ML20137H516

1996 Annual Rept of Afrri Triga Reactor

ML20137H516
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	12/31/1996
From:	Miller S ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9704020268
Download: ML20137H516 (159)

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ARMED FORCED RADIOZIOLO3Y RECEARCH INOTITUTE 8901 WISCONSIN AVENUE BETHESDA. MARYLAND 20899-5603

@5 5002 20 March 97 RSDR

SUBJECT:

Submission of Annual Report U.S. Nuclear Regulatory Commission

] Document Control Desk Washington, DC 20555 l

Dear Sir:

i Attached please find the 1996 Annual Report for the AFRRI TRIGA 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.

Attachment:

1. v teihen Miller as stated Rektor Facility Director 7

Cy Furn:

U.S. Nuclear Regulatory Commission ATTN: Mr. Marvin Mendonca, Mail Stop llB20 Washington, DC 20555 Regional Administrator U.S. Nuclear Regulatory Commission, Region I ATTN: Mr. Thomas Dragoun 475 Allendale Road King of Prussia, Pa.19406 ADW c:0055 I

3 9704020268 961231 PDR ADOCK 05000170 l$lll%Ellllkll,lll 1 R PDR -

1996

! Annual Report of the AFRRI TRIGA Reactor chu '~

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l Submission of 1996 Annual Report 1

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Submitted by

$ir. S[ephen Miller 24 /99

'Datl l Reactdf Facility Director I Approved i E.KEARSLEY W ~

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i Armed Forces Radiobiology Research Institute AFRRI Triga Reactor Facility )

1 January, 1996 - 31 December,1996 To satisfy the requirements of:

U.S. Nuclear Regulatory Commission, License No. R-84 (Docket No. 50-170),

Technical Specification 6.6.1.b.

The Reactor Facility Director would like to acknowledge the participation of the following ,

AFRRI personnel for their contributions to this annual report.

Edited by:

Mr. Robert George, Reactor Operations Supervisor With Contributions From:

LTC Ieonard Alt, Senior Staff Engineer Mr. Guy Bateman, Scientific Medical Illustrator SFC Danny McClung, Senior Reactor Operator Capt. Oscar I.essard, Senior Reactor Operator ENS John Ventura, Senior Reactor Operate,r Mr. John Nguyen, Senior Reactor Operator SFC Brian Cohill, Senior Reactor Operator Mrs. Emma Kephart, Health Physcists Submitted By:

Mr. Stephen Miller Reactor Facility Director Armed Forces Radiobiology Research Institute 8901 Wisconsin Ave.

Bethesda Md. 20889-5603 Telephone: (301) 295-1290 Fax: (301) 295-0735

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1996 ANNUAL REPORT OF THE l AFRRI TRIGA REACTOR i

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Docket 50-170 License R-84 Submitted By Stephen Miller Reactor Facility Director

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1996 ANNUAL REPORT  !

TABLE OF CONTENTS Introduction General Information i

Section I Changes to the facility design, performaace characteristics and operational procedures, results of surveillance tests and inspections i t

Section II '

Energy generated by current reactor core and number of pulses $2.00 or larger Section III l

. Unscheduled shutdowns ,

Section IV I Safety-related corrective maintenance  :

Section V Facility changes and changes to procedures as described in the Safety Analysis Report.

New experiments or tests during the year .

Section VI l Summary of radioactive effluent released Section VII Environmental radiological surveysSection VIII Exposures greater than 10% of 10 CFR 20 limits Attachment A Revised Reactor Administrative and Operational Procedures Attachment B ,

10 CFR 50.59 safety evaluations of modifications, changes, and enhancements to procedures or facilities Attachment C l

Appointment letters for various Reactor and Radiation Facility Safety Committee changes Attachment D Correction to SHD Data for CY95 Annual Report

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INTRODUCTION 1

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1996 ANNUAL REPORT INTRODUCTION In 1995 the AFRRI reactor facility was available for irradiation services throughout the year with only a brief non-operational period for the annual reactor maintenance shutdown. Members of the reactor staff were involved in a DNA Dispersion project at White Sands, New Mexico. This i

project utilized the reactor for multiple irradiations throughout the year. I i

The Reactor Facility Director, Mark Moore, stepped down as the RFD, pending his retirement

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in 1997, and was replaced by Mr. Stephen Miller. Mr. Miller has acted in the capacity of Deputy Reactor Facility Director for four years prior to this transition. Mr. Moore retains his NRC SRO license and provides assistance as requested by Mr. Miller to allow for a smoother I

transition.

1 i The microprocessor-based instrumentation and control console, fully installed in 1990 after a two year test period, completed its sixth year of successful operation as the primary control and I

safety instrumentation. The console operated throughout the year with only minimal malfunctions (See Section IV).

In 1995 AFRRI received a closure order from the Deputy Secretary of Defence. As of December 31,1996 no decision to reverse or to continue with the closure of AFRRI has been made. AFRRI has been funded to continue operating for the next several years. No funding has been allocated to facilitate the closure of AFRRI.

The Stack Gas Monitor electronics package was replaced in the spring of the year, and after the high voltage power supply was replaced (see Section 1 for description of 50.59 review) the new i electronics package performed as expected for the remainder of 1996. No further problems j occurred with the SGM throughout the year.

All existing procedures which had been written with Ventura Publisher software were converted l for use with Word Perfect software. The procedures were formatted to look virtually identical to their Ventura Publisher predecessors. The procedures were then saved to a read only drive on the reactor network with only the reactor management able to over-write previous versions of the procedures. Also added to each procedure was the storage location of the procedure to allow easier recovery of the electronic file.

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The annual reactor facility audit was held in November 1996. The auditors concluded that the

, reactor was being operated in a safe manner in conformance with the Technical Specifications and Operating License. No items of concern were identified during this inspection. Discussion of the inspection is covered in Section 1. E. The Nuclear Regulatory Commission did not inspect the AFRRI TRIGA Reactor during 1996, nor was this required.

The reactor core remained unchanged throughout the year.

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l Members of the reactor staff designed and built a Dynamic Reactivity Computer. A static reactivity computer was built several years ago using a PC type computer which worked well and produced results identical to the stopwatch method, but like the stopwatch method, required a critical core before each reactivity pull and was therefore very slow. The new Dynamic ,

Reactivity Computer uses the same PC computer, but uses a dynamic mathematical algorithm i to determine the reactivity. The new program worked much faster than its static predecessor because the dynamic computer looks at the rate of change in reactor power instead of step insertions from critical. The rod values measured using the dynamic computer fell within a couple cents of the static reactivity computers data an.i the stopwatch method. After extensive testing of the new computer, it was accepted and used to calibrate the control rods during the annual shutdown.

A pulse monitoring computer was built to help study reactor pulses as part of a research project by a University of Maryland student. The independent pulse monitoring computer provided data j that matched the reactor console pulse data and made individual data points available that could be manipulated for analysis by the student. The program will be used in the future for other projects which require capturing of pulse data. ,

J Reactor staff changes included the departures of ENS John Ventura and Capt. Oscar lessard.  !

No new members were added to the staff. SFC Cohill and Capt. Oscar Lessard were examined  !

and licensed by the NRC as Senior Reactor Operators. I Changes were made to the procedures and facilities during 1996. These changes were supported by an extensive safety review process in accordance with the provisions of 10 CFR 50.59.

These changes will be discussed fully in Sections I & V.

The reactor staff provided personnel to assist in conducting operational and safety inspections of the Fast Burst Reactor facilities at Aberdeen Proving Ground, Maryland, and White Sands Missile Range, New Mexico. At the request of Cornell University, an operations audit of their reactor facility was also conducted.

The remainder of this report is written in a format designed to include information items required by the AFRRI TRIG A Reactor Technical Specifications. Items not specifically required are presented in the General Information section. The following sections correspond to the required items as listed in Section 6.6.1.b of the AFRRI TRIGA Reactor Technical Specifications.

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GENERAL INFORMATION 1

Key Personnel i

Reactor and Radiation Facility Safety Committee I

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GENERAL INFORMATION All personnel held their positions as listed throughout the entire year unless otherwise specified.

1. Current key AFRRI administration personnel (as of 31 December 96) are as follows:

Director - CAIYT Eric E. Kearsley, USN Chairman, Radiation Sciences Department - CAPT James Malinoski, USN (15 October 1996)

AFRRI Radiation Protection Officer - CAPT C. B. Galley, USN ( 06 December 1996)

2. Reactor Senior Technical Manager:

l Reactor Facility Director - Mr. Stephen Miller (SRO) (23 October 1996)  ;

3. Current key Reactor Operations Personnel:

Reactor Operations Supervisor - Mr. Robert George (SRO)

Training Coordinator - Mr. Robert George (SRO) (13 November 1996)

Maintenance - Mr. John Nguyen (SRO) l Administration - SFC Danny McClung (SRO)

Senior Staff Engineer - LTC Leonard A. Alt (SRO)

4. Other Senior Reactor Operators: SFC Brian Cohill (SRO)

5. Operator Candidates: ILT Christopher Pitcher (31 December 1996)

6. Newly Licensed Operators: Capt Oscar lessard (10 October 1996)

SFC Brain Cohill (10 October 1996)

7. Additions to staff during 1996: ILT Christopher Pitcher (31 December 1996)

(Member of RSDD Staff)

8. Departures during 1996: ENS John C. Ventura (19 December 1996)

Capt. Oscar Lessard (20 December 1996) l

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j 9. There were several staff changes to the Reactor and Radiation Fa:ility Safety Committee

during 1996. Col. Curtis Pearson replaced COL David G. Jarrett as the chaliman of the committee.

! CAM Galley replaced Mr. Tom O'brien as the radiation protection officer. CAPT Malinoski

replaced CAPT Galley as the Chairman of the Radiation Sciences Department. Mr. Edward Herbert t

replaced Mr. James Caldwell as the Montgomery County Government representative for the Environmental Protection Agency. Mr. Stephen Miller replaced Mr. Mark Moore on the committee

when he assumed the position of reactor facility director. The appointment letters are shown in L Attachment C.

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The 1996 RRFSC consisted of the following membership in accordance with AFRRI Reactor 3 Technical Specifications (as of 31 December 1996):

Regular Members:

! CAPT Charles B. Galley, USN (AFRRI Radiation Protection Officer)

Mr. Stephen Miller (Reactor Facility Director, AFRRI)

} Mr. Marcus Voth (Director, Brenwale Reactor, Pennsylvania State j University)(Reactor Operations Specialist)

{ Mr. Mark A. Miller (Radiation Safety Officer, Naval Research Laboratory)

Chairman:

} Col. Curtis Pearson, USAF, MSC (Director's Representative)

] Special Members:

CAPT James Malinoski, USN (Chairman, Radiation Sciences Dept., AFRRI)

Non-voting members:

! Mr. Edward Herbert, Montgomery County Government, Environmental Protection Agency l Dr. Leslie McKinney (Radiation Pathophysiology and Toxicology Department, AFRRI)

Recorder:

l SFC Danny K. McClung l

As required by the reactor Technical Specifications, four meetings of the RRFSC were held:

l 18 March 1996 Full committee meeting

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i 11 June 1996 Full committee meeting i

i 16 October 1996 Full committee meeting l l

11 December 1996 Subcommittee meeting l

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SECTION I 4

l Changes to the Facility Design, l Performance Characteristics ,

i. and Operational Procedures. l f Results of Surveillance Tests i

and Inspections.

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l SECTION 1 A summary of changes to the facility design, performance characteristics, operational procedures, and results from surveillance testing are contained in this section. The revised reactor administrative and operational procedures can be found at Attachment A, while the 10 CFR 50.59 reviews can be found in Attachment B for all of the following changes.

A. DESIGN CIIANGES: I i

1. A malfunction of the Data Acquisition and Control system (DAC) hard drive (see )

malfunction, Section 4, 26 Feb 96) required a replacement hard drive be installed into i the reactor DAC. The DAC software was loaded onto an equivalent hard drive and I installed into the DAC computer. The drive was tested extensively with all software applications before the console was approved for normal operations.

2. The existing 2 pole switch located under the FIRE button on the reactor Control System ,

Console (CSC) was replaced with a 4 pole switch. The wires on bank 1 and 2 were l removed from one switch and attached to the new switch in exactly the same locations.

This change did not require changes to the drawings or the operation of the reac.or.

Additional wires.were added from bank 3 and 4 to a terminal strip in the back of the CSC. The additional outputs can be used to trigger other equipment to take action when the pulse FIRE button is pressed. In this case a piece of core monitoring equipment was under development and needed input from the FIRE button.

3. A drain line was installed in the bottom of the reactor cooling tower. The purpose of this drain is to allow the water in the cooling tower to be easily drained and replaced with fresh water as deemed necessary by a reactor staff member performing the monthly check of the cooling tower. During the summer months, water loss due to evaporation in the cooling tower is replenished through a float valve. As the concentration of the minerals in the water increases, !he minerals begin to plate out. By draining the water and replacing it with fresh water. the amount of deposition can be reduced. This addition to the cooling tower was m.ide due to the amount of deposits found in the sump of the cooling tower during the presious annual cleaning.

4. A more powerful power supply was '.:ubstituted for the Stack Gas Monitor (SGM) high voltage power supply. The power supply provided by the manufacturer of the SGM electronics appeared to be operating at its limit which was causing stability (drifting) problems. The replacement power supply can provide 100 times more current if necessary. Procedure C006 was also updated to provide a method for adjusting the high voltage for calibration of the unit.

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B. PERFORMANCE CIIARACTERISTICS:

No changes to the core occurred during 1996. All fuel, chambers and the CET remained l

in place for operations throughout the year. The performance characteristics of the core did not change.

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C. ADMINISTRATIVE PROCEDURES:  !

1. The boundaries for the restricted areas in the Physical Security Plan for the TRIGA Reactor Facility were changed from including the entire AFRRI complex to just the reactor areas. The entire AFRRI complex was designated as a DoD controlled access area. Removal of the formal designation of the entire AFRRI complex as a DoD Restricted Area will change the reference to the Internal Security Act of 1950, but will not change the manner in which the remainder of the Physical Security Plan operates to  !

protect the Reactor Facility Controlled Access Areas and Reactor Controlled Areas. This change in the officially designated security status of the Institute reflects the overall U.S.

Government trend since the end of the cold war toward greater openness in the operation of scientific facilities. Since the access control to the reactor areas remains unchanged, it will have no adverse impact on the physical security of the reactor facility. This i change was submitted to the NRC under a 10 CFR 50.54(p) review as required.

D. OPERATIONAL PROCEDURES:

1. Operational Procedure 8 Tab A (IAgbook Entry Checklist) was expanded to better define and clarify various steps in the procedure. The definition of Before-the-Fact is better described in Section 1 of the procedure. A better explanation of when the word SCRAM needs to be written in the logbook is defined in Section 5.a.(4). Section 5.a.(7) was rewritten to clarify that new operators should be logged onto the console before old operators are logged off of the console. Section 6 was expanded to show a samp6 logbook entry for a typical day with the correct ink color specified in the right colamn beside each en ry. Section 5.b.(2)(c) was added to also specify the color code for CET removal and insertion entries. These changes will help eliminate interpretation of color so that logbook entries can be made in a more consistent manner.

2. Procedure S010 (Measurement of Fuel Elements / Inspection) was changed to enhance the procedure by adding specifications such as doing a shutdown margin calculation before measuring fuel, and by explicitly stating that an SRO is in charge of fuel movements.

Also over a dozen steps in the procedure were expanded to add detail for good communications between operators during fuel movements. These changes were implemented to allow for easier training of, and easier reference by reactor staff.

3. Operational Procedure 1 (Conduct of Experiments) had sections added to specify the .

l'echnical Specification limits on experiments and to describe how to measure

experimental worths. These changes will assist operators in properly measunng '

experimental worth, and provide a reminder of the experimental limits as defined in the Technical Specifications. I

4. Operational Procedure 0 (Procedure 7hanges) was modified to include detailed instructions on how to write (including formatting instructions) and staff a new procedure. This procedure was rewritten to provide guidance for procedure writing so that the visual appearance and format of new procedures will match existing procedures.

This procedure specifies where to find a Word Perfect template with which to start new procedure. The procedure also specifies where to locate staffing sheets for procedures.

5. The Startup and Safety Checklists were changed to correct the wording for the new Stack Gas Monitor (SGM) electronics checks, and to change the normal ranges for two items

.vhich operate on air pressure. The SGM entries were changed due to the installation of the new SGM electronics package. The ranges for the air pressure instruments were changed to allow for normal fluctuations in air pressure. The previous ranges required unnecessary and frequent adjustments. The shutdown checklist was changed to allow for -

blank lines in part VI to verify that the night time high level alarm settings are checked. l Another line was added to the shutdown checklist to ensure that the auxiliary chart recorders are operating and tracing.

6. Procedure 10 (Stack Gas Monitor Procedure) was rewritten to coincide with the new analog electronics for the Stack Gas Monitor. The new electronics perform the same function as the old electronics, however, the method of testing has changed slightly from one electronics package to another.

7. Procedure C006 (Stack Gas Monitor Calibration) was rewritten to coincide with the new electronics unit for the SGM. The new procedure was adapted from the manufacturers calibration literature and clarified in .some areas to allow for case of procedure use.

Testing requirements from HPP 7.3 were also added to the procedure. This procedure l was rewritten because the calibration procedure for the old digital electronics would not function for the new analog electronics.

8. A new procedure was produced to formalize the method of determining the tank constant.

The tank constant at the reactor facility had been determined to be 1.48 at the startup of i the facility and is not routinely measured. This formal procedure was developed from old notes and historical knowledge to document the method of determining the tank constant for the training of future staff. The procedure was repeated several times to verify correctness and completeness.

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i E. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS:

All required maintenance and surveillance items were accomplished by the end of 1996 with the following exceptions.

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1. The SCBA Regulator Annual Flow Test was not accomplished during December as

] requested due to cold weather. The supplier who performs the flow test requires warmer

] weather to perform the test on site. Arrangements are planned to send the regulators to i j another facility where the test can be performed indoors. This' task will be performed

, within the required time frame of 15 months.

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2. The Annual Emergency Plan review was requested on time but as of 31 December a

, report had not been returned from the reviewer. The report is expected during January l 1997 which .s within the required time frame of 15 months.

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3. Security Personnel Training was delayed pending a new contract for the guard force. l Several dates have been scheduled for annual training of the guards during January.

4. The Physical Security Plan review was requested on time but as of 31 December a report

had not been returned from the reviewer. The report is expected during January 1997 l which is within the required time frame of 15 months.

! l All of the above items will be completed within the required time frame. ' Actions have been l taken on each of the above items but the tasks were not fully complete by the end of j December. Eae,1 of the tasks requires action from outside the reactor division.

j Malfunctions discovered are detailed in section IV.

l One outside audit was conducted during 1996. The audit was conducted by Dr. Walt l- Chappas and Mr. Vince Adams from the University of Maryland reactor facility. Dr.

! Chappas is the director of the University of Maryland reactor facility and Mr. Adams is a l licensed SRO at the facility. No safety concerns were fmmd by the auditors.

The Nuclear Regulatory Commission did not inspect the reactor facility during 1996, nor 1 were they required to do so.

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SECTION II i l

Energy Generated by Current Reactor Core and Number of Pulses i

$2.00 or Larger.

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SECTION II ,

Energy generated by the reactor core:  !

Month Kw-Hrs JAN 1069.8 FEB 1.881.5 ,

MAR 6605.7  :

APR 2532.6  !

MAY 3333.4 JUN 2042.7 JUL 619.1 AUG 433 2 4 SEP 205.9 OCT 2281.7 l NOV 70.4 -

DEC 28.4 l

TOTAL 21,104.4 i Total energy generated this year: 21,104.4 Kw-hrs j Total energy on fuel elements: 894823.8 Kw-brs i

Total energy on FFCRs: 162025.7 Kw-hrs '

Total pulses this year 1 $2.00: 32 Total pulses on fuel element 1 $2.00: 4191 Total pulses on FFCRs 1 $2.00: 79 i'

Total pulses this year: 322 Total pulses on fuel elements: 11159 Total pulses on FFCRs: 1394

4 SECTION III Unscheduled Shutdowns ,

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SECTION III Unscheduled Shutdowns:

There were no unscheduled shutdowns during 1996.

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k SECTION IV i

l; Safety-Related Corrective Maintenance 4

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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 equipment to its proper operational status.

20 Feb 96 On two separate occapons the Stack Gas Monitor (SGM) locked up. The CPU board was found to be locsing power. The CPU board was removed, the contacts were cleaned, and the board was reseated into the SGM. A channel calibration was performed and the unit was placed back on line. No further problems occurred with the SGM thereafter.

26 Feb 96 During new operator training on the Data Acquisition and Control unit (DAC),

the DAC failed to reboot. The problem was determined to be a hard drive failure. Discussions with General Atomics and other TRIGA reactor facilities provided instructions to create a new hard drive using an industry standard MFM hard drive. After a successful 50.59 analysis was performed verifying that the I new drive was equivalent to the previous drive, the General Atomics software was installed onto the drive. The new drive was installed in the DAC and extensively tested in all modes of operations, scrams, interlocks, and diagnostic programs for various I/O boards. After all tests were completed successfully, the console was placed back on line.

10 Apr 96 Reactor Radiation Area Monitor (RAM) R1 went into fail mode. The yellow lamp illuminated and the readings went to zero. The Safety and Health Department (SHD) and the calibration shop were notified and the R3 readout unit and detector (R3 is not required equipment) were switched with the R1 units.

SHD calibrated the new R1 unit, and the unit was placed back on line. The defective RAM readout and detector were sent to the calibration and repair shop.

A 555 timer was determined to be the problem with the electronics in the readout unit. After repair of the RAM, the unit was placed into the R3 location and calibrated by SHD. After a successful calibration the R3 unit was placed bac.k on line. RAM unit R1 is a required piece of equipment and was replaced and calibrated before any operations were permitted.

07 May 96 Security alarm point 6.1 (Entrance Door to Office Area) alarmed and could not be cleared. Diagnosis found that the Balanced Magnetic Switch (BMS) switch which provides this alarm had become misaligned through repeated use of the door. The switch was realigned so that the alarm point would operate properly.

The door was operated multiple times to verify that the alarm point was once again performing normally.

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i 30 Sept % During the daily startup, the REG rod drive failed to drive up. Diagnosis of this problem determined that there was a faulty connection in the REG rod drive mountin; tuot. The connection in the foot was repaired and the drive operated

]; normally. The REG rod drive was tested, and returned to service.

30 Sept % During repair of the previous problem, the console power was switched off and on multiple times for various diagnostic purposes. After one such power-up the i

low resolution monitor in the Control Systems Console (CSC) failed to come on.

Diagnosis of the rod drive problem was suspended until this problem was l corrected. The problem was diagnosed to be a broken monitor. A replacement l

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! computer monitor was disassembled and installed inside the CSC. The monitor {

} was plugged into the CSC computer and when the console power was turned on the monitor operated normally, displaying data as expected. The console was l

placed back on line, and diagnosis of the previous problem resumed.

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, 30 Sept % While still diagnosing the REG rod drive problem, after a DAC power up, smoke l came from the labmaster daughter board box. The smoke was discovered to have l j come from a capacitor inside the daughter board box which failed due to a higher I than expected voltage from a failed DC to DC voltage converter located inside l

- the box. A new daughter board, complete with power supply, was installed in l

the DAC. After repair of the previous CRT problem and rod drive problem, )

! several zero power pulses were fired with signals injected into the Safety / Pulse )

channel to verify the proper data collection capabilities and console data output.  ;

All systems operated properly and the daughter board unit and the console were

placed back on line.

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! 07 Oct 96 During the daily startup, the DAC failed to collect data during a zero power l i pulse. The problem was suspected to be with the software. The CSC and DAC j

] were rebooted as recommended by General Atomics, the manufacturer of the i i console. The zero power pulse was repeated and data collection occurred as )

1- normal. The consele was then operating normally and the startup was continued.

i 07 Oct 96 The core trailing boom had become disconnected from the core dolly. Between

the core dolly structure and boom there is a short metal arm which holds the boom at a fixed distance to prevent putting unnecessary strain on the wiring hanging from the boom. Onejoint on the metal arm had simply come unscrewed and had to be reattached. After reconnecting, and tightening of all the remaining connections, operations were resumed.

15 Oct % An operator found the reactor pool water to be warmer than normal after a weekend. The Reactor Facility Director suspended operations until the anomaly was investigated. The problem was discovered to be that a non-reactor air compressor had malfunctioned allowing the cooling tower bypass valve to shut off water flow to the cooling tower. A calculation Was performed to verify that the increased temperature of the reactor water was due solely to the mechanical heat added by the two cooling system pumps. Logistics was notified of the

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! problem and when an immediate fix to the air compressor was not possible, air j was resupplied to the bypass valve by opening a backup air supply valve which supplied air to the cooling tower bypass valve from the reactor air compressor.

! 23 Dec % The stack air flow rates were found to be running below normal during the daily startup, Investigation of the problem showed that an AFRRI house air '

compressor was supplying air pressure below the required 18 psi to the ventilation system dampers and the dampers were partially closed. This restriction in the

air flow through the system caused the lower than normal readings on the control room chart recorders. Logistics was notified of the problem and the pressure was

] corrected.

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SECTION V

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! Facility Changes and Changes to

! Procedures as Described in the  !

i Safety Analysis Report. New i

j Experiments or Tests During the Year.

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

A. The boundaries for the restricted areas in the Physical Security Plan for the TRIGA Reactor Facility were changed from including the entire AFRRI complex to just the reactor areas. The entire AFRRI complex was designated as a DoD controlled access area. Removal of the formal designation of the entire AFRRI complex as a DoD Restricted Area will change the reference to the Internal Security Act of 1950, but will not change the manner in which the remainder of the Physical Security Plan operates to protect the Reactor Facility Controlled Access Areas and Reactor Controlled Areas. This change in the officially designated security status of the Institute reflects the overall U.S.

Government trend since the end of the cold war toward greater openness in the operation of scientific facilities. Since the manner in which access to the reactor areas is controlled remains unchanged, it will have no adverse impact on the physical security of the reactor facility. This change was submitted to the NRC under a 10 CFR 50.54(p) review as required.

B. There were no new experiments or tests performed during the reporting period.

Attachment B contains the 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 Modification. This procedure utilizes a step-by-step process to document that there were no unreviewed safety questions and no changes were required to the Technical Specifications.

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I SECTIONS VI through VIII

Summary of Radioactive Effluent Released.

l Summary of Radiological Surveys.

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j Exposures Greater Than 10% of 10 CFR Limits

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SECTION VI Summary of Radioactive Effluent Released:

A. Liquid Waste: The reactor produced no liquid waste during 1996.

B. Gaseous Waste: There were no particulate discharges in 1996.

i The total activity of Ar-41 discharged in 1996 was 5.2 Curies. The i estimated average activity release was less than 1% of the permitted effluent concentration for unrestricted areas (Table 2 of Appendix B to 10 CFR 20).

I i Quarterly: Jan - Mar 1996 2.3 Ci Apr - Jun 1996 2.3 Ci Jul - Sep 1996 0.3 Ci Oct - Dec 1996 0.3 Ci C. Solid Waste: All solid radioactive waste material was transferred to the AFRRI byproduct license; none was disposed of under the R-84 License.

SECTION VII Environmental Radiological Surveys:

A. Environmental sampling of soil and vegetation reported radionuclide levels that were not above the normal range. The radionuclides that were detected were those normally expected from natural background and from long-term fallout.

B. The environmental monitoring (TLD) program reported the following results at a 95%

- confidence level for 1996:

1. The annual average reading of approximately 20 thermoluminescent dosimeters (T1.D) background stations, located 2 miles or greater from the AFRRI site, was determined to be 82 i 13 millirem.

2. The annual average reading of approximately 30 environmental TLD stations located on the NNMC grounds was determined to be 78 i 8 millirem.

i The data in this section was reported incorrectly in the 1995 Annual Report by the Safety and Health Depanment. The data given was reported as annual but was in-fact only the average quarterly readings. The correct data for CY95 can be found in Attachment D.

C. The reactor in-plant surveys, specified in HPP 3-2, did not exceed any of the action levels specified in HPP 0-2.

i SECTION VIII  :

l Exposures Greater than 10% of 10 CFR 20 Limits:

There were no doses to reactor staff personnel or reactor visitors greater than 10% of 10 CFR  ;

20 occupational and public radiation dose limits. i b

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ATTACHMENT A Revised Reactor Administrative and I

Operational Procedures  :

Procedure 0: Procedure Changes Procedure 1: Conduct of Experiments

, Procedure 8, Tab A: Logbook Entry Checklist i 4

Procedure 8 Tab B: Startup Checklist Procedure 8 Tab B1: Safety Checklist Procedure 8 Tab 1: Shutdown Checklist

Procedure 10

Stack Gas Monitor Procedure j Procedure C006: Stack Gas Monitor Calibration 1 Procedure S005: Tank Constant Verification Procedure S010: Measurement of Fuel Elements / Inspection l

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WW-t ' <<r y fY WN .r p.. N $NWDWYNWSYW'UWNWWW

~ bNmYA Sv sr .~ & Lw N* n s L i

Procedure Changes A l 1

GENERAL 3

4 This establishes procedures for ,

} 1. Introducing new procedures. '

! 2. Permanently or temporarily changing reactor operational or

, administrative procedures.

i j j SPECIFIC j l. Introducing a new procedure l 1. Using the current reactor word processor, generate a rough draft of the new procedure. The following should be addressed:

l

a. Title and Number of Procedure l

i. The title should be a short description of what the i procedure is.

l

) ii. The number has two parts, (1) the procedure type:

i (a) Administrative 4 (b) Operational l; (c) Maintenance, etc

(2) the next sequential number for that type of procedure in j the TRIGA Tracker

, b. General:

j i. Reference - Other documents that provide relevant information.

i- ii. Requirement - What is the requirement for this j procedure?

lii. Tools / Equipment - List of needed tools and equipment.

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. Revised: 16 APR 96 R:\PROCEDWP\0P_,0.WP6 Page 1 I ~

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y y , - - - - . . . .

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

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iv. Coordination - Personnel other than Reactor Staff

v. Staff Requirements - Reactor Personnel requirements vi. Safety Precautions - Self-explanatory O. Procedural Sequence

l. This section should clearly explain each step of the procedure that will be followed.

d. Associated Tables, Graphs, or Diagrams

i. This section should have blank copies of any tables which must be filled out for records, and samples of any graphs which must be generated.

ii. Any diagrams depicting how to setup equipment should also i be included in this section.

2. Use the current version of the reactor word processor to open the tile called: r:Vormsprocedur.frm

3. Modify the top two blocks and the footer of this file in accordance with instructions found in the file.

. l

4. Transfer the text of the new procedure (using " Paste") where indicated in  !

"procedur.frm" (The standard font for procedures is Arial 12 point.) j

5. Save the procedure as " number.xxx" where "xxx" represents the current word processor.  !

, for example, procedure M049 created in Wordperfect 6 l would be named: "M049.wp6" l l

6. - Print the procedure.

7. Open the file called  !

r:Vorms\procedur.dra

8. Modify the file as below;

a. Modify the top block to the appropriate type and number of procedure. For example:

4 Revised: 16 APR 96 R:\PROCEDWP\0P._0.WP6 Page 2

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i MAINTENANCE PROCEDURE M049

b. Add comments as desired below the signature block.

9. Save as sopropriate.

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10. Print out th 3 draft staffing sheet  ;

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11. This step may be performed to generate the 50.59 worksheet for this  !

procedure change. The 50.59 may also be done manually. l

a. Open the file called r:\ forms \ form 50.59 l

b. Fill out 50.59 worksheet, print, and save as appropriate l

12. Submit procedure to staff for comments with draft cover sheet, 50.59  !

worksheet, and table of contents, if needed, included in packet  ;

note: The Senior Reactor Staff Member should be the i last to review the document.  !

If applicable, staff in accordance with para 11.7, (i.e. Health Physics issues or cases involving building j modifications.) i I

13. Make appropriate corrections and re-staff with a new draft cover sheet. I

14. Repeat steps 12 and 13 until there are no further changes. i I

15. Open the file called r:\ forms \procedur.cov. {

Repeat step Ba.  ;

Print out final cover sheet.

16. Submit to RPO for approval if required. l Submit to FD for approval.  !

i

17. Staff final version for signatures.

18. Submit to RRFSC for review f i

19. Submit electronic copies of the procedure to Reactor Manegement for  !

storage on read-only network hard drive.  !

Revised: 16 APR 96 R:\PROCEDWP\OP 0.WP6 Page 3 [

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20. Submit procedure to TRIGA Tracker manager for table of contents update. '

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21. After table of contents has been updated, place procedure in TRIGA Tracker book.

11. Changing a procedure ,

i

1. Permanent changes are made by revising the entire procedure. The r revised procedures will be approved by the Reactor Facility Director (RFD) and reviewed by the Reactor and Radiation Facility Safety ,

Committee (RRFSC).  !

a. For procedures in 'Ventura Publisher" format, Load procedure into Ventura Publisher.

4 i.

The procedure should be located in r.\proced\) ,

! ii. Save procedure as an ASCil file.

f iii. Load procedure into current word processor. I iv. Make appropriate changes and manually add margin lines ,

i for changes.  !

v. Follow steps 4 - 21 from Section I above. '

b. For procedures already in proper format, j

i. Open the procedure file. (Located in r:\procedwp\)

ii. Modify the footer to reflect current date. <

, iii. Make appropriate changes, add margin lines for changes, and save document as appropriate.

iv. Follow steps 8 - 21 from Section I above.

2. Temporary changes will be documented on a separate sheet attached to the current procedure and implemented when initialed by the RFD, acting RFD, or Reactor Operations Supervisor (ROS) in accordance with Technical Specifications 6.3.2. These changes must be 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 and changes (temporary or permanent) will be staffed as

• drafts" for staff comments, and staffed again after being approved by the RFD. The final version will have a signature block for all operators and reactor staff members. Operators will review new or modified procedures Revised: 16 APR 96 R:\PROCEDWP\0P_0.WP6 Page 4

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l and sign the signature block prior to performing a new or modified procedure. When the block is completed, the procedure will be placed in the Reactor Procedures Binder and kept available for review in the control room.

5. All changes will be accomplished under the following guidelines:

a. The change will result in no decrease in the s afety of the actions being addressed.

b. The change will result in no decrease in the effectiveness of the procedure performance.

c. The change will not decrease the ability of the procedure to perform its intended function.

7. Procedures will be staffed under the following guidelines:

a. Procedures that may affect areas outside the reactor area, such as building changes, security, etc., will be staffed to the appropriate office (s) before further routing.

b. - Procedural changes that deal specifically with health physics procedures or radiation safety issues will be staffed through the AFRRI Radiation Protection Officer (RPO)

c. All changes will be staffed to the AFRRI TRIGA Reactor Facility staff 1

d. All changes will be reviewed by the Reactor and Radiation Facility 1 Safety Committee (RRFSC) i I

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Revised: 16 APR 96 R:\PROCEDWP\OP_0.WP6 Page5 1

-[ OPE R_ITIOR__.IN_E._56_Ei5E*7MEF"T"9"WT"TWu*r"ej acummammumammam m u ,__ P CONDUCT OF EXPERIMENTS t

1. 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 televisions l (CCTVs) in the exposure rooms and over the reactor pool can be used to meet this  !

I requirement.

2. All experiments will be set up so as to preclude movement unless the experiment l apparatus is designed for movement (such as rotators, etc.). l

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 I styrofoam surface to prevent sanitation problems from the animal waste.

l 4. The Reactor Staff will conduct a thorough inspection of all experiments to determine that no unauthorized materials are irradiated.

5. ALARA will be practiced during all experiments.

l

)

11. SPECIFIC I

1. A Reactor Use Request (RUR) is required for any experiments included under l

authorizations outlined in the Technical Specifications, section 6.4.2.a. and section l 6.4.2.b. RUR's are not required for reactor parameter measurements as outlined in the Technical Specifications, section 6.4.2.c. Any experiment perfcimed by the reactor staff (except under 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). Any irradiation of materials shall be in compliance with the limitations in the Technical Specifications, section 3.6. to prevent damage to the reactor or excessive release of radioactive materials.

2. Limitations on Experiment: The following limitations shall apply to the irradiation of  !

materials.

a. If the possibility exists that a release of radioactive gases or aerosols may occur, the amount and type of material irradiated shall be limited to assure the yearly i

Revised: 4 March 1996 R:\PROCEDWP\0P 1.WP6 Page 1 1

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compliance with Table 2, Appendix B, of 10 CFR 20, assuming that 100% of the gases or aerosols escape.

1

b. The total inventory of iodine isotopes 131 through 135 in the experiment is not greater than 1.3 cunes and the maximum strontium-90 inventory is not greater than  :

5 millicuries.  !

j

c. Known explosive material shall not be irradiated in the reactor in quantities greater  ;

than 25 milligrams. The pressure produced in the experiment container upon i detonation of explosive shall have been determined experimentally, or by i calculations, to be less than the design pressure of the container.

d. Samples shall be doubly contained when release of the contained material could  !

cause corrosion of the experimental facility.

e. The sum of the absolute reactivity worths of all expenments in the reactor and in the associated experimental facilities shall not exceed $3.00. 3

)

3. Determination of Experimental Worths: The experimental worth shall be measured as )

follows.  :

I

a. The movement of necessary fuel elements is required before conducting the j K-excess if one of the following limitations would be exceeded: )

(1) The absolute worth of materials, determined experimentally or by calculations,  ;

would exceed $3.00. l (2) The maximum available excess reactivity above cold critical with or without all  :

experiments in place would exceed 55.00.  !

I

b. Perform K excess without the experimental materials / setup.

t

c. Perform the K-excess with the experimental materials / setup.

d. The difference between K-excess performed in part b and c is the worth of  !

experiment, i.e. K-excess (b) - K excess (c) = worth.

4. Experiment Review (Processing of RURs):

a. Check the RUR for completeness (Section 1 should be filled out).

b. Forward the RUR to the Radiation Sciences Department, Dosimetry Division (RSDD) if dosimetry support is required.

c. Forward the RUR to the Safety & Health Department (SHD) for radiological safety Revised: 4 March 1996 R:\PROCEDWP\OP 1.WP6 Page 2 1 t

i coordination. l

d. Check experimental protocol against reactor authorizations,

e. Fill-in Section 4 of RUR with special instructions, as appropriate. Assign an RUR  !

sequence number. Write in estimated or measured experiment worth and the core position of the experiment facil@ to be utilized in the appropriate blocks (lower part [

of form).

i-  ;

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.

5. Conduct of Experiments. Perform setup and irradiation of experiments in accordance with the following procedures:  !

I

6. Exposure Room Entry -TAB A.

b. Core Evoeriment Tube (CET)-TAB B.  ;

c. Extractor System - TAB C.

d. Pneumatic Transfer System (PTS) - TAB D.

e. In-pool /In-core Experiments - TAB E.  ;

6. Complete the RUR by filling out Section 5 with the appropriate information, f 1

7. Attach form to clipboard in the control room. l i

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Revised: 4 March 1996 .

R:\PROCEDWP\0P_1.WP6 Page 3

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LOGBOOK ENTRY CHECKLIST l

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1. Operational Logbook

1. The reactor operations logbook is a "before-the-fact" record, that is, entries will be i logged whenever possible before the operator actually performs the operation. l Events, such as scrams, which may not be planned ahead of time, will be entered at the time of occurrence. Any late entries will be so noted. Entries about what you plan to do are not necessary, only actual events need to be logged in the logbook.

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 completion; 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 responsible for ensuring that the ROS is notified that the logbook is ready for RFD review.

4. All items 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: Most Operational Activities.

(1) Console locked and unlocked. The individual at the console will enter his/her name and the supervisory licensed operator's narne, if necessary.

Revised: 31 Jan 96 R:\PROCEDWP\OP_8A.WP6 Page1 l

l (2) Checklist number and completion time. j (3) Power level at criticality and subsequent power level changes.  ;

(4) Reactor SCRAM This entry to be added when the operational stamp I i

does not contain the SCRAM word in the last line. The K-Excess stamp does not specify SCRAM time. Also any time the operator deviates from normal stamps for multiple power changes for short periods of time, tho' final line should be SCRAM. Console locked does not futRIIthis i requirement even though the reactor scrams.

(5) Mode of operations. Use appropriate stamp or entry to designate the operation:

(a) Steady State.

(b) Square Wave (c) Pulse (6) Operation of reactor associated facilities such as lead shield doors, pneumatic tube systems, etc., unless such operations cause a change of reactivity (see 5.b.(2) below).

(7) Change of personnel at the console. When a change of operator is noted in the logbook, the name of the person replacing the person on console ,

should be entered first as "on console" before the person logging off of the console is entered as "off'.

(8) The operator in charge will be designated in the logbook whenever multiple 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 the log for the day.

! (11) Designation of the SRO on-call and physicist in charge (PlC).

(12) Reactor calibrations and data.

l (13) All changes to logbook entries (including line outs, error corrections, I changes to operations mode stamp lines, and end-of-page line outs) will Revised: 31 Jan 96 R:\PROCEDWP\OP 8A.WP6 Page 2 L

be initialed or signed by the operator.

f

]' b. RED: For items Which Change or Measure Reactivity (1) K-excess measurements, to include experiment worth determinations. ,

q (2) Actions which affect reactivity:

l j (a) Core movement. l t

] (b) Fuel movement. J (c) Control rod physical removal for maintenance. -

4 (d) Experiment loading and removal from the CET, PTS, pool, or core. l t

(e) Removal or insertion of CET into core.  !

c. GREEN: For Maintenance or Malfunctions '

(1) Any reactor malfunctions noted upon discovery / occurrence with a second l entry noting corrective action has been completed. I (2) Additional items entered at the discretion of the opera'.or such as addition of make-up water to the resctor pool, etc. l (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. i (4) Movement of detectors or chambers from above core. l

6. When an operation requiring entry into the logbook falls under more than one color l

code, the color to be used will be determined via the following order of precedence:  !

RED - GREEN - BLACK / BLUE-BLACK.  !

Sample Logbook Entries for a Typical Operational Day SRO/PIC/HP Stamp, Date Stamp BLACK ,

Startup Checklist Begun BLACK l Console unlocked by - BLACK I Startup Checklist #"" Complete BLACK Revised: 31 Jan 96 R:\PROCEDWP\0P_8A.WP6 Page 3 I

RAISING RODS TO GO CRITICAL RED i . CRITICAL AT WATTS FOR K-EXCESS j

TRANS $ SHIM $

SAFE $ REG $

(

CORE POSITION K-EXCESS $ i i

1  :

i  !

[.

Manual Scram BLACK  !

t l- Console locked by - BLACK l i SGM out of service for maintenance No operations GREEN l j SGM Back in service GREEN Console unlocked by - BLACK  !

Opening Pb Doors BLACK j

Moving Core to " RED ~t

! Closing PB Doom BLACK j Rabbit (containing ") inserted into CET RED l l

l RAISING RODS TO GO CRITICAL BLACK I TRANS T1 MAX C  !

SHIM T2 MAX "C l SAFE RUR#  ;

R E G ___ TOTAL KWHRS .

CRITICAL AT WATTS i SCRAM, TOTAL TIME MIN SEC ,

F Rabbit removed from CET RED Console locked by == BLACK Shutdown Checklist Begun BLACK Shutdown Checklist #" Complete BLACK Page lined out, Page signed by SRO BLACK  !

Revised: 31 Jan 96 R:\PROCEDWP\0P_8A.WP6 Page 4

ll. Malfunction Logbook All entries in the malfunction logbook should include the following information. For consistency, the bold words should be copied into the malfunction log prior to the information.

DATE, TIME, SIGNATURE OF PERSON DISCOVERING MALFUNCTION SYMPTOM:

This section describes how the system is acting or malfunctioning,i.e., channel went full scale, pump failed, keyboard stopped responding to keystrokes etc.

IMMEDIATE ACTION TAKEN:

This section is for denoting such things as Reactor Secured, SHD notified.

RFD NOTIFIED:

A remark should be made that the RFD or acting RFD was notified.

DIAGNOSIS : of problem A narrative description of what was discovered to be causing the problem, i.e.,

Which system was malfunctioning or which component failed.

SOLUTION: / repair A narrative description of what was done to correct the problem This could include both physical changes or administrative changes, i.e., a component was replaced and the unit was recalibrated, an additional backup system installed, an administrative prohibit on . . was initiated.

OPERATIONAL VERIFICATION AND/OR CAllBRATION:

A desciiption of what actions were taken to verify that the new unit / repair would indeed perform the function for which it was intended, i.e., a calibration signal, system actuated multiple times, system tested, system calibrated with a source.

Indicate whether the change will require staff training.

SIGNATURE RFD Revised: 31 Jan 96 R:\PROCEDWP\OP_8A.WP6 Page5

4Ypdkb6iVhI th') $ 9a htJ[ . 'gh j, lh> f .

DAILY OPERATIONAL STARTUP CHECKLIST Checklist No. Date l Time Completed Supervised by Assisted by I. EQUIPhENT ROOM (Room 3152)

1. Air compressor pressure (80 - 120 pa.g) .......... ......... .. ......

2. Water drained from air compressor . ........ . . . .. .. .

3. Air dryer operating .. . . ........ ...... ..... . . .. . .

4. Doors 231,231 A, and roofhatch SECURED .. .. .. .... ..... .

II. LOBBY AREA I

Lobby alarm turned off . . . . . . . . . . . . . ...... . ..... .... ... ....

III. EQUIPMENT ROOM (Room 2158)

1. Prefilter differential pressure (< 8 psid) . . ....... . . ..

2. Primary discharge pressure (15 - 25 psig) ...... .......... .. .... .

3. Demineralizer flow rates set to 6 gpm (5.5 - 6.5 gpm) . .. . . .......

4. Stack roughing filter (notify supervisor if > 1.0" of water) . . .. .

5. Stack absolute filter (notify supervisor if > 1.35" of water) . ... ... .

6. Visualinspection of area .................... .... .... . .. .

l

7. Door 2158 SECURED ... ....... ...... .... .. . . . ...  !

IV. PREPARATION AREA i

Visualinspection of area .. . . .. .. .. . . .. .. . . ... .

V. REACTOR ROOM (Room 3161)

1. Transient rod air pressure (78 - 82 psig) . .... . . .

2. Shield door bearing air pressure (8.5 - 11 psig) . . .. . .

3. Visualinspection of core and tank .. . .. .. . . .

4. Number of fuel elements and Fuel elements

• control rods in tank storage Control rods *

5. Air particulate monitor (CAM) i (a) Primary operating and tracing . .. . .. .. .... . .. . .

(b) Backmp operating . . . ... ... . ... . .. ..

(c) Channel test completed, damper closure verified .. ..

6. Channel test completed on SGM . . . . . . . . . . . . . . . . . . . . . . ........

7. Door 3162 SECURED . .... . ..... .. .. . . ..

• Numerical Entry AFRRI Form 62a(R) Revised: 01 May 96 R:\PRDCEDWP\0P_8B.WP6 Page 1 i

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VI. REACTOR CONTROL ROOM (Room 3160) l

1. Emergency air dampers reset . ...... ... ... . . ... . . .

2. Console recorders dated . . .. . .... .. ... .

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3. Stack flow and fuel temperature recorders dated . . . .. . ,

4. Logbook dated and reviewed ... . ... ... . ... . . .. . ..

l

5. Water monitor box j (a) Background activity (10 - 60 cpm) . . . . . . .. . .

(b) Water monitor box resistivity (> 0.2 Mohm-cm) . .. ......

• i (c) DM1 resistivity (> 0.5 Mohm-crn) .. ... .

(d) DM2 resistivity (> 0.5 Mohm-cm) ..

6. Stack gas flow rate (15 - 35 Kcfm) . ... . . .

7. Stack linear flow rate (1.0 - 2.0 Kil/ min) . . ... . . . .. . .

8. Gas stack monitor (a) Background (2 - 20 cpm) . . . . .

(b) Alarm check . . . .. . .

(c) High alarm set to 3.2 E-5 microCi/cc at stack top .

(d) SGM chart recorder operating and tracing .. .. ...

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

• 500 (a) R-1 (< 20)

(b) R-2 (< 10) 10

• 10 (c) R-3 (< 10)

• 100 (d) R-5 (< 20)

(e) E-3 (< 10) 10 (f) E-6 (< 10) 10

10. TV monitors on . . . .

I1. CAM high level audible alarm check . . . .

12. Water temperature (inlet)(5 - 35 *C) . ... ..

13. Water level log completed . ..

14. Console lamp test completed . . .. . .. .. .

15. Time delay operative . ..

16. Source level power greater / equal to 0.5 cps .

l'i. Prestart operability checks performed . .

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

19. SCRAM checks (at least one per rod)

(a) % Power 1 (h) Reactor key (b) % Power 2 (i) Manua!

(c) Fuel temp 1 (j) Emergency Stop (d) Fuel temp 2 (k) Timer (e) HV loss I (1) CSC Watchdog (f) HVloss 2 (m) DAC Watchdog (g) Poollevel

20. Zero power pulse ... .

' Numerical Entry AFRRI Form 62a (R) Revised: 01 May 96 R:\PROCEDWP\OP_8B.WP6 Page 2

ma,am3'M Qi:*ej% h:&Lj;)+Q$$klg Q Q & Njj"fjf%[ ,

l DAILY SAFETY CHECKLIST

< Checklist No. Date Time Completed Supervised by Assisted by I. EQUIPMENT ROOM (Room 3152)

1. Air compressor pressure (80 - 120 psig) ........... .. ...... .....

2. Water drained from air compressor ....................... .........

3. Air dryer operating . ................... ................ .....

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

II. EQUIPMENT ROOM (Room 2158) ,

1. Prefilter differential pressure (< 8 psid) .. ... ....... ..... . . ..

2. Primary discharge pressure (15 - 25 psig) .......... ... .... . .. ..

3. Demineralizer flow rates set to 6 gpm (5.5 - 6.5 gpm) . . . . . . . . . . . . . . . . . .

4. Stack roughing filter (notify supervisor if > 1.0" of water) .... ...... ....

5. Stack absolute filter (notify supervisor if> 1.35" ofwater) . .... .........

6. Visualinspection of area .... .. ..... . .......... ....... . ..

7. Door 2158 SECURED .... . ..... ........ ..... . ..........

III. PREPARATION AREA Visualinspection of area ........ .. ....... .... ... ......

IV. REACTOR ROOM (Room 3161)

1. Transient rod air pressure (78 - 82 psig) . .. ... . . .... .....

2. Shield door bearing air pressure (8.5 - 11 psig) .. .... .... ... ..

3. Visualinspection of core and tank . . . . .. . ..... ... . . ..

4. Number of fuel elements and Fuel elements control rods in tank storage Control rods

5. Air particulate monitor (CAM)

(a) Primary operating and tracing . . . ... ..... . .. .

(b) Backup operating . ... ..... ... ........ ..

(c) Channel test completed, damper closure verified ... ... .. . .

6. Channel test completed on SGM ........... .. ........... ....

7. Door 3162 SECURED . ..... . . .. . ... . .....

• Numerical Entry AFRRI Form 62b (R) Revised: 26 APR 96 R:\PROCEDWP\OP,_8Bl.WP6 Page 1

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

V. LOBBY AREA ^

i Lobby audio alarm tur ned off . .. . . . .. . . ... .. ... . . . . . ..... .. . . . . ... .. . . . ........ . . . .. . . .. . ... .. l 1

i VI. REACTOR CONTROL ROOM (Room 3160)

1. Emergency air dampers reset

2. Console recorders dated .... ........ ........... ... . ..... .

3. Stack flow and fuel temperature recorders dated . . . . . . . ..... . . .

4. Logbook dated and reviewed .. ... ........ ...... .. ..

5. Water monitor box (a) Background activity (10 - 60 cpm) . . . . . . . . . . . . . . . . . . . . . . . , , *

(b) Water monitor box resistivity (> 0.2 Mohm-cm) .. ... .... ..... *

(c) DMI resistivity (> 0.5 Mohm-cm) ....... ...... ....... ..

(d) DM2 resistivity (> 0.5 Mohm-cm) ...... . .. .. ......

6. Stack gas flow rate (15 - 35 Kcfm) .. .. ........ .. .. ........... *

7. Stack linear flow rate (1.0 - 2.0 Kft/ min) . . . . . . . . . . . . . . . . . . . . . . . . .

8. Gas stack monitor (a) Background (2 - 20 cpm) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . *

(b) Alarm check ... .......... ...... . . .. .. . . ..

(c) High alarm set to 3.2 E-5 microCi/cc at stack top . . . . . .......

(d) SGM chart recorder operating and tracing . . . . . . . . . .. . .. ..

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

(a) R-1 (< 20) 20 (b) R-2 (< 10) 10 (c) R-3 (< 10) 10 (d) R-5 (< 20) 20 (e) E-3 *

(< 10) 10 (f) E-6 *

(< 10) 10

10. TV monitors on .. .. ... .. .. . . .. . ... .. ... .

I1. CAM high level audible alarm check . .. . . . ... ..

12. Water temperature (inlet) (5 - 35 'C) .. . .. . . .. ... ..

13. Water levellog completed . . . .. .. ................ .... ......

14. Source level power greater / equal to 0.5 cps . ....... ..... . . .

1 1

• Numerical Entry I

AFRRI Form 62b (R) Revised: 26 APR 96 R:\PROCEDWP\0P_8Bl.WP6 Page 2 l

k ... - ., p ' k . r T4 . 7< k (+ I lh>c  ;-

DAILY OPERATIONAL SHUTDOWN CHECKLIST l

i. Checklist No. Date '

Time Completed Supervised by -

Assisted by j I. REACTOR ROOM (Room 3161) I j l

1. All rod drives DOWN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l

2. Carriage lights OFF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l i

3. Door 3162 SECURED . . . . . . ... .... ..................... ... ,

4. Channel test completed on both CAM's . . . . . . . . . . . . . . . . . . . . . . . . . . . '

5. Door 3161 locked with key . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

)

II. EQUIPhENT ROOM (Room 3152) 4

. 1. Distillation unit discharge valve CLOSED .. ...... . .... .. ... ..

! 2. Air dryer OPERATIONAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3. Doors 231,231 A, and roofhatch SECURED ........ .... ............

III. EQUIPMENT ROOM (Room 2158)

'I

1. Primary discharge pressure (15 - 25 psig) . . . . . . . . . . . . . . . . . . . . . . . . . . . . *

2. Demineralizer flow rates set to (5.5 - 6.5 gpm) . . . . . . . . . . . . . . . . . . . . . .

• l

3. Visualinspection for leaks ..... .. .......... . . .. . . ... .. 1

4. Door 2158 and 2164 SECURED . . . ....... ....... .... . .. ..  !

IV. PREPARATION AREA

1. ER2 plug door CONTROL LOCKED . .. . .............. ..

Door closed; and handwheel PADLOCKED . . . ... ....... . ..

2. ER2 lights ON and rheostat at 10% . . .. .... ....... .. . ..... .

3. ERI plug door CONTROL LOCKED ... ..... .... . .. ....... .

Door closed; and handwheel PADLOCKED .. .. . ... .. . ..

4. ERI lights ON and rheostat at 10% . . .......... ... . . . .. . .

2

5. Visualinspection of area . . . .. .... . . ..... ...

6. Warm storage doors closed . . . . ........ ... . ...... ...... . .

1 i

i AFRRI Form 62(R) Revised: 06 June 96 R:\PROCEDWPiOP._81.WP6 Page 1 i

1 l

1 V. LOBBY ALARM Lobby alarm audio ON ....... ...... ........................ ..... .

i VI. REACTOR CONTROL ROOM (Room 3160)

1. Reactor tank lights OFF ........ .......................... ...... ,

2. Console chart recorder pens raised ................. ................ _.

3. Steady-state timer OFF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4. Console LOCKED, and all required keys retumed to lock box .. .... .... .

5. Diffuser pumps OFF ...... .......................... ... .. ....

6. Purification, secondary and primary pumps ON . .. . . . .. . .. .

7. Reactor monthly usage summary completed . .... .. ........ ... ..

8. Auxiliary chart recorders operating and tracing ........ .. . ........ ..

9. Radiation monitors . ... .... ... ..... ... . ... ....... ...

MONITOR READING HIGH LEVEL ALARM SETTING (mrem /hr)

a. R-1

• 20

(<20) '

l

b. R-2 (<10) 10

c. R-3

• 10

(<10) d.R-5

• 20

(<20) e.E-3 (<10) 10

f. E-6 (<10) 10

g. R-6

• 10  !

(<10)

• Numerical Entry i

i 1

l AFRRI Form 62(R) Revised: 06 June 96 R:\PROCEDWP\0P_81.WP6 Page 2

o n -,,,n n 7p i

_FE M ))pNPRQQD,yggfibiagy M n . _ _ y m p e n m g m m m . ,, n m e' T. Q , ,gg W Q gre f9f,& 3 STACK GAS MONITOR PROCEDURE GENERAL This procedure specifies how to test the Stack Gas Monitor (SGM) to ensure proper operation of this monitoring device. This lnstrument is used to sample and measure the gaseous effluent Ar* in the building exhausi system.

SPECIFIC

1. DAILY START-UP (operational and safety)

A. Check the air sampling flow rate (should be between 3.5 and 3.9 cubic feet per minute with the chamber door closed). Change the particulate pre-filter if necessary. Adjust as necessary.

B. Press the RED, HIGH ALARM and BLUE, ALERT ALARM setting buttons on the front of the unit and verify the alarm point settings. (Limits +0 / -10%)

High Alarm Point Low Alarm Point C. Channel Test the SGM. Flip and hold the OP/ TEST Switch to the test position and verify that the analog mater reads 3600 ( range 3000-4200). If outside the range of 3200 - 3800, notify the ROS. Reset Alarms.

D. Remove the front cover of the detector shield and slowly insert the check source into the chamber. Verify that the alert lamp and alarm lamps illuminate at their designated alarm points and that the audible alarm sounds at the high alarm point.

E. Remove the source from chamber and press the alarm and meter reset buttons on the front of the unit.

Page 1 Revised: 17 May 96 R:\PROCEDWP\0P_10.WP6 1 1

1

. ( W M N (e @ i $ e N ( @ y @ p y )) $ ip p k g & s M l $ l[4] g M $ $ $ y M g .T. M STACK GAS MONITOR CAllBRATION

1. General:

. I.

Reference:

Tech Specs 4.5; HPP 7.3; NMC Stack Monitor Electronic Test and l Calibration Procedure P/N 0001020-1.

i

, 2. Requirement: The air particulate monitoring system (SGM) shall be calibrated annually, not to exceed 15 months. (HPP 7.3)

3. Tools: Crescent wrench, screwdriver, special calibration connectors.

4. Equipment: Oscilloscope w/ leads, voltmeter, pulse generator, pulse counter, plastic Ar-41 sample beaker w/ tubing provided by SHD

5. Coordination: With SHD to set a date for isotopic calibration and with the ROS/RFD to arrange a date on operations schedule with no other reactor operations.

6. Estimated time: One day

7. Safety precautions: Use caution when working around high voltage sources and minimize exposure to Ar-41 or Na-22 calibration sources.

8. General:

Tum off high voltage and main power before plugging or unplugging any of the circuit boards.

The " unit" refers to the rack mountable electronics section of the SGM.

II. Procedural Sequence:

1. Schedule date for calibration with RFD/ROS and SHD.

2. Assemble required tools and equipment

3. Produce Argon-41 (for argon calibration) with reactor.

Revised: 02 May 96 Page 1

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

Suggest: 2 syringes, 5040 cc P-10 gas irradiated for 5 min at 100 Kw.

1 ELECTRONIC CAllBRATION l

l

4. Turn off the power. Adjust the front panel meter to 10 cpm. l

5. Remove the CRA-14B/91 card and ensure switches SW1, SW2, and SW3 are I open.

6. Remove the IC-13 card and set the dip switches into the following configuration:

S1 Closed SS Open (10% Window) l S2 Open S6 Closed (20% Window)

S3 Open S7 N/C S4 Open (5% Window) S8 N/C

{ Gross counting mode S1 Closed, S2 Open, S3 Open.}  !

{ Spectrometer mode S1 Open, S2 Open, S3 Closed.}

Replace the IC-13 Card.

Place the card extension card into the CRA-14B/91 slot and attach the CRA-14B/91 card to the extension card. l i

7. Disconnect the detector cable from the back of the SGM unit and attach the test I cable to the SGM unit.

Power up the unit.

8. Verify 24 i 4 VDC between pins 19 and 20 on the terminal block inside the back l of the unit. If the voltage is outside this range, replace or repair the power supply.

9. Connect a pulse generator to the detector test cable.

10. Attach a test cable from the jack on the face of the AA-13A/91 plastic face mask to a volt meter. The access hole is just below the red high alarm button.

11 Set the pulse generator to create 16,666 cps (1x10' cpm). '

12. Adjust R33 on the CRA-14B/91 card to give -5.00 i 0.01 VDC.

13. Set the pulse generator to create 166.6 cps (1x10' cpm).

l

14. Adjust R32 such that the analog meter reads 10,000 cpm.

i i Revised: 02 May 96 Page 2 l

{

15. Set the pulse generator to create 16.66 cps (1000 cpm).

16. Verify 1000 cpm on the local analog meter.

17. Adjust the potentiometer on the 0-1 mA card such that the remote analog meter in the reactor control room reads 1000 cpm.

18. Adjust the potentiometer on the 0-10 VDC card to give 1000 cpm on the remote chart recorder in the control room.

19. Monitor the voltage through the AA-13A/91 mask test Jack, step through the following inputs, and verify the following outputs.

Pulse Generator Voiiage @ AA-13A/91 10 cpm 0.00 i 0.15 VDC 100 cpm 1.00

• 0.15 VDC 1000 cpm 2.00

• 0.15 VDC 10,000 cpm 3.00

• 0.15 VDC 100,000 cpm 4.0010.15 VDC l 1,000,000 cpm 5.00 i 0.15 VDC

20. Adjust the potentiometer, located above the yellow fail button, while pressing the yellow fail button such that the analog meter reads about 12-15 cpm. Set the pulse generator to 10 cpm. Verify that the fail alarm lamp illuminates when the analog meter needle drops below 12-15 cpm.

21. Press the meter reset and alarm reset buttons.

1

22. Press the alert and high alarm buttons to note the settings. Increase the count l output of the pulse generator to cross each of these alarm points and verify that the l respective lamps on top of the stack gas monitor illuminate at their set points and l that the sonalert alarms at the high alarm point.

23. Power off the unit. Remove the test cables from the front and back of the unit. j Replace the CRA-14B/91 card back into its slot. Attach the detector cable to the '

back of the unit.

24. Remove the IC-13 card and set the dip switches into the following  ;

configuration for spectrometer mode:

Revised: 02 May 96 Page 3

)

)

l S1 Open ,

SS Open (10% Window) l S2 Open S6 Closed (20% Window)

S3 Closed S7 N/C S4 Open (5% Window) S8 N/C  :

Replace the IC-13 card.

25. Power the unit back on. Turn on the high voltage. Press the meter reset button.

26. Insert the Sodium-22 source slowly into the chamber and verify operability of the unit.

t

27. Determine the proper high voltage and adjust as necessary to find the peak counts for the argon / sodium peak.

A. This is done with the sodium source or a sample of argon in the detector chamber B. Slowly adjust the voltage. Set the voltage such that the maximum counts are ,

read from the analog meter. Be sure that the peak selected is the Argon 41 J (1293 Kev) or the Sodium 22 ((1274 Kev) peak and not the sodium 22 (511 Kev) peak. graph the output vrs. voltage if necessary to find the proper peak.

28. Assist SHD, as needed, in the isotopic calibration using HPP 7-3.

29. After SHD provides the new alarm point numbers, adjust the alarm points.

A. The high alarm point is adjusted by pressing the high alarm button and adjusting the potentiometer located directly above the button to give the proper alarm point reading on the analog meter.

B. The alert alarm point is adjusted by pressing the alert alarm button and '

adjusting the potentiometer located directly above the button to give the proper alarm point reading on the analog rneter.

30. See that a new calibration sticker is placed on the SGM. Change the written alarm points at the appropriate locations (At SGM and Control Room Meter).

31. Obtain and file isotopic calibration report required by HPP 7-3 from SHD.

32. Create decay curve for Sodium-22 source to be used for semiannual source test.

33. Update TRIGA Tracker.

l i

l I

l Revised: 02 May 96 Page 4 i

_ . _ . _ _ _ . . _ _ _ ..m.._.. _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . .

MEMORANDUM TO FILES Re: Calibration of SGM The SGM was calibrated on . The results are as follows.

Step Point Es W ed As Left

8. Voltage TB19-20 24

• 4 VDC

12. AA-13A/91 Jack -5.00 t 0.01 VDC

17. . Control Room Meter 1000

18. Control Room Chart 1000

19. Inject Signals Pulse Generator Voltage AA-13A/91 Meter Reading 0 cpm- (O cps) V cpm 'l 100 cpm (1.666 cps) V cpm l 1000 cpm (16.66 cps) V cpm 10,000 cpm (166.6 cps) V cpm 100,000 cpm (1,666 cps) V cpm l 1,00,000 cpm (16,666 cps) V cpm J Fail Lamp Operates YES/NO Warning Lamp Operates YES/NO High Lamp Operates YES/NO Audible Alarm Operates YES/NO High Voltage Set Point VDC j Alert Alarm Set Point cpm High Alarm Set Point epm Counts Generated From Sodium-22 Source cpm Revised: 02 May 96 Page 5 7

~

1 J

'p 1

i

$$kd(%MGNA7JN{ M M M k% k 7hhd $$%

Tank Constant Verification General

Purpose:

This procedure is used to determine the Tank Constant for l j the reactor pool and core. I Tools / Equipment: 90 KW heater, heater power console, digital thermometer, binoculars, hand calculator, stopwatch, and clamp-on inductive ammeter.

Reference:

Calculator user's manual Coordination: Ensure no reactor runs are scheduled for the following  !

morning. I Staff Requirements: 2 Reactor staff members for setup. j Safety Precautions: 3 phase power hook-up to the heater. Special care should I be taken when working with high voltage.

Cover all electrical connections with non-conductive material to prevent accidental contact with wires or connections.

Procedural Sequence SETUP (the afternoon before the test)

1. Move core to position 700.

2. Retrieve Heater control console from above reactor office spaces.

3. Lower heater into pool, supported by 2 work grates, do not energize at this time.

4. Connect heater to heater control console in accordance with figure 1.

a. Do not close switch on power console Revised: 17 OCT 96 R:\PROCEDWP\S005.WP6 Page 1

i 1

b. Do not plug console into motor control panel  ;

c. Do not energize motor control panel socket l

l. 5. Lower stirrer into pool,' supported by 2 work grates. Place stirrer as close as ,

j reasonable to the heater.  ;

i  ;

i 6.- Lower temperature probe from digital thermometer into pool near core.  !

I i

7. Unplug tank lights, tum on stirrer and leave running overnight. i

8. Secure Primary, Secondary, and Purification system.  !

Tank Constant test  !

i

1. Leave stirrer running during test.

l l

2. Energize digital thermometer and verify operation.

l

' l Verify that heater control switch in not in the "on" position.

3.

4. Plug heater control console into special 175 amp receptacle on motor control psnel.

5. Energize electrical panel and close switch on heater control console.

]

6. Allow Heater to come to steady power and system to equilibrate for 15 minutes.

7. Begin taking temperature readings every 3 minutes, record this data la table 1.

8. Readings may be plotted as you take them either on graph paper or with a computer. Figure 2 is a sample plot of the data from table 2.

9. If plotting points by hand, calculate a linear regression at each point using the hand calculator.

l

10. - Continue to take readings until a steady slope can be determined from the i graph or regressions. (The first few data points may need to be ignored until a steady rise is observed. )This will take approximately 30 minutes.

11. This slope is in degrees per minute, multiple by 60 to get degrees per hour.

l. 12. The slope must then be corrected because the heater and stirrer do not output .

l I

Revised: 17 OCT 96 R:\PROCEDWP\S005.WP6 Page 2 _l

exactly 100 kW of thermal power. The true value for the power is determined by following the steps below,

s. Measure the current in each of the three legs running to the heater using the clamp-on inductive ammeter, not the hard mounted ammeters.

(The clamp-on ammeter is more accurate.)

b. Record the voltage between each pair of load wires using the mounted voltmeter on the heater control console.

c. The total power is equal to the thermal power from the heater plus the thermal power from the stirrer. Historical calculation shows that the thermal power from the stirrer is approximately 2 kW.

Heater power is determined from:

P = {3 I V 40 where P = average power in watts I = average current from ammeter V = average voltage from 3 voltmeters on Heater control console Total power is nominally 97 kW plus 2 kW from stirrer = 99 kW

d. To calculate the tank constant, divide the slope by the average power, then multiply by 100. This will yield:

Where Krc is the Tank Constant 1

? is the slope of the temperature vs time graph di P is the average power

13. Place graph and result for Krc in core physics log.

l Revised: 17 OCT 96 R:\PROCEDWP\S005.WP6 Page 3 i

Figure 1: Hookup of Heater to Control Console From line Ammeter leads not shown OOO v

A/s/s x -Hardwired inductive

) } } ammeter Breaker box

//

To heater (load) nma

1. Run 3 leads from power supply in through top 3 holes on Heater control console and into top of breaker box.

2. Run white ground wire from power supply to grounding screw on Heater control console.

3. Run 3 leads from heater (load) through inductive ammeters mounted on Heater control console, through lower 3 holes, into bottom of breaker box.

4. Run leads from voltmeter through lower 3 holes, into bottom of breaker box.

Revised: 17 OCT 96 R:\PROCEDWP\S005.WP6 Page 4

Figures and charts were produced using Microsoft Excel 4.0, however any appropriate software package may be used.

Figure 2 is Microsoft Excel file r:\procedwp\S005_f2.xl4 Table 1 is Microsoft Excel file r:\procedwp\S005_t1.xl4 Table 2 is Microsoft Excel file r:\procedwp\S005_t2.xf4 i

l Revised: 17 OCT 96 R:\PROCEDWPiS005.WP6 Page5

.bb.hNIbbk$f. bbk $ bh!!k b b b h$ b h b b f f

MEASUREMENT OF FUEL ELEMENTS /lNSPECTION l i

1. General:

1.

References:

Tech Specs Sections 4.2.5 and 5.2.2e l

2. Requirement: All fuel elements, to include Fuel Followed Control Rods,(FFCR's), shall be inspected for damage and  !

i measured for length and bow annually or after every {

500 pulses of $2.00 or larger, whichever occurs first  ;

3. Tools: Fuel handling tool

4. Equipment: Fuel measuring tool, standard fuel element l S. Coordination: RFDIROS [

6. Estimated time: 3-4 days to measure all fuel  !

7. Safety precautions / protection: The fuel should be kept under enough water to prevent any RAM's from alarming. Typically 4' of  ;

water will be enough. (Set R1 and R5 for 20 mrem /hr.) ,

II. Record Keeping: i This section describes how to make entries in the Operational Logbook and the Fuel Element Record Book.

Logbook Entries: (in red ink)

The operator should enter the time that the element was removed from the core and a statement such as the following in the logbook: Element #" number" ,

i removed from " grid position", measured, and replaced. Write " Verified"  ;

if the serial number was visually verified. f Fuel Element Record Book Entries:

f A. Stamp the date in the date column.

B. Enter the current length, (read from the measuring device). The Revised: 03 May 96 Page 1  ;

\

t i

f i

measurements are recorded in thousandths.  ;

C. Determine the difference between the initial measured length listed i in the top section of the sheet and the current length and enter the difference in the growth column.

D. If the element fits into the measuring tool it passes the bow test. This should be noted with a check mark on the fuel element record.

i E. Record the temperature of the pool water in the temp column. .

F. Record the current core grid position in the core position column.  !

G. Enter the current number of pulses for that element under the pulses column. If the element was removed during the lastyearit may have missed some pulses. Take this into account when calculating total pulses.

H. Enter the total core power in Megawatt-Hours under the total power column. If the element was removed during the last year take into account anypower the element may have missed during its time out of core. 1 1

1. The operator on console should watch for any element that has gone  !

over 8-10 years since its last verification and have that element {

verified. Write ' Verified" in the right column if the serial number was visually verified. >

lli. Procedural Sequence:

1. Determine total power and number of pulses on each element since last fuel measurement and update each fuel record sheet by adding to previous total.

Be careful to account for any fuel movements during the year that will affect the total energy on the element.

2. Perform a shutdown margin calculation. (Unloading 6 F-ring elements and 2 E-ring elements should provide an adequate SDM.) The shutdown margin calculation should be checked by the RFD or ROS.

Revised: 03 May 96 Page 2

3. Remove all chambers and experiments from vicinity of core. Place the operational channel outside the core but close enough to qualitatively measure the reactor power level.

4. Place 3 large work platform plates on rails to cover pool.

5. Place fuel measuring holder in pool by hanging it on the inside of the shroud just to the left of opening; secure in place with a C-clamp.

6. Using a fuel handling tool, place the standard element into the holder and allow them to come to temperature equilibrium in reactor pool. This takes about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

7. Turn off primary system pumps to minimize ripples in pool (optional).

8. Place notched end of fuel measuring extension rod on the top of the standard rod. Place the other end under micrometer extension rod (photo A).

9. Adjust micrometer to zero.

10. While fuel measuring / moving activities are being executed, a licensed operator must be on console and the fuel movement operation must be under the supervision of an SRO. The operator on console may be the SRO supervising the fuel movement. The operator on console is responsible to give permission to remove and insert fuel, to log fuel movements in console logbook, and to enter data into the fuel element record book. The key must be in the console and turned to the "ON" position for the chart recorders to operate.

11. Coordinating with operator on console, use the fuel handling tool to withdraw first element from the core and place it in measuring holder. The person moving the fuel should loudly and clearly state " COMING OUT" and " GOING IN" as they move fuel out of and into the reactor core such that the person on console can hear them over the i'ntercom, and "OPEN" and "CLOSE" to the person who is operating the handle of the fuel handling tool. The operator on console should acknowledge the movement of the fuel into and out of the core before the fuelis moved. The person operating the handle of the fuel handling tool should echo the words "OPEN" and "CLOSE" just prior to performing the Revised: 03 May 96 Page 3 l-l l

i operation. I i

12. Disconnect handling tool from element.

1 i

13. Place extension rod on trifiute as before. I

\

14. Read length measurement on micrometer and communicate the measurement '

j to the operatorin the control room. .i l 15. Control room operator enters data on record sheet for element being measured  !

and compares to previous years. A measurement of less than 5/1000 from tho'

~

l- initial measurement listed in the top section of the fuel element record page is l generally accepted without question. For a measurement differential greater

! l i

j than 5/1000 the operator on console should look at the fuel element record to r 4

j. see if there is a trend for that element to be not of the original length orif the  !

i new reading does not follow the historical measurement record. If the current i

i reading is out of line from the historical measurements then the person j measuring the element should be told to "TRY AGAIN". This should continue a

.j until the current measurement is accepted, even if it determines the element be I 4

bad. If the measurement is more than 12/1000 out of line from historical

{ measurements, the Reactor Facility Director should be notified. Growth greater 1

than +100/1000 from initial measurement data is considered defective and the element is not to be placed back into the core. Large length changes require

[ investigation.

4

16. Return the element to core. When placing a fuel element into the core, listen

for the characteristic double click of the element as it seats in the lower grid plate. Double clicks are also produced if the element hits an experiment hole i- in the lower grid plate or, if the element sits in an F-ring position, when the

j. element drops off of the side of the lower grid plate. To prevent hitting the i wrong hole or missing the grid plate, the operator returning the fuel to the core -

l should use care to place their hand directly above the grid plate position. Note that the round structural metal plate above the water has the same radius as j the core. The operator's hand should always be under this plate when seating  !

i j Revised

03 May 96 Page 4 1.

f i

fuel elements in the ocm. Also, the person's hand should be the same distana l from the center of this plate as the fuel element is from the center of the core.

17. Repeat steps 11-16 for each element in core. l

18. Rezaro micrometer (as in steps 8 and 9) after no more than 10 elements. The j operator on console is responsible for keeping track of when to rezero the instrument.  ;

19. Instrumented elements and FFCR's are handled by their extension tubes, not with the handling tool. Move the elements with a minimum of bending of the l extension tubes.

I

20. Check the serial numbers of fuel elements if the elements have gone more than about 8 to 10 years since the last verification. Check an element by resting it l l

at a 45' angle on the lip of the shroud or fuel measuring tool (with the fuel still j attached to the handling tool). j

21. Shine a bright light, if necessary, on the trifiute and observe with binoculars the  ;

number engraved on one side of the trifiute. Some adjustment of the light may l be necessary to provide the necessary shadowing of the engraved numbers.  !

22. The number must agree with the element listed for that core position. If not, try to determine the location where the element may have been moved to by studying the fuel element log. Begin verifying other fuel elements in other possible locations in the core until the error is found. If any of the switched j elements spent any time out of the core during the time they were switched, .

recalculate the Kilowatt-Hour usage on those elements. Notify the RFD of any fuel elements found out of place.

23. The control room operator will annotate in the logbook and fuel element record l which elements' serial numbers were checked by writing the word " Verified".  ;

i (Do not verify all the same elements as last year. Check elements which have j not been checked for several years).

24. Ensure that at least 10% of the elements are checked.

25. Elements in storage will be inspected just as the core elements were checked.

Revised: 03 May 96 Page 5  ;

t

(Do Not Disturb Any Damaged Elements)

26. The serial number must be checked on every element in storage to meet i requirements of Reactor Administrative Procedure A4. '

27. When measuring is completed, remove all equipment / tools from pool area and update TRIGA Tracker.

1 1

)

riprocedagn010 WPO

, Revised: 03 May 96 Page 6 I

I f

j ATTACHMENT B i

10 CFR 50.59 Safety Evaluations of

! Modifications, Changes, and Enhancements

to Procedures or Facilities i

l Procedure 8, Tab A: Logbook Entry Checklist

Procedure S010: Measurement of Fuel Elements / inspection

Replace DAC Hard Drive f Procedure 1

Conduct of Experiments Procedure 0: Procedure Changes Procedure 8, Tab B: Startup Checklist Procedure 8, Tab B1: Safety Checklist ,

Procedure 8, Tab 1: Shutdown Checklist Procedure 8, Tab 1: Shutdown Checklist Procedure 10: Stack Gas Monitor Procedure Procedure COOS: Stack Gas Monitor Calibration Replace FIRE Switch on Reactor Console Change Reactor Security Status in Physical Security Plan Install Drain Line in Cooling Tower Procedure S005: Tank Constant Verification Replacement of SGM Power Supply

FACILITY MODIFICATION

SUMMARY

SHEET 1996 NUM INITIAL TYPE LOCATION PROPOSED CHANGE WS# COMPLETE APPROVAL APPROVAL DATE CHANGE DATE DATE (RFD) DATE (RRFSC) 1 5 Feb 96 Procedure Procedure 8 Tob A Ctenfy Procedure end Add Detaile on Logbook Entries 2 8 Feb 96 6 Feb 96 18 Mar 96 2 6 Feb 96 Procedure Procedure S010 Add Detail to Sections of the Procedure 2 11 Jun 96 6 Feb 96 11 Jun 96 3 26 Feb 96 Facihty DAC Repisco DAC Hard Onve 2 26 Feb 96 26 Feb 96 18 Mar 96 4 4 Mer 96 Procedure Procedure 1 Add New Sections to Procedure 2 22 Mar 96 6 Mer 96 18 Mer 96 5 16 Apr 96 Procedure Procedure O Rewrite of Entire Procedure 2 11 Jun 96 16 Apr 96 11 Jun 96 6 16 Apr 96 Procedure Procedure 8 Tab B Menor Changes to Procedure 2 11 Jun 96 17 May 96 11 Jun 96 7 16 Apr 96 Procedure Procedure 8 Tab B1 Mmor Changes to Procedure 2 11 Jun 96 17 May 98 11 Jun 96 8 16 Apr 96 Procedure Procedure 8 Teb I Mmor Changes to Procedure 2 11 Dec 96 6 Jun 96 11 Dec 96 9 6 Jun 96 Procedure Procedure 8. Tab I Add Line Checking Aurihery Console Chart Recorders 2 11 Dec 96 6 Jun 96 11 Dec 96 10 17 May 96 Procedure Procedure 10 Rewrite Procedure to reflect new SGM Electronics 2 11 Jun 96 17 May 96 11 Jun 96 11 2 May 96 Procedure Procedure COO 6 Cahbration Procedure rewritten to reflect new SGM 2 11 Jun 96 17 May 96 11 Jun 96 Electronics 12 31 May 96 Facihty Reactor Console Replace " Fire

• Switch on Reactor Consolo 2 11 Jun 96 31 May 96 11 Jun 96 13 11 Jun 96 Both Security P!m Change AFRRt Security Status 1 11 Jun 96 11 Jun 96 11 Jun 96 14 9 Aug 96 Facility Cochng System Insteil Drain Line in Secondary Loop Cooling Tower 2 16 Oct 96 10 Aug 96 16 Oct 96 15 9 Sep 96 Ptocedure Procedure SOOS Informal Tank Constant Determination Formalized in 2 17 Oct 96 17 Oct 96 11 Dec 96 New Procedure 16 20 Jun 96 Facility SGM Repiece Internel SGM Power Supply with New 2 12 Sep 96 12 Sep 96 11 Dec 96 External Unit

/

Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: Procedure 8 Tab A  ;

Clarify Procedure and Add Details on Logbook Entries l

Modification to: Procedure _XX._ Facility Experiment Submitted by: George Date __5 Feb 96

1. Description of change:  !

Bef:re-the-fact record is better defined in section 1.

Section 5.a.(4) is expanded to specify that the word SCRAM will be used when the reactor is scrummed.

Section 5.a.(7) is rewritter to clarify how to log replacement operators on console.

S:ction 5.b.(2)(e) is added to specify that the CET Removal and insertion are to be red entries.

S:ction 6 is expanded to show a sample logbook entry for a typical day with the correct ink color specified in the right column beside the entry.  !

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 issue as defined in 10 CFR 50.59 (a) (2) . NO CHANGE

3. If change involves a facility modification, attach a drawing if appropriate. If structural f acility drawings need updating, forward a copy of changes necessary to Facilities. NONE

4. Determine what other procedures, logs, or training material may be affected and record below. NONE

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

NONE

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 RequiredIXX_

Reviewed and approved by RFD Mk e Date b kb RRFSC Notified e /#L Date UAQ lQ 1996

, ,p- .

J

mv "" EWE 3GiiWXs~AIS 6~f5EXiibiEiTERb2H5Di&E"?5PiW7??nns?".aumm*a"ndaawwa m~a.mosamam-%ua%wagews .usmaa wa sammaamsma LOGBOOK ENTRY CHECKLIST I. Operational Logbook

1. The reactor operations logbook is a "befoie-the-fact" record, that is, entries will be logged whenever possible before the operator actually performs the operation. Events, such as scrams, which may not be planned ahead of time, will be entered at the time of occurrence. Any late entries will be so noted.

Entries about what you plan to do are not necessary, only actual events need to be logged in the logbook.

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 completion; 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 responsible for ensuring that the ROS is notified that the logbook is ready for RFD review.

4. Allitems in GREEN (see below) that are not c'osed 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: Most Operational Activities.

(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 SCR AM. This entry to be added when the operational stamp does not contain the SCR AM word in the last line.The K-Excess stamp does not specify SC'R AM time. Also any time the operator deviates from normal stamps for multiple power changes for short periods of time, the final line should be SCR AM. Console locked does notfulfillthis requirement even though the reactorscrams.

Revised: 31 Jan 96 Page1

(5) Mode of operations. Use appropriate stamp or entry to designate the operation:

(a) Steady State.

(b) Square Wave (c) Pulse (6) Operation of reactor associated facilities such as lead shield doors, pneumatic tube systems, etc., unless such operations cause a change of reactivity (see 5.b.(2) below).

(7) Change of personnel at the console. When a change of operator is noted in the logbook, the name of the person replacing the person on console should be entered first as "on console" before the person logging off of the console is entered as "off".

(8) The operator in charge will be designated in the logbook whenever multiple 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 the log for the day.

(11) Designation of the SRO on-call and physicist in charge (PIC).

(12) Reactor calibrations and data.

(13) All changes to logbook entries (including line outs, error corrections, changes to operations mode stamp lines, and end-of-page line outs) will be initialed or signed by the operator.

b. RED: For Items Which Cnange or Measure Reactivity (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 .: ore.

(e) Removal or insertion of CET into core.

c. GREEN: For Maintenance or Malfunctions (1) Any reactor malfunctions noted upon discovery / occurrence with a second entry noting corrective action has been completed.

Revised: 31 Jan 96 Page 2

(2) Additionalitems entered at the disemtion of the operator such as addition of make-up water to the reactor pool, etc.

l (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.

(4) Movement of detectors or chambers from above core,

6. When an operation requiring entry into the logbook falls under more than one j color code, the color to be used will be determined via the following order of 1 precedence: RED - GREEN - BLACK / BLUE-BLACK. I Sample Logbook Entries for a Typical Operational Day I SRO/PIC/HP Stamp, Date Stamp BLACK i

Startup Checklist Begun BLACK '

Console unlocked by "**** BLACK )

1 Startup Checklist # *"* Complete BLACK .

i I

l RAISING RODS TO GO CRITICAL RED j CRITICAL AT WATTS FOR K-EXCESS  !

TRANS $ SHIM $

SAFE $ REG $  ;

CORE POSITION K-EXCESS $

1 M anual Scram BLACK )

l Console locked by *"**** BLACK  !

SGM out of service for maintenance No operations GREEN SGM Back in service GREEN Console unlocked by """*** BLACK l

Opening Pb Doors BLACK Moving Core to "* RED Revised: 31 Jan 96 Page 3 l

RED Moving core to ***

Closing PB Doors BLACK Rabbit (containing ****) inserted into CET RED RAISING RODS TO GO CRITICAL BLACK OC TRANS T1 MAX OC SHIM T2 M AX SAFE RUR#

REG TOTAL KWHRS CRITICAL AT WATTS SCR AM, TOTAL TIME MIN SEC Rabbit removed from CET RED Console locked by ******** BLACK Shutdown Checklist Begun BLACK Shutdown Checklist # **** Complete BLACK Page lined out, Page signed by SRO BLACK II. Malfunction Logbook All entries in the malfunction logbook should include the following information. For consistency, the bold words should be copied into the malfunction log prior to the information.

DATE, TIME, SIGNATURE OF PERSON DISCOVERING MALFUNCTION SYMPTOM:

This section describes how the system is acting or malfunctioning,i.e., channel I went full scale, pump failed, keyboard stopped responding to keystrokes etc.

IMMEDIATE ACTION TAKEN: -

This section is for denoting such things as Reactor Secured, SHD notified.

RFD NOTIFIED:

A remark should be made that the RFD or acting RFD was notified.

~

Revised:31 Jan 96 Page 4 1

DIAGNOSIS : of problem A narrative description of what was dis:: overed to be causing the problem,i.e.,

Which system was malfunctioning or which component failed.

SOLUTION: / repair A narrative description of what was done to correct the problem This could include both physical changes or administrative changes,i.e., a component was replaced and the unit was recalibrated, an additional backup system installed, an administrative prohibit on ... was initiated.

OPERATIONAL VERIFICATION AND/OR CALIBRATION: l I

A description of what actions were taken to verify that the new unit / repair would indeed perform the function for which it was intended,i.e., a calibration signal, system actuated multiple times, system tested, system calibrated with a  ;

source. Indicate whether the change will require staff training. '

SIGNATURE RFD 1

I i

2 Revised: 31 Jan 96 Page 5

b Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: Surveillance Procedun S010 Adding Detail to Sections of the Procedure.

Modification to: Procedure _XX_ Facility Experiment submitted by: George Date 5 Feb 1995

1. Description of change:

An entry was added between Step 1 and 2 to specify the perfonnance of a Shutdown Margin Cdculation.

Step 10 was expanded to specify that an SRO is in charge of fuel movement. Step 10 was also expanded with new sections on proper Logbook entries and Fuel Element Record Entries.

St:ps 11 through 25 were expanded to give more detall on verbiage used between the console cperator and the fuel movement personnel on the nactor deck. Detail is also added to assist cperators in verification of serial numbers.

2. Verify that the proposed change does not involve a change to the l Tschnical Specifications, the facility as described in the SAR, or '

procedures as described in the SAR, and does not produce an unresolved safety issue as defined in 10 CFR 50.59(a) (2) .

No Change l

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 Facilities. MONE

4. Determine what other procedures, logs, or training material may be affected and record below. NONE

5. List of associated drawings, procedures, logs, or other materials to be changed: NONE

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

l Action Sheet: Submitted Not Required _XX_

Reviewed and approved by RFD Date RRFSC Notified Date l

l J

2.-

EI5f E M K 6 M B If 0 if S 2 TS E 3 5 31 Ts E M Ifi?$2!U $ $5fi$ M El MEASUREMENT OF FUEL ELEMENTS / INSPECTION I. General: I

1.

References:

Tech Specs Sections 4.2.5 and 5.2.2e

2. Requirement: All fuel elements, to include FFCRs, shall be inspected for damage and measured for length and bow annually, or after every 500 pulses of $2.00 or larger, whichever occurs first

3. Tools: Fuel handling tool  ;

4. Equipment: Fuel measuring holder, standard fuel element,  !

i

5. Coordination: RFD/ROS l

6. Estimated time: 3-4 days to measure all fuel

7. Safety precautions / protection: Set R1 and R5 at 20 mrem II. Procedural Sequence:

1. Determine total power & number of pulses on each element since last fuel measurement and update each fuel record sheet by adding to previous total.

Be careful to account for any fuel moveraents during the year that will affect the total.

2. Perform a shutdown margin calculation. Unloading 6 F-ring elements and 2 E-ring elements should provide an adequate SDM. The shutdown margin calculation should be checked by the RFD or ROS.

3. Remove all chambers & experiments from vicinity of core

4. Place 3 large work platform plates on rails to cover pool l

5. Place fuel measuring holder in pool by hanging on inside of shroud just to  !

left of opening; secure in place with C-clamp

6. Using fuel handling tool, place standard element in holder and aJlow to come to temperature equilibrium in reactor pool.

7. Turn off primary system pumps to minimize ripples in pool (optional)

8. Place notched end of fuel holder extension rod. en the top of standard rod; place other end under micrometer extension rod (photo A)

1 1

9. Adjust micrometer to zero

10. While measuring fuel, a licensed operator must be on console and under the .

supervision of an SRO. The operator on console may be the SRO supervising the fuel movement. The operator on console is responsible to give permission to remove & insert fuel, log fuel movements in console logbook, and enter data into fuel element record book. The key must be in ,

the console and turned to the on position for the chart recorders to operate.

Log Book Entries: (In red ink)  !

The operator should enter the time that the element was removed from the l core and a statement such as the following in the logbook: Element l

1. " number" removed from " grid position" measured and replaced. Write

" Verified" if the serial number was visually verified.

Fuel Element Record Entries:

A. Stamp the date in the date column. l B. Enter the current length read off of the measuring device. The measurements are recorded in thousandths.

C. Determine the difference between the initial measured length listed in the top section of the sheet and the current length and enter the difference in the growth column.

D. If the element goes into the measuring toolit passes the bow test and should be noted with a check mark.

E. If the element fails either of the two previous tests the element should not be placed back in the core and the RFD should be notified.

F. Record the temperature of the pool water in the temp column.

G. Record the current core grid position in the core position column.

H. Enter the current number of pulses for that element under the pulses  ;

column. If the element was removed during the last year it may have j missed some pulses. Take this into account when calculating total pulses.

I. Enter the total core power in Megawatt Hours under the total power column. If the element was removed during the last year take into account any power the element may have missed during its time out of core.

J. Verify the serial number of any element that has gone over 8-10 years since its last verification. Write " Verified" in the right column if the serial number I was visually verified.

1

'1

l l

11. Coordinating with operator on console, use fuel handling tool to withdraw ~

first element from core & place in measuring holder. The fuel should be kept under enough water to prevent the RAMS from alarming. The person l moving the fuel should loudly and clearly state " COMING OUT" and i " GOING IN" as they move fuelinto and out of the reactor core such that the person on console can hear them over the intercom and "OPEN" and "CLOSE" to the person who is operating the handle of the fuel handling j tool. The person operating the handle of the fuel handling tool should echo l the words "OPEN" and "CLOSE" as they are doing so.

12. Disconnect handling tool

13. Place extension rod on trifiute as before

14. Read length measurement on micrometer and communicate the measurement to the operator in the control room.  !

)

15. Control room operator enters data on record sheet for element bemg  ;

measured & compares to previous years. A measurement ofless than 1 5/1000 from the initial measurement listed in the top section of the fuel I element record page is generally accepted without question. For a-  ;

measurement greater than 5/1000 the operator on console should look at i the fuel element record to see if there is a trend for that element to be not of the originallength or if the new reading does not follow the historical ,

measurement record. If the current reading is out ofline from the historical  !

measurements then the person measuring the element should be told to

. TRY AG AIN". This should continue until the current measurement is accepted. If the measurement is more than 12/1000 out ofline fro.m l historical measurements, the Reactor Facility Director should be notified.

G rowth greater than +100/1000 from initial measurement data is considered defective and the element is not to be placed back into the core.

16. Large length changes require investigation.

17. Returr. the element to core. When placing a fuel element into the core, listen for the double click of the element as it seats in the lower grid plate.

Double clicks are also produced if the element hits an experiment hole in the lower grid plate or,if the element sits in an F-ring position, the element drops off of the side of the lower grid plate . To prevent hitting the wrong hole or missing the grid plate, the operator returning the fuel to the core should use caution to place their hand directly above th'e grid plate position.

Note that the round structural metal plate above the water has the same l

radius as the core. The operators hand should always be under this plate when seating fuel elements in the core. Also the persons hand should be a i

I.

i commensurate distance from the center of this plate as are the fuel elements in the rings of the core grid plate.

{

! 18. Repeat steps 11-17 for each element in core

]. 19. Rezero micrometer after approximately 10 or fewer elements. The operator

, on console is responsible to keep track of when to rezero the instrument.

j 20. Instrumented elements & FFCRs are handled by their extension tubes, not

] with handling tool. Move the elements with a minimum of bending of the i extension tubes.

21. Check the serial numbers of fuel elements if the elements are more than about 8 to 10 years since the last verification. Check an element by resting it at a 45 angle on the lip of the shroud or fuel measuring tool (with the fuel still attached to the handling tool) 6

22. Shine a bright light if necessary on the trifiute & observe with binoculars  :

the number engraved on one side of the triflute. Some adjustment of the

light may be necessary to provide the necessary shadowing of the engraved numbers.

j 23. The number must agree with the element listed for that core position. If

] not, try to determine where the element may have been moved to by i studying the fuel element log. Begin verifying other fuel elements in other

possible locations in the core until the error is found. If any of the switched

j. elements spent any time out of the core during the time they were switched, recalculated the amount of Kilowatt-hour usage on those elements.

24. The control room operator will annotate on the logbook & fuel element l

record what elements were checked for number (do not use the same elements as last year. Check the older elements)  !

l 25. Ensure that at least 10% of the elements are checked

26. Elements in storage will be inspected just as the core elements were checked

27. The serial number must be checked on every element in storage to meet

requirements of Reactor Administrative Procedure A4

28. When measuring is completed, remove all equipment / tools from pool area and update TRIG A Tracker -

)

I I l

O Facility Modification Work Sheet 2 No 10 CFR 50.59' Analysis Required i Proposed Change: _ REPLACE DAC MARD DRIVE Modification to: Procedure Facility _XX_ Experiment Submitted by: GEORGE Date 1 % Feb 96 l

1. Description of change:

R place the DAC Hard Drive (MFM Type 13) with a different model Hard Crive (MFM Type 3). Hard drive is loaded with the current version of DAC C2ftware Vcrification of complete console operations will be performed before the rocctor is returned to fully operational status.

1 j 2. Verify that the proposed change does not involve a change to the Tschnical Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved asfety issue as defined in 10 CFR 50.59(a) (2) .

MFM is an industry standard technology. Both drives have the same p0rformance characteristics. Both interface using the same hardware.

3. If change involves a f acility modification, attach a drawing if cppropriate. If structural facility drawings need updating, forward a copy of changes necessary to Facilities.

{ NONE No drawings necessary 1

4. Determine what other procedures, logs, or traiMng material may be affected and record below.

NONE j 5. List of associated drawings, procedures, logs, or other materials to b3 changed:

NONE

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 quired iXX_

Reviewed and approved by RFD s L n Date '2_/2 </9 c RRFSC Notifled f 274 ff wf > gg'n t Date 18 MAR 1996

_ g' '

2

. y-- l Facility Modification Worksheet 2 l

No 10 CFR 50.59 Analysis Required j Proposed Change: Procedure Change- Operational Procedure 1 Modification to: Procedure X Facility _ Experiment _

. Submitted by: .Iohn T. Nguyen & SFC Cohlli Date: 04 Mar 96

1. Description of change:

i 3 A . sentence,"Any irradiation of matedals shall be in compliance with the l i

limitations in the Technical Specifications, section 3.6. to prevent damage to the reactor or excessive release of radioactive materials", is added into section II.1.

! New sections, section 11.2. Limitations on Expedment and section 11.3.

Determination of Experimental Worths, am added into the Operational l Procedure 1.

2. The proposed change does not involve a change to the Technical Specifications, 4

the facility as described in the SAR, or procedures as de, scribed in the SAR, and

does not produce an unresolved safety issue 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. Not Required  !

4. Determine what other procedures, logs, or training material may be affected i and record below. see #5 l

5. List of associated drawings, procedures, logs, or other materials to be changed: l see #1, Description of change -

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 X i Reviewed and approved by RFD [ Date h j RRFSC Notified yu,

/M M / [ fw.L_

~ -

wic Date ,

k.8 MAR 1900

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UPEE TION $yOC$Ddh5L;e3 .

• "I' ' >

- ,em mme um .- d $ Nocede d 1 E

, , m CONDUCT OF EXPERIMENTS  !

I. GENERAL ,

1. All experiments will be observed during irradiation with the exception of CET l

, experiments or those in which no movement is possible. The closed-circuit  !

televisions (CCTVs)in the exposure rooms and over the reactor pool can be  :

used to meet this requirement.

i 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 . I 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 determine that no unauthorized materials are irradiated.

5. ALARA will be practiced during all experiments.

II. 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. RUR's are not required for reactor parameter measurements as outlined in the Technical Specifications, section' 6.4.2.c. Any experiment j performed by the reactor staff (except under 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 i either the Reactor Facility Director (RFD) or the Reactor Operations I Supervisor (ROS). Any irradiation of materials shall be in compliance with the limitations in the Technical Specifications, section 3.6. to prevent damage to i the reactor or excessive release of radioactive materials.

2. Limitations on Experiment: The following limitations shall apply to the irradiation of materials:

Revised: 4 Mar % Page1

l

a. If the possibility exists that a release of radioactive gases or aerosols may l occur, the amount and type of materialirradiated shall be limited to l assure the yearly compliance with Table 2, Appendix B, of 10 CFR 20,  !

assuming that 100% of the gases or aerosols escape.

b. The totalinventory ofiodine isotopes 131 through 135 in the experiment is not greater than 1.3 curies and the maximum strontium-90 inventory is  :

not greater than 5 millicuries.

c. Known explosive material shall not be irradiated in the reactor in quantities greater than 25 milligrams. The pressure produced in the  :

experiment container upon detonation of explosive shall have been i determined experimentally, or by calculations, to be less than the design pressure of the container.

d. Samples shall be doubly contained when release of the contained material  ;

could cause corrosion of the experimental facility. j

e. The sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00.

3. Determination of Experimental Worths: The experimental worth shall be measured as follows:  !

a. The movement of necessary fuel elements is required before conducting i the K-excess if one of the following limitations would be exceeded: l (1) The absolute worth of materials,' determined experimentally or by calculations, would exceed $3.00. l (2) The maximum available excess reactivity above cold critical with or  ;

without all experiments in place would exceed $5.00.  ;

b. Perform K-excess without the experimental materials / setup. l

c. Perform the K-excess with the experimental materials / setup.  !

d. The difference between K-excess performed in part b and c is the worth of  !

experiment, i.e. K-excess (b) - K-excess (c) = worth. l

4. Experiment Review (Processing of RURs): l

a. Check the RUR for completeness (Section I should be filled out).  !

b. Forward the RUR to the Radiation Sciences Department, Dosimetry Division (RSD D)if dosimetry support is required. .

c. Forward the RUR to the Safety & Health Department (SHD) for radiological safety coordination. j

d. Check experimental protocol against reactor authorizations.

f p_  !

Revised: 4 Mar % Page 2 ~j

J

e. Fill-in Section 4 of RUR with specialinstructions, as appropriate. Assign {

an RUR sequence number. Write in estimated or measured experiment i worth and the core position of the experiment facility to be utilized in the appropriate blocks (lower part of form).

{

f. Have the RFD, acting RFD, ROS or acting ROS review and sign the {

fo m .

l

' g. Ensure the RUR form is placed in the reactor control room prior to the irradiation.  !

5. Conduct of Experiments. Perform setup and irradiation of experiments in accordance with the following procedures:  !

L

a. Exposure Room Entry - TAB A.

1

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.

{

l 6. Complete the RUR by filling out Section 5 with the appropriate information. l

7. Attach form to clipboard in the control room.

j i i I

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i e

4 l

4 1

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s 1

i 4

f Revised: 4 Mar 96 Page 3

{

D  !

Facility Modification Work sheet 2 No 10 CFR 50.59 1==1vsis maauired i

Proposed Change: Modifv Procedure 0 ;" Procedure Phanaes" to includa  :

In=tructinn= for writ 4na a new nrocedure and to l reflect channe to current word nrocessor format.

i Modification to: Procedure X Facility Experiment f Submitted by: na John c ventura Date: Anril 16, 1996 '

1. Description of change:

f Modify crocedure to include detailed instructions on how to write and staff a new nrocedure. Also include f instructions for new word crocessor format.

These ch=naes renresent assentially a rewrite of the entire i nrocedure, hence. no =mrain linas to indicate ch= nae are  ;

included. '

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 l safety issue as defined in 10 CFR 50.59 (a) (2) . I Verified I

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 Facilities.

None required

4. Determine what other procedures, logs, or training material may be affected and record below.

As each nrocedure recuires modification for content, it will be converted to Wordperfect format in accordance with this crocedure. l 1

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

be changed i

None at nresent, current format is still accentable l

6. Create an Action Sheet containing the list of associated work,  ;

specified above, attach a copy, and submit it to the RFD.  !

l Action Sheet: Submitted Not Required X Reviewed and approved by RFD Date !b RRFSC Notified Date b U _

&- +

)

Procedure Changes 1

GENERAL This establishes procedures for

1. Introducing new procedures.

2. Permanently or temporarily changing reactor operational or administrative procedures.

SPECIFIC

l. Introducing a new procedure l

1. Using the current reactor word processor, generate a rough draft of the new procedure. The following should be addressed:  ;

a. Title and Number of Procedure  :

i. The title should be a short description of what the procedure is.

ii. The number has two parts, (1) the procedure type:

(a) Administrative (b) Operational (c) Maintenance, etc 1

(2) the next sequential number for that type of procedure in 1 the TRIGA Tracker

b. General:

i. Reference - Other documents that provide relevant information.

ii. Reauirement - What is the requirement for this procedure? -

Revised: 16 APR 96 Page 1 l

iii. Tools / Equipment - List of needed tools and equipment.

iv. Coordination - Personnel other than Reactor Staff

v. Staff Requirements - Reactor Personnel requirements vi. Safety Precautions - Self-explanatory

c. Procedural Sequence

i. This section should clearly explain each step of the procedure that will be followed.

d. Associated Tables, Graphs, or Diagrams

i. This section should have blank copies of any tables which must be filled out for records, and samples of any graphs which must be generated.

ii. Any diagrams depicting how to setup equipment should also be included in this section.

2. Use the current version of the reactor word processor to open the file called: n\ forms \procedur.frm

3. Modify the top two blocks and the footer of this file in accordance with instructions found in the file. ,

4. Transfer the text of the new procedure (using " Paste") where indicated in "procedur.frm" (The standard font for procedures is Arial 12 point.)

5. Save the procedure as " number.xxx" where "xxx" represents the current word processor.  ;

for example, procedure M049 created in Wordperfect 6 would be named: "M049.wp6" l l

6. Print the procedure.

7. Open the file called n\ forms \procedur.dra

8. Modify the file as below:

a. Modify the top block to the appropriate type and numbbr of Revised: 16 APR 96 Page 2 l

procedure. For example:

i MAINTENANCE PROCEDURE M049 i y b. Add comments as desired below the signature block.

9. Save as appropriate.

.l 10. Print out the draft staffing sheet

11. This step may be performed to generate the 50.59 worksheet for this procedure change. The 50.59 may also be done manually.

a. Open the file called r:\ forms \ form 50.59

b. Fill out 50.59 worksheet, print, and save as appropriate

12. Submit procedure to staff for comments with draft cover sheet,50.59 worksheet, and table of contents, if needed, included in packet I

note: The Senior Reactor Staff Member should be the last to review the document.

If applicable, staff in accordance with para 11.7, (i.e. Health Physics issues or cases involving building ,

4 modifications.)

13. Make appropriate corrections and re-staff with a new draft cover sheet.

4

14. Repeat steps 12 and 13 until there are no further changes.

15. Open the file called r:\ forms \procedur.cov.

Repeat step 8a.

Print out final cover sheet.

16. Submit to RPO for approvalif required.

Submit to RFD for approval.

17. Staff final version for signatums.

I

18. Submit to RRFSC for review

19. Submit electronic copies of the procedure to Reactor Management for storage on read only network hard drive.  ;

i Revised: 16 APR 96 Page 3 I

i i

20. Submit procedure to TRIGA Tracker manager for table of contents update. ,

21. After table of contents tas ba3n updated, place procedure in TRIGA  !

Tracker book.

11. Changing a procedure 4

1. Permanent changes are made by revising the entire procedure. The i revised procedures will be approved by the Reactor Facility Director l (RFD) and reviewed by the Reactor and Radiation Facility Safety l Committee (RRFSC).  ;

a. For procedures in "Ventura Publisher" format, ,

i. Load procedure into Ventura Publisher. i The procedure should be located in r:\proced\) l ii. Save procedure as an ASCll file.

l iii. Load procedure into current word processor.

iv. Make appropriate changes and manus lly add margin lines ,

for changes. i

v. Follow steps 4 - 21 from Section I above.

b. For procedures already in proper format,

i. Open the procedure file. (Located in r:\procedwp\)

ii. Modify the footer to reflect current date. '

iii. Make appropriate changes, add margin lines for changes, and save document as appropriate.

iv. Follow steps 8 - 21 from Section I above.

2. Temporary changes will be documented on a separate sheet attached to the current procedure and implemented when initialed by the RFD, acting RFD, or Reactor Operations Supervisor (ROS) in accordance with Technical Specifications 6.3.2. These changes must be 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 and changes (temporary or pern.anent) will be staffed as

" drafts" for staff comments, and staffed again after being approved by the RFD. The final version will have a signature block for all operators and Revised: 16 APR 96 Page 4

l 1

I reactor staff members. Operators will review new or modified procedures ,

and sign the signature block prior to performing a new or modified I procedure. When the block is completed, the procedure will be placed in  ;

the Reactor Procedures Binder and kept available for review in the control l room.

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 effectiveness of the procedure performance,

c. The change will not decrease the ability of the procedure to l perform its intended function. '

7. Procedures will be staffed under the following guidelines:

i

a. Procertures that may affect areas outside the reactor area, such as building changes, security, etc., will be staffed to the

, appropriate office (s) before further routing.

b. Procedural changes that deal specifically with health physics procedures or radiation safety issues will be staffed through the AFRRI Radiation Protection Officer (RPO)  ;

c. All changes will be staffed to the AFRRl TRIGA Reactor Facility staff I d. All changes will be reviewed by the Reactor and Radiation Facility i

Safety Committee (RRFSC) 9 I i

4 b

Revised: 16 APR 96 Page 5 4

f S,~7, 2 Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: Chances to startuo, safety startuo, and Shutdown checklists Modification to: Procedure _XJ_ Facility Experiment

'ubmitted by:

S Georce / Alt Date 29 April 1996

1. Description of change:

Daily shutdown Checklist Lines added after Part VI. #8 High Level Alarm sottings.

Daily Startup Checklist: Wording on Part V. Line 6 changed from CGM quality assurance checked to Channel test completed on SGM.

A line was added to Part VI.8 (d) SGM Chart recorder operating and tracing.

Part I.1 range was changed to (80-120 psig)

Part V.2 range was changed to (8.5-11 psig)

Daily safety Checklist Wording on Part IV. Line 6 changed from 1 SGM quality assurance checked to Channel test completed on SGM.

A line was added to Part VI.s (d) SGM Chart recorder operating and tracing.

Part I.1 range was changed to (80-120 psig)

Part IV.2 range was changed to (8.5-11 psig)

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 issue as defined in 10 CFR 50.59(a) (2) .

NONE

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 Facilities.

NONE

4. Determine what other procedures, logs, or training material may be affected and record below.

NONE

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

NONE

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 (_JJ__

Rsviewed and approved by RFD DI Datell ib RRFSC Notified

, emu.2, g Date // Vb 7

l l

l Of ' k b D (Sk I $ I/a h ' % Sk .M , jh th;. . y ' . .,- ' ' 4.[. ' .. [l

, Ic _

h  ! .h 4

4 DAILY OPERATIONAL SHUTDOWN CHECKLIST Checklist No. Date Time Completed Supervised by Assisted by l I. REACTOR ROOh!(Room 3161) l

1. All rod drives DOWN . . . . . .... ..... ........... . . . . . . . .

i

2. Carriage lights OFF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3. Door 3162 SECURED . . . . . ....... ... ........ . . . . . . . . .

4. Channel test completed on both CAhi's . . . . . . . . . . . .. ... ..... .... .

5. Door 3161 locked with key . . . ........ . ... .. .. . ... . . . . .

II. EQUIPhfENT ROOM (Room 3152)

1. Distillation unit discharge valve CLOSED . ..., ....... ..... ........

2. Air dryer OPERATIONAL ... . .. .... . . .. . . ..... ....

3. Doors 231,231 A, and roof hatch SECURED . ........ .. ..... .....

III. EQUIPMENT ROOM (Room 2158)

1. Primary discharge pressure (15 - 25 psig) .... .. . ... . . . . . . .

l

2. Demineralizer flow rates set to (5.5 - 6.5 gpm) .. .. . . . . . .

3. Visualinspection for leaks ... ...... . . . . . .

4. Door 2158 and 2164 SECURED . . . . . . . . . . . .

IV. PREPARATION AREA

1. ER2 plug door CONTROL LOCKED . . . .. . .

Door closed; and handwheel PADLOCKED . . ... . . .

2. ER21ights ON and rheostat at 10% . ...... . .. . . .

3. ERI plug door CONTROL LOCKED .. .. ... .. .. . .

Door closed; and handwheel PADLOCKED .. .. . . . .

4. ERl lights ON and rheostat at 10% '

5. Visualinspection of area .. .. .... ... .. .... . . . .

6. Warm storage doors closed . . . .... ......... . . .. . . . . .

1 AFRRI Form 62(R) Revised: 01 May 96 Page1 I

V. LOBBY ALARM {

1 i Lobby alarm audio ON' ............. ................. .............  !

1 l VI. REACTOR CONTROL ROOM (Room 3160) i  !

4 .

t

1. Reactor tank lights OFF ......................... ................. l

2. Console chart recorder pens raised ......................... ........  !

3. Steady-state timer OFF ................................... ....... l l 4. Console LOCKED, and all required keys returned to lack box .., .........  !

! 5. Diffuser pumps OFF ..... ....... ..............................

l j 6. Purification, secondary and primary pumps ON . . . . . . . . . . . . . . . . . . . . . . . . j i 7. Reactor monthly usage summary completed ........ ............ . .. l

8. Radiation monitors .... ... ..... .. ................... .. ...  !

i i, i l MONITOR READING HIGH LEVEL ALARM SETTING )

(mrem /hr) i a.R-1

• 20 )

I' (<20)

b. R-2 (<10) 10 ,

c. R-3 i

d. R-5

(<10) 10 20

)

(<20)

c. E (<10) 10

f. E-6 (<10) 10

g. R-6 (<10) 10

• Numerical Entry t . w..;+ -

AFRRI Form 62(R) Revised: 01 May 96 Page 2

en n +,a: c, >;,,~m. .

-.::::9,s v a n DAILY OPERATIONAL STARTUP CHECKLIST Checklist No. Date Time Completed Supervised by 1 Assisted by I. EQUIPMENT ROOM (Room 3152)

• l

1. Air compressor pressure (80 - 120 psig) ... ... ....... .... . ....

2. Water drained from air compressor ...... ........... ....... . .

l 3. Air dryer operating ..... . ............... ... .. . .... . .

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

II. LOBBY AREA Lobby alarm turned off . . . . . .. ......... . . ... . .....

III. EQUIPMENT ROOM (Room 2158)

1. Prefilter differential pressure (< 8 psid) . .. ... . . .. .

l

2. Primary discharge pressure (15 - 25 psig) ....... . .. .. ... ..

3. Demineralizer flow rates set to 6 gpm (5.5 - 6.5 gpm) . ... . .

4. Stack roughing filter (notify supervisor if> 1.0" of water) ... .. .. ..

5. Stack absolute filter (notify supervisor if > 1.35" of water) . ......

6. Visualinspection of area ... .. ... ..... . ... . ...

7. Door 2158 SECURED .. ... . ... .. .. .. . ..

IV. PREPARATION AREA 1

Visualinspection of area .. . . ... . . . ...

i V. REACTOR ROOM (Room 3161)

1. Transient rod air pressure (78 - 82 psig) .. . ..

2. Shield door beving air pressure (8.5 - 11 psig) . .. . .

l

3. Visualinspection of core and tank . . . .

Fuel elements *

4. Number of fuel elements and control rods in tank storage Control rods

5. Air particulate monitor (CAM)

(a) Primary operating and tracing . . . . . . . ..

(b) Backup operating . . . . . . . . . .

(c) Channel test completed, damper closure verified . . i

6. Channel test completed on SGM . . . . . . ... . . . .. . .. ..

l

7. Door 3162 SECURED . . .. . ... . . .. . . ..

i a - n.m swu

• Numerical Entry AFRRI Form 62a (R) Revised: 01 May 96 Page1

VI. REACTOR CONTROL ROOM (Room 3160)

1. Emergency air dampers reset .......... .... .. ... ...... . ... 1

2. Console recorders dated .. . ... . .. ...... ...... . . .

3. Stack flow and fuel temperature recorders dated . . . . . . . ... ...

4. Logbook dated and reviewed . ..

5. Water monitor box (a) Background activity (10 - 60 cpm) . . ... .. . . . . .

(b) Water monitor box resistivity (> 0.2 Mohm-cm) . .. .. .. .

' (c) DM1 resistivity (> 0.5 Mohm-cm) . .... .... .. . .

(d) DM2 resistivity (> 0.5 Mohm-cm) .. . ...... .. .

6. Stack gas flow rate (15 - 35 Kcfm) . .. . . . . . .

7. Stack linear flow rate (1.0 - 2.0 KfVmin) .... . . . . .

8. Gas stack monitor (a) Background (2 - 20 cpm) .. .. . . . .. .

(b) Alarm check .. . .. . . . . . . .

(c) High alarm set to 3.2 E-5 microCi/cc at stack top . . .

(d) SGM chart recorder operating and tracing .

9. Radiation monitors l Monitor Alann Point Reading Alarm Setting Functional (mrem /hr) (mrem /hr)

(a) R-1 (< 20)

• 500 (b) R-2 (< 10)

• 10 (c) R-3 (< 10)

• 10 (d) R-5 (< 20)

• 100 (e) E-3 (< 10)

• 10 (f) E-6 (< 10)

• 10

10. TV monitors on ... .

I1. CAM high level audible alarm check .. . . .

12. Water temperature (inlet)(5 - 35 *C) . *

13. Water level log completed .

, 14. Console lamp test completed .

15. Time delay operative .

16. Source level power greater / equal to 0.5 eps .

17. Prestart operability checks performed . .

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

19. 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 I

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

20. Zero power pulse ... . ... .. . .. . . . . . . . ..

t en a w.

• Numerical Entry AFRRI Form 62a (R) Revised: 01 May 96 Page 2

1 1

n :. <. d +a;.i, a s ,m . .;; y ;.c ;a Giq y .

DAILY SAFETY CHECKLIST Checklist No. Date Time Completed Supervised by Assisted by  ;

?

I. EQUIPMENT ROOM (Room 3152)

1. Air compressor pressure (80 - 120 psig) ........ ...... .... ...... .

2. Water drained from air compressor .... ...... ......... .......... t

3. Air dryer operating ....................... ......................

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

l II. EQUIPMENT ROOM (Room 2158) f

1. Prefilter differential pressure (< 8 psid) ... ............ ........ . .

2. Primary discharge pressure (15 - 25 psig) .. ..... . . . .. .. .. ...

3. Demineralizer flow rates set to 6 gpm (5.5 - 6.5 gpm) . . . . . . . . . . . ..... ,

4. Stack roughing filter (notify supervisor if > 1.0" of water) ... ...... . ..

5. Stack absolute filter (notify supervisor if > 1.35" of water) . . . . . . . . . . . . .

6. Visualinspection of area .......... .. .. .. ........ . ...... ,

7. Door 2158 SECURED ..... . ........... . . ...... ... .....

III. PREPARATION AREA Visualinspection of area ... . . .. .. .... . . .. . .. ... .

IV. REACTOR ROOM (Room 3161)

1. Transient rod air pressure (78 - 82 psig) . . ..... . . ... .....

2. Shield door bearing air pressure (8.5 - 11 psig) .. .... ... ... . . .

l 3. Visualinspection of core and tank . .. . . .. ..... . ..

l 4. Number of fuel elements and Fuel elements 1

• control rods in tank storage Control rods

5. Air particulate monitor (CAM)

(a) Primary operating and tracing . .. . . ..... . . . . -l .  ;

(b) Backup operating .. ....... . ....... . . .....

(c) Channel test completed, damper closure verified .. ... . .

i

6. Channel test completed on SGM . . . . ... .... .. . . .. ..... l

7. Door 3162 SECURED ........ ... . ...... .. . ... ....... .

L t mmei m

• Numerical Entry AFRRI Form 62b (R) Revised: 26 APR 96 Page1  !

. - = . __ - - .

V. LOBBY AREA  ;

1 Lobby audio alarm turned off . . .. . . .. .. . . . . . . . . .. . .. .... . ... . .. . . . . .... .. .. .. . . . . ..... .. . . . . . .

VI. REACTOR CONTROL ROOM (Room 3160)

. 1. Emergency air dampers reset ................... ........... .....

2. Console recorders dated . ....

i 4

3. Stack flow and fuel temperature recorders dated . . . . . ........ ... ...

4. Logbook dated and reviewed ...... ... ... .... .. . . . . . . ....

5. Water monitor box (a) Background activity (10 - 60 cpm) .......... .. .. . .. .

(b) Water monitor box resistivity (> 0.2 Mohm-cm) *

(c) DMi resistivity (> 0.5 Mohm-cm) . .. .. ..

)i (d) DM2 resistivity (> 0.5 Mohm-cm) .. . ... .. . . .

I

{ 6. Stack gas flow rate (15 - 35 Kcfm) ... ... ... ... . . ... ..... .

• I

7. Stack linear flow rate (1.0 - 2.0 KfVmin) . ... .. . . ,. ..

, 8. Gas stack monitor

(a) Background (2 - 20 cpm) *

)

(b) Alarm check . . ... .. .... .. ... . ..

(c) High alarm set to 3.2 E-5 microCi/cc at stack top ... .

(d) SGM chart recorder operating and tracing . . . . . .

9. Radiation monitors [

Monitor Alarm Point Reading Alarm Setting  :

Functional (mrem /hr) (mrem /hr)

(a) R-1

• 20 l

(< 20)

(b) R-2 *

(< 10) .

10 l (c) R-3 (< 10) 10  ;

(d) R-5 (< 20) 20 l (c) E-3 *

(< 10) 10 (f) E-6 (< 10) 10

10. TV monitors on .

I1. CAM high level audible alarm check . . .

12. Water temperature (inlet)(5 - 35 *C) ... ..

13. Water level log completed . .. .. . . .

14. Source level power greater / equal to 0.5 cps . .. .

a_ m uasi =

• Numerical Entnf AFRRI Form 62b (R) Revised: 26 APR 96 Page 2

O i Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Prnposed Change: Add a line to shutdown checklist Modification to: Procedure _ZK_ Facility Experiment Submitted by: Georce Date 6 Jun 96

1. Description of change:

Add the line to section VI of the shutdown checklist

s. Auxiliary chart recorders operating and tracing

2. Verify that the proposed change does not involve a change to the Tschnical Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved esfsty issue as defined in 10 CFR 50.59(a) (2) .

NONE

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 Facilities.

NONE

4. Determine what other procedures, logs, or training material may be offected and record below.

NONE

5. List of associated drawings, procedures, logs, or other materials to ke ch.inged:

NONE

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 _XX_

0 e Rsviewed and approved by RFD < Date b i$

RRFSC Notified Date U

DD7 D 3iE id H H G @ '

l

DAILY OPERATIONAL SHUTDOWN CHECKLIST i 1

I

! I I

Checklist No. Date Time Completed Supervised by l' Assisted by i

I. REACTOR ROOM (Room 3161) i

1. All rod drives DOWN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .  ;

2. Carriage lights OFF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

3. Door 3162 SECURED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4. Channel test completed on both CAM's . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

5. Door 3161 locked with key . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

II. EQUIPMENT ROOM (Room 3152)

1. Distillation unit discharge valve CLOSED . .......... ..............

2. Air dryer OPERATIONAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3. Doors 231,231 A, and roof hatch SECURED ... ......... .. ........ I 1

III. EQUIPMENT ROOM (Room 2158)

1. Primary discharge pressure (15 - 25 psig) . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. Demineralizer flow rates set to (5.5 - 6.5 gpm) . . . . . . . . . . . . . . . . . . . . . . . . j

3. Visual inspection for leaks . . . . . . . . . , . . ..... ........ .. .... .

4. Door 2158 and 2164 SECURED . . .......... ... ... ..... .......

IV. PREPARATION AREA

1. ER2 plug door CONTROL LOCV.ED .. .. ............ ... .....

I Door closed; and handwheel FADLOCKED .. ............ ..........

2. ER2 lights ON and rhecstat at 10% .. ............................... l

3. ERI plug door COi4 TROL LOCKED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . j Door closed; and handwheel PADLOCKED .. .... ................. .

4. - ERI lights ON and rheostat at 10% . . , . . . . . . . . . . . . . . . . . . . . . . . . . . .

5. Visualinspection of area .............. .... . .... .......... .....

6. Warm storage doors closed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,

l I

I Revised: 06 June 96 - Page1 AFRRI Form 62(R)

V. LOBBY ALARM Lobby alarm audio ON ....................... .. ... ...............

VI. REACTOR CONTROL ROOM (Room 3160)

1. Reactor tank lights OFF .. .. .. .......................... ..

2. Console chart recorder pens raised . . ........... .... ............

3. Steady-state timer OFF . . . . . . .. .................... .. ..

4. Console LOCKED, and all requ' ired keys returned to lock box ........ ...

5. Diffuser pumps OFF ... .. .. .... ................... .........

6. Purification, secondary and primary pumps ON . . . . . . . . . . . . . . . . . . . . . . .

7. Reactor monthly usage summary completed ..... . ...... ..........

8. Auxiliary chart recorders operating and tracing .. .. ...................

9. Radiation monitors ..... .. ...............................

MONITOR READING HIGH LEVEL ALARM SETTING (mrem /hr) a.R-1 (<20) 20

b. R-2

• 10

(<10)

c. R-3 (<10) 10 d.R-5

• 20

(<20)

c. E-3 (<10) 10

f. E-6 (<10) 10

g. R-6 (<10) 10

• Numerical Entry av ,w  ;+

AFRRI Form 62(R) Revised: 06 June 96 Page 2

/0 Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Prcposed Change: Rewrite Operational Procedure 10 for the Daily Test of the New BGM Electronics Modification to: Procedure _XX_ Facility Experiment Submitted by: Georce Date _ 17 May 96 _

1. Description of change:

Prscedure was rewritten to coincide with the new analog electronics vs.

tho old digital electronics. The new electronics perform the same function however the method of testing them has changed slightly from one olectronics package to another.

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 cafety issue as defined in 10 CFR 50.59(a) (2) . l NONE ,

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 Facilities.

NONE

4. Determine what other procedures, logs, or training material may be affected and record below.

AJ OE- -womme= essex.

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

NONE

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 III_

Reviewed and approved by RFD Date /7 kyl RRFSC Notified Date // el f'd

-.- a ,

1

.. I STACK GAS MONITOR PROCEDURE l GENERAL l

This procedure specifies how to test the Stack Gas Monitor (SGM) to ensure proper operation of this monitoring device. This instrument is used to sample and measure the gaseous effluent Ar* in the building exhaust system.

SPECIFIC

1. DAILY START-UP (operational and safety)

A. Check the air sampling flow rate (should be between 3.5 and 3.9 cubic feet per minute with the chamber door closed). Change the particulate pre-filter if necessary. Adjust as necessary.

B. Press the RED, HIGH ALARM and BLUE, ALERT ALARM setting buttons on the front of the unit and verify the alarm point settings. (Limits +0 / -10%)

High Alarm Point Low Alarm Point C. Channel Test the SGM. Flip and hold the OP/ TEST Switch to the test position and verify that the analog meter reads 3600 ( range 3000-4200). If outside the range of 3200 - 3800, notify the ROS. Reset Alarms.

D. Remove the front cover of the detector shield and slowly insert the check source into the chamber. Verify that the alert lamp and alarm lamps illuminate at their designated alarm points and that the audible alarm sounds at the high alarm point.

E. Remove the source from chamber and press the alarm and meter reset buttons on the front of the unit.

R Wp 10 wpo Revised: 17 May 96 Page1

. //

! Facility Modification Work Sheet 2 i

No 10 CFR 50.59 Analysis Required Proposed Change: stack Gas Monitor Calibration Procedure C006 -

Chanced to Coincide With New SGM Electronics Modification to: Procedure A Facility Experiment i Submitted by: Georce Date 02 May 1996

1. Description of change:

The procedure C006 was rewritten to coincide with the new electronics unit for the SGM. The new procedure was adapted from the manufacturers calibration literature and clarified to allow for ease of procedure use. ,

2. Verify that the proposed change does not involve a change to the Tcchnical Specifications, the facility as described in the SAR, or procedures as described in the SAR, and does not produce an unresolved esfety issue as defined in 10 CFR 50.59(a) (2) .

. NONE

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 Facilities.

NONE

4. Determine what other procedures, logs, or training material may be effected and record below.

NONE

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

NONE

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 _XX_

Rnviewed and approved by RFD  % r Date h u,acQ(

RRFSC Notified Dat // f[

! sos 9c006.2wp g /

(*MY.IG11 #$!$ @ #MiB i[h% % %% M iPdy M Es STACK GAS MONITOR CALIBRATION

1. General:

1.

Reference:

Tech Specs 4.5; HPP 7.3; NMC Stack Monitor Electronic Test and Calibration Procedure P/N 0001020-1.

i

2. Requirement: The air particulate monitoring system (SGM) shall be calibrated annually, not to exceed 15 months. (HPP 7.3)

3. Tools: Crescent wrench, screwdriver, special calibration connectors.

4. Equipment

Oscilloscope w/ leads, voltmeter, pulse generator, pulse counter, i plastic Ar-41 sample beaker w/ tubing provided by SHD i 5. Coordination: With SHD to set a date for isotopic calibration and with the ROS/RFD to arrange a date on operations schedule with no other reactor operations.

l

6. Estimated time: One day l

7. Safety precautions: Use caution when working around high voltage sources and minimize exposure to Ar-41 or Na-22 calibration sources.

8. General:

Turn off high voltage and main power before plugging or unplugging any of the

. circuit boards.

The " unit" refers to the rack mountable electronics section of the SGM.

ll. Procedural Sequence:

l. Schedule date for calibration with RFD/ROS and SHD.

t

2. Assemble required tools and equipment

3. Produce Argon-41 (for argon calibration) with reactor.

Revised: 02 May 96 Page1

1 i i i

i

[ Suggest: 2 syringes, 50-60 cc P-10 gas irradiated for 5 min at 100 Kw. ,

ELECTRONIC CALIBRATION  :

I l- 4. Turn off the power. Adjust the front panel meter to 10 cpm.  !

i  !

! I

5. Remove the CRA-14B/91 card and ensure switches SW1, SW2, and SW3 are ,

l open.  !

6. . Remove the IC-13 card and set the dip switches into the following configuration

l 1' . S1 Closed S5 Open (10% Window)  !

S2 Open S6 Closed (20% Window)  ;

i S3 Open S7 N/C l I S4 Open (5% Window) S8 N/C j t i 2

i

{ Gross counting mode S1 Closed, S2 Open, S3 Open.}

{

l { Spectrometer mode S1 Open, S2 Open, S3 Closed.}  ;

1  ;

! Replace the IC-13 Card.

Place the card extension card into the CRA-14B/91 slot and attach the CRA- .

i 14B/91 card to the extension card. - j i i j ' 7. Disconnect the detector cable from the back of the SGM unit and attach the test - I i cable to the SGM unit. i

! Power up the unit. l l

l 8. Verify 24

• 4 VDC between pins 19 and 20 on the terminal block inside the back i of the unit. If the voltage is outside this range, replace or repair the power supply.

t I- 9. Connect a pulse generator to the detector test cable.

1 j 10. Attach a test cable from the jack on the face of the AA-13A/91 plastic face mask j i to a volt meter. The access hole is just below the red high alarm button. l l

11 Set the pulse generator to create 16,666 cps (1x10' cpm).

l

12. Adjust R33 on the CRA-14B/91 card to give -5.00 i 0.01 VDC.

j. 13. Set the pulse generator to create 166.6 cps (1x10d cpm).

14. Adjust R32 such that the analog meter reads 10,000 cpm. l i

i i-i Revised

02 May 96 Page 2 j  !

d

• i

i. 15. Set the pulse generator to create 16.66 cps (1000 cpm).

16. Verify 1000 cpm on the local analog meter.

17.. Adjust the potentiometer on the 0-1 mA card such that the remote analog meter

. in the reactor control room reads 1000 cpm.

t C

_18. Adjust the potentiometer on the 0-10 VDC card to give 1000 cpm on the remote chart recorder in the control room.

1

19. Monitor the voltage through the AA-13A/91 mask test jack, step through the following inputs, and verify the following outputs.

l Pulse Generator Voltage @ AA-13A/91 .

i 10 cpm 0.00

• 0.15 VDC j 100 cpm 1.00 t 0.15 VDC  ;

j 1000 cpm 2.00 i 0.15 VDC 10,000 cpm 3.00 t 0.15 VDC j 100,000 cpm 4.00

• 0.15 VDC j 1,000,000 cpm 5.00 t 0.15 VDC i

{ 20. Adjust the potentiometer, located above the yellow fail button, while pressing the

{ yellow fail button such that the analog meter reads about 12-15 cpm. Set the pulse j generator to 10 cpm. Verify that the fail alarm lamp illuminates when the analog

meter needle drops below 12-15 cpm.

i L 21. Press the meter reset and alarm reset buttons. .j i

! 22. Press the alert and high alarm buttons to note the settings. Increase the count j output of the pulse generator to cross each of these alarm points and verify that the j respective lamps on top of the stack gas monitor illuminate at their set points and  ;

that the sonalert alarms at the high alarm point. 1 i

F 23. Power off the unit. Remove the test cables from the front and back of the unit.

! Replace the CRA-14B/91 card back into its slot. Attach the detector cable to the

{ back of the unit.

1

[ 24. Remove the IC-13 card and set the dip switches into the following j_ configuration for spectrometer mode:

,. i i

Revised

02 May 96 Page 3 i

s i-

i i  :

S 1 l S1 Open S5 Open (10% ' Window) 1 j S2 Open S6 Closed  ;

(20% Window) 1 S3 Closed S7 N/C I j S4 Open (5% Window) S8 N/C Replace the iC-13 card.

l

25. Power the unit back on. Turn on the high voltage. Press the meter reset button.

{ 26. Insert the Sodium-22 source slowly into the chamber and verify operability of j the unit.

i

! 27. Determine the proper high voltage and adjust as necessary to find the peak counts for the argon / sodium peak.

A. This is done with the sodium source or a sample of argon in the detector i chamber, j B. Slowly adjust the voltage while monitoring the voltage through the face plate

! of the high voltage card. NOTE: the voltage can only be measured through the l face plate. Measurement of the voltage any other place will provide an inaccurate reading.

C. Set the voltage such that the maximum counts are read from the analog

meter. The voltage should be approximately 558 volts. Graph if necessary.

i

28. Assist SHD, as needed, in the isotopic calibration using HPP 7-3.

29. After SHD provides the new alarm point numbers, adjust the alarm points. I A. The high alarm point is adjusted by pressing the high alarm button and '!

adjusting the potentiometer located directly above the button to give the proper i

. alarm point reading on the analog meter.

B. The alert alarm point is adjusted by pressing the alert alarm button and i adjusting the potentiometer located directly above the button to give the proper j alarm point reading on the analog meter.

30. See that a new calibration sticker is placed on the SGM. Change the written alarm points at the appropriate locations (At SGM and Control Room Meter).

31. Obtain and file isotopic calibration report required by HPP 7-3 from SHD.

32. Create decay curve for Sodium-22 source to be used for semiannual source '

test.

33. Update TRIGA Tracker.

Revised: 02 May 96 Page 4

i l

MEMORANDUM TO FILES l, Re: Calibration of SGM ,

The SGM was calibrated on . The results are as follows.

P Step Point Expected As Left

8. Voltage TB19-20 24

• 4 VDC .

l

12. AA-13A/91 Jack -5.00

• 0.01 VDC i

17. Control Room Meter 1000 l

18. Control Room Chart 1000 'i i

19. Inject Signals l Pulse Generator Voltage AA-13A/91 Meter Reading  ;

O cpm (0 cps) V cpm j 100 cpm (1.666 cps)- V cpm '

1000 cpm (16.66 cps) V cpm i 10,000 cpm (166.6 cps) V cpm  :

100,000 cpm (1,666 cps) V cpm i 1,00,000 cpm (16,666 eps) V cpm Fail Lamp Operates YES/NO  ;

Warning Lamp Operates YES/NO High Lamp Operates YES/NO l Audible Alarm Operates YES/NO ,

High Voltage Set Point VDC Alert Alarm Set Point cpm High Alarm Set Point epm Counts Generated From Sodium-22 Source cpm ,.

Revised: 02 May 96 .Psige 5

FINDING THE ARGON-41 PEAK A source, either Argon-41 or Sodium-22, was placed in the detector chamber and the voltage on the detector was increased from approximately 500 Volts to 800 Volts. The output of the SGM electronics was graphed. The graph produced is the mirror image of the spectrum of the source used. The reason the graphed image is a mirror image of the source spectrum is because the new SGM has a pre-set window in the electronics and as the voltage is increased on the detector the spectrum is pushed through the window with the highest energy end of the spectrum being pushed through the window first.

~

The Argon-41 and Sodium-22 peaks were determined to be at 558 Volts.

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1. Insert source fully into SGM chamber and record the reading from the meter.

3

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ET-9272, REV. O I

Page 1 of 12 3

1 l

. i 1

l i, t-ELECTRONIC TEST AND CAllBRATION PROCEDURE t

for the '

CRM-51MS/91 2

COUNTING RATE METER I l P/N 0001020-1 i

v i

i ,

I 4

Nuclear Measurernents Corporation .

t 2460 North Arlington Avenue j indianapolis, IN 46218 l (317) 546-2415 ,

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, . E I

i ET-9272, REV. O  !

Page 2 of 12 .

RECORD OF REVISIONS i No Date ECO # Pages Prepared By Approvals [

i 0 03-06-96 - - . all D L Hewlett. M';

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. l ET-9272, REV. O Page 3 of 12 1.0 CHARACTERISTICS OF TEST All tests in this procedure shall be performed in an indoor environment at ambient temperature and humidity.

An electronic setup calibration and performance examination is necessary to demonstrate the electronic functional characteristics of the counting rate meter.

This procedure uses a certified pulse gerierator to test the circuitry of the module.

1.1 Equipment needed:

1.1.1 Certified digital voltmeter (DMM) and high voltage probe i 1.1.2 Certified oscilloscope 1.1.3 Certified test pulse generator, National Institute of Standards and Technology (NIST) traceable (NMC PG 21, PG-22, PG-81 or equivalent).

1.1.4 Small flat blade screwdriver 2.0 ACCEPTANCE AND REJECTION CRITERIA A tolerance value is specified for all test values in this procedure. Measurements or adjustments within this tolerance shall constitute an acceptable test result.

If a measurement is outside the tolerance band and is not correctable by adjustment, the circuit shall be rejected as unacceptable. The rejection of a plug in circuit module shall not be cause for rejecting the counting rate meter or an acceptable circuit module therein. A rejected circuit card may be repaired or replaced to continue testing.

3.0 SET-UP PROCEDURE:

~"

3.1 Before connecting power to the counting rate meter ensure that the following conditions are met:

a) . Master switch is off.

b) H.V. switch is off. -

c) OP-TEST switch is set to OP.

d) Meter on front panel is mechanically zeroed.

e) On the CRA-14B/91 card, set SW1 to be open, set SW2 to be open, and l set SW3 to be open.

1 1

ET-9272, REV. O Page 4 of 12 f) If equipment external to the counting rate meter is wired to the counting rate meter, ensure that the equipment is in good working order and is wired correctly.

g)

If extemal 24 VDC power is not required (pin 19 on terminal block) for the AA-13A/91, install a jumper between AA-13A/91 pin B and CRA-14B/91 pin B to supply the AA-13A/91 board with 24 VDC.

h) Configure the AA-13A/91 and CRA-148/91 cards as follows:

i AA 13A/91 '

J1 A-C U2 Delay 200 mS J2 I A-B U2 Delay 200 mS J3 AC U3 Delay 200 mS J4 AB U3 Delay 200 mS

{

J5 AC U4 Delay 1.6 S J6 A-C U4 Delay 1.6 S l

J7 A-C HIGH Alarm Latching J8 A-C ALERT Alarm Latching

.? A-C Fall Alarm latching J10 AB K1 (HIGH) Energized on Alarm J11 A-B K2 (ALERT) Energized on Alarm CRA-14B/91 ma J1 B;C 0-1 mA through Pin R J1 E-F 0-10 V through Pin S J2 DGND 5 Decade Module (10-10M cpm)

S1 0013201 0-1 mA Card Installed S2 00132'03-1

=r 0-10 VDC Card Installed 3.2 Connect the pulse generator (use negative pulses) to pins F(input) and A(gnd) of the counting rate mete.r input connector (Circular 8 pin MS connector). Do not turn on the pulse generator.

3.3 Connect a 1000 i1% chm resistor between terminal block pins 15 and 16 or connect the 0-1 mADC recorder to these output connections.

3.4 Connect 117 VAC power to the counting rate meter.

ET-9272, REV. O Page 5 of 12 3.5 Tum the master switch on.

a) The "ON" lamp should be illuminated.

b) The meter should show no needle deflection.

c) Confirm the following voltage at the terminal block: i

1. Between pins 19 and 20 24 2 4 VDC 3.6 The input Control card IC-13, incorporates several switches to set the counting mode and window width for the spectrometer mode "N". While testing, NMC i recommends setting the counting mode to "D1", (S1 closed, S2 & S3 open). This is the " gross counting" mode where all pulses above a fixed level are counted.

This is the easiest to set up. The spectrometer mode (pulse height counting) requires switch S3 to be closed (S1 & S2 open) and one of the three switches (S4, SS, or S6) closed. S4 is the 5% window , S5 is the 10% window and S6 is i the 20% window widths. l l

4.0 COUNTING RATE METER CAllBRATION PROCEDURE: I 1

" NOTE" If the circuit cards have been previously calibrated, skip to section 4.4. Then, if the counting rate meter fails section 4.4, attempt recalibration beginning with section 4.1. If the counting rate meter again fails section 4.4, effect a repair or replacement and re test.

" NOTE" All voltages in the remainder of this document are referenced to ground (pin 20 on the terminal block).

4.1 CRA-14B/91 circuit board

l a) Set the pulse rate on the pulse generator to 1x105 cpm. Depress the alarm reset pushbutton. Remove the front cover from the counting rate meter.

Monitor the Test Jack on the AA-13A/91 mask. Adjust R33 until the voltage is -5.00 i 0.01 VDC. * % % W M l-.- Lu b) Set the pulse rate on the pulse generator to 10,000 cpm. Adjust R32 I

such that the meter reads 10,000, cpm. O m ,y ,y C. sed 4.2 0 -1 mA circuit board:

a) Set the pulse generator to 1x105 cpm Adjust the potentiometer on the 0-1 mA card until the current through pin 15 of the terminal block is 1.00 i 0.03 mA. " /* " " ' o 4.3 0 -10 VDC circuit board:

i ET-9272, REV. O I Page 6 of 12 !

a) Set the pulse generator to 1x10' cpm. Adjust the potentiometer on the 0-10 i

i VDC card until the voltage at pin 14 of the terminal block is 10.00

• 0.30 VDC.  ;

4.4 i Monitor the current through pin 15(+), the voltage at pin 14(+), and the volta!

AA-13A/91 mask test jack, as the pulse generator is stepped through the ra inputs listed below. Confirm the appropriate current and voltage for each rsi Pulse generator Current pin 15(+) i Voltage pin 14(+) Voltage @ AA 13A/91 10 cpm 0.00:0.03 mA 0.00,0.30 VDC 0.00:0A5 VDC 100 cpm 0.2020.03 mA 2.0010.30 VDC

{

1000 cpm 1.00:0.15 VDC 0.40 0.03 mA 4.00 0.30 VDC i

10,000 cpm 0.60,0.03 mA 2.00$0.15 VDC 4 6.00e0.30 VDC 3.00.0.15 VDC

' ~

100,000 cpm 0.80:0.03 mA 8.0020.30 VDC 4.00:0,15 VDC 1,000,000 cpm 1.00:0.03 rnA I

10.00:0.30. VDC 5.0020.15 VDC i 5.0 - 1 COUNTING RATE METER FUNCTIONAllTY VERIFICATION:

5.1 On the alarm card, depress the high alarm setpoint pushbutton and adjust the potentiometer such that the meter indicates 100,000 cpm for the setpoint. In a l

similar manner, set the alert setpoint at 10,000 cpm, and set the fail setpoint at 100 cpm. Turn on the pulse generator and set the count rate to 1000 cpm. Then, depress the alarm reset button.  ;

a) Attach an Annunciator or equivalent across terminal strip pins 19(+) & 18(-).

b) NO ALARM STATE: Confirm the following conditions at the terminal block: j

1. Pin 10(+) 0 1 VDC -

j i

2. Pin 11(+) O e 1 VDC ' i

)- 3. Pin 12(+) 0 e 1 VDC

4. Annunciator off

5. Pins 1 and 2 open-circuit j 6. Pins 2 and 3 shorted .

7. Pins 4 and 5 open-circuit , .

i

8. Pins 5 and 6 shorted

9. Pins 7 and 8 shorted

10. Pins 8 and 9 open-circuit -

! c) Fall ALARM STATE: Monitor pin 12(+), and set the pulse gpnerator to 50 i

~ cpm. As the meter needle passes through 100 cpm (thereby entering fail alarm condition), verify that the voltage at pin 12(+) changes from 0 i 1  ;

p VDC to 24 t VDC. Then, confirm the following:

]

{  !

1

- ., - .--- m n - -e - -- .,-- n ,. e w

ET-9272, REV. O Page 7 of 12 l

1. Pin 12(+) 24 +/-4 VDC l

2. Pins 7 and 8 open-circuit

3. Pins 8 and 9 shorted l d) ALERT Al. ARM STATE: Monitor pin 11(+), and set the pulse generator to l 50,000 cpm (entering alert alarm condition). As the meter needle passes r through 10,000 cpm, verify that the voltage at pin 11(+) changes from 0 *1 .

VDC to 24 t 4 VDC. Then, confirm the following

1. Pin 10(+) 24

• 4 VDC  ;

2. Annunciator off ,

3. Pins 4 and 5 shorted

4. Pins 5 and 6 open-circuit e) HIGH ALARM STATE: Monitor pin 10(+), and set the pulse generator to  ;

200,000 cpm (entering high alarm condition). As the meter needle passes through 100,000 cpm, verify that the voltage at pin 10(+) changes from 0

• 1 VDC to 24 2 4 VDC. Then, confirm the following:

1

1. Pin 10(+) 24 i 4 VDC

2. Annunciator on

3. Pins 1 and 2 shorted

4. Pins 2 and 3 open circuit While stillin high alarm condition, depress the alarm reset button. Confirm the following:

5. Pin 10(+) 24 4 VDC *

6. Annunciator off

7. Pins 1 and 2 shorted

8. Pins 2 and 3 open-circuit f) NO ALARM STATE: Set the pulse generator back to 1000 cpm, and -

depress the alarm reset button. Confirm the following: .- r

1. Pin 10(+) 0 +/-1 VDC -

2. Annunciator off -

5.2 Place the input Mode Switch in the " TEST" position. Verify that the meter ,

indicates between 3000 to 4000 cpm. Place the Input Mode Switch back in the "OP" position.

4 l

l ET-9272, REV. O Page 8 of 12 5.3 Connect the high impedance probe with a DMM to pins B(+) and A(-) of the  !

counting rate meter input connector. Place the H.V. switch in the "ON" position, and verify that the voltage can be varied between 300 VDC and 1300 VDC. Set the high voltage to 500 VDC. Place the H.V. switch back in the "OFF" position.

5.4 Depress the Meter Reset button, and confirm that the meter indicates 10 cpm.

5.5 Tum power off. Remove the 1000 ohm resistor and the Annun::iator previously attached at the counting rate meter terminal block. This concludes the test.

5.6 Set the input Control card IC-13, to the desired counting mode. Review section ,

3.6 to determine the mode for this application. Denote in the data sheet the finat setting.

6.0 DATA SHEET SIGN-OFF '

Sign and date the data sheet. Have Engineering and Quality Assurance review the data sheet. Make copies for the customer and give the original data sheets to Quality Assurance.

I I

r k

l

! i F

1 i

ET-9272, REV. 0 )

Page 9 of 12 1 TEST DATA SHEET CRM-51MS/91  ;

Customer Date l Customer P.O. NMC S.O.

This Test Data Sheet contains the results of testing and calibration, and verifies that Nuclear M asurements Corporation (NMC) has successfully tested the subject equipment to all the conditions required by tne Test Procedure, with the following results: ,

Range 10 - 1,000,000 cpm Module serial No. -1020 1.1 TEST EQUIPMENT 1.1.1 DMM Model/ Serial No.:

Cal. Due: ,

1.1.2 Pulse Gen Model/ Serial No.: l Cal. Due:

3.0 SET-UP PROCEDURE 15 6 Description Required As Left [

"ON" Lamp On (Chk)

Meter Deflection None (Chk)

. Between pins 19 & 20 20 to 28 VDC VDC f

i COUNTING RATE METER CAllBRATION:

. t 4.1 CRA.14B/91 PC assembly  !

Description Required As Left  !

Adjust R33 / AA.13A/91 TJ -4.99 to .5.01 VDC i

ET-9272, REV. O Page 10 of 12 l

Adjust R32 for 1000 cpm (Chk) j 4.2 0-1 mA PC assembly (Location S1 on CRA Card)

Desenption Required As left t Adjust pot on PC / pin 15 of TB 0.97 to 1.03 mA mA 4.3 0-10 VDC PC assembly (Locaten S2 on CRA Card) k Description - Required As Left  !

Adjust pot on PC / pin 14 of TB 9.70 to 10.30 VDC VDC 1

4.4 Monitor Current 0 pin 15 (+) & Voltage O pin 14 (+)

Puita - lo Required lo As Left Vo Required Vo As Left Generator pin 15 (+) pin 14 (+)

10 cpm -0.03 to 0.03 mA mA -0.30 to 0.30 VDC VDC 100 cpm .17 to .23 mA mA 1.70 to 2.30 VDC VDC 1000 cpm .37 to .43 mA mA 3.70 to 4.30 VDC VDC 10,000 cpm .57 to .63 mA mA 5.70 to 6.30 VDC VOC 100,000 cpm .77 to .83 mA mA 7.70 to 8.30 VDC VDC 1,000,000 cpm 0.97 te 1.03 mA mA 9.70 to 10.30 VDC VDC

~

l 4.4 Monitor Voltage @ AA-13A/91 TJ

]

l Pulse Vo Required Vo As Left I t

Generator Test Jack I l

~

10 cpm -0.15 to 0.15 VDC VDC 100 cpm .85 to 1.15 VDC VDC 1000 cpm 1.85 to 2.15 VDC Vs0

. 10,000 cpm 2.85 to 3.15 VDC VDC 100,000 cpm 3.85 to 4.15 VDC VDC '

1,000,000 cpm 4.85 to 5.15 VDC VDC

c..

ET-9272, REV. O Page 11 of 12 5.0 COUNTING RATE METER FUNCTIONALITY VERIFICATION:

5.1.b) NO ALARM STATE: Confirm the following conditions at the terminal block:

Description Expected Actual Check if OK

(/)

1. Pin 10(+) 0 21 VDC

2. Pin 11(+) 0 e 1 VDC

3. Pin 12(+) 0 e 1 VDC

4. Annunciator off

5. Pins 1 and 2 open-circuit

6. Pins 2 and 3 shorted

7. Pins 4 and 5 open-circuit

8. Pins 5 and 6 shorted

9. Pins 7 and 8 shorted

10. Pins 8 and 9 open-circuit 5.1.c, d, & e) ALARM STATE: Confirm the following conditions at the terminal block:

SECTION Description Expected Actuai i Check if OK l (/)

5.1.c 1. Pin 12(+) 24 e 4 VDC FAIL ALARM 2. Pins 7 and 8 open-circuit

^

3. Pins 8 and 9 shorted 5.1.d 1. Pin 10(+) 24 e 4 VDC ALERT ALARM 2. Annunciator off AE '

3. Pins 4 and 5 shor1ed l

l 4. Pins 5 and 6 open-circuit

ET-9272, REV. O Page 12 of 12 SECTION Description Expected Actual

{ Check if OK

(/)

5.1.e 1. Pin 10(+) 24 4 4 VDC HIGH ALARM 2. Annunciator on

^

3. Pins 1 and 2 shorted

4. Pins 2 and 3 open-circuit

5. Pin 10(+) 24 2 4 VDC 6 Annunciator off 7 Pins 1 and 2 shorted 8 Pins 2 and 3 open-circuit 5.2 Meter indicates between 3000 to 4000 cpm

(/[

5.3 High Voltage range (300 - 1300 VDC)

(/D High Voltage set @ 500 VDC

(/))

5.4 Meter Reset is operable

(/)

5.6 input Control card settings:

GROSS GROSS SPECTROMETER MODE

'D 1 ' 'D2' S1 S2 S3 "N' S45% SS - 10% S6 - 20%

~^

6.0 Signatures Test Operator Date. _

Engineering Review Date ,

Quality Assurance Date

1

/2 - i i

Facility. Modification Work Sheet 2 l

No 10 CFR 50.59 Analysis Required Proposed Change: . Replace 2 pole " Fire" switch on Reactor Console with a 4 pole switch.

i Modification to: Procedure Facility X Experiment -

Submitted by: George / Ventura- Date: May 31, 1996

1. Description of change.

Replace the old 2 pole pulse fire switch with a 4 pole, but-otherwise identical, switch. Add two additional lines to serve i

~

cxternal uses such as a trigger for a new pulse computer.

Existing wiring replaced to same points on the new switch and additional wiring run to CSC TB5, ,

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 issue as defined in 10 CFR 50.59(a) (2) .

No Technical Specifications or SAR description change. [

Does not produce an unresolved safety issue. i

3. If change involves a facility modification, attach a drawing if  !

appropriate. If structural facility drawings need updating, forward a l copy of changes necessary to Facilities. l Attached: GA manual " Appendix C" change  ;

4. Determine what other procedures, logs, or training material may be ,

affected and record below. l' None

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

Wire Harness updated in GA manual " Appendix C"

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 X

~

0 Reviewed and approved by RFD i b - Date: May 31,1996 RRFSC Notified ,

' Date: 4w fd c/

____-_____.____.__--.__s- - -.---,q -,. , 5*g , ,- - , - -

s t

GA TRICA WIRE HARNESS

/'

UT CSC N/A VIRE DESTINATION I Tefzel SIGNAL i H22-1 l~B w- 3 5 (Oc- T u nus y ,o scra,c l l l H22-2 l- E u -n. i.

I I I H22-3 I I J l H22-4 l l l l H22-5 l

- 1 I I H22-6 I l l l H22-16 i

~ l I i H22-17 i l l I H22-18 I I I I E +12 V de RCC. TB5-12 l l l l H22-22 I CWh I R tv BRD " 5 L E. . , .-c x.,,.,,, orm Li , c )

i I I TBS-10. TBS-24 I l l l H22-23. TBS-22 IMERT ko Ehe*- 1 'c < c s, _...

l l l K3-1, K3-6 l l l l H22-25. TBS-16 I l l I TBS-15 I I I l K3-7 l l i l K3-12 TB5-32 I I I I K3-13. TBS-33 i

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,g i l l K5-6 I a9 I I i K5-5 I s.30 l l l K4-9, K4-1, K5-2 I I l l K4-10, TBS-32 I s.32 l l l TBS-31, TBS-18 l Ens.33 l l l TB5-34 TBS-19 l W rat.34 I I l TBS-33 l Br 5 35 l l l TBS-24 (+12 V de) l E nt.36 I I I K6-7 l-l 'r5-39 . 2 h.

WDS.37 I l l K6-6 lL'- 6 t4- '1' Ens.38 I i I K6-5 I ED5 39  ! l I K7-7 I' 5 56 ~ ' ' ~ '

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Ern3 43 I I l K8-6 lkr 5 E n_t.44 I I I K8-5 I E nt.45 i l l H21-50 I Reactor operate light rtn D .46 l l l E +12 V RCC l Reactor operate light +12 V de h t.47  ! l (92) l H23-24 1 r :-. ic y c rA LooD (RTN)

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EM 50 1 Tefzel (92) i H23-12 1/ K 7 0 JO C*r d

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l Facility Modification Work Sheet 1 10 CFR 50.54(p) Analysis Proposed Change: change AFRRI Security Status from DoD Restricted Area Submitted by: LTc Leonard A. Alt Date 11 JUN 96 l

1. Description of change:

Within this document the term " Restricted Area" refers to the d:cignation of an area for the purposes of security control, and does not c rry the meanings related to radiological contamination control as when tha term is used in a health physics sense. The formal designation of AFRRI as a DoD Restricted Area will be removed, with an anticipated offective date of 1 JUL 96. This will affect the requirement that

" Visitors must be escorted by permanent AFRRI staff members while in the complex...", as currently stated in section 3.1 of the Safety. Analysis Rsport dated 01 NOV 94. Under the new security system, visitors, after appropriate radiological training, will be granted card-key unescorted access to the AFRRI complex with the exception of the reactor area and esveral other critical areas. The reactor area will have coded card-key recders that will not permit unescorted access by visitors.

The Physical Security Plan for the TRIGA Reactor Facility at the ,

Armed Forces Radiobiology Research Institute is referenced as containing ]

procedures for safeguarding and ensuring the accountability of the l cpacial nuclear material by section 3.5 of the Safety Analysis Report. '

Tha Physical Security Plan currently designates the entire AFRRI complex as a restricted area in accordance with the Internal Security Act of 1950. This designation was originally made by the Institute's higher Command when AFRRI was established in the early 1960's. It is now being changed by higher Command to make the designated restricted areas only the referenced reactor facility areas and several other critical areas within the Institute. These areas will be posted as DoD Restricted Arcas. Removal of the formal designation of the entire AFRRI complex as a DoD Restricted Area will change the reference to the Internal Security Act of 1950, but will not change the manner in which the remainder of the Physical Security Plan operates to protect the Reactor Facility Controlled Access Areas or Reactor Controlled Areas.

It should be noted that the Physical Security Plan uses the term

" controlled access area" in two senses. Generally, it refers to the raactor facility areas and other Institute areas where visitor access is still denied without specific access control measures and escort. The term Controlled Access Area is separately defined by paragraph 3.b.(1) of the Plan to represent those reactor facility areas where,the highest degree of access control nust be maintained. This use of the term is consistent with that applied by the U.S. Nuclear Regulatory Commission.

2. Reason for change: This change in the officially designated security status of the Institute reflects the overall U.S. Government trand since the end of the cold war toward greater openness in the

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

i

.l operation of scientific facilities.. -It is expected to promote the Institute's mission by allowing greater access to collaborative

3. researchers.- Since the access control to reactor areas remains

unchanged, it will have no adverse impact on the physical security of the i reactor facility. l

. I l

I 1' 3.- Verify that the proposed change does not involve a change to the Technical Specifications or produce an unresolved safety issue as I

specified' in 10 CFR 50.59 (a) (2) . Attach an analysis to show this.

Analysis attached? Yes Y l

3 4. The proposed modification constitutes a changes in the facility or an i

operational procedure as described in the SAR. Describe which (check all i that apply).

[ Procedure X Facility Experiment 5

l

. The proposed change affects the way visitor access is handled in i areas of the Institute external to the reactor facility. It does not j affect the way visitor access is controlled in reactor areas.

j- 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

} One sentence in section 3.1 of the SAR (third paragraph of the

section) will be changed

i FROM:

j " Visitors must be escorted by permanent AFRRI staff members while in j the complex and persons carrying packages and briefcases are subject to i .. search" i

j TO: j i

" Visitors must be escorted by permanent AFRRI staff members while in j j j the reactor area and persons carrying packages and briefcases are subject l to search"  ;

s' j One sentence in section 3.b of the Physical Security Plan will be ,

j changed:  !

FROM:-

l q "This Plan designates the entire AFRRI complex as a restricted area  ;

j in accordance with reference (a)" l i

3 l t

i

To:

"This Plan designates the entire AFRRI complex as a DoD controlled access area, and the reactor area as a " Restricted Area".

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

Not applicable. l

7. Specify associated information.

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

None.

8. Create an Action Sheet containing a list of associated work specified in items #7, attach a copy, and submit another to the RFD.

Action Sheet: Submitted Not Required X

9. The Reactor Facility Director has determined this change does not decrease the effectiveness of the Physical Security Plan for the AFRRI TRIGA Reactor Facility and may therefore be implemented under 10 CFR 50.54 (p) .

Rsviewed and approved by RFD  % - Date Il Gu T 6 o

e RRFSC Concurrence Date // L%

/'

i

)

ANALYSIS (Per paragraph 3)

The Technical Specifications contain no requirements pertinent to the proposed Institute.

change in the designated physical security status of the There is therefore no change to the Technical Specifications.

The proposed change only affects the way visitor access is controlled in areas of the Institute external to reactor areas. Since the proposed change does not affect the physical security devices protecting the reactor, and does not affect the access control to reactor areas, there is no unresolved safety issue.

s Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: Install a Drain Line In Reactor Coolina Tower.

Modification to: Procedure Facility _XX_ Experiment Submitted by: Georce Date _ 9 Auc 96

1. Description of change:

This change is to install a drain line into the bottom of the reactor cooling tower. The propose of this drain is to allow the water in the cooling tower to be easily drained and replaced with fresh water as doemed necessary by the reactor staff member performing the monthly check of the cooling tower.

During the summer months, water loss due to evaporation is replenished in the cooling tower through a float valve. As the concentration of the minerals in the water increases, they begin to plate out. By draining the water and replacing it with fresh water, the amount of deposition can bs reduced.

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 issue as defined in 10 CFR 50.59 (a) (2) .

SAR does not say anything about a drain line from the cooling tower.

3. If change involves a f acility modification, attach a drawing if appropriate. If structural facility drawings need updating, forward a copy of changes necessary to Facilities.

Drawing of cooling tower will be modified.

4. Determine what other procedures, logs, or training material may be affected and record below.

M015 to specify occasional draining of cooling tower

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

M015, Cooling Tower Drawings.

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 fot Required IXX_

Reviewed and approved by RFD d DatelObdb RRFSC Notified Date b

.~ , y

/

/.6 Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: New Procedure to Perform Tank Constant Determination Modification to: Procedure X Facility Experiment Submitted by: ENS J C Ventura Date September 9, 1996

1. Description of change:

Informal process of determining Tank Constant has been formalized in this procedure.

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 issue as defined in 10 CFR 50.59(a) (2) .

No change or unresolved safety issue

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 Facilities.

No facility modification

4. Determine what other procedures, logs, or training material may be affected and record below.

No other changes

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

None s 6. Create an Action Sheet containing the list of associated work, specified above, attach a copy, and submit it to the RFD.

I Action Sheet: Submitted ,, _

Not Required X

? r

-iewed and approved by RFD .l Dat [bYld RRFSC Notified / [ '

Date [/ be c I(

yy

j

'r 5 g f k - $ h ,5 ' . , I '#.

Tank Constant Verification General

Purpose:

This procedure is used to determine the Tank Constant for the reactor pool and core.

Tools / Equipment: 90 KW heater, heater power console, digital thermometer, binoculars, hand calculator, stopwatch, and clamp-on inductive ammeter.

Reference:

Calculator user's manual Coordination: Ensure no reactor runs are scheduled for the following morning.

Staff Requirements: 2 Reactor staff members for setup.

Safety Precautions: 3 phase power hook-up to the heater. Special care should be taken when working with high voltage.

Cover all electrical connections with non-conductive material to prevent accidental contact with wires or connections.

Procedural Sequence SETUP (the afternoon before the test)

1. Move core to position 700.

2. Retrieve Heater control console from above reactor office spaces.

3. Lower heater into pool, supported by 2 work grates, do not energize at this time.

4. Connect heater to heater control console in accordance with figure 1.

a. Do not close switch on power console Revised: 17 OCT 96 Page1

b. Do not plug console into motor control panel

c. Do not energize motor control pan'.,1 socket

5. Lower stirrer into pool, supported by 2 work grates. Place stirrer as close as reasonable to the heater.

6. Lower temperature probe from digital thermometer into pool near core.

7. Unplug tank lights, turn on stirrer and leave running overnight.

8. Secure Primary, Secondary, and Purificat x system. ,

Tank Constant test

1. Leave stirrer running during test.

2. Energize digital thermometer and verify operation.

3. Verify that heater control switch in not in the "on" position.

4. Plug heater control console into special 175 amp receptacle on motor control panel.

5. Energize electrical panel and close switch on heater control console.

6. Allow Hester to come to steady power and system to equilibrate for 15 minutes.

7. Begin taking temperature readings every 3 minutes, record this data in table 1.

8. Readings may be plotted as you take them either on graph paper or with a ,

computer. Figure 2 is a sample plot of the data from table 2.

9. If plotting points by hand, calculate a linear regression at each point using the hand calculator.

10. Continue to take readings until a steady slope can be determined from the graph or regressions. (The first few data points may need to be ignored until a steady rise is observed. )This will take approximately 30 minutes.

11. This slope is in degrees per minute, multiple by 60 to get degrees per hour.

12. The slope must then be corrected because the heater and stirrIr do not output Revised: 17 OCT 96 Page 2

exactly 100 kW of thermal power. The true value for the power is determined by following the steps below.

a. Measure the current in each of the three legs running to the heater using the clamp-on inductive ammeter, not the hard mounted ammeters.

(The clamp-on ammeter is more accurate.)

b. Record the voltage between each pair of load wires using the mounted voltmeter on the heater control console.

c. The total power is equal to the thermal power from the heater plus the thermal power from the stirrer. Historical calculation shows that the thermal power from the stirrer is approximately 2 kW.

Heater power is determined from:

P = p i V 40 where P = average powerin watts i = average current from ammeter V = average voltage from 3 voltmeters on Heater control console Total power is nominally 97 kW plus 2 kW from stirrer = 99 kW

d. To calculate the tank constant, divide the slope by the average power, then multiply by 100. This will yield:

Where K rc is the Tank Constant dT

- is the slope of the temperature vs time graph dt P is the average power

13. '

Place graph and result for Krc in core physics log.

Revised: 17 OCT 96 Page 3

. . - . _ . - - - . - _ - - - . = - . - . . - _ . . - . . - . - - . . .

Figure 1: Hookup of Heater to Control Console i

From line Ammeter leads not shown OGG OOO r V

/

N/N /N /+-Hardwired i inductive

) ) ) ammeter Breaker box ,

To heater Coad) uom .

1. ' Run 3 leads from power supply in through top 3 holes on Heater control console and into top of breaker box.

2. Run white ground wire from power supply to grounding screw on Heater control console.

3. Run 3 leads from heater (load) through inductive ammeters mounted on Heater control console, through lower 3 holes, into iE'om of breaker box.

4. Run leads from voltmeter through lower 3 holes, into bottom of breaker box.

f Revised: 17 OCT 96 Page 4 ,

Figures and charts were produced using Microsoft Excel 4.0, however any appropriate software package may be used.

Figure 2 is Microsoft Excel file r:\procedwp\S005_f2.x14 Table 1 is Microsoft Excel file r\procedwp\S005_t1.x14 Table 2 is Microsoft Excel file r:\procedwp\S005_t2.xl4 Revised: 17 OCT 96 Page5

TABLE 1 Tank Constant Personnel: Date: l l Stirrer Heater la Va Core Ib Vb Wait time min Ic Ve heater power = KW 1(vg) 0.00 V(avg) 0.00 stirrer power = KW Power Level 100.00 KW slope slope corrected corrected Elapsed PoolTemp deg / min deg/hr slope elapsed slope 0.00000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 0.0000 0.0000 0 0.0000 deg I min slopes obtained by software regression Best Constant Slope = 0.000 s

S005_T1.XL4 9/27/96

EXAMPLE TABLE 2 I

Tank Constant '

Personnel: ENS J C Ventura Date: 16-Aug-95 Robert Georce Stirrer 500 Heater 500 la 116.50 Va 485 Core 750 lb i17.00 Vb 485 Wait time is min Ic 111.60 Vc 486 heater power = 96.70 KW 1(avg) 115.03 V(avg) 485.33 stirrer power = 2.00 KW Power Level 98.70 KW slope slope corrected corrected Elapsed Pool Temp deg / min deg / hr slope elapsed slope 15 25.390 - - - - -

20 25.558 - - - - -

25 25.646 - - - - -

30 25.766 0.02432 1.4592 1.4784 30 1.4784 35 25.900 0.02456 1.4736 1.4930 35 1.4930 40 26.011 0.02429 1.4575 1.4767 40 1.4767 45 26.140 0.02436 1.4614 1.4807 45 1.4807 50 26.249 0.02417 1.4503 1.4694 50 1.4694 56 26.425 0.02450 1.4702 1.4896 56 1.4896 60 26.508 0.02450 1.4701 1.4894 60 1.4894 65 26.623 0.02443 1.4660 1.4853 65 1.4853 70 26.744 0.02438 1.4630 1.4823 70 1.4823 deg / min slopes obtained by software regression Best Constant Slope = 1.482 S005_T2.XL4 9/27/96 EXAMPLE

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Facility Modification Work Sheet 2 No 10 CFR 50.59 Analysis Required Proposed Change: REPLACE STACE nas MONITOR POWER SUPPLY Modification to: Procedure Facility XX Experiment Submitted by: Lessard a Georce Date 20 June 1996

1. Description of change:

Replace the Stack Gas Monitor internal edge card power supply with an external power supply.

The replacement Power Supply voltage will be adjustable just like the existing power supply, but the replacement power supply will be able to supply more current than the existing edge card supply.

The existing power supply appears to be operating at its limit, which is causing stability (drifting) problems. The replacement power supply can provide 100 times more current if necessary. NMC claims that the current provided by the edge card power supply is enough to operate the unit.

Calibration will be performed before the stack Gas Monitor is returned to fully operational status 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 issue as defined in 10 CFR 50.59(a) (2) .

NONE 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 Facilities.

NONE

4. Determine what other procedures, logs, or training material may be affected and record below.

Procedure supply.

C006, Step 27 to be rewritten to reflect external power

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

Manufacturers drawings to be modified as per external supply.

6.

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

Action Sheet: Submitt II Not Required Reviewed and approved b RFD. S Date/

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RRFSC Notified '

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ACTION SHEET FOR REPLACEMENT OF STACK GAS MONITOR POWER SUPPLY Update Procedure C006, SGM Calibration Procedure Install new power supply and calibrate SGM Update manufacturers drawings to reflect new power supply location.

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i Attachment to " Replace Stack Gas Monitor power Supply 50.59" i

Current version of hx:edure C006

27. Determine the proper high voltage and adjust as necessary to find the peak counts for the argon / sodium peak.  !

A. This is done with the sodium source or a sample of argon in the detector chamber i B. Slowly adjust the voltage while monitoring the voltage through the face plate of ,

the high voltage card. NOTE: The voltage can only be measured through the face l plate. Measurement of the voltage in any other place will provide an maccurate reading. .

C. Set the voltage such that the maximum counts are read from the analog meter. I The voltage should be approximately 558 volts. Graph if necessary  ;

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Proposed version of Procedure C006 j l

27. Determine the proper high voltage and adjust as necessary to find the peak counts for

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the argon / sodium peak.

A. This is done with the sodium source or a sample of argon in the detector chamber B. Slowly adjust the voltage. Set the voltage such that the maximum counts are read from the analog meter. Be sure that the peak selected is the Argon 41 (1293 Kev) or the Sodium 22 ((1274 Kev) peak and not the wxiium 22 (511 Kev) peak. graph the output vrs. voltage if necessary to find the proper peak.

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SEP-10-SS TUE 14:50 NUCLEAR EASIREMERTS FAX NO. 3175434420 P.02 IN1FMM- NucAnarInstrumentation September 10, 1996 AFRRI Attn: Rob George Tel:(301)295-1221 Fax:(301)295-0735 l

Rob: i Here is a copy of the board level test procedure that we use in production testing. Most of the time, the board is checked in '

the instrument as part of an instrument calibration and check-out es opposed to a board level test. This procedure is really only used for spare parts situations where we don't have a complete instrument.

As far as a set of specifications for this card, I don't believe we have such a list; it was designed to meet the requirements of a typical NMC detector as part of a complete instrument design.

As you can see from the schematic drawing, this design was done in the early '70's and has been in production ever since. If you believe that you have a problem with this card, I would recommend that you return it to NMC for evaluation and/or repair, especially since it is still under warranty.

Sincerely I

.N- ~

Larry L. Crim Nuclear Measurements Corp' 2460 N. Arhngton Ave.

• P.O. Bos 18248 Phone (317) 5464415 indianapolis. Indiane 462184248 TWA (810) 3414137

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ATTACHMENT C 4

Appointment Letters for Current Reactor and Radiation Facility Safety Committee i

Changes i 1

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i ARMED FORCEO RADIOClOLOGY RECEARCH INSTITUTE

, 8901 WISCONSIN AVENUE

, BETHESDA, MARYLAND 20089-5603 l

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h 605.01 11 June 1996 ,

4 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Acting Chairman of Reactor and Radiation Facility Safety Committee 1

i

The following appointment is made

Captain Charles B. Galley, MSC, USN ACTION: Appointed as Acting Chairman of Reactor and Radiation Facility Safety f Committee i

AUTHORITY: Verbal orders, Director, AFRRI EFFECTIVE: 11 June 1996 I PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: All questions regarding A FRRI Reactor and Radiation Facility Safety Committee should be directed to Captain Charles B. Galley at 295 9261.

FOR THE DIRECTOR:

1 J

1 CURTiS W. PEARSON Colonel, USAF, MSC l Deputy Director for Administration l

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. ARMED FORCED RADIO 310 LOGY RECEARCH INi2TITUTE l

' 8901 WISCONSIN AVENUE BETHESDA, MARYLAND 20889-5603 J

605.01 1 RRFSC 16 October 1996 l l

MEMORANDUM FOR RECORD l

SUBJECT:

Appointment of Members of Reactor and Radiation Facility Safety Committee  !

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Effective this date, the following individuals are members of the Armed Forces Radiobiology Research l Institute (AFRRI) Reactor and Radiation Facility Safety Committee (RRFSC). Memberships are in  ;

cecordance with the Technical Specifications of Nuclear Regulatory Commir.3 ion license R-84. This cppointment memorandum supercedes all previous documer,ts pertaining to RRFSC membership.

PERMANENT MEMBERS Dr. David McKown, AFRRl, Radiation Safety Officer Voting Member .

Stephen I. Miller, AFRRI, Reactor Facility Director Voting Member j APPOINTED MEMBERS C.B. Galley CAPT, MSC, USN, AFRRI, Chairman Voting Member Dr. Marcus Voth, The Pennsylvania State University, Voting Member Reactor Facility Director Mark A. Miller, Naval Research Laboratories, Voting Member j Radiation Safety Officer SPECIAL MEMBERS J.W. Malinoski, CAPT, MSC, USN, AFRRI, Special Voting Member Head, Radiation Sciences Department Edward R. Herbert, Montgomery County Government, Special Non-Voting Member Environmental Protection Department Dr. Leslie McKinney, AFRRI, Special Non-Voting Member Radiation Pathophysiciogy and Toxicology Department RECORDER Danny K. McClung, SFC, USA, AFRRI FOR THE DIRECTOR:

CURTIS W. PEARSON COLONEL, USAF, MSC Deputy Director for Administration DISTRIBUTION:

1-each individual 1-RRFSC file

ARMED FORCEO RADIO'llOLOGY RECEARCH INSTITUTE

, 8901 WISCONSIN AVENUE BETHESDA, MARYLAND 20889-5603 605.01 16 rwrhar 1996 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member The following appointment is made:

Captain Charles B. Galley, MSC, USN ACTION: Appointed as Chairman of Reactor and Radiation Facility Safety Committee.

Colonel David G. Jarrett, MC, USA is replaced by Captain Charles B. Galley, MSC, USN.

AUTHORITY: Verbal orders, Director, AFRRl EFFECTIVE: 16 October 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: This appointment is made in accordance with the AFRRI l Reactor Technical Specifications of NRC License R 84. All questions regarding AFRRI l Reactor and Radiation Facility Safety Committee should be directed to Captain Charles B. I Galley at 295 9261.

FOR THE DIRECTOR:

D ,1 -

dM4 CURTIS W. PEARSON 1 Colonel, USAF, MSC l Deputy Director for Administration l

ARMED FORCEO RADIO 310 LOGY RECEARCH INSTITUTE 8901 WISCONSIN AVENUE eETHESDA, MARYLAND 20889-5603 605.01 16 Odober 1996 MEMORANDUM FOR DISTRIBUTION A 1

SUBJECT:

Reactor and Radiation Facility Safety Committee Member I The following appointment is made:

Mr. Stephen 1. Miller ACTION: Mr. Mark Moore is replaced by Mr. Stephen 1. Miller, Acting RFD, AFRRI AUTHORITY: Verbal orders, Director, AFRRI EFFECTIVE: 16 October 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: Mr. Miller is appointed as RFD (permanent voting member) to the RRFSC. This appointment is made in accordance with the AFRRI Reactor Technical Specifications of NRC License R-84. All questions regarding AFRRI Reactor and Radiation Facility Safety Committee should be directed to Mr. Stephen 1. Miller at 295-1290.

FOR THE DIRECTOR:

I aAC4 CURTlS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration

J ARMED FORCEO RADIO 2iOLOGY RECEAF CH INSTITUTE t

8901 WISCONSIN AVENUE I BETHESDA. MARYLAND 20889-5603 i

605.01 16 Ocabar 1996 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member The following appointment is made:

J Dr. David McKown i

ACTION: Mr. Thomas <rien is replaced by Dr. David McKown, Acting RSO, AFRRI AUTHORITY: Verbal orders, Director, AFRRI l EFFECTIVE: 16 October 1996

' i j PERIOD: Until superseded or rescinded SPECIA' INSTRUCTIONS: Dr. McKown is appointed as RSO (permanent voting member) to the RRFSC. This appointment is made in accordance with the AFRRI Reactor Technical Specifications of NRC License R-84. All questions regarding AFRRI Reactor and Radiation Facility Safety Committee should be directed to Dr. David McKown at 295-1285 FOR THE DIRECTOR:

7 uwa CURTIS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration l

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ARMED FORCED RADIO 2iOLOCY RESEARCH INSTITUTE 8901 WISCONSIN As 'NUE BETHESDA, MARYLAND 2 % 9-5803 605.01 16 October 1996 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member The following appointment is made:

Cap ain James W. Malinoski, MSC, USN ACTION: Captain C.B. Galley is replaced by Captain James W. Malinoski, MSC, USN AUTHORITY: Verbal orders, Director, AFRRI EFFECTIVE: 16 October 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: Captain Malinoskiis appointed as s special voting member to the RRFSC. This appointment is made in accordanca with the AFRRI Reactor Technical Specifications of NRC License R-84. All questions regarding AFRRI Reactor and Radiation Facility Safety Committee should be directed to Captain James W. Malinoski, at 295-1048.

FOR THE DIRECTOR:

n .. .

JAdM CURTIS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration

ARMED FORCEO RADIOEIOLOGY RECEARCH INOTITUTE l 5

8901 WISCONSIN AVENUE '

, BETHESDA, MARYLAND 20889 5603 i

605.01 9 December 1996 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member The following appointment is made:

Colonel Curtis W. Pearson, USAf MSC ACTION: Appointed as Chairman of Reactor and Radiation Facility Safety Committee.

Captain Charles B. Galley, MSC, USN is replaced by Colonel Curtis W. Pearson, USAF, MSC.

AUTHORITY: Verbal orders, Director, AFRRl

I EFFECTIVE: 9 December 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCT!ONS: This appointment is made in accordance with the AFRRI Reactor Technical Specifications of NRC License R-84. All questions regarding AFRRI '

Reactor and Radiation Facility Safety Committee should be directed to Colonel Curtis W.

Pearson at 295-0438.  ;

FOR THE DIRECTOR:

d <4 CURTIS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration a

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ARMED FORCED RADIO 2iOLOGY RECEARCH INSTITUTE 8901 WISCONSIN AVENUE BETHESDA, MARYLAND 20889-5603 605.01 9 December 1996

MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member The following appointment is made:

Captain Charles B. Galley, MSC, USN ACTION: Dr. David McKown is replaced by Captain Charles B. Galley, MSC, USN, Acting RSO, AFRRI AUTHORITY: Verbal orders, Director, AFRRI EFFECTIVE: 9 December 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: Captain Galley is appointed as RSO (permanent voting member) to the RRFSC. This appointment is made in accordance with the AFRRI Reactor Technical Specifications of NRL License R-84. All questions regarding AFRRI Reactor and Radiation Facility Safety Committee should be directed to Captain Charles B. Galley at 295-9261.

FOR THE DIRECTOR:

b SW CURTIS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration

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i ATTACHMENT D 1 1

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Correction to SHD Data for CY95 Annual Report I 1

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ARMED FORCES RADIOBIOLOGY RECEARCH INSTITUTE

's 8901 WISCONSIN AVENUE BETHESDA MARYLAND 20889-5603 4

605.01 16 Ocster 1996 MEMORANDUM FOR DISTRIBUTION A

SUBJECT:

Reactor and Radiation Facility Safety Committee Member

, The following appointment is made:

Dr. David McKown ACTION: Mr. Thomas O'Brien is replaced by Dr. David McKown, Acting RSO, AFRRI i AUTHORITY: Verbal orders, Director, AFRRI EFFECTIVE: 16 October 1996 PERIOD: Until superseded or rescinded SPECIAL INSTRUCTIONS: Dr. McKown is appointed as RSO (permanent voting member) to the RRFSC. This appointment is made in accordance with the AFRRI Reactor Technical Specifications of NRC License R-84. All questions regarding AFRRI Reactor and Radiation Facility Safety Committee should be directed to Dr. David McKown at 295-1285 ,

I FOR THE DIRECTOR:  !

pr.

uw4A CURTIS W. PEARSON Colonel, USAF, MSC Deputy Director for Administration i