ML20211C755

Forwards Responses to 860903 Review Visit Questions. Corrections to SAR & Revised Tech Specs Encl

ML20211C755
Person / Time
Site:	University of New Mexico
Issue date:	10/13/1986
From:	Busch R NEW MEXICO, UNIV. OF, ALBUQUERQUE, NM
To:	Dosa J NRC OFFICE OF ADMINISTRATION (ADM), Office of Nuclear Reactor Regulation
References
NUDOCS 8610220019
Download: ML20211C755 (42)

Text

- _ _ _ _ _ _ _ _

The University of New Mexico DEPARTMENT OF CHEMICAL AND NUCLEAR ENGINEERING Albuquerque, NM 87131 Telephone 505: 277 5431 October 13, 1986 John Dosa Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk Re: Docket 50-252

Dear Mr. Dosa:

Enclosed are the responses to your questions from the facility review visit of September 3,1986. Corrections to the Safety Analysis Report are noted in the response text. The Technical Specifications have been revised as directed and are resubmitted.

If you have any further questions, please call me.

Sincerely,

- W Robert D. Busch, Ph.D., P.E.

Chief Reactor Supervisor RDB/kml enc:

1 0610220019 861013 0 PDR ADOCK 05000252 9 p PDR 0 i

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l ADDENDUM TO LICENSE RENEWAL APPLICATION FOR UNM AGN-201M REACTOR DOCKET 50-252 1.A. Figure 10 of SAR - Campus Map 1.B. Additional part of Figure 13 with humidity and wind data l for Albuquerque.

1.C. Page A-1 of GAR should be corrected to read "would rise to about 170 C".

l 1.D. P3ge 32 of SAR should read "The placing of fissile material in the glory hole is strictly limited to a maximum positive reactivity of 0.4% AK/K."

! 1.E. Figure 2 of SAR should indicate "4 inch diameter access l ports". Page 3 of SAR is correct.

2. Total thermal output of reactor since its licensing in 1966 is about 3000 watt-hours 4 Burnup over that period is estimated to be 1.54 x 10 gram of U-235.

3. Population of Albuquerque in 1980 census was 332,336.

The UNM campus population in 1986 is about 28,000 students and staff with about 14,000 on campus at any given time. The reactor is located about 0.25 miles from the nearest residential area.

4. Description of ventilation system and schematics:

During normal operation the supply fan and the relief / exhaust fan is in operation. The supply fan has 4 - 2x4 fiberglass filters that filter the incoming outside air. This air is supplied to all rooms in the nuclear engineering lab. The relief blower (located in the Janitor's room) has duct work in all areas except the whole south side of the lab, (Reactor Room 069A & B) and 079. There are grills in all the doors to the above area and a motorized louver about the door in Room 079.

The relief blower takes a suction on all rooms located on the outside east, west, and north walls. The return air from other areas in the lab must be picked up by the intakes in those rooms. The return air is moved by ductwork to a set of absolute filters and then through the relief fan and is discharged through the ventilation stack located on the roof.

There is a 100% exchange of air (no recirculation). The relief fan is set to move slightly more air than the supply fan to give a negative pressure in the building.

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NORMALS, MEANS, AND EXTREMES ALBUQUERQUE. NEW MEXICO ELEVATION: Fi. Iged 5311 lasil 5314 TIME ZONE: MOUNTAIN kBAN: 23050 L A T I T UDE : 35 '03'N LONGifuDE: 106 *3 7 ' W tai JAN FEB MAR APR MAY JUNE JULY AUG SEP OCT NOV DEC YEAR TEMPERATURE *F:

Nor*als 57.2 48.0 70.3

-Caily "a . au - 47.2 52.9 60.7 70.6 79.9 90.6 92.8 89.4 83.0 71.7 25.9 31.7 39 5 48.6 58.4 64.7 62.8 54.9 43.1 30.7 23.2 42.1

-Ca.fy 9.nr+ue 12.3 44.0 35.6 56.2

-Montst, 34.0 19.4 46.2 55.0 64.3 74.5 78.8 76.1 69.0 57.4 E.tc..es 100 91 77 72 105

- ket or d H. m..s t 45 69 75 85 89 98 105 105 101 1971 196r, 1951 1980 1980 1979 1979 1979 1975 1958 JUN 1980

-fear 1971 1972 -7 3 -17

-17 -5 8 19 28 40 54 52 37 25

-Rec or d in~es t 45 1971 1980 1976 1974 JAN 1971

-Year 1971 1951 1948 1980 1975 1980 1979 1968 NORMAL DEGREE DAYS: 0 12 242 630 911 4414 Heating Ibase 65*FI 1 36 717 583 302 81 0 0 285 428 344 132 6 0 0 1254 Cooling Ibase 65"rl 0 0 0 0 59 77 80 83 76 76 79 79 77 72 76

% OF POSSIBLE SUNSHINE 45 72 73 73 MEAN SKf COVER (tentb31 4.3 3.5 3.5 4.0 4.6 4.3 Sunr i se . Su, set 45 4.8 4.9 5.0 4.5 4.1 3.3 4.5 MEAN NUMBER OF DAYS:

Sunrise to Sueset 12.1 13.7 16.9 17.6 15.2 13.8 170.4

-C l .ar 45 13.0 11.3 11.3 12.8 14.8 17.9 10.1 9.4 10.1 8.6 14.2 12.6 7.8 7.6 7.7 7.6 111.1

-Partty Cloudv 45 7.8 7.8 9.6 93.8

-Cloudy 45 10.2 9.2 9.6 7.8 6.1 3.5 4.7 4.8 5.3 5.9 7.1 Pr ec ip i t a t i on 4.0 3.3 4.0 59.4

.01 .nches or more 45 3.8 4.0 4.5 3.3 4.2 3.7 8.8 9.3 5.7 Snow.lte pellets 4.2 1.0 inches or core 45 1.0 0.8 0.7 0.2 0.* 0.0 0.0 0.0 C.0 0.0 0.4 0.9 0.1 0.3 0.9 1.5 3.7 4.9 11.2 11.1 4.6 2.5 0.6 0.2 41.5 Thunderstor s 45 Heavy Fog visioil6ty 0.6 5.5 1/4 alle er less 45 1.1 1.0 0.6 0.2 0.s 0.8 0.1 0.8 0.1 0.3 1.4 Te.oerature "F Ma.4 24 0.0 0.0 0.0 0.0 2.7 17.5 24.3 16.8 4.5 0.2 0.0 0.0 65.8 90 and aba.. 0.0 0.2 5.0 32" and below 24 2.5 0.6 0.2 0.0 0.0 0.0 0.0 0.0 0.0 1.5

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Hour 17 24 40 32 24 48 52 44 35 33 32 47 52 52 49 55 62 Hour 23 24 61 PRECIPlfATION 1 nchesi:

Water Equivalent 8.12 0.41 0.40 0.52 0.40 0.46 0.51 1.30 1.51 0.05 0.06 0.38 0.52

-Norail 1.45 1.85 3.33

-Masieu . Monthly 45 1.32 1.42 2.18 1.97 3 07 1.11 3.33 3.30 1.99 3.08 1948 1973 1942 1941 1967 1968 1967 1940 1972 1940 1959 JUL 1968

- Yo se 1978 0.00

-Mini va Monthly 45 r I i i i i 0.08 i i 0.00 0.00 0.00 1970 1984 1966 1972 1945 1975 1980 1962 1957 1952 1949 1981 CEC 1981

-Year 1.92 1.80 0.76 1.35 1.92

-Ma.la ua in 24 t-s 45 0.87 0 51 1.11 1.66 1.14 1.04 1.17 1.75 1962 1981 1973 1969 1961 1952 1961 1980 1955 1969 1940 1958 SEP 1955

- f e ar Sno'w. ic e pe l le ts 14.7

-Ma.i u. Monthly 45 95 0.2 13.9 8.1 1,0 i 0.9 9.3 14.7 1964 1973 1973 1979 1971 1979 1940 1959 DEC 1959

- Y e ar 1973 14.2 Ma.imum in 24 hrs 45 51 4.2 10.7 6.6 1.0 T 0.9 5.5 14.2 1946 1973 1973 1979 1971 1979 1946 1958 DEC 1958

- l e ar 1973 WINO: 8.3 7.8 7.7 9.0 Mean Speed l+on) 45 8.1 8.9 10.2 11.1 10.6 10.0 9.1 8.2 8.6 Pr .ailina Direc t ion N N SE SE SE SE SE through f963 N N SE S S S lastest Mlle N NW SE SE W E SE SE

- Dir ce t ion lit! 44 E NW NW S SE 82 68 61 62 66 57 90 90 speed iMPwl 44 61 68 80 72 72 1149 1944 1943 1946 1950 1946 1945 1951 1945 1959 1948 1943 DEC 1943 y e ar till See Reference Notes on Page 68.

Page 3

There are also two exhaust fans that take a suction on two vent hoods in Room 079. When in operation and exhausting air from the vent hood, the air passes through a set of absolute (4) filters and the fan exhausts to the roof stack. There is a set of filters.

for each exhaust fan. There are motorized louvers located in each duct just before the filters. When lined up to use one of the hoods the louver is closed in that duct so that all the air is drawn through the hood. In an emergency, the hoods may be set to open these motorized louvers so that they are open to the lab. In this mode, more air can be moved to remove anything dangerous or unpleasant in the air, (air-borne contamination, smoke, chemical spills, etc). The switches for these fans are located near the east entrance so the lab does have control of these fans.

Capacity: See Specifications.

Building Pressure - Slightly negative in normal operation -- highly negative if both hood fans are also exhausting air.

Type of Filters - Fiberglass on supply (oiled), absolute on intake to relief fans and intakes to both hood fans.

99.97% effective, 4 filters to each fan.

5. Electrical power and emergency power systems in the reactor room: Electrical power is supplied to the plug molds in the reactor room via main distribution panel-circuit breaker #18 located in the electrical room just inside the Nuclear Engineering Lab at the east door. The reactor console is hardwired into the south plug mold. The plug mold has a circuit breaker located on one end that will remove power from that plug mold and the reactor console. When the console power is turned off, the reactor automatically scrams so a power failure shuts the reactor down. Heat removal is done by natural conduction so no power is required for orderly, controlled shutdown. There is no emergency power as none is required for shut down.

6. The fuel is normally in a sealed core tank except when the approach to critical experiment is performed. This experiment is done under the supervision of the UNM Radiation Safety group. Anti "C" clothing is worn while handling fuel and a clean area is set up on the reactor top. The tank is vented through a charcoal filter. A clean area is also set up on the bench in the reactor room where the disks are temporarily stored during the experiment. One additional metal encased fission plate is stored in a storage cabinet in the reactor room. If the fuel would ever be removed from the reactor, it would be stored in the source Room 077 in an array with Keff < 0.9.

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7. Fire protection in the reactor room is provided by a fire extinguisher located on the north wall. A backup extinguisher is located outside the west reactor room door. Most of the material in the reactor room will not burn so the extinguisher is available to handle electrical fires.

8. We follow ALARA commitment so access at levels less than 0.9 watt is strictly supervised but accessible to make short duration measurements.

9. Page 39, section C.3 of SAR should indicate a radiation reading of 80 mrem /hr of gammas attenuated through 16" high density concrete block. The reduced dose should be about 1 mrem /hr.

10. The Reactor Laboratory's experimental program is directed towards training of graduate and undergraduate students in neutron physics, reactor kinetics, and neutron activation. All experiments are done under the supervision of a licensed reactor operator and a faculty member. There are 5 standard experiments performed:

1. Approach to Critical - involves removal of the top 5 fuel disks and roplacement with dummy fuel disks to simulate fuel loading.

2. Reactor period and rod drop - involves measurement of periods due to small and large changes (up to 0.20% A K/K) in positive and negative reactivity through rod manipulation.

3. Importance functions - involves insertion of various small foils into glory hole to analyze reactivity effects versus position in core.

4. Reactor neutron temperature - involves use of boron and cadmium to measure flux levels and calculate energy from 1/V relationships.

5. Reactor flux profile and power calibration - involves detector in the glory hole and activation use of DF,11s of gold f6 to determine power levels.

These experiments and activation of small foils (< 0.5g)  ;

of Ag, Au, In, U-natural, Mn, Cu, etc. are considered to

)

be routine. All operations on the reactor which are not defined as routine are non-routine. (See Operations  !

l Manual pgs. III-3 through III-5.) The experimental facilities consist of the glory hole and the access ports. Routine experiments tend to be those using the glory hole since this does not involve a change in configuration.

l l

l l

1

11. The only experiments allowed in the access ports are those which are less than 0.4% A K/K in positive reactivity and have the prior approval of the Chief Reactor Supervisor and the Reactor Safeguards Advisory Committee. All experiments in the access ports are subject to the same restrictions as those in the glory hole.

12. UNM is committed to the ALARA program with the responsibility for achieving ALARA resting in the occupational Safety Division, Radiological Safety Office (RSO). The RSO conducts radiation surveys of the reactor room on a monthly basis with at least two surveys per year performed while the reactor is operational. All reactor related personnel must pass a radiation exam and be certified as Radiation supervisors before they can be included in reactor licensing training programs. The Radiological Safety Officer has reviewed routine experiments for radiological protection needs and would have to review and non-routine experiment also.

13. See Table 1.

I I

14. Technical Specifications have been updated and l resubmitted. Changes are noted by a vertical line in the margin.

Table 1 - Summary of Annual Personnel Exposures (for reactor-related personnel)

Estimated whole body Number of individuals exoosure rance (rems) in each rance 1985 1984 1983 1982 1981 No measurable exposure 1 2 4 3 0

< 0.10 rem 6 4 2 2 4 0.10 to 0.25 rem 0 0 0 0 1 0.25 to 0.50 rem 0 0 0 0 0

> 0.5 rem 0 0 0 0 0 l

1

LICENSE NO. R-102 TECHNICAL SPECIFICATIONS FOR UNIVERSITY OF NEW MEXICO AGN-201M REACTOR (SERIAL #112)

DOCKET NO. 50-252 DATE: MAY 20, 1986 AS MODIFIED TO INCLUDE ANSI /ANS 15.1-1982 GUIDANCE i

I

TABLE OF CONTENTS PAGE 1.0 DEFINITIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety limits 2.2 Limiting Safety System Settings

(

3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits '

3.2 Control and Safety Systems 3.3 Limitation on Experir.ents 3.4 Radiation Monitoring, Control, and Shielding 4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactivity Limits 4.2 Control and Safety Systems 4.3 Reactor Structure 4.4 Radiation Monitoring and Control 5.0 DESIGN FEATURES 5.1 Reactor 5.2 Fuel Storage 5.3 Reactor Room, Reactor Control Room, Accelerator Room

.9 '

TABLE OF CONTENTS (CONT)

PAGE 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.2 Staff Qualifications 6.3 Training 6.4 Reactor Safeguards Advisory Committee 6.5 Approvals 6.6 Procedures 6.7 Experiments 6.8 Safety Limit Violations 6.9 Reporting Requirements 6.10 Record Retention I

1.0 DEFINITIONS The terms Safety Limit (SL) , Limiting Safety System Setting (LSSS),

and Limiting Conditions for Operation (LCO) are as defined in 50.36 of 10 CFR part 50.

1.1 Certified Operator - A certified operator is an individual authorized by the Nuclear Regulatory Commission (NRC) to carry out the duties and responsibilities associated with operation of the reactor.

1.2 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within acceptable range and accuracy, to known values of the paramater which the channel measures. Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip.

1.3 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification may include comparison of the channel with other independent channelo or methods measuring the same variable.

1.4 Channel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.

1.5 Control Rod - Any of the four moveable rods loaded with fuel which are manipulated by the operator to change the reactivity of the reactor. This includes safety type and regulating rods.

1.6 Excess Reactivity - The amount of reactivity above a k-effective of 1. This is the amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the eff = 1) point whare the reactor is exactly critical (K 1.7 Experiment - An experiment is any of the following:

a. An activity utilizing the reactor system or its components or the neutrons or radiation generated therein;

b. An evaluation or test of a reactor system operational, surveillance, or maintenance technique;

c. The material content of any of the foregoing, including structural components, encapsulation or confining boundaries, and contained fluids or solids.

1.8 Experimental Facilities - Experimental facilities are those portions of the reactor ausembly that are used for the introduction of experiments into or adjacent to the reactor core region or allow 1

beams of radiation to exist from the reactor shielding.

Experimental facilities shall include the thermal column, glory hole, and access ports.

1.9 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N.I.

Sax, third Ed. (1968), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the. National Fire Protection Association in its publication 704-M, 1966, " Identification System for Fire Hazards of Materials," also enumerated in the " Handbook for Laboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.

1.10 Maior Chance - Any change in reactor configuration which affects the probability or consequences of an event.

1.11 Measured Value - The measured value is the value of a parameter as it appears on the output of a channel.

1.12 Measurina Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring or responding to the value of a process variable.

1.13 Movable Exoeriment - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.

1.14 Operable - Operable means a component or system is capable of performing its intended function in its normal manner.

1.15 Operatina - Operating means a component or system is performing its intended function in its normal manner.

1.16 Potential Reactivity Worth - The potential reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

\

1.17 Reactor Comnonent - A reactor component is any apparatus, device, or material that is a normal part of the reactor assembly.

1.18 Reactor Oneration - Reactor operation is any conditicn wherein the reactor is not shutd.own.

1.19 Reactor Safety System - The reactor safety system is that combination of safety channels and associated circuitry which forms an automatic protective system for the reactor or provides information which requires manual protective action be initiated.

1.20 Reactor Secured - The reactor shall be considered secured whenever

a. cither: 1. All safety and control rods are fully with-drawn from the core, or 2

2. The core fuse melts resulting in separation of the core, and

b. The reactor console key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator.

1.21 Reoulatino Rod - A low worth control rod used primarily to maintain an intended power level. Its position may be varied manually.

1.22 Removable Experiment - A removable experiment is any experiment,

' experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.

1.23 Research Reactor - A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, educational, training, or experimental purposes, and which may have provisions for producing radioisotopes.

1.24 Safety Channel - A safety channel is a measuring channel in the reactor safety system.

1.25 Scram Time - The time for the control rods acting under gravity to change the reactor from a critical to a subcritical condition., In most cases,_this is less than or equal to the time it takes for the rod to fall from full in to full out position.

1.26 Secured Exoeriment - Any experiment, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor by mechanical means.

The restraint shall exert sufficient force on the experiment to overcome the expected effects of hydraulic, pneumatic, bouyant, or other forces which are normal to the operating environment of the experiment or which might arise as a result of credible malfunctions.

1.27 Shall. Should and May - The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission--neither a requirement nor a recommendation. _

1.28 Shutdown Marcin - Shutdown margin shall mean the minimum shutdcwn reactivity necessary to provide confidence that the reactor can be made subcritial by means of the control and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive condition, and that the reactor will remain subcritical without further operator action. '

1.29 Static Reactivity Worth - The static reactivity worth of an experiment is the value of the reactivity change which is measureable by calibrated control or regulating rod comparison 3

. 1 methods between two defined terminal positions or configurations of the experiment. For removable experiments, the terminal postions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.

1.30 Surveillance Time - A surveillance time indicates the frequency of tests to demonstrate performance. Allowable surveillance intervals shall not exceed the following:

a. Two-year (interval not to exceed 30 months)

b. Annual (interval not to exceed 15 months) c .'

d.

Semi-annual (interval not to exceed seven and one-half months)

Quarterly (interval not to exceed four months)

e. Monthly (interval not to exceed six months).

1.31 True Value - The true value is the actual value of a parameter.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to the maximum core temperature during steady state or transient operation.

Obiective To assure that the integrity of the fuel material is maintained and all fission products are retained in the core matrix.

Specification

a. The maximum core temperature shall not exceed 200 C during either steady state or transient operation.

Bases The. polyethylene core material does not melt below 200 C and is l expected to maintain its integrity and retgin essentially all of the fission products at temperatures below 200 C.

l The Hazards Summary l Report dated February 1962 submitted on Docket F-15 by Aerojet-General Nucleonics (AGN)gcalculated a steady state core average temperature rise of 0.44 C/ watt. Tge corresponding maximum core temperature would be below 200 C thus assuring integrity of the core and retention of fission products.

2.2 Limitina Safety System Settinas Applicability This specification applies to the parts of the reactor safety system j which will limit maximum power and core temperature, l

l 4

1 Obiective To as ure that automatic protective action is initiated to prevent a

! safety limit from being exceeded.

Specification

a. The safety channels shall initiate a reactor scram at the following limiting safety system settings:

! Channel Condition LSSS Nuclear Safety #2 High Power i 10 watts Nuclear Safety #3 High Power s 10 watts

b. The polystyrene core thgrmal fuse' melts when heated to a temperature of 120 C or less resulting in core separation and a reactivity . loss greater than 5% A k.

Bases i  !

Based on instrumentation response times and scram tests, the AGN Hazards Report concluded that reactor periods in excess of 30-50 milli-seconds would be adequately arrested by the scram system.

Since the maximum available excess reactivity in the reactor is less than one dollar the reactor cannot'become prompt critical and the corresponding shortest possible period is greater than 200 milli-seconds. The high power LSSS of 10 watts in conjunction with autogatic safety systems, and the average temperature rise of 0.44 C/ watt, and/or manual scram capabilities will assure that-the safety limits will not be exceeded during' steady state or.as a result of the most severe credible transient.

In the event of failure of the reactor to scram, the self-limiting characteristics due to the high negative temperature coefficiegt, and the to sting of the thermal fuse at a temperature below 120 C l wilg assure-safe shutdown without exceeding a core temperature of

200 C 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits l

Apolicability This' specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.

Obiective l

To assure that the reactor can be shut down at all times and that the safety limits will not be exceeded.

1 5 i

u . - _ _ ._ . _ . _ _

Soecification

a. The available excess reactivity with all control and safety rods fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% Ak/k.

b. The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1% A k/k.

c. The reactivity worth of the control and safety rods shall ensure subcriticality on the withdrawal of the coarse control rod or any one safety rod.

d. The excess reactivity with no experiments in the reactor and and the control and safety rods fully inserted shall not exceed 0.25% A k/k.

Bases The limitations on total core excess reactivity assure reactor periods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down without exceeding any safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive position.

3.2 Control and Safety Systems Apolicability These specifications apply to the reactor control and safety systems.

Obiective To'specify lowest acceptable level of performance, instrument' set points, and the minimum' number of operable components for the reactor control and safety systems.

Specification

a. The fine control rod, coarse control rod, and the two safety rods shall be operable and the carriage position of the regulating rods shall be displayed at the console whenever any rod is above its lower limit.

b. The total scram withdrawal time of the safety rods and coarse control rod shall be less than 1 second as inferred from strip chart data taken at high recording speed.

c. The average reactivity addition rate for each control or safety rod shall not exceed 0.065% A k/k per second.

6

1

d. The safety rods and coarse control rod sha'll be interlocked such that:

1. Reactor startup cannot commence unless both safety rods and coarse control rod are fully withdrawn from the core.

2. Only one safety rod'can be inserted at a time.

3. -The coarse control rod cannot be inserted unless both safety rods are fully inserted.

e. Nuclear safety channel instrumentation shall be operable in accordance with Table 3.1 whenever.the reactor control or i

safety rods are not in their fully withdrawn position.

f. A manual scram shall be provided on the reactor console, and the safety circuitry shall be designed so that no single failure can negate both the automatic and manual scram capability.

g. The shield water level interlock shall be set to prevent reactor startup and scram the reactor if the shield water level' falls 7 inches below the highest point on the reactor shield tank manhole opening.

h. The shield water temperature interlock shall be set to prevent reactor startup and scram ghe reactor if the shield water temperature falls below 18 C.

i. The seismic displacement interlock shall be installed in such a manner to prevent reactor startup and scram the reactor during a seismic displacement.

j. A loss of electric power shall cause.the reactor to-scram.

4 Bases The specification on operability of the rods assure console control over reactivity conditions within the reactor. Display of the positions of the regulating (fine and coarse) rods assures that the positions of these rods be available to the operator to evaluate the-configuration of the reactor.

The specifications on scram withdrawal time in conjunction with the safety system instrumentation and set points assure safe reactor shutdown during the most severe foreseeable transients. Interlocks on control and safety rods assure an orderly approach to criticality

, and an adequate shutdown capability. The limitations on reactivity addition races allow only relatively slow increases of reactivity so that ample time will be available for manual or automatic scram during any operating conditions.

The neutron detector channels (nuclear safety channels 1 through 3) assure that reactor power levels are adequately monitored during i

, 7

reactor startup and operation. Requirements on minimum neutron levels will prevent reactor.startup unless channels are operable and responding, and will cause a scram in the event of instrumentation failure. The power level scrams initiate redundant automatic i

protective action at power level scrams low'enough to assure safe

, shutdown without exceeding any safety limits. The period scram conservatively limits the rate of rise.of reactor power to periods which are manually controllable and will automatically scram the reactor in the event of unexpected large reactivity additions. The manual scram assures a method of shutdown without reliance on safety channels and circuitry.

The AGN-201's negative temperature coefficient of reactivity causes a reactivity increase with decreasing core temperature. The shield water temperature intgrlock will prevent reactor operation at temperatures below 18 C thereby limiting potential reactivity additions associated with temperature decreases.

Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation levels. The shield tank water level interlock will prevent reactor operation without adequate water levels in the shield tank.

The reactor is designed to withstand 0.6g accelerations and 6 cm displacements. A seismic instrument causes a reactor scram whenever the instrument receives a horizontal acceleration that causes a horizontal displacement of 1/16 inch or greater. The seismic displaceme'nt interlock assures that the reactor will be scrammed and brought to a subcritical configurction during any seismic disturbance that may cause damage to the reactor or its components.

I i

8

Table 3.1 Nuclear Instrumentation Channel No. Function Operatina Limit 1 Monitor- None

^

2 Low Power Scram -13 1 x 10 amperes High Power Scram' 200% of licensed power Short Reactor Period 5-second minimum period 3 Low Power Scram 5% of operating range High Power Scram 200% of licensed power 4

9

The manual scram allows the operator to manually shutdown the reactor if an unsafe or othewise abnormal condition occurs that does not otherwise scram the reactor. A loss of electrical power de-energizes the safety and coarse control rod holding magnets causing a reactor scram thus assuring safe and immediate shutdown in case of a power outage.

3.3 Limitations on Experiments Loolicability This specification applies to experiments installed in the reactor and its experimental facilities.

Obiective To prevent damage to the reactor or excessive release of radioactive materials in the event of an experimental failure.

Specification

a. Experiments containing materials (within the reactivity limits defined in specification 3.1) corrosive to reactor components or which contain liquid or gaseous, fissionable materials shall be doubly encapsulated.

b. Explosive-materials or materials which might combine violently shall not be inserted into experimental facilities of the reactor or irradiated in the reactor.

c. The radioactive material content, including fission products, of any doubly encapsulated experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation could not result in doses in excess of the equivalent annual doses stated in 10CFR20. This dose limit applies to persons occupying unrestricted areas continuously for two hours starting at the time of release, and occupying restricted areas during the length of time required to evacuate.the restricted area.

Bases These specifications are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from an experimental failure and to protect operating personnel and the public from excessive radiation doses in the event of an experimental failure. Specification 3.3c conforms to the regulatory position put forth in Regulatory Guide 2.2 issued November, 1978.

3.4 RADIATION MONITORING, CONTROL AND SHIELDING Acolicability 10

This specification applies to radiation monitoring, control, andreactor shielding required during reactor operation.

Obiective The objective is to protect facility personnel and the public from radiation exposure.

Soecification

a. An operable portable radiation survey instrument capable of detecting gamma radiation shall be immediately available to reactor operating personnel whenever the reactor is not secured,

b. The reactor room shall be considered a restricted area whenever the reactor is not secured.

c. The reactor room shall be' considered a radiation area whenever the reactor is operated.

d. Whenever the reactor is operated, the top of the reactor shall '

be considered a high radiation area and the access stairs to the top of the reactor shall be equipped with a control device which shall energize a conspicuous audible alarm signal in such manner that the individual using the stairs'and the reactor operator are made aware of the entry.

e. The following shielding requirement shall be fulfilled during reactor operation: 1 L

.The thermal column shall be filled with water or graphite except during a critical experiment (core loading) or during other approved experiments requiring the thermal column to be empty.

f.- The core tank shall be sealed during reactor operation.

Bases Radiation surveys performed under the supervision of a qualified health physicist have shown that the total gamma, thermal neutron, j

and fast neutron radiation dose rate in the reactor room, at the closest approach to the reactor, is less than 50 mrem /hr at reactor power levels less than 5.0 watts (i.e., without access to reactor top).

~

The facility shielding in conjunction with radiation monitoring, control, and restricted areas is designed.to limit radiation doses i to facility personnel and to the public to a level below 10 CFR 20 limits under operating conditions, and to a level below criterion 19, Appendix A, 10 CFR 50 recommendations under accident conditions.

11

4.0 SURVEILLANCE REOUIREMENTS Actions specified in this section are not required to be performed if during the specifed surveillance period the reactor has not been brought critical or is maintained in a shutdown condition extending beyond the specified surveillance period. However, the surveillance requirements must be fulfilled prior to subsequent startup of_the reactor.

4.1 Reactivity Limits Anolicability This specification appies to the surveillance requirements for reactivity limits.

Obiective To assure that reactivity limits for Specification 3.1 are not exceeded.

Specification

a. Safety and control rod reactivity worths shall be measured annually.

b. Total excess reactivity and shutdown margin shall be determined annually,

c. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before or during the first startup subsequent to the experiment's first insertion.

Eases The control and safety rod reactivity worths are measured annually to assure that no degradation or unexpected changes have occurred which could adversely affect reactor shutdown margin or total excess reactivity. The shutdown margin-and total excess reactivity are determined to assure that the reactor can always be safety shutdown with one rod not functioning and that the maximum possible reactivity insertion will not result in reactor periods shorter than those that can be adequately terminated by either operator or automatic action. Based on experience with AGN-reactors, significant changes in reactivity or rod worth are not expected within a 12 month period.

4.2 Control and Safety Systems Applicability This specification applies to the surveillance requirements of the reactor control and safety systems.

12

Obiective To assure that the reactor control and safety systems are operable as required by Specification 3.2.

Specification

a. Safety and control rod scram times and average reactivity-insertion rates shall be measured annually.

b. Safety and control rods and drives shall be inspected for proper operation annually.

c. A channel test of the following safety channels shall be performed prior to the first reactor startup of the day or prior to each reactor operation extending more than one day:

Nuclear Safety #2 and #3

d. A channel test of the seismic displacement interlock shall be performed semiannually.

e. A channel check of the following safety channels shall be performed daily whenever the reactor is in operation:

Nuclear Safety #2 and #3

f. Prior to each day's reactor operation or prior to each reactor operation extending more than one day, safety rod #1 shall be inserted and scrammed to verify operability of the manual scram system.

g. The period, count rate, and power level measuring channels shall be calibrated and set points verified annually, but at intervals not to exceed 16 months.

h. The shield water level interlock and shield water temperature interlock shall be calibrated by perturbing ~the sending element to the appropriate set point. These calibrations shall be performed annually, but at intervals not to exceed 16 months.

! Bases The channel tests and checks required daily or before each startup will assure that the safety channels and scram functions are operable. Based on operating experience with reactors of this type, the annual scram measurements, channel calibrations, set point verifications, and inspections are of sufficient frequency to assure, with a high degree of confidence, that the safety system settings will be within acceptable drift tolerance for operation.

13

l 4.3 Reactor Structure _l l

Applicability l This specification applies to surveil' lance requirements for reactor components other than control and safety rods.

Obiective The objective is to assure integrity of the reactor structures.

Specification Visual inspection for water leakage from the shield tank shall be performed annually. Leakage sufficient to leave a puddle on the floor shall be corrected prior to subsequent reactor operation.

Bases Based on experience with reactors of this type, the frequency of inspection and leak test requirements of the shield tank will assure capability for radiation protection during reactor operation.-

4.4 Radiation Monitorina and Control Apolicability This specification applies to the surveillance requirements of the radiation monitoring and control systems.

Obiective To assure that the radiation monitoring and control systems are operable and that all radiation and high radiation areas within the reactor facility are identified and controlled as required by Specification 3.4.

Specification

a. All portable radiation survey instruments assigned to the reactor facility shall be calibrated under the supervision of the Radiological Safety Office annually. The reactor area Remote Area Monitor shall b3 calibrated annually with the internal check source.

b. Prior to each day's reactor operation or prior to each reactor operation extending more than one day, the reactor access high radiation area alarm (Ref 3.4d) shall be verified to be operable.

c. A radiation survey of the reactor room, shall be performed under the supervision of the Radiological Safety Office annually, but at intervals not to exceed 16 months, to determine the location of radiation and high radiation areas corresponding to reactor operating power levels.

14

Bases The periodic calibration of radiation monitoring equipment and the surveillance of the reactor access high radiation area alarm (Ref 3.4d) will assure that the radiation monitoring and control systems are operable during reactor operation.

The period radiation surveys will verify the location of radiation and high radiation areas and will assist reactor facility personnel in properly labeling and controlling each location in accordance with 10 CFR 20.

5.0 DESIGN FEATURES 5.1 Reactor

a. The reactor core, including control and safety rods, contains approximately 667 grams of U-235 in the form of 20% enriched UO 2 dispersed in approximately 11 kilograms of polyethylene.

The lower section of the core is supported by an aluminum rod hanging from ga fuse link. The fuse melts at a fuse temperature of about 120 C causing the lower core section to fall away from the upper section reducing reactivity by at least 5% Ak/k.

Sufficient clearance between core and reflector is provided to insure free fall of the bottom half of the core during the most severe transient.

b. The core is 3 surrounded by a 20 cm thick high density (1.75 gm/cm ) graphite reflector followed by a 10 cm thick lead gamma shield. The core and part of the graphite reflector are sealed in a fluid-tight aluminum core tank designed to contain any fission gases that might leak from the core.

c. The core, reflector and lead shielding are enclosed in and supported by a fluid-tight steel reactor tank. An upper or

" thermal column tank" may serve as a shield tank when filled with water or a thermal column when filled with graphite.

d. The 6 1/2 foot diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick fast neutron shield.

The fast neutron shield is formed by filling the tank with approximately 1000 gallons of water. The complete reactor shield shall limit doses to personnel in unrestricted areas to levels less than permitted by 10 CFR 20 under operating conditions.

e. Two safety rods and one control rod (identical in size) contain less than 15 grams of U-235 each in the same form as the core material. These rods are lifted into the core by electromagnets, driven by reversible DC motors through lead screw assemblies. De-energizing the magnets causes a spring-driven, gravity-assisted scram. The fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw. This rod may contain fueled or unfueled polyethylene.

15

~

i 5.2 Fuel Storace

] Fuel, including fueled experiments and fuel devices not in the reactor, shall be stored.in locked Room #077 in the Nuclear ,

Engineering department laboratories. The storage array shall be {

such that K is no greater than 0.9 for all conditions of t moderation $bk reflection.

5 . 3' Reactor Room. Reactor Control Room. Accelerator Room 4 a. The reactor room houses the reactor assembly and accessories required for its operation and maintenance, and the reactor control console.

4

b. The reactor room is a separate room in the Nuclear Engineering Laboratory, constructed with adequate shielding and other i radiation protective features to limit doses in restricted and

{ unrestricted areas to levels no greater than permitted by

10 CFR 20.

c. The' access doors to the reactor room shall contain locks.

4 6.0 ADMINISTRATIVE CONTROLS I

6.1 Oraanization-The administrative organization for control of the reactor facility

, and its operation shall be as set forth in Figure 1 attached hereto. .The authorities and responsibilities set forth below are designed to comply with the intent and requirements for administrative controls of the reactor facility as set forth by the Nuclear Regulatory Commission.

2 6.1.1 President The President is the chief administrative officer responsible for the University and in~whose name the application for licensing is made.

i 6.1.2 Dean. Collece of Encineerina j

~

The Dean of Engineering is the administrative officer responsible for-the operation of the College of Engineering.

6.1.3 Reactor Administrator i

Provides final ~ policy decisions on all phases of reactor operation and regulations for the facility. He is advised on matters concerning personnel. health and safety by the Radiological Control Officer and/or.the Committee on Radiological Control. He is advised on matters concerning safe operation of the reactor far the Reactor Operations Committee and/or the Reactor Safeguards Advisory Committee. He designates Reactor Supervisors and names the Chief Reactor Supervisor. He approves all regulations, instructions and 4

f 16 l

1' FIGURE 1 Administrative Organization of the University of New Mexico AGN-201M Reactor Facility NRC License R-102 i

i 1

i Level 1

! Level 1. President of 1 the University Committee on Radiological Control

) Dean, College of Engineering i

Level 2 Level 2 Reactor Administrator i .

Radiological Reactor Safeguards Safety Officer Advisory Committee Level 3 l Level 3 Chief Reactor -

l Supervisor

_ Reactor Operations I Commi ttee Level 4 j Level 4 Reactor -

Supervisors

?

Reactor I Operators I

f Reactor Assistants Authorized Operators t

3 4

l 4-I 17 i

e -- , - -:rm..-ww,, --- w,--,e.f ,. ,m,3, ...,--.----r---, _,.r-+m.- r e- m-. m e- ----v-mmvv.-- - . - - - -y-- g 1 - --

- .x..  :- _ =-u a a n . . . - - - -

procedures governing facility operation. He submits the annual report to NRC. He issues and changes the code on the cipher locks of the Nuclear Engineering Laboratory Building at the beginning of each semester. He is the Chairman of the Nuclear Engineering Department and shall hold a graduate degree in Engineering.

6.1.4 Radiolocical Safety Officer .

1 Normally represents the Committee cn Radiological Control in matters concerning the radiological safety aspects of reactor operation. He is available for advice and assistance on radiological safety problems. He is the emergency director if an emergency involves radiation safety, a

6.1.5 Reactor Safeauards Advisory Committee l'

Reviews and evaluates reactor operations and procedures to ensure that the reactor shall be operated in a safe and competent manner.

There shall be at least four members on the RSAC with at least two

, members from organizations outside the University. The committee is available for advice and assistance on reactor operation <

problems. They must approve any major change in the facility. They meet semi-annually. They audit the physical security plan every two years.

! 6.1.6 Reactor Operations Committee Consists of the Reactor Supervisors with the Chief Reactor l Supervisor. Other qualified persons may also be members. They are directly. responsible to the Reactor Administrator for the preparation and submission of detailed procedures, regulations, forms, and rules to ensure the maintenance, safe operation,

competent use and security of the facility. The Committee ensures J

l that all the activities, experiments, and maintenance involving the facility are properly logged and are in accordance with established j local and U.S. Nuclear Regulatory Commission regulations. ' They l review all proposed changes in procedure or changes in the facility i

and must approve any minor change before the change is implemented.

6.1.7 Chief Reactor Suoervisor i

Holds a senior operator's license issued by the NRC. He is l responsible for the distribution and enforcement of rules, regulations and procedures concerning operation of the facility. He has the authority to authorize any experiments or procedures which

, have-received appropriate prior approval by the Reactor Operations Committee, the Reactor Safeguards Advisory Committee and/or the Committee on Radiological Control (or the Radiological Safety i- Officer) and have received prior authorization by the Reactor i Administrator. He shall not authorize any proposed changes in the facility or in procedure until appropriate evaluation and approval

, has been made by the Reactor Operations Committee or the Reactor i Safeguards Advisory Committee and authorization given by the Reactor Administrator. The Chief Reactor Supervisor is directly responsible ,

i 18

- - - , . . . . , - . . - . _ . _ . . _ _ _ . . _ _ _ _ - - . _ _ _ _ _ _ _ . _ _ _ . _ - , , m._ _.-___-_______.,_.-__m. . _ . . _ . - _ , , __-.---.m,.

for enforcing operating procedures and ensuring that the facility is operated in a safe, competent and authorized manner. He is directly responsible for all prescribed logs and records. He is the emergency director for emergencies not involving radiation.

6.1.8 Reactor Supervisors Shall hold valid senior operator licenses issued by the Nuclear Regulatory Commission. A Reactor Supervisor shall be in charge of the facility at all times during reactor operation and must witness the startup and intentional shutdown procedures. The Reactor Supervisors are directly responsible to the Chief Reactor Supervisor. The Reactor Supervisors do not have to be present other than when the reactor is going critical, being intentionally shutdown, or when reactor experiments are loaded or unloaded.

However, the location of the supervisor must be known to the Reactor Operator at all times during operation so that it is possible to contact him/her if required.

6.1.9 Reactor Operators Must hold a valid operator's license issued by the NRC. They must conform to the rules, instructions and procedures for the startup, operation and shutdown of the reactor, including emergency procedures. Within the constraints of the administrative and supervisory controls outlined above, a reactor operator will be in direct charge of the control console at all times that the. reactor is operating. The reactor operator is required to maintain complete and accurate records of all reactor operations in the operational logs.

6.1.10 Authorized Operators These are individuals who are authorized by the Reactor Supervisor to operate the reactor controls and who do so with the knowledge of the Supervisor and under the direct supervision of a licensed Reactor Operator.

6.1.11 Reactor Assistant These are individuals who are present during a reactor operation to provide assistance to the Operator as needed, with the exception that a Reactor Assistant does not operate the controls of the reactor. In an emergency they may push the Reactor Scram button.

6.1.12 Operatina Staff

a. The minimum operating staff during any time in which the reactor is not shutdown shall consist of all of the following:

One licensed Reactor Operator in the reactor control room.

One other person in the reactor room or reactor control room qualified to activate manual scram and initiate emergency procedures.

19

l l

One licensed Senior Reactor Operator readily available on call. This requirement can be satisfied by having a licensed. Senior Reactor Operator perform the duties stated in paragraph 1 or 2 above or by designating a licensed Senior Reactor Operator who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes,

b. A licensed Senior Reactor Operator shall supervise all

reactor maintenance or modification which could affect the reactivity or the reactor,

c. A listing.of reactor facility personnel by name and phone number shall be conspicuously posted in the reactor control room.

6.2 Staff Oualifications The Chief Reactor Supervisor, licensed Reactor Operators, and technicians performing reactor maintenance shall meet the minimum-qualifications set forth in ANS 15.4, " Standards for Selection and Training of Personnel for Research Reactors". Reactor Safeguards Advisory Committee members shall have a minimum of five (5) years experience in a technical profession or a baccalaureate degree and two (2) years of professional experience. The Radiological Safety Officer shall'have a baccalaureate degree in biological or physical science and have at least two (2) years experience in health physics.

i 6.3 Trainina-

. The Head of the Department of Nuclear Engineering shall be responsible for directing training as set forth in ANS 15.4-1977,

" Standards for Selection and Training of Personnel for Research Reactors". All licensed reactor operators shall participate in requalification training as set forth in 10 CFR 55.

6.4 Reactor Safeauards Advisory Committee 6.4.1 Meetinas and Ouorum l The Reactor Safeguards ~ Advisory Committee shall meet as often as i deemed necessary by the Reactor Safeguards Advisory Committee 1

Chairman but shall meet at least once each calendar year. A quorum for the conduct of official business shall be three (3) members.

6.4.2 Reviews The Reactor Safeguards Advisory Committee shall review:

i a. Safety evaluations for changes to procedures, equipment or systems, and tests or experiments, conducted without Nuclear l Regulatory Commission approval under the provision of 10 CFR 50 paragraph 50.59 to verify that such actions do not constitute an unreviewed safety question.

1 20 l

b. .

Proposed changes to procedures, equipment or systems that change the original intent or use, and are non-conservative, or those that involve an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.  ;

c. Proposed tests or experiments which are significantly different from previous approved tests or experiments, or those that

' involve.an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.

d. Proposed changes in Technical Specifications or licenses.

e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety.

g. Reportable occurrences.

h. Audit reports.

6.4.3 Audits

. Audits of facility activities shall be performed at least annually -

under the cognizance of the Reactor Safeguards Advisory Committee but in no case by the personnel responsible for the item audited.

These audits shall examine the operating records and encompass but shall not be limited to the following:

a. The conformance of the facility operation to the Technical Specifications'and applicable license conditions, at least '

annually.

'b. The Facility Emergency Plan and implementing procedures, at

least.every two years.

c. The Facility Security Plan and implementing procedures, at least every two years.

! d. Operator requalification program and records, at least every two years.

e. Results of actions taken to correct deficiencies, at least annually, j 6.4.4 Authority

~ The Reactor Safeguards Advisory Committee shall report to the Reactor Administrator and shall advise the Chief Reactor Supervisor i on those areas of responsibility outlined in Section 6.1.5 of these '

Technical Specifications. l l

i i

21

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6.4.5 Minutes of the Reactor Safeauards Advisory Com3ittee One member of the Reactor Safeguards Advisory Committee shall be designated to direct the preparation, maintenance, and distribution lof minutes of its activities. These minutes shall include a summary of'all meetings, actions taken, audits, and reviews. Minutes shall be distributed to all RSAC members, all administrative levels, and the Radiation Safety Officer within 2 months after each meeting.

6.5 ADorovals The procedure for obtaining approval for any change, modification, or procedure which requires approval of the Reactor Safeguards Advisory Committee shall be as follows:

a. The Chief Reactor Supervisor shall prepare the proposal for review and approval by the Reactor Administrator.

b. The Reactor Administrator shall submit the proposal to the Reactor Safeguards Advisory Committee for review and comment,

c. The Reactor Safeguards Advisory Committee can approve the proposal by majority vote.

d. The Reactor Administrator can provide final approval after '

receiving the approval of the Reactor Safeguards Advisory Committee.

6.6 Procedures There shall be written procedures that cover the following activities:

a. Startup, operation, and shutdown of the reactor.

b. Fuel movement and changes to the core and experiments that could affect reactivity,

c. Conduct of irradiations and experiments that could affect the operation or safety of the, reactor.

d. Preventive or corrective maintenance which could affect-the safety of the reactor.

e. Routine reactor maintenance.

f. Radiation Safety Protection for all reactor related personnel.

g. Surveillance, testing and calibration of instruments, components, and systems as specified in Section 4.0 of these Technical Specifications.

h. Implementation of the Security Plan and Emergency Plan.

The above listed procedures shall be approved by the Reactor Administrator 22

and the Reactor Safeguards Advisory Committee. Temporary procedures which do not change the intent of previously approved procedures and which do not involve safety question may be employed on approval by the Chief Reactor Supervisor.

6.7 Experiments

a. Prior to initiating any unapproved reactor experiment, an experimental procedure shall be prepared by the Chief Reactor Supervisor and reviewed and approved by the Reactor Safeguards Advisory Committee.

b. Approved experiments shall only be performed under the cognizance of the Chief Reactor Supervisor.

6.8 Safety Limit Violation The following actions shall be taken in the event a Safety Limit is virlated:

a. The reactor will be shut down immediately and reactor operation will not be resumed without authorization by the Nuclear Regulatory Commission (NRC).

b. The Safety Limit Violation shall be reported to NRC Region IV Office of Inspection and Enforcement, the Director of NRR, the Reactor Safeguards Advisory Committee, and Reactor Administrator not later than the next work day.

c. A Safety Limit Violation Report shall be prepared for review by the Reactor Safeguards Advisory Committee. This report shall describe the applicable circumstances preceding the violation, the effects of the violation upon facility components, systems, or structures, and corrective action to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the NRC, and the Reactor Safeguards Advisory Committee within 14 days of the violation.

6.9 Reportina Reauirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director, Office of Nuclear Reactor Regulation, USNRC, Washington D.C., 20555, Attention: Document Control Desk; with a copy to Regional Administrator, Region IV.

6.9.1 Annual Operatina Report Routine annual operating reports shall be submitted no later than ninety (90) days following June 30. The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience having safety significance that was gained 23

during the year, even though some repetition of previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear.

Each annual operating report shall include:

1. A brief narrative summary of

a. Changes in facility design, performance characteristics, and operating procedures related to reactor safety that occurred during the reporting period.

b. Results of major surveillance tests and inspections.

2. A tabulation showing the hours the reactor was operated and the energy produced by the reactor in watt-hours.

3. List of the unscheduled shutdowns, including the reasons therefore and corrective action taken, if any.

4. Discussion of the major safety related corrective maintenance performed during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for the corrective maintenance required.

5. A brief description of:

a. Each change to the facility to the extent that it changes a description of the facility in the application for license and amendments thereto.

b. Changes to the procedures as described in Facility Technical Specifications,

c. Any new or untried experiments or tests performed during the reporting period.

6. A summary of the safety evaluation made for each change, test, or experiment not submitted for NRC approval pursuant to 10 CFR 50, paragraph 50.59 which clearly shows the reason leading to the conclusion that no unreviewed safety question existed and that no technical specification change was required.

7. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge.

a. Licuid Waste (summarized for each release)

1. Total estimated quantity of radioactivity released (in Curies) and Total volume (in liters) of effluent water (including diluent) released.

24 i

b. Solid Waste (summarized for each release)

1. Total amount of solid waste packaged (in cubic meters)

2. Total activity in solid waste (in Curies)

3. The dates of shipments and disposition (if shipped off site).

8. A description of the results of any environmental radiological surveys performed outside the facility.

9. Radiation Exposure - A summary of radiation exposures greater than 100 mrem (50 mrem for persons under 18 years of age) received during the reporting period by facility personnel or visitors.

6.9.2 Reportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actions and measure to prevent recurrence, shall be reported to the NRC as described in Section 6.9.1. Supplemental reports may be required to fully describe final resolution of the occurrence. In case of corrected or supplemental reports, an amended licensee event report shall be completed and reference shall be made to the original report date,

a. Promot Notification with Written Followuo The types of events listed below are considered reportable occurences and shall be reported as expeditiously as possible by telephone and confirmed.by telegraph, mailgram, or facsimile transmission to the Director of NRC Region IV. Office, or his designated representative no later than the first work day following the event, with a written followup report within two weeks as described in Section 6.9.1. Information provided shall contain narrative material to provide complete explanation of the circumstances surrounding the event.

1. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reached the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.

2. Operation of the reactor or affected systems when any parameter or operation subject to a limiting condition is less conservative than the limiting condition for operation established in the technical specifications j without taking permitted remedial action.

25

3. Abnormal degradation discovered in a fission product barrier.

4. Reactivity balance anomalies involving:

a. Disagreement between expected and actual critical rod positions of approximately 0.3% A k/k.

b. Exceeding excess reactivity limit.

c. Shutdown margin less conservative than specified

, in technical specifications.

I

d. .If sub-critical,<an unplanned reactivity insertion of more than approximately 0.5% k/k or any unplanned criticality.

5.- Failure or malfunction of one (or more) component (s) which prevents or could prevent, by itself, the i

fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the Safety Analysis Report.

'6. Personnel error or procedural inadequacy which prevents,

could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with l accidents analyzed in the Safety Analysis Report.

7. Unscheduled conditions arising from natural or man-made

events that, as a direct result of the event, require i reactor shutdown, operation of safety systems, or other ,

. protective measures required by Technical Specifications.

8. Errors discovered in the transient or accident analyses or

, in the methods used for such analyses as described ~in the Safety Analysis Report or in the bases for the Technical Specifications that have or could have permitted reactor j operation in a manner less conservative than assumed in'

! the analyses.-

4

9. Release of radiation from site above allowed limits.

10. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analysis in the Safety Analysis Report or Technical Specifications that require remedial action or a corrective measures to-prevent the existence or development of an unsafe condition.

I 4

26

9 6.9.3 Special Reportq Special reports which may be required by the Nuclear Regulatory ,

commission shall be submitted to the Director of the NRC Region IV Office within the time period specified for each report. This includes changes in level 1, 2 or 3 administration which must be reported within 30 days of such a change.

6.10 Record Retention 6.10.1 Records to be Retained for a Period of at Least Five Years

a. Operating logs or data which shall identify:

1. Completion of pre-startup checkout, startup, power changes, and shutdown of the reactor.

2. Installation or removal of fuel elements, control rods, or experiments that could affect core reactivity.

3. Installation or removal of jumpers, special tags or notices, or other temporary changes to reactor safety circuitry.

4. Rod worth measurements and other reactivity measurements,

b. Principal maintenance operations.

c. Reportable occurrences.

d. Surveillance activities required by technical specifications.

e. Facility radiation and contamination surveys.

f. Experiments performed with the reactor.

This requirement may be satisfied by the normal operations log book plus,

1. Records of radioactive material transferred from the facility as required by license.

2. Records required by the Reactor Safeguards Advisory Committee for the performance of new or.special experiments.

g. Records of training and qualification for members of the facility staff.

h. Changes to operating procedures.

27

6.10.2 Records to be Retained for the Life of the Facility

a. Records of. liquid and solid radioactive effluents released to the environs.

b. Appropriate off-site environmental monitoring surveys.

c. Fuel inventories and fuel transfers,

d. Radiation expocures for all personnel.

e. Drawings of the facility.

f. Records of transient or operational cycles for those components designed for a limited number of transients or cycles.

g. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

h. Records of meetings of the Reactor Safeguards Advisory Board, and copies of RSAC audit reports.

28 r -r yr - m- w