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Evaluation of Idaho State University Subcritical Assembly
	ML20236R032
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Site:	07001374
Issue date:	12/31/1973
From:	Jotikasthira P; IDAHO STATE UNIV., POCATELLO, ID
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{{#Wiki_filter:g..h.,%- l l l THE EVALUATION OF THE IDAli0 STATE UNIVERSITY SUBCRITICAL ASSEMBLY i l. l l by j Prinya Jotikasthira l A thesis submitted in partial fulfillment of the requirements for the degree of f4 ASTERS OF SCIENCE IN NUCLEAR SCIEf1CE Af!D Ef1GIllEERING 1 i IDAll0 STATE Uf11VERSITY 1973 u 9007210197 990715 l PDR ADOCK 0700 3 4

fa y f ~ l To the Graduate Faculty: The members of the Committee appointed to examine the thesis of Prinya Jotikasthira find it satisfactory and recommend that it be accepted. d N.N C ts Major Advisor l L t-f. f. I $Wkf~- l C. ..h AA NAAl d ] 4 I --iei.----i-i-ei

... ' / ACKNOWLEDGEMENTS Among the many individuals who generously offered their help and suggestions on my thesis I wish to especially acknowledge and express my sincere gratitude to the following: Dr. A. E. Wilson, the chairman of the Department of Architecture and Nuclear Science and Engineering and also my advis~or during my l length of study at Idaho State University, who assigned my research i l problem and who aided me with much of the experimental work. f l Jeanne~ Wilson, wife of Dr. A. E. Wilson, who kindly volunteered to correct the gra:nmar and format of my thesis. Dr. P. T. Charyulu, Associate Professor of Nuclear Science and Engineering, who patiently tolerated my presence in many of his classes over these past four years and who so helpfully served as my unofficial advisor in coordination with Dr. Wilson. Mrs. Malati Charyulu, wife of Dr. Charyulu and a former instructor of mine at Idaho State University, who not only was especially kind and understanding while I was in her classes, but further took the trouble to review the final copy of my thesis. l i t l l

f, i p m V [ I t i 1 TABLE OF CONTENTS 1 Page I. INTRODUCTION.................... 1 II. HISTORICAL. BACKGROUND................ 3 III. DESCRIPTION OF COMPONENTS OF THE SUBCRITICAL ASSEMBLY.................... 5 The Core Tank................. 5 Thermcl Column 6 Fuel Plates.................. 11 Fuel Spacing Assembly............. 11 Core Support Devices.............. 24 l Water Handling System............. 24 Cri ticality Al ana............... 31 Emergency Shut Down System. 31 IV. DESCRIPTION OF THE SUBCRITICAL ASSEMBLY CORES.... 33 S-1 Core.................... 33 S-2 Core.................... 34 S-3 Core.................... 34 D-1 Core.................... 34 D-2 Core.................... 34 T-1 Core.................... 35 V. MATHEMATICAL MODEL................. 37 DISNEL Input.................. 39 Result of the Calculation........... 43 VI. APPROACH TO CRITICALITY EXPERIMENT......... 49 General Theory................. 49 Procedure for Approach to Criticality..... 51 Safety Considerations............. 57 Instrumentation................ 59 Result of the Criticality Approach Experia.ent. 62 VII. EXPERIliENTAL PROCEDURE FOR THE USE OF THE IDAHO STATE UNIVERSITY SUCCRITICAL ASSEMDlY..... 73 Introduction................... 73 l Sa fe ty it ul e s.................. 74 l Procedure for Operation.. 76 APPENDIX......................... 73 REFERENCE 89 j l

i/ LIST OF TABLES 1. Specification of the subcritical assembly cores... 36 II. DISNEL 19 group energy structure.......... 38 III. Experimental data from approach to criticality Experiment of T-1 Core (Run 1)......... 64 IV. Experim?ntal data from approach to criticality Experimr.t of T-1 Core (Run 2)......... 65 V. y radiation observed in Run 2 of T-1 Core...... 67 l VI. Experimental data from approach to criticality Experiment of S-1 Core............. 68 VII. Experimental data from approach to criticality Experiment of S-2 Core............. 70 j l l l l j I { l i

~ 88 r. 1 l LIST OF FIGURES i Figure Dage 1. The core tank with the thermal colurn. 7 2. East view of the thermal colunn............ 8 3. North side of the thermal column........... 9 -4 North side of the thermal column........... 10 5. The subcritical assembly fuel plates fabricated by H & C Nu cl ea r Co................. 12 6. The single plate spacing assembly........... 13 7. The double plate spacing assenbly........... 14 8. The triple plate spacing assembly........... 15 9. The top single plate spacing grid........... 16 10. The bottom single plate spacing grid......... 17 11. The top double plate spacing grid........... IP 12. The bottom dcuble plate spacing grid........ 19

13. The top triple plate spacing grid........... 20 14.

The bottom triple plate spacing grid......... 21 15. The top view of the fuel spacing frame........ 22 16. Side view of the fuel spacing frame.......... 23 17. Bottom enre support structure............. 25

18. The core side support brackets............

26 19. Core suoport devices in the core tank......... 27 20. Schenatic of the water handling system........ 29

(.J c>, 'r LIST OF FIGURES (cont.) Figure Page 21. The wiring diagram of.the. water handling system.... 30 l l 22.- K-eff vs. nunber of fuel plates. for "S" type core from 1 DIS'!EL calculations............,.. 45 23. K-eff vs... number of fuel plates for "D" Type core from DISNEL calculations................ 46 24. K-eff vs. number of fuel plates fcr T-1 core from DISNEL calculations............... s :' 25. Comparison between DISHEL code calculation and Rutgers calculati n.................... 48

26. -Components of number 1 detector system........

61 27. Conponents of number 2 detector system........ 61 28. Components of number 3 detector system........ 61 U '29. One over subcritical multiplication vs. number of fuel plates in T-1 core............. 66

30. ' One over subcritical multiplication vs. number of fuel plates in the modified S-1 core.......

69

31. One over subcritical multiplication us. number of

. fuel plates in the S-2 core........... 71

32. The modified S-1 Core.................

72 L _i I I l I

I j.? i,c 'F l I. INTRODUCTION l 7[ [ In a nuclear reactor, the ratio of number of fissions in any one a;.

f generation to'the number of fissions in the i m ediately preceding n

2 generation is defined as the multiplication factor, K. If the nul- .J. triplication factor is exactly equal to unity the reactor is said to be critical and a neutron chiin reaction within the system will be sustained at a constant rate. If K is less than unity the system is in a subcritical state and the chain reaction cannot be self-sustaining. In order to maintain a self-sustaining chain reaction for a given composi-i tion of fuel and other materials, a certain amount of fuel. compatible with the geometry of the arrangement of the fuel-material composition, is needed. The minimum amount of fuel necessary to maintain a self-sustaining chain reaction is called the critical mass, and the corres-ponding geometry of the arrangement of fuel and other materials is called the critical geometry or critical size. A suberitical assembly or a subcritical reactor, as it is some-times called, could therefore, be considered to be a portion of the core of a critical reactor. It usually consists of a geometric arrange-ment of nuclear fuel, moderator and possibly coolant. The assembly is small enough in volume, or contains such a small amount of fuel that it cannot sustain a chain reaction without the aid of an external neutron source. The external neutron source offsets the loss of neutrons (by leakage) from the sides of the assembly, and thus a constant neutron L

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h h. Page 3 Mu ) % q' R'y !W I.I. HISTORICAL BACKGROUND ]h ) The history of the subcritical assembly can be traced for three billion years, at which time the oldest known rocks were formed. (1) }y At such time natural uranium contained six percent of the uranium 235 i isotcpe, compared to 0.728 percent present today. It is quite likely that a water moderated system with six percent U235 es ' fuel' existed h d !Vi in nature. However, man-made subcritical assembly began only thirty-LI:l' four years ago when Halban, Toliot and Kowarski of France and Sizilard, Q RL and Fermi of Columbia University, started generating fission neutrons. r!!g 7 9 Haban's neutron source amplification experiment utilized heavy water f I H ! 1 as moderator and thus laid the foundation for the D 0 systems. Initially f] 2 nh Y l Fermi worked with light water and later on, turned to graphite as mod-i i erator. The assembly in which Fermi observed a ten percent increase in t i ik L ! i neutron production over the source density, consisted of 200 Kg. of I 3 U03 8 encased in fifty-two tinned iron cans. These cans were 5 cm. in f h diameter and 60 cm. in length. The 60 cm, cylindrical lattice was sub- ) merged in a 540 cm cylindrical tank which contained a solution of f 3 a MnS0. Fermi made the following statement in a paper describing the 4 experiments with the above assembly. id "From this result we may conclude that chain reaction could be 1 % ltJ maintained in a system in which neutrons are slowed down with-r- @d{ out much absorption until they reach thermal energies and are th"1 rostly absorped by urar.ium rather than by another element. It

nains an open question, however, whether this holds for a Lj system in which Hydrogen is used for slowing down the neutrons."(2) h-jl

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I Page 4 i i Shortly after, an " Intermediate Pile" was assembled at Columbia y. University. In 1954, New York University started to construct their subcritical assembly named "Pickel Barrel." The original name surely must have come from the twenty-five dollar wooden barrel. An Italian olive importer sold the University the assembly tank and coopers were hired to assemble it. It seems that the barrel had to be taken apart because it was too large for their building entrance. From this date on Universities around the country began building their own subcritical assemblies. Most o.f them utilized natural uranium as fuel. Only two institutions constructed enriched uranium subcritical assemblics. The University of Texas designed a homogeneous enriched assembly. Rutgers, The State University of New Jersey, designed a heterogeneous enriched system which " traveled west and settled" at Idaho State University. At Idaho State University, this suberitici1 3 assembly will be referred to as the Idaho State University subcritical assembly. I I l

ow h v ?., an ~h bg Page 5 1 m l N( III. DESCRIPTION OF C0!!PONENTS OF THE SUBCRITICAL ASSEMBLY The Idaho State University subcritical assembly was originally designed and assembled by Dr. F. J. Jankowsky and his associates at o Rutgers in the year 1961. This assembly, unlike most others, consists of a water moderated and reflected enriched uranium system. The enrichment mckes it possible to have less fuel mass. This assembly has been designed to accommodate a variety of core configurations with (( u. the same type of fuel plates. The subcritical assembly was reassembled fj .. v at Idaho State University essentially as it existed at Rutgers.. A Ij a water handling system, a shutdown system, a working platform, and a 1 VU core lifting device were added to the original components. Also, in j$ il accordance with the United States Atomic Commission regulations, a M NJ gamma criticality alarm was installed for safety. M ih The Idaho State University subcritical assembly consists of an v q1 hj aluminium core tank (placed on a graphite thermal column), three sets M of fuel spacing assemblies, core support devices, and one hundred and ny fifty' fuel plates. A } The Core Tank N (Q@jl, A cylindrical aluminium tank is used to house the core and the + {h{p > moderator. This tank is large enough to accommodate additional water around the core to act as a reflector. The tank wall is made of a 1/4 in. .A aluminium sheet, rolled and welded. The bottom of the tank and the 1: removable lid for the top of the tank are fabricated from a 3/8 in. I

d y'. g Page 6 f.p aluminium sheet. The tank is 36.0 in in diameter and 39.0 in, high Other dimensions of the core tank are shown in Figure 1. The aluminium top can be locked to the core tank and thus the core tank can be used s I to store the fuel plates. L Thermal Column [ The Therrral Column is used primarily to house the neutron sources I h and supply the core with neutrons of thermal energy. The thernal column is formed by stacking eight criss-cross layers of graphite blocks as shown in Figure 1. Each layer consists of twelve blocks i l arranged side by side. Each graphite block is 4 in, square and 48 in. l lilj long. The total weight of the thermal column, comprised of ninety-six blocks, is approximately 4500 lbs. One of the blocks, labeled (y8, x1), [, l 4 was accidentally broken into two pieces prior to arrival at Idaho f, i State University. The block labeled (y2, x6) is in two sections made b out of two identical pieces each 24 in. long. Eleven other blocks y are specially made to accommodate neutrons sources, or 1/4 in. fission Qti ]Nl counter and enriched U235 foils. Figures 2, 3 and 4, show the original positions for these special blocks. In Figure 2, the block (y7, x6) is shown with two enriched uranium foils. Next to it is a blank, a h block without special machining. Directly below, a block (y5, x6) J .n is shown with the fission counter in place. Figures 3 and 4 show eight $1 ,I other special blocks. Blocks (y6, x6), (y6, x7) and(y4, x5) are used M to house cylindrical neutron sources of 1.02 in. diameter and 1.45 in. i! l high'.. Blocks (y3, x6) and (y4, x6) are mechined to hold foils. i l. M u

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l: } p I. ! W Page 11 Each of these eleven blocks has a 3/8 - 16 tapped hole to facilitate handling. Since the positions of any two blocks can be interchanged, a versatile source and measuring system is available. Fuci Plates A total of 150 fuel plates are provided with this assembly. Each plate is.08 in. thick, 3.0 in, wide, and 26.0 in long. The plates -are constructed with a fuel bearing region of 0.04 in. thick, 2.75 in. wide and 24.0 in. long, clad with aluminium. A typical plate is shown l in Figure 5. 4 The fuel region consists of a uranium-aluminium mixture with the f 6 uranium enriched to 20, U235 by weight. The total amount of uranium fllk used is 7614.79 gm. with 1510.27 gm. being U235 It is assumed that fjf;

qu all the plates are identical and that the uranium is distributed equally gj Qi among them.

Fuel Spacing Assembly [g s The fuel plates can be arranged in three different lattice config-if hf' urations using the appropriate spacing grids. The fuel plates may be ~ 4 i used singly to form a quasi-homogene'ous lattice, or in groups of two 1s or three to obtain a more heterogeneous lattice. Each lattice requires a set of grids which is -bolted to a comon frame to complete the fuel h[y spacing assembly. The three fuel spacing assemblies are shown in ii ti Figures 6, 7 and 8; the dimensions of the grids are shown in Figures 9 to 14; Figures 15 and 16 give dimensions of the frame. A detailed l l description of each type of core is discussed in the next chapter. l w--________--_____

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J Page 24 k Core Support Devices A honeycomb structure shown in Figure 17, is used to support the core at the bottom of the tank. To secure the core in its proper place relative to the sides of the core tank, brackets are provided which are bolted to the rim of the tank. This arrangement is shown in Figure 18. Figure 19 shows the picture of both devices in place, j Water Hardling Systen Distilled water is used in the subtritical assembly as a moderator and reflector. The water is stored in three inter-connected poly-ethylene drums of 55 gallon capacity each. At the beginning of each i experiment the water is pumped from the drums into the core tank. At the end of the experiment the water is drained back into the storage 1 4 drums. Schedule 40 polyvinylchloride pipes and fittings are used. Half inch pipe is used to feed water into the assembly tank and three quarter inch pipe is used to return the water to the storage drums The pump chosen employs a magnetic drive mechanism with a maximum flow capacity of 940 gallon per hour. Ideally, a pump with the above flow rate will i fill the core tank within thirteen minutes. However, because of pipes, i fittings, valves and the 5 ft. head in the system the time required to fill the tank has almost doubled. The present system requires twenty-five minutes to fill the core tank with 33 in. of water. A"normally open" solenoid valve i:; installed to control the return flow of water to the storage drums. "Normally open" inplies that the return flow is cut off only when the coil is

A 4 : 5. l f, Page 25 5 g i' l tj [i [,hk'. I 1 o 'A g YY ,,..nay - a, T (Y Y"' 'Y(n.^ Y YYYT,' e r T qq' f% I L- ! II ? b; {i ) l ll f-i l d- [ ~' E".!., Q d l, d I, Figure 17: Bottom core support structure. 1 .;. r, s 8 9 ...... _ _ _. - - -. A

i e c .1 Page 26 gjy n. ?k I 3 O @'*3/c g o u.. a u k j'), .l l.

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3 core tank a I l Figure 18: The core side support brackets. L __

>l 9 i l Page 27 4 t l .e i A 6 F 4g, [, s; I . s .y.,f 8 Y Fa. 5 ,h, 4 'I ..A i J.- psi :a-v ,a ,.s.,, ~ ;- .x W .N ,,f \\ A .y) 1

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7' .- y a %'6 i;.t \\ */ t y 4 4 7 - G I 1 I .V. Figure 19: Core support devices in the core tank. t M

Page 28 I energized. Under operating conditions, the coil is energized and the valve is closed. In case of power failure the solenoid coil is de-energized and the valve is automatically opened to ensure the safety i of the system. Two tee junctions are provided so that a demineralized can readily be connected in parallel with the pump if necessary. The water level in the core tank is indicated by a pressure gauge calibrated in inches of water. This gauge is located at the edge of the graphite pedestal. The operating level of water in the core tank has been set at 33 in.; additional water will automatically flow back into the storage container through the overflow line. A complete schematic of the water handling system is shown in Figure 20. J The power to the solenoid valve and pump is supplied through normally open contacts of the double pole double throw relay. The t[ 'yj normally open contacts of this relay are closed by the 18 volt-AC

J signal from the criticality alarm.

During alarm condition the 18 VAC ' Ild signal is turned off by the alarm, thereby cutting power to both solenoid i I, valve and pump. A second relay is used to insure that power does not return to the pump and solenoid upon power return follcuing a power f failure. This relay is wired as an electrically latched switch and must l] l be reset following any power interruptions. A mechanical switch is provided so that the pump may be turned off when the core tank is filled. The complete electrical wiring diagram for the system is shown'in Figure 21. h

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i i ll lI ))ll).!jl !-i1l1 ll!;!l !1l:l!l. 4- [- m g x 2W uo 7 = y =. t hg t i 3 t. P t t s e f T i m t I P - r t E l e a t r o e s t APc y i h p n e mi m t o s P P w g s e n .vl i av l d d n h i a c o t h o i e L w lo r s hM-C v S M e A t R g a d t [./ w S N a r v e F e h to t r I y ]l t r f_ I - t u o o .t m A-f a O r N g-a 3 V . d i } g R n OT i S r I S i e s. M 'R I w i / e e e h s e T 0 m ) 0 Y 1 A 1 A 2 L Y I T P I e L T r A L C u F g O T I i C R F 8 0* C AV 0 l- + 2 e m.x 1 = l e [ ll1.l

>: 0 3 l w t [ Page 31 d i { f Criticality Alarm 9 The United States Atomic Energy Cormiission requires that each <j l' licensee who is authorized to possess in excess of 500 gm. of contained d> lF M w//or 330 7. of Plutonium and/or 30') 7,. of U ? 3, provide a is -criticality alarm in the area in which the special nuclear material is handled, usad or stored. (3) The criticality alann installed in the subcritit.al assembly room / is manufactured by Ludlum Measurements Inc. This instrument is called

Ql the model 300 area monitor.

The model 300 area monitor utilize 115V p 60 Hz single phase electrical line as power source. Stand-by batteries d{ are incorporated within the unit to provide power in case of line power I[ q ~1 failure. 1';}{ The garmia field strength is displayed on a three decade logarithmic jdp 4p scale meter covering the range of 0.02 MR/hr to 20 MR/hr. This unit is Tq mounted on the wall approximately six feet away from the surface of the I + assembly tank and approximately 15 feet away from the fuel storage ..,] n+ container. b 3: Q ,a Emernency Shut Down System { During normal operation of the subcritical assembly, with one "p t 1 j ,1 curie Pu-Be neutron source placed in the middle of the most reactive core, a combined dose rate of 3.0 HR/hr from source neutrons, fast neutrons, thennal neutrons, prompt and delayed gantna is expected. Calculations are shown in Appendix A. However, in the event of a radiation level reaching 20 MR/hr the solenoid valve opens automatically to drain water from the core tank back into the water storage container. M

f..

j

).. .( c Page 32 [ The automatic draining is initiated by the high limit alarm signal from the model 300 area monitor, as explained previously, in the water handling system section. Draining of the water will reduce multipli-cation of the core to such an extent that a continued criticality would not be possible. i e .) q,.a !d M 4 70f ajig 7 ,1 .i r ! 4 '). 2 .. 3:)' s '4 4, 1 ~ I I m. l.

r.. t d j-A l Page 33 } IV. DESCRIPT!0ft 0F THE SUBCRITICAL ASSEMBLY CORES Six different core configurations can be formed with the fuel-spacer assemblics presently available. Other core configurations could also be realized if new fuel spacing grids are fabricated. All the six cores are of rectangular parallelpiped geometry with different base dimensions. The height of the core, however, is, fixed at twenty-four inches. Using the single plate spacing assembly three different lattices can be obtained which, in the present work, are referred to as the S-1 core, the S-2 core and S-3 core. With the double fuel spacing ~ t assembly two fuel plates together are used to act as one fuel element. In this manner, the thickness of the fuel and the moderator in each differential slab is increased, while the fuel density remains similar G to the corresponding size "S" type core. In other words, heterogeneity of the core has been increased. Two core sizes can be formed with the double plate grids, which are referred to as the D-1 core and the D-2 core. To assemble a more heterogeneous core, three fuci plates are used as one fuel element in the triple plate spacing assembly. This core will be referred to as the T-1 core. S-1 Core To form the 5-1 core the fuel plates are loaded singly in every slot of the single plate spacing assembly. The resulting fuel span is .175 inches. To obtain a core size of 9 x 9 x 24 in.3 a loading of three plates (in a row) by fifty rows is used. The effective multipli-i cation factor of the S-1 core is calculated to be 0.73. I )

.o I a. Pag.' 34 S-2 Core The S-2 Core utilizes every other slot of the sinole plate spacinq assembl y. The fuel span in this configuration is twice that of the S-1 cera. TV fel n11tos 'en ire m ai for nlates in e rm % thirtv-seven rows. Thus, 148 fuel plates are used in this S-2 core resulting in a core size of 12 in. by 13 in. by 24 in. The calculated value of ~ K-ef f for this core is 1.865. S-3 Core To increase the amount of noderator in the core the fuel plates are placed in every third slot in the single plate spacing assembly. All the 150 fuci plates are utilized in a five plate per row by thirty l rows configuration. The core obtained in this manner is 15 in. by 16 in, a and the calculated K-eff is 0.825. j D-1 Cnre The dual fuel grids are cut in such a manner that two fuci plates are held together as one in each slot. The fuel span of the plate is 1 i 0.65 in. The core formed is 12 in, wide by 12.75 in, thick by 24 in. I l [ long and uses a total of 144 fuel plates in a four plate per row by ) f 18 row thick lattice. The K-eff of this core is calculated to be 0.870. t D-2 Core The D-2 core is the largest core which can be assembled at present in the Idaho State University subtritical assembly. The size of the v D-2 core is 15 in, by 19.5 in. by 24 in. with a fuel span of 1.3 in. All 150 fuel plates are used. This core yields a calculated K-eff of 0.78.

)L:

g) y_

H w -R (:- !'s n page 35 .ja: p T-1 Core j Only one core can be fomed with the triple plate grid. One-it ~ 4 hundred and forty seven fuel plates are used in sets of'three to form jj ,g a core of 12.5 in. 1y 12.5 in. liy '4 in. This tera 1. %crwtrt.nthr 3 core of the subcritical assembly with a calculated K-eff of 0.87. ,f .y furthe-specification'is of the cor9s such as the width of tha .jj), moderator channel are given in Table I together with the dimensions ,Iji 1f in.each core, the fuel span and the calculated and measured K-eff.

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de ta 'l f 3 5 5 0 0 3 u f 5 6 2 7 3 7 C c e 7 8 8 8 7 8 l a K 0 0 0 0 O 0 L d 1 e 6 5 5 r f 0 7 2 u f 8 8 8 s e y a 0 0 t e K 0 N isre v r ) n ot ee f ol s i b anh 8 6 4 e h rnc 9 7 5 9 4 0 2 4 4 er t ea n 1 3 t o ddhi i oC( 0 0 1 1 aC t WM. Sy l l eb e h n ou ) ae tf s ds e I s rr h 8 6 4 A ee c 7 5 3 5 2 f tt nn 1 3 5 6 3 6 ol nnai 2 a eep( 0 1 1 nc C cs oi it ti 5 5 iS c x x ar e ) 0 7 5 3 1 ce z s ib i e 9 1 6 2 9 f u s h 1 1 1 x c e n x x x 5 e r i 0 5 2 5 2 p o ( S C 9 2 1 1 1 1 1 s - e I ftoa E l L r pd B e e 0 8 0 4 0 7 A bl s 5 4 5 4 5 4 T meu 1 1 1 1 1 1 uu Nf E R 1 2 3 1 2 1 O -C S 5 S 0 D T

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l y Page 37 yl V.

MATHEMATICAL MODEL ll Although the subcritical assembly was designed and operated for 1 lt t a ')rief ;+rine of tir' at htcors, The ',tr.te t'niversity, tnre S n noen 'I very little information published about the assembly. The only avail-a

)

i able infmmtion being an article by Jankewski (4) and the special 'H e h! Huclear Materials License Application submitted by Rutgers (5). Both !- lJ [ the articles were directed more towards the economy and safety aspects of the assembly than systematic and accurate analysis of the criticality <!h worth of the different cores. This investigation is directed toward ,b 1 j$,0 f1 a detailed and accurate reactor analysis of the core by means of a com-ic u V putor code and experimentally verifying the results obtained from the ?Q

d. M l}}[e computer calculations. The computer code used in this study is called t4

'ik{n DISNEL. The acronym DISNEL is derived from the name of the code: }l " Diffusion Iterative Solution for Nineteen Energy t_evels." The language jy, 7]~ of the computer program is in FORTRAN-IV and is designed for use on an 1 IBM 360-75 computer system. DISNEL is an operating code, in use at d N present at the National Reactor Testing Station, and is written by Kunze, - d, / et. al. (6) of Aerojet Nucicar Corporation (formerly Idaho Nuclear Corporation). The DISNEL code is a "one-pass," one-dimensional diffusion code. l The nuclear cross sections required for the calculations are provided in the code itself. A fixed 19 group energy structure, as shown in Table 11 is used in the code. The group structure in the resonance region f or computing resonance self-shielding corrections is essentially I

V q +;;: s...

l ty/$ 1 til Page 38 y p M k 1h J('qh TABLE II. DISHEL 19 groups energies. ' i: Group No. Lethargy Width Ener9y in (ev) ji yr Upper Lower y Vfij 1 0.25 1.000 x 107 2.788 x 106 (3:;j 2 0.50 7.788 x 106 4.724 x 106 gM 3 0.50 4.724 x 106 2.!ES x 100 4' O.50 2.865'x 106 1.738 x 106 y 5 0.50 1.738 x'106 1.054 x 106 M WI 6~ 0.50 1.054 x 106 6.393 x 105 )[gj 7_ 0.50 6.393 x 105 3.877 x 105 .q,p,; d~ 8 0.50 3.877 x 10s 2.352 x 105 m;;. i 9 1.25 2.352 x 105 6.738 x 10 10 2.00 6.738 x 10 9.119 x 103-iq 11 2.00 9.119 x 103 1.234 x 103 hs ?3 12 2.00 1.234 x 103 1.67 x 102 7 j 13 2.00 1.670 x 102 2.260 x 101 l 14. 1.75 2.260 x 101 3.928 ,j <;g 15 1.25 3.928 1.125

,y

.m 16 1.00 1.125 4.14 x 10-1 g 17 1.00 4.140 x 10-1 1.523 x 10-3 18 1.03 1.523 x 10-1 5.437 x 10-2 _-s 19 Maxwell Boltzman Thermal Average.

Page 39 identical to that given in ANL 5800 (7). The element cross-sections I l I are from a Nuclear Data Tape compiled at General Electric, NMP0, Cincinnati. DISNEL is suitable for use with four different core geo-metries, viz., infinite slab, cylindrical in radial direction, spherical l and cylindrical with radial and axial calculations cross-matched on buckling, so that, in effect, it is a two dimensional cylindrical geo-metry. The machine requirements for this code are 35K words of usable memory with some of the variables in the diffusion calculation being I double precision system words. The limitations of this code are that I i the mesh points be limited to 120 for any one dimensional calculation and that the number of different elements, compositions.and regions I be limited to 20 each. DISNEL Input As mentioned previously, all the six cores of the subcritical assembly are of parallelpiped type. Unfortunately, this code is not amenable to calculations using this geometry. Therefore, a choice must be made among the four kinds of options of geometry available. If the infinite slab geometry is used for the calculations, the width of the 1 core will be the only variable dimension, and leakage from the finite ) thickness as well as the length, would have to be represented by one l buckling term. As a result, the reflector savings in two directions of the core will be no more than an approximation. A more accurate calculation would be a radial-axial cross-matched calculation. If the cylindrical geometry option is used for the calculations, then the actual m_______

Page 40 shape of the system which is parallelpiped, will have to be represented 3 by an equivalent cylindrical geometry. A cylindrical core is always a . ;}' more reactive than a parallelpiped core of the same height, volune and E composition. Therefore, the cylindri' cal core geometry used for the calculations would result in a larger value for K-eff than the actual rectangular parallelpiped core. A decision was made to use the cylindrical geometry with a radial-axial cross-matched option. Af ter one complete cross-i match calculation was performed, a radial-only calculation was perforned s on the same core with a fixed axial buckling term. A value of 2.0 x 10-3 cm-2 was used as the axial buckling term in the subsequent calcula-tions. This number was suggested by one of the authors of the code (8). The K-eff obtained from this radial-only calculation was only 0.09%. f less than the one obtained from the radial-axial cross-matched calcula-tion. Therefore, the cylindrical radial direction calculation with an I axial buckling of 2.0 x 10-3 cm-2, was finally used in calculation of T_' other cores. In order to use cylindrical approximation with plate-type fuel it was necessary to assume a homogeneous core and then apply a self-shiciding correction factor in the calculation. The self-shielding correction factor used in this calculation is outlined in the DISNEL User's Guide (9). The formula is foi + S 1-C = op 1.27 4NV 1+0.lc { b

7

p.,.,

.g ~ [ Paco 41 1~ y where l! 1 = Potential scattering cross-section per absorber atom. c i f = An empirical correction factor. s=corface of *c 'nol al o "an t V = Volume of the fuel element 1 t i N = Atom density of the absorber } C = The Dereof f correctini fa-tnr i I i p. f ol 4n (1.25 x 10-3 (A)1/3 cm)2 BARN f XTDT-f 10-24.cm f cAgg 2 F f =.700 for Oxygen it i C = 2E3 (I d) kb s .rn j The potential scattering cross-section is the surface area of the absor-3ji t p: i ber nucleus. The empirical correction factor "f" was given as.700(10) .p ' tid ( l for oxygen bearing moderators. The Dancoff correction factors differ for 4g 4~ l each fuel casing geometry. The formula presented above is the Dancoff '? l i. [. correction factor for.a slab type fuel element (11), Where "d" is the F 7 moderator channel width, Is is the macroscopic scattering cross-section a.- y Y ^ t of the moderator, and E3 (x) is the generalized exponential integral .T j function whose values are tabulated in reference (12). I' f The self-shielding factors for U238 and U235 can also be found in reference (13). For each core of the subcritical assembly, the K-eff of five core sizes was computed by the code in an attempt to generate enough data for a simulation of approach to criticality experiment. For example, the

j l 1,, l k' l Page 42 i "l-S-1 has the dimensions of 9 in. x 9 in. x 24 in. The radius R of an e 7 equivalent cylindrical core is given by: r; (9.0 in. x 9.0 in. )S 2.54 cm = 12.83 cm. x Tii R* = (r)D The first core calculated by the code is 8 cm. in radius, then F f$. 10 cm.,12 cm.,14 cm., and 16 cr. Phile the heiqht and composition of n IE the core would be held constant. If the results of Vs-eff versus core p [ size are plotted on a graph, the values for a core of 12.83 cm. in radius P( can then be read from the graph. O; ? The composition of each core was calculated in the following . [t y ). manner. 4 L g Total core area - area occupied by fuel plates 1, g Volume fraction of the water = Total Core Area a f: 5 b Area occonied by aluminium $s. Atom Density of Aluminium lietal = Total core area 0 ~

r density of Aluminium jj:

X Atomic weight of Aluminium M

? 6.02 x 1023 Atom / gm-atom i 10-2s. Barns /cm2 7 U Amount of U235 2 .602 Atom cm / Barns - g3-atrim_ l 2 Atom Density of U 35 = Vol. of core X A of U235 l Anount of U230 2 .602 Atnm cm / Barns -Jm-atom 2 X Atom Density of U 39 = Vol. of core A of U235

k. '

p! a l Page 43 [ [;s As the assembly core increases in size,the reflector thickness of f the core decreases by the sane amount. Therefore, for ease in calcula- 'i 4-] tions each core has been assumed to have 15 cm, of water as radial reflec- [ tor. Since the 'tean free path er a neutron in water is nnly.tM cm., t 15 cm. of water will be an infinite reflector in the calculation. The top and bottom reflector thickness are incorporated within the axial buckling term. Parar:eters used in the calculations of' the core are listed in Appendix B. 1 Results of the Calculation '~ f The results of the calculations provide an insight for the approach to criticality experiment. As mentioned previously, the geometry used for the mathematical r;odel is different from the real core geor.etry. !0 4 Consequently, the results of the calculation will indicate that the 'j! system is more reactive than it actually is because the surface to volume ratio in the model is smaller. This results in lower neutron leakage [ than in the actual system. The effective multiplication constant (K-eff) ( obtained from the calculations is plotted in Figures 22, 23, 24 and also r plotted in rigure 25 which compares these results with those obtained ^ by Jankowski (14). The X-eff obtained from the DISflEL code calculation is slightly less than Jankowski's calculation except for the S-1 core. The deviation in the two calculations can be explained by the combina-t tions of different assumptions made in Jankowski's calculations. These assumptions were: I I i l C M

j. .p it n l-Page 44 1 .:, h 1. A fixed reflector saving of 5 cm. was used in all cases.

ji 2.

The U238 (content) in the core was ignored. 3. .07 in. thick plates with 9.45 gm. of U23s were assumed. 4. R?c.t w.7uhr ra alleipir-j g.,eratry was used. 4 5. Two group calculations were employed. 0 1 h I 1 ..+ of 1g i - h. . i }l' 1 11: j,di,- h!

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h o uC ...s j u < w., mor f er O. c ' e-e p y t v, D 0 '0 3_ 2 ro ) f e r s o e t c 0 a 8 e l ' 1 h p -x t w l n eu -y i f s f s e e r 0 t on o 6 a o ri c 1 l m p et + N 2 b a l ml p D e uu u nc e f .la g s.~. i 4 o v 0 f sc 1 L r fE e fN e m r b eS I o r a KD c un 0 1 2 ( w. i 1 3 D 2 e r o u g i 3 F s_ 0 '1 w t + a + 0 S. 7 6 5 9 1 0 0 0 0 0 s ?s ~" ?O ts. :

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.a ~ Page 49 VI. APPROACH TO CRITICALITY EXPERIMENT Operations of the subcritical assembly at Rutgers has not revealed any criticality hazard. However, for safety reasons, criticality experi-ments will be performed on all six cores of the subcritical assembly. I i Additionally, these experiments are required by the Idaho State Univer-sity Special Nuclear Material License application of the fuel for the suberitical assembly. General Theory (15) In a subcritical assembly, the introduction of an external neutron source will cause the neutron population of the assembly to increase and then reach an equilibrium level as long as the source strength is constant. Suppose that, starting at some arbitrary time zero, No neutrons from an external source are introduced into a subcritical assembly, l 4 o L Each subsequent generation the neutron population in the assembly will t, I be as follows. j' .o i Time in Neutron Generation Neutron Population 'i 1 0 No (source population) 1 No + No K-eff 2 No K.cfr 2 No + No K-eff 4 No (1 + K-ef f + K. err +,,,4)?.cf r } 2 ti .= j.-

I i b Page 50 One can observe that the increase in the neutron population is a power series which can be written as follows: f5 1 - K-eff (m+1) IIo~ 1 - L-ett where m is the number of generation since t = 0. In a subcritical assembly, where K-eff is less than unity and (m) is sufficiently large, the subcritical multiplication M is equal to one divided by one minus K-eff. In the core where the system is just critical, the neutron popula-tion would continue to increase without limit. If the subcritical multiplication can be measured for various fuel loadings, and its reci-procal plotted vs. the amount of fuel used, the curve obtained nay be extrapolated to 1/M = 0 to predict the mass required for criticality.. 1 Neither N or N may be measured direc~tly, but a neutron detector ) o placed at a suitable location in or near the core should provide a count rate which is proportional to the neutron density. A count rate ~l g o without fuel, Co, is first measured and then a count rate, C, is neasured .with fuel in place. The value of C/Co obtained corresponds to M for this particular loading. Additional fuel is then added and a new value of C measured for this fuel loading. The. observations are repeated and a graph of 1/M vs. amount of fuel is plotted. Selecting a suitable locatirn for the detector is complicated by the fact that most neutron detectors are sensitive to thermal neutrons while the neutron energy spectrum from the source is generally different from the fission spectrum produced in the core. Placing the detector

h l Page 51 in the center of the core is usually the best location. If this is not possible, the closest location to the core, within the reflector will provide the best results. Procedure for Appraoch to Criticality (1) Three independent neutron detection systems were installed in the subcritical assembly area for this particular experiment to make sure that the multiplication of the system is knovnat all times. Each of the systems consisted of a boron trifloride filled neutron proportional counter coupled to a suitable pulse-counting system. A detailed descrip-tion of each pulse-counting system is included in the instrumentation section of this procedure. The detector for system number one was located inside the core tank next to the fuel-spacing assembly with the detector center approximately the same level as the core center. Detector number two was located a few inches above the water covering the core. 1 It was attached to a movable steel member which can be removed for core ] access and replaced at a precisely located position prior to obtaining data. Detector number three was located within a water-tight aluminium tube next to (or inside of) the core. The center line of detector number three was approximately at the lower boundary of the fuel. The aluminium tubing which houses detector number three was attached to the fuel spacing assembly in such a manner as to remain fixed during the experiment. Detec-tor number three was taken out of the aluminium tubing during lif ting or lowering of the core and replaced when the core was properly secured inside the core tank.

!? a. E f Page 52 l (- (2) After the three neutron detectors had been placed in their proper locations as indicated in step 1, a neutron source was placed in the graphite block beneath the core tank. The source consisted of 18 gm. of plutonium mixed with beryllium so that the neutron emission q rate is'2.2 x 106 neutrons per second. The source was placed in a ] l hole previously prepared in the graphite blocks designated Y6, X6 and Y6, X7, which was then inserted into the graphite assembly. The n source was approximately eight' inches below the very center of the core ' tank which houses the uranium fuel. The source was inserted into the l ll prepared hole using tongs long enough so that the operator can never i be closer than three feet to this source. Calculations have shown s that the fast neutron dose rate to the operator will be less than three q Mr/hr during the loading. Since it is anticipated that the loading of ( the source will take only a few minutes, exposure to the operator is f a s l-neg11gible. 1 (3) All th ee of the neutron detection systems were energized at

f. ; ~.

i this time and checked to make sure they were counting neutrons from the 1 . neutron source. A 0.02 thick sheet of cadmium was placed between each 1 L-l. of the detectors and the source location to insure that the pulses

r i

obtained from each of the detectors are indeed due to the arrival of --l

t' thennal neutrons and not due to spurious' noise.

No further operations l were performed until it was evident that each of the three channels was ' operating properly. 1, i i II l' j

_g .a 9 [ Page 53 (4)- The water pump was energized and the water was pumped from the storage tank into the core tank until the water level reached the overflow point of the tank. Pumping was. maintained long enough to make sure that the overflow line was cicar and that water was indeed draining back into the storage tanks. The water pump _was then stopped and the fuel spacing assembly for the core was attached to the lifting mechanism and lowered, without fuel, into the water-filled tank. It was observed that the displaced water drained automatically back into -the storage tanks through the overflow line. It is necessary that this line remain open at all times so that the water level between the 1 core and detector number two remains constant throughout the experiment. (S) With this configuration the count rate at each of the three detectors was determined in order to estimate the sensitivity of each ild of the three detectors. From previous experience, it was anticipated }; that detector number two would provide a count rate in excess of two j a counts per second in its location in the critical assembly. It was i anticipated, therefore, that a counting time of ten minutes would be adequate to provide a statistical accuracy in the count rate of approxi-i b mately +2.5%. itowever, it should be pointed out that the statistical L, accuracy of this initial count is not a determining factor in the -l, success of the experiment since extrapolation technique will be used in determining the value of the effective multiplication constant. As j i the effective multiplication constant approaches unity, the overall

Y b Page 54 i multiplication of the system becomes infinite, and the extrapolation of reciprocal multiplication approaches zero. Any error in the initial count rate, therefore, becomes negligible as the count rate from the system becomes large. It was felt that the other two counters, being placed closer to the sour:e than counter number two, provide count rates such that their statistical accuracy would be adequate in the ten minute count. In any event, the counts obtained from systems one f and three would be such that the statistical accuracy should exceed +10 per cent for these systems. (6) The fuel spacing grid was lifted out of the water with a winch assembly and bars placed below it so that there was no possibility of it accidentally falling back into the core tank. At this time one-third of the fuel, or 50 plates, was placed in the fuel spacing assembly l designated for the particular core. Thus, one-third of the fuel was loaded in the initial loading. r! (7) The fuel spacing assembly containing the 50 fuel plates was ([r l slowly lowered into the water-filled tank with constant monitoring of the instrumentation indicating neutron count rate. It was noted in the description of the instrumentation that counter one has a count rate i meter which gives a continuous indication of the neutron count rate. I The lowering of the fuel assembly was such that.the fuel would not be ~ ( lowered more than one foot in three time constants of the number one rate meter. If the count rate meter attached to counter one indicates that the count rate is higher than a factor of 3 above the initial count ) l ~ _ __ _ _ / ~.

l t Page 55 i rate, lowering of the core enould stop until the effective multipli-cation constant is determined bj obtaining an accurate count from l, detector one. Lowering of the fuel assembly continues until it is i folly submerged or until one or both of the count rate meters indicate I more than three times the count from the previous readings. If sub-mersion has to be stopped because of high count rate, the assembled core should be lifted out of the water, 20 fuel plates removed from the core and the remaining lowered into the core following the same procedure. A multiplication of approximately three was predicted for this loading. (8) When the fuel assembly had reached a stable position resting on the bottom support, a count was obtained from each of the three j detectors of such magnitude that the statistical accuracy was better than +3 per cent for each. Using these counts, and those obtained i previously with no fuel, a graph was initiated for each of the detectors, plotting the reciprocal of the system multiplication versus the number of fuel plates in the assembly. Each of the graphs was extrapolated to a value of zero for reciprocal multiplication to indicate the number of plates required to achieve a critical system. A straight line extra-polation was used even though detector placement indicates that a con-cave or convex curve would be anticipated. That curve indicating the minimum number of plates necessary for criticality was used for deter-mining the assembly procedure in the following steps. C

r Page 56 I (9) The core assembly system was lifted from the water using 3 the wi0ch, and additional fuel plates were added to the assembly. The number of plates to be added shall not exceed more than half of the j plates required from the lowest prediction of the previous step, or 25 plates, whichever is smaller. The plates added were placed in such a manner in the grid assembly so as to maintain as uniform a loading across the core as possible. (10) The core assembly was lowered back into the water at a rate of less than one foot per three time constants of the number one rate meter as explained previously. Submersion was terminated if the count rate exceeds 2.5 times that obtained from the previous steady state measurement. Should the count rate exceed this limit, the core motion j was stopped, an accurate count will be determined and the multiplication j factor estimated from an extrapolation of the previously initiated e.urves for the detectors one and three. One half of the plates loaded in step 1 nine was removed and step ten repeated. l (11) When the fuel spacing assembly containing the fuel plates was in place and resting on the core supports, count rates were deter-mined for each of the three detectors. Sufficient counting time was allowed to make sure that each of the count rates was within +3 per cent i statistical accuracy. From these counts', the multiplication of the assembly was determined and a point placed on each of the three graphs of the reciprocal multiplication versus the number of fuel plates. A straight line extrapolation was again used, using the previous point and I L

s- .. / l Page 57 i the point just located to determine the number of fuel plates necessary to achieve criticality. The lower value of the statistical uncertainty limit was used in the new point for determining this extrapolation. (12) Steps nine, ten and eleven, were repeated until all 150 i fuel plates have been added to the assembly and lowered into the core ? tank or until the multiplication factor for the system exceeds 0.93. Since calculations indicate that the effective multiplication factor with all 150 plates in the most reactive (T-1) configuration would not exceed 0.88, it is assumed that all 150 fuel plates will be in position following the assembly. Safety Considerations Y While it is anticipated that during the initial assembly of the 'N l ISU subcritical reactor, no safety problems will arise, each of the { participants in the assembly experiment did observe standard safety precautions. Each person wore a film badge containing a neutron film ',] andagammasensitivefilm. In addition, each participant wore.a dossimeter sensitive to gamma radiation. All dossimeters were read .i inuediately prior to the experiment and subsequent to the experiment to determine if any gamma radiation of significance had been received t by any participant. Film badges will be processed by the commercial film badge supplier for Idaho State University and records kept of the exposure received by any participant. In addition to the personnel a b

F, Page 58 I -t monitoring badges and dossimeters worn by the participants, two port-able gama ray monitors werb. in operation in the assembly room during the experiment. There is also a criticality alarm which is permanently attached to the voll of th? room. These three gamma ray instruments, together with the neutron detection' equipment used in the subtritical

f experiment, insuied the participant that their exposure is below acceptable values.

Should either of the two portable ganna ray instruments or the wall-mounted criticality alarm indicate excessive gamma radiation levels, or should any of the three neutron monitoring instruments indicate ex- 'I cessive neutron levels, the experiment will be terminated and the room evacuated. Upon evacuation, the project director shall be responsible for shutting off the power to the control console. Removal of this power deactivates the solonoid valves and allows all the water from the q 'i core tank to drain back into the water storage system. Draining of this water reduces multiplication from the fuel assembly to such an extent Io l' that continued criticality would not be possible. The project director b will direct one member of the experimental team to leave through the p

1 g

emergency exit of Room 24 so that the automatic alarm in the hallway b of the building will be sounded. The pro. ject director himself will I L [ 1 eave through the corridor next to the health physicist's office and N will turn off all building air circulation equipment at the emergency j u station there when exiting from the area. The project director shall take one ganna ray monitoring instrument with him during this exit and when reaching the hall shall determine whether the radiation level is

r ( Page 59 I. safe for continued occupancy. Should the gamma ray level at this point still exceed acceptable values, the project director will trip the build-ing fire alarm system on his way up the stairs which will insure evacua-tion of the entire building. All other members of the critical assembly 1 team will leave the building as quickly and directly as possible follow-ing the initiation of this emergency procedure. If at any tirce during this experiment eny person feels that an unsafe condition exists for any reason, he shall contact the project director. The project director shall order the experiment to be terminated until investigation has shown that the situation is safe. At no time during I the assembly shall any member present be subjected to excessive radiation levels as indicated by any of the portable monitoring instruments present

.\\,

i in the assembly room. Instrumentation The primary instrumentation fdr the critical assembly experiment for the 150 subcritical assembly will consist of the following three ' systems, the locations of which have been previously described. l l System One: The detector for system one shall be a Reuter Stokes i proportional counter filled with BF. gas to 70 cm. of mercury pressure. 3 The boron in this detector is enriched to 93 percent boron 10. The active dimensions of this detector are approximately thirteen inches long and two inches in diameter. A Mech-Tronics Corporation Model 253 high voltage supply with a maximum voltage of 3000 volts will be used to power the detector. Canberra Industrigs Model 806 preamplifier will be used toget-her with a Harsha11 Model NA-16 linear amplifier to amplify pulses from fL

P .l Page 60 detector one. A Metronic Nuclear Company scaler Model 709 containing an integral timer will be used to determine the counts from the counter number one. A count rate meter will also be used to provide continuous count rate dits from cointe" number one. The entire system number one will be housed in a " mini-bin" meeting AEC specifications. i System Two: System number two shall consist of a Reuter Stokes proportional counter filled with BF.to 40 cm. of mercury with the boron 3 again being in excess of 93 percent boron 10. The active dimensions of the counter for system number two are approximately eight inches long by two inches in diameter. The high voltage supply for this system will be a Power Design Pacific Inc., Model 2K-10, with a maximum voltage of 1 2000 volts. A Metronic Company preamp with a Canberra Industries linear i I amplificr Model 1410 will amplify the pulses from counter number two. A Canberra time model 1492 coupled to a Me.tronic Nuclear itodel 700 scaler determines the count rate. This system will be housed in a " mini-bin" meeting AEC specifications. System Three: System three will consist of a Reuter Stokes detector Model RSN-127A with active dimensions of six inches by one inch in diameter. 'I This detector will be connected to Ludium Model 2200 portable scaler-rate meter with integral high voltage suoply, power supply and timer. The manner in which the neutron monitor systems one, two and three I are set up are shown respectively in Figures 26, 27 and 28. i ~ t I

i Page 61 I n Rate Meter i A Ca,barra 806 g P1rshew NA-16 Detector ) Pre-Amp Linear Amplifier )_________. q V flech-Tronics l t i (Timer Model 253 Scaler l i HV Supply I Mech-Tronics Nucleer Model 709 _j g 1 Figure 26: Components of Number 1 System j T Canberra 1410 Detector j Pre-Amp Linear Amp l v r; Power Design Canberra 1492_ Mech-Tronics Pacific Model 700 s Model 2k-10 Timer Scaler HV supply j Figure 27: Components of Number 2 System j a P Ludlum 2200 HV Supply 3 Scaler / Rate Power Supply with (Detector) Meter Standby Batteries 4 Scaler, Pre-Amp, Amp Rate Meter Figure 28: Components of Number 3 System 7 l

9 t. 1 Page 62 Result of the Criticality Approach Experiment k a T-1 Core Calculations have shown tha.t this T-1 core was the most reactive . core. For this reason this core was used for the first assembly to f insure that no hazards exist in utilizing most of the fuel plates in the subcritical assembly. Two complete approachs to criticality experi-ment were performed with this core. The difference between the two runs was the manner in which the fuel plates were added to the core. During the first run the initial 49 plates were placed in each slot of the triple plate fuel spacing assembly. Subsequent additions of fuel plates were placed contiguous to the plate of the previous loading. In run number 2, three plates were inserted at the same time, in each slot of the triple plate fuel spacing assembly. Thus, the core size was increased with each loading while a coristant fuel to moderator ratio was maintained within the core. Consequently, run number 2 gave a more uniform increase in the values of K-eff. The measured K-eff of.825 was obtained in both runs. l S-1 Core q The detector number 3 in this experiment was placed inside a water I tight aluminium tubing. This tubing was attached to the top single plate grid. In doing so, the configuration of the S-1 core had to be modified. 2 was chosen. A quasi cylindrical core with a hollow center of 2 x 3 in i Because the pre-er.perimer,tal calculation utilized a cylindrical geometry there was speculation that the obtained experimental results would better The loading configuration for the modified S-1 follow the predictions. = l------------- l

I., l i Page 63 i core is shown in Figure 32. The location of the detectors 1 and 2 has I not been moved.. The highest K-eff value for this core is.806. S-2 Core The S-2 core w>s asserrbled essentially the same as was described in Chapter III. However, five plates had to be taken out near the middle \\) of the core and added to one side of the core. The incore space was used for detector 3 tubing. Also, all 150 plates were used in this -l 1 experiment instead of the planned 148 plates. Detectors 1 and 2 have 0 1 not been relocated since the experiment with T-1 core. The highest measured K-eff of.875 was obtained from detector 3. From the experiments it was evident that detector-location is the y most critical factor. This phenomena can be explained by the small d suberitical multiplication of the subcritical assembly core. Also, l calculations have predicted that T-1 core was the most reactive of the a cores. However, experimental results contradict the calculations by j showing a K-eff of 0.875 for the S-2 core, 0.805 for the modified S-1 core, and 0.82 for the T-1 core. This contradiction can be explained by difference in the detector location in the three experiments. In the T-1 experiment the detector was located outside of the core while in the S-2 and S-1 experiments the same detector was in or very close to the center of the core. By placing the detector inside the core more fission neutrons are detected. As a result, a higher subcritical multi-1 i j plication is observed. It is therefore, very difficult to conclude which i of the cores is the most reactive. However, the experiments have demonstrated that criticality cannot be achieved by the T-1, S-2 and modified 5-1 core. I

x . E* I ~ l - ev 8 1 e e 7 2 7 4 cl a 3 fe 4 3 7 9 1 1 5 3 rm ua ss ee hh tt ~. t y al '4 dt e 2 7 1 4 2 8 ea y 0 0 3 cm t 2 ai 2 3 l x i 5 l po a r c sp ap i t wa ir) 1 re C1 rt 6 2 on on 9 8 M 4 3 2 6 6 t e t u 9 5 6 4 cc 0 6 6 0 5 3 R 4 e 9 9 2 h( 0 9 7 1 3 3 t s c 2 1 1 l or et ae 1 p - h t di 1 ro 1 pC eh A1 ti E w. t ,r L mT n e B o ~ 5 ek t 5 0 ~- M A rf 4 4 6 6 6 4 1 2 5 mn n 7 9 T Fo 1 8 0 0 6 7 7 2 0 i a e ~ at 0 6 rt c 8 9 4 7 1 2 3 e 1 1 t n 2 ae ~ pee Dm xr r eoo .i l r c c ae s t p i ee nx hh h meE tt t i 9 2 ~ 0 8 1 nf. s r 4 3 9 1 4 4 6 3 9 I o a 0 0 7 e 7 4 9 5 1 5 0 5 9 p 8 3 1 2 x 1 1 1 e E ~ t o 3 3 N 0 0 M f i i i o/ 0 0 S f l 1 4 I o e u n n n rC CI r C k CF .s o o ro C C f e C ot E a rl c cc ep r ce r e r .M b - o 7 s o 5 s o /S m t 3 t 1 t 5I u c1 90 c2 30 c3 5 N e 50 e 80 e 0 t 11 t 14 t 1 e = e =. e = D o D o D o C C C ii

= = w w o$S 8 7 7 0 0 9 3 4 8 2 1 0 0 4 7 1 5 1 9 2 6 0 8 7 6 9 2 0 8 4 3 9 1 0 1 1 6 1 3 3 6 1 4 4 8 5 2 1 3 2 5 2 4 5 7 1 0 5 6 0 7 8 5 0 3 y 0 3 8 9 2 3 t 1 0 1 1 5 3 i lac i t i) r2 C n 1 ou 4 0 0 1 7 9 tR 8 5 1 2 0 0 4 3 6 8 ( 5 8 6 0 0 3 6 6 3 h 3 8 1 4 2 ce 7 0 1 1 1 4 3 ar oo rC p p1 A -T V m I of 7 ro 3 5 6 3 8 6 E F 6 8 1 5 9 1 5 3 3 5 L t 8 7 6 0 0 0 3 2 4 B an 1 5 8 4 3 A te 6 0 1 1 1 3 3 T am Di r l e ap t x nE e m a i 9 2 2 6 9 5 r 4 5 6 9 7 4 0 8 5 e 5 8 6 0 3 0 8 5 p 1 6 8 2 7 x 5 0 2 1 2 3 E p, CC o. c o e c s e C q r s r e i0 l n0 o n o nM r CI CF h o r i0 o r i o u 0 r 0 r 5 r f C1 E C4 E C1 r s E fe c ot e a 6 s 1 e rl 7 3 e 1 eo r 3 0 r 4 s r 1 M b o 5 0 o 8 o 1 m t 3 0 t 1 0 t 1 5 u c 1 c 0 c 1 N e1 = e2 = 4 e3 = t n t t n e o i e o n e o i D C D C i D C g I{lI

1. I 41ll 4i; \\ l 7 E ~ 0 ~ b 3 0 8 e 2 r o c 1 0 5 T 2 e h t O n '4 i 2 se o. 0 a t 2 l 0 '2 p i t l n_ e e l u u 0 f c b 1 2 e f 3 r o o 4 c r e 0 e b 5h m 1t u n ~. n i s 0 u ~ '6 s s l e r ~ t e a v S __O l p n 1 4l i 0 o r 1 e t o u a t -f c c e i f l t 0 o p e 9 2 i D 1 r t ~ e l ~ b u m m 1 u oN l O o d a 2 i c 1 g i n t n u u R i 1 r R 0 c 8 b O u s e h a r 0 t 6 re 0 v I o b e 1 4 n 3 O r o t 9 c 5 2 e 2 t e e r D u i g ~ F 0 0 0 0 0 0 0 0 0 0 0 9 8 7 6 5 4 3 2 1 1 0 0 0 0 0 0 c 0 0 x?

7 4 1 rr Mh 4 5 2 7 1 0 g rr rr n Mh Mh i 9 re 9 3 4 ur Do 0 0 C d e1 v - rr V rT rr Mh e Mh E se 4 8 L bh 7 3 3 0 0 B Ot A T nf oo i t2 rr a rr Mh i n Mh rF du 5 iE aR 9 3 3 f R 4 0 0 5 y dnu rr rr o Mh Mh r Oc 1 5 k 0 1 ca 0 0 B r l e o ~ e nc nf nt u ea eo ece f kf e k k ev arr ap ati f e t uo t or t et oo sc t e D c ~ t g g t g e r3 nee nea net g e1 ihh ih w id on b n dttk dt di ri ~ ar a n a e a s pb u etfa eth en u l f R aot R at Ri3 t

0 0 6 0 5 3 9 5 2 5 1 2 6 8 5 3 9 1 9 0 6 0 0 7 0 9 7 2 4 6 9 1 0 1 1 1 7 9 2 = 0 3 1 8 3 4 8 3 7 3 6 7 8 3 5 6 1 8 7 6 0 0 8 5 0 5 y 1 8 4 2 6 0 0 1 1 1 6 2 t 7 i lac i te i r ro CC 4 o1 9 9 8 1 7 t - 0 0 8 6 7 1 6 0 4 5 S 9 8 1 6 0 0 8 3 5 9 h 0 6 3 7 5 0 1 1 1 4 cd 6 2 ae L oi I rf V pi pd E Ao L M B m. 8 A of 5 4 5 6 9 9 T ro 0 2 6 8 0 0 6 6 6 5 F-5 4 4 0 0 6 6 6 1 1 .t 0 7 5 9 7 n 0 1 1 1 2 3 an 5 te am Di r l e ap c c t x e e nE s s e M Cy m 3 0 o C 0 0 5 i -s 1 CI o o-CI o r 4 r o 1 r CF r e e p t n r n o n o x a i r i r i r E l r r P C C E C E l e u F f c o 1 7e 3s 2 8c 3 r r 6 7 e 3 e o 70 r 7 s r 3 b t 30 o 7 o 5M m c 0 t 19 t 1 u e =1 c 0 c 5 N t e =4 e =1 e on t t D Ci e on e on D Ci D Ci l'

2 S d o 0 m 6 2 e h t 0 n 4 i i 2 se ta 0 l 2 p )2 l eu f n 0 o i t 0 f t 2 o e a r r l o e u c b c 0 m la i8 e u 1 h n c t L s n u E i s N 1 0 r S i 6 s e I 1 e v r D o t a n tc l o e 0 p i t I 1 l a 4 t e D e c u i f lp 0 f i I 2 c t 1 l r u e m bm l 0 u a 4 0 N c 1 i t I i re 0 b I 8 u s e h k i 6 r t 0 3 e I r v o o tc e e i 0 n 4 O te D 0 3 I 0 2 erug F i 0 0 0 0 0 0 0 0 0 0 0 9 8 5 4 2 7 6 3 1 1 0 0 0 0 0 0 0 0 0 3

3 0 9 5 8 9 0 5 6 3 2 6 8 4 2 7 9 1 2 0 0 5 5 0 7 5 8 2 7 9 2 1 1 5 0 1 1 1 2 1 5 2 2 4 9 4 r 5 1 5 8 1 7 9 8 3' 3 1' 7 5 5 0 0 7 9 3 1 9 2 7 0 0 1 0 1 4 1 1 2 1 1 6 1 7 8 1 8 5 8 1 7 5 2 7 5 2 6 y 1 9 2 5 5 0 0 7 3 6 3 0 7 9 1 t 1 5 0 1 1 2 i 2 l 1 1 ac i t ir C 4 4 3 7 4 9 0 7 8 5 2 7 8 4 3 oe 1 2 7 5 0 5 0 0 1 8 3 0 t r 9 7 7 7 2 0 1 1 o 0 2 hC 1 1 c a2 o-f rS I p V pf Ao E 5 8 0 2 3 0 L mt 6 9 7 0 7 5 6 6 3 4 B on 6 3 3 6 T, 8 0 8 1 9 A re 2 4 9 3 7 4 0 1 1 2 T Fm 9 i ar te ep Dx E la 5 0 8 7 4 1 t 0 3 2 1 4 2 7 0 0 8 n 5 9 8 6 0 3 8 0 0 0 e 3 4 0 4 7 3 0 1 1 3 m 8 i 1 re px E 3 h0 0 0 M c c 5 e e 1 r SC s o r o s ns n CC o n CC o e t r i r a r r l E C E P leu F 1 5 c 2 6 c 3 f 5 e 0 e 2 r t 4 3 'e, 6 s r 2 o r 4 s o 0 7 o 5 M 'c 1 0 t 1 e c 0 0 c 5 b e = 1 = 4 e = 1 m t t t u e o n e o n e o n N D C i D C i D C i

. V t-Page 71 CJ ~OM O h .M H 64 0 U 1 N s .O to c N O .C U C O M ~4N mWu 4 H O D.

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~ - - - i ) pags 72 b 3r j a -[ 3.0 en 2 f son' W l l li.. .'*--3.o m % 1-n. 5 g 15 PLAtts to I 9 M i 9 g l-I l 1 l i. b.jb f I I 20 PLATES / 'l o i 4 //Y' / 4 4 }l. l' ~ 3.0 in e g l' s Debscler.5 I [. g l i r i [? g t g l 1 1 1 { i i !.Il j i 1 e 1 (- ) j ! .g'. .4 3. -1 l b ]. 1 1 i f.) Figure 32: The modefied S-1 core 1 ['i 1 . r- - (- I 6.- l i

.s Page 73 VII. EXPERIMENTAL PROCEDURE FOR THE USE OF THE IDAHO STATE UNIVERSITY SUBCRITICAL ASSEMBLY introduction The Idaho State University subcritical assembly consists of an aluminium core tank 36 in. in diameter and 39.0 in, high. The core tan'< is plac?d on ter of a graphite themal colurrn of 4.0 ft. square and 32 in, high. The subcritical assembly fuel consists of 150 alum- ) inium cladJ uranium-aluninium plates. These fuel plates are 3.0 in. wide,.08 in. thick and 26 in. long. The fuel bearing portion of the i plates is 2.75 in, wide,.04 in. thick and 24.0 in. long containing 10 gm. of U235 These fuel. plates may be assembled in a variety of geometric configurations to form the core of the suberitical assembly. Several different grid arrangements are available for maintaining j proper spacing of the fuel plates in the core tank. l Moderator and reflector is provided by water which is pumped into the core tank from storage containers located nearby. The operat-ing level.of the core tank has been set at 33 in. of water. The con-stant operating level is maintained by the valve and overflow line. Figure 20 shows a schematic view of the core tank, the thermal column, -l the pumping assembly and the water storage container. Figure 21shows a schematic diagram of the shut down mechanism associated with the water handling system.

W Page 74 Safety Rules 1 It has been shown by both calculations and initial assembly experiments that no danger is present in the operation of the sub-l critical essemMy. Mcwenr, sera pre:autions.ve imposed to perons involved with the operation of the suberitical assembly in order to minimize the accidental possibilities and ensure personal protection during an accident. These rules are: 1. The following materials are not to be taken into the subcritical assenbly area without the permission of Professor A. E. Wilson. Permission may be granted for only one material at a time and cannot exceed the limitation stated: Graphite 8 lbs., Beryllium 2 lbs., Beryllium 0xide 4 lbs., Heavy Water 2 lbs., Fissionable Isotepes (U-235, U-233. Pu239) 3 gm of any one or a" combination. This limitation applies to other chemical forms or mixtu7e containing the above material, detailed plans of the experiments to be conducted must be sub-mitted before approval will be granted. Under no con-ditions are larger quantities of these materials to be j carried into the subcritical assembly area. 1 l 1 2. Handling of fuel plates and foils should be performed with water-proof gloves or other appropriate hand covering. 4 3. Neutron source should be handled with source holder or i tongs with a minimum length of 1 m. 4. A minimum of two persons must be.present in the room while the assembly is in operation. One of the two persons must be an authorized operator. 5. Personnel inside Room 23 during the operation must carry on person a personal monitoring instrument either a pocket dossimeter or y-neutron sensitive film badge. The dosage received by persons wearing the dossimeter should be re-corded and kept in file.

Page 75 6. Chemicals or materials which could cause a class B fire should not be stored unattended in Room 23. i 7. No beverage or food should be taken inside Room 23. 8. In the event of an emercency whe-o evacuation nf personrel is needed the procedure below will be observed under the direction of the person responsible for the assembly at that particular time. either the' faculty member present at such time or the student receiving permission to use the facility. All power to the pump and solenoid drain valve will be cut off. (This step should already be performed by the criticalityalarmsignal.) For added precaution the drain valve and the pump power switch will be turned off. The switch is located on a console between the exit door and the assembly tank. The portable gama survey meter will be taken out of Room 23 if it is felt that no danger will result frcm carrying out this action. After evacuation, the door of Room 23 will be closed and the alarm attached to the emergency exit door of Room 24 activated. If radiation level in the immediate outside area is above normal, the building fire alarm will be activated to evacuate the entire building. Also, the building ventilation system will be deactivated. This action can be achieved by pushing the " Penthouse Power Emergency Trip" switch. This switch is located on the wall facing the health physicist's office. The person in charge will be responsible for the evacua-tion of all personnel persent in the basement or the building if the entire building must be evacuated. He will find a safe location for the personnel to stay during the emergency. No one will be allowed to reenter the dangerous area. The health physicist and the Chairman of the Department will be informed immediately. Their names and phone numbers are written on the door of Room 23 and on the 4 health physicist's door. i l 1 .)

Page 76' 9. An operating y monitoring instrument must be placed on the graphite pedestal during any experiment performed with the subcritical assembly. i

10. At the end of each experiment the neutron source must be stored in the paraffin filled storage container inside the storage vault of Room 22.

The fuel plate ny be stored in the locked corc teri with solenoid valve deenergized or stored in the locked fuel container in Room 23. It is the responsibility of the person in charge of the experiment to check that all fuel plates and neutron sources are securely locked in their appropriate places. Procedure for Operation 1. All the fuel plates which will be used in the experiment are loaded in the desired fuel spacing assembly. This loading should be performed with the fuel spacing assembly properly supported by core j support devices. Inside the core tank brackets on the side of the f tank and a hoi.ay-%no structure at the bottom of the tank are used to support the core. Also, the fuel may be loaded with the fuel spac-ing assembly above the tank. If this loading manner is preferred the i fuel spacing assembly should be secured to the core lifting device and i resting positively on iron bars placed across the core tank. Upon completion of loading, the core is lowered into the core tank such i [ that the core rests on the core support devices, f f 2. At this point the neutron souce can be brought into Room 23, l and placed at the desired location for the experiment. The neutron 1 source can be placed in several locations inside the thermal column as well as inside the core itself. The paraffin filled source storage container is mounted on wheels. Therefore, it is recommended that the source container be brought into the roo'n before the source is removed from the storage container.

-..o -, c Page 77 3. At this point the gamma survey meter is turned on'in the appropriate range and placed on the thermal column. 4. The power to the solenoid valve and pump is restored. This is acecTrlished by pushing the" valve and pump power recet" switch. Then the manual pump switch is turned on. Approximately 25 minutes is required to fill the assembly tank with 33 in, of water. 5. If, during the course of the experiment, the gcmma survey meter or the criticality alarm meter indicated a gamma field of 20.0 Mr/hr, the assembly should be shut down and evacuation procedure is observed. l ) i 1 j i v i [ l s ) 4 4 _________________.s

m O O S g i i 'j APPENDIX l I I I r P

{.,. j Page 78 APPENDIX A 1 00SE CALCULATIONS na ted Pw-r K-eff =.874 i source neutrons = 2.1 x 106 n/sec 1 i ti (subcritical multiplications) = 1 7.9a = 1-K-eH Neutron Population rate = Source neutrons x M 2.1 x 106 n x 7.94 = 1.67 x 107 i = n/see sec Fission Neutron population rate = 1.67 x 107 rn 2.1 x 106 n 1.46 x 107 n sec sec sec Fission Rate = 1.46 x 107 n/sec 5.90 x 106 Fission 2.47 n/ fission see Rated Power = L9x106 fission /sec .19 x 10 3 watts. 3.1 x 1010 fission watt-sec Fast Neutron Dose Rate Calculation In this calculation, bulk shielding facility data will be used. This data is published in ANL 5800 as Figures 7 - 11 on page 464.

Page 79 l At 30 cm. (or 12 in.) from the core surface a ' number of 15 ergs /gm-hr-watt is reported for fast neutrons. Conversion of this number in Rad yield br-watt 1.5 x 10-1 Rad or 1.5 Rea with fast neutr'ons RBE (Relative t hr-watt br-wa tt Ciological Effectiveness).of 10. { Therefore i DR(f) 1.5 Rem x 1.9 x 10-4 watts (ratedpower)=.285f1 REM hr-watt hr Thermal fleutron Dose Rate Calculation Again, BSF data will be used. At 30 cm. from the core surface ) a number of 2.5 x 105 n (thermal) is obtained from BSF data. j 2 cm -sec-watt -i 2.5 x los n x 1.9 x 10-4' watts = 47.5 n 2 cm -sec-watt 2 em -sec I A conversion 8.6 x 10 4 Mr/hr n(.625ev) is given in ANL 5800 page 466. 2 cm.see DR(th)=47.5 n x 8.6 x 104 fir /hr 2 =.0403 MREM /hr. cm -sec n(.025 ev) 2 cm.5cc l I l L_ L_________

l Page 80 y Photon Dose Rate Calculation l 1 BSF data gives 1.3 x 103 ERGS at 30 cm from the core surface h gm-hr-watts 4 1.3'x 103 _ ERGS x 1.9'x 10-4 uatts = 2.56 x 10-1 _ ERGS -gm-br-watt gm-hr If a Conversion of IRAD is used. 100 ERE 9'n I DR(g)= 2.56 x 10-l' ERGS 2.56 *E AD = 2.56 MREM = A gm-hr Fr-- hr ll - y Source Neutron Dose Rate , J. ' \\ Assuming that 2.1 x 106 n/sec PuBe neutron source is located l { in the center of the core tank. The flux at the surface of the core tank will be: i, 1 2.1 x 108 n x geometric attenuation x water shielding sec .i; <, = 0.3. 63 n cm2-sec q.: .h, i i where geometric attenuation = 1 = 3.76 x 10-5 cm-2 g g2 (core tank) y and water shielding = e raR (core / tank) = 4.75 X 10' i r = Macroscopic neutron cross section for 4.5 Mev neutrons = n a p .117 co-l. ANL 5800 page 466 gives a conversion factor of.11 mrem /hr n/cm -sec e for 5 Mey neutrons. Therefore, S! n; . - ]3 (, j

'Page 81 DR(Source)=.363 n x.11 Mrem /hr z cm sec n/cm -sec -[ e l: h =.04 MREM i Fr i 9. Tctal Dose Rate !!' i DRt =.285 MREM +.0408: MREM + 2.56 MREM + 04 IGtEM ! ' 'l hr hr. hr hr i .t.

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Page 89 LIST OF REFEREt:CE I-1. Borst, L. B., 1962. " Historical Aspects of Suberitical Assemblies." New York Universitv.. United States Atomic Energy Conrnission Renr t in 70FL pp.1 L 2. Fermi, E. 1939. (cited by) Eerst, L. B., 1952. " Historical Aspects of Subtritical Asseiblies." New York University. United States Atomic Ene gy Commission Report TID 7619 pp. 2 l I 3. Code of Federal Regulations Title 10, Part 70.24 (1). 4 Jant:owski, F. J.,1962. "A Heterogeneous Enriched Subcritical l- ' Assembly." Retgers, The State University of New Jersey. l U.S.A.E.C.' Report TID-7619, P(P73-9). 5. Rutgers, The State University of New Jersey, "Rtugers, The State University Special Nuclear Material License. U.S.A.E.C. ~ Report Docket'70-461. l 6. Kunze, J. F.,1971. " User's Guide for Reactor Physics Computer Code DISHEL (D',ffusion ' Iterative Solution for Nineteen Energy i Levels)." Reactor Development Branch Idaho fluclear Corpora-l tion (presently Aerojet Nuclear Corporation.) 7. " Reactor Physics Constants",1963. Second Edition. U.S.A.E.C. l Report ANL-5800, pp. 526-7

8. ~Kunze, J.

F., 1973. Personal Telephone conversation. l 9. Kunze, J. F.. 1971, " User's Guide for Reactor Physics Computer Code DISNEL (Diffusion Iterative Solution for Nineteen Eneroy Levels)".pp. 17-18. Reactor Development Branch Idaho Nuclear Corporation (presently Aerojet Nuclear Corporation.) ] 10. Ibid, pp.17 11, '" Reactor Physics Constants" 1963. Second Ed. U.S.A.E.C. Report ANL 5800 pp. 281. 12., Case, K. M.,~ 0cHoffman, F., Placzeck, G. " Introduction to the Theory I . of Neutron Diffusion." (Washington, D. C., Government Printing . Office.1953.) pp. 155-9. I e_

,;, o..y.o ~ ~ - ~ 1 i. i Page 90 13 " Reactor. Physics Constants," 1963. ANL-5800 pp.526-9. Second Ed., U.S.A.E.C. Report f- '14 Jankowski, F.-J., 1962. 1 "A Heterogen2ous Enriched Suberitical Assembly." Rutgers, The State University of New Jersey, i i, .U.S.A.E.C.-Report; TID-7619.. pp. 78.

15. :Glowe., D.0.

Jc65. "Er erire t t i.' h*rtn M W w u McGratt4 Hill, New Yo'rE pp. 2T2. ', ' " " * ' " # 8 '.. 1. i I b

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'3 '.g s :..:.- ( F- ' EMERGENCY PLAN FOR THE NUCLEAR FACILITY AT. 1 LILLIBRIDGE ENGINEERING LAB AT-IDAHO-STATE UNIVERSITY 3-s April-26, 1994 (Revision 5) 'l l I f;' I ,!i 1 i ..^. s--------_--------------------_-- J m

L.: g. l'. CONTENTS SECTION 1. INTRODUCTION..................................... 2 SECTION 2. DEFINITIONS...................................... 4 SECTION 3. ORGANIZATION AND RESPONSIBILITIES................ 5 SECTION 4. EMERGENCY PROCEDURES............................ 9 N u c l e a r Eme rg e n cy......................................... 12 B o m b Th r e a t.............................................. 14 F i r e o r Exp l o s i o n.......................................... 16 Thef t or Attempted. Thef t of Special Nuclear Material...... 17 i C i v i l D i s o rd e r........................................... 17 SECTION 5. EMERGENCY FACILITIES 'AND EQUIPMENT.............. 19 SECTION.6. MAINTAINING EMERGENCY PREPAREDNESS.............. 2 0 APPENDIX 1. . FLOO R P LAN.................................................... 2 2 Figure 1 t LEL First Leve l Floor Plan................................ 2 3 i Figure 2 LEL Second Level Floor Plan............................... 2 4 Figure 3 Reactor Labora tory Floor Plan............................. 2 5 Figure 4 Sub-critical Assembly Laboratory Floor Plan............... 26 APPENDIX:2. NOTI FI CATION ROSTER........................................... 2 7 i I APPENDIX 3. l-EMERG EN CY EVACUATION PLAN...................................... 2 8 i EP, Rev. 5 p. 04/26/1994, Page 1 [ [_ __________-.______________------U

i j SECTION 1. INTRODUCTION This emergency plan shall be used as a plan of action to follow in the event of a nuclear incident at the nuclear facility located at. Idaho State University, Pocatello, Idaho. It shall be used in conjunction with the State of Idaho Radiation Emergency Response Plan for all events involving a radiation accident, i.e., any release of radioactivity which may injure or contaminate a person. The Idaho plan delineates the actions to be taken by the Idaho State Police, Department of Health and Welfare, and the agreements with the local hospitals and the fire department in the event of a radiation accident. Those events which do not involve a radiation accident shall be the j responsibility of the local operating staf f and the administration of Idaho State University. The nuclear facility consist of an AGN-201 nuclear reactor ~ manufactured by Aerojet General Nucleonics in 1956. It is owned by Idaho State University and is operated under License Number R-110. The maximum power it is licensed to operate at is 5 watts. The fuel consists of uranium enriched to 19.88% uranium 235. The AGN-201 reactor system consists of two basic units, the reactor and the control console. The reactor unit includes the core consisting of uranium dioxide dispersed in polyethylene, a graphite reflector, and the lead and water shielding. Fuel loaded control and safety rods are installed vertically from the bottom of the reactor unit, passing by the instruments which measure the power level. The rods are inserted by control mechanisms which r, EP, Rev. 5 04/26/1994, Page 2 \\..... - _ _

provide safe and ef ficient operation of the reactor. The weight of the reactor unit, with the water shield, is 20,000 pounds; the weight of the console unit is 800 pounds. The AGN-201 is located in the basement of the Lillibridge Engineering Labcratory at Idaho State University. Refer to Appendix 1 for the floor plans of the laboratory. 1 EP, Rev. 5 04/26/1994, Page 3

I.,. r l i SECTION 2. DEFINITIONS Emeraency plannina zone - Rooms 14, 19, 20, 22, 23, and 24 on the first level of the Lillibridge Engineering Laboratory. Nuclear incident Any unusual circumstances or occurrence that could lead to or cause damage to the reactor and/or sub-critical facility nuclear fuel or nuclear fuel cladding. Operations area - The area inside room #20 and #23. Radiation accident - Any release of radioactivity which may injure or contaminate a person. Operations boundary ~ The walls, ceilings, and doors of rooms #20 and #23. Nuclear facility - Consists of a AGN-201 nuclear reactor and the sub-critical assembly area. Operations team - Consists of NRC licensed operators and health physicists. Nuclear emeraency - Any emergency which combines a radiation accident with any nuclear incident. EP, Rev. 5 04/26/1994, Page 4

SECTION 3. ORGANIZATION AND RESPONSIBILITIES l Table 1 is a diagram of the local operating organization and l l shows the relationship to the State of Idaho radiation emergency 1 response organization. The Idaho plan contains the responsibilities and duties of the auxiliary organizations who are committed to respond to a radiation accident, i.e., Idaho State Police, Department of Health and Welfare, the local fire department and hospitals. The Idaho State University personnel who will respond to a nuclear incident are the Reactor Administrator, Reactor Supervisor, Radiation Safety Officer, and Campus Security. They will be advised by the School of Engineering faculty. There are no other local support organizations directly committed to respond to a nuclear incident other than the normal response provided by the local fire and police departments. In order for the emergency plan to function as intended, it is essential that all coordinating personnel at Idaho State University be aware of their areas of responsibility and assure that their facilities and equipment are available and operational. The following is a list of University personnel and their areas of responsibility: 1. The Reactor Administrator and/or Reactor Supervisor are responsible for: a. Operations at Idaho State University should a nuclear incident occur. b. Notification of the State Police and Idaho Department of EP, Rev. 5 [ 04/26/1994, Page 5

3 Health and. Welfare in the event of a radiation accident. c. Requests for medical assistance or notification of a r, applicable hospital to prepare for patient care. d. Safety -regulations and practice within the nuclear facility. e. Internal' operations.and assignments. f. Routine checking of safety equipment and safety within the facility. and assuring' that employees are knowledgeable in equipment operation. g. . Requests for additional fire fighting assistance, and instructing the fire marshall concerning the hazards of the nuclear facility. h. Evacuation plans and assembly areas. i. Maintaining up-to-date notification roster of appropriate personnel and agencies. 'j. Personnel accountability procedures at the.Lillibridge Engineering Laboratory. k. In their absence the Radiation Safety Officer shall assume their' duties. 2. The Radiation Safety Officer is responsible for: Health physics assistance at Idaho State University. a. l. b. Authorizing volunteer emergency workers to incur radiation exposure in excess of normal occupational limits. EP, Rev. 5 04/26/1994, Page 6 J

c. Manning check points or control points for surveying personnel and equipment. d. Health physics at Lillibridge Engineering Laboratory and scheduling of personnel and working times in radiation areas. e. Monitoring teams and environmental sampling at Idaho State University, analysis of samples, and maintenance of records. f. Decontamination procedures and control. g. Health physics for contaminated personnel until they are attended to by the proper medical personnel. h. Insuring that all necessary health physics information is communicated to the appropriate agencies. 1. Personnel monitoring, personnel radiation records, and for scheduling of personnel for the operations team. j. In his/her absence the Reactor Supervisor or the Reactor Administrator will assume these duties. 3. Idaho State University Campus Security will be responsible for: Establishing area control and manning of check points, a. f b. Traffic control and traffic counting. i Assistance in communications and information dispersal. c. d. Assisting State Police in the event of a radiation accident. / EP, Rev. 5 04/26/1994, Page 7

c... = Refer to'. Appendix ' 2 for a list of organizations who have. indicated they can provide assistance upon request either via the State of Idaho Radiation Emergency Response Plan or as normal duty of their organization. l-l i l l EP, Rev. 5 04/26/1994, Page 8 i

i ., o i i SECTION 4. EMERGENCY PROCEDURES Table 2-shows the emergency classification system for . potential emergency situations which may occur in order of-increasing severity. i Emeraency Class Tvoe of Purnose Incident 1. Unusual incident Civil Disturbance To secure the area. l Bomb Threat Investigate the { Theft of SNM situation. Alert the police. 2. Personal injury Fire or Explosion Same as 1 plus alert firemen and minimize damage. 3. Personal injury Nuclear Emergency Same as 2-plus with contamination activate the. State Radiation Emergency Plan. Table 2. Emeraency Classification Table, i The emergency planning zone, EPZ, Rooms 14, 19, 20, 22, 23, 24 on the first level of the'Lillibridge Engineering Laboratory,'see . figures 1-through 4. This emergency plan shall apply to the EPZ. There are no postulated accidents for the AGN-201 Reactor which would result in exposure of 1 rem whole body or 5 rem thyroid beyond the operations boundary. -In the event of an incident which requires evacuation of the

building, i.e.,

fire or explosion, all personnel within the EPZ shall proceed to the shop area near the double doors to be accounted for and if radioactive contamination is suspected the EP, Rev. 5 04/26/1994, Page 9

potentially contaminated personnel will be separated from all others. In the event of a radiation accident the State of Idaho Emergency Radiation Response Plan will be initiated. The emergency exposure guidelines are the same. as the radiation dose standards for individuals in restricted areas as 'specified in 10 CFR 20.101. These guidelines are sufficient when the size and postulated radiation accidents are considered for the nuclear facility at Idaho State University. The facility will maintain emergency procedures for dealing with various emergencies including nuclear emergencies, bomb threats, fires or explosions, theft or attempted theft of special nuclear material, and civil disorder. The response procedures describe the type of response to be accomplished, the duties and responsibilities of the security organization and the management involved in the response. An up-to-date notification roster will be maintained in the Reactor Supervisor's office, in the School of Engineering administrative office, and the Campus Security office. The notification roster ' indicates the names and telephone numbers of those who will be notified immediately of any emergency and also the names and telephones numbers of those who may be called upon to assist. -Refer to Appendix 2 for the notification roster. A number of radiation monitoring devices are maintained at the l- . Radiation Safety Office. The Radiation Safety Officer will determine which devices are to be used. A Geiger Counter will be the standard device for monitoring dose rates and contamination EP, Rev. 5 04/26/1994, Page 10

levels around the facility. Radiation monitoring devices are also maintained in the chemistry and Physics buildings in the event that the Radiation Safety Office is not accessible. These devices include pencil dosimeters, hand held Geiger Muller counters, s c i n t i l l a t i o n c o u n t e r s', and thermoluminescent dosimeters. All personnel entering a radiation area or a suspected radiation area shall have some method of determining the radiation field and personal dose. No person shall enter a suspected radiation area unless under the direction of the Radiation Safety Officer. The . exception shall be for the police and the fire departments. If it is necessary that the police or firemen enter a radiation area without personal radiation monitoring devices the Radiation Safety Officer shall be informed immediately and will then survey the affected person for contamination and arrange for a whole body assay if necesnary. EP, Rev. 5 04/26/1994, Page 11

c-- l+,, i Nuclear Emeraency A nuclear emergency shall be any emergency which combines a radiation accident with any other nuclear incident. Any emergency which includes a radiation accident is a sufficient condition to initiate the State of Idaho Radiation Emergency Response Plan by calling the Idaho State Police, Region V. Evacuation of the Lillibridge Engineering Laboratory may or may not be required for l a nuclear emergency. If the emergency is strictly a radiation accident and not combined with fire or explosion, building evacuation will be ordered if the radiation levels are above 10 mR/hr outside the operations boundary or if there are airborne radioactive materials. In the improbable event that the nuclear reactor laboratory must be evacuated two exits are accessible from the reactor laboratory itself. These are: (1) through the double doors to the reactor laboratory and (2) an emergency escape hatch located in the roof. Two sets of stairs and an elevator lead to the ground level floor from which three exit points are available; one to the Southeast and two to either side of the display foyer. A ladder leads to an escape hatch in the ceiling of the reactor laboratory opened only from the inside. The emergency exit sequence shall be: (1) Personnel in adjacent laboratory spaces shall be warned by operators to initiate evacuation; (2) The first person to reach the Emergency Ventilation Cut Out Switch _(Located on the south wall, across from the health f physicist's office shown in figure 1 will trip all ventilation off f EP, Rev. 5 04/26/1994, Page 12

- _ _ _ _ - - - - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ = the line in the building preventing any further air exchange; (3) If time permits, radiological monitoring equipment will be taken from the reactor laboratory. If this is not possible other monitoring equipment has been stored at the Physical Science Building for emergencies; (4) The building fire alarm will be sounded to evacuate the entire building. The locations of the local fire alarms are at the bottom of the staircase on the south side (or on the way to the staircase on the north side) of the building; (5) The last person leaving the reactor laboratory area will shut all doors. All persons leaving the EPZ area shall proceed immediately to the northeast corner of the building near the large overhead door of the machine shop to be accounted for and be checked for radioactive contamination; (6) Staff personnel shall proceed to the Fire Department Emergency Command Center (if established) and provide information and assistance to the on scene commander; (7) The University Administration will be notified as soon as is practicable. The Reactor Administrator and the Reactor Supervisor will be notified immediately. They will, in turn, determine which State and Federal agencies shall be notified. The Radiation Safety Of ficer will be responsible for thorough radiation monitoring. The nuclear reactor laboratory will be reentered as radiological levels permit and then only by authorization of the Reactor Supervisor or his designated representative. The reactor system will be checked for damage. Refer to Appendix 3, Emergency l Evacuation Plan. EP, Rev. 5 04/26/1994, Page 13 L

+. ' t- .g. l Spub Threat -1. The : person receiving the. threat should obtain as much information as possible. Ask the following questions: ca. Where is the bomb? b. What kind of bomb is it?- c. What time will it go off? d. Why are you doing this? 2. ' Notify: a. Idaho State University Campus Security b. Pocatello Police Department 3. The security of ficer, upon being notified'of the threat, will . proceed immediately-and notify the following offices: a. Chief of campus Security b. Pocatello Police Department c. University Administration d. Reactor, Administrator e. Reactor Supervisor '4.- Shut down the reactor. S. The security of ficer will record the name and location of the I person receiving the threat. EP, Rev. 5 04/26/1994, Page la C1__. ::_ __ __ ___ _ _ _ ___

6.

The removal or transfer'of any radioactive material will be the responsibility of the Reactor Administrator and/or-the -Reactor Supervisor. l

7..

School of Engineering staff will assist'with any subsequent . searches of the Lillibridge Engineering Laboratory. i is l-i- l l l-l l-l- EP, Rev. 5 04/26/1994, Page 15

Fire or-Explosion 1. Scram the reactor by pushing the power off button and check that:the: rods have scrammed by any of the following methods: a. . Rods engaged lights out, b. Decreasing current trace on channel 2 and 3 strip charts. 1 c. Period meter pegged low. 2. As soon as the scram is verified evacuate the building. 3. Notify the Pocatello Police Department by the quickest -available means, i.e., radio, fire alarm, telephone. 4. The Fire Department-will: a. Proceed to the area. b. Notify.'Pocatello Police Department.and Idaho State University Campus Security for-traffic control. 1 .i 5. ISU-Campus Security will notify, the campus maintenance department. The maintenance. department will provide a person to secure or activate-building systems and alarms as .necessary. l 6. Notify'the Reactor Administrator and/or Reactor Supervisor who L 1 L will-in turn notify: j a. Idaho State University Radiation Safety Officer. EP, Rev. 5 04/26/1994, Page 16 (

b. Idaho State Radiation Control Section. c.. U.S. Nuclear Regulatory Commission Region IV. Theft or Attempted Theft of Special Nuclear Material 1. If an indication of a thef t or an attempted thef t exists in or around Rooms 20 or 23 of the Lillibridge Engineering Laboratory, immediately notify the Campus Security who will in turn notify: Chief of campus Security a. b. Pocatello Police Department c. Reactor Administrator d. Reactor Supervisor 2. The Reactor Administrator and/or the Reactor Supervisor will proceed immediately to the Lab and inspect and inventory all special nuclear material. If a theft or attempted *. heft of l special nuclear material has occurred, the following.will be 1 l notified immediately: l a. .U.S. Nuclear Regulatory Commission Region IV. b.. Idaho State Radiation Control Section. l l Civil Disorder 1. Notify the Chief of Campus Security. 2. Campus Security will post guards in the basement of the lab. t I i EP, Rev. 5 04/26/1994, Page 17

,a 3. Notify the Pocatello Police Department for riot and incident control. l i 1 I i l l EP, Rev. 5 04/26/1994, Page 18 f' I

SECTION 5. EMERGENCY FACILITIES AND EQUIPMENT The emergency support center will be in the northeast corner of the machine shop near the large overhead door, Figure 2. Emergency control directions will be given from this area. An ambulance shall be called for any person who may be l injured. If that person is also contaminated with radioactive l materials, the State of Idaho Emergency Radiation Response Plan shall be initiated by calling the Idaho State Police, Region V, and the receiving hospital shall be informed the injured person is potentially contaminated. If a person is not injured but contaminated with radioactive q materials, the State Plan will be initiated and decontamination procedures will begin under the direction of the Idaho State University Radiation Safety Officer in accordance with the State Plan. The only emergency communications system in addition to the normal telephones are the hand-held radios which are used by campus security. Emergency communications will have to be by word of mouth if the telephone system is inoperable or the radios used by Campus Security are unavailable. EP, Rev. 5 04/26/1994, Page 19

SECTION 6. MAINTAINING EMERGENCY PREPAREDNESS The. Reactor Administrator and the Reactor Supervisor are responsible to. ensure the proper execution of the Emergency Preparedness Plan. The training of University personnel who are responsible'to act under this emergency plan is the responsibility of the Reactor Administrator and the Reactor Supervisor with the assistance of the 4 W Technical Safety Office in the area of radiological control. For those personnel and agencies not a part of Idaho State University, training is a responsibility of the Department of Health and Welfare, State of Idaho. u The Idaho State University Reactor Administrator and the ~i f Reactor Supervisor will provide a training program at least once a year to train other University personnel who may be called upon to assist in the improbable event of a nuclear incident. University personnel who would be involved in a nuclear incident will be tested by annual drills. This will be accomplished by the unannounced initiation of a drill by the l M ' Reactor Administrator or by the Reactor Supervisor with written s j d ' permission from the Reactor Administrator. Outside agencies will i be contacted in advance and informed of the drill. University L personnel will carry through with this action as though it were an actual' emergency. Records of these drills will be entered into the facility operating records by the Reactor Supervisor or a licensed p. Senior Reactor Operator or Reactor Operator. tn 5. The Emergency Preparedness Plan shall be audited under the M cognizance of the Reactor Safety Committee at least once every two years. They shall evaluate the effectiveness of the plan and note EP, Rev. 5 04/26/1994, Page 20

the results of the evaluation in their minutes. They shall also approve any changes which may be made to the plan. Emergency equipment used for fire

fighting, radiation detection and air sampling shall normally be checked for proper operation annually, but in no case shall the check be greater than 16 months.

Batteries in portable equipment shall be checked prior to each use and annually unless previous experience dictates a more frequent check is required. A complete stock of replacement batteries shall be available for all battery powered emergency equipment. Emergency equipment will be inventoried annually. EP, Rev. 5 04/26/1994, Page 21

APPENDIX 1. FLOOR PLAN LILLIBRIDGE ENGINEERING LABORATORY I u l l i l EP, Rev. 5 l 04/26/1994, Page 22 ._. _ _ _ _ _ _. _ - - _ ~

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c. i l \\... 1.;. ..!... >. : w....:..._,,..._. .? .;..a..;s.......-,...,. ...;e :., i \\ i. I c i .e I a a \\ SUBCRITICAL ASSEMBLY LABORATORY a i i I 4 I Subcra.tical l}, Assembly Security l cabinet I t 1 i ); ? L C_m y e i.i ,J. '.*r b 4 \\ Floor Plan of Suberitical Assembly Lab Figure 4 ~ l EP, Rev. 5 04/26/1994, Page 26 l t

,.,.i APPENDIX 2 NOTIFICATION ROSTER REACTOR ADMINISTRATOR J. BENNION HOME: 208-233-3239. WORK: 208-236-3351 REACTOR SUPERVISOR T. GANSAUGE HOME: 208-234-0862 WORK: 208-236-3637-RADIATION SAFETY OFFICER T. GESELL .HOME: 208-237-1076 WORK: 208-236-3669 ISU CAMPUS. SECURITY .208-236-2515 POCATELLO. POLICE DEPAltTMENT 911. -POCATELLO FIRE DEPARTMENT 911 SUPPORT NOTIFICATION ROSTER -IDAHO STATE POLICE 208-236-606'6 -BANNOCK REGIONAL MEDICAL CENTER 208-239-1800 POCATELLO REGIONAL-MEDICAL CENTER 208-234-0777. 1 ISU ADMINISTRATION 208-236-3440, NUCLEAR REGULATORY COMMISSION 301-816-5100 ~ ) L UPDATED S/6[7d 4 e I l[- M EP,-Rev. 5 .. j-); 04/26/1994, Page 27 .___u_______-

APPENDIX 3. EMERGENCY EVACUATION PLAN TO BE FOLLOWED IN THE EVENT OF A NUCLEAR EMERGENCY WHICH HAS POTENTIAL OF CAUSING INJURY 1. The licensed reactor operator is cognizant of the detailed emergency plan. HE/SHE WILL BE IN CHARGE OF EVACUATION. 2. Use the normal room exit and building exits if possible. The escape hatch located in the roof is to be used only if normal exits are blocked by fire or radiation. Make sure exits to lab are closed after all persons are out. 3. The radiological monitoring instrument on the reactor console and the reactor log book will be brought from the laboratory room by the reactor operator. i 4 4. If the radiation levels are above 10 mR/hr outside the operations area of the Nuclear Reactor Laboratory, the reactor operator will order building evacuation. 5. The first person to reach the Emergency Ventilation Cutout Switch (located on the south wall, across from the health physicist's office) will trip all ventilation off the line. l I EP, Rev. 5 04/26/1994, Page 28

o, 6. Tha reactor opsrator shall initiate building evacuation by . tripping one of the building fire alarms located at the bottom of the staircase on the south side (or on the way to the staircase on the. north side) of the building. 7. The reactor. operator - shall notify the Reactor Supervisor and/or the Reactor Administrator immediately. 8. The Reactor Supervisor and/or the Reactor Administrator shall be.-in charge of all building reentry. i EP, Rev. 5 i. 04/26/1994, Page 29

. ? PROPRIETARY INFORMATION NOT FOR PUBLIC DISCLOSURE CONTAINS'10 CRF 2.790 (D) INFORMATION WITHHELD FROM PUBLIC DISCLOSURE EMERGENCY PROCEDURES (continued) e. Reactor Administrator Work: 208-524-0905 Home: 208-526-4907 f. Reactor Supervisor Work: 208-236-3637 Home: 208-233-1173 4.- The security officer will record the name and location or the person receiving the threat. 5. The removal or transfer of any radioactive material will be the responsibility of the Reactor Administrator and/or the Reactor Supervisor. 6. Engineering Department Staff will assist with any subsequent searches of the Lillibridge Engineering Laboratory. i B. Fire or Explosion at the Lillibridge Engineering Laboratory l. Notify the Pocatello Fire Department by the quickest available means, i.e., radio, fire alarm, telephone 911. .2. The Fire Department will: a. Proceed to the area, but be sure that the fire fighters are wearing protective clothing and breathing devices. b. Notify Pocatello Police Department, Bannock County Sheriff, and/or Idaho State University Security to post guards around the area, and to keep out any j unauthorized vehicles and persons. REVISION 3 FEBRUARY 23, 1990 I L L r L

PROPRIETARY INFORMATION NOT FOR PUBLIC DISCLOSURE CONTAINS 10 CRF 2.790 (D) l INFORMATION WITHHELD FROM l PUBLIC DISCLOSURE ENCLOSURE 1 OPERATING ORGANIZATION POSITION ,NAME TELEPHONE NUMBER Reactor Administrator A. Stephens Work: 208-524-0905 Home: 208-526-4907 Reactor Supervisor R. Clovis Work: 208-236-3637 Home: 208-233-1173 Radiation Safety Officer T. Gesell Work: 208-236-3669 Home: 208-237-1076 REVISION 3 FEBRUARY 23, 1990 1 l l l l Enclosure A4 Figure Al. Locations ofISU Accelerator Centet sites rulative to the Nuclear Facility. l d

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