ML20214R950

Technical Evaluation Rept Re Renewal of OL for Univ of New Mexico, Informal Rept

ML20214R950
Person / Time
Site:	University of New Mexico
Issue date:	03/31/1987
From:	Carolyn Cooper EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20214R940	List: ML20214R937 ML20214R950
References
CON-FIN-D-6010 EGG-NTA-7467, NUDOCS 8706090066
Download: ML20214R950 (42)

Text

- _ _ _ _ _ _ _ _ _ _ _ _ .

EGG-NTA-7467 March 1987 l

l INFORMAL REPORT

?. - i l

t

I

/daho TECHNICAL EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE i National UNIVERSITY OF NEW MEXICO Engineering i Laboratory 1

Managed C. H. Cooper by the U.S.

Department ofEnergy I

f4" EGRGio.a. Preoared for the w, y,,_,,,,

U. S. NllCLEAR REGULATORY COMMISSION l DOE Contract No. DE-AC07 76lD01570 8706090066 870310 PDR ADOCK 05000252 P PDR

e .

DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal hability or responsability for the accuracy, completeness, or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infnnge pnvately owned nghts. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or impiv its endorsement, recommendation, or favonng by the United States Govemment or any ageccy thereof. The views and opinions of authors expressed herein do not necessanly state or reflect those of the United States Government or any agency thereof.

O O

-- . , _ . . o

EGG-NTA-7467 TECHNICAL EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY OF NEW MEXICO Docket No. 50-252 C. H. Cooper March 1987 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415

, Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. D6010

i-A8STRACT This Technical Evaluation Report for the review of the application filed by the University of New Mexico (UNM) for renewal of Operating License No. R-102 to continue to operate its research reactor has been i prepared for the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Comission. The facility is located on the campus of the University of New Mexico in Albuquerque, New Mexico. The INEL concludes that the reactor can continue to be operated by the University of New Mexico without endangering the health and safety of the public.

I 1

I 1

i s

FIN No. 06010 Casework and Non-Power Reactor Reviews I

i 11

'l i

i FOREWORD i

This report is supplied as part of the Evaluation of Application I

Assistance for the Non-power Reactors Program being conducted by the Idaho National Engineering Laboratory for the U.S. Nuclear Regulatory Comission,

]

Office of Nuclear Reactor Regulation.

l The U.S. Nuclear Regulatory Comission funded the work under the authorization of FIN No. 06010.

I' i

l PREFACE ,

This report summarizes the safety review performed by the NRC Technical Assistance Division, Idaho National Engineering Laboratory, of q the technical portions of the license renewal application submitted to the a U.S. Nuclear Regulatory Commission (NRC) from the University of New Mexico for the continued operation of their research reactor. As this document is i

to be the basis of select sections in a formal Safety Evaluation Report j (SER) to be published by the NRC before final licensing action, the section numbers in this report correspond to their appropriate positions in the final SER.

The NRC is responsible for writing the following sections of the SER.

i

1. Introduction

2. Site characteristics ,

3. Design of structure, systems, and components

13. Conduct of operations

15. Technical specifications

16. financial qualifications

17. Other license considerations

18. Conclusions.

l Thus, this report consists of Sections 4 through 12 and 14, plus the ,

4 references applicable to these sections.

i 1

4 iv

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

f i

I CONTENTS A8STRACT .............................................................. 11 FOREWORD .............................................................. 11)

. PREFACE ............................................................... iv

, 4. REACTOR .......................................................... 1 4.1 Reactor Core ............................................... 1 4.2 Moderator .................................................. 1 4.3 Reflector and Shielding .................................... 6 4.4 Safety and Control Rods .................................... 7 4.5 Physics and Reactivity Control ............................. 9 d

4.5.1 Excess Reactivity anc Shutdown Margin .............. 10 4.5.2 Conclusion ......................................... 10 4.6 Operating Procedures ....................................... 11 4.7 Conclusions ................................................ 11

5. REACTOR COOLANT AND ASSOCIATED SYSTEMS ........................... 12

6. ENGINEERED SAFETY FEATURES ....................................... 13 I

, 7. CONTROL AND INSTRUMENTATION SYSTEMS .............................. 14 l \

7.1 Systems Summary ............................................ 14 ,

4 l 7.2 Nuclear Control System ..................................... 14  :

7.3 Instrumentation System...................................... 15 l

! 7.3.1 Nuclear Instrumentation ............................ 15  !

7.3.2 Process Instrumentation ............................ 16

! 7.4 Conclusions ................................................ 16

8. ELECTRIC POWER ................................................... 18

,, 8.1 Offsite Power .............................................. 18 9

8.2 Emergency Power ............................................ 18 l 8.3 Conclusion ................................................. 19 i

e v

I

- - - - - .- . = ._ ._-

9. AUXILIARY SYSTEMS ................................................ 20 9.1 Fuel-Handling and Storage Systems .......................... 20 9.2 Fire Protection System ..................................... 20 9.3 Communications System ...................................... 20 9.4 Ventilation System ......................................... 21 9.5 Conclusion ................................................. 21 -

10. EXPERIMENTAL PROGRAMS AND FACILITIES ............................. 22 10.1 Experimental Programs and Reviews .......................... 22 10.2 Experimental Facilities .................................... 23 10.3 Conclusion ................................................. 23 l 11. RADI0 ACTIVE WASTE MANAGEMENT ..................................... 24 i 11.1 Waste Generation and Management ............................ 24 11.2 Conclusions ................................................ 24

12. RADIATION PROTECTION PROGRAM ..................................... 25 12.1 A L A R A C o nn i t me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 12.2 Health Physics Program ..................................... 25 12.3 Radiation Source ........................................... 27 12.4 Routine Monitoring ......................................... 28 12.5 Personnel Monitoring ....................................... 28 t

, 12.6 Potential Dose Assessments ................................. 29 (

i  !

12.7 Conclusion ................................................. 29  !

14. ACCIDENT ANALYSIS ................................................ 30 14.1 Maximum Hypothetical Accident .............................. 30 ,

14.2 Operator Error ............................................. 31 i .

1 14.3 Conclusion ................................................. 32 i

i i 19. REFERENCES ....................................................... 33 1

1 i

.j' vi i

J

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

TABLES 1

2 4.1. Principal Design Parameters ......................................

1 7.1 Scram-Producing Safety Channels .................................. 17 i

FIGURES i

i l 4.1 AGN model 201M reactor ........................................... 3 i

4.2 Schematic of the reactor (looking from above) .................... 4 4.3 AGN-201M core tank and contents .................................. 5 i

4.4 Control rod ...................................................... 8 1

1

i j

i I

l e

l

)

i j -e t

k d

f, vil i

1 l

4. REACTOR The reactor, located on the University of New Mexico (UNM) campus, is a small homogenous thermal reactor (AGN-201M) manufactured by Aerojet General Nucleonics, which is regularly used for operations training and student laboratory experiments at powers less than 5 W. The reactor is comprised of an enriched U0 c re, p lyethylene moderator, graphite 2

reflector, lead and water shielding, and safety and control rods, all located inside a stainless steel tank. The reactor power is controlled by inserting and withdrawing control rods that contain enriched uranium. An overall view of the reactor is given in figures 4.1 and 4.2 and the principal design parameters are listed in Table 4.1.

4.1 Reactor Core The cylindrical reactor core (Figure 4.3), which is 25.6 cm in dia, by 24 cm high, consists of nine separate polyethylene discs that contain particles of (U0 )2enriched to 20% U-235. The core is contained in a gas-tight aluminum cylindrical tank (32.2 cm diameter and 76 cm high). A 2.54 cm (1-in.) dia, tube (glory hole) passes through the center of the core to allow experimental access. The total fuel loading is 667 g of U-235. A basic safety design feature of the reactor core is the core polystyrene fuse that supports the bottom three fuel discs of the core.

The fuse contains a fuel density twice that of the fuel discs so that during operation, more fission heat is generated in the fuse than in the remainder of the core. Thus, any reactivity transient would heat the fuse twice as fast as the core. Should the temperature of the fuse reach 100*C, the polystyrene fuse will melt allowing the three bottom discs to separate from the remaining discs and thus shut the reactor down neutronically.

4.2 Moderator

~

, Neutron moderation is performed by the polyethylene in which the UO 2

is homogeneously dispersed. This moderator material is a radiation- '

l' stabilized polyethylene that has a lifetime of 100 W-y.

1 j

TABLE 4.1. PRINCIPAL DESIGN PARAMETERS Reactor type AGN-201M Maximum licensed power level 5W Fuel element design .

Fuel-moderator material UO2 -Polyethelyene Uranium content 6 wt%

Uranium enrichment <20% 2350 Shape disk Thickness of fuel varying from 1 to 4 cm Diameter of fuel 25.6 cm Number of fuel disks 9 Source type 2 C1 P u 8, Excess reactivity, maximum 0.65% Ak/k Excess reactivity with glory hole empty 0.25% ak/k Number of control rods 4 Coarse control rod 1 Fine control rod 1 Safety rods 2 Total reactivity worth of rods 4% Ak/k

! Reactor cooling Natural convection of pool water S effective 0.0075 o

4 2

i

f

~

1.98 meters MANHOLE AE C M TANK H

OR WATER TANK p

pf7 10 cm , . b h g. ..7 CORE )

LEAD SHIELD ,

5 k$ 5 2 cm 2.5 cm diam. , ,

GLORY HOLE g g4

,_ j 20 cm GRAPHITE 10 cm diam. - >

y-g y REFLECTOR i ACCESS PORTS i M y g

,] \ g g CONTROL RODS l

~

2.89 meters BASE CONTROLLED

~

ACCESS AREA

. u i

~

i i

i l Figure 4.1 AGN-201M research reactor l

i

.- __ - - _ _ _ _ . _ _ _ - _ . . - =_ - _ _

i

\

EAD g WATER TANK I

WATER REACTOR TANK GRAPHITE i CORE TANK

^

s 4s .

s

'Q X ^ .K ACCESS PORTS 4

x h '/

,-( ws -

COHE ..

2 IONIZATION CHAMBER (linear) (ch.3)

IONIZATION CHAMBER FISSION CHAMBER (log) (ch. 2) (ch.1)

MANHOLE COVER GLORY HOLE Figure 4.2 Schematic of the reactor (looking from above)

. e e e

n _ __ -.= --. . - . . . . . . . . -pCORE TANK COVER IS~ $

l - '., '$ ' ,' ' , ** ,' ' , ' , ' .

h //,< GRAPHITE REFLECTOR

• '%, .',,;': ,'.,0-l.  :. '/l  ;, ,,',/ //,,,,<

'a, g ,i,'// ,// / ,. '

/

' 'dQ

..-:.-9 l'l*:?.':::::'i$::';::'7,s',/ , ' . ,g:/,:/

FUEL C' l .

.-l.-lIf f

M u

f Immys*.;?% **'*teewar *JKM4 so A

,~.-

s g3 mw,x mem smexwv  ;:

GLORY HOLE

, W "tlI I IP'#'E4y/44 i P4X j . ,

] 4 l16 g_ af; - 7,' ~ ' '  % x

{,,p ,

-THERMAL FUSE i '

AN [M l ,Y $ W% CORE TANK 4

.' , s N...

b',,%).'

.g.

N M

)g a

h{ -

g

,' ( i I -

' N CONTROL & SAFETY l, N [

g ,

ROD THIMBLES L N

%k.

b '

s

.< CORE SUPPORT ASSEMBLY a '9 .

% d r- -

• .W r,;;, y-u l

M-CORE TANK COVER l

l i I i

i i

l I

Figure 4.3 AGN-20lM core tank and contents 5

l i

_ - . - . . . - , _ - _ , _ _ _ _ _ . _ _ _ _____.,__.__ _ __._ -_ _ _ ._...,___.._. .-. -.___ _ ..-.,.,. _ __,--- - - . ~ . - . . ~ . . .

Because this reactor does not operate above 5 W or on a continuous basis, the actual core life is expected to significantly exceed the duration of the license.

4.3 Reflector and Shielding 4.3.1 Reflector .

As shown in Figure 4.3, the reflector consists of graphite surrounding the core. It is 20 cm thick with a density of 1.75 g/cm . Part of the graphite is in the aluminum core tank and part is outside. For experimental purposes, there are four 10-cm dia. access holes that pass through the graphite to the exterior of the water tank.

4.3.2 Shielding The shielding of radiation and neutrons from the reactor core is accomplished by three materials: Icad, water, and concrete. The graphite reflector is surrounded by a 10-cm thick lead shield. The lead shielding, graphite reflector, and core are enclosed and supported by a thick steel reactor tank (47.5 cm radius). The stainless steel reactor (Figure 4.2) tank acts as secondary containment for the core tank and is fluid tight.

The removable thermal column tank located above the core (Figure 4.1) is provided to permit access to the core tank. The thermal column tank is normally filled with water to provide biological shielding, but can be filled with graphite if a thermal column is desired.

The water tank provides effective neutron shielding during reactor operation and is the third and ou'ermost of the fluid-tight containers. It is 198 cm in diameter, made of stainless steel, and holds 3785L (1000 gal) ,

of water to form the fast neutron shleid.

6

Finally, there is a 60-cm thick concrete shield on the front side of the water tank in the direction of the control console with 40 cm thickness on the other sides. There is no concrete shielding on the top of the water tank.

4.4 Safety and Control Rods The reactor uses two safety rods, one coarse control rod, and one fine control rod to control reactivity during the conduct of reactor operations. Criticality can only be achieved with addition of tiio fuel contained in the safety and control rods. Reactivity increases as these rods are inserted because they contain U0 2 dispersed in polyethylsne.

The safety rods and the coarse control rod are magnetically coupled to a carriage and compress a spring as they are driven into the core. Thus, the removal of the electromagnet current deactivates the magnet and withdraws the rod by gravity, with an assist from the compressed spring.

Figure 4.4 shows the control rod mechanism. A scram signal doenergizes the magnets on the safety and coarse control rods so that they fall by gravity, assisted by compressed springs, to a full-out safe position. The fine control rod must must be driven out of the reactor, because it is mechanically connected to its carriage.

Safety Rods Each of the two safety rods is 5 cm in diameter and contains 14.5 g of U-235 dispersed in polyethylene with an active length of 15 cm. The active fuel is doubly encapsulated in aluminum containers, which isolates the fuel in the rods from the core. The total travel length of the safety rods is 24 cm and the full length insertion time is 40 to 50 s. The scram removal time is approximately 200 ms for full removal of the rods. The reactivity l warth is 1.25% Ak/k per rod.

1 1 l l

1

)-

e I I - ,

L_ C ,_.,

-- b _

p CORE l l

~

l l g

I l

/,gCAPSULE l

' I

! H e l

I i l I i i i l __J L T-6-

l I BEARING 4 MAGNET h

I LEAD SCREW

", Mi l

I l

1 DASH POT l

!,/

CHAlN ORIVE

> -MOTOR

'v'g:um. /

SYNCHROS mr

=ww Figure 4.4 Control rod

Coarse Control Rod The coarse control rod is 5 cm in diameter and has a 24 cm travel length. The active length is 15 cm and the rod can be inserted at high

(~1/2 cm/s) or low (~1/4 cm/s) speed. It contains 14.5 g U-235 dispersed in polyethylene and the fuel is doubly encapsulated in aluminum.

Insertion or withdrawal of the coarse control rod is normally performed at the high speed and takes 40 to 50 s. Scram time is approximately 200 ms.

The reactivity worth is 1.25% ok/k.

The rod carrtages are mechanically driven to full-out pos1tton following a reactor scram.

fine Control Rod The fine control rod is 2.5 cm in diameter and contains 2.71 g of U-235 in dispersed in polyethylene. The fine control rod fuel is doubly encapsulated and mechanically coupled to the carriage. The fine control rod can be inserted a distance of 24 cm at high or low speed. Normal I

operation of the fine control rod is at high speed, which results in an insertion / withdrawal time of 40 to 50 s. The slow insertion rate (1/4 cm/s) is one-half the fast insertion rate (1/2 cm/s). The fine control rod cannot magnetically decouple, but is driven out at the fast l withdrawal rate upon a scram. Total reactiv1ty warth of the fine rod 1s t 0.25% ok/k.

4.5 Physics and Reactivity Control l

The operation of the AGN-201H is accompitshed by manipulating control rods in response to observed changes in measured reactor parameters such as neutron fluu (reactor power). Interlocks prevent inadvertent reactivity additions and a scram system initiates rapid, automatic shutdown when trip set points are reached. Because the U0 2 is dispersed in the polyethylene, the reactor exhibits a strong negative reactivity feedback 9

(temperature coefftetent of reactivity is -2.5 x 10'*/'C) due to rapid core expansion. This inherent nuclear control feature enhances stability and safety and is ef fective even if control rods or the safety instrumentation should fati to perform their intended functions.

4.5.1 Excess Reactivity and Shutdown Parain The Technical Specifications limit the maximum excess reactivity of the reactor to 0.65% ak/k, and excess reactivity with no experiments in the reactor and the control and safety rods fully inserted to 0.25% ak/k. These Ilmits are well below the value necessary to go prompt critical. The shutdown margin with the most reactive safety or control rod fully inserted is at least 1% ak/k and the reactivity addition rate for each control or safety rod cannot exceed 0.065% Ak/k/s.

l The limitations on core excess reactivity ensure that the reactor would not go prompt critical and the reactor periods would be suffletently long so that the reactor protection system and/or the operator would be able to shut the reactor down before any safety limit could be reached.

The shutdown margin and control and safety rod reactivity limitations ensure subertticality, even if the rod of highest reactivity worth fatis to scram and remains in the reactor.

4.5.2 Conclusion The Idaho National Engineering Laboratory (INEL) concludes that the limitations on total core excess reactivity and reactivity insertion rates ensure reactor periods of sufficient magnitude that the reactor protection system will be able to shut the reactor down before the any significant core temperature would be reached. The shutdown margin and control and ,

safety rod reactivity limitations are sufficient to ensure subcriticality even if the highest worth rod fatis to scram. .

10

In addition to the electromechanical safety controls for normal and off-normal operation, the negative temperature coefficient provides an inherent backup safety feature.

In accordance with the above and the details presented in Section 7, the INEL concludes that the reactivity control systems of the AGN-20lM are designed to function adequately to ensure safe operation and safe shutdown of the reactor under all normal and off-normal operating conditions.

4.6 Operatino procedures The University of New Mexico has implemented a preventive maintenance program that is supplemented by a detailed preoperational checklist to ensure that the reactor is not operated at power unless the appropriate safety-related components are operable. The reactor is operated by NRC-Itcensed personnel in accordance with explicit operating procedures ,

prepared by the Reactor Operations Committee, which include specified responses to any reactor control signal. Before installation into the reactor, all proposed experiments are reviewed by the Reactor Safeguards Advisory Committee for potential effects on the reactivity of the core, damage to the reactor, and possible effects on the health and safety of the staff, students, and the general public.

4.7 Conclusions The INEL review of the University of New Mexico AGN-201M has included the study of its specific design and installation, controls and safety instrumentation, and operational limitations as identified in its Technical Specifications. On the basis of its review of the AGN-201M, the INEL concludes that the AGN-201M is designed to good industrial practices and there is reasonable assurance that the AGN-201M is capable of continued safe operation as limited by its Technical Specifications.

I 11

l l

1. 4 l 1

} 5. REACTOR COOLANT AND ASSOCIATED SYSTEMS 1

The AGN-201M reactor operates at very low power for short periods of time and for this reason does not require an active coolant system.

However, the reactor core is inside a steel cylindrical tank 1.98 m in "

diameter which contains 3758L (1000 gal) of water (for fast neutron

, shielding) and any heat rejection would be accomplished by natural ,

I convection. When the reactor is operated under maximum steady-state conditions at S W, the bulk pool temperature would rise a negligible amount before achieving thermal equilibrium.

The INEL concludes that the design of the coolant system in i

conjunction with the S W power level ensures adequate cooling capability j for the AGN-210M when operated within its licensed power limits and for its fission products' decay heat.

.l.

r 1

4 i

i

(

i i

4 I

1

)

i l 12 i

_ . - - . - ~ . _ . - ____,,m____._._,__.. ___....-m_ .-...- . ~ . _ _ _ .....- . - - - . - ,- -- _ _ ~ --

1

6. ENGINEERED SAFETY FEATURES Engineered safety features (ESFs) are systems provided to mitigate the radiological consequences of design-basis accidents. Because the University of New Mexico reactor operates at a maximum power level of 5 W, the fission product inventory is very low. In addition, the analyses of accidents in Section 14; including the maximum hypothetical accident, indicate that there will be no significant radiological releases.

Therefore, no ESF systems are provided at the University of New Mexico facility.

The INEL concludes that operating the University of New Mexico reactor without any ESF systems does not pose a radiological hazard to the public or to the environment in the event of an accident.

L

}

i i

1 l

l 13

.I

.g . - . .. - - -- - - - - -

T i

7. CONTROL AND INSTRUMENTATION SYSTEMS 7.1 Systems Summary The nuclear control and instrumentation systems for the AGN-201M reactor are similar to those generally used in other research reactors of a similar size in the United States. Control of the nuclear fission process ,

3 is achieved by using coarse and fine control rods and two safety rods. The

.c ,

control and instrumenta' tion systems are interlocked to provide automatic and manual scram capability in case of reactor malfunction and to provide the means for operating the various components of the reactor in a manner consis tent with design objectives. The scram-producing safety channels, functions, and set points are shown in Table 7.1. The licensee's Technical Specifications require that the reactor safety channels shall be operable in accordance with Table 7.1, whenever the reactor control or safety rods are not in their fully withdrawn position.

7.2 Nuclear Control System i,

The nuclear control system is composed of the nuclear equipment designed for operation in case of failure or malfunction of components essential to the safe operation of the reactor.

4 A detailed description of the nuclear control system that consists of the safety rods, fine and coarse control rods, and their associated drive mechanism'5\1s presented in Section 4.

The safety rods and the coarse control rod are interlocked so that (a) only one safety rod can be inserted at a time, (b) the coarse control V.

'l rod cannot be inserted unless both safety rods are fully inserted, and (c) reactor startup cannot commence unless both safety rods and the coarse control rod are fully withdrawn from the core.

. \1 s 14

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

g

The rods are controlled by manual holddown (spring-return) switches located on the control console. The positions of the control rods are indicated to the nearest 0.01 cm on the console. The location of any one of the three magnets at either its upper or lower position is indicated by an appropriate light on the control console. Contact between the magnets and their associated actuator rods also is indicated by control console lights.

7.3 Instrumentation System J

The instrumentation system is composed of both nuclear and process instrumentation circuits.

2 7.3.1 Nuclear Instrumentation The instrumentation discussed below provides the operator with the necessary information to properly manipulate the nuclear controls.

The source-range channel (Nuclear Safety Channel 1) uses a gas-filled U-235 fission chamber to monitor reactor startups. The high voltage on the chamber is automatically switched off when the signal from Nuclear Safety Channel 2 exceed 10' A.

The log power and period channel (Nuclear Safety Channel 2) comprises a compensated ion chamber, power supply, logarithmic picoammeter, recorder, period signal, and scram. This channel covers the power range from source level to full power and will produce a scram if the power level is greater than 10 W. This channel also will cause a scram if the reactor period is less than 5 s.

The linear power channel (Nuclear Safety Channel 3) comprises a compensated ion chamber, power supply, linear picoammeter with a range.

switch, recorder, and meter. This channel has scrams at 5 and 95% of indicated range (as well as 2X licensed full power).

15

Both compensated ion chambers are located in watertight cans submerged in a water tank above and to the side of the core. The appropriate controls and meters and recorders are located on the control console. A manual (operator-controlled) scram also is located on the control console.

7.3.2 Process Instrumentation The instrumentation discussed below senses and monitors nonnuclear parameters associated with the pool water and provides, as appropriate, startup prohibits or scrams.

The pool water level monitor consists of a float switch and the associated circuit. This monitor provides an audio and visual alarm at the control console and initiates a scram if the pool water level drops to 17.8 cm (7 in.) below the reactor tank top.

The tank water temperature monitor consists of a resistance bulb thermometer that senses the bulk pool temperature. Temperature indication is provided on the control console. A starup prohibitor scram is initiated if the bulk pool temperature falls below 18'C.

7.4 Conclusions Based on review of drawings, reports, and a site visit, the INEL concludes that the control and instrumentation system at the University of New Mexico research reactor facility is well designed and maintained.

Redundancy in the crucial areas of power measurements is ensured by overlapping ranges of the log power and linear power channels. The control system is designed so that the reactor shuts down automatically if electric power is lost. In addition, the procedus for reactor room, pool water, and personnel are adequate for the proposed reactor operating conditions.

On the basis of the above analysis of the control and instrumentation systems, the INEL concludes that both systems are adequate to ensure the safe operation of the facility. l 16

l TABLE 7.1. SCRAM-PRODUCING SAFETY CHANNELS Device Function Set points Channel 1 Monitor startup flux None Nuclear Safety Channel 2 High power $10 W Nuclear Safety Channel 3 High power 10W 1

Reactor tank water level Protect shielding 7 in, below reactor interlock capability tank top Pool water temperature Limit reactivity 18'C or less interlock addition Seismic displacement interlock Seismic protection In place Manual scram Normal shutdown Scram on operator decision e

17 l

s

8. ELECTRIC POWER l

8.1 Offsite Power During operation, the electric power requirements for the University

~

of New Mexico AGN-20lM are supplied by the Public Service Company of New <

l Mexico, which services the University. The reactor facility requires 110 V ,

ac power.

Electrical power is supplied to the plug molds in the reactor room via a circuit breaker located in the main distribution panel located just inside the Nuclear Engineering Building 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 for that plug mold and in turn remove power from the reactor panel.

8.2 Emeroency Power t

4 No emergency power is provided for the University of New Mexico

. AGN-20lM reactor operation. In the event of electric power failure, the control system is designed to be fall-safe and scram the reactor. In addition to this feature, an emergency light powered by batteries is installed in the console and operates automatically.if the power is cut off. Hand-held battery powered radiation monitors are available at the l console. The terminology " fail-safe", as far as power failure is concerned

! means that during critical operation, energy being drawn from the power

system is used to maintain the coarse control rod and_two safety rods in the reactor by means of electromagnets. Loss of power deenergizes the magnets and the rods are accelerated downward by means of springs and gravity to their safe, stable positions out of the core.

i c

4 l

5 l

18

.--_ , _ - - - - ~

8.3 Conclusion The electrical power system at the University of New Mexico is a standard electrical supply system designed and constructed to specifications similar to those at other low-power research reactor facilities. This, coupled with the fact that the reactor will scram in the event of a power failure, supports the INEL conclusion that the electrical power system is acceptable for continued safe operation of the University of New Mexico AGN-201M reactor.

e i

19

9. AUXILIARY SYSTEMS 1 l

9.1 Fuel-Handlina and Storace Systems j Because of the extremely low reactor power (5 W), periodic fuel replacement for the AGN-201M is unnecessary. The fuel is normally in a sealed core tank except for one laboratory experiment (approach to ,

critical). A clean area is set up on the reactor top and the tank is vented and pressure equalized by drawing an air sample through a cloth and charcoal filter. A clean area is set up to receive the top five fuel disks when they are removed from the core tank and placed on the reactor room bench. This experiment is done under supervision of the UNM Radiation Safety Organization. Protective clothing is worn while handling fuel.

The only other fuel for the reactor is an additional fuel disk, which is stored in a storage cabinet in the reactor room or in a source storage room with a k,ff <0.9.

9.2 Fire Protection System The components of the AGN-20lM reactor are basically nonflammable, as is the building in which the reactor is located, and in the event of fire no special precautions would be required. A carbon dioxide fire extinguisher is located in the reactor room. In case of a fire, the reactor would be shut down and locked and the supervisor or his alternate would be notified. Normal fire procedures that are in place are expected to eliminate accumulation of flammable material in the building to reduce the probability of a fire.

9.3 Communications System The reactor room is serviced by the University phone system, which .

allows communication to and from outside sources.

4 20 i

f 9.4 Ventilation System The ventilation system is composed of an air-handling system designed for the Nuclear Laboratory Building. The system is designed to provide a negative pressure in the building so that all exit air passes through a set of high efficiency particulate air filters and out the ventilation stack located on the roof. Outside air is supplied through intake filters to all rooms in the Nuclear Engineering Laboratory and a relief blower takes suction on all the rooms and transfers the air to the stack.

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 the negative pressure in the building. In an emergency, motorized louvers can be opened to the lab to increase the amount of air being exhausted. The switches for the exhaust fans are located near the east entrance to allow the laboratory to have control of the fans.

9.5 Conclusion The INEL concludes that the auxiliary systems at the University of New Mexico reactor facility are designed and maintained adequately and are capable of performing their intended functions to help ensure the safe operation of the facility.

21

1

10. EXPERIMENTAL PROGRAMS AND FACILITIES 10.1 Experimental Programs And Reviews The University of New Mexico AGN-20lM reactor provides support to the ,

t nuclear engineering undergraduate and graduate programs. Various experimental programs of the staff and students involve use of the . !

reactor. Most of the experimental work involves activation of various materials and their subsequent analyses. These irradiated materials may be foils or small samples used to evaluate reactor parameters or material composition (neutron activation analysis), or used as tracers in various studies. Materials activated in the reactor are short half-life nuclides.

These materials are surveyed for activity and contamination before use in the laboratory.

All proposed experiments are reviewed by the University's Radiation Safety Officer (RS0) and Reactor Safeguards Advisory Committee. The

, reviews by the RSO and the Reactor Safeguards and Advisory Committee are performed to:

1. Ensure that accidents causing changes in composition and geometry

, of the experiments will not cause positive changes or ramps in reactivity that might place the reactor on unsafe periods 1

2. Provide assurance of mechanical integrity, chemical compatibility, and adequate protection against any other potential hazard

3. Ensure any experiments containing materials corrosive to_ reactor components or which contain liquid or gaseous fissionable ,

material are doubly encapsulated 1

22

4. Provide assurance that in the event of an accident, the postulated complete release of all gaseous, particulate, or volatile components from the experiment will not result in doses which exceed 10 CFR 20 limits

5. Ensure that explosive materials are not used.

10.2 Experimental Facilities The horizontal glory hole has diameter of 2.54 cm and goes through the reactor core. Samples may be placed in the glory hole at varying positions in the core and reflector. Samples may also be placed in the access ports that pass through the graphite to the outside of the core.

10.3 Conclusion The INEL concludes that the design of the experimental-facilities, together with the limitations for experiments delineated in the Technical Specifications, and the safety review process ensure proper and safe experimental programs.

l 1

i l

23 i

11. RADI0 ACTIVE WASTE MANAGEMENT 11.1 Waste Generation and Management Because of tr.e low power level and limited operating schedule of the

~

University of New Mexico (UNM) AGN-20lM research reactor, there has been negligible generation or release of radioactive waste either airborne, ,

solid, or liquid.

Materials activated in the glory hole or access ports are short half-life nuclides for student laboratory use. Activated samples are surveyed for activity and contamination before use in the laboratory.

Records of radionuclides produced are documented in the reactor log book.

Transfers of radioactive materials to other licensees are rare and conducted in accordance with appropriate state and federal regulations.

All radionuclides to be removed from the reactor facility are transferred using the UNM radiation permit issued by the New Mexico Environmental Improvenent Division. All such material is then handled in accordance with University guidelines, a copy of which was attached to the Emergency Plan and approved therein on June 11, 1985.

11.2 Conclusions The INEL has reviewed the operational history of the UNM AGN-20lM and concludes that any airborne radioactivity released from operating the reactor at 5 W will be insignificant. The INEL also concludes that the waste management activities of the UNM have been and are expected to continue to be conducted consistent with 10 CFR 20 and ALARA (as low as reasonably achievable) principles.

4 24

i

12. RADIATION PROTECTION PROGRAM The University of New Mexico (UNM) has developed a radiation protection program with adequate staff and monitoring equipment to ensure detection, control, and documentation of occupational exposure.

Health Physics for the AGN-20lM reactor is provided by the University Radiological Safety Office (RS0), which consists of several senior health physicists plus technicians and the full scale of monitoring instrumentation. The radiation monitoring instrumentation is calibrated on 4 a regular basis to ensure that accurate readings'are taken during the >

periodic radiation surveys conducted by the RSO.

12.1 ALARA Commitment The UNM President has instructed the Reactor Operations Committee to formally establish a policy that operations are to be conducted in a manner to keep all radiation exposures ALARA. All proposed experiments and ,

procedures at the reactor are reviewed for ways to minimize the potential exposures of personnel. All unanticipated or unusual reactor-related exposures will be investigated by both the RSO and the operations staff to develop methods to prevent recurrences.

12.2 Health Physics Program Health physics activities at the reactor are performed by the RSO staff who are available for consultation in all matters concerning radiological safety. The RSO staff conducts radiation surveys of the reactor room on a monthly basis with at least two surveys per year j performed while the reactor is in operation.

The INEL believes that the radiation safety support is adequate for the research efforts within this facility.

25

i-o i

! 12.2.1 Procedures f-Written procedures have been prepared that address the radiation safety support that is provided to the operations of the AGN-201M. These procedures identify the interactions between the operational and experimental personnel. They also specify numerous administrative limits and action points, as well as appropriate responses and corrective actions ,

[ if these limits or action points are reached or exceeded. Copies of these procedures are readily available to the operational and research staffs and  !

administrative personnel.  ;

i i 12.2.2 Instrumentation i

The University of New Mexico has a variety of detecting and measuring f instruments available for monitoring potentially hazardous ionizing radiation. The instrument calibration procedures and techniques ensure that radiation of any significant magnitude will be detected promptly and l measured correctly.

Radiation monitoring instrumentation available to the reactor operator

includes a console-mounted meter and portable survey meter. These and other such instruments available within the reactor laboratory are j calibrated periodically by the RSO of the University. There are remote area monitors with automatic alarms installed to monitor the Reactor Room and the building exhaust stack.

12.2.3 Trainina  ;

t 1

All reactor-related personnel are given an indoctrination in radiation f

! safety before they assume their work responsibilities. Additional ,

radiation safety instructions are provided to those who will be working  :

i directly with radiation or radioactive materials. The training program is .

] designed to identify the particular hazards of each specific type of work j to be undertaken and the methods to mitigate their consequences.

I Retraining in radiation safety is also provided. All reactor operators are 26 l

-- , - - , - - ,-,.-.---n,-,n ,,,,,n..., - - - - .,- --, - , - , , , - . - , , , , , -

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

given an examination on health physics practices and procedures at least once every 2 y. The level of retraining given is determined by the examination results. This radiation safety training appears to be appropriate for the facility.

12.3 Radiation Sources 12.3.1 Reactor The only source of radiation directly related to reactor operations is the radiation from the reactor core. Operation at 5 W results in a very small fission product inventory that decays to a low level in a few days.

The level in the polyethelyne is so low that a fuel disk can be handled without shielding 24 h after shutdown. All of the fission products-generated by reactor operation are contained in the polyethylene.

Radiation exposure from the reactor core is reduced to acceptable levels by water, lead, and concrete shield'.ng.

12.3.2 Extraneous Sources

Sources of radiation that may be considered as incidental to normal i reactor operation, but associated with reactor use, are activated foils or samples. Per sonnel exposure to radiation from intentionally produced radioactive material, as well as from the required manipulation of activated experimental components, is controlled by rigidly developed and reviewed operating procedures that use the normal protective measures of time, distance, and shielding.

G 9

4 27

12.4 Routine Monitorina 12.4.1 Fixed-Position Monitor The AGN-201M reactor room has one fixed-position radiation area ,

monitor (RAM) on the reactor room wall. The monitor has an adjustable alarm set point and provides an audible alarm to the operators at the ,

control console if the radiation level exceeds a 10 mR/h set point.

12.4.2 Experimental Support The University of New Mexico RSO participates in experiment planning by reviewing all proposed procedures for methods of minimizing personnel exposures and limiting the generation of radioactive waste. Approved procedures specify the type and degree of radiation safety support required by each activity. As an example, procedures require that changes in experiment configuration include radiation surveys by health physics personnel, and all items removed from the reactor room be surveyed.

12.5 Personnel Monitorina The University of New Mexico personnel monitoring program is described in its Radiation Safety Manual. Personnel exposures are measured by the use of film badges assigned to individuals who might be exposed to radiation. Instrument dose rates and time measurements may be used to administratively keep occupational exposures of other personnel below the applicable limits in 10 CFR 20. The following provides the latest annual summary of doses to reactor related personnel.

Estimated whole body exposure rance (rems) Number of individuals in each rance 1985 1984 1983 1982 1982 .

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 28

12.6 Potential Dose Assessments Natural background radiation levels in the Albuquerque area result in an exposure of about 100 mrems/y to each individual residing there. At least an additional 8% (~8 mrems/y) will be received by those living in a brick or masonry structure. Any medical diagnosis x-ray examinations will add to these natural background radiations, increasing the total cumulative annual exposure.

Exposures from potential airborne or liquid releases from the AGN-201M have been estimated by the INEL and are considered to be negligible.

Conservative estimates indicate radiation exposures from operation of the f reactor at 5 W to individuals in both restricted and unrestricted areas are j well below 10 CFR 20 limits.

12.7 Conclusion i

The INEL concludes that radiation protection receives appropriate l support from the University of New Mexico administration. The INEL further 7

t concludes that (a) the program is staffed and equipped properly, (b) the reactor radiation safety-related staff has adequate authority and lines of

! communication, (c) the procedures are integrated correctly into the ,

research plans, and (d) surveys verify that operations and procedures

achieve ALARA principles.

Additionally, the INEL concludes that the University of New Mexice

radiation protection program is acceptable because there have been no

] instances of reactor-related exposures of personnel above applicable

) guideline values and no significant releases of radioactivity to the environment have been identified. There is reasonable assurance that the personnel and procedures will continue to protect the health and safety of 3 , the public during routine reactor operations.

i 29 i

14. ACCIDENT ANALYSIS l

Two accidents are considered for the AGN-201M reactor. First the 4

maximum hypothetical accident to place an upper bound on radiological

! consequences and second, operator error to demonstrate adequate protection

~

during normal operation.

, 14.1 Maximum Hvoothetical Accident 4

The maximum hypothetical accident (MHA) considered for the AGN-201M reactor is the insertion of fissionable material into the reactor core via i the glory hole. The consequences of the scenario is dependent on the amount of fissionable material inserted and the insertion speed. Argument i is given in the applicant's documentation to justify that a 2% step increase in reactivity will easily encompass any possible fissionable material insertions.

Three assumptions are used as a basis for calculating the power generated in the accident.

1. At time zero, a 2% step increase in reactivity is inserted with

the reactor at 5 W power 1

2. Also at time zero, the energy in the core is negligible compared with the energy liberated during the accident

3. No heat is removed from the core during the excursion.

During the MHA, the reactor would reach 75.0 MW peak power and have a total energy release of 2.41 MJ in approximately 150 ms. The resulting average temperature rise would be 100.7*C, and the temperature at the center of the core would be about 150*C. The total dose to a person ,

i standing next to the reactor during the MHA would be about 1 R. The prediction that the core does not melt except for the fuse is reasonable,

because polyethylene does not melt below 200*C. The fission products are i

30

. . . . - . . - _ - - . , . . , _ . - . , = = - , . _ , . . _ _ . - . . . . . - . . - , _ . . ,_ - .._- -. -.- - -

.__ _ - . .- . =_ _ . - . _ . _ . _ _ _ _ _ _ _ _

l l

contained within the core and primary and secondary containers and the  ;

small amount of gaseous fission products that would be released from the fuel in the fuse when it melted would be small. The power excursion is self-limiting because of core expansion due to the temperature rise. This excursion is strongly dependent on the magnitude of the temperature coefficient of reactivity.

^

The use of polyethylene fuel elements makes the reactor extremely safe. The inherent safety of these elements , even for reactor transient periods as low as 4 to 6 ms, is based primarily upon the arrangement of the fuel within the moderator. The homogeneous dispersal of the fissile material in the polyethylene results in a negligible time delay in transferring the heat from the uranium dioxide particles to the surrounding plastic moderator. The prompt heating of the polyethylene increases the average thermal energy of the neutrons, which results in a decreased fission-to-capture probability for these neutrons. In addition, the prompt expansion of the fuel elements causes a decrease in the density of the moderator, which increases the leakage of the neutrons out of the core.

4 The net result of this prompt heat transfer is that the reactor shuts down-safely.

14.2 Operator Error In general, an operator error occurring during the normal operation of the reactor would be rectified before any unsafe condition could result.

Interlocks ensure that the proper procedure is being followed by the operator during the startup of the reactor. Abnormal conditions caused by human error will automatically shut the reactor down. In addition, scrams i

are initiated by the following events that could be initiated by operator error.

1

1. Exceeding a maximum preset power level

2. Placing the reactor on a period which is less than 5 s

3. Lowering of the shielding water level to less than 7 in.

31

4. Loss of-electrical power

5. Pressing the sensitrol reset button

6. Reaching a minimum preset power level

7. Disconnecting the electrical cables to the safety and control rods ,

8. Not keeping reactor power within 5 and 95% of Channel 3 scale.

14.3 Conclusion Based on the applicants documentation and our independent assessment, the INEL concludes that the maximum hypothetical accident will not melt the polyethylene surrounding the uranium fuel. Consequently, insignificant fission products would be released and the direct radiation would be a small fraction of 10 CFR 20 guidelines. Therefore,-the postulated MHA poses no risk to the health and safety of the public or to the reactor personnel. In addition, the INEL concludes any abnormal condition caused .

by operator error will be safely controlled by automatic protection systems and no risk to reactor personnel or the general public will occur.

4 t

t l

32 i

l

19. REFERENCES AGN-201M Reactor Operation and Operator Training Manual, Department of Chemical and Nuclear Engineering, The University of New Mexico, July 1984.

Reactor Hazards Evaluation Report and Site Survey for the AGN-211 Nuclear Reactor, October 1957.

Technical Specifications for the University of New Mexico AGN-201M Reactor, May 1986.

Safety Analysis Report for the University of New Mexico AGN-201M Reactor Facility, May 1986.

American National Standards Institute /American Nuclear Society (ANSI /ANS) 15 Series.

Code of Federal Requ1ations, Title 10. " Energy," U.S. Government Printing Office, Washington, D.C.

2 l

33

I at' cay NwU.Ed 'desass er fiOC esa trov Ad. r aneirs N#C FOAM 33 uS.NuCLEAAk41ULafOmvs W assom

'b'#,' BIBUOGRAPHIC DATA SHEET EGG-NTA-7467 til iggravCTiCNS ON rwe nasEast 3 L''*"

2 rirLt a,no sustif ta TECHNICAL EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY OF NEW MEXICO a o.rt #E80mf coMPLETEo Wo%r= vt.m i .w r-c. s.

flarch 1987

+ o.repeacar'avio C. H. Cooper oo~r. ....

l March 1987

. oact r. ..c.. ..r ~ua ..

, . . s a...~o c.a.,,,1. r io, ,... .~o .. . u~o .o o. . . ,a,,,,e e. C ,

NRR & I&E EG8G idaho, Inc. * ,.~o ca.~r~u ...

P. O. Box 1625 Idaho Falls, ID 83415 06010 via Tvtl op stront

'O S4%5ameNG cmG.Niz. riom %.wt .NO M. sung .cDats& dsaewee te C. set Standardization & Non-Power Reactor Projects Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ' ' ' " ' c "' " " * ' " ~ " " ' '

washington, DC 20555 t a s ,.t. .s r.. . so v .

, , . . r..e m -. r ,

This Technical Evaluation Report for the review of the application filed by the University of New Mexico (UNM) for renewal of Operating License No. R-102 to continue to operate its research reactor has been prepared for the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located on the campus of the University of New Mexico in Albuquerque, New Mexico. The INEL concludes that the reactor can continue to be operated by the University of New Mexico without endangering the health and safety of the public.

, I t a occwve%f .s.6vses .. nevwomos cascmiercas i t .v. . a. rv r Unlimited Distribution

i. stCum,7v CLaS3+8'CAfloh t ros sa,e#

, ,o. sri,,....c.n i~e o i s Unclassified sine reeern Unclassified l

,,~e....o,..ois 14 9 Rect