTRODUCTION ,

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f' PLUG 000R Cutaway View of AFRRI TRIGA Reactor

1991 ANNUAL REPORT INTRODUCTION:

In 1991, the AFRRI reactor staff accomplished two important milestones in continuing efforts to improve the operational capabilities and reliability of #

the reactor facility. Af ter Nuclear Regulatory Commission (NRC) approval of the installation through Amendment 21 to the f acility operating license (Attachment C), the reactor staf f installed three new f uel-follower standard control rods, s i

new longer air-follower transient control rod, and related equipment during an exter.ded maintenance shutdown period of November-December 1991. Significant post-installation testing and calibration was performed to verify the validity of the original safety analysis and refueling plan (Attachment A) and the reactor was returned to regular steady-state operation on 16 December 1991.

Approval for resumption of pulsing operations was withheld by the Reactor Facility Director pending completion of an testing program of incremental sized pulses in early 1992. That testing program will include pulses up to 82.S0, but the current administrative pulse size Iimit of $2,00 remains in offect for normal operations.

Also, the microprocessor-based instrumentation and control consolo installed in 1990 completed its first fu!I year of successful operation. Two modifications mndo to the console to improve its capabilities are discussed fully in Sections I and V and Attachment E.

The Reactor Facility was inspected by the Defense Nuclear Agency Inspector Osneral fren 25 to 27 September 1991. The inspection found that the AFRRI TRICA Rsactor facility received an everall grade of SATISFACTORY with no significant and discrepanciesnotedinorganization, sofeguards. There were no reportableoperations,materialhandlinglntenance deficiencies found in the ma records, operational log books, or training records. During the inspection, four minor deficiencies were documented. None of the cited deficiencies were considered significant and, neither singly nor in aggregate, impaired the performance or degraded the safety of the TRICA nuclear reactor operations.

Concurrently with that inspection, the reactor facility also received a Nuclear R actor Security Inspection by representatives of the Assistant to the Secretary of Defense, Atomic Energy. That inepection also gave the reactor facility an overall rating of SATISFACTORY.

The Reactor Facility was also inspected by NRC personnel from Region I on 23-2S April 1991. No vioistions were identified during this inspection.-

Changes were made to the procedures and facilities during 1991. These_ changes ccre sepported by an extensive safety review process in accordance with the provisions of 10 CFR 60.59. The changes will be discussed fully in Sections I and V.

No trainees war.,added to the reactor staff during 1991. Two former trainees cbtained Senior Reactor Operator licenses and one former trainee received his Reactor Operator license. One Senior Reactor Operator, Mr. Thomas Wright, d; parted curing the year. Also, one Senior Reactor Operetor, Mr. Stephen Holmes, chose license was terminated on 29 November 1990 was relicensed in 1991.

R: quests from non-AFRRI investigators continued to supplement the substantial inhouse experimental work load. These experimenters included representatives from the National Institutes of Health (NIH), Smithsonian Institution, National Institute of Standards and Technology-(NIST), Naval Medical Research Institute 2

_ - ~ - . _ _. _ _ _ _ _ _ _ _ . _ _ _ . . _ - _ _ _ _ _ _ _ , - _ ., _

(NMRI), and the University of Waryland at Baltimoro. Tho reactor staff was also tasked with providing personnel to assist in conducting inspections of the Fast Burst Reactor facilities at Aberdeen Proving Grounds, Maryland and White Sands Wissile Range, New Wexico. At the request of Cornell University, an operations audit of their reactor facility was conducted.

The AFRRI documentation on financial assurance for decommissioning was accepted by the NRC on 11 February 1991.

No Licensee Event Reports were submitted during the year, however there were several malfunctions and unplanned shutdowns that are discussed in Sections III and IV.

The remainder of this report is written in a format to include notification items required by the AFRRI TRICA Reactor Technical Specifications. Items not specifically required but of general informational value are presented in the General Information section. Each section following the general information corresponds to the required section as listed in Section 6.6.1.b of the AFRRI TRICA Reactor lechnical Specifications.

3

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GENERAL INFORMATION Key Personnel Reactor and Radiation Facility Safety Committee.

l l

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l

CENERAL INFORMATION:

All personnel held their poaltions as listed throughout the entire year unless otherwise specified.

1. Current key AFRRI personnel (as of 31 December 1991) are as follows:

Director - CAPT Rabert L. Bumgarner, USN (ef fective 30 August)

Chairmate, Radiation Sources Department - CAPT C. B. Calley, USN Chairman, Safety and Health Department - Mr. Thomas J. O'Brien and AFRRI Radiation Protection Officer (ef fective 1 June)

2. Reactor Facility Director - Mr. Mark Moore (SRD)
3. Current key Reactor Operations Personnel:

Reactor Operations Supervisor - Capt Matthew Forsbacks, USAF (SRO)

(effective 31 May)

Training Coordinator - CPT Christopher Owens (SRO ef fective 7 May)

Walntenance/ Procurement - Wr. Robert George (SRO) '

Administration - MSO Harry Spence (SRO)

Other Senior Reactor Operators - Wr. John Nguyen (SRO ef fective 7 May)

Wr. Stephen Holmes (SRO ef fective 15 Oct)

Reactor Operator - SFC Wichael Laughery (R0 ef fective 7 May)

4. Senior Reactor Operator Candidates: None
5. Departures during CY 1991:

Mr. Thomas Wright (SR0 license terminated 31 May)

6. There were several changes to the RRFSC during the 19g1 calendar year. Mr.

Thomas O'Brien replaced Mr. Douglas Ashby as Chairman, Safety and Health D partment effective 1 June. CDR Joseph E. DeCicco of the Navy Dosimetry Center sorved as a regular member from 1 January until his retirement on 30 November.

Dr. Samuel Levine was appointed as a special member effective 9 December to assist in revlow of the fuel-follower concrol rod installation. Mr. John Dickson and Ms.-Leslie Moore of the Calvert Marine Museum were appointed as special members on 10 September to perform special projects related to reactor pool cater quality.

The 1991 RRFSC consisted of the following membership in accorcience with

~ AFRRI Reactor Technical Spe::lfications (as of 31- December):

Chairman: Col Nicholas Wanderfield,-USAF (Director's Representative) 4

__o

i Regular Wembers:

Mr. Thc. mas o'Brien (Chairman, Safety and Health Department, AFRRI)

Wr Mark Woore (Reactor Facility Director, AFRRI)

Dr. Marcus Voth (Director, Breazeale Reactor snd Professor of Nuclear i Engineering, Pennsylvania State University)

Wr. Ron Luerson (Safety Directorate, Naval Research Laboratory)

Special Members: j CAPT C.B. Calley, USN (Chairman, Radiation Sources Dept)(Certifled HP)

Dr. Samuel Levine (SHL Nuclear Associates)(Nuclear Engineer-Reactor Specia list)

Mr. John Dickson (Calvert Marine Museum)(Water Quality Specialist)

Ws. Leslie Moore (Calvert Marine Museum)(Water Quality Specialist)  :

Non-voting Observer:

Mr. James Caldwell (EPA, Montgomery County, WD)

Recorder:

Ws. Carol King W:etings of the RRFSC were held:

20 February 1991 27 June 1991 (Subcomittee) 24 September 1991 17 December 1991 4

\

5

SECTION I Changes to the Facility Design Performance Characteristics and Operational Procedures.

Results of Surveillance Tests and Inspections.

4 M

SECTION I Changes to the facility design, performance characteristics, operational procedures, and results from surveillance testing are contained in this section.

A. DESIGN CHANCES:

1, As previously discussed in the Introduction section, three new fuel-follower sta dard control rods, a new air-follower transient control rod, and associated equipment were installed during the year. Also, the control rod drive support structure was modified to permit easier access to the transient rod drive mechanism for ma;at.enance. (Attachment E-1)

2. The pool water gamma activity monitoring system was redesigned to replace vacuum tube technology with a more reliable solid stato system (Attachment E-2). Also, the primary water conductivity readout modules in the control room were replaced by newer models. The sensors and overall systyn design remain the same and no changes to the flow of the primary coolant system were made. (Attachment E-3)
3. The air particulate monitor (CAM) automatic damper closure system was modified by adding an override circuit to allow an operator to open the reactor room air dampers while a CAM is alaraing. This temporary modification would allow controlled venting of the reactor room should a cladding failure occur when it would be necessary to pull air through the absolute filters, for example, during pulse testing of the new fuel-follower control rods. (Attachment E-4)
4. The steady-state timer originally installed on the new reactor console was a count-down timer. That timer was replaced with a count-up timer to facilitate determination of steady-state run lengths since runs are often terminated based on dosimetry measurements and not at predetermined elapsed times. (Attachment E-5)
5. The lead-acid oattery back-up system that previously provided standby power to the criticality monitor and radiation area monitors (RAM.) was replaced by a larger uninterruptible power supply (UPS) necessary for providing more stable current and decreasing maintenance requirements.

Replacement of thc lead-acid batteries also eliminates a hazardous chemical problem and hydrogen explosion hazard.

6.

Hardware and software modifications were made to the control console to install the new low pulso display option and new pulse mode scram timer (Attachment E-6). The pulse timer allows the operator to adjust the amount of time the transient rod remains up during a pulse as per earlier preapproved console installation plan. The new small pulse display option allows high sensitivity pulso data acquisition and a more readable display of pulse characteristics on the console CRT.

B. PERFORMANCE CHARACTERISTICS:

Installation of the fuel-follower control rods and readjustment of the reactor core loading resulted in an increase in the nominal control rod worths by an average of 80.40 (0.0028 Ak/k) and increased the k-excess measurement at infinite water by 10.67 (0.0047 Ak/k). The negative 6

l l

l temparate,o co3fficient of reactivity and the reactor tank constant were not affected. Spwific details of the installation and testing program will be presented in the Startup Report to be submitted as required by the reactor Technical Specifications o early 1992.

C. ADVINISTRATIVE PROCEDURES:

Several changes to the Reactor Administrative Procedures were approved and implementad during the year. A complete set of current administrative procedures is included at Attachment B.

Administrative Procedure Al was changed to clarify the recording requirements when an operator is using over-the-counter medications by requiring operators to check possible side effects in the Physician's Desk Reference, notify the RFD, and log the medication to a Drug Log.

Administrative Procedure A2 was changed to delete the escorted access roster for the exposure room prep area. All persons allowed in the area now receive trainirig in radiation safety to allow them unescorted access.

Administrative Procedure A3 was revised to simplify the 10 CFR 50.59 worksheets to implement recommendations from various NRC inspections.

Administrative Procedure A4 is a completely new procedure developed in conjunction with NRC inspection recomnsdations on inventory procedures for special nuclear matorial.

D. OPERATIONAL PROCEDlRES:

Numerous changes were made to the operational procedures to improve clarity and to account for the new fuel-follower control rods, new low pules display, and other facility modifications. The changes trade during the major revision in May 1991 at listed at Attachment F and subsequent 1991 changes are summarized below. A complete set of current operational procedures is at Attachment B.

1. Procedure 4, Personnel Radiation Protection, was revised to specifically indicate that AFRRI Instruction 6055.8, Occupational Radiation Protection Program, is the radiation protection program followed by the reactor.
2. Procedure 8, Reactor Operations, was extensively revised as follows:
s. Basic: The basic procedure was revised to clarify that retpirator equipment is intended for use only during emergency conditicns, not on a routine basis. Changes were also made to record the SP.0 on-call at the beginning and end of each day or if a change occurs during the operational day instead of on each logbook page. Also, e new designation, physicist in ch-ege (PIC), is added to clarify the individual in charge physienliy present at the reactor.
b. Tab A: The logbook entry chneklist was changed to show that SRO on-call and PIC entries would bs in black ink.
c. Tab C: The nuclear instrumentation setpoints were adjusted for the new reactor control console and to delete references to the obsolete stack particulate monitoring system.

7

d. Tcb 0: The k-execes procedure was revised to change the standard power level for performance from 15 watts to 5 watts. This allows better readability of the power level on the reactor console linear power indicator.

E. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS:

All required maintenance and surveillar.et tems were accomplished as '

required. Malfunctions discovered ar e o. . a; lod in Section IV.

4 s

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p_ _ __..____-_..__m---- - - - - - - - - - -- 2-----

SECTION II Energy Generated by Current Reactor Core and Number of Pulses

$2.00 or Larger.

4 -

SECTION II Energy generated by the reactor core:

MONTH KwHrs JAN 1883.3 FEB 3115.1 MAR 4417.6-APR 5993.6 VAY 1747.0 JUN 4715.9 JUL 1575.3 AUG 2303.6 SEP 2078.8 OCT 1343.0 NOV 1509.4 DEC 1370.9 TOTAL 32053.5 Total energy generated this year: 32053.5 KwHrs Total energy on fuel elements: 734168.6 KwHrs Total energy on FFCRs: 1370.9 KwHrs Total Pulses this year 2 82.00: None-Total Pulses on fuel elements 2 82.00: 4112 Total Pulses on FFCRs 2 82.00: None Total Pulses on fuel elements: 9765 Total Pulses on FFCRs: None a

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3 I

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(

SECTION III Unscheduled Shutdowns l

1

i SECTION IfI Unscheduled Shutdowns:

There were four unscheduled shutdowns during this reporting period.

1. During a 1 MW steady-state run on 7 March, a reactor scram occurred when a high voltage drop in the NP-1000 channel exceeded the trip point of 10%.

Investigation showed that the trip point had been set more conservatively than the required 20% limit. The trip point was readjusted to 20% and operations continued.

2. An "NPP High Voltage Low" scram occurred during a 1 MW steady-state run on 1 August. The voltage circuit was tested and no problem-could be found.

The run was resumed and the reactor operated normally.

3. During a 1 MW steady-state run on 25 October, a reactor scram occurred because of a momentary loss of site electrical power affecting a broad local area. All required safety instrumentation continued to operate on back-up uninterruptible power supplies and normal power was restored in only a few seconds.

i

4. During control rod calibrations on 4 December at a power level of 5 i watts, a reactor scram occurred with a " Timer Scram" indication on the control console CRT screen. At the time, the timer scram was not enabled nor was the steady-state timer running. Testing of both the related console hardware and software did not indicate any cause for the scram and calibrations resumed with no further problems.

10

SECTION IV Safety-Related Corrective Maintenance.

l

SECTION IV Safety-related corrective maintenance:

The following are excerpts from the malfunction logbook during the reporting period. The reason for the corrective action taken, in all cases, was to return the failed (quipment to its proper operational status.

17 Jan 91 Problem: While performing k-excess, the console scrammed when the Auto mode button was pushed. A " Console Scram Button' message appeared on the CRT.

Solution: System was tested but the malfunction could not be made to reoccur. Reactor operated normally thereafter.

18 Jan 91 Problem: While moving the core from position 700 to position 250, a

" Console Scram Button" message appeared on the CRT. The reactor was not at power at the time.

Solution: System was tested but the malfunction could not be made to repeat. Reactor operated normally thereaf ter.

05 Feb 91 Problem: While raising control rods to begin a steady-state run, the transient rod scrammed while being raised. No message appeared on the CRT , no other rods scrammed, and the transient rod drive did not drive down until the DOWN button was pressed.

Solution: System was tested but the malfunction could not be made to reoccur. Reactor operated normally thereaf ter.

04 Mar 91 Problem: While preparing for a steady-state run, an NM-1000 high power trip occurred and would not allow the rods to be raised. The CRT scram message could not be cleared.

Solution: NM-1000 scram clearing sequence was performed, but the scram would not clear until a complete console prestart check sequence was run. Reactor operated normally thereaf ter.

22 Apr 91 Problem: The console would not pass the automatic prestart check sequence due to a '0IS064 Timeout' message on the CRT.

Solution: A defective power supply in the CSC ccmputer expansion chassis was replaced, the system was tested, recalibrated, and returned to full service.

25 Apr 91 Problem: Criticality monitor R-5 alarmed at a steady-state power level of 70 KW, lower than expected.

Solution: The calibration was checked and appeared to have drifted slightly. The monitor was recalibrated with a known source and returned to service. The calibratire was checked periodically to ensure r.o further drift was observed.

06 May 91 Problem: During morning startup procedures, multiple scram messages, DOS error messages, and timecuts occurred.

11

Solution: The problem was traced-to a possibility of excessively high humidity in the control room.-The ventilation system was adjusted and when the humidity was lowered below 804 all systems returned to normal.

15 May 91 Problem: The secondary air particulate monitor (CAM)-exhibited a

~

very erratic trace while the primary CAM showed no fluctuation-in effluent radiation concentration.

Solution: The problem was traced to a faulty chart lamp. The lamp was replaced, the CAM was tested with a known source, and returned to service.

31 May 91 Problem: While raising control rods for k-excess measurements, the three standard control rods scrammed but the rod drives did not drive down. The transient rod did not scram and no message appeared on the CRT.

Solution: A loose relay contact.in the NP/NPP high voltage power supply was secured and the reactor operated normally. ,

10 Jun 91 Problem: A malfunction of the reactor area security system computer caused the computer to not accept keyboard comands and to not' respond to alarms. Reactor Facility Director notified as well as AFRRI security and logistics sections.

Solution: The reactor area backup system was turned on so that reactor-related alarms would appear on security monitors at the front desk which is mar.ned at all times. The priraary system was repaireo by a contractor and returned to full service on 12 July. No loss of security effectiveness occurred while the backup system was operational.

4 17 Jun 91 Problem: During the daily startup procedures, a large volume of air began escaping around the transient red anvil. The RFD was notified.

and an investigation begun. 1 2

Solution: The reactor steff disassembled the transient _ rod _ system and deteralnoa that a teflon seal on the piston 'had worked -loose due to the lack cf a required locking pin in the retaining collar.:This pin was left out by the-contractor during rebuilding.of the drive as part of the fuel-follower control rod installation program. A new taflon seal was machined and a.new locking pin inserted. Cleaned and lubricated piston and barrel. Removed, inspected, and replaced 0-ring seal:at bottom of barrel. Reassembled drive and performed. rod drop tasts and recalibrations as' required._ Notified General Atomics that locking pin had not' bean installed during rebuilding of drive _

assembly at General: Atomics factory.

02 Jul-91-- Problem:-When the console CRT-screen was 'reconfigured after.-_

displaying the graph of'a pulse, the digital linear power readout showed 2 mi'lliwatts while the CRT bargraph had a' scale of-0-1.2-milliwatts.

=_ _

12

Solution: Trua rsector power determin2d to be 2 milliwatts based on independent monitoring of counts out of fission chamber. As soon as the power was raised enough to require a change of bargraph c,cale (to 10-120 milliwatts) the system operated normally. General Atomics notified of possible software problem in CRT reconfiguration program.

05 Jul 91 Problem: After completing automatic console prestert check sequence, a " Pulse Power High' scram message would not clear. While repeating prestart sequence in an attempt to clear the scram, the safe and shim rod drives started to drive up. There was no magnet power to the drives and all control rods remained in the core.

Solution: The console was tested and the malfunction could not be made to reoccur. The reactor operated normally thereafter.

09 Jul 91 Problem: None of the console keyboard function keys used for alternate CRT displays (pulse graph, accumulated operator time, etc) would work.

Solution: The console computer was reinitialized and all keys then worked normally.

16 Jul 91 Problem: While the reactor was secured, an 'NPP High Voltage Low" scram message appeared on the console CRT screen.

Solution: When the key was inserted into the console the scram message cleared. No cause for the scram could be found and the reactor operated normally.

03 Oct 91 Problem: While closing the ERI plug door, the door stopped moving and a loud grinding noise was apparent in the gear reduction box.

Solution: The ER2 gear box was moved to ER1 and the door operated with no more problems. The defective gear box was returned to the factory for repairs.

07 Oct 91 Problem: During prestart checks, a malfunction of'the DRIVE UP microswitch caused the safe rod drive to drive up instead of down when the reactor was scrammed. The magnet released and the rod itself dropped correctly. When the drive continued past the upper limit, the position indicator belt and flexible wire guide broke.

Solution: The drive was removed; belt, wire guide, and DRIVE UP.

microswitch replaced; drive assembly -inspected, tested, and-reinstalled. Rod travel was adjusted, rod drop time checked, and rod-worth curves checked.

NOTE: The reactor was out of service for the annual maintenance shutdown period 12-27 September. No malfunctions occurred during that time. The reactor was also taken out of service for installation of the fuel-follower control rods on 8 November. No malfunctions occurred af ter the reactor returned to steady-state cperation on 16 December.

\

13 I

SECTION V Facility Changes and Changes to Procedures as Described-in the Safety Analysis Report. New Experiments or Teets during the Year.

SECTION V Changes to the facility and procedures as described in the Safety Analysis Rsport and new experiments or tests performed during the year are contained in this section.

A. As previously discussed in the Introduction section, three new fuel-follower standard control rods, a new longer air-follower transient control rod, and related equipment were installed during the year, Also, the control rod drive support structure was modified to permit easier access to the transient rod drive mechanism for maintenance. P ktachment E-1)

B. The pool water gamma activity monitoring system was redesigned to replace vacuum tube technology with a more reliable solid state system (Attachment E-2).

Also, the primary water conductivity readout modules in the control room were replaced by newer models. The sensors and overall system design renain the same and no changes to the flow of the primary coolant system were made. (Attachment E-3)

C. The air particulate monitor (CAM) automatic damper closure system was modified by adding an override circuit to allow an ope:ator to open the reactor room air dampers while a CAM is alarming. This temporary mdification would cllow controlled venting of the reactor room should a cladding failure occur then it would be necessary to pull air through the absolute filters, for example, during pulse testing of the new fuel-follower control rods. (Attachment E-4)

D. The steady-state timer originally installed on the new reactor console was a count-down timer. That timer was replaced with a count-up timer to facilitate determination of steady-state run lengths since runs are often terminated based on dosimetry measurements and not at predetermined elapsed times. (Attachmer.t E-5)

E. The lead-acid battery back-up system that previously provided standby power to the criticality monitor and radiation area monitors (HAMS) was replaced by a larger uninterruptible power supply (UPS) necessary for providing more stable current and decreasing maintenance requirements. Replacement of the lead-acid batteries also eliminates a hazardous chemical problem and hydrogen explosion hazard. The installation of these items was approved along with two other UPSs that were installed in 1990. (See Attachment C-5,1990 Annual Report)

F. Hardware and software modifications were made to the control console to install the new small pulse display option and new pulse mode scram timer (Attachment E-6). The pulse timer allows the operator to adjust the amount of time the transient rod remains up during a pulse as per earlier preapproved console installation plan. The new small pulse display option allows high sensitivity pulse data acquisition and a more readable display of pulse characteristics on the console CRT.

C. There were no new experiments or tests performed during the reporting period that are not encompassed in the Safety Analysis Report. However, all previously cpproved Routine Reactor Authorizations were replaced by a revised set approved by the RRFSC on 24 September 1991. The revised Authorizations do not include any cxperiments that were not in the previous authorizations. The revision served primarily to update format and references and to obtain approval from the current RRFSC membership. A set of the Authorizations is at Attachment D.

14

l - Attachments E-1 through E-6 are a summary of safety evaluations made for changes not submitted to the NRC pursuant to the provisions of 10 CFR 50.59. Each modification was described and qualified-using Administrative Procedure A3, Facility'Wodification. This procedure utilizes a step-by-step process to document the fact that there were no unreviewed safety questions and no changes required to the Technical Specifications.

15

SECTIONS VI through VIII Summary of Radioactive Effluent Released.

Summary of Radiological Surveys.

Exposures Greater Than 25% of 10 CFR Limits.

> 1

SECTION VI Summary of radioactive effluent released:

A. Liquid Waste - The reactor produced no liquid waste during CY 1991.

B. Caseous Waste - There were no particulate discharges in CY 1991. The total Ar-41 discharges in CY 1991 were 13.408 Curies.

On a quarterly basis: Jan - Mar 1991 49E9,9 mci Apr - Jun 1991 3139.5 mCl Jul - Sep 1991 4195.4 mci Oct - Dec 1991 11BS.4 mci C. Solid Waste - All solid material was transferred to the AFRRI byproduct license; none was disposed of under the R-84 license.

SECTION VII Environmental radiological surveys:

A. The environmental sampling of soil, water, and plant growth reported radionuclide levels that were not above the normal range. The radionuclides that were detected were those normally expected f rom natural background and f rom long-term fallout.

B. The environmental monitoring (dosimetry) program reported the following results for CY 1991.

1. The average background of 19 thermoluminescent dosimeters (TLD) located outside a 15 mile radius from the AFRRI site was determined to be 84.40 2 1.69 millirem.
2. The average reading of approximately 30 environmental stations located on the AFRRI site was determined to be -0.45 2 0.31 millirem above background.
3. The single highest environ,wntal station reading was 14.39 6.07 millirem above background. This station is approximately 500 meters from the AFRRI reactor building.
4. The above results are expressed at a 95% confidence level.

C. The in-plant surveys, including analysis of effluent filters,-showed no measurable activity (except as reported in this section) in all areas outside the restricted-access areas.

D. Tnere were no special environmental studies cor. ducted during the year.

SECTION VIII Exposures greater than 25% of 10 CFR 20 limits:

There were no exposures to staff or visitors greater than 25% of 10 CFR 20 limits.

16

ATTACHMENT A 10 CFR 50.59 Analysis and Refueling Plan for Fuel-Follower Control Rods

I Facility Modification.Worksheet 1 10 CFR 50.59 Analysis Install Fuel Follower Control Rods Proposed Change Subinitt:d by: Cape Forsbacka Date 23 OCT 91

. Description of change:

Replace current standard control rods with fuel follower control rods (FFCRs).

A complete description of the FFCRs is included in the attached report " Max-imum Temperature Calculation and operational Characteristics of Fuel Follower Control Rods for the AFRRI TRICA Reactor Facility"(AFRRI TR91-1).

2. Reason for change:

FFCRs will be installed to overcome long werm burnup effects and increase the excess reactivity of the reactor core.

l

3. Verify that the proposed change does not involve a change to the Technical-Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.

This modification DoES INVOLVE A TECHNICAL SPECIFICATIONS CHANGE. See l

attached approval from NRC. Full documentation is in Reactor File 605.06.

Analysis attached? Yes X AFR21 TR91-1 is the safety analysis report regarding FFCR installation and use.

4. The proposed modification constitutes a :hanges in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).

! Procedure X Facility 'X Experiment Revised: 15 May 91 Page 3

Facility Modification Worksheet 1

5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this description need not contain the final SAR wording.

The following sections in the SAR will require modifcation to reflect the installation of FFCRs:

- Section 4.9, Fuel Elements

- Section 4.10, Reactor Control Components

6. For facility modifications, specify what teating is to be performed to assure that the systems involved operate in accordance with their design intent.

See attached plan, "AFRRI TRIGA Refueling Plan".

Revised: 15 May 91 Page 4

Facility Modification Worksheet 1

7. Specify sssociated information.

New drawings are: Attached x (Blueprints of FFcas are in che reactor blueprint file)

Does a drawing need to be sent to Logistics? Yes No x Are training matcrials effected? Yes _x_ No Will any Logs have to be changed? Yes No x Are other procedures effected? Yes No s

List of items affected:

Training Materials:

- New SRO Training Package will need modification to reflect presence of FFCRs in the core.

Maintenance Procedures:

- Annual ehutdown checklist will have to be modified because FFCRs will not be brought out of the pool once they have been irradiated.

- Elongatation and lateral bend will have to be measured on the FFCRs.

8. Create an Action Sheet containing a list of associated work specified in item #

7, attach a copy, and submit another to the RFD.

Action Sheet: Submitted x Not Required

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Reviewed and approved by RFD /A - Date N RRFSC Concurrence Date Revised: 15 May 91 Page 5 E_ __ .. _ . . .

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. p* ~%

UNITED STATES I , ,. . c

!- ,I NUCLEAR REGULATORY COMMISSION o, c WASHINGTON. D.C. 2005 i

s f October 8, 1991 Docket No. 50-170 Colonel George W. Irving, III, BSC, USAF

! Director Armed Forces Radiobiology Research Institute Bethesda,- Maryland 20814-5415

Dear Colonel Irving:

SUBJECT:

ISSUANCE OF AMENDMENT NO.- 21 TO FACILITY OPERATING LICENSE NO. R-84' - ARMED FORCES RADI0 BIOLOGY RESEARCH INSTITUTE (AFRR The Cosmission has issued the enclosed Amendment No. 21Ito Facility Operating 4

i License No. R-84 for the AFRRI TRIGA Research Rear. tor.. The amendment :onsists of changes to the Technical Specifications in respnse to_your_ submitcal of April-30, 1990, as supplemented on December 17, 1990. March 5 -1991, May 17 1991, August 16, 1991, and September 10, 1991.-

The amendment (1) corrects errors in typography and grammar,_ (2) increases the maximum licensed steady state reactor power to 1100 kilowatts, (3) author-m izes installation of fuel follower control rods,-(4)-clarifies the transfer of

! Reactor and (5) allows Facilityoperational Director (RFD) flexibilityresponsibilities in performing in-thesurveillance absence of testingthe RFD,f o l

4 the ventilation _ system for the reactor facility.

Enclosure 2 is a copy of. the related Safety Evaluation supporting Amendment

, No. 21.

! Sincerely,

( ]Y-

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/-

Alexander Adams,- Jr., Project ; anager Non-Power Reactors, Decommissioning and:

Environmental Project Directorate

! Division of Advanced. Reactors

-and Special Projects -.

Office of. Nuclear- Reactor Regulation

Enclosures:

1.- Amendment No. 21 2._ Safety Evaluation-cc w/ enclosures:-

See next page f I

}

., _ -n - . + -n "w,.-s, 4r-t+- ,-e-m e- w - r w-a---ew . x y,a.- - <,c,- w r, w-.y- w er , ,.- --

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AFRR1 TRIGA Refuellag Plan Receiving and Storage of FFCRs Prior to Installation Prior to receiving the FFCRs, the Hot Cell will be prepared for fuel storage. The room will be reasonably free of dust and debris, a gamma detector (criticality monitor) will be in place, a high security lock will be installed, and cradles for holding the fuel will be available.

The FFCRs will enter AFRR1 through the shipping and receiving department (LOGS). A memorandum will be prepared (See attachment 1) to instruct LOGS personnel to not open the containers holding the FFCRs. LOGS personnel will be instructed to call RSDR and SHD immediately upon receipt of the FFCRs. SHD will perform a radiological survey in accordance with HPP-0 3 of external radiation and external contamination of the packaging in the loading dock area prior to moving the FFCR containers to the Hot Cell.

Once the FFCRs have been cleared by SHD to be moved to the Hot Cell, RSDR staff will conduct a series of tests within the Hot Cell to determine if any uranium is on the outer surface of the FFCR claddmg. The testing methodology is similw to series of measurements outlined in 10 CFR 70.39 which deals with the certification of caixation or reference radiation sources.

The following tests will I;c conducted:

1. Dry wipe test. 'lle entire surface of the FFCRs will be wiped witn fiher paper

) using moderate finger pressum. Any radioecovity on the fiker paper will be detenrdned by measunng the radaanon levels using SHD's counting lab.

2. Wet wipe test. The entire surface of the FFCRs will be wiped with filter paper, moistened with water, using moderate finger pressure. Any rachoactivity on the fiher paper will be determmed by measuring the radia:.lon levels using SHD's countag lab.
3. Water Boil Test. De FFCRs will be @y immersed in boihng water for one hour. The residue obtained by evaporating the water will then be moutored for the presence of uranium using the SHD en=*= lab.

If measurable quanuoes of radinactivity are present following any of the above tests, the FFCRs will be thoroughly elemaad and the tests will be iW. In the event that radioactivity is found after repeated tests and c62 g, the shipment will be rW and returned to the manufacturer.

Following the cuumlatinn of mermful tests showing no contamination of the FFCR extenor cladding with uranium or other radioactive enatammanet, the FFCRs may be moved to storage in the Reactor Room.

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l 4 Core Preparation 4

Prior to shutdown and FFCR installation, neutron activation foils / wires in will be used in ERI l with the bare core configuranon at a fixed position one meter from the core. Determine neutron energy spectrw, save data for leer comparison with FFCR loaded core. Next, completely unload all fuel from the reactor core. Following the removal of fuel, the standard control rods I and transient control rod will be rernoved. The rod drive support structure will then be modified as required to allow for the ease of transient rod drive maintenance. Next, the new transient rod .

will be installed.

! The FFCR conrxctmg rods will be ret up to allow for the removal of FFCRs without bringing them out of the reactor pool. Once the FFCRs have been irrwliatad, they will become highly-radioactive due to their fission product inventories, so it will be important to keep them under wut.r for shielding purposes.

i installation of FFCRs into AFRRI TRIGA Reactor Core 4 Measuremaart of the requirej connecting rod lengths will be made by measuring the entire length of the ==aderd rods connecad to the barrel. Correcting for differences between the FFCRs and i the enadard tods, the connacting rod lengths for the FFCRs will be determined. The Wag rods will then be asenchad a the FFCR and the entire nait will be insamilad into the reactor core, i

The connecting rods will then be attached to the banel assembly to complete the imenilmenon of i i the FFCRs.

1

! Following inenilmaina, the FFCRs will centered in their grid locations using the set screws for i the berrel assemblies. The goal is to minimiae any rubbing of the FFCR as it travels up and down. Once the FFCRs have been centered, drop time tests will be j '.. ;4 to insure

! compliance the technical specifications.

Refseling AFRRI TRIGA Ranctor Case A conservative approach to refaeling the teactor core will be taken. These instructions will suppnessent Reactor Oper=einmal Procedure VII. Once the critical loadmg has been achieved, l excess reactivity will be n=elemaad using the transient rod until there is enough excess reactivity  :

to na'wm rod worth curves. Excess reactivity will be detenmined aher each fuel loadmg step I until the optJetional coe83 erdion is achieved. Care will be takaa to ensure that the $5.00 mariamm allowed cIcces reactivity is not **caadad, Following Reactor Operational Proceduit VII, install the thermaramplad fuel alamamen and load i the B-ring. Place neutron sourse in source holder and BF, neutros detectors in F-7 and F-18, 1 hen load *lammaet for grid lacasinas C-1, C-5, C-6, C-9, C-10, C-11 D-14, D-15, E-18, E- l

19,F-22, and F-23. Loading these alanwmet will allow for the asutronic coupling of the FFCRs and the neutmn source, At this point the rods will be withdrawn as described in step 2.a. of the

i r

procedure, and the first subcritical multiplication measurements will be taken.

Complete loadmg the C-Ring, load D-ring elements 2, 6, 8,12, and 18, and repeat the suberitical multiplication measurements. Complete loading the D-ring, and load E-ring elements 5,6,14, 15; perform suberitical mukiplication measurements. Lead E-ring elements 1,2,8,9,10,16, 17, and 24; perform suberitical multiplication preearements. Load E-ring elements 2, 23,7, 11,13, and 20; perform subciiiical multiplication measurements. Lead the following sets of elements and perform suberitical multiplication maanirements after each step (note: this load

(

pattern may be modified to accommodate instrumentation or other items that may obstruct fuel loading): E-4 and E-12; E-21 and E-22; F-1 and F 2; F-3 and F-30; F-4 and F-29; F-5 and F-28; F-6 and F-27; F-26 and F-16; F-15 and F-17; F-14 and F-18; F-13 and F-19; F-12 and F-20; F-il and F 21; F-10; F-9; and F-8. This ined.ng pattern allows for the FFCRs to exercise a high infhwner over the neutron population while the core is still very subcritical.

Core Calibration Once the core has been loaded to the operational excess reactrvity, core calibrations will proceed in the same mannar as followag an annual shutdown. Differential and integral reactivity worth curves will be generated for each rod in core poetions 250,500, and 750. Up to this point all testing has been done at very low powers, so the fission product inventory in the FFCRs will be low. Since the prnhahility that the FFCR cladding may fail is highest shortly after innallahna, a senes of pulses will be perfanned to stress the FFCRs before the fission product inventory has much of a chance of building up. During this pulsing operation the water will be closely momtored for any fission fragments. In the event that fission fragments are found in the pool water, all activities will stop, the NRC and GA will be notified, and the leaking element (s) will be found and isolated.

A thermal power cahbration will be perfonned in the usual mammer. Next the power coefficient of reactivity curve will be generated followed by the reflection coefficient measurements in positions 250, 500, and 750. The neutron energy spectrum experiment in ERI and ER2 will be i@ to ensure that the characesr of the radiation field has not been snochfied.

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1 l Refacilag Checklist i

Pre-shipment 4  !

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1. Clean Hot Cell.

r 2. Inspect Hot Cell, ensure area is reasonably dust free.

l 3. Inemil high security lock on Hot Cell.

4. Ensure radiacon monstor in hot cell is functional. i
5. Prepare fuel cradles, install in Hot Cell.

. 6. Send memoranda to SHD and LOG on receipt of FFCR shipment.

i 3

Receipt of shipment i

7. SHD performs radiological surveys of exterior of package in accordance with their procedmes, i 8. FFCR packages are transferred to Hot Cell.

i 9. Open FFCR shippie4 packages in Hot Cell.

l 10. Place FFCRs in prepared cradles.

Hot cell testing
11. Perform dry wipe test.
12. Perform wet wipe test.

! 13. Perform water boil test.

14. Repest wet wipe test.
15. If all easts me ==ccm==41, FFCRs sany be moved to the Reactor Room for storage.

! Core preparation I

l 16. Use neutron acavation wire set from Reactor Experiments to establish base line neutron j- energy spectrum using LAdasy's set up.

17. Uniond all feel from the core in accordance with Procedure Vll.

I

18. Modify the rod drive support structure.

FFCR instausdos

19. Instau new transiset rod.
20. Fabricate commacong rods for FFCRs.

l 22. Install FFCRs, insure that they are not " bottomed out" with they are fully down.

23. Measure FFCRs agaiset FFCR standard, record resuks.

I -. . . . - - .- .. _ . . . ... - .. - . _ . - . . -

24. Reinstall FFCRs, install rod drive motors, and center FFCRs in their core grid locations.
25. Perform drop time tests to insure compliance with Technical Specifications.

Core refueling

26. Place neutron source in its holder, and place BF, or fission detectors in core grid locanons F-7 and F-18.
27. Load the thermocoupled elements irto core grid locations B-5 and C-6.
28. Complete loading the B-ring and load C-1, C-5, C-6, C-9, C-10, C-11, D-14, D-15, E-18, E-19, F-22, and F-23. Perform subcritical multiplication maniments.
29. Complete loading the C-ring. Im1 D-ring locations 2, 4, 6, 8,10,12,14,16, and 18.

Perform subcritical multipbcation inneenmemst.

30. Complete loading of D tmg, and lend E-ring elements 5,6,14, and 15. Perform subcritical muloplicadon --wements.
31. Load E-ring h 1, 2, 8, 9,10,16,17, and 24. Perform subcritical nmitiplication nietsurtNwnet.
32. Load E-ring elements 15,19, F-22, and F-23, place neutron source into its holder, and perform subetincal multiplication enessurements.
33. Load E-ring elan ==en 2, 23, 7,11,13, and 20. Perform subcritical muinplication mensmements.

'^

34. Load E rieg alaama*= 4 med 12, and perfona subcritical -- ir=ba enemrements.
35. Load E-ring ela===*= 21 and 22, and pwform subcritical andtiplication numenements.
        • 140TE ****

When critical coanguration is achieved, estimase excess reactivity using the transient rod. Fa=*ia=* loading until there is enough excess reactivity to perform a control rod worth measurement

36. Load F-ring enemments 1 sad 2, perform subcritical - ti&r'wlexcess reactivity -

inensurements.

37. Imad F-ring . 3 and 30, perform subentical >==lriplicatinalexcess reactivity mensarements.

J l

1 38. Imd F ring elements 4 and 29, perform suberitical multiplicadon/ excess reactivity l measurements.

39. Load F-ring eicments 5 and 28, perform excess reactivity measurements.
40. Load F-ring elements 6 and 27, perform excess reactivity measurements.
41. Load F ring elements 16 and 26, perforin excess reactivity measurements.

l l 42. Load F-ring elements 15 and 17, perform excess reactivity measurements.

. 43. Imd F-ring elements 14 and 18, paform excess reactivity measuremcots.

  • * *
  • NOTE * *"

Do not exceed $5.00 excess reactivity!

. . . . . . Per.,xin thamal power calibration at 500 to insure proper placement of operational channel. Perform rod worth curves for all rods in position 500 when opermeinmal loading is achieved

44. Imf F-ring elemanet 13 and 19, perform excess reactivity measurements.
45. Imd F-ring ah 12 and 20, perform excess reactivity measurements.
46. Load F-ring elements 11 and 21, p L- excess nectivity inessurements.

i l 47. Load F-ring elements 10, perform excess reactivity measureements.

48. Load F-ring elements 9, perform excess reactivity measurements.
49. Load F-ring ch 8, perform excess rescarvity . . , . . . .

. Core calibration

< $0. Install all cora instrm=nammina into their permanent positions.

i 4

51. Perform rod worth curves for all rods in core poutions 250, 500, and 750.
52. Perform thennel power calibration and s2 safety chambers,
53. Fire $1.10 pense, take pool water sample, manknr for finnan fragsments.
54. Fire $1.20 pulse, take pool water u:.nple, =nnianr for fission fragments.
55. Fire $1.30 pulse, take pool water sample, mandar for fission fragmesus.
56. Fire $1.40 pulse, take pool water sample, mouitor for fission fragments,
57. Fire $1.50 pulse, take pool water sample, monitor for fission fragments.
58. Fire $1.60 pulse, take pool water sample, monitor for fission fragments.
59. Fire $1.70 pulse, take pool water sample, monitor for fission fragments.
60. Fire $1.80 pulse, take pool water sample, monitor for fission fragments.
61. Fire $1.90 pulse, take pool water sample, monitor for fission fragments.
62. Fire $2.00 pulse, take pool water sample, monitor for fission fragments.
63. Fire $2.00 pulse, take pool water sample, monitor for fission fragments.
64. Fire $2.00 pulse, take pool waar sample, maniw for fission fragments.
65. Fire $2.00 pulse, take pool wanar sample, maniw for fission fragments.
66. Fire $2,00 pulse, take pool water sample, maniw for fission fragments.
67. Fire $2.00 pulse, take pool water sample, maaiw fw fission fragmeans.
68. Fire $2.00 pulse, take pool water sample, monitor fw fission fragments.
69. Fire $2,00 pulse, take pool wahr sample, mnaiw for Assion fragments.
70. Genersee power coefficient of reactivity curve, take pool water sample and monitor fw fission fragments.
71. Fire $2.00 palme, take pool water sample, monitor for fission fragments.
72. Opersee reactor at 1.0 MW fw 30 minness, take water semple and amaitor fw fission fragments.
73. Fire $2.00 paise, take pool water sample, monitor for fission fragments.
74. Fire $2.00 puhe, take pool water maple, manieor fw fission fragsments.
75. Repeat neutron energy spectra measurements in ENand ER2.
76. Perform messarements of reflecaor coeffbeients in panisinsu 250,500, and 750.

I i

e ContrW NurnDW sunlect "YtOlE"" 8"*""" N / A Installation of FFCRs T4Te' 23 OCT 91 Action neewired RFD APPROVAL AND CONCURRENCE

-- wome, noum for neeere. iosecnee trosy the requeements eac=grouno ano acton tamen or recommenoeo uust ce su e cen'ty cetaaea to

'dentity tBe 4Chon w@out f9 Course to other sourtee )

1. Request that we plan to install FFCRs -in mid to late November,1991 in a:cordance to fuel reloading plan attached.
2. Training caterials will require an audit to ensure that they will reflect the reactor configuration with FFCRs installed.
3. Formal maintenance procedures involving FFCR surviellance should be developed in the next few months.

I (Conenue on osan bond) l I caenneenwee ae.=en ]

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TECHNICAL REPORT l I

t Maximum Temperature Calculation

and Operational Characteristics

, of Fuel Follower Control Rods for the AFRRiTRIGA Reactor Facility

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Contents IHtroduCtIon...............................................................................................................I General Deserlption of Fuel Follower Control Rods..............................................1 F F CR Muimum Fuel Temperat ure Calculat10n.................................................... 2 Power Density in F F CR F uel Elemen t ................................ ........ .... ......... ...... 3 Maximum Temperature in FFCR Fuel Element............................................. 4 Fuel Temperature in Pulse Mode Operation............................. ...................... 0 F F C R O per atlon al C h ar ac t erls t le s .. ..... .. . . . .... . .. .. . . ........ .... ... . . .. ... . .... .... . .. ... ... . . .. .... . . . . 8 Conclus10n..................................................................................................................9 NeferenCes..................................................................................................................9 Appendix A: Determination of Free Convective Heat Transfer Coefncient.....11 Appendix D: Reactor Core Loading and Unloading ..........................................15 111 1

i INTRODUCTION l Operational requirements of the Armed Forces Radlobiology Research  !

Institute (AFRRI) TRIGA reactor facility necessitate the implementation of

. Fuel follower control rods are like the -

4 fuel standard follower TRIGAcontrol rods control rods as(FFCR's) described in section 4.10.1 of the TRIGA Safety Analysis Report (SAR) except that they have a fuel filled ,

follower ratl.er than an alt or aluminum follower. The primary purpose of the FFCR's is to offset the long term effects of fuel burnup.

requires that The modificationsCode ofofFederal a portionRegulations (s) of a licensed (CFR; Titleas10, facility, Part 50.50)in _the facility described SAR, be documented. with a written safety analysis. The SAR ensures that all safety issues associated with the implementation of FFCR's have been reviewed. This technical report will show that implementing FFCR's will allow the standard control rods to function in their intended purpose and will restore core reactivity economically. FFCR's have been implemented in approximately a dozen TRIGA reactors and have been used for over 20 years without reported failure. -

This report has been submitted to the AFRRI Radiation Facility Safety Committee to ensure that all safety questions have been reviewed before submission to the U.S. Nuclear Regulatory Commis;lon (NRC), as required under 10 CFR 50.50.

GENERAL DESCRIPTION OF FUEL FOLLOWER CONTROL RODS The current AFRRI TRIGA standard control rods were installed in 1964. -

The standard control rod consists of a ser. led aluminum tube (0.065 inch thick) approximately 1.25 Iriches in diameter and 31 inches long. . The upper 15.25 inches of the tube contain a compacted borated graphite rod D with 25 percent free ~ boron or other boron compounds), which functions a (as C neutron absorber or poison. The lower end of the tube contains a 15,25 inch long and 1.125-inch diameter solid aluminum rod called the aluminum follower. The follower functions as a mechanical guide for the control rod as it is withdrawn from or inserted into the reactor core.

The proposed FFCR's differ from the current standard control rods in the following respects:

  • The aluminum cladding is replaced by smooth stainless steel fSS304) cladding with a wall thickness of 0.020 inch. The inner anc outer-diameters are 1.085 inches and 1.125 inches, respectively.-
  • The length of the control rod is increased to 37.75 inches; the absorber

. and fuel follower section are both nominally 15 inches long.

  • The outer diameter of the absorber section and the fuel follower are both 1.085 inches.
  • The fuel follower has a solid zirconium rod as its central core with an outer diameter of 0.225 inch.

l

- _ _ _ . _ . , , x_.._.._ _ _ . _ _ . . _ . , _ .. _ _ _ .. . _ . _ , _ _ . .

The absorber or poison material of the proposed FFCR's is, however, identical to the standard control rods presently installed.

) The fuel contained in the FFCR consists of a fuel moderator element in '

which zirconium hydride is homogeneously mixed with partially enriched

! uranium. The FFCR fuel element contains 12 percent uranium by weight and has a nominal entlehment of 20 percent in the 235U isotob>e. The FFCR fuel element contains about 30.0 grams of ::sU this is 79,o of the 335U loading of a standard AFERI TRIGA fuel element. The nominal hydrogen-to-zirconium ratio in the FFCR fuel element is 1.7 with a range between 1.0 The and 1.7. The FFCit fuel element contains no burnable poison.

stainless steel eladding on the FFCR fuel element has a hardness greater than the aluminum control rod guide tubes, so wearing will occur on the guide tubes rather thsn on the FFCR fuel elements.

FFCR MAXIMUM FUEL TEMPERATURE CALCULATION A thermal hydraulle analysis of the FFCR fuel element to determine the maximum fuel temperature uses the following model:

  • The neutron mean free path for neutrons of all energies is smaller than the diameter of the TRIGA fuel rods, so the reactor must be treated as a heterogeneous reactor. Thus, the active volume of the core is taken to be the volume of fuel contained within the reactor core.
  • The ratio of power in a fuel element with 12 wt % uranium versus 8.5 wt-% uranium is 1.21. This is determined by General Atomics design calculations.8 '
  • The reactor is operating at a steady state power level of 1.0 hiW, and the heat flux across the fuel element is described by Fourier's law of thermal conduction:8 q(r) = -kVT(r) (1) where q" = heat flux at por! tion r k =(r)t hermal conductivity T(r) = temperature at position r For steady state heat transport, the heat production rate and the rate of energy loss due to heat transport are equal. This can be generally expressed as q' ' ' (r) = V* g' ' (r) (2) where q"'(r) = volumetric heat rate (heat production rate) at position r.

Substituting equation equation of thermal (1) ion;into equation (2) yields the time independent conduct q' ' ' (r) = -Y* kVT(r) (3) 2 l

Equation is, thus, the second-order ordinary differential equation that must be so(3) lve d to determine the maximum temperature attained in the fuel portion of the FFCR.

Using this model to determine the maximum fuel temperature divides the analysis into two separate tasks: determining the power density in the FFCR in a D ring grid position and solving equation (3) for the given power density.

Power Density in FFCR Fuel Element The anticipated fuel loading for the AFRRI TRIG A reactor core with FFCR's installed will consist of 77 standard Tn!GA fuel elements and the three FFCR fuel elements. Presuming that the control rods are fully withdrawn to achieve a power level of 1.1 MW, the total active fuel volume will be 30,507.9 cm 3. Thus, the average power density at 1.1 MW will be 36.0 W/cm3.

The maximum fuel temperature is the itnportant parameter, so only the radial variation of the core centerline power density is considered. To determine the maximum power density in the D ring location of the FFCR fuel element, the following calculations are made:

For the AFRRI TRIGA, the radial and axial peak-to average power ratios are 1,55 and 1.30, respectively.3 Thus the maximum power density (heat rate) will be q' ' ', = (1.55) (1.go) q ' ' ',y, (4)

= 72.4 W/cm To determine q"'om relative to q'" it is useful to compute a scalin6 factor from the gro,s,'b variation of TI[e,rmal neutron flux in the radial normalized radial flux distribution for the AFRRI TRIGA core i)sdirection best (the represented by a Dessel function of the first kind of order zero:

ftherm " O (2.405r R } (}

e where R* = 21.78 cm, the extrapolated core radius r = 11.99 cm, radial position of D ring element and the Bessel function scaling factor is J, (1.3240) = 0.6074.

The power density for thn D ring is thus computed to be 3

q' ' D-ring = (0.6074) q' ' ' = 44,0 W/cm (6)

Because the FFCR fuel element differs from the standard fuel element in concentration of uranium, the power density in an FFCR fuel element is 3

greater than the power density in a standard fuel eleinent by a factor of 1.21.8 Taking the above scaling factor into account, the power density of an FFCR fuel element is found to be q' ' 'FFCR = (1.21) q' ' D-ring (7)

= 53.2 W/cir.3 Note that the calculation of q'" takes into account the most limitin condition for power peaking in a [Ov/o fuel element versus an 8.5 w o fue1 element.

As developed in equation

, q"'r is still considerab

  • less as (7)determi'neS in equation 4. less than the theoretical roachmaximum would have q"'lso a accounted for the reduce (d ) volume in conservative an FFCR fuelapp!cment.

e Maximurn Ternperature in FFCR Fuel Eleinent Equation 3 azimuthal (sy)mmetry (see Figure 1): takes the following form for cylindrical geometry with 1[dr.dr)+q=0 r

k (8)

The boundary conditions required that constrain equation (8) are as follows:

dT

- = 0 at r - Rg and T = Tg at r = Rg (9)

J

, Te\ ,

'~

+Tj E .Tg d,

Figure 1. Cross section of FFCR fuel element.

4 i

Integrating equation (S) twice yields a general solution of the form 2

T (r) = -q' ' ' -

+ Cg in(r) + C2 (10)

Applying the boundary conditions to solve for the temperature difference between the outer edge of the rircoelum rod and the inner surface of the cladding, qRI 2 O g= (11) 2k g C2=Ti- (In(Rg ) - )

2 qR R R I

Tg - T, s [( )2 - 21n ( ) - 1) (12) f 1 i To account for the transfer of heat from the fuel through the cladding to i the coolant, we must consider the heat conduction between the inner and and the heat conduction from the outer outer surfacesurfaces of the cladding of thetocladding, the coo qaTa,nt, qWe make the assumption that no heat is produced in the cladding or SN3e coolant, so the heat conduction and gnaa*

from the conduction The heat outer surface of the leaving the fuel fuel,is qua,iven g bymust be q,i,,

equal to 2 2 ,,'

qfuel = r(R,2 -R g )g4,,, = r((R,/R g)2 - 1)LR g q (13) where L = length of fuel element dT 2rke L(T -T) 9 clad = -kch (14) dr clad R +c In( )

R, qfluid = W T, - Tf ) = 2r(R, + c)Lh(T c -T)f Note that the area, A, used in computing qa,, is the logarithmic mean area of the cladding. Recall that 9 fuel " 9 clad " 9 fluid (15)

So we can solve for the temperature differences in the above equations in terms of q"':

5 i

2

+c R Ri R

l T -T c = In (- ) ((J)2 , g) q,,,

R R 2 I (16)

R E)2 , g) Ri

  • q,,,

1 T -

C f = (R,+c)h ((R g 2

with equation (12 Adding expression equations for the max (16)imum temperature in the FFCR. ,) and solving for T, gives us the I

l q,R1 2. R R (E)2 - 21n( 2) - 1 (17)

Tg=Tf+ 4k g ,R g R ,

1 2 +c qRI R 1 R 1 l +

[(R 2)2 -1)(k, In( )+ )

2 g R, h(R, + c.) ,

where T i = maximum fuel temperature T, = bulk coolant te'nperature R' = 1.38 cm (radius of FFCR fuel element) eR= 0.280

= 0.051 cmcm (radius (cladding of Zr rod))

thickness k = 0.18 W/cm 'C k* = 0.138 W em.'C(thermal conductivity of UZril)'

g h = 1.339 W cm .'C (thermal conc'uctivity of SC304)8 3

q"' = 53.2 W/ m 3 (from(free equation (7)) convective heat transfer coefficien Note that the free convective heat transfer coefficient, h, was an l experimentally derived quantity. The method by which h was determined is presented in Appendix A, Solving equation (17) using a volumetric her.t rate of $3.2 W/cm3 and a bulk water temperature of 48.0'C (the conditions at which h was determined) yields a maximum fuel temperature of 210.2'C.

The maximum temperature achieved in the FFCR is nearly 180*C less than the normal temperature of 390'C in a standard fuel element in the B-ring during a 1.0 hiW steady-state power operation.

Fuel Temperature in Pulse Mode Operation The Nordhelm Fuchs model predicts the maximum fuel temperature achieved in a pulse mode operation 5

The fundamental assumptions of this model are as fohows:

  • The neutron flux in the reactw la separated into a spatial component (shape factor) and a time-dependent component (amplitude factor), such that f(r,t) = vn(t)t(r) (18) l l

6

, l where v = neutron velocity n(t{ = neutron density (amplitude factor), proportional to power f(r, = shape factor The shape factor is assumed to remitin constant during a pulse. This is called the point reactor model.

  • The production of delayed neutrons and the effects of source neutrons l are neglected,
  • The pulse from a thermodynamic standpoint is adiabatic, so dT

= Kn(t) (19) dt

$vhere T = fuel temperature K = reciprocal of heat capacity From the first and second assumptions we can write the time dependent neutron density as dn p-/

- a n (20) dt E where R = mean lifetime of neutrons in the reactor p = delayed neutron fraction p = reactivity ,

To account for a step insertion of reactivity, we can write -

p*p y -

uaT (21) where a = negative of the temperature coefficient of reactivity pg= step-Insertion of reactivity Taking the derivative with respect to time of the above equation and

! substituting the result from equation (18),

= -aKn (22) i dt Applying the chain rule to equation (20) yields d" . . (# ~ 4) dp aK1 (23) 1 7 1

i e, - g .

-9.+w. e. 9 --*w , w-., r. , , --w w- .--. y- .--- .y

i I

l l

Integrating equation .(23) and solving for the constant of integration gives us the result

-i 1

n= ((p, - /)2 , (p , p)2] (24) i 2aK1 f The pulse is terminated when n becomes negligibly small. This occurs when p=2p-p, (25)

Equation (25 gives us the condition for the total energy release from the pulse, which ) manifests itself as a temperature rise in the fuel element w it is substituted into equation (21).

I 2(po - #)

core, ave

" (0) a Calculations by General Atomics show that a complete core of 12 w/o fuel erature coefficient of reactivity, a, that is- 75% of the would for value have ana8.5 tempw /o fueled core.1 The value for a is taken to be

- 80.0118/' C. (This value was experimentally verified,,w,dh a series of 23 of pulses ranging C-within from $1.30 8.5% of the to $2.00-published thatvalue. resultedThe in an effect averafe o add a'm,,g,Niree 80.0128/'FFCR's -would, however, have a -neg)ligible effect on the overall 12 w/o-temperature wellicient of reactivity for the entire core.1

Applying the Nordhelm Fuchs model for self limiting power exultsions, we ,

can determine the maximum average increase in temperature for the entire core using equation (26). For a maximum allowed reactor- pulse with a i

84.00 step insertion of reactivity, the maximum attained average temperature rise is calculated to be 333'C. Applying the power peaking factors from- the

! previous section, the maximum calculated temperature rise in ' an FFCR

the temperature rise for _ an FFCR for a $4.00 pulse would be 1.48'AT is calculated to IEc "83'C. - Assuming an initial temperature of 25'C, the maximum temperature value would be 518'C. Note that even in the.

limiting case, neither the technical specification safety limit of 1000'C nor j the limiting safety systene setting of 600'C is violated.

FFCR OPERATIONAL CHARACTEPISTICS FFCR's are a standard design offered as a stock item by General Atomics and have been used in several TRIGA- reactors for over 20 years. FFCR's are currently implemented in approximately a dozen TRIGA reactors.- There '

has been no reported evidence of fuel failure as a result of FFCR use in the

' United States, . The operational issues- to be resolved are the effects of-l burnup' on the - FFCR and- the influence of FFCR's on the temperature -

coefficients of reactivity, shut-down margin, and rod worth.

FFCR- control rod worth curves-.will be generated the same way that standard control rod worth curves are, Since the poison section of the FFCR will be the same as that of the currently installed standard control 8

l _. . _ _ _ _ _ .

- . .- - .= _ . - -

rods, the worth of the poison section of the FFCR will be that of the currently installed control rods. Measurements made by AFRH1 reactor staff of control rod worth of the currently installed standard control rods yielded a nominal rod worth of $1.90. The fuel follower is expected to add at least (0.70 of reactivity' when the control rod is fully withdrawn, so the total rod worth for an FFCR is estimated to be $2.00. The transient control rod and its follower have been measured and have a total nominal worth of

$ 4.01. The shutdown margin, as established by ANS/ ANSI 15.1, is computed as follows:

Total rod worth $11.81 k , y , ,, (maximum) - $ 5.00

$ 0.S1 Worth of TRANS rod - $ 4,01 Shutdown margin 8 2.80 The shutdown margin with the most reactive control rod removed from the reactor is $2.80--well in excess of $0.50 minimum allowed value.

Once operational rod worth curves are established and power monitoring channels have been calibrated by the thermal power calibration method, power coefficient of reactivity curves will be generated. The issues regarding the measurement of shut-down margin and excess reactivity are addressed in Appendix D. Reactor Core Loading and Unloading.

Structural changes in the FFCR's will be monitored on an annual basis as part of the annual shutdown and maintenance. Specific effects to be monitored are the elongation and lateral bending of the fuel. FFCR fuel elements that have an elongation greater than 0.100 inch or a lateral bend greater than 0.0025 inch will be removed from service.

CONCLUSION The analysis in this report shows that installing FFCR's in the AFRRI TRIGA reactor core will not result in an unsafe condition or violation of technical specifications. - The primary parameter of interest, the maximum fuel temperature, was computed to be 210'C in the limiting case for steady-state operation and 518'C in the limiting case for pulse operation.

Operational issues regarding maximum excess reactivity, shutdown margin, and burnup have also been addressed, and it has been determined that sufficient surveillance capabilities exist to prevent any unsafe or illegal condition.

REFERENCES 1.

General Atomics, letter to M. Moore on fuel follower control rods, 28 October 1988.

2. Ei-Wakil, M. M., Nuclear Heat Transport, The American Nuclear Society, Lagrange Park, IL,1978.

9

t e

1 4 3. Defense Atom t e Support Agency, ATRRl/USAEC Facility License R 94, Complete teith Applications and Amendments, Dethesda, MD,1902.

4. Wallace. W. P., and Simnad, M. T., Afetallurgy of TRIGA Fuct Elements, G A 1949, General Atomics, San Diego, CA,1901. l i
5. Hetrick, D. L., Dynamics of Nuclear Reactors The University of Chicago Press, 1971.

! 6. DNA Contract DNA001-80 R 0030 to General Atomics for fuel follower j control rod construction.

) 7. Jaluria, Y., Natural Convection Neat and Afass Trans/c r, Pergamon Press, 1980. j i

)

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APPENDIX A: DETEllS11 NATION OF FRFE CONVECTIVE IIEAT TRANSFER COEFFICIENT Introduction We can measure the bulk water temperature within the AFRRI TRIGA core to determine the average free convective heat transfer coefficient of the cooling water. This experiment involves inserting a temperature measuring t probe between the D. and C-ring fuel elements while the reactor is operating at a steady state power level of 1.0 h!W and measuring the water temperature at various axial positions. Once the bulk water temperature has been determined, Newton's law of cooling can be used to calculate the average free convective heat transfer coefTicient.

Experimental Apparatus and Procedure The equipment used in this experiment consists of two approximately 18 foot lengths of chromal alumel thermocouple wires fused together at one end, encased in a 10-foot long, 0.375 inch diameter aluminum (Al) tube, and the thermocouple display readout on the AFRRI computerized reactor control

  • console (Figure A 1).

The potential difference generated at the thermocouple junction as the water is heated by the reactor is amplified and- displayed by the thermocouple circuitry in the AFRRI computerized reactor control console. The thermocouple is initially inserted into the core to correspond to position I.

The thermocouple resides in each region for several minutes to allow it to attain thermal equilibrium. Once thermal equilibrium is attained, ten temperature readings are taken at 10 second intervals. After each temperature measurement, the thermocouple is withdrawn to the next position, and the temperature measuring procedure is repeated.

Figure A 2 shows that the temperature is messured in five axial' positions:

(I 3 inches below midpoint-(14 inches of thermocouple wire inserted- into midpoint in axial dimension; (III halfway between midpoint th)e core); (II) f graphite slug; (IV) at top of ) fuel re and bottom o .

above top of fuel region.

Thermocouple junction To new conde

= G -.

Al tube s

Figure A 1. Experimental apparatusc 11

~ ~

4 s

- 4 IV 4 lit

- <;~ n 4 i H;0 Figure A 2. Axial measuring points.

Safety Considerations There are two safety considerations associated with this experiment: radiation streaming and an unintentional positive change in reactivity if the thermocouple wires are rapidly withdrawn from the reactor core while it is at power. Radiation streaming is avoided by flooding the aluminum tube with water and bending the tube so that it is at an angle not normal to the top of the core. The thermocouple wire displaces only 0.043 imh3 of water when it is fully inserted in the core, so using the void coefficient of reactivity, the thermocouple wire represents a negative reactivity insertion of only 0.001 cents. If we were to estimate conservatively that the thermocouple wire had the same neutron-absorbing properties of a control rod, the maximum negative reactivity would be only 0.01 cents. Thus, there is no possibility of a reactivity accident associated with the apparatus used in this experiment.

Data Table A-1 summarizes the data gathered during a 1.0 MW steady-state run of the AFRRI TRIG A reactor. The variation in the temperature measurements is most likely due to variance in the radial position of the temperature probe in the channel.

Table A 1. Bulk Water Temperature at Each Axlal Position in the AFRRI TRIGA Reactor Core Axlal inlet Measured core position temp ('C) bulk water temp ('C)

I 22 72.9 11 24 65.0 111 25 48.6-IV 26 51.6 V 27 59.7 l

l l

12

Analysis and Conclusion The purpose of this experiment is to determine the bulk water temperature within the core shroud; thus, it is the iowest measured value of the water temperature that is sought. Figure A 3 illustrates the temperature variation within a cooling channel.

Table A 1 shows that the measured value that most closely represents the bulk water temperature within the core shroud is 48.0'C.

The free convective heat transfer coefficient, h, is found by solving equation (8) for boundary conditions given by a standard TRIGA fuel element.

Equation (A-1) gives the solution in terms of h.

,,, 2 , , .-1 (Tg-T)-g ( )2 - 21n( )-1 1 4k g . rg rt 1 r In( , + c,)

h=( ) -

(A-1) qr(

r, + c, E T 2,[ c o 2 , rg ,

where Ti = measured fuel temperature at 1.0 MW T, = mea.sured bulk coolant temperature in the core r, = fuel outer radius,1.810 cm ri = fuel inner radius, 0.229 cm e, = cladding thickness, 0.051 cm k, = thermal conductivity of fuel, 0.18 W/cm.'C k' = thermal conductivity of clad, 0.138 W/cm *C qi " = volumetric heat rate.

l'he measured fuel temperature in the B ring at 1.0 MW steady state power level is 300'C, and the calculated volumetric heat rate is 65.9 W/cm 3.

Using the measured value 3 of the bulk coolant temperature of 48.0'C yields a value of 1.339 W/C'-cm for the free convective heat transfer coefUcient.

a

'\ r 5 W i

1 w

,p .-

Figure A 3. Ternperature variation within a cooling channel.

13

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)

Newton's law of cooling expresses the linear relationship between the heat transfer rate, Q, and the temperature difference between the clad surface temperature, T,, and the bulk water temperature, T,, as Q = hA(Tf - T,)

where h is the overall convective heat transfer coefficient and A is the area of the fuel element.' The value of h determined for the AFRR1 TRIGA is unique in that it takes into account the flow configuration, fluid propertice, and the dimensions of the fuel elements. Assuming that the dependence of Q on the temperature difference T - T h, computed using data from B rin,g ents elem,is roughly with their higherl'near, heat transfer then the value o rate and local temperature difference, will be close to the value for h for D-ring positions, t

t l 14 1

r. , ,

c , - , . . , - - ,... - , . , - - , - - - . , .. - n-,,. ~ ,, - - - , - - ~cn,- ,,,v-+, ~

-.= -. .-- - - --

APPENDIX D: REACTOR COltE LOADING AND UNLOADING General Loading and unloading of the reactor core shall be performed under the supervision of the Reactor Facility Director or the Reactor Operations Supervisor.

I Specific

1. Setup
a. Ensure that at least one nuclear instrumentation channel is operational,
b. Ensure that the source is in the core.
c. Ensure that an operator monitors the reactor console during all fuel movements.
d. Check new FFCR's before insertion into the core; this includes cleaning, visual inspection, and length and bow measurements,
e. Install all control rods.
f. If irradiated fuel elements are to be removed unshielded from the pool, obtain a Special Work Permit Department SHD do not remove f(uel elements with aSWP) fromhistory the Safety and Hea (greater than (1 KW)in the previous 2 weeks from the reactorpower pool.
2. Core Loading
a. After each step of fuel movement perform the following:

(1) Record detector readings.

(2) Withdraw control rods 50%; record readings.

(3) Withdraw control rods 100%; record readings.

(4) Calculate 1/M.

(5) Plot 1/M versus number of elements (and total mass of U).

(6) Predict critical loading.

(7) Insert ALL rods; continue to next step.

b. Load elements in the following order:

(1) Load the B ring and C-ring thermocouple elements.

15

( l 1

1 i

(2) Connect thermocouple outputs to reactor control console display.

] (3) Install any other thermocouple elements. 1 an d C ring elements (total of 18 standardelements (4) Complete plus 3 loading FFCR's).of D ,

! (5) Load D ring (total of 33 standard elements plus 3 FFCR's).

) (0) Load the following E ring elements in order:

10,17,18, 20, 6, 8, 9,10 (total of 41 standard elements plus 3

~

FFCR's).

) (7) Complete the E-ring by loading the following elements in order:

15, 21, 11, 5, 14, 22, 4, 12, 13, 1 (total of .51 standard
elements plus 3 FFCR's). 1

. (8) Load the following F-ring elements in two elements per step until criticality is achieved, using the following loading order:

! 22, 23, 24, 21, 20, 25, 26,-27, 28, 29, 30, 1, 2, 3,_4, 5, 19, 18, ~

4 17, 16, 15, 14, 13, 6, 12, 7, 11, 8, 10, 9. i I Once criticelity has been achieved, perform control rod worth measurements *

! at core position 500 by rod drop technique. Calculate shutdown margin

{ (SDM):

l SDM = total control rod worth - K,y,,, - TRANS rod worth d

(9) Load core to $2.00 excess reactivity by loading two elements per step using the loading order in instruction 8.

1

(10) Verify control rod worth using rod drop techniques; calculate '

i SDM, i

(11) Load the core to achieve' a K that- will allow calibration of-f the TR.ANS rod based on the last available*Torth- curve of the- TRANS rod

- (approximately $4.00). Calculate the reactivity value of each element as it '

is added.

(12) Calibrate all control rods.

(13) Calculate SDM.

(14). Estimate K,,,,,,- with a fully loade'd core (must not _ exceed

[

15 instructi(on)8, and recalibrate all control reds. Load core to fully operational Calculate SDM.

16

_ _ _ _ _ . . _ , . _ - _ . , , - - _ . _ - . _ . _ . . _ . . ~ _ . ~ . _ . _ , , _ - , _ _ . _ - . _ . . _ . . .

- - - - , - - . - . - - . - - . - . - . ~ . . - . -. . . _ - . . . - _ . - -

!f necess(ary.16) Recalibrate Adjustallthe corerods.

control loadingCalculate patternSDM.

to meet operations requirements

3. Core Unloading
a. Unload the reactor core starting with the F. ring and ending with the D ring,
b. Remove the fuel elen.;nts individually from the reactor core, identify them by serial nutaber, and place them in the fuel storage racks or a shipping cask. '
c. If elements are to be loaded into a shipping cask, clean the cask completely, and check for rad'ological contatnination before placing the cask in or near the pool. Load cask in accordance with procedures specific to the cask,
d. Once the cask is loaded, perform an air sample and survey; check temperature and pressure inside cask, if necessary.
e. If elementa are placed in temporary storage away from core monitoring, ensure that criticality monitoring in accordance with 10 CFR 70 is in place.

l 17

l l

l l

1 1

ATTACHMENT B t 1

Current Reactor Administrative and Operational Procedures i

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i PROCEDURES For The AFRRI Reactor Facility O C' 4< < >

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TABLE of CONTENTS r

ADMINISTRATIVE PROCEDURE Revised Date Procedure A1 Fitness For Duty ........ ............................. ............................. 15 May 91 Procedure A2 Personnel Passage nrough The Prep Arca.......................... 15 May 91 Procedure A3 Facility Modification ........................................ .................... 15 M ay 91 Procedure A4 Special Nuclear Material Accountability............................... 15 May 91 OPERATIONAL PROCEDURE Proce dute 0 Procedute Changes .................................................................... 15 M ay 91 Procedure 1 - Conduct of Experiments ........................................................... 15 M ay 91 Procedure 1, TAB A Reactor Exposure Room Entry .................................... 15 May 91 Procedure 1, TAB B - Core Experiment Tube (CET) ..................................... 15 May 91 Procedute 1, TAB C - Extractor System........................................................... 15 M ay 91

, Procedure 1, TAB D Pneumatic Transfer System (PTS) ............................... 15 May 91 Procedure 1, TAB E - In. Pool /In Core Experiments ........................................ 15 May 91 Procedute 2 - Reactor Staff Training ............................................................... 15 M ay 91 Procedute 3 Maintenance Procedures ........................................................... 15 May 91 Procedure 4 Perscanel Radiation Protection ............... . ............................... 15 Nov 91 Procedure 5 Physical Security ....................................................................... 15 M ay 91 Procedure 6 - Eme rgency Procedure s ............................................................. 15 M ay 91 Procedute 7 Core Loading and Unloading .......... ........................................ 15 May 91

1 TA

-.BLE of CONTENTS Revised Date.  ;

1 Proce d u re 8 R eactor C pe r ations ......... ............. ............ ...... . ...... ................ 15 Nov 91 Procedure 8, TAB A - legbook Entry Checklist .... ........... . ..... . ..................... 22 Jul 91 Procedure 8, TAB B Daily Operational Startup Checklist ..... .... ................ . 15 hiay 91 i Procedure 8, TAB B1 Daily Safety Checklist .............................. .. .............. 15 May 91 Procedure 8, TAB C Nuclear instrumentation Set Points ........ .....................15 hiay 91 Proce d u r e 8, TAB D K.Exce ss ..... .............. . .................... ....................... ... . . 15 M a y 91 Procedure 8, TAB E Steady State Operation ....................... .......................... 15 May 91 Procedure 8, TAB F1 Square Wave Operation (Subcritical) ......................... 15 May 91 Procedure 8, TAB F2 Square Wave Operation (Critical) ............................... 15 May 91 i Procedute 8, TAB G1 - Pulse Operation (Critica1) ........................................... 15 May 91 -

Procedure 8. TAB O2 Pulse Operation (Subcritical) ..................................... 15 May 91 Procedure 8, TAB H Weekly Operational Instrument Checklist ................... 15 May 91 Procedure 8, Tab I Daily Operational Shutdown Checklist ........................... 15 May 91 Procedure 9 Reactor Roorn S afe ty ................................................................. 15 M ay 91 Procedure 10 - Staek Gas Monitor Procedure ................................................. 15 May 91 Procedure 11 - Air Particulate Monitor Procedure .......................................... 15 May 91 f

I

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Lhocedure A1 ADMINISTil

, .. . - T.iVE![i m _ __ PROC _SDUREE I i

t This procedure has been approved by the Reactor Facility Director '

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Reactor Facility D fctor FT KaGG u

' DatI Reviewed by the Reactor Staff fd/S h 2-9 New4/

Wright ' Date L (s.,w 24 n1(

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Spence Date

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FITNESS FOR DUTY l OENERAL The AFRRI Reactor Facility is a drug free work place. The use of illicit drugs by any RSDR staff member is prohibited. Personnel using over the counter or prescription drugs which cause drowsiness or otherwise alter one's state of con-sciousness will not be permitted to operate the AFRRI TRIGA reactor. In addi-tion, reactor operators, operators in training, and management will be monitored for attitude and behavioral changes that may impact an individual's reliability.

SPECIFIC

1. RSDR staff members shall participate in drug free awareness programs spon-sored by AFRRI. Military and civilian staff members shall submit to drug screen-ing programs conducted by their respective services. If a staff member's drug screening test yields a positive result, that staff member shall not be permitted to operate the reactor pending verification of the test. The Reactor Facility Director (RFD) is required to ensure that the cutoff levels for alcohol or controlled sub-stances as established in 10 CFR 26 are not exceeded by NRC licensed personnel.

Any staff member determined to be a drug user will be terminated.

2. Personnel are instructed to inform their physician of their job description and requirements prior to being issued a prescript!on medication. They are instructed to inquire about any medication side effects expected and the physician's opinion regarding interference with safe job performance. This information shall be re-layed to RFD as soon as possible.

Personnel are encouraged to minimize their use of non prescription over-the-counter drugs for self medication purposes. Specifically, sedatives, cough and cold preparations, appetite suppressants, and pain relievers have central nervous system side effects. If these medications are used in any quantity, an operator must in.

form the RFD or ROS and~be relieved from operating on that day or until any side effects have resolved once the medication has been discontinued.

Personnel are instructed to read the information in the Physician's Desk Refer-ence (PDR) concerning medication they are taking. If the PDR indicates that the medication will adversely affect an operator's abil!ty to safely perform his/her du.

ties, he/she must inform the RFD or Reactor Operations Supervisor (ROS) that he/she must be relieved from operating on that day.

Revised: 15 May 91 Page 1

i 1

3. The RFD shall continuously monitor the reliability of individuals under his/her i command by the following criteria-l
  • Any court martialor civilconviction of a serious nature. Minor traffic l violations are not a consideration.
  • Negligence or delinquency in duty performance.
  • Significant mental or character traits, or aberrant behavior, sustained by medical authority, that might affect the reliable performance of duties.
  • Behavior patterns that s'..ow or suggest a contemptuous attitude toward the law or regulations _

+

Drug abuse or alcohol misuse.

  • Poor attitude, lack of motivation toward assigned duties, or financial it.

responsibility.

The RFD will be observed by his superiors to ensure his/her adherence to reliability criteria. Individuals who exhibit any of the listed behavions or actions will be removed from licensed activities, t

Revised: 15 May 91. Page 2.

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r This procedure 'has been ap roved .y the Reactor Facility Director

)41b/to L MJu#t/ '

Rea6tn't Facil'ity DVgftor Dale I h i A /l .2 7 t% v t/ l Chairnfan, Safety and Health Department D(ite l Reviewed by the Reactor Staff i

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> y.a ,ggy y;ppy i PERSONNEL PASSAGE THROUGH THE PREP AREA GENERAL Access to the Reactor Prep Area is lirnited to personnel who are granted access in accordance with the reactor physic 6 security plan and Operational Procedure 1.

The Reactor Facility Director is responsible for mairaining an unescorted access roster for the Reactor Prep Area and for providing a Prep Area briefing to all persons listed on that roster. This administrative procedure does not recapitulate the operational procedure. Rather,it presents specific guidelines for Reactor Prep Area pascage for individuals who are authorized access.

SPECIFIC There are three specific conditions when Reactor Prep Area passage is considered appropriate.

1. ROUTINE PASSAOE:(EXPOSURE ROOM DOORS CLOSED)
a. Perronnel who are authorized unescorted access to the Reactor Prep Area may pass through the Prep Area but are required to radiologically frisk themselves when exiting the Prep Area,
b. Personnel who are being escorted through the Reactor Prep Area may pass through the Prep Area with their escort. Both individuals must radiologically frisk themselves when exiting the Prep Area.
c. Only appropriate personal dosimetry that has been issued by SHD or AFRRI Security is requi;ed for routine passage through the Reactor Prep Area. There is no requirement to wear a pocket chamber in addition to the AFRRI issued TLD for routine passage.
2. CONTROLLED PASSAGE:(EXPOSURE ROOM DOOR OPEN)
a. Only personnel associated with the experiment / operation being performed are normally authorized access to the Reactor Prep Area during an exposure room operdng. These personnel will be required to wear the AFRRI TLD dosimeter (issued with their AFRRI badge), pocket chamber (if dose rate at face of door is 2 mr/hr or more), and the following special dosimeters if they enter the exposure room:

AFRR1 wrist dosimeter or finger ring dosimeter.

Revised: 15 May 91 Page i

4 All personnel who enter an exposure room will log their pocket chamber reading ia the pocket chamber log prior to :ntering the room for the first time that day and will enter the final pocke chamber reading following their exit from the exposure room at the end of the day. Each individual who enters an exposure room is responsible for monitoring his accumulated dose throughout the day to ensure the he/she doc not exceed the AFRRI daily exposure limits of 50 mR/ day or 100 mR/ week. Extremity dosimetry is required only if work is to be performed on an experimental array or within 1 meter of the core projection.

b. Personnel authorized unescorted access to the Reactor Prep Area or personnel being escorted through the Reactor Prep Area may pass through the Prep Area when an exposure room is open with permission from the Reactor Staff person in charge during the opening if the following guidelines are met:

(1) The person desiring passage must stop just inside the Prep Area door upon seeing that an exposure room door is open and request permission from the Reactor Staff member in charge before proceeding. At that time, the Reactor Staff and Safety Sta" members monitoring the Exposure Room opening will determine if the radiation level at the uutside entrance to the exposure room in direct line of sight with the core projection is less than or equal to 2 mR/hr. If this reading is less than or equal to 2 mR/hr, the Reactor Staff Member may grant passage permission. If the reading is greater than 2 mR/hr, passage will be denied.

(M3 .rsonnel who pass through the Reactor Prep Area must tadiobgically frisk. themselves before exiting the Prep Area. There is no requhement for these individuals to wear a pocket chamber just to pass through the Prep Area.

3. OPEN PASSAGE:(NON-ROUTINE)

The prep area may be opened for passage by personnel traveling between buildings at AFRRI when maintenance is being performed on the normal connecting a hallw;y. This is not a routine occurrence and warrants written approval from the Reactor Facility Director with concurrence from the Chairman, Safety and Health Department. In addition, the Prep Area must be monitored at all times by appropriately train:d personnel. Prior to the opening of the Reactor Prep Area for open passage, the Safety and Health Department shall conduct a radiological survey of the area and cenify that no radiation areas exist within the Prep Area and that the non painted areas of the Prep Area floor are free of contamination.

There is no requirement for personnel who pass through the Prep Area to wear pocket chambers or frisk themselves during periods of open passage. Finally, open passage will be suspended during exposure room openings.

~

Revised: 15 hfay 91 Page 2 )

^M!psygggaftnygg gth W & .

gr>ropdLQA3 IThis procedure has been approved by the Reactor Facility Director

,I _

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[. Reviewed by the Reactor Staff

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$1 ' , Wriglit Date

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Date G e dr y\l (l {lf

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FACILITY MODIFICATION GENERAL i Changes to the Reactor Facility and operational procedures must comply with requirements specified in the Reactor License, and 10 CFR 50.59. It is required '

that modifications to the facility or procedures as described in the Safety Analysis Report (SAR) be documented with a written safety analysis. Under 10 CFR 50.59, a licensee may make changes to the facility provided there are no changes made to the Technical Specifications, there are no unreviewed safety questions, and that a proper safety analysis is carried out, documented, and reviewed.

Applicability:

The Facility Modification Procedure applies to proposed facility changes or changes in the operating procedures.

The referenced procedure will not cover routine replacement of parts or components with equivalent parts or components.

DESCRIPTION This administrative procedure consists of these instructions, the Facility Modifi-cation Worksheet Guide, and two worksheets to facilitate.a 10 CFR 50.59 review of modifications and to determine if a detailed safety analysis is necessary. The in-structions in the Facility Modification Worksheet Guide are used determine which worksheet must be completed for the modification. One of three conclusions re-garding the proposed facilMy modification will be reached:

1. The modification requires prior approval or a license amendment from the USNRC,
2. The modification may be made according to the provisions of 10 CFR 50.59(a)(1) (Facility Modification Worksheet # 1), or
3. The modification does not require a 10 CFR 50.59 safety analysis (Facility Modification _ Worksheet # 2).

l Revised: 15 May 91 - Page1 E _ _ . . _. ._ ~ ._ _ -.- _ _ .- . - .

Facility Modification -Worksheet Guide.

1. Technical Specification Change: If the proposed modification requires a changes in the Technical Specifications, a license amendment is required prior to-making the change. NRC approvalis required; do not implement the change with-out this approval.
2. Unreviewed Safety Question: If an unreviewed safety question is created by the proposed change as defined in 10 CFR 50.59(a)(2) such that the change increases the probability of occurrence or severity of an accident described in the SAR, can malfunction in a manner that can cause an accident of a different type than de-scribed in the SAR or can decrease safety margins as d: fined in Technical Specifica-tions, then NRC approvalis required. Do not implement the change without this approval.
2. If the proposed modification makes a change in the facility ss described in the SAR or changes a' procedure as described in the SAR, the change can be per-formed under a 10 CFR-50.59 analysis with a safety review, if there are no un-reviewed safety issues (10 CFR 50.59(a)(2)). The change may be made following a review by the RRFSC. _Go to Facility Modification Worksheet # 1.

~

3. If the proposed modification does not make a change to the facility 'as described in the SAR or to a procedure as described in the SAR and does not pose an un-reviewed safety issue, a 10 CFR 50.59 analysis is not required. Go to Facility Mod-ification Worksheet # 2.

i Revised: 15 May 91 Page 2 l

. - . . ~ ~ _ . . _ . -_ _ . _ . - _ _ ... . _ _ . . _

d Facility Modification Worksheet 1 10 CFR 50.59 Analysis Proposed Change ,

Subrnitted by: Date

1. Description of change:
2. Reason for change:
3. Verify that the proposed change does not involve a change to the Technical Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.

Analysis attached? Yes _

4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).

Procedure Facility Experiment Revised: 15 May 91 Page 3 -

l

Facility Modification Worksheet 1

5. Specify what sections of the SAR are applicable, In general terms describe the.

necessary updates to the SAR. Note that this description need not contain the final SAR wording.

i

-l

6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.

1 i

Revised: 15 May 91 Page 4

Facility Modification Worksheet 1

7. Specify associated information.

New drawings are: Attached Not required Does a drawing need to be sent to Logistics? Yes No Are training materials effected? Yes No Will any_ Logs have to be changed? Yes No Are othe; procedures effected? Yes No List of items affected:

8. Create an Action Sheet containing a list of associated work specified in item #-

7, attach a copy, and submit another to the RFD.

Action Sheet: Submitted Not Required-1 Reviewed and approved by RFD. D ate -

RRFSC Concurrence - Date Revised: 15 May 91 Page 5-

. . - . - . _ . . _ . _ . . - - _ . c. . - . _ . . . - . . _ . _ . _ . - 1._ . . . _ . . . . . _ _ . _ _ . . _ u .. ..

1 1

Facility Modification Worksheet 2 No 10 CFR 50.59 Analysis Required

Proposed Change Modification to: Procedure Facility Experiment Submitted by: Date
1. Description of change:

f i

2. Verify that tre proposed change does not involve a change to the Technical Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved safety question as defined in 10
CFR 50.59(a)(2).
3. If change involves a facility modification, attach a drawing if appropriate. If-structural facility drawings need updating, forward a copy of changes necessary to
Logistics.
4. Determine what other procedures, logs, or training material may be affected 4

and record below.

5. List of associated drawings, procedures, logs, or other materials to be changed

! 6. Create an Action Sheet containing the list of associated work specified above, attach a copy, and submit it to the RFD.-

Action Sheet: Submitted Not Required Reviewed and approved by RFD Date RRFSC Notified Date i

Revised: 15 May 91 Page 6

Y h D N Y N $ N $$ $ NIIdMNAONEI[$$D 1$2E E ?dddf5!$$5 0 SI This procedure has een approved by , e Reactor Facility Director

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Reactor Facility Dirpo/ g W /[Ll } ate

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Chair" man,Sifety and H6aftli Department Ifate Special Concurranc . , , , . ,

/7#7 91

/ Director, I Dat Reviewed by the Reactor Staff Wni/ r/zy/7/

l Wright Date Lke Mf4l 21

-Fo acka. Date

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Date Date Date Reviewed by RRFSC 24 S EP 1991 Date

, - . _ - - , ~ , , , -

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wa cws stig g g g g ] g'g4g SPECIAL NUCLEAR MATERIAL ACCOUNTABILITY

1. PURPOSE This procedure prescribes the responsibility for all work involving the receipt, use, and disposal of Special Nuclear Material (SNM) at AFRRI.
2. REFERENCES a.10 CFR 70 Domestic Licensing of Special Nuclear Material b.10 CFR 73 Physical Protection of Plants and Materials c.10 CFR 74 Material Control and Accounting of Special Nuclear Material
d. NUREG/BR-0006/7 Instructions for Material Balance Reports and Transac-tion Reports
e. AFRRI Health Physics Procedures
3. RESPONSIBILITY
a. Overall responsibility for all SNM at AFRRI rests with the Reactor Facility Director (RFD). The RFD designates individuals to be in charge of supervising the use of and accountability for specific types of SNM.
b. Members of the reactor staff will use SNM procured under license R-84 in ac-cordance with the reactor Technical Specifications and Rerctor Operating Proce-dures. Users of SNM listed under the byproduct material license will be

! governed by Health Physics Procedures (HPP 5-X series). The RFD will desig-l nate a member of the reactor staff (Inventory Officer) to conduct inventories and maintain accountability records for all SNM as required by Reference d,- i this procedure, and HPPs.

c. The inventory officer will receive training in fuel records,-accountability pro-l cedures, and inventory procedures prior to assignment as inventory officer. This will be documented on the appointment memo. (Appendix A)
4. RECEIPT AND TRANSFER OF SNM Receipt of SNM at AFRRI and transfer af SNM from AFRRI are regulated by 10 CFR and 49 CFR and performed in accordance with HPPs 0-3 and 0-5. When any SNM is received or transferred, the responsible inventory officer will ensure that DOE /NRC Form 741 is submitted as required by Reference d and 10 CFR 74.15.

Revised: 15 May 91 Page1

The RFD must authorize any receipt or transfer of SNM Approval is also re-quired from the RXSC for any SNM on the byproduct material license.

5. ACCOUNTABILITY AND INVENTORY PROCEDURES
a. Safety and Health Department (SHD) personnel will conduct leak tests as re-quired by HPP 5-2. A physical inventory of all sealed SNM sources on the by-product license will be conducted as of March 31 and September 30 of each year. This semiannual inventory will be conducted to ensure annual compliance with 10 CFR 70. The specific inventory method will follow ALARA considera-tions. Reports required by References c and d (DOE /NRC Form 742/742C) will be prepared and submitted within 30 days (10 CFR 74.13).
b. A current reactor fuelinventory sheet (Appendix B) will be maintained. It will give the core loading (core position) for all in-core fuel elements and FFCRs. During each inventory , the inventory officer will sign the inventory sheet verifying loading. This document will be updated if changes occur and at least 10% of the in core elements will be verified by serial number as to location annually.
c. Inventory procedures for reactor fuel elements and fuel-follower control rods I

will be as follows (conducted as of March 31 and September 30 each year by the l inventory officer): (1) Perform K-excess to verify the core is loaded as per the fuel invea-tory sheet. (2) Sign the fuel inventory sheet to certify correct loading and attach to summary accountability sheet. (Appendix C) (3) Inventory by serial numbers fuel elements in stcrage except for en-tombed damaged elements (previously recorded) (4) Account for any fuel elements or FFCRs received or shipped out during the reporting period (5) Ensure number of elements and FFCRs present matches the num-ber of fuel element record sheets and that Fuel Element Record (AFRRI Form 255) sheets are current (6) Transfer data from fuel inventory sheet to summary accountability shect and verify correct total number of fuel elements and FFCRs (7) Calculate fuel burnup in grams based on kilowatt-hours generated during the reporting period and subtract from the previous material balance Revised: 15 May 91 Page 2

(8) Complete and submit DOE /NRC Forms 741, 742, and 742C using procedures in Reference d. (9) File the fuelinventory sh:et, the summary accountability sheet, and copies of all DOE /NRC forms in reactor files,

d. Fission chambers and other SNM on the R-84 license will be inventoried by piece count in conjunction with the fuel inventory, recorded on the summary ac-countability sheet, and reports submitted as in (8) above. All smear testing /han-dling will be in accordance with Reference e.
6. RECORDS All records and reports pertaining to any SNM held at AFRRI will be maintained in reactor facility files by the inventory officer. Duplicate receipt and transfer files, radiological survey documents and leak test results will be maintained in SHD files.
7. REPORTS OF SNM INCIDENTS Reports of loss, theft, attempted theft, or diversion of SNM will be submitted as re-quired by 10 CFR 74.11. Other incidents involving SNM may be reportable under reactor Tec'mical Specifications or 10 CFR 20.402-20.405
8. SIGNATURE AUTHORITY All reports and other outgoing correspondence dealing with SNM will be signed only by the Reactor Facility Director or his designee (see paragraph 3a).

f Revised: 15 May 91 Page 3

Appendix A INVENTORY OFFICER is appointed Inventory Officer to perform duties as per Ad-ministrative Procedure A4. He has received proper training in accountability proce-dures, inventory procedures, and fuel records. Reactor Facility Director D ate Revised: 15 May 91 Page 4

APPENDIX B - FUEL INVENTORY WITHDRAWN FROM PUBLIC DISCLOSURE I l l l 1 l l I I 1 l l l l 

 . .. .                         ~ . .. .. ... .-            . - ..                       --

I f Appendix C \ l Summary Accountability Sheet Number Weight Fuel Elements In Core Fuel Elements In Tank Storage Total Fuel Elements Fuel Follower Control Rod (Location) Chambers Foils Sources l L i l 1 i Revised: 15 May 91 Page 11'

f. 1 

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R$Ryj{Q $$ g [{ j{j{pjjjjjj p g } ] hygfggjyygffgy{ g m 0', PROCEDURE CHANGES GENERAL This establishes procedures for permanently or temporarily changing reactor oper-ating procedures. SPECIFIC

1. Permanent changes are made by revising the entire procedure. The revised pro-cedures will be approved by the Reactor Facility Director (RFD) and reviewed by-the Reactor and Radiation Facility Safe _ty Committee (RRFSC).
2. Temporary changes may be made in pen and ink on the current procedure when initialed by the RFD or Reactor Operations Supervisor (ROS). These changes must be documented, approved by the RFD, and reviewed by the RRFSC at the next scheduled meeting.
3. Temporary procedures may be established by the RFD for a specific situation.
4. All procedures (temporary or permanent) will have an signature block for all op-erators and reactor staff members. Operators will review new or modified proce-dures and sign the signature block prior to operating the reactor console When the block is completed,' the procedure will be placed in the Reactor Procedures Binder and kept available for operator review.
5. All changes will be accomplished under the following guidelines:
a. The change will result in no decrease in the safety of the actions being addressed.
b. The change will result in no decrease in the efficiency of procedure performance,
c. The change will not affect the ability of the procedure to perform its intended function.

l l Revised: 15 May 91 Page 1 

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7. All changes will be staffed to the following:
a. Chairman, Safety and Health Department (SHD) ,
b. Reactor and Radiation Facility Safety Committee (RRFSC)
c. AFRRI TRIGA Reactor Facility staff Procedures that may affect other areas such as building changes, security, etc.. will .

be staffed to the appropriate office (s) prior to routing to Chairman, SHD

  • NOTE: Procedural changes that _do not deal specifically with health physics pro-

', cedures or radiation safety issues need not be staffed through Chairman, SHD. 4 4 a i 5 e i e Revised: 15 May 91 Page 2 i

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1. All experiments will be observed during irradiation with the exception of CET experiments or those in which no movement is possible. The closed-circuit televi-sions (CCTVs) in the exposure rooms and over the r actor pool can be used to meet this requirement.
2. All experiments will be set up so as to preclude movement unless the experiment apparatus is designed for movement (such as rotators, etc.).
3. All animal experimental arrays (shielding)in the exposure rooms that are set up on wooden tables or on styrofoam will have an absorbent pad placed over the wood or styrofoam surface to prevent sanitation problems from the animal waste.
4. The Reactor Staff will conduct a thorough inspection of all experiments to de-termine that no unauthorized materials are irradiated.
5. ALAR A will be practiced during all experiments.

SPECIFIC

1. A Reactor Use Request (RUR)is required for any experiments included under authorizations outlined in the Technical Specifications, section 6.4.2.a. and section 6.4.2.b.. RURs are not required for reactor parameter measurements as outlined in the Technical Specifications, section 6.4.2.c. Any experiment performed by the reac-tor staff (except T.S. 6.4.2.a) for the purpose of determining information to be used to enhance, define, ascertain, or develop methods to expand the performance of the reactor will not require an RUR. Facility tours will not require an RUR but will require verbal approval of either the Reactor Facility Director (RFD) or the Reactor Operations Supervisor (ROS).
2. Expr.timent Review (Processing of RURs):
a. Check the RUR for completeness (Section I should be filled out).
b. Forward the RUR to the Radiation Biophysics Department, Opera-tional Dosimetry Division (BRPD) if dosimetry support is required.
c. Forward it to the Safety & Health Department (SHD) for radiologi-cal safety coordination.

Revised: 15 May 91 Page 1

l

d. Check experiment protocol against reactor cuthorizations.
e. Fill in Section II of RUR with special instructions, as appropriate, Assign an RUR sequence number. Write in estimated or measured ex.

periment worth and the core position of the experiment facility to be t utilized in the appropriate block (lower left hand corner of form),

f. Have the RFD, acting RFD, ROS or acting ROS review and sign the form.
g. Ensure the RUR form is placed in the reactor control room prior to the irradiation.
3. Conduct of Experiments. Perform setup and irradiation of experiments in ac-cordance with the following procedures:
a. Exposure Room Entry - TAB A.
b. Core Experiment Tube (CET) - TAB B.
c. Extractor System - TAB C.
d. Pneumatic Transfer System (PTS) - TAB D.
e. In pool /In core Experiments - TAB E.
4. Complete the RUR by filling out Section IV with the appropriate information.
5. Attach form to clipboard in the control room.

l i (. x I l: L Revised: 15 May 91 . Page 2. L

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0(W M Q f & g &.nyewyg:4QQMQantTfDyg*3 REACTOR EXPOSURE ROOM ENTRY

1. REFERENCES a.10 CFR 20, " Standards for Protection Against Radiation"
b. USNRC licenses: R-84,19-08330 02
c. AFRRI Instruction 6055.8B
2. GENERAL
a. PURPOSE: This procedure specifies all safety and security procedures for ac-tivities involving entry into the AFRRI TRIGA Reactor exposure rooms, currently designated exposure rooms 1 and 2 (rooms 1123 and 1122).
b. AUTHORIZED ENTRY: Both AFRRI picture badge and U-badge person-nel, may enter a reactor exposure room under the supervisica of the Reactor Facil-ity Director (RFD) or his representative. Visiting personnel (V badge) require special authorization by both the Chairman, Safety and Health Department (SHD) and RFD to enter either exposure room. In general, permission to enter the expo-sure rooms will be granted personnel whose duties require such entry; however, per-mission may be denied to personnel for serious or repeated safety or security violations, or for safety reasons emanating from conditions in the exposure rooms.

All personnel who are granted unescorted access to the prep area or warm storage will receive a special prep area safety briefing prior to being granted access. Only personnel who have been granted unescorted access will be given the combination to the prep area . The RFD is responsible for maintaining a roster in the prep area tw personnel who have been granted unescorted access. Other personnel re-quiring unescorted access to the prep area or warm storage for a specific purpose or time period may be granted special access in writing by the RFD with concur-rence of SHD. However, these personnel who are granted special access from the RFD will not be given the combination to the prep area,

c. ER ENTRY INSTRUCTIONS - All personnel will:

(1) Know the Reactor staff representative is in charge of all operations in the prep area. Obtain permission to enter either exposure room from the Reactor staff representative. (2) Wear AFRRI TLD whole body badge and pocket dosimeter. Revised: 15 May 91 Page 1

l l l (3) Wear wrist or finger dosimeter if work is to be performed on an experi-mental array or within one meter of the core projection. (4) Wear booties, eye protection, gloves and coat. (5) Check and log pocket dosimeter reading on log in prep area prior to entry. (6) Familiarize themselves with approximate radiation levels in the room, based on radiological surveys performed and data obtained by SHD. (7) Ensure that all materials removed from the exposure room are properly la. beled and entered on the exposure room entry log (AFRRI FORM 130) and the ac-tivated materials control log. (8) Glove and coat requirements may be waived by the Reactor Representa-tive on an individual basis for personnel who will not be touching anything in the exposure room. There must be a specific reason for waiving;such requirements.

d. DEPARTURE FROM REACTOR EXPOSURE ROOM ENTRY PROCE-DURES: Any departure from the following procedures will require a special work permit (SWP). Exceeding any radiation dose limits will require a written justi-fication from the supervisor of the research project which must be approved by the Head, SHD.
3. SHD EXPOSURE ROOM SURVEY
a. EXPOSURE ROOM CAM: Prior to opening either exposure room, the re-spective CAM must read 2000 cpm or less, above background. If the CAM reads 2000 cpm or greater above background, change the filter of the C. M. If 10 min-utes or more have elapsed since the end of the reactor run, the door may be opened to the first step to facilitate radioeffluent clearance in the room._ Then check the CAM after 1 minute and if the reading is below 2000 cpm above back-ground, proceed with the exposure room opening. Ifit is above, change the filter and wait another minute. If the CAM alarms during;or immediately after a run,--

change- the filter and reset the CAM.

b. DOSE RATE AT FACE OF DOOR: If the dose rate at the face of the plug-door in the direct line of sight of the reactor tank bulge reads greater than 100 ,

mr/hr, the door will be closed sufficiently to preclude access. The plug door will be reopened upon agreement of the- SHD and RFD representatives for reevaluation of radiation levels.

c. DOSE LEVELS IN ROOM: Exposure rates will be measured at specific sites-in the rooms. These measurements will be given to both the reactor representative-and the personnel entering the room. Additionally the readings will be entered in -

Revised: 15 May 91 Page 2:

I the room entrance log (AFRRI FORM 130) and kept in the prep area. The levels will be measured at: (1) The reactor door face in the direct line of sight of the reactor tank bulge (2) At the contamination line in the entrance of the room (3) The middle of the room (4) One meter from the tank wall or shield (5) Contact with the tank wall or shield (6) The area (s) where individual (s) will be working for an extended pe-riod of time and any other place deemed necessary by the SHD or reac-tor representatives,

d. ROUTINE ENTRY: Entry is routinely permitted only when the maximum reading in any occupiable area is 1 R/h or less. Entry may be permitted iflevels are 1-5 R/h, but no work will be permitted in fields over 1 R/h. If personel are working in a specific area for an extended period of time , the dose rate in that area will be measured.

(1) Readings over 100 mR/hr (closed window) will be reported to the Reactor representative by the SHD monitor. These areas of the exposure room will be iden-tified to the Reactor representative and entry personnel. When appropriate, after consultation with the SHD and Reactor representatives, stay times will be assigned for entry personnel. All personnel entering will be assigned a stay time if they will be working in the high radiation area. AFRRI limits of 100 mR/ week and 50 mR/ day are to be used as the basis of stay time determinations. (2) All exposure room entries will be checked by the SHD monitor for compli-ance with radiation safety aspects of applicable Reactor Use Requests (RURs). If not, non compliance will b reported to the RFD and to SHD.

e. FILLING OUT THE SURVEY OF EXPOSURE ROOM OPENING LOG:

The exposure room opening log sheet must be filled out completely for each open-ing of an exposure room. - Care must be taken to fill out each blank on the entry log sheet. If a section is not applicable to the particular opening, N/A should be filled in the blank.

4. NON MONITORED OPENING:
a. Personel may enter the exposure rooms without a SHD monitor present if ALL the following conditions hold:

i (1) The reactor has not been to power in that ER since the last survey. Revised: 15 May 91 Page 3

I (2) Survey meter readings at the door indicate safe entry conditions (should be less than 1 mR/hr). (3) The ER CAM should be observed, and its reading (net) should be - less than 200 cpm above background.

b. An entry will be made in the exposure room log by a reactor staff member, with a note that the survey has been waived.
c. SHD must be notified if any radioactive materials or equipment are to be re-moved from the prep area.
5. PERSONNEL PROTECTION PROCEDURES
a. Dosimetry and protective clothing requirements are given in paragraph 2.c, ER Entry instructions.
b. Entry is permitted only after the SHD monitor has completed the survey and reported results to those about to enter (excluding non monitored openings - Refer-ence Paragraph 4, above).
c. All personnel shall record initial dosimeter reading in the prep. area dosimeter log prior to entering the exposure room for the first time each day.' Personnel shall read dosimeters when leaving the exposure room and record a final dosimeter read-ing in the prep area log at completion of daily operations. Net doses over 10 mrem -

must be reported to the SHD Monitor,

d. Protective clothing will be removed in such a way as not to contaminate
            " clean" areas by items from " dirty" areas,
e. All personnel will " frisk" themselves before leaving the prep area.
6. SPECIFIC ACTIONS TO OPEN EXPOSURE ROOM DOORS-
a. Turn up exposure room lights (this can be waived for experiment needs).
b. - Check plug door tracks for obstructions; ensure all obstacles are clear of the door (including ropes).
c. Ensure that only authorized personnel (see 2.b.) are present in the reactor prep area during: exposure room openings.
d. When facility safety interlocks and opening procedures have been satisfied, in-sert key into exposure room door key panel and open door. DO NOT LEAVE KEY IN LOCK UNATTENDED.

t Revised: 15 May 91 - Page 4 '.

e. Open door in accordance with entry procedures. Ensure all required data is logged in entry log.
f. Ensure that individuals who will be moving lead, bismuth, or other heavy ma-terials are wearing steel toed shoes,
g. Limit exposure times of all personnel entering the exposure rooms based on the results of the radiation survey.
7. ACTIVATED MATERIALS
a. PLACING MATERIAL IN EXPOSURE ROOM: Before placing any equip-ment or material in an exposure room for irradiation the following will be ob-served:

(1) Equipment tagged as AFRRI property: a DF must be sent to both the RFD and the AFRRI property officer. The DF must state that the equipment is knowingly being irradiated and therefore request that it be removed from the prop-erty books. It must also state that should the material remain byproduct material after a reasonable amount of time it will be disposed of as radioactive waste. The-DF must contain all nomenclature as well as an . adequate description of the equip-ment in order for it to be identified.on the property book. (2) Non-tagged AFRRI equipment or material (to be returned): a DF. or state-ment on the reactor RUR must be sent to the RFD giving the~ kinds and. amounts of byproduct material expected to be produced (that is the material that the experi-menter wishes to be returned) and a copy or number of their radionuclide authori-zation number. The DF or RUR statement must be specific and contain an - accurate description of the material being exposed (converted to. byproduct). Other information will be required from personnel before any material is allowed ,

to be removed from the prep or warm storage areas-(see next section of this proce-dure 7.b. and .7.c.)

(3) Non-tagged equipment or material (not to be returneil): A DF or state i ment on the-RUR that the experimenter understands that byproduct material pro- . duced as a result of their irradiations will be disposed of as radioactive waste, and additionally any' material not specifically requested to be held, will be disposed of-as radioactive waste in.the next shipment, i (4) Non-AFRRI owned equipment / material: A signed memorandum from the - responsible property. owner that they understand- that byproduct materials gener-l ated in excess of their license will be disposed.of as radioactive waste unless prior-arrangements have been made with the reactor /SHD staffs for storage. Any mate-Revised: 15 May 91 Page5 

                ._            _.         _ _ . _ _ _ _ _ _ . . . _    _.--. _ . _          _   m. __ .         _ _ . . . _ _

                                                                                 +

l l rial not removed within a reasonable amount of time will automatically be dis-posed of as radioactive waste.

b. SURVEY OF MATERIALS COMING OUT OF EXPOSURE ROOM (1) All materialleaving the exposure rooms :..ust be surveyed for activation or .

contaminaticn Survey meter readings will be used to determine dose levels. Smear surveys may be used, if the SHD representative deems them necessary. All materi- ,i , als will be labeled appropriately in accordance with HPP 0-2. I (2) All special equipment that has been activated such as chambers,;otators, motors, meters, etc., will be stored under the control of the reactor license or the AFRRI byproduct license in warm storage-or the prep area. Removal'of items ' from the prep area will only be allowed in accordance with HPP 31  ;

c. DISPOSITION OF ACTIVATED MATERIALS All activated or contaminated materials will be under the control of the reac -

tor license while such materials remain in the reactor controlled area. Removal of j any radioactive materials from the reactor controlled area will be done in accor- i dance with HPP 31.

8. COMPLETION OF ENTRY '

i

a. The Reactor Staff Representative will check to see that all personnel have left -l the exposure room before the plug door is closed. In the event that the warning horn in either exposure room is disconnected, for testing-or experiment require-ments, the exposure room plug door shall not be closed wil at least two (2) li- l censed reactor operators visually inspect the room to ensure .that no personnel l

remain in the room. To ensure compliance with the reactor Technical Specifica-tions, the names of these licensed operators present at the exposure room closing shall be entered into the reactor operations logbook and on AFRRI FORM 130.- At the completion of the test or experiment, the warning horn shall be reconnected and tested. All actions regarding the warning horn shallin entered in GREEN ink-in the reactor: operations- logbook.

b. The SHD monitor will not leave the area while the plug door is open without  !

notifying the Reactor -Staff Representative.

c. Lock the exposure room door control panel; reset lights, if appropriate.
d. Resecure the prep area on departure.-

Revised: 15 May 91 - ' Page 6

i OPF.RATjoNAL PROCEDURil ' hocedure ID Thb procedure has be n approved by he Reactor Facility Director l N hev j . 

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e Reviewed by RRFSC 24 SEP 1991 Date a $ Elar jr js .

OPERAT10NAL PROCEDURL Procedure 1, TAB D CORE EXPERIMENT TUBE (CET) } GENERAL ALARA principles will be practiced during CET operations. SPECIFIC

1. CET Insertion into the core:
a. Ensure a reactor operator is monitoring the reactor console,
b. Ensure a reactor staff member is present in the reactor room,
c. Establish comrnunications between the reactor room and the control room.
d. Test fuel handling tool for operability,
c. Lower the fuel-handling tool into the core and attach to the desired element. Notify operator on th; console that you are prepared to lift fuel element. When acknowledged, lift fuel element from the core.
f. Transfer ehment to a storage rack location and secure fuel handling tool cable,
g. Loosen CET bracket bolts and remove CET bracket.

h While the CET is held down, cut cable tie from around the CET.

i. Lift CET from the storage rack location and transfer to the reactor carriap.
j. Notify the ;onsole operator that you are prepared to lower the CET into the core; when acknowledged, lowe: 'he CET into the core ensur-ing that it is properly seated in the lower grid plate,
k. With a downward pressure on the CET to keep it seated, secure the CET bracket with the two bolts.
1. Ensure appropriate entries are made in the operations logbook and the fuel book, and that the reactor core pegboard is updated.
2. Irradiation:
a. Clean the rabbit (s) using alcohol and water.

Revised: 15 May 91 Page1

b. Once clean, do NOT handle the rabbit except with gloves, Kimwipes, or handling tools.
c. Ensure that the rabbit cap is secured tightly,
d. The rabbit may,be inserted into the CET either before the run begins or while the reactor is at power .
c. If the rabbit is to be inserted while the reactor is at power, then after notifying the reactor operator on console, drop or lower the rabbit into the core WITH THE CAP UP. Ensure that you spend a minimum amount of time in the vicinity of the carriage. Do NOT lower the rab.

bit with the extractor tool while at power.

f. Complete irradiation and shut down reactor.
g. Ensure appropriate entries are made in the operations logbook r.nd the CET logbook.
3. Rabbit Retrievals:
a. Ensure that a reactor staff member and a Safety & Health Depart-ment (SHD) monitor are present in the reactor room. Any staff mem-ber who will be handling the sample following the irradiation may be required to wear a pocket chamber and appropriate extremity dosime-try depending on the radiation levels of the irradiated sample. If the CET is in the core, a reactor operator must monitor the console during the retrieval,
b. Test the rabbit extractor (" fishing pole") for operability,
c. Insert the extractor head mechanism into the CET and reel out cable until you reach the low end indicator painted on the cable.-
d. Drop the extractor head firmly on the rabbit,
c. Ensure the SHD monitor has a teletector positioned near the CET top to monitor the rabbit.
f. If the CET is in the core, notify the reactor operator that the rabbit is being pulled and continue when acknowledged.
g. Ree) in the cable at a rate commensurate with radiation levels; lower the rabbit back into the CET if the rabbit is excessively hot.
h. Stop when the rabbit is visible at top of CET; have SHD take an ac-curate radiction reading,
i. If radiation ievels are acceptable, swing rabbit away from carriage i

and have another individual grab it with a handling tool 'If the radia-tion levels are not acceptable, lower the rabbit back into the CET. The rabbit will again be withdrawn for teevaluation of radiation levels Revised: 15 May 91 Page 2

I when the SHD and RFD representatives concur on w aceptable radi. i ) ation level in accordance with ALARA and mission nquirements. 

                                                                                                                                                                                   >~

j, Release extractor head and detach rabbit from head.

k. Unless working with the rabbit. er radiation levels are low, store rab.

[ 1 bit or irradiated material in a lead pig or storage cask, i 1. .%Iake appropriate entries in the operations and CET logbooks. l

4. CET Removal from Core:

.1

a. Complete steps la c above.
b. Loosen the CET bracket bolts while holding the CET downt remove the CET bracket. i
c. Notify the console operator that you are prepared to remove the CET from the reactor core.
o. When acknowledged, transfer the CET to the storage rack, ensuring 1 that it is kept as low in the water as possible. ,
e. Secure the CET with cable ties.

4

f. Secure the CET bracket with the two bolts.
g. Remove the fuel element from the stork;. rack and transfer to core. '

Notify the console operator and receive acknowledgment prior to inser-tion of the element .

h. Ensure the element is properly seated in the lower grid plate by lis-tening for the " double clicks".
i. Make appropriate entries in the operations and fuel logbooks and up.  ;

date the reactor core pegboard. ' t ( Revised: 15 May 91 PageS 

 - _ , , . . , . . , . . . _ . , - . . . _ , . - - , . - -.,_..,,,,_-,,,,..__...--,._..,,.,m..-_

_ r.,.. , _ , _ . _ . , _ . , . . _ _ . . < - . = . . . . . . - . .m. 

                                   '                            ~

OPERATIONA1. PROCEDtJRE': s Procedure 1C r This procedure has been approved by the Reactor Facility Director AA b hdhmIY- . _ _ ? hdif Re- 

                     .or Faiility Dirpoff            D            j!       [ D' ate '

4 A /YA 

               /Chaiffnan,      ~
                                              ~

At% 9/ Safety ahttrealth Departrnent' Date Reviewed by the Reactor Staff 1R4+ S*. 2 9 t/ Wrisht Date I!.4<. . , s c3-7n=il Date 

                              &Ceort e / [- '24A9/

For; acka Date A ch M/k,fff Spence Date 

                               %%M               J9 a%91 Laugher'y 0              D ate
                                  /Atm %s % , h mui91 uyen ' "           D ate' L. htt+w 34ny+

Dwens Date Date Date Reviewed by RRFSC24 SEP1991 Date 4 g

OPERATIONAL FROCEDURE Procedure 1. T AB C EXTRACTOR SYSTEM GENERAL The extractor s, ya VM s iried for operability prior to the initial experiment for the day. SPECIFIC

1. Assembly of the extractor system:
a. Inside the exposure room:

(1) Move the inside receiver section into position in front of the core; screw tube supports to the floor . (2) While holding the appropriate connecting tube in position, tie the strings in the tube to the two ends coming out of the exposure room wall and to the two ends in the receiver section. (3) Align the ends of the tubes and slide the clamp over each joint. (4) Place the alignment tools into the appropriate holes to check the tube alignment; tighten down the clamps. (5) Remove the alignment tools,

b. Outside the exposure room:

(1) Remove tube plug. (2) Move the receiver section close to the tube projecting from the wall. (3) While someone else is pushing the table toward the wall, insert two screws into the holes on the securing bracket (beneath the table). (4) Tie the string from the end of the small tube to the end of the wire cable. (5) Pull the string in the large tube slowly while having someone inside the room guide the string. (6) When the cable is all the way through both tubes, thread the cable through the receiver tube while moving the receiver table into final po-sition against the wall (if necessary, add an additionallength of cable to the take up reel). Revised: 15 May 91 Page 1

i i ! (7) Position and tighten clamp over the joint; position carrier in tube j and connect cable to each end; remove the tape on the take up teel. i (S) Pull back on the drive motor assembly untilthere is no slack in the ! cables; tighten the adjustment bolts on the drive assembly. , f (9) Connect the electrical cables to the motor, control unit, and limit switches. ,

2. Disassembly:

i a. Reverse the order of the above with the following changes: ) (1) Defore loosening the motor assembly, place tape on the cable drum

to keep the cable from moving (ensure the carrier is in the receiver sec.

tion). i (2) Before pulling the cable through the tubes, attach a new string to it. (3) Leave enough slack for disassembly inside the exposure room. (4) Cut the string at thejointsin the room and tape the ends to the tubes or tie the ends together .

b. Ensure the tube plug is in place, and the control unit is secured.

e 3. Operations:

a. On the motor control, initially set controls as follows:

(1) Power switch:"OFF". (2) Torque control:"OFF". (3) In/out switch: " BRAKE". l (4) Speed control:"0%".

b. Plug motor controlinto AC outlet; switch the power switch to "ON".
c. Switch in/out switch to appropriate position.
d. Slowly increase speed to an appropriate level; as the carriage approaches its-i full in/out position, decrease the speed slowly to "0%"..
c. Turn the in/out switch to " BRAKE".
f. During retrieval operations, the Safety and' Health' Department (SHD) moni. -

tor will be present.

Revised: 15 May 91' Page 2
    . . - . . . - - . . - . - . . - . = - . - .                   . _ . - - . . .   . . . -. -     ._-- . .- - - . - ..- .. - . - ,- :

OPERATIONAL) i PROCEDtJRE" 1 m, ,a w . . u.- , N - Procedure 1D . r This procedure has en approved by the Reactor Facility Director 

                                /bf([j ;~ ~ - nnm04hw19r Re   ,t o r _ Facility pec]                                U
                                                                                                 'j              Dale
                       /L A&,                                                               MNcey 9(
                    "Chait' man, Safety and f(alth Department                                       ' D ate Reviewed by the Reactor Staff T /e/,rdl/                                 929    *p / -

Wright' Date L C-<. q C-79Al Cyeotje Date 4'Y%/[jjAL 79 A,4; F sba'cka Date L( 29 +7 % / I Spence Datc h to.A~ o 14 %ff Laughcly0 date 

                                     /h4w Nww 'h m% 9 /
                                    $gyed'                                             D'at e.

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                                                   ~

Owens Date D ate . Date Reviewed by RRFSc 24 SEP 1991 D ate - 

 ?!                                                                                   k                                         $h?"

OPER AT10NAt.' PROCEDtJRE , W ' ' Ptocedure l, TAD D PNEUMATIC TRANSFER SYSTEM (PTS) . GENERAL

1. This PTS procedure is inactive. If the PTS Facility is reactivated, then this pro-cedure must be reviewed and approved by the RRFSC and the Reactor Facility Di-rector.
2. ALAR A principles will be prscticed during PTS operations.
3. All PTS operations will be directly supervised by a reactor operator present in the Hot Lab.

SPECIFIC

1. PTS Setup:
a. Position core at 700 (inside region 111).
b. Ensure communications are established between the hot lab and the control '

room.

c. Inspect rabbits to be used in the PTS for cracks or other damage,
d. Aluminum rabbits must be diverted to the Hot Cell and therefore may only be used on the "A" system.
e. If the anticipated radiation level of an." returned rabbit is greater than 1.0 R/hr at 1 meter, take the following precautions:

(1) Use the remote control unit, unless experiment requirements dic-tate otherwise. (2) Place a radiation survey meter next to the receiver / sender station so that it can be monitored from the remote control unit. (3) The rabbit will be irradiated in the "A" system and then diverted to the Hot Cell or returned to the irradiation location.

2. Manual Operations:
a. Ensure all switches on both the local and remote control units are in the "OFF" position; place the local / remote switch in the desired position,
b. Place blower switch in the "ON" position.

l l Revised: 15 May 91 Page 1 i 

            ._. , . , _ . _ . . _ , _ - . _ _ . _                                          . . - _ . . ~ _ , . _ . _ _ _ . - _ _ _ . _ _ . . . _ _ . . ~
c. Insert key into local control unitt turn key to "0N" position.
d. Ensure tubes are empty,
c. Set mode switch (man /off/ auto) to "hf AN" position. Blower will start,
f. Set in/out switch to the "OUT" position and the tube on/off switches to "ON"; allow the system to run for a short time,
g. Set tube on/off switches to "OFF" and turn in/out switch to "lN".
h. Load samples into tubes.
i. Check communications with reactor operator at the reactor console.

J. When the reactor is at the designated power level, set the tube on/off-switches to "ON" one at a time, to send rabbits into the irradiation location.

k. Begin stopwatch or timer.
1. Tur". sube on/off switches to "OFF" and turn in/out switch to "OUT".
m. Ensure a Safety & Health Department (SHD) monitor is present during re-trievals.
n. Set on/off switch to "ON" one at a time; rabbits will return to sender / receiver station,
o. Set all switches to "OFF", and remove key from control unit.
3. Automatic hiode:
a. Complete steps 2a d above,
b. Set mode switch to " AUTO" position. Blower will start.
c. Complete steps 2f i above. '
d. Set timer (0 to 5 minutes) by turning the red and black arrows to the desired irradiation time,
c. When the reactor is at the desired power level, briefly push the timer push button and release. The rabbits willleave the receiver / sender station and will au.

tomatically return at the end of the preset irradiation period. The timer will au-tomatically reset.

f. Turn all switches to "OFF" and remove kr.y from control unit.

Revised: 15 May 91 Page 2

i i - l 4 i

4. Diverting Samples: l r
a. Diversion of sarnples to the Hot Cell may only be made using the "A" systern.  ;

i b. After the rabbit has returned to the receiver / sender station, set the di.  ! vert / send switch to " DIVERT" and hold it until the loading port handle trips to l the rear position, i i

c. Send the divert / send switch to " SEND" and hold for a few seconds. The rab.

bit will leave the receiver / sender station and travel to the Hot Cell. l t

  • i 3

1 l 1 l l Revisedi 15 May 91 Page 3 1 l............._......-._._...-,.-...--.-.-,_,-,,_.,.~, , , , , ,

i OPERAT10NAC JPROCEDURB'. Procedure 1E i This procedure ha been approved by the Reactor Facility Director 1A($ta): ----_ b' A u C- f Reacfor Facifity Dj/pt D Of ly (l D/te LA ",l

  • Au9/

Chairn6n, Safety and H61rif' Department 'D ate Reviewed by the Reactor Staff itM.4 T- M ' V/ Wrigtft Date 

                                 /LW.,#                 9 -1 9-41 l    I.iL 21/              /(

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                                !)rdsu 90.,.su 79 M A)

Date Sj{u)'eVi b ta J9MWff 

                                                       $1 Owens "                    Date Date Date Reviewed by RRFSC24 SEP 1991 Date
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                                                                           'F OPERATION 5LIPROCEDURC@,

ar mm am m.w an T ds~ Procedure 

                                                                                                                - ~-a 1, TAB E IN-POOL /IN-CO}iE EXPERIMENTS GENERAL ALARA principles will be followed during these experiments. These procedures apply to all in pool or in core experiments except CET operations (See Procedure 1 Tab B).

SPECIFIC

1. All operations will be supervised by a licensed operator
2. Actions willbe taken to prevent damage to the reactor core or aluminum Mnk.
3. Ensure that a member of the reactor staff and a SHD representative are present in the reactor room during the removal of samples from in pool or in core loca.

tions.

4. The removal of experimental materials from the pool or core will be monitored with a radiation survey meter; additionally, a reactor operator will monitor the re-actor console during insertion and removal of in core /in pool experiments.

1 1 ' i I l Revised: 15 May 91 Page'l _. . _ _ _ _ . . _ _ . . _ . . . _ _ . _ . . - . - ~ . _ _ _-__._.._.__.._u_..,.___, _ _ _ -

1 1' a.J 

                                                                                                                                                         ~
                                                                                                            ^                             ^
                                                                                                                                                                                     . Procedute 2
OPERA _TIONALS
                                    . - . .           -a ~.                  iPROCEDUiLEf mn                                                                              m-       4 1

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This procedure has been approved by the Reactor Facility Director i 0 ,

C11 Uf@ifi.. i i / L 0 59 i l O/ M A<J D ale W ' j Reactor Facility Directg l Reviewed by the Reactor Staff I , rh%[/ V- 2 7 ?/ Wright Date a M <. . . C-1 Hl f f 4 74 M Fgrsbadka Datc ) 9i % m 2679?MI [ Spence tate 

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f> 94 i j Owens ~ D ate - I Date Date a 

                                                                                                                                                                                                                            ~

Reviewed by RRFSC 24 SEP 199I Date- 

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x. OPER ATION Ali PROCEDURE 3.. ,c Procedure 2 REACTOR STAFF TRAINING

1. The reacter staff training is delineated in the current "AFRR1 Reactor Opera-tor Requalification Program".
2. The Reactor Facility Director (RFD) determines who is allowed into the train-ing program. As part of the training /requalification program, the following will be performed:
a. A training file will be maintained for each trainee / operator,
b. When a section of training is completed, it will be annotated on the training checklist in each file.

Revised: 15 May 91 Page1 I _____A

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PERApop p [ j g pR{[;_ g % } '@d%Q " Procedure 3 This procedure has been approved by the Reactor Facility Director i 

                                                                                        ?'ji
                                                   /                    I   h5)2
                                                                            'T10 l/r.4, '

AlV / M Date A~ 4/7 Reactor Facility Directg/ Reviewed by the Reactor Staff 1 town / T. t 9 ?/ Wright' Date 

                                            #.G<, e                       C-79. 41 Geor      ej j                       Date jI                           '2 9 A ,e,/
                                        /F f /L/

Fqjba'cka r Date 

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Slence Date M1~alot JFMn9/ Laughefy O _ date Ar&1_,Ywie Vinuyet t uyen D' ate 

                                             ]D //hs $4/Wf/

w' ens Date Date Date 

                                             .                      - Reviewed by RRFsc 24 SEP1991 Date
                      - '  *f       '

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                                                                                                                   ..     . Procedure 3 MAINTENANCE PROCEDURES GENERAL Maintenance procedures are provided in other references.

SPECIFIC

1. Preventive maintenance procedures for each item of the reactor systems are pro-vided in the maintenance logbook and console systems manuals.
2. Annual shutdown procedures are given in the Annual Shutdown Checklist which is revised each year by the Reactor Operations Supervisor (ROS) and ap-proved by the Reactor Facility Director (RFD).
3. Malfunctions are annotated in the Malfunction Logbook. Entries are normally made by the operator who discovered the deficiency. When corrective actions have been made and annotated in the m:lfunction logbook, the RFD or ROS shall re-view and initial the entry.
4. Procedures for maintenance of specific equipment are provided in the manufacturers' literature.

Revised: 15 May 91 Page 1 i

I 

   $f-  '-     -

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                                                                           .l"gjp;41:

i i 1 i 

 'This procedure has been approved fy the Reactor Facility Director
                    -    'i             ot,:n  wn      n=   m (6 #N9 [

R Facility D yv ' L Date 

                                          )

t / h _CL 

                     - A            m                             11-20-9]

Chalt in fafety and Health Department Date i Reviewed by the Reactor Staff 

                           )Wh                 ).A N %'

Date Holmes M &- I4 p.e}l YhYL~ l7iv! P Date 

                          '%ptsble'kavAvnw trhr1)

SpeTee ' Date M J$~ Jut. JrAln 1/ r Date flAs3 augherylaw()11MovAI 

                                       '~

[pye Date IW > lla. % ft 0"yjns o f"f ,lo j Date 

                            > _ W
  • v.G *V Y (2 Id k Eni
                             ~

8t'd Date Reviewed by RRFSC 17 DEC 1991 Date b' 

     ?E%J{0ffQ[gRp ppyg{ "
  • 7 7" Pn,ahn 4 PERSONNEL RADIATION PROTECTION GENERAL All activities performed in areas of potential personnel radiation exposure will be done in accordance with ALARA principles. These areas are the reactor room, upper equipment room (3152), lower equip: at room (2158), ivarm storage, prep area, exposure room 1, exposure room 2, and the hot lab / cell. AF R R1 Instruction 6055.8B, Occupational Radiation Protection Program, is the radiation protection program followed by RSDR.

SPECIFIC

1. Reactor Room:
a. CET Operations: See Procedure 1 Tab B.
b. When working inside chained area around pool: The reactor operator on the console shall be respor.sible for controlling entry into the chained area during operations
2. Warm Storage: See HPP 31.
3. Prep Area: See Prep Area Briefing.
4. Exposure Rooms: See HPP 31 and Procedure 1 Tab A.
5. Hot Lab / Cell: See H PP 3-5 and Procedure 1-Tab D.
6. Upper and Lower Equipment Rooms:
a. No written radiation protection procedures are required for entry into these rooms,
b. Access to these areas is controlled by the AFRRI Reactor Physical Security Plan.
7. Personnel Dosimetry and Monitoring: See HPP 3-1,3 2, and the Prel' Area Briefing.

l - 1 Revised: 15 Nov 91 Pace 1

OFERATIONAtMPPhcEDURENWA 

               . a.     .m . 4             maa :          '      m, 7' As.e r. -               *       -

Procedure 5 [This procedure has been approved by.,,,.,.j.. 

                                                                          ,           the )q}eact,or, Facility Director
                                           "                                             !      h.h id (w Reactor Facility Dirq#or nn 's.       date Reviewed by the Reactor Staff 7%N                   7-19-9/

Wrigh't Date _ L 6, ,,n 9-WI 

                                                               /   p--      29h
                                               ~F      ba'cka                   Date     i w           m%nt S1fence                        tate w %2n                     D9m.4/

Laughe{J () tiate. 

                                               &#uo guye'n '
                                                            /fa%' c 79 DN 'lI date
                                                  ]9 hhs., Jfw 4f

_Owens Date Date Date Reviewed by RRFSC 24 SEP Igl . Date 6 

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b;; . . .; - - Y g j3 ; y 1 

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yggg;g. gig.g ;sjr; PHYSICAL SECURITY GENERAL Physical Security requirements are given in the AFRRI Reactor Physical Security Plan. SPECIFIC

1. The reactor control room and the reactor room will be secured if no reactor -

staff member is present for a prolonged period of time during duty hours.

2. Control of keys is delegated to the Reactor Operations Supervisor. Key inven-tories will be performed annually, not to exceed 15 momhs..
3. The Physical Security Plan will be reviewed annually.

l l l Revised: 15 May 91 _Page1

3 OPERAT10NAlc.,,,_m,_-,,

PROCEDURE . - Procedure 6 i .w -

4 l i This procedure has been approved by the Reactor Facility Director i j\ Q w r V u.. 

                                                                                                                                          ~x,
                                                                                                               ~ ~ ~           u il
                                                                           $)             .

jb Tbc.a 9/  ; Reactor Facility Direep D' ate I 1 Reviewed by the Reactor Staff  : stL}/ T 2 '/ 1/ Wrighi Date ' AL ,# v- r 9- 9l 

                                                                                    /4        /       \.       29                                                                           t
                                                                               ,Fppbadka                                Date
                                                                              %m                              Wrfif/

Spence 'D ate 9Mfa.sLe E F L f/ Laughir date 

                                                                              /)hb [y <tu       U_ m 29 nto..,9/                                                                            -
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Owens Date. Date Date i Reviewed by.RRFSC i Date 

                                                                                                                                                                                     ~

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  . _ _ . _ - . _ _ _ . _ . . .                                         .m__ . _ m. .. _                                  _ _ . . _ . . . _ . , . _ _ _ _ _ . _ . - . _ _ . . _ . _

l OPERAyONAL]ROCEDURE,7 hocedet 6 

                                              =_                                    -.

EMERGENCY PROCEDURES GENERAL The reactor ernergency organization, emergency classes, and emergency action lev-els are set forth in the AFRRI Reactor Facility Emergency Plan and its implement-ing Procedures. SPECIFIC Perform the following, as appropriate (need not be done in order).

1. Reactor Emergency:
a. SCR AM reactor,
b. Check radiation monitors; use portable survey instruments to assess situa-tion, if necessary.
c. Notify ERT Commander of situation,
d. Activate emergency organization.
2. AFRR1 Complex Emergency Evacuation:
a. SCR AM reactor,
b. Secure any exposure facilities which are in use so that personnel access to that facility is not possible,
c. Remove logbook, emergency guide, radios, teletector, tool kit, and keys; re-port to ERT.
d. Ensure reactor area doors are secured upon departure.
3. Proper classification of emergency situation: All SROs must review the vefer-enced Emergency Plan documents and be able to properly classify the events as they occur. Below is a tabulation of emergency classification to be used as guid-ance.

Revised: 15 May. 91 Page 1

4 i EMERGENCY Radiation Activate AFRRI Complex Activate Emergency . CLASS Alarms Emergency Evacuation Response Team (Unanticipated) l l Class O Fire Alarm (non reactor) Yes Yes Class 1 R 1 > 1 min. Yes. R2> 1 min.

  • Yes R3 No No R5 > 1 min.
  • Yes R6 No No E3 > 1 min.
  • Yes E6 >1 min.
  • Yes SOM> 1 min.
  • Yes Reactor Stack Fan Monitor No No Fire Alarm (reactor) Yes Yes Class 2 CAM > 'l min.

concurrent with R1, R2,-

RS, and/or SGM
  • Yes NOTE:
  • A decision to evacuate the Institute will be made by the ECP Commander based on input from the ERT Commander.

l 

 --Revised: 15 May 91-                                                         Page 2

i l i OPERATIONAL ' PROCEDURE ? Procedure 7 d r I il'his procedure has been approved by the Reactor Facility Director j I - .- . .. 

                                                                    /                         !! n'
                                                                   .-,fu':ofh, lA k l & ,. N                                            2711               ci ,

Reactor Facility Diredfor F6 ate r Reviewed by the Reactor Staff 17sCO r. 29-?/ Wright' Date L C c.,# _ s'-7 H f

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spence bate m /~ihn 29n V Lau IfetyU Ifate b An uks., l'"? >Vs V q ')} Uyen' ' ~ Diite d 9Aiut 1Mheet Owens Date Date Date Reviewed by RRFSC24 SEP 1991 Date 

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. REACTOR CORE LOADING AND UNLOADING GENERAL Loading and unloading of the reactor core shall be perforrned under the supervi-sion of the Reactor Facility Director or the Reactor Operations Supervisor. SPECIFIC  :

1. Setup
a. Ensure that at least one nuclear instrumentation channelis operational.
b. Ensure that the source is in core.
c. Ensure that an operator monitors the reactor console during all fuel move.

ments.

d. Check new FFCRs before insertion into the core; this includes cleaning, vi. .

sual inspection, and length and bow measurements,

c. Install all control rods,
f. If irradiated fuel elements are to be removed unshielded from the pool, obtain a Special Work Permit (SWP) from the Safety and Health Department (SHD);

do not remove fuel elements with a power history (greater than 1 KW) in the previous 2 weeks from the reactor pool.

2. Core loading
a. After each step of fuel movement perform the following:

(1) Record detector readings.

(2) Withdraw control rods 50%; record readings.

(3) Withdraw control rods 100%; record readings. (4) Calculate 1/M. (5) Plot 1/M versus number of elements (and total ma.es of U-235). , (6) Predict critical loading, i (7) Insert ALL control rods; continue to next step.

b. Load fuel elements in the following order:

(1) Load the B ring and C ring thermocouple elements. Revised: 15 May 91 Page 1 L. - . . = a. - - - - - - . - - - -  ;-------"

1 (2) Connect thermocouple outputs to reactor control console display. (3) Install any other thermocouple elements. (4) Complete loading of Ib and C ting elements (total of 18 standard el-ements plus 3 FFCRs). (5) Load D ring (total of 33 standard elements plus 3 FFCRs) (6) Load the following E ring elements in order: 16,17,18,20,6,8, 9,10 (total of 41 star.dard elements plus 3 FFCRs). (7) Complete the E ring by loading the following elements in order: 15,21,11,5,14,22,4,12,13,1 (total of 51 standard clernents plus 3 FFCRs) (8) Load the following F ring elements in two elemems per step until criticality is achieved using the following loading order: 22,23,24,21,20,25,26,27,28,29,30,1,2,3,4,5,19,18,17,16,15, 14,13,6,12,7,11,8,10,9. Once criticality has been achier,:d, perform control rod worth measure-ments at core position 500 by rod drop technique. Calculate shutdown margin: SDM = Total Control Rod Worth K excess TRANS Rod Worth (9) Load core to $2,00 excess reactivity by loading two elements per step using the loading order in instruction 8. (10) Wrify control rod worth using rod drop techniques, calculate SDM (11' aac the core to achieve a K excess that will allow calibratiori of the TRANS rod based on the last available worth curve of the TRANS rod (approxirnately $4.00). Calculate the reactivity value of each ele-ment as it is added. (12) Calibrate all control rods. (13) Calculate the shutdown margin. (14) Estimate K excess with a fully loaded core (must not exceed $5,00). (15) Load core to fully operational load using loading order in instruc-tion 8, and recalibrate all control rods. Calculate the shutdown margin. (16) Adjust the core loading pattern to meet operational requirements if necessary. Recalibrate all control rods. Calculate the shutdown mat-gin. Revised: 15 May 91 Page 2

8. At the end of each dry in which a Daily Operational Startup Checklist or Daily Safety Checklist has been completed, perform a Daily Operational Shutdown Checklist (Tab I),

9. Complete the monthly summary .
10. Respirator equipment will not be used on a routine basis. Respirator equipment is provided for use during emergency conditions only.

l l l Revised: 15 Nov 91 Page 2 4

m M 5L4.M3il'!8 - This procedure has been approved by the Reactor Facility Director ( M MW9 

                                                                                                                     ~

Reactor Facility Dirgr Date P Reviewed by the Reactor Staff A & W p snwij ~ G orge s Date 

                               //fot ///M                                                               7%,
                               'Fo sbhb '                                                                   Date usa 2 Fee 1/

Speilco' Date LL J2hA/ Lauhhei9 Date WT.Nw 492 4/ 

                                                           //                                          hA          )

Holmes 

                                                                                                            /3Nf/
                                                                                                            - Date Date Date Reviewed by RRFSC17 DEC 1991 Date

0 M d.ff t 6 

                     ,--                 REACTOR      OPERATIONS GENERAL Logbook entries will be made in accordance with the Logbook Entry Checklist (Tab A).

SPECIFIC

1. The names of the individuals who supervised and performed the daily and weekly checklists will be shown at the top of the checklist. Checkmarks or num-bers, as appropriate, will then be entered on each checklist line as that item is formed.
2. Perform reactor Daily Operational Startup Checklist (Tab B), utilizing appro ate nuclear instrumentation set points (Tab C). In the case of no planned opera-tions, a Daily Safety Checklist (Tab B1) may be performed.
3. Record at the beginning of each day in the reactor operations logbook the SR on-call for that date. At the end of the "zeding hy, WRW1
    =ce!! fer +he gec : ; - gh+, "feeend, er he!!&y _(Med 'he m:ee                                           /J W 1 2.

of 'S

4. At the begining of each working day, also record
  • ame of the physicist in charge (PIC)present at the reactor facility. If th " C vised entry will be made in the logbook. 9 changes during the day, a re-either the PIC or SRo on-call
5. Perform K-excess measurement (Tab D).
6. Perform operations in accordance with the following:
a. Steady state operation (Tab E).
b. Square wave operation (Tab F).
c. Pulse operation (Tab G).
d. CET operations (Procedure 1, Tab B).
e. Pneumatic Transfer System (r'rocedure 1, Tab D).
f. In-pool /in-core experiment (Procedure 1, Tab E)
7. Perform Weekly. Operational Instrument Checklist once during each calen week (Tab H).

i Revised: 15 Nov 91

Page 1
                                              , , _ .  .~ .            -   -~' ~ ~~~~~ ~   ~~~

l 6e g. e 

             -c     .

n e. .a m;.q gggg y;j: g gg, g g g ,ae d This procedure has been approved by the Reactor Facility Director b d [.ctrt<_ E4 , V 7/ Reactor Facility Director 'D at e

Reviewed by the Reactor Staff 4W Ilusj Get!e Date
                                   /$Yo 1~l N                    tisJint Pors ek'a                      Date
                                             %                 //4L4/
                                    ~Spe'nce                    " d ate 41 M wkull,D4 9/

Laughery 4 Q bane h/w M up, R Jd q i Owens Y lC Date fj Date l Date Date Reviewed by RRFSc 2 4 SEP 1991 Date l i 9 - , - ,w - . . , , e

00 lIY$$$NNbb$$$$$$$3diESN5$345$?^B~^ LOCBOOK ENTRY CHECKLIST

1. The reactor operations logbook is a before the-fact record, that is, entries will be logged before the operator actually performs the planned function. Any late en-tries will be so noted.
2. The operations logbook will have a hardbound cover and will be sequentially numbered by volume. The pages will be dated at the top of each page and each page will be sequentially numbered.
3. The Reactor Facility Director (RFD) will review each logbook upon its comple-tion; he will make an appropriate entry in the back of the logbook and sign the entry. The operator who makes the final entry at the end of a logbook is responsi-ble for ensuring that the ROS is notified that the logbook is ready for RFD review.
4. Allitems in GREEN (see below) that are not closed out during the working day will be carried in GREEN at the end of the day and again at the beginning of the next operational day.
5. The entries will be made in ink and in accordance with the following designated color code:
a. BLACK and BLUE-BLACK:

(1) Console locked and unlocked. The individual at the console will enter his/her name and the supervisory licensed operator's name, if necessary. (2) Checklist number and completion time. (3) Power level at criticality and subsequent power level changes. (4) Reactor SCRAM. (5) Mode of operations. Use appropriate stamp or entry to designate the op-eration: (a) Steady State (b) Square Wave (c) Pulse Revised: 22 JUL 91 Page1  ; 

                                                                                                     \
                                                                                     .---_m__._._-.a

(6) Operation of reactor associated facilities such as lead shield doors, pneu-matic tube systems, etc., unless such operations cause a change of reactivity (see 5.b.(2) below). (7) Change of personnel at the console. Name of personnel v.ill be entered along with the licensed operator present in the control room, if the person at the console is not a licensed operator. (8) The operator in charge will be designated in the logbook whenever multi-ple operators are signed on the console. (9) Completion of the daily startup and shutdown checklists and weekly checklist. (10) Signature of reactor operator to close out '.ne log for the day. (11) Designation of the SRO on call and physicist in charge (PIC). (12) Reactor calibrations and data. (13) Allline outs, entry errors, changes in mode of operation stamp lines, and end of page line outs will be initialed or signed by the operator,

b. RED:

(1) K-excess measurements, to include experiment worth determinations. (2) Actions which affect reactivity: (a) Core movement. (b) Fuel movement. (c) Control rod physical removal for maintenance. (d) Experiment loading and removal from the CET, PTS, pool, or core.

c. GREEN:

(1) Any reactor malfunctions noted upon discovery / occurrence with a second entry noting corrective action has been completed. (2) Additionalitems entered at the discretion of the operator such as addi-tion of makeup water to the reactor pool, etc. (3) Any Technical Specification required equipment taken out of service for any reason. A second entry is made when the unit is returned to service. , 6. When an operation requiring entry into the logbook falls under more than one

color code, the color to be used will be determined via the following order of prece-dence
RED - GREEN - BLACK / BLUE-BLACK.

Revised: 22 JUL 91 Page 2

i ', OPERATIONAL -c_m. [.2:PifC'CEDUREN- ~- ", M ,""- ' m ~...m ET_f" T71.P}ocedure 

                                                                                            <                m:            ,-   8. TAB B-l This procedure has been approved by the Reactor Facility Director                                                                                 i
                                                                                                .              .c,..
                                                                                          .h 5 $ NU N 0 $

' d1L IQ Reactor Facility Dir 'c 'r w xn, D a te' Reviewed by the Reactor Staff i? Wu Er' 5- 2 er- 1) Wright ' Date 44,,,< T-71 -9/ Geor e Date 

                                                  //lt F sba'cka b 21%)/

Date w M %'ll Yp'ence 'D ate M 'imla.. a??#fcar f/ Lau h(ryg Date a Mm., " 79 h 4f yen # D ate - i 

                                                        . Jhu Of#Wfs Owens                                    Date i

Date ( Date 

                                                                                . Reviewed by RRFS$4 SEP 1991 Date bhb5?$$l$?ll ll%$?AblNQQ jl}l$f[

OPERATIONA1.--- ' PROCEDtJR@ m, m 

                                                                    ~ m,                                   '

Procedure 8. TAB B a DAILY O_PERATIONAL STARTUP CHECKLIST Checklist number Date SRO On Call Supervised by Assisted by Operators Time completed I. EQUIPMENT ROOM (Room 3152)

1. Air comp ressor pre ssu re (psi) . ... ... .... . ..... ..... .. ..... .. .. . .. .. . .. . . ..  ;
2. Air compressor water trap drained .......... .... ....................... ..

3 . Ai r d ry e r Op e ra t in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4. Doors 231,231 A, 3152, and roof hatch SECURED.................

II. LOBBY AREA Lobby andio ala rm tu m e d off . . .. .. .. . . .... . . . . . . . .. . . .. . ...... . ... . .. .. . . ... . . 111. EQUIPMENT ROOM (Room 2158)

1. Pre filte r diffe re ntial pressure ......... ... .. .. ..... . ... .........................
2. Primary discharge pressure (psi) ..............................................
3. Deminetalizer flow rates set to 6 gpm ........................... .........
4. Staek roughing filter (inches of wate r) .....................................__.
5. Stack absolute filter (i.m e s o f wat e r) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6. Visu al inspeetion of are a .. . ........ . .. . . . . . ... ... . ... . . .. . . ... ... . .. . .. . ..
7. D oo r 215 8 S E C U RE D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

IV. PREPARATION AREA Visu al i n spection of are a .. .. .. . . . . .. . . .. .. . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . .. . . . . . . V. REACTOR ROOM (Room 3161)

1. Tt ansient rod air pressure (psi) .............................. .................
2. Shielding doors bearing air pressure (psi) ................................
3. Visual inspection of core and tank ...........................................
4. Number of fuel elements and fu el elem en t s . .... .. .......... ..

control rods in tank storage cont rol r o d s .... ........ ...... ....

5. Air particulate monitor (CAM)

(a ) Op era tin g a n d Tra cin g . .. .. .. .. .. .. . . . . .. . .. . .. . . . . . . . . .. .. .. .. . . . . . . .. . . .. .. (b) Alarm test completed, damper closure verified...................

6. D oo r 316_2 SECU RED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7. Stack gas monitor quality assurance checked ........ .................

_J AFRR1 FORM 61a (R) Revised: 15 May 91 Page 1 1 l

I 4 i VI. REACTOR CONTROL ROOM (Room 3160)

1. Emergency ait dampers teset ......................................................
2. Console recorders dated - ..........................................................  !
3. Staek flow and fuci temperature recorders dated .........................

l

4. Logbook dated and reviewed ........................................................ j
5. Water monitor box (resistivity must be > 0.5 Mohm cm) 1 (a) Background aetivity(epm) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

(b) Water monitor box resistivity [Mohm cm) .................................... (c) D M 1 r esistivit y [M o h m cm ) ....,,. .......... .. .... . ... ........ . .. . ...... ... ... .. . .. (d) D M 2 resisti ity (M o h m cm] .... ...... ... .. .......... .. .. .... .. .... .... .. ... . ..... .

6. S t a ck g a s flo w r a t e [K cfm ) . . . . . . . . . . . . . . .. . . . . . . . . . . .. .. . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . .
7. S t a ek lin ca r flow ra t e (K ft/ min) ........... .. .. .. .. .... .. ... ... ... .. ... . . ..... . .. ... . . .
8. Gas stack monitor (a) Background (cpm) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

(b) Alarm check . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . (c) H igh ala rm set t o 800 M P C Ar 41 .................................................

9. Radiation monitors Monitor Alarm Point Reading Alarm Setting Functional (mR/hr) (mR/hr)

(a) R 1 500 (b) R 2 10 (c) R 3 10 (d) R 5 50 (c) E-3 10 (f) E 6 10

10. TV monitors on .............................................................................
11. CAM high level audible alarm check . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . __
12. Wa t er t em p e r a t u re (in let) . .. . . . . . . .. .. .. .. .. .... .. .. .. .. .. . . .. .. .. .. .. .. .. .. .. .... .. .. .. . .
13. Water levellog completed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
14. Time delay operative ....................................................................
15. Source level power grcater/ equal to 0.5 cps. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
16. Prestatt operability checks performed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

! 17. Interlock Tests (a) Rod raising, SS mode (c) I kW/ Pulse mode (b) Rod raising, Pulse mode (f) NM-1000 HV (c) Source- RWP (g) Inlet Temp - _ (d) Period RWP 

                   '18. SCRAM checks (at least one per rod)

(a) % Power 1 (h) Reactor key 

                      - (b) % Power 2 (i) Manual                                                                                                          '
                       -(c) Fuel temp 1                                                                              (J) Emergency Stop

, (d) Fuel temp 2 (k) Timer l (e) HV loss 1 (1) CSC Watchdog (f) HV loss 2 (m) DAC Watchdog (g) Poollevel

19. Z ero p o wer p ulse . . . . . . . . . . .. . . . .. . . . .. . . .. . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . .

L l l AFRRI FORM 61a (R) Revised: 15 May 91 Page 2 

           . ._               _      _ . _ ~___.                                                  _ . _ - -                                               . . _ _ , _                                     _ . _ _      _ -    _ _ . _ -

OPERATIONA13 uuam.a w%-. IPROCEDU.REi_d w .u :a aew, m: _ O .e--,".3 . m&x - .. 

                                                                                                                                                                   ' Y'Piscidurf 8, TAB B1 l

This procedure has been approved by the Reactor Facility Director 

                                                                                                                                                                                                         ^ ~
                                                                                                                               "!b=::ngg g%,'        %
                                                                                                                                       . ; 7 y';j' yli I                4 9/

Reactor Facility Dir(dtor /Date Reviewed by the Reactor Staff TWalI f. 2M-9) Wright Date it. c u m., r- 2 Mi cor e Date 

                                                                                                                 .. k 248,1/

Fo sbaMa Date n n%4/ fpence ' Date ' 

                                                                                                   #1 L s}m                    AMhr 9/

Laughet'y CT tate-O G k wyn 79 nau,9' Date fpuyen'o is b ut v dYr J W 7/ Owens Date Date Date 

                                                                                                                              ' Reviewed by RRFSc 2 4 SEP 1991 D ate 4
                                                                                                                                                    ,       ,             'c                    8 '

1 

 ,        ..           m
                                                                                                                                                       -      %--   , , .    --..r.-,     -,ev,     -m-- , -w-      -

 . - . ... . - -                         . ~ . . ~ . - - ~                                 _ - - .                         . - .           . . -                 . - - -                 .           . -

l , .1 1 l 1 OPERATIONAIF PROCEDU,REi , m a um w a d  %.*IE . m-iI ._ s._ 

                                                                                                                                                                ~ Procedure 8. TAB B1 DAILY SAFETY CHECKLIST Checklist number                                                                                             Date                                                      _

SRO On Call Supervised by Assisted by Operators Time completed I. EQUIPMENT ROOM (Room 3152)

1. Air compressor pressure (psi) ............................................ .....
2. Air compressor water trap drained ..........................................
3. Ai r d ry e r Op e ra t in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,
4. Doors 231,231 A, 3152, and roof hatch SECURED.................

II. EQUIPMENT ROOM (Room 2158)

1. Prefilter differential pressure ....................... ............................
2. Primary discharge pressure (psi) ..............................................
3. Demineralizer flow rates set to 6 gpm .....................................,
4. Stack roughing filter (inches of water) .....................................
5. Stack absolute filter (inches of water) ......................................
6. Visual ins pe etio n o f area ............... .... . ........ .. .. ... ... ... ............... ..
7. D oo r 215 8 S E C U R ED . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

III. PREPARATION AREA Visual ins pe etion of are a ....... ............................. ........................ .. IV. REACTOR ROOM (Room 3161)

1. Transient rod air pressure (psi) ................................................
2. Shielding doors bearing air pressure (psi) ................................
3. Visual inspection _ of core and tank ........................................ ..
        -4. Number of fuel elements and                                                         fu el elements ...................

control rods in tank storage cont ro l ro d s . . . . . . .... ... . .. .. .. ..

5. Air particulate monitor (CAM)

(a) Op e ra t in g a n d Tr a cin g ..... .. ........ . . .. .... .. . . .... .. .. .. . . . . .. .. .... .. .. .. (b) Alarm test completed, damper closure verified...................

6. Doo r 3162 - SECURE D . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . .
        -7. Stack gas monitor quality assurance checked ..........................

AFRRI FORM 61b (R) Revised: 15 May 91 Page 1- 

                                                                                                                                                         ... .              ... .. ~ _ - .._ ._._ -..

i V. LOBBY- AREA I Lo b b y a u d io a la r m t u r n e d o ff . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I VI. REACTOR CONTROL ROOM (Room 3160)

1. Emergency ait dampers teset ......................................................  !
2. Console recorders dated ................................................................ 1
3. Stack flow and fuel temperature recorders dated ..........................
4. Logbook dated and reviewed
5. Water monitor box (resistivity must be > 0.5 Mohm cm)

(a) B a ckgro u n d a etivit y(cp m) ..... .. ... ....... ... .. .... .... . ...... .. ....... . ... .. ... . . . (b) Water m onitor box resistivity (Mohm-cm) .................................... (c) D M 1 r esist ivit y [M o h m -cm] . .............. ... ... . .. ... ...... ...... ........ . ... .. .. .. . (d) D M 2 r esistivit y [M o h m cm ) . .... .. .... . ... .. .... .. ... .... .. ...... ....... ............

6. S t a ck ga s flo w ra t e [K cfm ] . . .. . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .
7. S t a ck lin ea r flow r a t e (K ft/ min) . ... .... .............. ....... .. ... . .. . .. ... . ..... .. . .... . .
8. Gas stack monitor (a) Baekground (cpm)

(b) Alarm check (c) High alarm set to 800 M PC Ar-41 ................................................

9. Radiation monitors Monitor Alarm Point Reading Alarm Setting Functional (mR/hr)

(a) R 1 (mR/hr) 500 (b) R 2 10 (c) R-3 _ 10 (d) R-5 50 (c) E-3 (fj E-6 10 10

10. TV monitors on
11. CAM high level audible alarm check
12. Wa t er t em p era t u re (inle t) ........ ... ... ...... .. .. .. .. . .. ... .. .. ...........................
13. Water level log completed
14. Source level power greater /equa1 to 0.5 cps. .....................................

1 { I AFRRI FORM 61b (R) Revised: 15 May 91 Page 2 

                                                     ~

OPERATIONAL Procedure 8. TAB C 

               'PROCEDURH[

i This procedure has been approved by the Reactor Facility Director 

 \
                                                                               .;i -
                                                                 - .      . .0 V M, a , i i     I 2RAEm l/

Reactor Facility Directp ' D a t e' l Reviewed by the Reactor Staff 7'. w u L t. S*-'4- 9 I Wright ' Date IL G.,a 9 -71 9 b l (L 79 1% F rsbacka Date , _ nce- 2144y'11 1 pence ~ Date 

                                           & l m /a 29 Mle,f/                                                             '

Laughdy(/ bate Adu , % ruin, ln mau 9( uyen # Dite bobran d7n7N4/ 

                                          '0 wens                  D ate Date Date Reviewed by RRFSC 24 SEP1991 Date Ik      b                              bhhhhf}bhhhhhh$$[hfhIbi,

w;. , . 

                           ~ . . 3. s. . . . .gggg;ggy                             ggg NUCLEAR             INSTRUMENTATION          SET    POINTS               l GENERAL 1

These set points may be adjusted for a specific operation by the RFD or ROS but in no case may they be set at a point non conservative to the technical specifications. SPECIFIC The following are channel or monitor set points (alarm, scram, rod withdrawal prevent),

1. Scrams:
a. Fuel Temperaturc 1 & 2: 575 degrees C
b. High Flux 1 & 2: 110% (1.1 M W) 1
c. Safe Chambers 1 & 2 HV Loss: Loss of 20%
d. Pulse Timer: Less than 15 seconds
e. Steady State Timer: as necessary
2. Rod Withdrawal Prevents:
a. Period: 3 seconds b.1 KW (Pulse Mode): 1KW
c. Source: 0.5 CPS
d. Water Inlet Temperature: 60 degrees C
e. Fission Chamber HV Loss: 20 %
3. Alarms:
a. R AM S: As directed in procedures
b. CAMS: 10,000 CPM
c. Stack Gas: 800 MPC Ar-41
d. Water Monitor Box Gamma: 7000 CPM
e. Criticality Monitor (RS): 50 mR/hr day -

l 20 mR/hr night Revised: 15 May 91 Page 1

OPERATIONAISf-PROCEDUREiQ@:MW3CM'J' 

      - :~..un   .a e     . x:a c:
                                                          @IE"
                                                             -~
                                                                   . IdfNPiEcEdhre58,TABD
                                                                    . . w w.,a ,

c l l i This procedure has been approved by the Reactor Facility Director 3 .,..,z .77 l j ,[ } j' g7 4 Reactot' Pa~cility Dis /ctor Date Reviewed by the Reactor Staff 744 {l- >K l9-9/ Wright ' Date ILL<., n T-h-9 

                              //               !   b 79         s, o sbacka              Dafe v .       %            Eff)lsfif fpence                   Date hf_$~rla             09 'mm f/

Lau thir>C) D' ate d 4e Lunt Wh >vu 4l gyen'f' Date L J J k , k & n d w h *7/ O'wens ~ Date Date Date Reviewed by RRFSC24 SEP1991 D ate 

   ,           y      , *          "* 4
                                   .        *,                                         4 F

1 1 ggq,,. . \ ;L ::.g; - 

                                                     ~ yp;ygg; ggg. nun wgw       aggwymrg ..;;;;..y                     y g,.
                    . w.
                                                                               .suzu                        .. . a.

2 . . K-EXCESS  ! 4 E FF ECTI V -

1. Withdraw SAF and SHIM rods 100% ar.d ..i:h& . :he TR AMS :od 25 u. leTBu9
2. Use the REG rod to bring the reactor to cold critical at 5 watts. If criticality can not be reached with the REG rod full out, use the TRANS rod to bring to crit-ical.
3. When power is stabilized at 5 watts, record rod positions in reactor operations
logbook, entering all information in red ink.
4. Using rod worth curves, compute K-excess for the core position used and re-cord in the reactor operations logbook .

Note: Use the curves for infinite water when doing K-excess between positions 300 and 700, 4 4 i i t l Revised: 15 May 91 Page1; l

OPERATION ^1hfPRDCEDURE" a -~ ,u,r s .. -.  :,ws .. Procedure 8. TAB E This procedure has been approved by the Reactor Facility Director 

                                                                                                 *:n-i llo a.m
                                                                                                             $bN!'h zq {y, Reactor Facility Directo9                               Date Reviewed by the Reactor Staff 7* Wa kt         7.29-9/

Wright' Date il kw s-M -St 

                                                                                      ;         74         ,

Fgo sbacka D' ate QRW 2W/14rtl fpence 1 ate-Dt d--rb Mer~ fl Laugifer6 Date W k a u m 2 % 1 fl en' ' 'Date A u hAd e k H/h M /- U~ wens. Date Date D ate - Reviewed by RRFSC24 SEP1991 D ate - 

                                                                                                                               ^       ' '

_ hQ. .-3 

                                                                                                                   ---A--        2-       - - - - - -     " -

OPERA~. 110NAlf? ey y~ .'i t Pr60edure 8, TAB E m . .~P.ROGDUREt 

                                           .au .'w. ' ,<

a.x ,';w$w$ .-- STEADY STATE OPERATION GENERAL The reactor shall not be operated at a power greater than '.l.0 MW. SPECIFIC

1. Set the mode switch to manual mode and clear allwarning messages and scrams.
2. For runs greater than 200 KW, adjest alarm points on R-1 and R-5 to full scale.
3. Raise control rods with the appropriate banking, taking into consideration the location in the pool, power level, and experimental requirements.
4. If final approach to criticalis to be made in Auto mode, perform the following:
a. Set the the thumb wheel dials to the desired power.
b. Raise the rods to the appropriste banking.
c. Select the rods that are to servoed.
d. Make sure that all rods that will be servoed have been raised at least 5%
e. Enter Auto mode.
5. Scram the reactor at the end of the run using the manual or timer scram.
6. Ensure the appropriate entries have been made in the operations logbook.
7. If no further steady state runs, square waves or pulses are anticipated, adjust R-1 and R-5 alarm points to their normal settings.

Revised: 15 May 91 Page1 

                                                                                ^~              ~#

OPERATIONAlfd_PROCEDUREP 

                              . - . - - +               - . - - . _ . ,            ^                            "" _ M.,, j PrClure   S TAB F1
                                                                                                                                  - _ .~   -_

i This procediare has been approved by the Reactor Facility- Director 

                                                                                              , .,y -            .,
                                                                                                 ', l k . - [ , j ; L.

_ ' 2+4 %i-Reactor Facility Director // _ 5 -Date 2, /,, Reviewed by the Reactor Staff 7 Wx$4 ., . m . ii Wright Date 

                                                                            & L<, -                     'i~79-4I L                    74             st sbacka                        Date w

2pyA Jl spence 15 ate 9M L As M W ~ f/ Laujh(ryU tate 

                                                                          /Mrc8Uwn - 11 ho4(

fju *e'n e" Dite l' 1Au 01av AMM4s dGens - D ate Date D ate -- , Reviewed by RRFSO 2 4 SEP1991 l Date 4 

                          - ,                     -                  ._..=

I'ERANg]Rg@pyR j ', jg,]pfmyM M SQUARE WAVE OPERATION (Suberitical) GENERAL The square wave mode will not be used above a demand power of 250KW. SPECIFIC

1. Set R1 and R5 to full scale
2. Determine the transient rod critical position using the core position, the final transient rod position, the rod curves and the equation below. Note that a square wave insertion can not exceed 75 cents.

CRITICAL POSITION (S) = FINAL POSITION (S)-INSERTION (S)

3. Apply air to the transient rod and raise the anvil to the citical position that was calculated above.
4. Bring the reactor cold critical using the three standard contral rods; use a rod configuration commensurate with the core position and experimental requirements.

If Auto Mode is used, select the rods to be used. Ensure thet 6ese rods have been raised at least 5% before entering Auto Mode. Set the cold critical power level on the Power Demand thumb wheels and enter Auto Mode.

5. Stabilize the reactor in Manual Mode.
6. Set power demand thumb wheels to desired power level.
7. Select the standard control rods to be servoed. Make sure that all control rods to be servoed have been raced at least 5%.
8. Scram the transient rod.
9. Raise the anvil to the desired final position.
10. Allow the power level to fall below 10 watts.
11. Switch into Square Wave mode.
12. Depress Fire button.
13. As the power level approaches the power demand level, the console will switch into Auto Mode. If power can not reach the demand power,it will automatically change to manual mode. At this time, either switch to Auto Mode or bring the re-actor to the desired power level manually.

1 Revised: 15 May 91 Page1

l

14. Scram the reactor at the end of the run using the manual or timer scram.
15. Ensure all pertinent information has been entered in the reactor operations log-book.
16. If no forther steady state runs, square waves or pulses are anticipated, adjust R 1 and R-5 alarm points to their normal settings.

Revised: 15 May 91 Page 2

OPERATIONild:JPdOCEDORE$38 

 .   ,. __.au, a;- w. ;uz                m - # ^ #: 'l2.                                  '
                                                                                            ' ' 93thsdure m 2x ,-

8, TAB F2 This procedure has been approved by the Reactor Facility Director A c_ kh , L 9( Date Reactor' Facilit) Directo{ Reviewed by the Reactor Staff 

                             ;Cf4 fg                                           f 4 9-t/

I Wrigh't Date sk.u '5- 7 4 -91 k55ll T 2?? , Date. Wysbadka e cYokn/ 5' pence TTate 91d-c)<e 2k 9/ Laujht'ryC/ T) ate 

                             % e h utu, %hiuv{ni nye'n/ '                                         date A J es                                        bipri Owens                                                 D ate Date D ate Reviewed by RRFSC 24 SEP 1991 Date
$kkk!h ) ;[N k$ h'$. k           J-[ 5/ h h ((i $ 3 Q                                                         fik

I l PERgTig g [ fRgg p p { [ i n 6 y > , Wfm 8, TAB F2 SOUARE WAVE OPERATION (Critical) GENERAL The square wave mode will not be used above a demand power of 250KW. SPECIFIC

1. Set R-1 and R-5 to full scale.
2. Bring the reactor cold critical using the three standard control rods; use a rod configuration commensurate with the core position and experimental requirements.

If Auto Mode is used, select the rods to be used, ensure that these rods have been raised at least 5% before entering Auto Mode, set the cold critical power level on the Power Demand thumb wheels, and enter Auto Mode.

3. Determine TRANS rod anvil setting for desired insertion. Insertion cannot ex-ceed 75 cents. Raise the anvil to that setting.
4. Stabilize the reactor in Manual Mode.
5. Set power demand thumb wheels to desired power level.
6. Select the standard control rods to be servoed. Make sure that all control rods to be servoed have been raised at least 5%.
7. Switch into Square Wave mode.
8. Depress Fire button.
9. As the power level approaches the power demand level, the console will switch into Auto Mode. If power can not reach the demand power, it will automatically change to Manual Mode. At this time, either switch to Auto Mode or bring the re-actor to the desired power level manually.
10. Scram the reactor at the end of the run using the manual or timer scram.
11. Ensure all pertinent information has been entered in the reactor operations log-book.
12. If no further steady state runs, square waves, or pulses are anticipated, adjust R-1 and R-5 alarm points to their normal settings.

Revised: 15 May 91 Page 1

oPER5TIONM25_; mm ma.mm u padmm_dE6ifRB39li" 

                         - -                                 '*' '_Mi';?
                                                                       =: f]6Eyn91$_jf:a{Pi6ddi..ii82
                                                                                      ...m.       L a        a.x . m -

TAB 01 - This procedure has been approved by the Reactor Facility Director f !du u UL@ y .J 

                                                 ~ k % J/                                          c;. 2c/          97 Reactor Facility Director ~                                                                 Dati: -

Reviewed by the Reactor Staff R/d O S' 27 ti Wright Date-LC.- 9-T %I e c f Date - 

                                  /L [ L.                         Vl-- ?t A st F sba6ka.                                           Date
                                                      +%                           2*)fWA l fpence                                            9 ate k /mb4                                           2 G m ,f/

Laughir)() B' ate SA Ytiwat M Ynau 91 yen! " D' ate AshllAu Jt/JFve/ Qwens - D ate : D ate > Date - Reviewed by RRFSC 24 SEP1991 

                                                                                                                     - D ate -
                                                                                                                           t 8            .
                                                      *d

_ _ _ . _ o

OPERATIONAL l PROCEDURE' f DPlocedure 8,' TAB G1 - 

          . ,~._ u . -, n .w a z;;_                                ...   , .w .-

PULSE CyERATION (CRITICAL) GENERAL Pulses above S2.00 must be approved by the RFD (prior to pulse initiation). Spec-ification on the RUR may be used to meet this requirement. SPECIFIC

1. Set the alarm points on R-1 a:-ri R-$ (criticality monitor) to full scale.
2. Bring the reactor cold critical using the three standard control rods; use a rod configuration commensurate with core position and experimental requirements.

Note: A series of repetitive pulses may be fired using the same rod positions on the same day as long as the reactor power is not increasing and is less than 1 kW.

3. Stabilize in the manual mode.
4. Raise the transient rod anvil to the desired pulse position. (This position is ob-tained from the control rod worth curves for the appropriate core operating posi-tion)
5. Select the proper pulse detector according to the table below. If the Cerenkov detector is selected, turn off the reactor room and tank lights.

Detector 1 = Pulse Ion (Maximum insertion = S2.00) Detector 2 = Cerenkov (Maximum insertion = S4.00)

6. Enter Pulse Mode and enter an identifying string at the prompt, The power level must be below 1 kW to enter Pulse Mode.
7. Fire the pulse by depressing the " Fire" button on the reactor console.
8. Record the appropriate data in the reactor operations logbook from the pulse display.
9. Reset R-1 and R-5 to their normal alarm points when pulsing, square wave, or steady state operations are complete.

l Revised: 15 May 91 Page 1 

                                                                                                                                                            ?

OP~ERXTION 

                      ~ owaaw-a~ew        th@ROCEDUREsl[., #              1
                                                                                                - J~IT. . . .. ...ya.. ,s#3n.PiMurai i
                                                                                                                                 .aw .aSiTAB G2 This procedure has been approved by the Reactor Facility Director n n , . . c .- ;,           +
                                                                                                .h              I       lk 7 w -----

7 g , /p Reactor Facility Directffr Date 4

Reviewed by the Reactor Staff fibL Z/ T- t 9 9 /

Wriglft Date i C4. . 4-71-t Geo Date 4%dr e/ML zu, ba'cka Da'te As M94// Sfence 6 ate 

                                                                   ~
                                                                     , -i ' us- ebs 0f//fmf/
                                                                                                       ' D6te Laugherf dw Mwst () M nw, ri j
                                                                  $ uyed "                                D6te t

gAwhjau N/nn4/ Dwens D ate - 

                                                                                                          -Date Date Reviewed by RRFSC 24 SEP1991 D ate .

R 

                                           - *    ,s   , ' '            '

OPERSUONACO.PROCEDOREIMM _. m m eam-au ,7* " [n -;kM_N[Pidedarc ..m 8, TAB z G2 h PULSE OPERATION (SUBCRITICAL) G2NERAL Pulses above S2.00 must be approved by the RFD (prior to pulse initiation). Spec-ification on the RUR may be used to meet this requirement. SPECIFIC

1. Set the alarm points on R-1 and R 5 (criticality monitor) to full scale.
2. Given a core position, set the transient rod at a position corresponding to the dollar value determined by the following equation:

S Value = Total worth (S) Transient rod - Desired pulse (S) Wlue

3. Bring the reactor cold critical using the three standard control rods, use a rod configuration commensurate with core position and experimental requirements.

Note: A series of repetitive pulses may be fired using the same rod positions on the same day as long as the reactor power is not increasing and is less than 1 kW.

4. Stabilize in the manual mode.
5. Select the proper pulse detector according to the table below, If the Cerenkov detector is selected, turn off the reactor room and tank' lights, Detector 1 = Pulse Ion (Maximum insertion = S2.00)
Detector 2 = Cerenkov (Maximum insertion = S4.00)
6. Scram the Transient rod.
7. Raise the Transient rod anvil to 100%.
8. Let the power decay to approximately 1 watt or less.
9. Enter Pulse Mode and enter an identifying string at the prompt.
10. Fire the pulse by depressing the " Fire" button on the reactor console.
  - 11. Record the appropriate data in the reactor operations logbook from the pulse .                  1 display.
12. Reset R-1 and R-5 to their normal alarm points when pulsing, square wave, or steady state operations are complete.
 . Revised: 15 May 91 Page1
                                                                                         .,w--t     -c

i l OPERATIONAL ' PROCEDUR!! g. . Procedure Si TAB H This procedure has been approved by the Reactor Facility Director . , r . . . .a __ kl j 29 [9p Reactof Facility Director / Date  ! Reviewed by the Reactor Staff 

                                         .                         l 170k4/               '/ 29-9/

Wrigh't Date Lw ,w "- 21 9 l G eor e Date Ib \ l _ 7 4 mat 

                                                 ~

sbacka Date 

                                . st/tu._            2744/7/
                                $ bence                  # tate 9A da b>              Mpfm f/

Laugh 1!ry j B' ate Mu M nya " 21 %y4 f fyyen#' Date { T/m dw 226 2 4/ Oivens Date Date l Date Reviewed by RRFSC 24 SEP1991 Date h bb hkkkkhhhh[hhhhhh![U 

                        ?!$ M *                                                                                                                                   $1h WEEKLY OPERATIONAL INSTRUMENT CHECKLIST CllECKLIST #                                                      DATE, SUPERVI5ED BY ASSISTED DY ,_                                                    REVIEWED DY                                                  _._
l. WATER LEVEL INDICATOR I

l A, in pool, cast sidc. deptess float on watc' level indicator...................... ....... l B . O b ser v e ser a m o n co n so le . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .! . . . . . . . . .

11. WATER RESISTIVITY <

List resistivity readings for previous week from daily startup checklists. Deterrnine the average at each point is > 0.5 Mohm cm. MON TUE WED THU FRI AVO Monitor Box DM1 - ~ ~ DM2 ' a l 111. RADIATION ALARMS A. Test alarm functions for high level and failure Monitor Failure alarm functional HIGH Level alarm functional R1 __ R2 R 5 (criticality) E3 E6 Reactor Room CAM Oas Stack Monitor 

                                                                                                                        ~

B . R e set ala r m s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . IV. OTH ER A. Top lock key seals at Security Desk and at LOG verified intact............... B. Ch a n ge Filt er in t h e St aek O a s M onito r .................................................._ _ I T AFRRI Form 66(R) Revised: 15 May 91 i i

OPERATIONAL PRONDUREi Procedure 8,TABI I Trhis procedure has been approved by the Reactor Facility Director r ( , blfh A __ E 

                                                                                                                                       /

Reactor Facility Directf/ Date Reviewed by the Reactor Staff 

                                                                                              ,weM          $~ 2.9. 't/

Wriglft Date 

                                                                                           /, G<, , a         T-74 -9 I Date
                                                                                         /idor I r/L w -s F gbacL4a                Date
                                                                                         ,          u.. 8%'lty6/

Spence 'Date 9M $nid4 197/or4/ Laugh 6fyQ tate 0 V w k nr " 1 % u Al Da%e [q'guyenihtthdra ??nar1/ Owens Date Date Date Reviewed by RRFSC 24 SEP1991 Date 

                                                               $3$$$$$[N((Mh5NE[hfh8NikfE$!D$2iE6ddi,                                            1
                                                                                   ..                     -                                        i

t Ot'EP ._.T10_2{AITT,- PROCEDURB. 

                                                - -                                                                                         l-~--em.-
                                                                                                                                              .: 4 m 4. Procedure 8. TAD I D AILY OPF.7 ATIONAL SHUTDOWN CIIECKLIST Checklist No.                                                                                                               Date Time Completed ,_                                                                                                           Supervised by Assisted by          ,,
1. REACTOR ROOM (Room 3161)
1. A 11 r o d d r i v e s D O W N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2. C a r r la ge ligh t s O F F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .
3. D 0 0 r 316 2 S E C U R E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4. D o o r 3161 lo c k ed wit h k e y .. .... .. .. . . . . .. . . . . . . . . . . .. . . . . .. .. . . .. . . . . .. . . . . .. . . .. .. . . .

FM

11. EQUIPMENT ROOM (Room 3152)
1. Distillat10n unit discharge valve CLOSED .....................................
2. A ir d rye r O P E R AT I O N A L . .. .. .. .. . . .... . . . . .. .. . . .. . . . . . . . . . . . . . . .. .. .. . . .. .. . . .. . . .
 ,    3. Doors 231, 231 A, 3152 and Roof hatch SECU R ED .....................

111. EQUIPMENT ROOM (Room 2158)

1. Prim a ry disch a rge p r essu re (P SI) ................ .............................. .....

L. D emineralizer flow ra tes set to 6 G PM .......................................... ,,,,,,

3. Visu a l in sp eet to a fo r le a k s ... .. .. .. .. . . .. .. .. . . .. . . .. . . .. .. .. . ... .. . . .. . . .. .. . . . . .. .. ...
4. D o o r 215 8 S E C U R E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

IV. PREPARATION AREA

1. ER 2 plug door CONTROL LOCKED; D oor closed; and h a ndwh eel PA D LOCK ED ................................
2. ER 2 ligh ts ON a nd rh co sta t a t 10% ..............................................
3. ER 1 plug door CONTROL LOCKED; D oor closed; an d handwheel PAD LOCKED .............................,.
4. E R 1 ligh t s O N a n d r h eo st a t a t 1(, 4 ......... .. .... .... ... . ................ .......
5. Visu a 1 in sp eet io n o f a r ea .............. ..... .. .... .... .. .. .... . . . . .. . . .. ... . .. . . .. .. .. . . ..

AFRR1 Form 62(R) Revised: 15 May 91 Page1 

                                                                                                                                                                                      \

V. LOBOY ALARM LobbyaIarmaudtoON........................................................................ L_._... l 1 VI. REACTOR CONTROL ROOM (P am 3160) I 1. R e a et o r t a n k ligh t s O F F .. .. ...... .. .. . . . .. . . .. .... .... .. ... . .. .. . . .. . . . . . . . . . . . .

2. Con sole ch a r t r ecor d er pen s r aised ................................................. i 3.TVmonitorsOFF...........................................................................
4. Console LOCKED, and all required keys returned IoloCkboX......................................................................................
5. D iffu ser a n d seco n d a ry pu mp s O F F ................ .............................. .
6. Pu :ificat ion a n d p rim a ry p u m ps O N ............................... .. .. ...........
7. R eaet or mon t hly u sage su mm a ry complet ed ..................................
8. Exposure rcom camera power supply turncd OFF ........................
9. R a d l a t t o n m o n it o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

MONITOR READING HIGH LEVEL ALARM SETTING (mr/hr) 'j

a. R 1 20 _
b. R.? _. N/A
c. R 3 N/A
d. R 5 20
c. E 3 N/A
f. E 6 , N/A
g. R 6 N/A s

AFRRI Form 62(R) Revised: 15 May 91 Page 2 

                                                                                                                                                                             )

l 

            - _ . .. .- _..--_                -     _ ~ - _ _         = . _ . -       .-       -. . _ _ .       .      _.-

OPERATIONAL PROCEDtJRE Procedure 9 

  )

This procedure has been approved by the Reactor Facility Director I 

                                                                                         'I!p Pjj;                         l f
  ;                                         [A                'bt                                     fL9 f9 r l                               Reactor Facility Director g                                             Date i

Reviewed by the Reactor Staff l f%4vN f LY* '// Wrigfit Date ett.<,,w G-71^) 

                                                                        ?tt ,st F sbacka                           Date
                                                 .:fl %              N/f%1l Sjience                       "D a t e M $nda, DG95% Q Lau h6yj                         4 ate
                                             & M1 hwsw

Ynva, A) 14 u'en#~ Dilte i s h AYC7i Owens Date Date Date Reviewed by RRFSC 24 SEP1991 Date M4 = I * * 

                                                                                                                   .,j

                                                                                                                            *T55Ea '****

0?E?^?2 W. OWO9$ *iL , . REACTOR ROOM SAFETY GENERAL The following safety procedures will be observed while in the reactor room. SPECIFIC

1. lioist Operations: Perform the following before/during any hoist operations:
a. Inspect any lifting equipment (ropes, cables, etc.) for wear or dam-age prior to use,
b. rinsure that the hoist has n current load testing (within last 12 months) ,
c. Ensure areas beneath the hoist are clear of personnel when opera-tions are underway. This is particularly important when using the hatches between several floors.
d. Each time a load approaching 10,000 pounds is handled, test the brakes by raising the load a few inches, applying the brakes, and check-ing for slippage.
c. Ensure a load is not lowered below the point where two full wraps of cable remain on the drutn.
f. Ensure no tools or poles longer than 10 feet are raised vertically in the reactor room while the power rails are energized.
2. Mercury or mercury compounds in any form are not allowed in the reactor room at any time.

Revised: 15 May 91 Page 1 

                                  .. v+

LO - ckDifR ad., ~ OPERXTIWA,a,s..pk6.a.v.ws_d?jW C,.en JE~IOV.w;FM:W%aL,c 

                                                                                                                    ,-        aman s Pro,cedure e           10 j                                                                                                                           i

. iThis procedure has been approved by the Reactor Facility Director 

                                                                   '[A                  Ia                          fMllidli61[L9
                                                                                                                    ;   aua u m Date          1/

Rea or Facility DirectoPa 

                                                        -   A                     A 8A'                                      2 9' h e 9f Chair 6an, Sfafety'anii fferth Department ' Date Reviewed by the Reactor Staff 1:wL]f                             T. Ls9/

Wright' Date Ilb,,,a C' 24 -91 d( k?9 ,, a f sbackh - D at'e Ee 21f%4f fpenJe 'Date 9HJa d< MA f/ Laugher Diite [ Q1 ion" + MhtvAl yen V Da(e L kl.n NOhaf / OWens Date Date Date Reviewed by RRFSC 24 SEP1991 l Date r, y . 

                                                                                                                                               ,gggg 74:etg4g

_ . _ . . _ . , . . _ _ ~ . . _ . _ _ . - , . _ _ 

                                                                             ..c. .a 3 STACK GAS MONITOR PROCEDURE GENERAL This procedure specifies all the requirements for operation of the Stack Oas Monitor (SOM)in the reactor room. This instrument is used to sample and measure the gaseous effluent in the building exhaust system.

SPECIFIC A quality assurance check (QA) is performed daily, prior to reactor operations, as part of the reactor start.up. This check is performed in the following manner:

1. The particulate filter is changed if necessary.

2 The front cover of the detector shield is removed and the check source is inserted all the way in to the face of the detector. The blue alert light should come on as the count rate rises above the alert setpoint.1 he red high level alert light and bell should come on as the count rate rises above the high level set point. The audible alarm can be silenced by pushing the rec button on the front of the SOM cabinet.

3. The detector voltage system set point is checked by pressing menu function 5 and then menu function 1. The high voltage set point is displayed under itern 7 on the display screen. Press 0 twice to return to the main menu.
4. The air sampling flow rate (should be greater than 3.5 cubic feet per minute).
5. The counts per minute reading shown on the display should be checked against the plot of counts per minute versus Julian date to determine if it falls within the plus or minus 5% deviation lines for the detector and check source. Ifit does, the cheek source should then be removed and the detector cover replaced.

if the counts per minute consistently fall outside the +/ 5% window, it is considered an abnormality and should be reported immediately to the Reactor Facility Director and to the Safety and Health Department.

6. The SGM alarms will be acknowledged by pushing the "ACK" button on the SGM keyboard.

t Revised: 15 May 91 Page1

1 ortxArio. -o NAG.o_m.enoc.._rp0d@_.!.";f T > , %+ w' 1^~T#'~"!. o w h . ~ 2M !W x 2. M . ,s f 1i _:rioc.a , , This procedure has been approved by the Reactor Facility Director [_ 17 . Jh in y ,$/'2.9 [77 Reac r / Facility Direct & . JUS U J=U Date ikA$ 

                                                     'hairian, Safet'y a'nd Trealth Department # Date
                                                                                                            **c7 h k u 9/'

Reviewed by the Reactor Staff f ubX4 T l *! 1) Wright Date 4 c<,e s t4 9j co Date 4 dr e/ 1 / 4 w a ,, F9 sbdeka Date 42>;+rra.- e#7kA'/ 

                                                                         ~ 3l 5ence      ,             'D ate M 1.-ric< M//fu4/

L u hfry -Date i f.4 h.Mmw4f ~ 6 uyefi ' ' Date 

                                                                          / h AL '26sv t

Gwens Date Date Date Reviewed by RRTSc 24 SEP199' Date

0(ERADON.:RPf5e M iE2 W HFiIIE 2 E?E N B E hr*F AIR PARTICUL YE MONITOR PROCEDURE GENERAL This procedure specifies how to test the CAM to ensure proper operation of this monitoring device. SPECIFIC This procedure uses a radioactive source to test the alarm set points of the CAM,

1. OPERATING and TRACING Observe to see that the CAM is operating and tracing.
2. ALARM TEST WITH SOURCE Open the detector chamber door and slowly bring a radioactive source near the detector. Observe the meter on the fron' of the CAM. The yellow light will oc come on at approximately 4,000 counts per minute. The red light will come on at approximately 10,000 counts per minute, the abrm will sound and the dampers will close. Reset the alarrr., close the chamber door and replace the source in the drawer.

Revised: 15 May 91 Page1

ATTACHMENT C - Amendment No. 21 to Facility Operating License 

                             -- _ - - _ _   u

f * %, 

                 /                        o,
              !                      / i                                                    UN*TED STATES
             !)                  ,.. ,4( [3                     NUCLEAR RELLLATORY COMMISSION
               *, 84                                                                  WA$HINGTON. D C. 2tF66 ARMEDFORCESRAD10B10LOGYRES{ARCHINSTITUTE DOCKET NO. 50-170 AMENDMENT TO FACILITY OPERATING LICENSE
  • Amendment No. 21 License No. R-84
1. The U.S. Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment to facility Operating License No. R-84 filed by the Armed Forces Radiobiology Research Institute (the licensee), of April 30, 1990, as supplemented on December 17, 1990, March 5, 1991, May 17, 1991,' August 16, 1991, and September 10, 1991, complies with the standards and re Energy Act of 1954, t.s amended (the Act)quirements of the Atomic 

                                                                                                                             , and the Consnission's        rules and regulations as set forth in 10 CFR Chapter I; B.      The facility will o)erate in conformity with the application, the provisions of t1e Act, and the rules and regulations of the Consnission; C.      There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comunission's regulations; D.       The issuance of this amendment will not be inimical to the coninon defense and security or to the health and safety of the public; E.       The issuance of this amendment is in accordance with 10 CFR Part 51 of the Consnission's regulations, and all applicable requirements have been satisfied; and F.       Prior notice of this amendment was not required by 10 CFR 2.105(a)(4),

and publication of notice for this amendment is not required by 10CFR2.106(a)(2).

2. Accordingly, paragraph 2.C.(1) of Facility Operating License No. R-84 is hereby amended to read as follows:-

(1) NaximumPowerLevM AFRRImayoperatethereactoratsteady)statepowerlevelsup to a maximum of 1100 kilowatts-(thermal , and at pulse power levels not to exceed a pulse reactivity insertion of 4.00 dollars. 

             ~                 _        _              _ , _ . _ _ _ _ _ . . _ _ -

2-

3. Accordingly, the license is amended by changet. to the Technical Specifications as indicated in the enclosure to this license aniendment, and paragraph 2.C.(2) of License No. R-84 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications co.itained in Appendix A, as revised through Amendment No. 21. are he eby incorporated in the license. The licensee shall operate the fa;ility in accordance with the Technical Specifications. 4 This license amendment is effective as of its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Thomas . Nur , Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Changes Date of Issuance: octot>er 8,1991 

  ._._ _ __-                            ___r                      _ . . _ . . _ _ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ - _ _.

ENCLOSURE TO LICENSE AMENDMENT NO. 21 FAClll1Y OPERATING l! CENSE NO. R 84 DOCKET NO. 50-170 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. 1 Remove insert i i 2 2 7 7 8 8 22 22  ; , 23 23  ; 25 25 26 26 27 27 28 28 29 29 30 30 , 31 31 32 32-34 34 35 35-f 9 v e-e--= --r --e-..--a,- r-, ww .a-v,v *w, --,re,--,v,v- .-y,,,- v.,my- ,rw-,-w,>,v,,,w , y-w - ,w , , ~,v,,-,vwry~,,eiew r-,o w gv ,mvr -c' g me~- Sw=t- g c, , %yg,'s,--: w

TECllNIC AL SPECirtC.4TIONS Foil Tile AfitRI ItEACTOR FACILITY LICENSE NO R 84 DOCKET #50170 TABLE OF CONTENTS M i0 DErlNITIONS 1.1 ALARA 1 1.2 Channel Calibration 1 1.3 Channel Check i 1.4 Channel Test i 1.5 Cold Critical 1 ' l.6 Core Grid Position 1 1.7 Experiment i 1.s Experimental Facilities 1-1.9 Fuel Element 2 1.10 instrumented Element. 2_ 1.11 Limiting Safety System Setting 2 l 1.12 Measured Value 2 ! 1.13- Measuring Channel 2 l 1.14 On Call 2 Operable 1.15 2 1.16 Pulse Mode 3 1.17 Reactor Operation 3 1.18 Remeter Safety Systems 3 1.19 Reactor Secuted 3 1.20 Reactor Shutdown 'S 1.21 Reportable Occurrence 3 1.22 Safety Channel 4 1.23 So.fety Limit 4 1.24 Shutdown Margin 4 1.25 Standard Control Rod 4l 1.26 Steady State Mode 4 1.27 Transient Rod 4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 3 i 2.1 Safety Limit . Fuel Element Temperature 5 i 2.2 . Limiting Safety System Setting for Fuel Temperature 6 - - - 3.0 LIMITING CONDITIONS FOR OPERATIONS 7 3.1 Reactor Core Parameters -- T . 3.1,1 Steady State Operation 7 3.1.2 Pulse Mode Operation 7 

                 - 3.1.3 Reactivity Limitations                                              6 3.1.4 Scram Time                                                           8 1

Amendment NJ. 21

e. Henetor Pool
d. Core Experiment Tube
e. Portable Detm Tubu
f. Pneumatic Transfer System
     .. Incore Locations 15    117,L EL Ehf ENT A fuel element is a single TRIG A fuel rod, et the fuel portion of a fuel follower controi rod.           l t 10 INSTRUMENTED ELEMENT An instrumented element is a special fuel element in which sheathed chromal/alumel or equivalent thermocouples ue embedded in the fuel.

1.11 LIMITING S AITTY SYSTEM FPTTING limiting safety system settings are settings for automatic protective devices related to those variables having significans safety functiott 1.12 MEASURED VALUE A meuvred value is the magnitude of a variable u it appears on the output of a aneuuring channel. 1,13 MEASURING CllANNEL A measuring channel is that stombinatien of sensor, interconnecting cabla or linn, amplifiers, and output device that ue connected for ti.e purpoet of measu*ing the value of a variable. 1.14 E CALL A person is considered on call if

a. The individual hu been specifically designated and the operator knows of the designationi
b. The individual keeps the operator poeted as to his/La whernabouts and telephone number; and
c. The individual is capable of getting to the reactor facility within 80 minuta under normal circumstantw.

1.15 OPERABLE A system channel. device, or component shall be considered operable when it is capable of performing-its intended function (s) in a normal manner. 3 Amendment No. 21 1 . 

                                                                                                                   .1

30 LIMITING CONDITIONS FOR OPEft ATIONs 3: ItE ACTOH CORE PAR AMETEltS 3.t.1 fTEADY STATE APEH ATION A tohe abilit y This specification applies to the manimum reacter power attalned during steady state operation. ON*etive To assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a set point for the high flus litniting safety systerns, so that automatic protective action will prevent the safety limit from being reached i during steady state operations, Erecificatleris l The remetor steady state power level shell r.

  • exceed 1.1 rnegawatte. The normal i

steady state operating power limit of the reactor should be 1.0 megawatt. For l l purposes of testing and calibration, the rtactor may b, operated at power levels not to exteed 1.1 megawatts during the testing period. buit Thermal and hydraulic calculationr and eterational experience indicate that TRIG A fuel may be safely operated up to power levels of at leut 1.5 megawatts with natural convective cooling. 3.1.2 PULSE MODE OPERATION Arrhes.kjjity. This specification applies to the muimum thermal energy produced in the teactor u a result of a prompt critical insertinn of reactivity. Obieetive The objective is to usure that the fuel temperature safety limit will not be exceeded. Specification The maximum step insertion of reactivity shall be 2.8% Ak/k (84.00) in the pulse mode. Dbtil Based upon the Fuchs-Nordheim rr athematical model (cited by C.E. ClifTord et al. in the April Reador"), 1961 GA an insertion ReF. of 2.8, rt #2119, Ak/k "Model results in of theaverags a maximum AFRRI TRIGAfuel temperature of less than 550'C, thereby stayin,g within the limiting safety settings that protect the safety limit. The 50 C margin to the Limiting Safety 7 Amendment No. 21

System Betting and the 450'C margin to the safety limit amply allow for uncertainties due to extrapolation of meputed data. accuracy of meuured data, and location of instrumented fuel elements in the core. l

3.1.3 REACTIVITY 1.lMITATIONS 1

A r+1ie nbilit y 4 These specifications apply to the reactivity condition of the reactor and the i ] reactivity worths of control rods and experiments. They apply for all modes of l l operati)n. > Obiective e The objective is to guarantee that the reactor can be shut down at all timo ' and that the fuel temperature safety limit will not be exceeded. St-iMettlug 

                                                                                                                                                                            ?
a. The reactor shall not be operated with the mutmum available excus remetivity sbove cold critical with or without all experiments in place greater than 86.00 (s.s% Ah/k),
b. The minimum shutdown margin provided by the remaining control rode with the most reactive control rod fully withdrawn or removed shall be 80.60 (0.365 Ak/h) for any condition of operation.

l 11A111

a. The limit on available excess reactivity utablishn the maximum power if all control elements are removed.
b. The shutdown margin usurn that the reactor (an be shut down from any operating condition even if the highut worth control rod remains in the i fully withdrawn position or is completely removed.

3.1.4 SCRAM TIME Aeollembility The specification applit to the time required _to fully inert any control rod to a-full down position from a full up position. , Oblective > The objective is to achieve rapid shutdown of the reutor to prevent fuel damage. Specification The tisne from scram initiation to the full insertion of any control. rod from a full up position shall be less than I second. f k 8 Amendment No. 21 L. ,_.u__ _ _ _ . _ . _ . _ , _ _ _ , . _ _ . _ _ . _ . _ _ . _ _

E,kffjfit ation Functional checks shall be made annually, but not to exceed 15 months, to insure the following:

a. With the lead shield doors epen, neither exposure room plus door can be electrically opened,
b. The core dolly cannot be enoved into position 2 with the lead shidd doors cloud e The wuning horn $all sound in the exposure room before opening the lead shield door, which allows the core to move to that exposure room unless cleued by two licensed operators.

U.hNL These functional checks will verify operation of the interlock system. Experience at AFRRI indicatn that this is adequate to insure operability. 4.2.5 REACTOR FUEL ELEMENTS Acelleability This specincation applies to the surveillance requiremerits for the fuel elements. Obiettive The objnlive is to verify the integrity of the fuel element cladding. Soetifications All the fuel elements prennt in the reactor core, to include fuel follower control rods, shall be insputed for dart. age or deterioration, asi meuured for length and bow at intervals sepusted by not more than 500 pulsa of inurtion greater than 

           $2.00 or annually (not to exceed 15 months), whichever occur .nret . Fuel elements in long term storage need not be measured until returned to core; however, fuel elements routinely moved to temporary storage shall be meuvred every 500 pulses of insertion greater than 82.00 or annually (not to exceed 15 months), whichever occurs fint.

Elst The frequency of inspection and meuvrement is bued on the parameters most likely to affect the fuel cladding of a pulse reactor, and the utilisation of fuel l elements whose characteristics are well known. The limit of transverse bend has been ebown to rwult in no dimculty in disusembling the core. Analysi4 of a woret case scenario in which two adjacent fuel elements suffer sufficiently severe tranevme bends to twult in the touching of the fuel elements has shown that no darnage to the fuel elements will roult via a hot spot or any other known mechanism, 22 Amendment No. 21

4.3 COOL ANT FYSTEMS Attlitatilitz This spec:ication applies to the surveillance requirements for monitoring the pool water and the water conditioning system. Objective The objective is to usure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating possible radiation huards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other component, in the primary water system. S t.*eific ations

a. The pool water temperature, u meuurid near the input to the water purification system, shall be meuured daily, whenever operations are planned.
b. The conductivity of the water at the output of the purification system shall be meuvred weekly, whenever operations are planned.

EAdi Dued on experience, observation at then intervals provides acceptable surveillance of l limits that usure that fuel clad corrosion and neutron activation of dipolved materials will not occur. 4.4 VENTIL ATION SYSTEM Aeolienbility This specification applies to the facility ventilation system isolation. Obleetive The objective is to usure the proper operation of the ventilation system in controlling the releue of radioactive material into the unrestricted environment. Eggtification The operating mechanism of the positive naling dampees in the reactor room ventilation system shall be verified to be operable and visustly inspected at leut monthly (interval. not to exceed six weeks). Datig Experience accumulated over years of operation hu demonstrated that the tests of the ventilation system on a monthly basis are sufficient to assure proper operation of the system and control of the releue of radioactive material. 23 Amendment No. 21-l

50 DESIGN FEATUllCS 5i SITE AND FACILITY DEscitII' TION Acclieability This specification applies to the building that houses the reactor. Obi etive The objective is to restrict the amount of radioactivity releued into the environment. So,eifiestions

a. The reactor building, u a structurally independent building in the AFRill complex, shall have its own ventilation system branch. The effluent from the reactor ventilation system shall exhaust through absolute filters to a stack having a minimum elevation that is 18 fut above the roof level of the highest building in the AFRRI complex.
b. The reactor room shall contain a minimum free volume of 22,000 cubic feet,
c. The ventilating system air ducts to the reactor room shall be equipped with positive sealing dampers that are activated by fail safe controls, which will-automatically close off ventilation to the reactor room upon a signal from the reactor room air particulate monitor.
d. The reactor room shall be designed to rwtrict s.it leakage when the positive sealing dampers are closed.

EElill The facility is designed so that the ventilation system will normally maintain a negative pressure with respect to the atmosphere, so that there will be no uncontrolled leakage to the ervironment. The free air volume within the resctor building is confined when there ! is an emergency shutdown of the ventilation system. Building construction and gukets around doorways help rwtrict leakage of air into or out of the reactor oom. The stack height insures an adequate dilution of effluents well above ground level. The separate ventilation system branch lasures a dedicated air flow system for reactor effluents.

  • 5.2 BEACTOR CORE AND FUEL 5.2.1 REACTOR FUEL Arcticability These specifications apply to the fuel elements, to include fuel follower control rods, used in the reactor core.

Obiective The objectiva are to (1) usure that the fuel elements are dwisned and fabricated in such a manner u to permit their use with a high degree of reliability assure thatwith twpect the fuel to their elements usedphyelcal in the coreand nuclear characteristics, are substantially those analyse and (2) d in the Safety Analysis Report. Soeeifications The individual nonirradiated standard TRIG A fuel elements shall have the following characteristics: 25 Amendment No. 21

i n. Uranium content: Muimum of 9.0 weight percent enriched to leu than 20% uranium.386. In the fuel follower, the maximum uranium content will be 12.0 welght percent enriched to less than 20% uranium.235, b, llydrogen.to elttordum atom ratio (in the Zril ): 1.0 Zr atoms with a range between 1.6 and Ih. Nominal 1.7 li atoms to

c. Clatiding: 604 etainlese steel, nominal 0.020 inch thick.

d. Any buenable poleon und for the specific purpose of compensating for fuel burnup or lont. term teactivity adjustmente shall be an integral put of the manufactured fuel elements. Undt A muimum uranlurn content of 9 weight percent in a standard TRIGA element le greater than the design value of 8.6 weight percent, and encompuses the maximum probable variation in individual elemente. Such an increue in bding would result in an increase in power density of len than 6%. An increue in local power density of 6% in an-individual fuel element reduces the safety mugin by 10%, at most. The hydrogen.to elrconium ratio of 1.7 will produce a rnaalmum prenure within the cladding well below the rupture strength of the cladding. The local powe, denalty of a 12 0 weight percent fuel follower is 21% steater - than an 8.5 weight percent standud Tit!GA fuel element in the D Ring. The volumefuel TRIGA of element. fuel in a fuel followed rod is $6% of the volume of a standud Therefon the actual power produced-in the fuel followed rod D-ring. the is 83% less than the power , produced in a standed TRIGA fuel element in 6.2.2 REACTOR CORE Adelienbility These specifications apply to the configuration of fuel and in core experimente. Obiective The objective le to restrict the arrangement of fuel elemente and experimente so as to provide moeurance that excessive power densities will not be produced. Soecifications a. The reactor core shall consist of standard TRIGA reactor fuel elemente in a close packed array _ and a minimum of two thermocouple instrumented TRIG A reactor fuel elemente. 

              -b.

Ther, shall be four single core positions occupied by the three standud control rode and transient rod, a neutron start up source with holder, and positions for poselble in core experimente, e. The core shall be cooled by natural convection water bw, s d. In core experimente shall not be placed in adjacent fuel positions of the B-ring and/or C ring. 26

Amendment-No. 21.

I

4

e. Fuel elemente indicating e.n elonsstion great:r thin 0.100 inch, a lateral bending treater than 0.0625 inch, or significant visible damage shall be considered damaged, and shall not be und in the reactor core.

Datia Standard TRIG A cores have bun in un for years, and thelt safe operational characteristice are well documented. Experience with TRIGA reactore hu shown that fuel element bowing that could result in touching hu ouvrred without deleterious effects. The elongation limit hu been specified to (a) usure that the cladding matnial will not be subjected to strence that could caun a non of integrity in the fuel containment, and (b) nasure adequate coolant flow. 5.2.3 CONTROf RODS Arctic ability These specifications apply to the control rode und in the reactor core. Obiettive The objective is to usure that the control rode are dulgned to pumit their un with a high degree of reliability with respect to their physical and nuclear characteristice. Specific ations

a. The standard control rode shall have scram capability, and shall contain borated graphite, B4C powder, or boron and its compounde in solid form as a poison in aluminum or stainless eteel cladding. Thue rode moy have an aluminum, sir, or fuel follower. If fuel followed, the fuel region will conform to the Specifications of 5.2.1.
b. The transient control rod shall have uram capability, and shall contain borated graphite, B C powder, or boron and its compounde in polid form u a poleon in alumindm or stainlue stul cladding. This rod may incorporate an aluminum, poison, or air follower.

Eatig The poleon requirements for the control rods are estisfied by using neutron. absorbing borated graphite, B C powder, or boron and its compounds. Thue materials muel be contained i,n a evitable cladding material, such u aluminum or i etainless steel, to insure mechanical etability during movement and to leolate the ' poison from the pool water environment. Scram capabilities are provided by the rapid insertion of the control rode, which is the primary operational omfety feature of the reactor. The transient control rod is designed for use in a puleing TRIGA reactor. 27 , Amendment No. 21 1

6.3 S'ECI AL Nt.' CLEAR MATERI AL STOR AGE Mdicat1hty This specification applies to the storage of reactor fuel at times when it is not in the remeter core. Obieetive The objective is to usure that stored fuel will not become critical and will not reach an Unskfe temperature. St et(fit tti9n All fuel elemente not in the teactor core shall be stored and handled in accordance with appliemble regulations. Inadiated fuel elements and fueled devices shall be stored in an array that will permit sufficient natural convective cooling by water or air, so that the fuel element or fueled device temperature will not exceed design values. Storage shall be such that groups of stored fuel elements will remain suberitical under all conditions of moderation. 

       !!alig The limits imposed by this specification are conservative and usure safe storage and                          l handling. Experience shows that approximately 67 fuel ele:nents are required, of the design used at AFRRI, in a closely packed array to achieve criticality. Caicuistions show that in the event of a full storage rack failure with all 12 elements falling in the most reactive nucleonic configuration, the mus would be less than that required for criticality.

Therefore, under normal storage conditions, criticality es,nnot be reached, t 28 1 Amendment No. 21

eo M)MINISTR ATIVE CONTROLS e! QRG AN17 ATIOS 6 1.1 ST R t'CTt'R E The organisation of personnel for the management and operation of the AFRRI reactor facility is shown in Figure 1. Organtiation changes may occur, bued on Institute requireniente, and they will be depicted on internal documents. However, no changes may be made in the Operation, Safety, and Emergency Control Chain in which the Reactor f acility Director hu direct responsibility to the l Director, AFRRI. thrector. AFRRI g, , 4% ,,, Ore'aumal. Safet). Reactor and Radiauen Factluy Safety Commutee Safety and Health Expt. comrut Y.$'8"7 , l l Chaarnian. I { {

  • Ra&auon i Sources Dept. {

{ Advm Abitv) l l l Rextor Factlary Dusetor ,,,,,,,,,,,,,,j l Reactor Opersoon Supervisod I Reactor Operanons Sta!P rine. t. o ri..i..ii.. .: r er.....i v.e u...e.... ..a o,. . i.. .e 36. A rax: n..si., r..iiit y. A.y e..et.t st.ff m.mb.e bu useos 4. ch. IMr.et.: f.e m .46et. .f e.foty. 29 Amendment No. 21 3

e.l.2 EIMPONSIBILITY The Director, facility. The ReactorAFRHi,Facility shat ave Directorlicense resp)asibility (RFu for the reactor shall be responsible for administration and operation of the Reactor Facility and for determination of applicabihty of procedures, experiment authorientions, maintenance, and operations. The RfD may designate an individual who meets the requirements of Section 6.1.3.1.a to discharge the RFD's responsibilities in the RFD's absence. During brief absences (periods less than four hours) of the Reactor Facility Director and his designee, the Remetor Operations Supervisor shall discharge these responsibilities. 6.1.3 STAFFING 6.1.3.1 Selection of Personnel

a. Reactor Facility Director l At the time of appointment to this position, the Beactor Facility Director shall have 6 or more years of nuclear experience liigher education in a ecientific nuclear engineering field may fulfill up to 4 years of experience on a onedor one basis. The Facility Director must have held a USNRC Senior Reactor Oprator license on the AFRRI reactor for at least 1 year before appointment to this position.  ;
b. Reactor Operations Supervisor (ROS)

At the time of appointment to this position, the ROS shall have 3 l years nuclear experience. Illgher education in a science or nuclear engineering field may fulfill up to 2 years of experience on a one-for one bula. The ROS shall hold a USNRC Senior Remeter Operator license on the AFRRI reactor. In addition the ROS shall have 1 year of experience as a USNRC licensed Senior kenetor Operator at AFRRI or at a similar facility before the appointment to this position.

c. Reactor Operators / Senior Reactor Operators At the time of appointment tc this positlen, an individual shall have a high school diploma or equivalent, and . " possess the appropriate USNRC license. ,
d. Additional staff u required for support and training. At the time of appointment to the reactor staff, an individual shall poseus a high school diploma or equivalent.

6.1.3.2 Ooerations

a. Minimum staff when the reactor is not secured shall include: ,
1. A licensed Senior Reactor Operator (SRO) on call but' not necessarily on site
2. Radiation control technician on call
3. - At leut one licensed Reactor Operator (RO) or Senior. Reactor Operator (SRO) present in the control room
4. Another person within the AFRRI complex who la able to carry out written emergency procedures, instructions of the operator, or to summon h(lp in case the operator becomes incapacitated 30 Amendment No. 21

_ _ _ _ . _ _ _ _ _ _ . . . .. _ _ _ . _ _ _.s.. __ .. ~. u _

b. Maintenance actisiues that could affect the reactivity of the reactor .

shall be accomplished under the supervision of an SRO.

c. A list of the names and telephone numben of the following personnel i shall be readily available to the operator on dutyi
1. Management personnel (Reactor Facility Director, AFRR1 Director)
2. Radiation safety personnel (Head, Safety and Health Department) l;
8. Other operatione personnel (Reactor Staff ROS) 6.1.4 TRAINING OF PERSONNEL A training ana retraining program will be maintained, to insure adequate levels of proficiency in persons involved in the reactor and reactor operations. >

6,2 RgviEW AND AUDIT . Tile REACTOR AND R ADIATION FACILITY SAFETY ' COMMITTEE fREFSCI 6.2.1 COMPOSITION AND QUALIFICATIONS 6.2.1.1 Ccmoosition

a. Regular RRFSC Members (Permanent Memben)

(1) The following shall be members of the RRFSC. i (a) Chairman, Safety and Health Department, AFRR1 l (b) Reactor Facility Director, AFRR1 (2) The following shall be appointed to the RRFSC by the Director, AFRRI: (a) Chairman se appoln'ed by the AFRRI Directorat., i (b) One to three non AFRRI members who we knowledgeable in fields related to reactor safety. At least one shall be a Reactor Operations Specialist, or a Health Physics Spetlalist,

b. Spelai RRFSC Memben (Temporary Memben)

(1) Other knowledgeable pereone to serve as alternates in item a(2)(b) above u appointed by the AFRRI Director. l l (2) Voting ad ja memben, invited by the Director of AFRRI, to anoist in review of a particular problem.

c. Nonvoting memben as invited by the Chairman, RRFSC.

31 Amendment No.=21 

  - _ __                                    - . - . -         __ _                                                       --.m.-                         __ . _ _-           _ . . _ _ . _ . .

I 6.2.12 Quahfit11}Qat The minimum qualifications for a person on the RRFSC shall be 6 yeus of profeusonal experience in the discipline ce specific field represented. A baccalaureate degree may fulfill 4 years of experierce. 62.2 FUNCTION AND AUT110RITY 6 2 2 i Funetlen The Reactor and Radiation Facility Safety Committn is directly responsible to the Director, Afiliti, The committee shall review all radiological health and safety matters concerning the reactor and its usociated equipmtnt, the structural reactor facility, and thon items listed in Section 6.2.4. 6.2 2 2 Authority The RitFSC shall report to the Director, AFRill, and shall advise the Reactor Facility Director in those ueu of responsibility epecified in Section 6.2.4. 6.23 CIIARTER AND RULES 6 2.3.1 Alternates Alternate members may be appointed in writing by the RRFSC Chairman to serve on a temporary buis. No more than two alternatn shall participate on a vollng buis in itRFSC activities at any one time. 6 2.3.2 Meettnr Freauency, The RitFSC or a subcoramittu thereof shall meet at leut four timu a calender year. The full RRFSC shall mut at leut semi annually. l 6.2,3.3 Quorum A quorum of the RRFSC for review shall consist of the Chairman (or designated alternate) and two other members (or alternate members), one of which must be a non AFRRI member. A majority of thon prennt shall be regular members. 6.2.8.4 Votine Rules Each regular RRFSC member shall have one vote. Each spe<lal appointed member shall have one vote. The majority is 61% or more of the regular and special members pruent a,nd voting. 6.2.3.6 hiinuttu Minutes of the previous meeting shall be available to regular members at leut I week before a regular scheduled meeting. 32 Amendment No. 21

f. Any other area of Facility operations considered appropriate by the RRFSC or the Director /AFRRI.
g. Reactor Facility AL ARA Program. This program may be a seet n of the total AFRRI program.

6s PNoCEDURES 6A 1 Written instructions for certain activities shall h approved by the Reactor Facility Director and reviewed tr the R:actnr and Radiation Facility Safety Committee (RRFSC) The pre:: lures shall be adequate to assure sJe operation of the reactor, but shall not peelude the use of independent judgment an I attion as deemed necessary. These activities are as follows: l

a. Conduct of irradiations and experiments that could affect the operation and safety of the r:.ctor.
b. Reactor start. training program,
c. Surveillance, testing, and calibration of instruments, components, and systems involving nucless safety.
d. Personnel radiation protection consistent with 10 CFR 20.
e. Implementatica of required plans such as the Se urity Plan and Emergency Plan.
f. Resetor core loading and unloading.
g. Checkout startup, standard operations, and securing facility.

6.3.2 Although substantive changes to the above procedures shall be .nade only with approval by the Reactor Facility Director, temporary changes to the procedures that do not change their original intent may be made by the ROS. All such temporary changes shall be documented and subsequently reviewed end approved by the Reactor Facility Director. 6.4 REVIEW AND APPROVAL OF EXPERIMENTS 6.4.1 Before issuance of a reactor authorisation, new experiments shall be reviewed for radiological safety and approved by the following:

c. Reactor Facility Director
b. Safety and Health Department
c. Reaceor and Radiation Facility Safetv Committee (RRFSC) 6 4.2 Prior to its performance, an experiment shall be included under one of the following types of authorisations:
a. Soecial Remeter Antheriatlan for new experiments or experimente not included in a Routine Acactor Authorisation. These experimerts shall be performed under the direct supervision of the Reactor Facility Director or designee.
b. Routine Reactor Authorisation for experiments safely performed at least once. These experiments may be performed at the discretion of the Reactor Facility Director and coordinated with the Safety and Health Department l when appropriate. These nothorisations do not require additional RRFSC review.

34 , 1 i Amendment No. 21 1 1 i i

c. hjp? Parameters Authorisation for routine measurements of reactor parameters, routine core measurements, instrumentation and calibration E cheth, maintenance, operator training, tours, testing to verify reactor outputs, b-other reactor testing procedures. This shall constitute a single authorisattoa. Then operations may be performed or der the authorisation of the Reactor Facility Director or the Resetor Operat . sne Supervisor. 6.4.$ Substantive (reactivity worth more than 280.25) change to previously approved experiments shall be made only after review by the RRFSC and after approval (in writing) by the Reactor Facility Director or designated alternate.- Minor changes thst do not significantly alter the experiment (reactivity worth of Ins than 180.25) may be approved by the ROS. Approved esperiments shall be carried out in accordance with utablished procedures. l 6.5 J1EQUIREI; ACTIONS 6.5.1 ACTIONS TO BE TAKEN IN CASE OF SAFETY LIMIT VIOLATION a. The reactor shall be shut down immediately, and reactor operation shall not be resumed without authorisation by the NRC, b. The safety limit violation shall be reported to the Director of NRC Region-1, Office of Inspection and Erforcement (or designate); the Director, AFRRl; and the RRFSC not later than the next working day, e, A Safety Limit Violation Report shall be prepared,. This report shall be reviewed by the MtFSC, and shall dweribe 11) appiitable circumstances l- i preceding the viola ion, (?) effects of the violation on facility components, ' structures, or systems, and (3) corrective action taken to prevent or reduce the probability of recurrence. d. The Safety Limit Viobtion Report shall be submitted to-the NRC; the Director, AFRRl; and the RRFSC within 14 days of the violation. l 6.5.2 REPORTABLE OCCURRENCES Reportable occurrenen as defined in 1.21 consequences, corrective actions, and measu(res to prevent recurrence) shall b reported to the NRC. Supplemental reports may be required to fully deseribe the final resolution of the occurrence.

a. Promot Notification With Written Followan. The types of events listed .

below shall be reported as soon as possible by telephone and confirmed by 35 Amendment No. 21

4 y %M. ATTACHMENT D 4 v-yp soutine neactor Authorizations

i l Routine Reactor Authorization #1 September 1991 latteduction: The Reactor and Radiation Facility Safety Committee has reviewed and approved t!,e operations described below. These operations have been performed many times in the past and are now considered part of the routine operations for use of the AFRRI reactor. Authorintion: As permitted by NRC license R-84, the Reactor Facility Director, Reactor Operations Supervisor, and NRC licensed operators may perform measurements, conduct operations, effect maintenance, perform experiments, conduct tours, and conduct operatoc training within the scope of AFRRI procedures, Technical Specifications, and NRC regulations. ALARA principles will be followcx! at all times during the conduct of operations.  ; Approved: bl>-r2/ Reactor Facility' 

                                                                                   /            .
                                                              $NMi         ,, SHD R              ~

l NM. Manderfield COL /iiSAF, MSC - Chairman, Reactor and Radiation Facility Safety Committee - [2cactor Radiation F ility Safety Committee mw M. L re CDR J.E. DeCicco, MSC, USN Y~ 0 et A.u [s. Ashby - / g. s CAPT C.B. Galley, fSC, USN A Luersen 

   ##h1
23. VM

i Routine Rcoctor Authorization #2 September 1991 latI9 duction: The Reactor and Radiation Facility Safety Committee has reviewed and approved the dosimetry operations dewribed below. These operations have been safely perfctmed over many years and are classified as routine. Authunzations: As permitted by NRC license R 84, dosimetry instrumentation and other measuring instrumentation may be used alone, attached to, or included with experiments in the reactor radiation facilities within the site boundary, subject to the limitations imposed by Technical Specifications. All dosimetry devices, (quipment or experiments will be used or performed under the supervision of the RFD or his designee. In particular, the dosimetry devices and the experiment shall: o not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table 11 Appendix B, of 10 CFR 20 are exceeded, a not create inventories of I 131 through 1135 greater than 1.3 curies and Strontium-90 inventories greater than 5 millicuries, o limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o be doubly encapsulated if the release of the contained material can cause corrosion to the radiation facility, o have an absolute worth less than $3.00, and o have been inspected and approved by a reactor operator prior to performance of the experiment. The term " dosimetry instrumentation" shall include, but not be limited to: fission chambers, ionization chambers including those with flowing gas, thermoluminesent devices, foils, tablets, and phantoms meeting the limitations listed above. Such dosimetry instrumentation may also be irradiated separately for the purpose of dosimetric calibration or device evaluation and testing. ALARA principles will be followed at all times during the desi and conduct of experiments. Approved; h,.

  • ODv-Reactor Facility Dtrdptx Chairman 'S14D h///$

N.M. Manderficid COL (USAF, MSC Chairman, Reactor and Radiation

Facility Safety Committee

Recctor and Radiation Facility Safety Committee M.I ___oore __/_ M OA 33). Ashby - Ell $d R. Luttsgn l Y6 M. Votti ~ CDR J.E. DeCicco, MSC, USN bb CAPT C.B. Galley, M ,USN

Routine Reactor Authorization #3 September 1991 IntrodKuon: The Reactor and Radiation Facility Safety Committee has reviewed and approved the reactor expenment described below. This experiment has been performed on numerous occasions in past years and is considered routine. Authorintuon5: As permitted by NRC license R-84, The Reactor Facility Director may permit a principal investigator to irradiate animals, animal tissue and other biological materials in the reactor irradiation facilities subject to the limitations imposed by Technical Specifications. This authorization includes shiciding and support materials, sensors, control devices and the use of phantoms and other dosimetry instrumentation authorized under Routine Reactor Authorization #2. In particular, the experiments shall: o not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o not create inventories of I-131 through I-135 greater than 1.3 curies and Strontium-90 inventories greater than 5 millicuries, o limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o be doubly encapsulated if the release of the contained tnateriel can cause corrosion to the radiation facility, o have an absolute worth less than $3.00, and o either have movement precluded or be monitored by a Senior Reactor Operator. ALARA principles will be followed at all times during the ign and conduct of e periments. Approved. h/ ' Reactor Facility Dir@ 

                                                           -yan,gHD
                                                                                  $N        J N.'WI. Manderfield COL, U$AF, MSC Chairman, Reactor and Radiation Facility Safety Commi* tee
                                                           ,                        e

Reactor and Radiation Facility Safety Committee M. [ hQ ' f kore pD. Ashby , LA&J R. Luersp J M. Voth _ CDR J.E. DeCicco, MSC, USN O a o CAN C.B. Galley, ApC, USN i 

                                            =_ -

Routine Reactor Authorization #4 September 1991

== Introduction:== The Reactor and Radiation Facility Safety Committee has reviewed and approved the operations described below. These experiments have been performed on numerous occasions in the past and are considered routine. Authorizations: As permitted by NRC license R-84, the Reactor Facility Director may irradiate biological or non biological materials with an atomic number less than 93 and any experiment stnictural suppon and experiment containers in the reactor irradiation facilities subject to the limitations imposed by Technical Specifications and applicable procedures. !n particular, these experiments shall: o not be able to cause the release of radioactive gases and aerosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o not create inventories of I-131 through I-135 greater than 1.3 curies and Sr-90 inventories greater than 5 millicuries, o limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o be doubly encapsulated if the release of the contained materia _1 can cause corrosion to the radiation facility, o have an absolute worth less than $3.00, and o either have movement precluded or be monitored by a Senior Reactor Operator. ALARA principles will be followed at all times during the design and conduct of ex 'ments. Approved: Reactor Facility Di(eplor pirman, D . 

                                                                                   /4 NW. Manderfield COli USAF, MSC Chairman, Reactor and Radiation Facility Safety Committee

l 2eactor and Radiation Facility Safety Committee I r 

    .       hx)re

_C1  % Ashby[, 

           .1 id[/w
   . Luerip i\      Al
t. votti lDR J.E. DeCicco, MSC, USN

[b e U APT C.B. Galley, MS% USN g ( l

RSDR Forsbacka/mjf/51221 Date: 19 SEP 91 MEMORANDUM FOR RRFSC

SUBJECT:

Request Concurrence to Classify Krypton Gas Irradiation Experiment Protocol Under Routine Reactor Authorization #4 1.

Purpose:

The Naval Medical Research Institute (NMRI) has a requirement for a radioactive inett gas for use in the study of tissue inert gas exchange kinetics and bubble dynamics. Krypton when irradiated produces a isotope, namely Kr-85m, which is useful in this research. Inert gasses under pressures up to 2 atm have been irradiated safely under Routine Reactor Authorization # 102 dated September 4,1990.

2. Experiment

Description:

See attached letter from LCDR Novotny dated 16 SEP 91.

3. Safety Analysis: A worst case scenario accident analysis involves the complete release of the irradiated Krypton gas after a 1 MW-hr irradiation. The gas irradiation vessel has a volume of 446 ml. At a gas pressure of 40 psig, the vessel will deliver 1214 ml of gas at STP (0.0542 moles). The Neutron Activation Tables (Ed. Gerhard Erdtmann,1976) allow us to determine the quantities of radioactive Krypton produced after an irradiation at 1 MW for 1 hour. The composition of the gas is as follows:

Isotope Abundwe On==+ity in Veael Product Half Life "Kr 0.35 % 1.498E-2 g "Kr 34.9 hr "Kr 2.25 % 9.872E-2 g - "Kr 13.3 sec 

             '2Kr               11.6 %                    0.5218 g                   "Kr            1.86 hr "Kr              57.0 %                      2.656 g                    "Kr           4.5 hr-
             "Kr                17.3 %                    0.8157 g                   "Kr           76 min         i The heutron Activation Tables tell us the decays /second per microgram of target element given a 1 hour irradiation in a thermal neutron flux of-10" n/cm'sec, so the total amount of radioactivity produced is as follows:

Radioisotope Total Ona=*ity Produced Concentration in Recctor Room of Released "Kr 9.6 nucroci 1.05E-8 microCi/ml - 

              "Kr                                   1.98E4 microCi-           2.16E-5 microCi/mi "Kr                                   7.39E5 microCi           -8.05E4 microCi/ml "Kr                                   5.48E4 microci            5.97E-5 microCi/ml s
              "Kr                                   7.01E3 microci            7.64E-6 microCi/ml The free volume of the reactor room is 32,400 ft', so the concentranon of radioisotope in the reactor room is found by dividing the total quantity produced by the free volume. Adding the above quantities of radioactive matenals yields a total of 0.82 Ci of radioactive laypton.

l 10 CFR 20, Appendix B specifies the allowed concentrations of radioactive mHerials in air in restricted and unrestricted areas based on inhalation for 40 hours per week for a 13 week period. These limitations are as follows: Allowed Concentration Allowed Concentration Radiniiolene in RestricleiArca ULUntestricted Area 

          Kr                  3E-9 microCi/mi                       lE-10 microCi/mi
          'Kr                    IE-6 microCi/ml                     3E-8 microCi/ml "Kr                    IE 6 microCi/ml                      3E-8 microCi/ml "Kr                   6E-6 microCi/ml                       IE-7 microCi/mi "Kr                    !E-6 microCi/ml                      2E 8 microCi/ml Given the condition of the total release of the irradiated krypton gas, it would take 15 half lives (27.9 hours) for "Kr and 10 half lives (45 hours) for "Kr to decay to below the specified allowed concentration in an unrestricted area. After 48 hours, "Kr concentrations will still exceed the 10 CFR 20, Appendix B limit to an unrestricted area by a factor of 50. To resume normal operations, it will be necessary to slug discharge the remaining gas through the reactor room dampers and dilute it with the air passing through the reactor gas stack. AFRR1 TR83-1, " Safety Analysis of Modifications to Upgrade the Reactor Ventilation System at Armed Forces Radiobiology Research Institute" specifies that 3430 cfm of air is exhausted from the reactor room into the reactor gas stack which has a total flow rate of 35,000 cfm. Simply opening the dampers dilutes the gas 10 a factor of 10. Opening the dampers for one minute in ten minute intervals lets out approximately one tenth of the gas and further diltas the remaining gas in the room. Since the air system is designed to change the air in the reactor room 4.4 times per hour, the dampers will need to be opened for 14 one minute periods to completely clear the room of        t krypton gas.

A straight forward approach to ensure the safe irradiation of krypton gas would be to perform the irradiation with the reactor room air dampers closed to isolate the air in the reactor room. In the event of a gas vessel failure, the room will remained isolated for 48 hours to allow most of the radioactive gas to decay to acceptable levels for release to an unrestricted area. Since the design basis accident radiation release referred to in the facility safety analysis report is approximately 7 Ci for a mptured fuel element (Safety Analysis Report for AFRRI TRIGA Mark-F Reactor, page 6-19), the reactor confinement should easily contain the 0.82 Ci that could be released in a gas vessel rupture. After a 48 hour decay period, the dampers will be opened for one minute at ten minute intervals for 2.5 hours to sufficiently dilute the remaining radioactive gas as it is released to an unrestricted area. These precautions are extremely conservative as 10 CFR 20.103 specifies that the concentration levels are based on a 40 hour per week exposure for 13 weeks. Because the gas will be effectively completely decayed in 14 days, there is no chance of exceeding the quarterly limit. The hazard is further mitigated due to the fact that radioactive krypton is reble gas and only presents a submersion dose hazard unhke fission products which are far more dangerous. l

4. Experiment Procedure: The following steps shall be taken during the irradiation of krypton gas in the gas irradiation vessel described earlier:
1) Manually close reactor- room dampers, ensure good seal of dampers with a vane anemometer. (DELETED 24 JAN 92)
2) Load krypton gas into irradiation vessel.
3) Irradiate krypton gas at 1 Mw for I hour.
4) Monitor gas pressure in vessel to ensure gas has not leaked.

A) If gas has pressure has decreased, stop irradiation immediately.. Notify . ROS/RFD. Continue to monitor gas pressure. Assume worst case scenario and I (SEE ATrACHED GWIGE) isolate the reactor room for 48 hours. Following the 48 hour waiting period, open the dampers for one minute at ten minute intervals for 2.5 hours. Leave the dampers open after the 2.5 hours of opening and closing the dampers.'.' At least one reactor operator will be present in the reactor facility until the gas has been released from the reactor room. B) If gas has been irradiated without incident, the gas will be cryogenically transferred into transfer vessel. Once the gas is outside of the reactor facility boundaries, the dampers shall be reopened. (OfANGED 24 JAN 92) Submitted by: j af Matt Forsbacka Capt, USAF-' Reactor Operations Supervisor i Approved by Reactor Facility Director: Approved- by Chairnaan, Safety and Health Cew a;: M Moore Reactor Facility Director pDotig; , 

                                                                                           'y
                                                                                     , Safety and Health Dtpartment -

1 _ - _ - _ _ - _ _ _ _ . Y

l I l Approved by Reactor and Rs fiation Facility Safetv Co ntnittee: Oh ( # N.W. Mandertield - Colonel. USAF, MSC Chairman, RRFSC C7 w 4 ugfshby Chairrhan, Safety and Health Department r dhbw Mark Moore [' Reactor Facility Director kb uh Mark Voth C Jofieph DeCicco, MSC, USN mm aiv d h b ,, Ron Luersen CAPT C.B' Galley,/4SC, USN 1 l l .: _ - _-_ . _ - - _ . _ _ _ . ...~..-~....._._.-._--._;_.- . - - --. - . . - ...-

Routine Reactor Authorization #5 September 1991 Initeduction: The Reactor and Radiation Facility Safety Committee has reviewed and approved the operations desenbed below. These experiments have been performed ot. numerous occasions in the past and are considered routine. Authorizations: As permitted by NRC license R 84, the Reactor Facility Director may irradiate up to 1000 gm of active or passive electronic components in the reactor irradiation facilities subject to the limitadons imposed by Technical Specifications and applicable procedures, in particular, these experiments shall; o not be able to cause the release of radioactive gases and acrosols such that the annual isotope concentration limits of Table II, Appendix B, of 10 CFR 20 are exceeded, o not create inventories of I-131 through I-135 greater than 1.3 curies and Sr-90 inventories greater than 5 millicuries, o limit known explosive materials to less than 25 milligrams and its explosive potential shall be determined to be within the design limits of its container, o be doubly encapsulated if the release of the contained matenal can cause corrosion to the radiation facility, o have an absolute worth less than $3.00, and o either have movement precluded or be monitored by a Senior Reactor Operator. ALARA principles will be followed at all times during the design and conduct of experi nts. Approved: Reactor Facility Direcief 

                                                                                      ~

6dehairman S .D i NW. Manderfield col., USAF, MSC Chairman, Reactor and Radiation Facility Safety Committee -

Reactor and Radiation Facility Safety Committee [* -- GDLI-M,I oore _ . _ . _ . _ 17 Le

9. Ashby/,

(24W R. Luerso,n I cu hi, Voth CDR J.E. DeCicco, MSC, USN 

      $1 alley, DAPT C.B.

A$MS$,u USN

ATTACHMENT E 10 CFR 50.59 Safety Evaluations of Modifications, Changes, and Enhancements to Procedures or Facilities (Other than Fuel-Follower Control Rods) 1 

                                                                )

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Facility Modification Worksheet 1 10 CFR 50.59 Analysis Proposed Change Modify Transient Rod Drive Support Structure i capt Forsbacka Date U CI 9I Submitted by:

1. Description of change:

Raise transient rod drive to position approximately 60 inches above its current position. This will require a connecting rod which is approximately 60 inches longer than the current connecting rod.

2. Reason for change:

The purpose of this modification is to facilitate the maintenance of the transient red drive mechanism and to unclutter the space between the pool water and bottom of the carriage. This modification will allow for the complete change out of the transient rod drive mechanism without disrupting the other control rods.

3. Verify that the proposed change does not involve a change to the Technical-Specifications or produce an unresolved safety issue as specified in 10 CFR '

50.59(a)(2). - Attach an analysis to show this. This change does not involve a Technical Specifications change. Analysis attached? Yes - X

4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).

Procedure Facility X Experiment 

      ; Revised: 15 May 91                                                                                      Page 1       j A TTAcHMEwi~ 5-l   - - _ - _ _ _ _ _ _ -

Facility Modification Worksheet 1

5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this description need not contain the final SAR wording.

Section 4.10 Reactor Control Components will need to be modified to describe the new physical position of the transient control rod drive mechanism.

6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.

No modification to the actual rod drive mechanism will be made. This modification. is simply to change the physical location of the control rod drive mechnaism. The transient rod drive mechansim will operate in accordance with its design intent. Revised: 1$ May 91 Page 4 l

Facility Modification Worksheet 1 l l~ i

7. Specify associated information.

New drawings are: Attached ,jt.JPhoto is attached, drawing to be produced - 

                                                               ** * ***** d***)

Nm giM Does a drawing need to be sent to Logistics? Yes X No Are training materials effected? Yes No x Will any Logs have to be changed? Yes No x Are other procedures effected? Yes No -v List of items affected: Q%& M yab. I

8. Create an Action Sheet containing a list of associated work speci6ed in item #-

7, attach a copy, and submit another to the RFD. Action Sheet: Submitted X Not Required-l Reviewed and approved by' RFD Date A Ni[ RRFSC Concurrence - , - Date l Revised: 15 May 91 Page 5 L

k L Safety Analysis of Transient Rod Drive Support Modification The transient rod drive will be raised approximately 60 inches above its current location. The purpose of this modification is to facilitate maintenance of the transient rod drive mecimanism. No modification of the function of the transient control rod drive will be made. All normal testing will be performed to insure that the systems will operate in accordance with its design intent. The photographs below illustrate the modification made: k 6', s 4 e

Action Sheet  !, subiect omce symoon DCBD sv. pen.. ] Modification of Transient Control Rod Drive Support 23 OCT 91 ACtlen A44Wi red Concurrence and Approval uemer.noom for meewa,icescru oneay me reouvemems. escurouno ano accon tuen or recommenoso vust ce su ,c>ency r cernoea to ,conety me ecnon *imout recourse to emw sources i

1. The transient control rod drive mechanism needs to be raised by about 60 inches above its current location to allow for ease of maintenance.

LOGD needs to design and appropriate support structure to keep the rod drive mechanism securely in place above its current location.

2. Once built, we'll need to make a drawing of the finalized design for our as-built drswings. Autocad would probably be best.
3. No training materials should be effected by this modification, but we should brief the staff on why the change was made.

on LA OCttA4. /h O CACud ' h . Jh k4CkP- . 

             %           au % D/ as cweJJ4                                    .
                                                       <C ..eee .     ,

i 1 c ne n. e_: Office peesne Phone in nenD w. 9 wmn . s-179n 7/ o(ruf 6 h r- 4 Q fe r. s i in o V Disoetenes (Dtg Show Aadiuonal Coorenstkm on Reverse $de er Conenuetten Sheet Actmn Otticer (Name, grace, t,;amne ord signature) Matt Forsbacka. Capt USAF. 5-1221 eD3" Hegroooo when Seoersted from caserned Deamnt

ATTACHMENT E-2

Attachment I hEFERENCE: ADMINI3TRATIVE FROCEDURE I. FACILITY MODIFICATIONS ANALYJ13 OF FROPOSED MODIFICATION Modification Nomenclature: REPLACEMENT OF WATER GAMMA ACTIVITY MONITORING SYSTEM Analysis: SFC LAUGHERY Date: 12 February 1991 SECTION A

1. Document analysis to determine if a change to the Technical Specification is required. Include 10 CFR and/or Technical Specification references as applicable.

NO CHANGE TO THE TECHNICAL SPECIFICATION IS REQUIRED

2. If your analysis determines that a Technical Specification change is required, go to SECTION B.
3. If a Technical Specification change is not required, document an analysis to determine if the prvposed modification would constitute a change to the facility as described in the SAR. Include 10 UFR ind r Ab references as applicable.

NO CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR IS REQUIRED. The proposed modification is the replacement of the vacuum tube operated monitoring system with a solid state system. This modification is necessary because replacement vacuum tubes are no longer available.

4. Document an analysis to determine if the proposed modification would constitute a change in a procedure as described in the SAR. Include license and/or SAR references as applicable.

PROPOSED MODIFICATION DOES NOT CONSTITUTE A C11ANGE IN A PROCEDURE AS DESCRIBED IN THE SAR

5. Document an analysis to determine if the proposed modification would constitute a change in the tests or experiments as described in the SAR. Include tieens-and/or SAR references as applicable.

PROPOSED MODIFICATION DOES NOT CONSTITUTE A CHANGE IN THE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR

6. If the proposed modification does not constitute a change to the facility, procedures, tests or experiments as B TTRC HMEn T E- L
  . ECT10t1 C
1. tio 10 CFR 50.59 is required. The analysis in SECTIO!15 A.3. A.4, A.5 provide the bases for this determination that:
a. no change in the Technical Specifleation is required.
b. no change in the facility as described in the current SAR is proposed.
c. no change in the procedures as described in the current SAR is proposed, and
d. the proposed test or experiment:

(1) coincides with those described in the current SAR; c (2) is permitted by the Technical Specifications, and (3) has been previously reviewed and conducted.

2. This modification has been reviewed in accordance with ALARA principles. Comments (if necessary) are as f ollo -

1 b j Mhh

4. Tw . - _ approved by RFD i _ _ D.,: t.fg /fi/
5. RRFSC Concurrence /

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Attachment i 1*Ei EhEth;E : ADMINISThATI' E i-hOaDUhE I. FAtlLITY MvDIFICnTION6 ANALYbid OF EhulO5ED MODIF1'AT1uN Modirteattun ik.me nc l a tu re REPLACEMENT OF WATER CONDUCTIVITY MONITORING SYSTEM Ana ly s u : SFC LAUGHERY Date: 12 February 1991 

    .iECTION A
1. Document analysis to determine if a change to the Technicil
             .3pectfication is required.        Include 10           CFR and/or Technical Specification references as applicable.

NO CHANGE TO T!!E HNICAL SPECIFICATION IS REQUIRED

2. II your analysis c . ermines that a Technical Specification change is required, go SECTION B.
3. If a Technical Specification change is not required, document. an analysis to determine it the prvposed modification would constitute a change to the tacility as described in '!r .4 h .

Include 10 CFR and/or SAh reIerences as applicable. NO CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR IS REQUIRED. The proposed modification is the replacement of the present monitoring system with a new system by. the same manufacture This modification is necessary because replacement parts for the old system are no longer available.

4. Document an analysis to determine if the proposed modificatien would constitute a change in a procedure as described in the SAR.

Include license and/or SAR references as applicable. PROPOSKD HODIFICATION DOES NOT CONSTITUTE A CHANGE IN A PROCELURE AS DESCRIBED IN THE SAR

5. Document an analysis to determine if the proposed modification would constitute a change in the tests or experiments as described applicable.

in the SAR. Include license and/or SAR references as PROPOSED MODIFICATION DOES NOT CONSTITUTE A CHANGE IN THE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR

6. If the proposed modification does not constitute a change to the facility, procedures, tests or experiments as described in th-SAR TAnswer to SECTIONS A.3, A 4 and A,6 is 'Nv" in all casesi.

Go to SECTION C. '7T/MHME]YT Ed 1 _ _ _ _ _ _ - - - - - .J

dECT M!i C 1 !J . 10 CFh 50.03 $s required. The analysis in IE' f l0!1.3 A.3. A.4, A.5 provide the bases for thi.e de t e rmi r.s t i . n thst:

a. no change in the Technical Specifteation is required.
b. no change in the facility as describ-;d in the current SAR is proposed.
c. no change in the procedures as described in the I current SitR is proposed, and I
d. the proposed test or experiment (1) coincides with those described in the current 4
                                                                  '6 AR ;

(2) is permitted by the Technical Specifications, and (3) has been previously reviewed and conducted.

2. This modification has been reviewed in accordance with ALARA principles Comments (if necessary) are as f ollows:
4. Reviewed and appr >ved by RFD Ilb > -

Date1[lfh/

5. RRFSC Concurrence . d# ,([u __ _ Date.2fffl 9

E ATTAOiMENT E-4

_ _ _ _ _ _ _ _--_- - - - - ~ Facility Modification Worksheet 2 No 10 CFR 50.59 Analysis Required Proposed Change " "*" " # 1 ' " '

  • 8 Y " ' * * * * " ' d
  • Modification to: Pr ocedure . Facility " Experiment Submitted by: ceor2e Date s % m -
1. Description of change:

that will allow operators to override the cam dampor closure system to - - the dampers to open when the came are alaming. System will only be operated upon the orders of the RFD. will be detected before any daily operations. Operation of the ems damper closure s THIS FFCRS. SYSTEM IS TEMPORARY AND WILL BE RDt0VED AFTER SUCES - USE oF THIS SYSTEM VILL BE ALLOWED oNLY BY DIRECT AUTHR 2 Verify that the proposed change does not involve a change to the Technical 

                                                                                                                                                                         =

Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved safety question as defined in 10 CFR 50.59(a)(2), intended.

3. If change involves a facility modification,t attach a drawing if a structural Logistics, facility drawings need updating, forward a copy of changes necessary to n/A
4. Determine what other procedures, logs, or training material may be affected and record below.

Staff bziefi

5. List of associated drawmgs,ng procedures, logs, will or beotreautred. her materials to be changed:

Sea attached drawing of system for reactor f.iles.

6. Create an Action Sheet containing the list of associated work specified above, attach a copy, and submit it to the RFD.

Action Sheet: Submitted x Not Required _ Reviewed and approved by RFD RRFSC Notified 

                                                                                                                             /       Date[b I/

[ Date 17 DEC f991 

                   ^

Revised: 15 May 91 Page 6 HTi?%HM6tvr E-9 _ _ _ - - _ - - - - - - - _ _ . _ _ _

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l TRAINING POR DAMPER OVERRIDE BOX GENERAL The damper closure system located in the penthouse requires a closed loop to cause the dampers to close. When the loop is open the dampers can be opened 'nd will remain open. A two conductor wire is run from the penthouse to a relay Ln the motor control center. This relay is operated by AC voltage from the primary CAM. If the primary CAM alarms, AC power will be turned off to the relay in the motor control centnr. If the backup CAM alarms it will break the AC power with a relay in the black box in the reactor room under R2. Under normal operation power from the primary cam energizes the relay in the motor control center. This opens the loop to the penthouse damper closure system which will allow the dampers to be open. DAMPER OVERRIDE: The Cverride system consists of a two conductor wire with a plug together. on the end of each wire so that they can be plugged The wire breaks into the two conductor wire which comes from the penthouse and simply acts as an extension of that wire. One end of the new wire connects into the penthouse wire at the relay in the motor control center. The l other end of the wire, the end with the connector, is located in 3152 on the wall in a labeled box above the air compressor. Under normal operations the connector will always be connected. will operate This properly.will act like a normal wire and the system In the event of an emergency and if the RFD orders it, the connector can be disconnected. Opening this connector will open the circuit to the penthouse thus allowing the dampers to be opened. The orders connector from the should RFD. never be disconnected without direct 

                                                                   ~

contro Numoer Action Sheet ' swo,0., omee ,mt.o. .w.,en.e , n em a  ! 

                                                                                                                        >: : s Temporary Installation of CAM Damper Override                                    sete                              q System                                                                                      e DEC 91               j AC16on Sequired l

Training of Staff in use of override system, removal of system when no longer raquiri womorenowm eor necore. icesence oneay ene roouvemeets eackroun o ano action taaen or 'ecommencea vust te sec eavy cetnoea +o cents v ecton .cout recovise to omer sources i

1. A CAM damper override system is proposed for the purpose of allowing the air dampers to be opened if necessary during the pulse testing of the FFCRs. THIS SYSTEM IS TEMPORARY, AND USE OF THIS SYSTEM MAY ONLY BE AUTHORIZED BY THE RFD.
2. Once the FFCRs are fully tested, this override system will be removed.

l l I, tConenue on can bono) w__ Coormaoiseas a,,rweie OWs  % Pheae inattale Date RSDR Mark Moore. RFD 5-1290 Dispatched (Dtt) 

 ^

Show Adeltional CoordinstNm on AWeree Ede er Continuetton Sheet Action Omcor (Narne. graoe. phone and segrature) Matt Forsbacka Capt USAF. 5-1290 DNA Form 547 4 Jan as nega$ed when Seperstvo mun Clase#Aed Docurnent J JW d

4 

,                    ATTACHMENT E-5 pu iiiiin

4e , 4 , 

                         . r ni Mi F        ,1 N I N I . o b /' ' 'F    t     ".'""1        -  I'.4it   W ' t ' I t' l ' % 1   '

N. 

                         ..h              y i fe }     .:4 1, Mt t iicic;A T 10 N Mc 1 i t i c *s t t on N : mere; i si a re Au x Lil a ry_ Ti toe t a n_. H o tL C oli ti ol e An a !.v 3 i s Ofil 14nu11hu rY                Date 2Q 30 m 'et 1990 J,ECTION A

, 1 Document analycis t,o determine if a change to the Te c hn a -:31 5pecification is required. Incjude 10 CFR and/or Technical 

                                .3pucification references as spplicable.

NO CilANGE TO Tile TECllNICAh SPECIFICATION IS REQUIhED. 

                         .;      if your analysis determiner., that a Tevhnim i Stua1 tie ,tlun change is reanir+d, go to CECTION B.

J. If a Technical 6pecifluation ;hange 1 nut requate,s. 1.4 om-nt an snaLys)e to determine ti the proposed modifientiva would constitute a change to the facility as described in the f. A R . Include 10 CFR and/or SAR references as applicable. TIIERE IS NO CilANGE TO THE FACIh1TY AS DESCRIBED IN Tile GAR.

4. Document an analysis to determine if the proposed modification would constitute a change in a procedure as described in the SAR. Include license and/or SAR references as applicable.

T!fERE IS NO C11ANGE IN ANY PROCEDURES AS DESCRIBED IN Tile SAH.

5. Document an analysis to determine if the proposed modification would constitute a change in the tests or experiments as described in the SAR. Include lict;nse and/or SAR references as applicable.

THERE ARE WO CHANGES IN TIIE TESTS OR EXPERIMENTS AS DESCRIBED IN THE SAR.

6. If the proposed modification does not constitute a change to the facility, procedures, tests or experimants as described in the SAR (Answer to SECTIONS A.3, A.4 and A.S is 'N0" in all cases). Go to SECTION C.

e 77t]C HMsgr E~ ~l 

             .Ec110N O 1              N.                              p; i: Fis 00.59 te re;uired. The an4lytis in 3ECT MN A A . 3.

is 4. A . '- prcvide the b a s e t.; f or this detertnirnt ien t hM

s. no change in the Technical Opeuitioiti n 1*

required,

b. no change in the facility is dv e c. r i be d in ' h --

current SAh in proposed. e no change in the procedures as des 'ribed in the current SAR is proposed. and

d. the proposed test or capt ringmt (1) evincides with those described in the current 6AR;

( !: is permitted by the Technical ap"eification*, t3) has been previously reviewed and conducted. J. This modification has been reviewed in accordance with ALAhA principles. Comments (if necessary) are as follows: The ne.w reactor console has a COUt1T DOW11 timer. The auxiliary timer is a COUt1T UP timer which facilitates record keeping during experiments by displaying the actual time the reacter was at power. 3. W Reviewed and approved by, R03-_ . _____,___ Date L M t._fi

4. Reviewed and approved by RFD,.b b g _ Date L U d 7/
5. RRFSC Concurrence _ #ff

_ _ 4 __ Date @ [f 7/ t

ATTACfWENT E-6 0

1 Facility Modification Worksheet i 10 CFR 50.59 Analysis Proposed Change t'p g r a d e pulse dispiny capabilitie. and install variable pulse timer as specified in origin.?1 contract with General Atn~fc. __ Submitted by: capt Forsbacka Date 13 SUV 91

1. Description of change: _

Modification of both NPP-1000s for small pulse operation. This modification allows a gain change for high sansitivity pulse data aquis4. tion. Addition of an Eagle dig timer and relay logic to scram the reactor after initiating a pulse.

2. Reason for change:

To fulfill original GA contract obligations.

3. Verify that the proposed change does not involve a change to the Technical Specifications or produce an unresolved safety issue as specified in 10 CFR 50.59(a)(2). Attach an analysis to show this.

Analysis attached? Yes Not Required

4. The proposed modification constitutes a changes in the facility or an opera-tional procedure as described in the SAR. Describe which (check all that apply).

Procedure Facility Experiment Revised: 15 May 91 Page 3 B TT/9 CHM EN T 6-6

Facility Modification Worksheet 1

5. Specify what sections of the SAR are applicable. In general terms describe the necessary updates to the SAR. Note that this det.cription need not contain the final SAR wording.

Section 4.14. Scram Logic Circuitry: Add description of pulse mode scram t ic:e r .

6. For facility modifications, specify what testing is to be performed to assure that the systems involved operate in accordance with their design intent.

Botti modifications are for the purpose of ensuring that the console operates in accordance with its design intent. f I Revised: 15 May 91 Page 4

Facihty Modification Worksheet 1

7. Specify associated information.

New drawings are: Attached x - Not required Does a drawing need to be sent to Logistics? Yes . No

  • Are training materials effected? Yes X No Will any Logs have to be changed? Yes No x Are other procedures effected? Yes v No List of items affected:

Training materials will have to reflect the ability to view posses in two modes. Procedure 8. TAB G1 and C2. will have to be modified to instruct the operator whisn mode of pulse display should be uced.

8. Create an Action Sheet containing a list of associated work specified in item #

7, attach a copy, and submit another to the RFD. Action Sheet: Submitted X Not Required Reviewed and approved by RFD Date/3 /V [ h RRFSC Concurrence Date 17 DEC 1991 s 

 ~

Revised: 15 May 91 Page 5 

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                       ,,                        UNITED STATES
        !-      'O ) i                NUCLEAR REGULATORY COMMISSION
        ',, Sg F 4                             W A$HINGTON. D C. h?A
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  • cctober 8, 1991 Docket No. 50-170 Colonel George W. Irving,111. BSC, USAF Director Armed Forces Radiobiology Research Institute Bethesda, Maryland ?0814-5415

Dear Colonel Irving:

SUBJECT:

ISSUANCE OF AMENOMENT NO. 21 TO FACILITY OPERATING LICENSE NO. R ARMED FORCES RAD 10 BIOLOGY RESEARCH INSTITUTE (AFRRI) The Commission has issued the enclosed Amendment No. 21 to Facility Operating License No. R-84 for the AFRRI TRIGA Research Reactor. The arnendment consists of changes to the Technical Specifications in response to your submittal of April 30, 1990, as supplemented on December 17,1990, March 5,1991, May 17, 1991, August 16, 1991, and September 10, 1991. The amendment (1) corrects errors in typography and grasunar, (2) increases -- - the maximum licensed steady state reactor power to 1100 kilowatts. (3) author-izes installation of fuel follower control rods, (4) clarifies the transfer of Reactor Facility Director (RFD) responsibilities in the absence of the RFD, and (5) allows operational flexibility in performing surveillance testing of the ventilation systese for the reactor facility. Enclosure 2 is a copy of the related Safety Evaluation supporting Amendment No. 21. Sincerely, c r Alexander Adams, Jr., Project Ianager Non-Power Reactors, Decommissioning and Environmental Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

1. Amendment No. 21
2. Safety Evaluation cc w/ enclosures:

See next page C.- 

                                                                                        **""""**~

Action Sheet s.e, t om.e s,m- ....en.e D R B P' N/A ' Installation ot upgraded pulse display and variable p.to 13 NOV 91 i pulse timer. A tion Required Modify Procedure 8. TAB G1 und TAE CL womeranowm tw socore tooecnte trear tee reownements. escaroung saa octon teen or recommenoso vwat t4 *w"wey mareo to eenm tne acton .itnout recouru to othw sourcee 6

1. The new pulse display will allow us to select high or low resolution, depending on the size of the pulse. The low resolution will be used for large pulses and tha high resolution will be used for small pulses. We need to inodify procedure
8. TAB G1 and TAB G2 to reflect this option.

(Contrive on > tonc) l I Coertiencuene Appreenis Otsce ifeme Phone Inty Date AC w._ RSDR Mark Moore. RFD 5-1290 ' 

                                                                                                             /?PD       fjM*f(

WQ hk u su :L Diepesched (Ots) Show Additional Coeralinet6en on A* verse Side er Corttmusten Sheet Acuen ovner m grooe, onone ano egne,ures Matt Forsbacka, Capt USAF. 5-1290 P*NAfonn 53 547 Megrened when separet d from Closetw Document

FCOM 12/06<B4 15: 46 P. 1 l FAX TRANSMISSION GENERAL ATOMICS TRIGA REACTORS FROM: a111 nood PAGE 1 OF 2 TO: narry spence, ArnaI

SUBJECT:

Work Performed week of 11-18-91 DATE: 12-06-91 QLD._1URIMEEEI a) Installed Pulsa Mode Scram Timer. Added a Eagle digital timer and relay logic to scram the reactor a preset time after initiating a pulse. The timer is mounted behind the chart recorders in the CSC, and is intertacnd to the CSC computer and scram loop through a pair of relays. The timer is only active in PULBE mode, and is started by firing the pulse rod. There is no provision to disable the timer in pulse mode. The console software provides for a maximum scram delay of 15 seconds in pulse. If the operator wishes to scram the reactor prior to the 15 cocond cottware scram timeout, the desired time must be entered to the eagle timer prior to initiation of the pulse. Note that due to the software configuration the Csc will not post the Pulse Mode Scram Timer scram message until all pulse data acquisition is completed. Rob and Mike were given a copy of the Pulse Mode Scram Timer and the updated CSC scram loop schematics, b) Replaced the serial data cable between the CSC computer and the Tektronix high resolution monitor. The new cable is a two piece cable, which allows the disconnection of the high resolution monitor at the rear of the computer chassia. The AFRRI vire list was updated to reflect the wiring change (see .0 Rob or Mike for a copy of the list). T 

                                                                                .0 EHEREM't. ORDBRB t a) Fuel Follower Control Rods. You        should have all of the information on this subject already.

       *0M                                           12 esesa i5 as P. I b) small Pulse Data Acquisition.
                 }!ARDWARE: Modified both NPP-1000's for the small pulse option.

This modification allows a gain change for high sensitivity pulse data acquisition. Also modified DOH-32 relay output #18 (in the DAC) to enable the gain change in pulse mode. Rob and Mike hava ocpien of the changes. SOFTWAREl Installed now software to allow the acquisition of both normal and small pulsos. The high sensitivity pulso data acquisition will acquire up to 1000 MW pulses (full scale) . 5 r t __ TOTAL P. 2

ATTACHMENT F n May 1991 Summary of Changes to Administrative and Operational Procedures i t s

Summary of Changes to Administrative and Operational Procedures introduction in May,1991 RSDR stalf completely updated the Administrative and Operational Procedures to implement grammatical changes, a new format for the procedures, and corrections throughout including updating the core position references. Administrative Procedures { A1. Add reporting requirements to (2). A2 Delete reference to ' escorted access roster *. A3. Complete revision / simplification of 10 CFR 50.59 worksheets. A4. Completely new procedure developed in conjunction with NRC recommendations. Operations] Procedures E O. Change ' initial block' to " signature block".

1. Add " acting ROS" to 2.f.

lA. Delete escorted access roster (2.b.). Change "open* to " enter" (4.a.). Delete material that duplicated HPP 3-1 (7.c.). Update badge types for entry (2.b.).

18. Change " element F28' to ' desired element" (l.c. and 4.g.).

Delete reference to " upper end indicator

  • painted on cable (3.b.).

IC. Delete requirement place lead bricks on tube supports (1.a.(1)). Delete references m former ' limit switch" (l.a.(5)). Delete reference to having prep area ' scaled off" during operations (3.f.). 1D. No changes. IB. Change "SRO' to "!W operator' (1.). Add "in the reactor room' to (3).

2. Delete statement 'a record of operatious will be kept for each trainee / operator" (2.c.),

d

3. Add *and conwle system manuals
  • to (1).
4. add 'during operations' to (1.b.).
5. Add (3).
6. No changes.
7. Revise references to Shutdown Margin throughout.
8. Add (5.f.).

8A. Delete reference to " Mode 1, I A, etc' (5.a.(5)). 8D. Delete measurement of pool level in 3161 (V.3.). Change ' conductivity" to ' resistivity" (VI.5.). Change ' pool" to " inlet" (VI.17.g.). f,H l . Sarne changes as in 88. 8C. No changes. 8 D. Change 15 Watts to 5 Watts. Delete reference to monthly summary sheet (4). 8E. Update title. Change ' servo nxxle' to " auto irode' throughout. Change 800 Kw to 200 kW in (2). Add (7). 8Fl. Change ' servo' to " auto

  • throughout.

Add general staternent. Add specific (1). Add second sentence in (13). Add (16). 8F2. Change ' servo" to " auto

  • throughout.

Add general statement. Add specific (1) and (12). Add second sentence in (9). 301. Add mnimum pulse sitcs to (5). Add different types of operations to (9). 802. Add pulse sires to (5), g Add different types of operations to (12).

811. Add (IV.b.). Remove referes i 1.0 to automatic scram reset, change "conouctivity to 

                                  ' resistivity".
81. Update (VI.2.) and (VI.4.).
9. Add 'while the power rails are energized
  • to (l.f.).

Add 'or mercury compounds in any form

  • to (2).
10. Moved from Procedure 8. Tab K to new Procedure 10. Add additional instructions to step 3 to better illustrate how to check the high voltage setting of the SGM l
11. Formerly was attachment to Procedure Vill Tabs B and Bl.

_ _ - _ _ - _ _ - _ _ - _}}

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