ML20216B685

Technical Evaluation Rept for Renewal of OL for Purdue Univ Reactor, Final Informal Rept

ML20216B685
Person / Time
Site:	Purdue University
Issue date:	05/31/1987
From:	Carpenter W, Carolyn Cooper EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20216B656	List: ML20216B651 ML20216B685
References
CON-FIN-D-6010 EGG-NTA-7527, NUDOCS 8706300132
Download: ML20216B685 (67)

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OlSCLAIMER This book wee prepared as an acccant of .vork sponse ed by an agency of the United States Government. Neither tne Wited .S'.ates Governrnent nor any agency thereof, nor any of their employees, makes my warr0nty, express or implied, or assumes any

. lega! liability or responsibility for the accuracy, completeness, or usefulness of any infontotion, apparatus, product or process disclosed, or represents that its use would net infrk.ge privateh owned nghts. References herein to any specific commercial

, Mtw process, or service by trade name, tradem A t. manufacturer, or otherwise, d&as r,rt, necessanly constitute or imply its endorsemettt, recommendation, or favoring . ,

by tre Orated ?).ates Government or any agency thereof. The views and opinions of I authort espressed herein do not necessarily state or reflect those of the United States . J Govurnment or any agency thereof.  ;

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EGG-NTA-7527:

iL iTECHNICAL; EVALUATION REPORT FOR THE RENEWAL

-1 1: . OF THE OPERATING. LICENSE FOR THE PURDUE. UNIVERSITY REACTOR r ;. ,

W. R'..Carpenter C. Hr Cooper?

- I Published May 1987' EG8G-Idaho,;Inc.

-Idaho Falls, Idaho.' 83415-

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- - 4 -o U;S. Nuclear Regulatory Comission.c

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- FIN No..-D6010

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y H l NBSTRACT l'

.This Technical. Evaluation ' Report for 'the application- filed by Purduet University .for the renewal of Operating License No. R-87.:to ' continue operating. its'research reactor has been prepared for the Office' of' Nuclear:

Reactor Regulation'of the U.S. Nuclear Regulatory Commission.

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The. facility' ..f!

is located'on the. campus of Purdue University in West Lafayette, Indiana.

i The INEL-concludes.that the reactor can continue to be operated'by Purdue University without. endangering'the h'ealth and' safety of the"public.

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1 FIN No. D-6010 Casework and Non-Power Reactor Reviews:- -

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FOREWORD-t ThisreportissuppliedaspartoftheEvaluationoflApplicationfor Nonpower Reactors program being conducte' d for the U.S. Nuclear Regulatory' Commission (NRC), Office of Nuclear Reactor Regulation, by the NRC Technical- Assistance , Group, Idaho Nationa1' Engineering Laboratory. 'The NRC-

~

funded this work under the authorization of FIN No. D6010.

D This report summarizes the safety review, of.the' technical portions of the license' renewal application: submitted to the NRC from Pur'ue d University for' the continued operation of their research reactor. As thisLdocumentLis to be the basis of select sections in a' formal Safety Evaluation' Report (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 ofLthe~SER:

1. Introduction

2. Site characteristics'

3. Design of structure, systems, and components

13. Conduct of operations

15. Technical specifications

16. Financial qualifications- l

17. Other license considerations '!

18. Conclusions. il 1

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

i references applicable to these sections'. '

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CONTENTS  !

ABSTRACT .............................................................. 'ii -

J FOREWORD .............................................................. iii f

.4. - REACTOR ...........................................'...............

1 4.1 Reactor Core ............................................... 1-4.1.1 Fuel Elements ........................................ 6-

'4.1.2 Control Rods ....................................... 6 4.1.3 Neutron Source ................... .................. 6 4.2 Reactor Pool'and Biological Shield ......................... ~7.

4.3 Grid Plates and Core Support Structure . . . . . . . . . . . . . . . . . . . . . . 7' 4.4 Reactor' Instrumentation .................................... 8 4.5 Dynamic Design Evaluation .................................. 8

4. 5.1' Excess' Reactivity and Shut'down Margin ............... 8-4.5.2 Assessment ......................................... 9 4.6 Functional Design of Reactivity Control-Systems ~............ 10 4.6.1 Control Rod Drive Asserblies ....... ............... . 10 4.6.2 Control Rod Ci rcuitry, and Interlocks . . . . . . . . . . , . . . 11 4.6.3 Assessment............ ............... .... ......... 12 4.7 Operational Procedures .....................................- 12-4.8 Conclusion ............................,.................... .13-

5. REACTOR COOLANT AND ASSOCIATED SYSTEMS ......,.........,........... 114~

5.1 P ri ma ry C o o l i n g Sy s t em . . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 14~

5.2 Process Water System ................ ...................... 14 5:3 Primary Cool ant Makeup Water System . . .a . . . . . . . . . . . . . . . . . . . . 14' 5.4 P rima ry Cool ant Chil l e r Sy stem . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . '16 r ..

5.5 Conclusion .................................. ............... . .16

6. ENGINEERED SAFETY FEATURES ............. 4 ,................'....... 17.

.J 6 .' l~ Ventilation. System ..........-...............'................ 17: 1 a 1

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6.2- ' Drain' System ................................................ '17 1

6.3 Conclusion ......................................-............ 19  !

7 .' CONTROL AND INSTRUMENTATION SYSTEM ................................ 20-7.1 Reactor Control. System'..................................... 20 f 7.1.1 Control Rod Dri ve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '20

" ~

7.1.2  : Servo Control System....................-... 3....... 22' -

7.1.3- ' Neutron Source' Drive ............................... u22

.7.1. 4 - Fission Chamber Drive .............................. 22 1

. 7.1.5 Annunciator and Alarm Systems ....................... 122 { 4 i '

7.2- Reactor Instrumentation ..'.................................. 23

.7.2.1 Channel No. 1 - Startup Channel .................... 23 j 7.2.2 Channel. No. Log N and , Period . Channel . . . . . . . . . . . 23  ;

7. 2.- 3 - Channel No . 3. Li near Power . . . . . . . . . . . . . . . . . . . . . . . 23' -

7.2.4 24 Channel.No. 4L- Safety Channel .....................

7.2.5 Temperature and Water. Monitor: Channel s . . . . . . . . . . . . . 24 7.2.6 Radiation Monitoring Instruments ................... -24' 7.3 SCRAM Sy stem and Inte rl oc ks . .. . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . 24- l 7.4 Conclusion .................................................. 25 1

8. ' ELECTRIC-POWER .................................. ................ 27.

8.1 El e ct ri c al P owe r Sy st em . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .. . . . . 27 8.2 Emergency Power ............................................... 27 8.3 Conclusion .................................-................ 27

9. AUXILIARY SYSTEMS .................. ............................... 28-9.1 Ventilation System .......................................... .

28-9.2 Fi re P ro te c ti o n Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . :. . . . . . . .

. 28 9.3 Fuel Storage ................................................ . .

28 9.4 Heating and Air Conditioning System ........................'.. 29' 9.5 Crane System ................................................ 29 9.6 Conclusion ........-.........................................: 29

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10o'EXPER! MENTAL PROGRAMS ............................................. .

- 30 .1 l10.1 Ex p e ri me n ta l Fa c i l i t i e s . . . . . . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . 30;  !

10 .1.1 R e f l e c t o r T u b e s . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . .. . . . . . . 30-  !

10.1.2 D r o p T u b e s . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130' r I

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10.2 Experimental Review ........................................

31 1 p 10.3 Conclusion ....... .......................................... 31 ..

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11. 'RADI0 ACTIVE WASTE MANAGEMENT .....................................-

32 i 11.1- ALARA Commitment ...........................-................. 332 L11.2;' Waste Generation and. Handling Procedures ...................- 32 1

11.2.1, Solid Wa'ste...................s........~.............. 32 l 11.2.21 Liquid Waste ....................................... 33 4

'11.2.3^' Airborne Waste. ...................................... 133 J

-11.3' Conclusion ................................................. 34

. 1 d

12. RADIATION PROTECTION PROGRAM .......................,............... .

35 .

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1,2 .1 ' A LARA C omm i tm e n t - . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

35 j 1

.i 12.2L Health Physics. Program .....................................- 35 1 12.2.1- Procedures ..................................'....... 136 :

12.2.2 Instrumentation ...........,...:... .........'....... . 36 9 12.2.3 Training ....... ................................... 136 12.3 Radiation Sources .......................:......'.............. 38-12.3.1 Reactor .......................................'..... . -38' 12.3.2 Extraneous Sources ......................... ...'.4.; . L38 12.4 Routine Monitoring ........,...... .......'..................: 387

'12.4.1 Fi x ed-Po si ti on Moni tors . . . . . . . . . . ; . . . . . . . . . . '. . . . . . 38- ,

112.4.2 Experimenta1'................s...................... '39:

12.5 Occupational Radiation Exposu'res ................... .......

394 s 12.5.1 Personnel Moni toringi Program' . . . . . . . . . . . . . . . . . . . . . . 39-

12. 5 . 2 = P e r s o n n e 1 ' Ex p o s u re s , . . . . . . . . . . . . . . . .. , . . . . . . . . . . . . ;-:. . ~ 39 ' .

12.6 , Effluent Monitoring .................................'....... s 40:

12.6.1 Airborne Effluent ...............:....................: 40

12.6.2 Li q u i d E f f l u e n t s . . . - . . . . . . . . . . . ' . . . . . . . . '. . . . . . . . . . . - -40J vi-t

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' i b 12. 7 i Envi ronmental Moni tori ng l . . . . .>. .'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 141 1 o '

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12.8? Potential Dose' Assessment ..............s.-.................. 41- l l 1

i. 12.9- Conclusion ....................'.............................. '41 j l

14; ACCIDENT' ANALYSES ................................................ 42' 1 T14 ' 1 Fuel El ement Handl i ng Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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• 14,1.1 Scenario. ....................................... ....

. 42 j 14 1.2 Technical Assessment ............................... 43 1

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14.'2? Maximumc Rea'tivity Insertion ................. .............. ;43 l

il 14.'2.1 . Scenario ............................................ 43 i

'14.2.2. Technical. Assessment ............................... 47-  !;

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. 14.3 : LFloodings o f ~ an Irradiation Facility a'nd Failure 'of a .

d Mov abl e Ex p e ri me n t . . . . . . . . . . . . . . . .i. . . . . . . . . . . . . . . . . . . . . . . . . 491 14.3.1. Technical Assessment ................................ -49' 14.4. Loss of Coolant. Accident.................-..............c,.....- .

49.1 i 14.4.1 Scenario ............ -

50 .l 14.4.2 . Technical Assessment ................................ 50 14.5 Maximum Hypothetical Accident ... ..........................

. :50-9

- 14. 5.'1 Scenario ................................. . . . . . . . . . . . . 51 1 14.5.2 Technical Assessment ..... ............... ......... 51'  ;

14.6' Conclusion ..........................'......................... " 52 i

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19. REFERENCES ......................................................... 56.

1 FIGURES' i 1

4.1 Facility layout ..................... ...................:.......... 4 ,; g

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4.2 Core configuration ............................-................... e y

5.1 Reactor Water process system .........................-............ 15L j 6.1-' Reactor room, ventilation and cooling system ..............c....... 18 ,]

7.1 Reactor' control system .........................................-.- L21 - .

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12.1 Organi zati onal structure : fo rLPUR-1 operati ons . . . . . .. . . . . . . r . . .. . . . . ;37H a

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I TABLES i

4.1 PUR- 1 pri nci pal desi gn parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1

14.1 Results'of the power transient analysis with ramp insertion  !

of control rod ................................................... 45 j t i 14.2 Results of the power transient analysis with no control rods ..... 46 i

14. 3 Compari son of important fuel data . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . 48 14.4 Dose rates in the reactor' room from a failed fuel experiment ..... 53 .  ;

14.5 Dose rates at 100 meters from a failed fuel experiment . . . . . . . . . . . 55 l

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o l4; . REACTOR'-

. The Purdue ' University Reactor (PUR-1) was built by ;Lockheed Nuclear

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Products and initially' attained criticality in August 1962. This reactor.

uses MTR type 93E enriched U-235 aluminum-clad fuel plates that are L assembled into fuel assemblies and 'placed into'a graphite-reflected region j S form the reactor; core'. The reactor. co're is immersed in1an open tank of; light water-that serves as the neutron moderator,-coolant, and. shield. The.-

. ' reactor' operates _at a maximum power.-level of 1_kW. .The' reactor power is?

regulated by inserting or withdrawing; neutron-absorbing control rods.

The~ reactor'is used as a neutron source for. activation analysis studies,,

, academic research,'and_the;1imited production of radioactive. isotopes. It also isfused as:a' training facility for the nuclear engineering: educational program. The PUR-l'.is operated for'an average;of about:13..kWh/y. The:-

principal 1designparametersfor.the' current'coreconfigurationare$listedi in Table'4.1.

The PUR-1 facility layout in the Duncan Annex of the Electrical Engineering Building isLshown!in Figure 4.1.

4.1 Reactor Core The core of the reactor is 30.48 cm. square and~60'96 cm.<high'. .It consists of 13 fuel assemblies and 3 control rod assemblies. Each fuel assembly consists of up to 10 aluminum-enriched uranium a.lloy.. plates' Each control rod assembly consists of up to six plates and'two1 aluminum l guard-plates with space for control rods. Adjustments'.tolensure ^that: maximum excess reactivity is not exceeded are effected.by substituting dummy fuel) plates for uranium plates. The reactor is cooled and moderated'by a; pool ,

of light water. The 4'x.4 array of fuel elements isi eflected r on a11' sides:

.. with graphite-reflector; elements and:en.cthe top and bottom with water; :The 20' reflector elements are composed. of graphite waterproofed with epoxys

' resin and are contained in standard fuel! element can'./s One row offsix' graphite-reflector. elements is designed to hdid ' samples (for isotope production (see Figure' 4.2).

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LTABLE 4.1. .PUR-1 PRINCIPAL ~ DESIGN PARAMETERS-

. Maximum power? level- 1 kW LGeometry of core 1 0.3 x 0.3 x 0.6 m; JModerator-coolant , Light. water Maximum' excess' reactivity 0.6% Ak/kL

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-6 Promptneutronkiifetime-  ; 77.2 10 s-

' Fuel assemblies

. Number . . L 16 Standard. .

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? Cont'rol : a ssemblies ' 13 Numberiof-plates per: standard assembly 10-Number,of plates-per control assembly? 6- .

7.0 x 64:x 0'.15 cm

' Plate dimensions.

Active fuel length' 59.4 cm.

.:U-235'per plate- '

.16.5 g-Water. gap E0.53'em.

Cladding 0.020 aluminum

. EnHchment 93%; ,

Reflector '

Material ontsides ~

.Graphi.te1 Number <of graphite assemblies" 201

Control rods and' drives .

Number ofz: regulating rods- 1 Number ofJshim safety. rods . 2:

Total number of' control rods 3 .,

Measured worth of control rods. ,

Regulating rod - .

'O.26% Ak/kI Shim-safety rod No. 1- .5.0%'Ak/kl Shim'safetyirod No. 2' '2.4% Ak/k' '

Rod' speed-out-

-Regulating rod - 45.0Lem/ min. ,

' Shim safety rods - 11.2icm/ min. . - '

-Scram time?for comp.lete inser' tion. 1 s;

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- TABLE.4.;1.- (continued)-

Material Regulating' rod H'ollow st'ainless steel

, Shim-safety. rods Solid borated stainless' steel Size Regulating: rod - .1.3 x 5.7 x 64.8 cm ,

Shim-safety rods 1.3 x.5.7 x'64.8 cm

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Maximum rate of_ reactivity change Regulating rod. 0.006% Ak/k/s.

Shim-safety rod No. 1 0.031% Ak/k/s Shim-safety radLNo; 2 0.013% Ak/k/s a

Average' rate of-reactivity change

-Reg'ulating rod_

0.0031% A/k/ Ds-Shim-safety rod No._1- 0.015% Ak/k/s' Shim-safety rod No. 2 0.007% Ak/k/s Reactivity effects-Temperature coefficient 2

(Calculated) -2.1-10 %Lak/k per *C-

-2

-3.4 10 % Ak/k per.*C (Measured)

Void coefficient (measured) -2.6:10 -2

% Ak/% void Process water resistivity >330,000 ohm-cm-pH 55-

. 1-Flow rate 1.89 L/s; 4

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'Fiqure 4.2 Core configuration  :

5' l'

I 4.1.1 Fuel Elements  :

The MTR-type fuel plates are 93% enriched U-235 metal alloyed with-1100 aluminum alloy and clad with 0.051 cm 1100 aluminum alloy, with a i total thickness of 0.152 cm. These flat MTR-type fuel plates are then inserted in aluminum canisters. Up to 10 fuel plates, 7 x 63.8 x 0.152 cm overall dimensions, are contained in each of 13 fuel elements. Up to six .  ;

plates plus two guard plates of 6061 aluminum alloy are contained in each I of three control rod assemblies. The number of fuel. plates in the fuel -

assemblies can be adjusted to provide for a maximum excess reactivity of j 0.6% Ak/k. )

l.

4.1.2 Control Rods Three control' rods are used to control and regulate the power levels in the PUR-1: a regulating rod and two shim-safety-rods. Each of the three rods operates within a hollow guide tube. The neutron absorber in the j regulating rod is stainless steel, and the neutron absorber in the shim-safety rods is borated stainless steel. Each control rod is 64.7 cm g long and has a vertical travel of ~61 cm. The cross-sectional dimensions j are 1.3 x 5.7 cm for all the rods. The maximum rate of withdrawal for the control rods corresponds to 0.031% ak/k/s and 0.013% Ak/k/s for the two  ;

shim-safety rods and 0.006% ak/k/s for the regulating rod.

l The neutron-absorbing sections of the shim-safety rods are supported by I electromagnets that release the rods in a scram. The scram time for complete insertion of these shim' safety rods is 1 s. The regulating rod is mechanically connected to its drive and does not scram.

4.1.3 Neutron Source The PUR-1 utilizes a 5 Ci Pu-Be neutron startup source. The source is -

located in a special reflector element source holder adjacent to and just outside of the graphite reflector (see Figure 4.2). The source can be-withdrawn from its in-core position manually by means of an attached steel 4 1

6

5, s .- ,

e .

.cableithatiis connected.to the top:,of the~ source holder cap'. 'An indicatorL

l

1.ightLcoupledtothestartupmeter'atthe'controi:consoleshows'whetherthei j

. source.i's'inLor'out of the core.

'4'.2' Reactor Pool and Biological Shield TheLreactor core.isslocated within two coaxial tanks'that form the reactor

' pool. The outer tank rests onL a' concrete pad 4.'6 m below floor level.

~

+ '

The reactor pool 'is' built below floor. level except for' the one' meter wall a 1

that serves as a: biological shiild for the' operators and' experimenters..

' The poolu is contained in a cylindrical < tank 5.3 m' deep. andL2.4. mLin.

. diameter. The' core isl located totone- side (to give additionali experiment space. ' Opposite from the reactor core, two' fuel storage racks:are. mounted-on the tank- floor. Thes.e ' fuel stoEage. rack's are-fabricated of aluminum and-' :1 contain aLboral sheet:in'theirLeenters'as a. neutron-absorption material.

The : supports for the d' rive mechanisms' for the control Jrods,sthe fission  ;

chamber.and the source,- and the" neutron detectors (are fastened to the>

support plate 1at the top of the'. tank. A trave'rsingimechanism was mounted; on the top of the-reactor pool wall after the-reactor was b'uilti tA lightweight,' portable aluminum bridge can be. placed across ;the. pool. for -

maintenance.and fuel-handling operations..

Shielding over the core is provided by 4'm.of water,.which'r'educesithe radiation level at the top of the pool to less than 1 mrem /h.when the core 3

~

is operating at 1 .kW. The concrete pad, reactor' tank, and distance reduce the maximum radiation level at the. control. console area'.to:less'than' O.1-mrem /h at 1 kW.

4.3 Grid Plates and Core Suoport Structure A 7 x_11-position grid plate supports theJ16 fuel elements and 20 reflecton'

.. and. isotope production elements. The'approximatefactive core dimensions.

4

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.are 30.5 x'.30.51x;61 cm.

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' 7..

Thel core' structure..ikcentered'approximately.

[76"cm fron( the' center of the reactor tank Land '9 cm from the bottom of the

. tan'k'.

e --4,41IReactor Instrumentation The nuclear (operation of the PUR-1 isimonitored.b'y fourfneutron sensitive

. channels'(two of:which are always on; range).that: indicate. thermal power; q

sleVelio.ver)the entire operating range of the: reactor. These channels -

. initiate ' scram signalsiif preset neutroniflux : levels areireached. The. bulk

. reactor. coolant.temperatureismea'suredmanual1hwithia;thermometerplaced

'inftheipoolwater.-l.TheinstrumentationlandLeontrolsystemsarediscussed

indetallfinS'ebtion7.

-4.5 Dynamic Desian Evaluation-Y The' PUR-l'.is; operated b'y manipulating control' rods in response. to changes '

-in .the neutron flux'(power) measured by the instrument channels. There are c

interlocks to prevent: inadvertent. reactivity additions and a. scram: system to' initiate a rapid. shutdown (reactor scrarA) if al preset:; poweril.imit' has? '

been reached. Additionally, the measured temperature Leoefficient.'ist negative'over the operating temperature range. Inlthe"unlikelpfeventof.

inadvertent high power operation leading to high: temperatures',uthis.

negative temperature coefficient of reactivity'will. tend to limit'the' reactor power. i

4. 5.1- Excess Reactivity and Shutdown Margin; . ;. .;.

i u j x

t;  ; :l Excess reactivity is defined as that value of reac.tivitylthat'.would' occur- j if'all control rods were completely' removed from.the1 reactor: core./ ,

j Reactivity is measured'for a given core-loading _ starting from'a; d just-critical cold,, clean core. A, designated core loading mayfincludel '

irradiati'on facilities, such as.the. isotope production elements, or'other;

~

j facilities of,such nature that they-become a'portionLof?the! core'when *

- ~2) installed. s

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i Excess reactivity must be built into the reactor core in order to compensate for a number of reactivity losses. Also, a sufficient ,

. reactivity must be available to allow for an adequate reactor period for )

l the PUR-1. This reactivity value has been determined to be not more than 0.6% Ak/k, and is the maximum allowed'under any operating condition by the {

Technical Specifications.

The Technical Specifications require that the control rods provide a' 1 1

shutdown margin greater than 1.0% with the highest-worth control rod fully withdrawn and with the highest-worth experiment (0.4% Ak/k for each secured' )

experiment or 0.3% Ak/k for each movable or unsecured experiment) in its j most reactive state under any conditions of operation. This is to provide assurance that the reactor can be shutdown safely even'if one control' rod did not insert.

The current core configuration has an excess reactivity of 0.48% Ak/k. The ]

individual control rod worths are shown in Table 4.1. The total ~ rod worth ]

is 7'.66% Ak/k. The shutdown margin for the current core configuration the.

highest-worth rod fully withdrawn is 2.18% Ak/k. Therefore, the current 4 core configuration meets both the shutdown and the excess reactivity requirements. With all rods fully inserted, the core is suberitical by l 7.18% Ak/k.

. i 4.5.2 AssessmejLt j The Technical Specifications require that at least three control rods be operable and the reactor can be brought to a subcritical condition even if the highest-worth control rod is totally removed from the core. These requirements ensure an adequate shutdown margin and provide sufficient redundancy in the unlikely event of a control assembly malfunction.

Limiting the total excess reactivity of the core plus installed experiments to less than 0.6% Ak/k allows for adequate reactor control under normal circumstances and prevents a prompt power excursion under any postulated'  !

. abnormal circumstances.  !

9

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On basis of the above considerations, the Idaho National Engineering Laboratory (INEL) concludes that excess reactivity will be limited sufficiently and adequate redundant shutdown capability is provided to ensure safe, controlled operation of the PUR-1.

In addition, the 1.0% ak/k shutdown margin with the highest-worth rod fully withdrawn and the highest-worth experiment in its most reactive position .

provides reasonable assurance that the reactor can be shut down adequately under all postulated operating conditions. ,

4.6 Functional Desion of Reactivity Control Systems f i

The power level of the PUR-1 is controlled by three control rods: two shim-safety rods and one regulating rod. The two shim-safety rods can be scrammed and use solid borated stainless steel as the neutron absorber.

The rods are connected to their drives by electromagnets. If electrical power to these electromagnets is interrupted for any reason, the armatures f of the two shim-safety rods are released and they fall by gravity into the core. The single regulating rod uses hollow stainless steel as the neutron- {

1 absorber. This rod is mechanically attached to its drive.and does not fall into the core during scram. The core locations of.the three control rods are shown in Figure 4.2. ]

l Each rod-drive system is energized from the control' console through its own f independent circuits. A manual scram at the control console is possible l for the two shim-safety centrol rods, or they can be scrammed automatically by the safety circuits. Although the regulating rod cannot be scrammed, {

certain manual or automatic trips can cause the regulating rod to be driven i downward into the core. {

l 4.6.1 Control Rod Drive Assemblies The tubular control rod-drive assemblies mounted in a vertical position.

over the core. Motion is imparted to a control rod by a positive upward .

- l) force on the extension rod, which is coupled to an electromagnet (except 10 q 1

.c ifor the' regulating' rod)', and' to a screw mechanism, ehich' is rotated through a fixed nut by-the drive ' motor at the upper end of the tube. All rod drives are supplied with instant reversing induction motors. Downward motion results from the positive down-drive of thefscrew mechanism by the drive motor.

Limit switches are provided on the drive mechanism for the rod-drive.

. units. The following switches are;provided: Jam,'up, 2/3 up, and down.

. . -. In-addition, the safety rods.are provided with bottom and magnet-engage j switches. One coarse and one fine position indicator located on the 1 console indicates,the position of each rod. These indicators are located. -1

.on,the console and provide readings'to 0.01 cm.

4.6.2 Control Rod Circuitry and Interlocks J

l Three identical control channels are used for the two shim-safety rods and j the regulating. rod. A push-button switch selects an individual' rod to be i controlled. All controltrods'can be inserted simultaneously into the. core

)

by a gang-lower switch when shutdown is desired. This switch cannot cause d 1

the control rods to be gang raised under any circumstance. Control-console j indicators for each rod include upper limit and lower limit. Additionally, j engage (magnet coupled) and shim range (rod not' bottomed) indications are provided for the, shim-safety rods.

Rod positions are indicated on a coarse vertical scale and on a selectable digital readout device having a resolution of 0.01 cm. The true value'of the rod position is known to 0.03 cm, which is equivalent to a maximum  ;

reactivity uncertainty of less than 0.01% ak/k.

. i Two types of automatic action are incorporated into the reactor safety  !

system to correct abnormal or. unplanned conditions; trip'and rod insert.

In a trip, the shim-safety rod or rods are dropped by . removing the current,  ;

from the magnets. A rod insert (set back) will cause a11'three rods.to-  !

- dive' downward into the core. Both actions are of the. latching ~ type and l

l l

l 11 i

+

manual reset of the safety system is required to return to the normal conditions. A complete description of the scram' system with set points is contained in Section 7.

4.6.3 Assessment I

The PUR-1 is equipped with safety and control systems, control rods, rod ,

drives, scram-logic circuitry, and interlocks that have performed reliably-and satisfactorily in the PUR-1 for 24 years. .

The control systems allow for an orderly approach to criticality and for {

safe shutdown of the reactor during normal and abnormal conditions. There )

1 is sufficient redundancy of control rods to ensure safe reactor shutdown, j even if the most reactive rod fails to insert upon receiving a scram f signal. Interlocks prevent inadvertent rod withdrawal and, thus, inadvertent positive reactivity changes. A manual scram button allows the j operator to initiate a scram independently for any conditions deemed to f require a prompt shutdown. )

i i

On the basis of the above discussion, the INEL concludes that the reactivity control systems of the PUR-1 are designed adequately and w.ill function to provide a reasonable assurance of safety, i

4.7 Operational f rocedures )

i The PUR-1 operates under Technical Specifications that direct the {

operation, audit, and surveillance of the reactor and provide procedural reviews for all safety-related activities. Written procedures have been i established for safety-related and operational activities that include i i

reactor startup, operation, and shutdown; maintenance; and calibration of equipment and instrumentation. In addition, the reactor is operated by trained NRC licensed personnel in accordance with the above-mentioned ,

procedures and Technical Specifications.

12 i

I.8? Conclusion'

..The'INEL~ review'of the PUR-1 facility;has included' studying its. specific design'and insta11ation',. control.and. safety. systems,;and operational limitations, as identified in the Technical ~ Specifications.- The INEL.

con'cludes that the PUR-1 reactor was designed and built'according to^ good industrial practices. The INEL furtherLconcludes that there is sufficient-shutdown margin to ensure that the PUR-1 reactor can be adequately l shut-

~

down under all anticipated normal and Labnormalz operating conditions.

The' design features of the-PUR-1' reactor:are similaritoithose of many.

pool-type research; reactors operating in many countries: of the world.'

Based on its review of the PUR-1 reactor and~1ts experience with similar facilities,' the INEL concludes that this1 reactor is capable.of safe operation,'as limited by its Technical Specifications.

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b 5 REACTOR COOLANT AND ASSOCIATED SYSTEMS i

5.1 Primary Coolino Syste'm The energy produced in the core'is dissipated to the pool water as heat by' the natural convection of the approximate 22,700 L of demineralized water .

in the reactor pool. The pool water is maintained at the ambient temperature'of.the environment by heat conduction to the ground and air and ,

by'some evaporation of. water from the pool surface. In order to raise the ,

temperature of the pool water 10 C, the reactor.would have to operate continuously for more than 200 h at full power (1 kW), assuming no heat loss. Thus, pool heatup poses no constraint on the anticipated operating schedule of the 'PUR-1 reactor (~13 kWh/y).

1 5.2 Process Water System

'l' The process water system is assembled in one unit and contains a pump,>

filter, demineralizer, valves, flow meters,;and a heat exchanger (see a Figure 5.1). The heat-removal capacity of.the heat' exchanger is 10.5 kW.

It was designed to maintain the reactor pool temperature at 75 F'during -

continuous operation at 10 kW. The demineralizer contains a removable cartridge that is monitored continuously for radioactivity buildup'. This-system limits the aluminum corrosion rate, corrosion product buildup, and ]

neutron activation of impurities in the coolant by the use of filters'and  !

ion-exchange resin. .

1 5.3 Primary Coolant Makeue Water System ]

l Makeup for the pool.is taken batchwise from the Purdue University water line and passed through the demineralizer enroute to the pool. ALvacuum i breaker. excludes any possibility of siphoning pool water into the supply- ..

line. The pool makeup water system, in addition to the demineralizer, also-includes a norma 11y' closed manual shutoff and throttle' valve and a check' -

]

valve.

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LI5 -

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5.4.. primary Coolant Chiller System 9

' Although the chiller is net needed for present operations, it remains

available if required. Calculations. indicate that the temperature rise L

rate while operating the PUR-1 at a power level of 1 kW would be

~

[

-2 C/h, based on the mass of water equal to 1.85 10 4 4.65110 kg. This takes no credit for heat loss to the surrounding sand and gravel or loss by'

~

evaporation. Experimentally, no temperature increase.has been observed with the pool = thermometer following 8 h of operation at 1 kW. 'The chiller is designed with three loops to prevent the spread of radioactive contamination from the primary loop.to.the heat dump. The. pool water -  :

passes through the primary loop while a freon .refrigeranttis in the secondary loop. . The third loop uses campus water to remove the heat and is-  !

l discharged into the campus' sewer. system. Radioactive contamination cannot i i

l pass through the three-loop system, unless at least two pipe failures were j to occur simultaneously with significant abnormal radioactivity in the l

primary coolant, and the chiller system was infoperation.

5.5 . Conclusion 4

The INEL concludes that the reactor coolant system at the PUR-1 reactor is  !

of proper size, design, condition, and is maintained properly to ensure j adequate cooling of the reactor at the power level.specified in the PUR-1 '!

~

operating license. Also, the process water systam can limit both corrosien' I and radioactivity problems associa'ted with coolant contamination.

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L . 6.

ENG1NEERED SAFETY. FEATURES 1

Engineered' safety features are those features or systems that-mitigate the j p_otential consequenbes of accidents. The only systems that could be l considered-as engineered safety features associated with the PUR-1 facility- l 3

are the' ventilation system and drain. system'. These systems are designed.to {

limit.the unco'ntrolled release of airborne and solid radioactive materials '!

during normal' operating conditions as.well as' accident conditions.

6.1 -Ventilation System The outside air supply and room exhaust are passed through high efficiency particulate air-(HEPA). filters (see Figure 6.1). The reactor room is:

maintained at negative air pressure (minimum 0.13 cm of water). All doors to'the' reactor room have foam rubber seals. Steam heat is-used to heat the.

room and a room air conditioner circulates and cools the reactor' room air.

t Curing emergency conditions, the exhaust system 'and the air conditioners ';

are shut off and the sealed room will prevent the rapid spread of )

contamination. I 1

6.2 Drain System j

I The only floor drain to the sewers is sealed except for a. vent opening. j This~ vent is raised about 1.2 m above the floor and has a_ filtered inverted 2 opening. Condensate from the air conditioner is released to this drain

~

through an opening 3.66 m above the floor. During an emergency, the valve on the drain from the condensate hold-up tank.is shut off with the same i switch that shuts off the exhaust system. The condensate is. held until it j '

is tested for radiological contamination before releasing it:to the sewer.

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i? J0n the basis ofEthe evsluation o'f potentialtaccidents atLthe PUR-1:thatfare ,

J, l , discus' sed in Section 14Lof thisLSER, the INEL concludes th'at no'significant 4

amo'unts of: airborne : radioactivity: or liquid ; waste' would heL released -.in1!o : or ,

c.outiofsthe reactor room.' Therefore,'the:a'ailable' v ventilation system;is ' g s

considered to be.' acceptable;.:

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E7h LCONTROL AND INSTRUMENTATION SYSTEMS; ,

E ,

o The major components of the PUR-1 control.and. instrumentation systems, j includi.ng rod controls, annunciators, pen recorders, and meters, are' '

.i located:in the control console. The control console is designed to provide

' maximum visibility _of the instruments and accessibilit9;to the controls and ,

indicators, All indicators and controls _necessary for startup and, shutdown l

operationsare-lodatedinonegroup11nfrontofthel operator. ,' . j q

.,7 ;l 'leactor Control System ,

The reactor controlisystem.at the.PUR-1 facility, consisting of'_'both' nuclearand process instrumentatior', prtvides' reactor _ control' during normal operations ;and ensures safe shutd'own in the event of abnormal operation l(see Figuref 1). InterloAks'areprovidedb'etweentheinstrumentation system'and trc(scram system to provide. positive control of the reactor and' t essentially-eliminate the chances of accident initiation.

7.1.1 Cont;o1 Rod Drives Three identical and independent control channels are used-for the-shim-)afet.y and regulating rod drive systems. A push-button; switch selects an indiv;dt.nl rod to be controlled. All control rods can be inserted simultaneously into t.he core by a gang lower switch when' shutdown is desirt<. This switch cannot cause the control rods _-to be gang raised under Gny cficumstance. A complete description of the PUR-1 Control Rod' Drive; Sys'.ect is given fn Section 4 -

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-7.1.2 Servo Control Sysgem A servo control system provides automatic control once the reactor has reached the desired power level. The servo control system senses i deviations from an adjustable set point on the Channel No. 3 linear power i recorder and adjusts the position of the regulating rod to maintain the y reactor at.a constant puwer level. Servo permit circuitry actuates the .

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console alarm buz:er if the reactor power deviates by more than 5% from the j set point, indicating a malfunction of the system. A deviation meter is , 1

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located on'the console. {

l 7.1.3 Neutron Source Drive A! motorize'd . neutron source drive is provided to raise the source through a travel of approximately 1.8 m to the " full out" position. The two position system is operated by-raise-lower switches at the console with limit switches to indicate the source " upper limit" or " lower limit" positions.

s 7.1.4 Fission Chamber Drive i

i Controls for the motor drive system for the Channel No.1 fission detector. I are also located on the console, with both a coarse position' indicator and a selectable fine position _ indicator. The drive. system is selected'and coupled to the drive switch in the. same manner as the control rod drives.

Indicator lights note the upper and lower _ limit positions, however, the detector may be placed at any position within its range. .

7.1.5 Annunciator and Alarm Systems i

When a system trip occurs, or when other abnormal system conditions are sensed, an alarm (buzzer) sounds and an illuminated indicator is lighted on the control console, indi'cating the' source of the trouble. An annunciator ,.

acknowledge button may be used to reset the buzzer.

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7.2'. Reactor Instrumentation, 1

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The' function of the reactor instrumentation is to-provide adequate- l r

information for._ the operator and generate signals to control the reactor or s

initiate trips. The nuclear instrumentation consists of. a fission. chamber, j a compensated ion chamber, and two uncompensated-ion chambers. All neutron' l detectors are arranged near' the reactor core to obtain 'high- sensitivity _ and'

.fa'cilitate repair, maintenance, and repositioning. The detectors'are in l watertight aluminum tubes, The f.iss. ion chamber is provided with a

-: motor-driven positioning mechanism and position indication system; the ,

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-i other detectors:are manually adjustable, j d

7.2.1 Channel No. 1 - Startup Channel' The startup channel' is used to monitor.the neutron. flux. The. channel consists of a movable fission chamber, preamplifier,. pulse amplifier,

.scalerforaccuratecounting,'logcountrateandperiodampiifier, count i rate recorder, and shares a period recorder with Channel No. 2. The range of_this equipment is from 1 to 10 4-counts /s (about 1.5 x'.10 -5 to.

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I 1.5 x 10 ~1 W) with periods from ',0 to +3 s.

i 7.2.2 Channel No. 2 - Loo N pd Period Channel Thc %g N channel indicates the reactor power level over the range from '  ;

0.0001 to 300% power level (10 ~3 to 3 kW). .The detector.for this; channel-

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is a compensated ionization chamber followed'by a. log N amplifier plus  ;

period instrumentation with outputs to the log N recorder and to--the period  ;

recorder shared with Channel No. 1. j i

7.2.3. - Channel No. 3 - Linear Power - ,

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The' linear level channel is capable of measuring neutron flux in a reactor- , q operating range from about 10-5'W (shutdown)to >100 kW. The sensing;

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f element is a BF ionization chamber coupled to a micro-microammeter.

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7.224 Channel No. 4 - Safety Channel

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l The' safety channel: utilizes a BF3 ion chamber and: feeds directly into the safety' amplifiers. The sensitive' range of this instrument is from'a few percent .to at' least 150% of power, linearly. Its-output.is indicated on the instrument chassis (instrument panel). The purpose of this channel is' solely' to provide reactor trip. ' .

. 7.2.5 ' Temperatur'e - and Water Monitor Channel s ,

.Water temperat'ure in the PUR-1 pool is measured by;a thermometer suspended-by a string in.the pool. Water level is measured by a scale immersed in the pool. Water conductivity is displayed in.the console-by two meters -

registering the output of two conductivity cells that measure the ' pool water.'be' fore and after passing through the demineralizer.

7.2.6 R'adiation Monitoring Instruments 2

The radiation monitoring system consists of three fixed position' remote I area monitors (RAMS) and one continuous air monitor (CAM). All of 3he RAM alarms initiate a reactor scram. The CAM detects airborne particulates and  ;

alarms. A continuous sample is' drawn from the reactor room;through the CAM filter, which is checked bimonthly for gross beta gamma activity.

7.3 Scram System and Interlocks Three types of action'are incorporated into the control jystem to correct:

for. abnormal reactor conditions:

Fast scram--is initiated by short reactor, period on Channel No. 2'

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o or high flux on Channel No. 4--interrupts the current to the-control rod magnets electronically ,, ;

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'o Slow scram--initiated.by short reactor period on Channel No.1 or No._2 or'by'high' flux on' Channel No. 2 or No. 3--causes the rods to drop by removing power.to the magnet power supply by.

means of a relay o Rod Insert--(setback)--initiated by short reactor period on Channe1LNo. I or No. 2 or by high flux on Channel No. 3 or No. 4.--causes all three rods to be driven downward into he

... core.

All three actions are of:the latching type and' require' manual. reset before return to the normai operating condition.

In addition to high flux and reactor period initiated scrams, reactor scram or rod insert at PUR-1 is initiated by any of the following signals:

o Manual push button o CIC power supply failures o- Safety amplifier trouble o Area radiation. .;

7.4 Conclusion ..

t The control and instrumentation systems at the PUR-1 are designedito provide reliability and flexibility. There is' adequate redundancy and _

4 diversity in the nuclear flux (power) monitoring circuits ~. In .'particul ar,

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nuclear. power measurements are overlapped.in the ranges of the startup,-

log-N, linear power, and percent power -level-(safety) channels. In view of 1

, the simple nature of the open pool, the' temperature and water monitor instrumentation is considered adequate. On the basis of~the above

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-,.v. Jsystemslat?the~PUR-1complywith.therequirementsTand-performance.

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' 8.1' Electrical Power System l

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The electrical power for building lightingLand reactor instrumentation is

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single phase, 60 Hz,.120/240 V, which is furnished th' rough a transformer. j and several control panels located throughout the building. I

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8.2 Emergency Power The reactor will scram'in the case of an electrical power interruption -

because th'e control rods are supported by electromagnets. Because the- j decay heat generated in the core following a scram is not enough to cause fuel damage (see Section.14), emergency power is not required to' maintain- j the reactor in a' safe shutdown condition. Power.for the facility. Intrusion detectors is supplied by a 12V battery that is checked monthly and replaced ]

biannually. In the event of an electrical outage,.this battery would

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i supply the necessary power for the'se instruments for at'least 24 h. j Battery powered emergency lighting is also available to facilitate ]

personnel movement during a power outage. In the event of a' power outage, no radiation monitors would be operating except for.the portable _ hand-held.

battery powered type.

8.3 Conclusion

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The INEL concludes that the design of the electrical power system, coupled .

with the fact that the reactor will scram in the event of a power-failure, is acceptable for continued operation of the PUR-1.

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9l.-AUXILIARY' SYSTEMS The.. auxiliary systems considered are the' ventilation system, the provisions for -fire protection, the fuel storage system, the heating and air conditioning systems, and the crane' system.

9.1 Ventilation System The' ventilation system is considered to be an engineered safety feature and is discussed in'Section 6 of this report.

,9.2 Fire Protection System-The PUR-1' reactor and.the building where'the reactor is located is intrinsically fireproof and in the event of a fire no special ' precautions are required. The fire protection system for the reactor facility is-typical of a prudent university program and of low p'ower research j reactors. There are two portable fire extinguishers in.the' reactor room: .l located at either end of the room. In the eventof al fire, the reactor would.be shut down and the supervisor,' or alternate, would be notified.

Normal-fire procedures for the building are in' place and are expected to -

eliminate accumulation of flammable materials. A fire ' station is located  !

on campus-(~1/2 mi from reactor) and would be available on short notice to j assist in case of a' fire.

9.3 Fuel Storace The only fuel.at the PUR-1 reactor facility is the existing core. The. fuel is only handled outside the core configuration for experimental or-inspection purposes. During some experiments, fuel ~ elements are stored..in one of the two fuel racks located inside the vessel on the' bottom. These -

racks employ poison (boral steel plates) and geometry to ensure suberiticality. An-annual fuel inspection.only' involves removal of one

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element'outside the vessel at a time.

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[9'.4?'HeatingandAirConditioningSystem. ,

'. . l The PUR-1: reactor facility heating andL air conditioning system is . integral y ito the ventilation. system discussed in Section 6 of.this: report. The air j L:t operated dampers in this system are supplied pressurized air from the-E j

University system. The actuators'are designed to close the dampers if a l

l loss of air should occur.

qe-. l e, 9.5. Crane System  !

-i The PUR-1 reactor facility. has a 2 ton' crane which on'ly runs along one ' axis , -

over the center of the reactor: tank and .is. normally positioned at the end-offthe track". This crane is only occasionally usedLunder the supervision. ,

of.the,reactorsupervisorfor.~installingspecialexperimendsand'isin' good 1

? working order.  :

9.6' Conclusion The INEL concludes that the auxiliary systems.r.t.the Purdue Universityi reactor' facility are designed operated, and maintained adequately.and are:

q capable of performing their intended function of helping to ensure thefsafe operation of the PUR-1 facility. l I

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.1'0i$EXPERIMENTAFPROGRAMS; The PUR-1 supports' educational programstin physical,.biologica1 Eand-l pharmaceutical? sciences and is used the training ~ engineering students 'on--

the'Purdue campus.: The': reactor also is faisource of ionizing : radiation and '

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. neutrons used for!varioussresearch programs 2 ..

10.E ' Experimental Facilities'-

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t 110.1.'11 Reflector Tubes' .

> - Six! positions in t'he graphite reflector < grid /along .'one side 'of. the coreTare' utilized for:lspecialfisotope production elements(see Figure 4.2)c These elements are. identical to the graphite . reflector elements except for central-. access' holes to accommodate: samples'up-to 3.8 cm diameter and.up to

61 cm:long. ' Currently, the sample capsules loaded into' these~ locations are j ,

' 7.6 cm'longc x 2.5 cm-outside diameter and are fabricated of aluminum.

. Experiments: located 'in these tubes are considered securediin: that, they .are not moved during' reactor' operation.

'10.1.2 Drop Tubes c Three drop. tubes are currently utilized at the PUR-1. 'Their positions:are shown- in Figure 4.2. All of these. tubes extend from'the level of:the?

active core to a height above the pool waterflevels sufficient?toprevent ,i inadvertent flooding.. Reactivity addition,' 'shouldLaccidental[ flooding occur, is co'vered in Section 14. The inside diameters of'the.tubesJare 1.6, 7.6, and 12.7'cm. The 7.6 cm tube is fabricated of PVC;.theMother two '

are fabricated of stainless steel. ' Experiments contained in these tubes q are properly secured to a suitable . tether 'and then lowered .into ' the. tube.-

Because they may be moved during -reactorfoperation, theyLare, classedf as. .,,pJ nonsecured experiments. j!

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experiments' proposed for the'.P'UR-1 are reviewed and< approved.by the--

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h Committee'on.ReactorOkeration(CORO)LtoTestablish1fthey.fallwithinlan

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lendelope of previously accepted reactivity a'ddition and radiological: j q consequences. j 1

l 10.3. Conclusion

1. v j Th'e. INEL; concludes that the design of the; experimental' facilities, combined )

withthereviewandadmihistrative.'proceduresapplied.toallresearch Jactivities, is

adequateito ensure that' experiments (a) are not'lik~ely..t'o :

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fail, (b):.are'unlikely.to release.significan't radio' activity to'the V S

environment,:and.(c) are unlikely to causeLdamage tocthe rea'ctor:sy'stemstori<

the fuel., Operating experience provides'furiher assurance that the" ,

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. experimental _ program ati PUR-1 will be. conducted . safely in the - future, Therufore the.INEL concludes-that reasonable provisions have been:made so' d

'the experimental programs and . facilities do not-; pose' a _significant. risk' to; h the reactor core or radiation' exposure to.the public.

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L11.j RADI0 ACTIVE LiASTE MANAGEMENT l The .PUR-1; reactor' produces essentially no radioactive; waste'during normal operation because of the low power level and limited operating' schedule.

1 i '

11.1 ,ALARA: Commitment The; University's commitment to ALARA (as low as reasonably achievable) was established by the Radiological-Contro1' Committee.in'1951 and'was'recently. .

(November 19,1986) restated tiy the' President.of Purdue University. Purdue.

University. is committed _to a policy of. making 'every reasonable effort 1to' L

keep; radiation. exposures as far below the;specified regulatory.' limits.as readily' achievable. Thus, the' underlying philosophy of-the radiological control. operations of'the University will be to maintain 7 radiation exposures "as low as reasonably achievable" which islin Lkeeping.with the recommenciations of the National / Council on . Radiation Protection and Measurements, the National Academy'of Sciences-National Research Council, and other independent scientific organizations. The principle of ALARA is.

also codified as.part of the NRC regulations,in-Section 20.1'(c) of" Title'10,'Part 20,. Code of Federal' Regulations 'which states that licensees shoul'd " ...make every reasonable effort'to maintain radiation exposures,-

andreleasesofradioactivematerialsineffluentstounrestrictedl areas;, '

as far below the limits specified in the part as practicable."'

11.2 Waste Generation and Handling Procedures.

i 11.2.'1' Solid-Waste ,

The disposal of high-level radioactive waste in the form:of spent fuel iss not anticipated for the. term of_the license. Low-level solid wa'st'e 1 generated at the facili_ty consists of potential.ly contaminated paper and:

gloves and' solid samples produced for experiments and-is usually less than. -

3 1 ft /y. Th'ese wastes are co'11ected in specially marked containers and Lare disposed under the University's;By-Product License. .T.he; water process.

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l system uses demineralized resins to remove impurities from the primary coolant and collect any radioactive ions in the water. These resins are continuously monitored and are periodically replaced. To date, no radioactive materials have been detected before shipment.

11.2.2 Liouid Waste Normal reactor operations produce no radioactive liquid waste except those ,

that might be produced from student or faculty experiments. The pool water l is analyzed periodically for radioactivity. Any detected short-lived activity is allowed to decay and would not constitute disposable liquid j i

radioactive waste. Any wastes generated by research activities are disposed under the University's By-Product License. -

11.2.'3 Airborne Waste j l

The airborne waste that could be present at the PUR-1 facility is composed j of argon-41, tritium, nitrogen-16, and activated dust particles. l l

Argon-41 is produced by thermal neutron activation of argon-40 in the air l

dissolved in the pool water. No detectable traces of Ar-41 from air j dissolved in the water has been observed or are expected at the PUR-1 l facility.

The most likely source of tritium (H-3) is the pool water. From monthly water samples, the level of tritium has been found to be'very low (much  ;

less than the most restrictive maximum permissible concentration).  !

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The main possible source of nitrogen-16 is from the fast neutron l interaction with oxygen in the pool water. The nitrogen must then diffuse j to the surface of the pool before it is released to the atmosphere. In

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normal operation, currents that might be established in the reactor pool j j

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would be very small and with the short half-life (7.14 s), the nitrogen-16 decays before reaching the surface. No nitrogen-16 has been detected in l 1

the reactor room. i i

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ii' 0 ' Th'e reactor room is' sampled continuously lfor'partiEulateslby a C M.

Filters.are changed and analpzed everyf.2. months for ' gross -alpha and-beta activity using-a windowless _ flow proportional' counter; 11.3 , Conclusion

.The:INEL has reviewed the operational 1 history';of theiPUR-1 and concludes ,

'that-no significant. wastes are generated ~as aJresult of the normal

. operation of the PUR-1 reactor.. However, should any waste ever be-

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generated, acceptable provisions.for.the radioactive'wa'ste management activities at:the PUR-l'have been adopted and are expected to continue to

. ensure' consistency with the guidelines of 10 CFR'20 and the ALARA principle, ,

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12. RADIATION PROTECTION PROGRAM purdue: University has a.st'ructured radiation-safety program. Policies for the program are determined by the Radiological. Control Committee-established by the President of the Univ'ersity (See' Figure 12.1). The program is administered by the Radiological Control Officer and his staff.

The staff is equipped.with radiation detection instrumentation to i determine, control, and document occupational radiation exposures a't the reactor facility and all laboratories.using radioisotopes at the University under the'By-Product ^ License 13-02812-04(Broadscope). Routine surveys are-performed of the reactor room and include analysis of the reactor pool and reactor room air. l 12.1 ALARA Commitment The University is committed to the ALARA principle and the-Office of Radiological and Chemical Control makes every' effort to keep doses as low as reasonably achievable (ALARA). All unanticipated or. unusual exposures ,

are investigated by the Radiological Control Committee and the. operations staff to develop methods to prevent recurrences.

12.2 Health Physics Program At present, the University has a full-time health physics st'a'ff consisting of a Radiation Safety Officer, Assistant. Radiation Safety Officer,.two Health Physicists,_ Environmental Waste Technician, and appropriate secretarial support. The Health Physics staff.' performs all routine surveys and is available for consultation in all matters concerning  !

radiation safety.

There are no documented procedures to ensure consultation on matters  ;

i concerning radiation safety as opposed to ensuring' availability of the.

health physics staff. However, specific reactor operating procedures require that radiological . control be present when specified operations are 4

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? performed; Becauselthe Radiological. Control Officer.is'a member of the j Committ'ee on; Reactor Operations (CORO), this ensures the' relationship is more' .than' casual- ( see Figure' 12.1).

q 12.2.1 procedures I

Written procedures have been. prepared that address routine health physics - .

- monitoring.at the University's research reactor facility. These procedures .

I identify the interactions between the operation and. health ~ physics ,

personnel and the' administrative limits to. control exposure. Copies of-these procedures are available to the operational and research staffs and administrative. personnel. ,

12.2.2 Instrumentation ,

The University-has a variety of detecting and measuring instruments :for monitoring potentially hazardous ionizing radiation. Instrument; I calibration procedures and techniques are available to ensure that any I

credible type of radiation' and any.significant intensities will be detected' l promptly and measured correctly. I 12.2.3 Training All reactor-related personnel are required to attend 'a radiation safety 1 training session before they begin work at the reactor. Additional training is provided to those personnel working directly1 with~ radioactive materials to illustrate the ALARA principle and methods to minimize i exposures. Retraining for reactor operators in radiation. safety is;also provided periodically. .;

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Radiation from the the reactor core is the primary source of radiation j directly related to reactor operations. Radiation exposure rates from the ,

reactor core are reduced to acceptable levels by the water in the pool and concrete shielding. -

i 12.3.2 Extraneous Sources j l

Sources of radiation associated with reactor use include radioactive q isotopes produced for research, activated components of experiments, and j activated samples. Personnel exposure from these sources is strictly f controlled by fully developed procedures that employ normal health physics principles.

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12.4 Routine Monitoring ,

12.4.1 Fixed-Position Monitors i l

The PUR-1 has three fixed position remote Radiation area monitors (RAMS)'

with adjustable alarm set points and 1 Continuous air monitor (CAM) in the reactor room. The CAM air filters are changed and analyzed semimonthly.

The RAMS are set to alarm at 7.5 mR/h. This set point was calculated using' the limit set in 10 CFR 20 of 3 R per quarter maximum whole body dose and assuming 10 weeks per quarter and 40 h per work week. Instrument calibration is performed semiannually by Radiological Control. The. RAMS are calibrated for exposure rate and the CAM is calibrated for detection efficiency. -

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12 J.2' Experimenta1' -l 1

i Wipe' tests of exposed surfaces _of the reactor room are made_ monthly. Water samples a're. taken and counted monthly. All-samples and material removed-i from the reactor are checked for levels :of. activity. and wip'e tests made for loose' contamination.

12.5 Occupational Radiation Exposures 12.5.'1 Personnel Monitoring Program Film badges and TLD finger rings are assigned _to all approved reactor 1 person'nel . In addition, self-reading pocket dosimeters and dose-rate instruments are'used to. administrative 1y. keep occupational: exposures below regulatory limits in 10 CFR-20. Students and_ visitors are provided self-reading pocket dosimeters. .

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'12.5.2 personnel Excesures -

1 Approved reactor. personne1' are monitored with' film badges and TLD finger rings, which are read quarterly. If. doses are >100 mrem, a record is.made-l of how and why the dose was received. Exposures are. generally much less- q than 100 mrem, except during the annual fuel plate inspection when a_ finger -

q

. ring dose of >100 mrem may be experienced by the plate _ inspector. Bec'ause the reactor personnel and fast breeder blanket facility. (FBBF)l personnel ,

are the same'and use one. personnel dosimeter-system, it_is'difficul't to determine how much of the " facility"~ dose'is a result of the reactor operations. However, this:is not considered necessary because the actual-dosages are so low. A summary of the 1ast.5.y1whole body. exposure.to reactor personnel is provided below, i

, 4

. a

.{

l 39- 1 q

History-of Personnel Radiation' Exposure at'Purdue University Reactor Facility c

Whole Body ,

Number of Individuals  !

1 Exposure Range (rem) 1981 1982 1983 1984 1985 ., q l

<0.1 8 9 6 8 7 . )

1

>0.1 None j i

1 12.6 Effluent-Monitoring l

'l l

12.6.1 Airborne Effluent j Potential releases of ' air borne effluents are monitored with a CAM in the reactor room.

The CAM has never detected an unexplainable level.'of activity from either the accumulation of particulates or immersion exposure'(i.e., immersion gas).

Analysis of the CAM filter generally indicates a concentration less than 3 x 10 -16 ci/cc for particulates, which is well below the most '

restrictive Maximum Permisable Concentration (MPCa) for an unknown alpha emitter of 6 x 10 ~13 uci/cc.

l 12.6.2 Liquid Effluents >

The reactor generates no detectable radioactive liquid wast'e during normal operation. Before any release-of potentially contaminated water to the j sewer system could occur, samples are collected and analyzed by standard ,.l techniques. I

~ b l;

l 40 i

.j

_ J

1 12.7 Environmental Monitoring j The environmental monitoring program consists of monthly samples of the reactor pool water, semimonthly analysis of reactor room CAM air samples, one TLD placed ir: the reactor room, and one TLD in a classroom above the j ro ph a gr tor r o ir m es r a d for gross alpha and gross beta. Results indicate nothing beyond natural background has been detected on the reactor room air and reactor pool water j samples. Typical exposures have ranged from minimum detectable to 30 mrem.

12.8 Potential Dose Assessments j i

i Natural background radiation levels in the West Lafayette area result in an ]

q average exposure of about 100 mrem /y. The maximum potential nonreactor '

room dose is less than 1 mrem /y based on film badge data in the classroom above the reacter, so there is no significant contribution to the 1 background radiation in unrestricted areas.

I 12.9 Conclusion  !

The INEL concludes that radiation protection receives appropriate support ' l from the Purdue administration. The INEL further 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 Purdue PUR-1 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 the public during routine or off-normal. reactor operations.

41

W i

14. ACCIDENT ANALYSES r

l The consequences of potential accidents in the PUR-1 are limited by the low .I power level (1 kW) at which the reactor is operated. The low power levels {

and low use of the PUR-1 result in a very small accumulation of fission .j products during normal operations and a correspondingly' low level of decay , f heat and' radioactivity stored in the fuel elements. Therefore, some accidents'normally postulated for nonpower reactors, such as loss of tank , f water and handling of irradiated fuel, do not constitute a major' hazard in I the PUR-1. To pose a significant hazard, an accident must generate and release a'significant amount of fission product. 4

- l The licensee and INEL evaluated the potential accident consequences j res'ulting from (a) fuel handling accident, (b) maximum reactivity insertions, (c) reactivity insertions from experiments, (d) loss of  ;

coolant, and (e) failure of a fueled experiment.  !

14.1 Fuel Element Handling Accident l

.)

Fuel element maneuvers are always conducted in the reactor pool. _The fuel elements are removed from the core and moved into the storage space, one.at a time, using a hand-held fuel handling tool. Normally, fuel.is not-removed from the pool except for an annual inspection of a fuel element., A fuel element weighs about 3.18 kg (7 0 lb) in air and only about 2.0 kg.

(4.4 lb) in water. ,

l 14.1.1 Scenario i

Three potential fuel handling cases are considered. Case A assumes the- i element is dropped on top of the core; Case B assumes the element drops back into the core position as it is being raised; and Case C' assumes the , j element falls flat on top of the core. j i

l 1

42 ]

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pg ,

+ , q p -l

, .j v -

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p l14.li2RTechnicalLAssessmenti

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.For. Case.A,.i f a^ fuel element:should-fall'from the'ha'ndling tool during.1ts- j

-transfer underwater it is not heavy enough to cause' any considerable 1 ]

- damage. The most severe damage..likely to occur would be some denting'of-the end fittings because the fuel' element', being:an elongated. object,~would=

' tend to fall-in water in a rather upright position _. For Case B,~no damage 7 1 s

to the element'would be expected if.the element fell back into its original:

,< position' because of the small distance (60.1L em)~ 1t would. fall . - - Al so,' noL i reactivity. concerns 'are present because the core .would have.had sufficient. 4 j

j a

shutdown margin present before fuel' element removal. lFor Case: C, if the element'were to fall flation top of the core,;1ess. force would be placed on:

the element then falling on the endtherefore,

no damage would be expected. 3 i i 14.2. Maximum Reactivity Insertion This hypothetical accident'begins with the step insertion of.the. maximum l, licensed excess reactivity of 0.6% Ak/k into'the critical . reactor operat.ing

]

at the maximum licensed power of 1.0 kW. .)

14.2.1 Scenario- f W

i Two general case are considered: Case'A. assumes that the safety.contro11 2

j circuitry is operating and is activated by.the reactor power exceeding the : l 120% (1.2 kW) power set point. SCRAM redundancy is provided under.theseL '

conditions, because the period would be about.1 s and the"short period tr p. l would also initiate a reactor scram. ,0uringtheiscram,itLisassumedthatL -l the most reactive shim-safety rod,is stuck- and-the :second shim-safety rodL -j drops into the' reactor providing 'a shutdown: margin. offl.8% Ak/k.c It is - l, assumed that all negative reactivity' provided by~ the' shim-safety ' rod-is~ l added as a linear ramp in I s, the time specified in the? Technical- 'f

.. Specifications.for the rods to be fully inserted. ,

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la l

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Dy ...

. Case-~B considers the worst. failure'of the' safety system where both rods fail-.to scram and the reactor is controlled only by the negative .

temperature feedback of the moderator. In this ca'se, the calculated value of the moderator coefficient is used because it is the lessor of the two

values '(calculated vs. measured). reported in Table 4.1~ and thus assures conservatism.

'Th'e results for Case A show reactor power would rise to'about 2.8 kW within-55 msTfollowing the step insertion of 0.6% Ak/k, primarily due to the prompt jump. ' The 120% (1.2 kW). trip, however, would initiate l scram within l p about.3 ms.of the step insertion. Between this 3 ms scram initiation-and j the' 55 ms peak power, the scram reactivity is insufficient to control the {

power rise; however, once the prompt-jump to'2.8_kW is completed, the scram reactivity.. becomes. controlling and the PUR-1 is quickly shut down.- The  !

l calculated results for Case A are summarized in Table' 14.1. As'shown, the power'is above 'its initial value for less than 0.5 s and the resultant l energy release is about'O.5 kW-s. This results in a negligible temperature- J I rise in the PUR-1.

[

The results for Case B, where no control rods are inserted and the negative l I

moderator temperature coefficient is considered as the'only shutdown ,

1 mechanism,-are presented in Table 14.2. In.this case, the analysis shows;

that the reactor power rises over a period of about 3 min to a level of j about 380 kW. At this time the 0.6% Ak/k step reactivity insertion causing  !

the power rise is completely quenched by the . negative moderator, temperature .,

coefficient. At' this point in time or before, it.is reasonable _to assume I another' scram or operator-initiated scram occurs:and the power, level.is-quickly reduced below 1 kW. These results are consistent >with-a number of-'  ;

the excursion experiments performed at the BORAX AND SPERT Facilities.

Some of the results of the SpERT-1 experiment using the DU-12/25; core are applicable to the analysis of the PUR-1 reactor because j

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i TABLE 14,1. RESULTS OF THE POWER TRANSIENT' ANALYSIS WITH RAMP'INSERTIONL j

. OF CONTROL ROD j

,> ;l

' i

'd i

Case A'  ;

4 l Initial Conditions.

a Power . level ..(kW) . . 1' 1 j

Temperature ('C) ~20 ,

l Flow- rate (kg/s) . OL 1

-4

. Input Values Reactivity (step) added.(% Ak/k) 0.6' .

SCRAM' time (s) :1.0 .

1 q

SCRAM reactivity ramp (% ak/k/s) -2.40-.. )

Calculated Results Maximum power level (kW)' -2.85 4

.... 1 Elapsed time to maximum power (ms): .55.

R Elapsed time while P>1 kW-(ms)- c275 Total energy released while P>l kW (kW-s).. ..

t0.5 j 1

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., J

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i

~ TABLE 14.2. RESULTS. OF POWER TRANSIENT. ANALYSIS. WITH NO. CONTROL RODS

, Initial ~ Conditions Case B1 Power. level: (kW) 1 Pool temperature (*C)- 20 1

20.6' Reactor . temp lerature .(*C) - ,

-2 Moderator. temperature' coefficient (% Ak/k/*C)

-2.10 10 Input Values Reactivity (step).added (Ts^Ak/k) ' 0.6l Calculated Results- ..i.

Maximum power level (kW) . 380:

/

Elapsed time to max (min) .~3

- Total'energyreleasedin3' min-($-s). '~5'4l -

+

. ., \

1. -

t m, .

.6-

'oi

ro x

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theifuel geometry and composition are very similar., A comparison of the j

reactor characteristics 1s-given in Table"14.3. A' series of self-limiting -

power excursion-. tests was carried outLin SPERT-1 using five core loadings. j The input variable. referred to in these experiments was the' reactor pericd, j induced.by a'st'ep-wise reactivity insertion; The results'of'the- i calculations.of 'the PUR-1 experiment:are consistent with the observed l results of the SPERT-1 experiments using long periods'ontthe order of'1-3 s

~

~

. wherefabsolutely no. fuel damage was-observed. In' fact,'the 3 PERT-1 A

e'xperiments showed,that the fuel could withstand transients with periods'.as-short as 14 ms with no apparent damage.to the fuel.

.e

. t Because< the results in' Table.14.2 are consistent with long period '

SPERT-1 tests where ne fuel failure was' observed and because the availsble j excess-reactivity is much less than was added to the SPERT-1 du, ring the. e

!i short period tests (as low as 14 ms with no fuellfeilure), cit is conchide'd t L

30 that even> during the very unlikely event of a safet'y system. failure during -

the maximum credible reactivity accident, the fuel would not melt and mo. '

fission. products would be released. j

'14.2.2 Technical Assessment The INEL considers that the reactivity insertion accidents considered by, -)

the Pu'rdue are representative of the most severe transientsLthat can credibly. occur at'the PUR-1. The.INEL has reviewed'the licensee's accident' assumptions and'calquiations and finds them conservative, reasonable, and- I acceptable. The INEL, therefore, concludes that it is uniikely that a- s i

cre'dible nuclear excursion in the PUR-1 would lead to fuel meltinglor cladding failure and,. consequently, such a transient would not pose:a 0-significant hazard to the public.

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. , TABLE 14.3..: COMPARISON.0F .IMPORTANTh FUEL DA.T. A 1

PUR-1' SPErtT - ..

.y.

. Geometry- s P1 ate' .

~ Plate" i+

$ E. Length (cmf 61' -

61L - .

'M

~

y < '

Wi.dth[(cm) 5., .7,0 7. . 6 ~

Thi ckne's s ' ( cm)'. 0.'15 ' . .0.15i ,

Watergap.(cm) 0.53 :0.45l k ,. 'i.

4 L  : Fuel

~ iMateri al '. U-Al 'U-A1:.

n Enrichmunt (*.')- 93

^

'93 Thickness (mm) 0 ~. 51 - 0.51 L ,

Cladding' .

~

Material All ' Al '

Thickness (mm).. 0.51< i

0.51.

s m

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I '

<) 6 t 0 j t. (

s s s

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sw 1 -

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f1 1

14.3 Floodino of an Irradiation Facility and Failure of a Moveable Exoeriment A sudden replacement of a voided, i.e., air-filled, space next to the core by water, such as the. flooding of an experiment . tube, would cause a i

{

stepwise reactivity insertion, its magnitude depending on the void volume J e being replaced and its position relative to the core. Experiments have ]

'shown that flooding of the 12.7 cm irradiation tube located outside of the 4 graphite reflector element F6 with water adds 0.3% Ak/k to the core.

Another identified mechanism for suddenly adding reactivity to the critical-

) core at the PUR-1 is the failure of.a movable experiment. Similarly, the  ;

maximum stepwise reactivity addition is limited by Technical Specification Requirements to 0.3% Ak/k.

14.3.1 Technical Assessment It is shown in Section 14.2 that a sudden reactivity insertion of 0.6% Ak/k into a critical core of the PUR-1 can be tolerated with a sufficient safety margin. Therefore 0.3% Ak/k would be enveloped by that analysis.

[14.4 Loss of Coolant Accident U

\T)e reactor pool is designed to prevent unintentional drainage. The pool

, "i s constructed of a stainless steel liner and set in a second steel tank wish the interstitial region f.111ed with sand. The tank rests on a i 4  ;

7 g ,

concrete pad about 4.6 m below the floor of the reactor room, which is in the. basement of the building. The pool has no drains or coolant pipes that ff

• could open or break. Therefore,-a sudden loss of coolant is considered to -

p be extremely unlikely. Furthermore, if the pool drained instantaneously,

. while the reactor was operating, the loss of water (moderator) would shut.

the reactor down.

49

,y l

- 14.4.1: ' Sconario

' Any reasonably conceivable leakage of. water from the reactor pool is l-y expected to be rather slow. In'such a case,: the radiation area monitor t mounted directly above the core would detect any additional radiation coming from the core due to.a' decreasing pool water level. Because the pool water level-is checked during; daily. routine operations, any .

. significant leakage would be detected before reactor.startup.for the day.

14.4.2 Technical Assessment Although extremely unlikely, if the core were,to'become immediately

-uncovered.following a 1 kW' power run for 24 h, heat transfer would occur by

-natural: convection of ambient air. The. decay. power of the PUR-1

- immediately after shutdown from full power (1 kW) is about 65 W. The decay

- power rapidly decreases, being about 35 W after.1 minutelof decay'and l 0.87 W in 24 h. For this case, the amount of heat removed.is proportional

.to the cladding temperature. No significant_ temperature increase of the.

fuel would be expected because heat transfer would' occur first.to the

-aluminum fuel assembly' cans and then by convection'to the air. Even assuming adiabatic conditions for 24 h,.the fuel. temperature rise.would?  ;

only be 9.5 C.

14.5 Maximum Hyoothetical Accident In this section, an analysis is performed'to assess the hazard associated..

with the failure of an experiment where fissile-material has been irradiated in the reactor. In the scenario of this accident, it' is assumed that a capsule containing irradiated fissile material = breaks and a portion of the fission product inventory becomes airborne.. The consequences of'the-release.are analyzed for both the reactor staff.and general'public.

- Because the potential impact of this postulated. accident _is greater than'in -e any other accident analyzed, the failure'of a fueled. experiment is designated ~as_the' maximum hypothetical accident of the PUR-1.

50 u_m_z__m .m_ .. .. . . . . . . . _ . . . .i s . . .. . i..si...ide.n...L

. .,,..;..i....,,,..o...

+-

.. . ,,ii.. .- ,,, r. ..i.,,. - -.

-w -,- . , --i .--.. - ..y

14'.5.1' Scenario-In this analysis the' consequences of a failed experiment generating 1 W were studied. The capsule containing the experiment is assumed to break as it is removed from the reactor. The fission products expected to become airborne are the noble gases and elemental iodine. Other fission products and actinides are not volatile at the fueled experiment temperature (which is essentially at room temperature). All of the noble gases and 25% of the radiciodine are assumed to be released, which is consistent with Regulatory Guide 1.25. No credit for the absorption of iodine in water is taken because of the designation of this event as the maximum hypothetical. ,

A conservative assumption that the irradiation time was infinite was made in this analysis. Therefore, the fission inventories used in the analysis

'for some long-lived radionuclides, e.g. , Kr-85 or even I-131, are overly conservative. Furthermore, it was assumed that the fission products are instantaneously. released and uniformly distributed in the 424 M3 reactor room air volume.

14.5.2 Technical Assessment The calculated saturation activtby for each respective radioiosotope and its concentration in the reactor room after experiment failure is shewn in Table 14.4 for an experiment of 1 W. This experimental specimen power level corresponds to the amount of fuel that could be allowed by the relevant Technical Specifications. Also shown in this table. 3re the calculated dose rates for the whole body, skin, and thyroid. Under these, conditions any one of the RAMS would cause an automatic reactor shutdown and audible and visual alarms in the control room. From past experience, the reactor building can be evacuated within 1.5 min. Therefore, it is assumed that the exposure time to the members of the reactor staff is 1.5 min which results in radiation doses of: whole body--17 mrem, skin--11 mrem, and committed thyroid--830 mrem. These doses are <10% of

. the dose limits as stated in Regulatory Guide 2.2 for doubly encapsulated experiments.

51 i.

n This radiction exposure is .less than the limits (410% of the equivalent'

annual. dose stated in 10 CFR 20) established in-the Tc anical Specifications, Section 3.5.f.for a single. encapsulated experiment. This experiment corresponds to the irradiation of 1.1 gm of U-235 in. the mid plane of the isotope irradiation tube located in position F6. ]

For the radiation calculations outside of the reactor building it was

~

. assumed that'all fission products released in the reactor building would leak out:within 24 h. Because the reactor' room.does not-have.any windows ...

and has only a few doors and emergency procedures call for turning off the"

-air exhaust system, the leak rate assumption is considered. to be reasonable.

Other conservative assumptions in the analysis-include:

o No radioactive decay (no. decrease..in source-strength) o No building wake effects and horizontal pl'ume meandering o Release at ground level versus out the 15.2 m smoke stack o Average wind speed of 1 m/s versus actual average.of 3.4 m/s.

The calculated dose rates at 100 m (distance assumed;as a reasonable-

~

distance evacuation and control could be accomplished with 2 h) for an experiment-power of 1 W are as shown in Table:14.5. If it:is assumed that an individual is located at this point for 2 h following the ; fission product release from a postulated experiment' failure, thenlhis/her '

resulting radiation dose to the whole body would be 0.51 mrem and 20 mrem committed dose equivalent to the thyroid. These doses are less-than 10% of the' dose limits as stated in Regulatory Guide 2.2.

14.6 Conclusion ,

The~INEL has reviewed the credible potential accidents for.the PUR-1. 'On a the' basis of thisLreview, none of the postulated accidents are-expected to-t i

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-Q >=  ;

@. N. M. m.4. s m. N. A. C. . m. A M. c. %. N. M. D. .

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t 2.= g .a3-'g

-Q --

E$ = M M A @ P= C .'. <

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• • N M X e M c eM P=

fu1 M MtoMmM. M *= M *= M *=M s=M.enM.** em M' ,

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? rele.,aseisignificant fission 1productsdo'thetenvironment or'excesshe doses

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,- 1 .

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.Lto th'e7 reactor personneli oritofthepdblic..;Therefore,the11NEL1 concludes': ;c that:.there,.isfreasonableiassurance that:;any accident resultingLfrom, 1

m continuedoperation'~of:the'.lPUR-1 Mould'not.posezasignifica'ntr.isk'tothei

~

fhsalihand:s'afety'ofthepublic. .

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.,.[

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> / _ l m

' t N I

b i

1

-TABLE'14.5. DOSE RATES AT 100 METERS FROM A FAILED. FUEL EXPERIMENT l

.(POWER = 1 W) j m

DR gamma DR-beta DR-thyroid ,

Isotope: -(mrem /h)- (mrem /h)- -(rem /h)

I-131~ 1.95 E-03~ 9 06 E-04' 6.98 E-03 I-132 1.78.E-02 2.91'E-03 .'6.49 E-05 l

.I-133 5.01 E ,3.88 E-03' 2.51 E-03  !

.I-134 2.28 E-02 4. 67 :. E-03 '1.71 E-05 I-135 1.65.E-02 2.50.E-03' 5;41 E-04'  ;

~

Kr-83m '1.21 E-05 -4.20'E-05 .--

.Kr-85m 1.52 E 1.96 E-03 --

-: Kr-85 '4.23 E-06 3.90 E-04 ---

Kr-87. 2.17 E-02' 1.45.E-02 '

Kr-88 4.30 E-02 '7.36'E-03 --

Kr-89 5.02'E-02 3.64 E-02 --

.3 .

Xe-131m 4.01 E-06. 2.45 E-05: --

Xe-133m 3.48 E-04 1.45 E-04 --

)

-Xe-133 1.30'E-03' 5.54 E-03 --

Xe-135m. 5.91 E-03 . 1.19 E-03 --:

Xe-135 1.02 E 1.17 E-02/ --

Xe-137 6.31 E-03 5.03-E -- o Xe-138 4.85 E-02 3.08 E --

l 2.53 E-01 1'.73 E-01 1.01 E ,

o 4

t

.}

55 i

r k #.'.I 1

19. REFERENCES

1. American' National Standards' Institute /American Nuclear Society .

(ANSI /ANS), 15 Series.

2. American Nuclear Society ( NS) Standard 5,1 " Decay Heat Power in Light Water Reactors." 1978/

3. American.Nuclea'r Society (ANS), 15.1, " Standard for the Development of Technical Specifications for Research Reactors," 1982.

4. --

,L ANS .15.4, Selection and Training of Personnel for Research ,

Reactors,".1977.

5. -- , N401-1974/15.6, " Review of Experiments for Research Reactors," 4 l

1974~. 1

- 6 '. -- ,15.11, " Radiological: Control at Research Reactor Facilities'," 1977.

. .r

- 7. Regulatory Guide' 1.25, Assumptions Used. for Evaluating. the Potential j Radiological. Consequences of a Fuel Handling Accident, March 1972. i f

.a

8. Code of Federal-Regulations, Title'10, " Energy," U.S. Government 1 Printing Office,. Washington, D.C. l

9. Regulatory. Guide 2.2, Development of Technical Specification for. 1 Experiments in Research Reactors, November 1973.

10. J. R. Dietrich, Experimental Determinations of' the Self-Regulation' and'.

Safety of Ooerating Water-Moderated Reactors, Argonne National.

Laboratory,.A/ Conf.8/P/481, June 30, 1955.

11. S. G. Forbes et al., Instability in the SPERT I Reactor, Preliminary Reoort, I00-16309, October 1956. I

12. R- N. Miller et al., Reoort of the SPERT I Destructive Test'Procram on an Aluminum, Plate-Tyoe, Water-Moderated Reactor, 100-16883, June 1964.

13. W. E. Nyer, S. G. Forbes, F. L. Bentzen, G. _0. ' Bright, F. Schro' der,.

and T. R. Wilson, Exoerimental Investigations of Reactor Transients, ID0-16285, April 1956.

1 I

56'

7 m

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, A.,om No . 4 ,- r, c. <. N. . .,,

. g,",' g', sieUOGRAPHIC DATA SHEET EGG-NTA-7527  ;

i 848 INstavCTIONS ON fNa atvente 2 TITLS AND SUGfif L8 J LGAV4 SLANa

. TECHNICAL EVALUATION REPORT FOR THE RENEWAL OF THE-

. OPERATING LICENSE FOR:THE PURDUE UNIVERSITY REACTOR

. oati a , oar co ario i momrw reaa l.

• Auraca's> May '1987

' . oArea., oar,muso W. R.~ Carpenter moNTN vsAa C. H. Cooper j May 1987 ]

7 i P0 amino CaGAMi4Ar 0N NAMG ANO MAsLING A00#ta tsammeto C.mA 4. pit 0 JECT,7Asaaucast WNaf Nyusea NRR & I&E Support j

.EG&G Idaho, Inc. . iN oa oaANvmu m en  ;

P. 0.-Box-1625/

-Idaho. Falls, ID -83415 to 8,0N80meNo onGANi4Aviose asAME A8e MAaLING A0.nen ,ssemeto c.ms s te, fveg 08 atP0af OfficeofNuclea$ReactorRegul_ation V. S.. Nuclear Regulatory Commission

'"'"**""""""~'**"'

Washington, D.C. ;20555 o su,,u=NrAav oru is Aosta ACT (Jd0 weren .r 'sesJ . .

~ This Technical . Evaluation Report for the application, filed. by-Purdue: University . I for the renewal of Operating License No. R-87 to continue operating its research reactor; i 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 Purdue University in o West Lafayette, Indiana. The INEL concludes that the reactor can continue to'be l operated by Purdue University without endangering'the . health and' safety of the public.

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