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TECHNICAL REPORTS SERIES No. 34

1 F teactor

Shielding

INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1964

R E A C T O R SHIELDING

The following States a te Members of the International Atomic Energy Agency:

AFGHANISTAN

ALBANIA

ALGERIA

ARGENTINA

AUSTRALIA

AUSTRIA

BELGIUM

BOLIVIA

BRAZIL

BULGARIA

BURMA

BYELORUSSIAN SOVIET SOCIALIST

REPUBLIC

CAMBODIA

CAMEROUN

CANADA

CEYLON

CHILE

CHINA

COLOMBIA

CONGO (LÊOPOLDVILLE)

CUBA

CZECHOSLOVAK SOCIALIST REPUBLIC

DENMARK

DOMINICAN REPUBLIC

ECUADOR

EL SALVADOR

ETHIOPIA

FINLAND

FRANCE

FEDERAL REPUBLIC OF GERMANY

GABON

GHANA

GREECE

GUATEMALA

HAITI

HOLY SEE

HONDURAS

HUNGARY

ICELAND

INDIA

INDONESIA

IRAN

IRAQ

ISRAEL

ITALY

IVORY COAST

JAPAN

REPUBLIC OF KOREA

LEBANON

LIBERIA

LIBYA

LUXEMBOURG

MALI

MEXICO

MONACO

MOROCCO

NETHERLANDS

NEW ZEALAND

NICARAGUA

NIGERIA

NORWAY

PAKISTAN

PARAGUAY

PERU

PHILIPPINES

POLAND

PORTUGAL

ROMANIA

SAUDI ARABIA

SENEGAL

SOUTH AFRICA

SPAIN

SUDAN

SWEDEN

SWITZERLAND

SYRIA

THAILAND

TUNISIA

TURKEY

UKRAINIAN SOVIET SOCIALIST REPUBLIC

UNION OF SOVIET SOCIALIST REPUBLICS

UNITED ARAB REPUBLIC

UNITED KINGDOM OF GREAT BRITAIN

AND NORTHERN IRELAND

UNITED STATES OF AMERICA

URUGUAY

VENEZUELA

VIET-NAM

YUGOSLAVIA

The Agency's Sta tute was approved on 23 October 1956 by the Conference on the Statute of the

IAEA held at United Nations Headquarters, New York-, it entered' into force on 29 July 1957. The

Headquarters of the Agency are situated in Vienna. Its principal objec t ive is " to acce le ra te and enlarge

the contribution of a tomic energy to peace , heal th and prosperity throughout the world".

© I A E A , 1 9 6 4

Permission to reproduce or translate the information contained in this publication may be obtained

by writing to the International Atomic Energy Agency, Kärntner Ring 11, Vienna I, Austria.

Printed by the IAEA in Austria

October 1964

T E C H N I C A L R E P O R T S SERIES No. 34

REACTOR SHIELDING

R E P O R T OF A P A N E L ON REACTOR SHIELDING HELD IN VIENNA,

9 - 1 3 MARCH 1964

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1964

International Atomic Energy Agency. Reactor shie lding. Report of a Panel on Reactor

Shielding held in Vienna, 9 - 1 2 March 1964. Vienna, the Agency, 1964.

164 p. (IAEA, Technical reports s e r i e s no. 34)

6 2 1 . 0 3 9 . 5 3 8 .

REACTOR SHIELDING, IAEA, VIENNA, 1964 S T I / D O C / 1 0 / 3 4

FOREWORD

Reactor shie lding i s n e c e s s a r y that people m a y work and l ive in the vicinity of reactors without receiving detrimental biological e f fects and that the n e c e s s a r y m a t e r i a l s and instrumentat ion for r e a c t o r operat ion m a y function proper ly .

Much of the n e c e s s a r y theoret ica l work and experimental measurement h a s been a c c o m p l i s h e d in recent y e a r s . S c i e n t i s t s have deve loped s o m e very sophisticated methods which have contributed to a more thorough under-standing of the p r o b l e m s involved and have produced s o m e v e r y r e l i a b l e r e s u l t s leading to s ign i f i cant reduct ions in sh i e ld conf igurat ions .

A panel of experts was convened from 9 to 13 March 1964 in Vienna at the Headquarters of the International Atomic Energy Agency and under the chairmanship of Mr. E. P. Bl izard to d i scuss the present status of reactor shie lding. The part ic ipants w e r e prominent shie lding exper t s f rom m o s t of the l a b o r a t o r i e s engaged in th i s f i e ld throughout the world . They p r e -sented status reports descr ib ing the past h i s tory and plans for further development of reactor shielding in their countr ies and much valuable d i s -cuss ion took place on some of the most relevant aspects of reactor shielding. All this mater ia l i s presented in this report, together with abstracts of the supporting papers read to the Panel .

C O N T E N T S

STATUS REPORTS

Achievements and projec t s in Belgium in the f ie ld of reac tor shielding 1 H. Dopchie

Canadian reac tor shielding . 3 A. C. Whittier

Status report - Czechos lovakia 14 B. Stoces and A. Zbytovsky

Shielding r e s e a r c h in the Federa l Republic of Germany 21 D. Biinemann

Shielding r e s e a r c h in France 28 P. Lafore

Working and p r o g r a m m e s connected with shielding p r o b l e m s in Italy . 32 A. M. Moncassoli

Status report of shielding invest igat ions in Japan 37 M. Shindo

Summary of shielding work in Norway 57 U. Tveten

A summary of shielding r e s e a r c h in Sweden 63 E. Aalto and J. Braun

Status report on reac tor shielding r e s e a r c h in Switzerland 79 J. M. Pictet and A. Etemad

Shielding r e s e a r c h in the United Kingdom 87 J. Butler

Reactor shielding in the United States of A m e r i c a 98 M. Grotenhuis, P. S. Mittelman and E. P. Blizard

Some problems of b io log ica l shielding in reac tors 108 Yu. A. Egorov

Status report - ENEA 135 H. В. Smets

Status of shielding r e s e a r c h at EURATOM Ispra 136 H. Penkuhn

SUMMARIES OF DISCUSSIONS

Sources of radiation in reac tors 139 M. Grotenhuis, H. Dopchie, H. Smets and U. Tveten

Experimental fac i l i t i e s and experimentat ion 142 J. Braun, A. Aalto, G. Richter and A. C. Whittier

Methods for calculating radiation attenuation in sh ie lds 146 J. Butler, D. Biinemann, A. Etemad, P. Lafore, • A. M. Moncassoli, H. Penkuhn, M. Shindo and B. Stoces

Engineering des ign problems 152 j . M. Pictet, C. Cortissone, Yu. A. Egorov, P. Mittelman, J. Rastoin and A. Zbytovsky

GENERAL CONCLUSIONS 157

E. P. Blizard (Chairman)

ABSTRACTS OF SUPPORTING PAPERS 159

List of part ic ipants 163

STATUS R E P O R T S

A C H I E V E M E N T S AND P R O J E C T S IN B E L G I U M IN THE F I E L D O F R E A C T O R SHIELDING

H. DOPCHIE

Four reactors are at present operating in Belgium: the BR-1 and BR-2 which are r e s e a r c h and m a t e r i a l s test ing r e a c t o r s , the B R - 3 which has a power output of 11 MW(e) and the B R - 0 2 which i s the nuclear m o c k - u p of the B R - 2 .

In 1965 the BR-3 will be operating on the spec tra l - sh i f t principle , the moderator cons i s t ing of a var iable mixture of'heavy and light water . The VENUS reactor , the cr i t ical mock-up of the future BR-3 core , should com-mence operation in April 1964. By the end of 1964, the University of Ghent wil l have at i ts d isposal a swimming-pool type reactor . The fast reac tors MASURCA and HARMONIE were studied on behalf of the French Atomic Energy C o m m i s s i o n and EURATOM, jointly. F inal ly , there i s the VULCAIN project, for the purpose of developing a spectra l - sh i f t reactor for ship pro-pulsion. The land prototype i s now in the design stage.

Most of these reac tors r a i s e no shielding problems other than the con-ventional ones . The VULCAIN prototype shie ld wil l be designed in accor -dance with the spec i f i ca t ions character i s t i c of naval r e a c t o r s , in which an opt imizat ion of vo lume , weight and cos t i s of pr imary in tere s t . Study of these problems has begun. The radiation damage of the BR-3 pressure v e s -s e l i s a prob lem when equipped with the future s p e c t r a l - s h i f t c o r e . The l eve l and spectrum of neutron irradiation will be experimentally determined in the VENUS facil i ty.

The B R - 2 i s a very high-f lux r e a c t o r , the core tank of which i s con-tained in a pool. The horizontal b e a m - t u b e s , 30 cm in d i a m . , h a v e had to be sea led by means of shie lds which presented certain problems as regards plug heating and mechanical des ign, all of the tubes being removable under water.

The gamma heating of the BR-2 in-pi le experiments , reaching a maxi-mum leve l of 20 W/g , i s continuously studied. A probe has been developed for the measurement of this gamma heating inside the core while the reac -tor i s in operation.

Of course , designing the shield for the dismantling and observation ce l l s , for the shipping conta iners , e tc . , posed many prob lems which go beyond the scope of the present subject. It is interesting, nevertheless , to examine F i g . 1 which shows the a d m i s s i b l e point g a m m a act iv i ty in the B R - 2 d i s -mantling ce l l s as a function of the energy of the emitted photons. It will be noted that for energies below 2.5 MeV, the activity of the sources to be dealt with i s res tr i c ted by heating requirements rather than by-doses outside the shielding (2.5 m r / h ) . The wal-ls are of barytes concrete , 1.35 m thick and clad throughout with 10 m m s tee l .

All shie lding studies w e r e carr ied out by conventional methods and in no case was any research project, either experimental or theoretical, under-taken along these l ines . Works were accepted on the bas i s of their practi-

1

E ( M e V )

F i g . l

Max imum to le rance g a m m a - s o u r c e ac t iv i ty in the BR-2 dismantl ing cells

cal utility and no spec i f i c experiment for the ver i f icat ion of the theoret ical studies was ever carr ied out.

Among the main theoret ical problems encountered in shielding design, the following should be cited:

(1) G a m m a and neutron attenuation along s tra ight and s tepped ducts; (2) Back-scat ter ing of gammas and neutrons from air (skyshine), sol id

and liquid mater ia l s , and shielding of the ref lec ted radiations; (3) Gamma and neutron propagation at long distance in air; (4) Capture gamma s p e c t r a as a function of the energy of the incident

neutron; and (5) Gamma radiat ions from neutron ine las t i c scat ter ing . In conjunction with a F e r r a n t i - M e r c u r y computer, we use the Engl i sh

codes RASH B, RASH D and BURP for neutrons and GASLIT and GASH for gammas . We have drawn up the ET AR programme for the study of gamma attenuation in reactors and reactor shielding, in a two-dimensional cylindri-cal geometry.

2

CANADIAN R E A C T O R SHIELDING

A.C. WHITTIER

1. INTRODUCTION

It has been the pol icy in Canada to build radiation measur ing fac i l i t i e s into reac tor bulk sh ie lds when the r e a c t o r s are under construct ion. Radi-ation attenuation through the NRX shield was measured by means of a special plug [1]. T h e s e m e a s u r e m e n t s formed the b a s i s for the des ign of the NRU and NPD shie lds . Des ign calculations [2 ] and measurements [3] ontheNRU shie ld have been published. The m o s t r e c e n t sh ie ld m e a s u r e m e n t s w e r e c a r r i e d out on NPD [4].

NPD i s a D2O moderated and cooled,22 MW(e) (gross ) power reac tor . The core c o n s i s t s of 132 s i t e s fue l led with natural U 0 2 spaced at 26.0 c m in a square array . The coolant and fuel are contained in 0 . 4 2 5 - c m thick Z i r c a l o y - 2 s t r e s s tubes the outer radius of which i s 4.55 cm. Each s i te i s insulated from the moderator by a 0 . 5 3 - c m air gap between the s t r e s s tube ând a 0 . 1 3 7 - c m thick aluminium calandria tube. This core i s contained in a double -wal l ed a luminium calandria as shown in F i g s . 2 and3 . F i g u r e 3 shows how the inner wall i s tapered to conserve D 2 0 . Between the two walls, the inner one of which i s 0.635 cm thick and the outer 1.27 cm thick, i s H2O which acts both as a secondary neutron re f l ec tor and as a radiation shield.

The concrete primary shield i s 213.4 cm thick and has a 0 .318-cm thick m i l d - s t e e l l iner on the reac tor s ide . The c o a r s e and fine aggregates con-s i s t of i lmenite ore. The density of the concrete i s 3.57 g / c m 3 . In addition to the concre te sh ie ld on the s i d e s of the r e a c t o r , there are rotat ing end shie lds of i lmenite concrete poured into a s t ee l cas ing, and a laminated top shield of concrete , s t ee l and masoni te .

There i s no separate thermal shield, but the inner 30 cm of concrete i s cooled by water flowing in mi ld s t e e l pipes as shown in F i g s . 2 and 3. The outs ide d i a m e t e r of the p ipes i s 1.91 c m and the wal l t h i c k n e s s 0 .29 c m .

Steel rods are used to re inforce the concrete , but a large area in front of the centrel ine of the radiation measurements i s left free of rods.

During construction of the concrete shield m i l d - s t e e l s l e e v e s were im-bedded in it as shown in Fig. 4. The ends of the s l e e v e s were located along the l ines marked of radiation measurements" in F igs . 2 , 3 and 4.

P l u g s to f it the s l e e v e s w e r e made by pouring i lmeni te concre te into m i l d - s t e e l cas ings . Detai l s of the plugs are shown in Fig . 5. The ends of the plugs were fitted with a s m a l l aluminium capsule to hold f i l m or fo i l s . Plugs 1, 2, 3 and 4 also had a cavity to hold a smal l ionization chamber; this i s shown in F ig . 5. In plugs 1 , 2 , 3 and 4 was a cable tube which sp ira l l ed down to the ionizat ion chamber cavi ty .

Measurements were a lso made with a "solid" plug in hole No. 1 which had no ionization chamber cavity or foi l capsule at i ts end. This i s shown as plug No. 1* in F ig . 5. Slots just wide enough to take w i r e s w e r e cut at 7 . 6 - c m interva l s along this plug.

3

Fig. 2

End view (sect ion X-X) of NPD and shield (All dimensions in cent imetres)

2. RADIATION MEASUREMENTS

The radiation measurements were made over a period of about one year. During this t ime there w e r e changes in moderator l eve l , moderator purity and reactor power. Values of radiation f ie lds plotted in the various graphs have been normal ized to full power operation.

2.1. Thermal neutron flux

The absolute value of thermal neutron flux was obtained from irradiation of cobalt w i r e 0.0127 m m in d iameter at the end of plug No. 1*. Measure -ment of the variation of thermal neutron flux throughout the shield was done by m e a n s of manganese w i r e , gold w i r e and gold f o i l s . Some wire m e a -

4

Fig .3

Plan view (sect ion A- A) of NPD and shield ( All dimensions in cent imetres)

surements were obtained by fixing the short wires at 7 .6 -cm intervals along a flat twin- lead t e l ev i s ion cable and then running the cable down the sp ira l tube in plug No. 1. The ratio of the bare to cadmium-covered gold wire ac-t i v i t i e s was a l so m e a s u r e d by aff ixing cadmium-wrapped gold to the cable at 7 . 6 - c m i n t e r v a l s . Other w i r e m e a s u r e m e n t s w e r e obtained with plug N o . 1 * .

The thermal flux in the lower flux outer reg ions was measured by gold fo i l s attached to the inner ends of plugs 2 to 7.

5

OF RADIATION

/ MEASUREMENTS

[508

© ®

15.08 15.08 5.08 5.08

¡5.08 #

/—ELEVATION OF

/ REACTOR $

LOOKING INTO WALL FROM OUTSIDE SURFACE

MEASUREMENTS

Fig. 4

Plan view of the a r rangement of access holes

Numbers in circles are for ident i f ica t ion

(All dimensions in cent imetres)

Figure 6 shows a typical distribution of thermal neutron flux through the concrete shield. Dis tances are measured from the face of the concrete along a radius f rom the core axis . The large r i s e in flux at the outer edge of the concrete i s due to radiation leakage f rom nearby ion-chamber beam ho le s .

The rat io of bare to cadmium c o v e r e d gold i s shown in F ig . 7 and the ep i thermal gold act ivat ion in F ig . 8.

2.2. Fast neutron flux

The fast neutron flux was m e a s u r e d by irradiat ing sulphur pe l l e t s in the c a p s u l e s at the ends of the p lugs . The i rrad ia ted p e l l e t s , nominal ly 0.5 g in weight, were burned and the P 3 2 activity measured. Figure 9 shows

6

PLUGS 5. 6. 7

PLUG 1*

Fig. 5

Plugs used for the measurements (All dimensions in cent imetres)

the variat ion of fast flux as measured by the sulphur activation. To obtain flux f r o m act ivat ion it was a s s u m e d that the S 3 2 (n ,p) P 3 2 c r o s s - s e c t i o n was 0.06 b.

During the reac tor c o m m i s s i o n i n g period and before full power oper-ation had been reached, the H2O re f l ec tor was drained temporari ly . While the HoO was out one sulphur irradiation was made. The observed flux leve l was 1.5X109 n cm "2 s .

2.3. Gamma radiation

Gamma radiation was measured by means of a Geiger counter* , a graphi te -wal led ionizat ion chamber and f i lm. The Geiger counter was 1.27 cm long by 0.635 cm in diam. It was surrounded by 0.58 m m of lead. It was p laced at the fu l l - in pos i t ion of the plug being used and drawn back 7.6 cm at a t ime. Like the Geiger counter, the graphite ionization chamber was drawn back f rom the innermost posit ion of plug No. 1 in 7 . 6 - c m s teps

Phillips, Type 18529.

7

DISTANCE WTO NPD SHIELD [ c m )

Fig. 6

T h e r m a l neutron flux

• Gold wire • Gold foil X Gold foi l О Gold wire Д Gold foi l

8

22

20

18

16 ZI 14

8

4 к

0 10 2 0 30 ( 0 SO 60 7 0 80 90 100 110 120 130

DISTANCE INTO NPD SHIELD C " n )

Fig. 7

Ratio of bare to c a d m i u m - covered gold ac t ivat ion

• Gold wire О Gold foi l

CO

10»

107

10е

*• 5 t 105 >

о < Ы g 10' « _J bJ cr

103

102

10 I ' l ' I M I N M n i l l i l M I I M i n i M M H I I H I I I i i n H I I M I I I I I I I I I 0 10 20 30 < 0 SO 60 70 6 0 9 0 100 110 120

DISTANCE INTO NPD SHIELD ( c m )

Fig. 8

C a d m i u m - c o v e r e d gold ac t iva t ion

• Gold wire X Gold foi l

(Normal ized at 31 cm)

DISTANCE INTO NPD SHIELD ( c m )

Fig. 9

Fast neutron flux

о = 0 . 0 6 b

and read ings taken at e a c h s tep . It was u s e d only in the v e r y high f i e l d s at the inner end of plug No. 1. The X - r a y f i l m p a c k s * * w e r e enc losed in lead boxes 3.18 cm X 2.54 cm X 0.318 cm; the lead was 0.635 mm thick. They were placed in the capsules at the ends of the plugs.

The gamma ray attenuation i s shown in Fig. 10. The large r i s e in field at the outside surface of the concrete i s due to radiation leakage from nearby ion-chamber beam holes . While the H 2 0 r e f l e c t o r was out, a g a m m a - r a y measurement was made throughout the shield with the Geiger counter. This i s shown in Fig. 11.

3. CALCULATION

Calculat ion of the radiat ion f i e ld s through the NPD shie ld i s being carried out using removal -age-d i f fus ion theory. Both MAC [5] and a s e r i e s of p r o g r a m m e s wr i t ten at AECL [6] are being u s e d in this work. P r e l i -minary ca lculat ions indicate that whi le agreement between calculated and m e a s u r e d g a m m a radiat ion and fas t neutron flux i s good, there i s lack of a g r e e m e n t in the c a s e of the t h e r m a l neu t ron flux. This work i s continuing.

* * Dupont, Type 544.

10

DISTANCE INTO NPD SHIELD Ce™)

F ig .10

G a m m a radiation

• A • V Geiger counter Д Ionizat ion chamber Y Film badge

11

DISTANCE INTO NPD SHIELD С с m l

Fig. 11

Gamma radia t ion

(H 2 0 ref lector drained)

» O ® Geiger counter

12

R E F E R E N C E S

[1] BELL, R. E. , MILLAR, С. H. and ROBSON, J. M. , unpublished Rpt (1951). [2] HENDERSON, W.J. and WHITTIER, A . C . , Handbook of Shielding and Heat Production Calculations

for the NRU Reactor, Rpt. AECL 403, revised (1957). [3] ROBSON, J. M. , The Attenuation of Neutrons by the Side Shield of the NRU Reactor, Rpt AECL 932 (1959). [4] WHITTIER, A.C. and HILBORN, J. W. , NPD Bulk Shield, Rpt R64CAP 15 (9 Mar. 1964). [5] PETERSON, E. G., MAC - "A Bulk Shielding Code", Rpt HW-73381 (Apr. 1962). [6] FOWLER, A. G., Bendix G-20 Computer Programs for use in Reactor Shield Design, Rpt AECL 1678 (1962).

13

STATUS R E P O R T - C Z E C H O S L O V A H A

B. STOCES AND A. ZBYTOVSKY* '

The work in the f ie ld of reactor shielding done up to the present in the Czechoslovak Social ist Republic i s connected c lose ly with the Nuclear Power Plant des ign p r o g r a m m e . The bas ic sh ie ld des ign was e laborated in the USSR. Some specia l problems which arose during a more detailed elabora-t ion of the sh i e ld d e s i g n w e r e then s o l v e d in the Czechos lovak S o c i a l i s t Republ ic .

Great attention was given especial ly to the calculation of heat generation in r e a c t o r s t r u c t u r e s , with s p e c i a l r e g a r d to the p r e s s u r e v e s s e l and the a luminium h e a v y - w a t e r container , and in the s t e e l thermal sh ie ld . Data on the acces s ib i l i ty of different parts of the reactor after shut-down for in-spection, repair and dismounting were also obtained and special requirements for permissab le amounts of impurit ies in structural mater ia ls were pointed out.

The b io log ica l sh ie ld of this type of h e a v y - w a t e r moderated and g a s -cooled reactor involves s o m e spec i f i c problems, one of which i s the design of the top sh ie ld . The deformat ion of the r e a c t o r - v e s s e l head l i m i t s the p e r m i s s i b l e length of the fuel charge tubes. The top bio logical shie ld has to be placed, there fore , within the v e s s e l under the v e s s e l hèad.

The following essent ia l requirements must be fulfil led by the top shield: neutron dose in the reactor hall must be reduced to a permiss ib le level; the p r e s s u r e v e s s e l head has to be protected against activation; the shield ma-ter ia l m u s t not contain hydrogen because of the poss ib i l i ty of i sotopic ex-change with the moderator v ia the coolant gas , and the shie ld has to be as thin as p o s s i b l e because of the l imi ted length of the p r e s s u r e v e s s e l .

Two variants were considered: (1) s tee l , graphite and boral slabs; and (2) s t ee l and graphite s labs . A s e r i e s of calculations were carried out using the RASH В r e m o v a l mult igroup method [1]. A sh ie ld cons i s t ing of s t e e l and graphite s labs containing 25 vol.% of s tee l was shown to be sat isfactory [2]. Some of the r e s u l t s of these calculations are shown in F i g s . 12 and 13. The dependence of neutron flux at the ex terna l boundary on the t h i c k n e s s of the f i r s t iron layer of a two- layer iron-graphite shie ld of constant thick-n e s s of 140 c m i s shown in F i g . 14. F l u x e s in a sh i e ld of the s a m e con-figuration as in Fig . 13 with borated graphite ( Ea = 5 cm"1 for thermal neu-trons) are shown in F ig . 15.

The calculations were checked experimental ly on a mock-up which was introduced into the r e s e a r c h reactor WWR-S instead of the thermal column. The re lat ive attenuation of fast , epithermal and thermal neutrons was mea-s u r e d . F a s t neutrons w e r e m e a s u r e d with thresho ld d e t e c t o r s of p h o s -phorus and sulphur pé l le t s , and with scinti l lation detectors using ZnS(Ag) or ZnS(Ag) in paraffin. Epithermal and thermal neutrons were measured with activa-tion fo i l s of Аи.ОугОз and Pbl . The mock-up is placed on the thermal co-lumn carriage of the reactor WWR-S. Some measurement results are shown in Fig. 16.

Presented by B. Stoïes.

14

Fig .12

Neutron fluxes in i ron-graphi te shield ca lcu la t ed with RASH В

As there are annular gaps for cooling gas around the plugs in the channel of the top shie ld (Fig.17) , neutron t r a n s m i s s i o n through these gaps was m e a -sured by m e a n s of sc in t i l la t ion d e t e c t o r s with a 1 - m m s c i n t i l l a t o r . The detector moved ver t i ca l l y in the s y m m e t r y plane behind the channel. The resu l t s of fast neutron t r a n s m i s s i o n m e a s u r e m e n t s using the ZnS(Ag) sc in-t i l lator in one variant of plug des ign are shown in Fig. 18.

To avoid neutron s treaming where the s t ee l plugs p ierce the top shield, the use of a nickel alloyed s t e e l was cons idered. At present the s treaming of neutrons of energy about 26 keV in s t a i n l e s s s t ee l , cas t iron and carbon s tee l i s compared experimental ly . It i s assumed it wil l be poss ible to avoid the use of nickel s tee l .

The measurements on a mock-up of a charge-d ischarge machine shield are prepared us ing burnt-up fuel e l e m e n t s of the r e s e a r c h r e a c t o r a s

15

Fig .13

Neutron fluxes in lamina ted i ron-graphi te shield

мп u n iron i i graphite

a g a m m a - r a y s o u r c e . A c o n c r e t e b io log i ca l sh i e ld ¿n p lace of d i f ferent ducts wil l be checked with planary gamma s o u r c e s a lso consist ing of burnt-up fuel e l ements .

In the f ie ld of spec ia l shielding mater ia l s the powder meta l technology of a luminium-coated boral plates of a complex f o r m was developed. S e r -pentine concretes are being developed using local serpentine as well as spe-cial heavy concre tes .

As it i s a s sumed that other power reac tors of the HWGCR type will be built in C z e c h o s l o v a k i a , attention i s concentra ted on the deve lopment of neutron attenuation calculation methods suitable for non-hydrogenous shields, or for aluminium, heavy-water , s t ee l , concrete s lab sh ie lds . In detection methods attention i s given to the measurement of fast neutron flux and spec-

16

THICKNESS OF THE IRON LAYER f e m )

Fig. 14

External boundary fluxes for d i f ferent thicknesses of the iron layer of a shield of overa l l thickness of 140 c m , consisting of two layers: iron and graphi te

The thickness of the first layer (iron) is changed in 2 0 - c m steps from 0 - 140 c m О = values of ex te rna l boundary fluxes of a l amina ted shield, consisting of 20 c m iron and eight

a l ternat ing slabs of 1 0 - c m graphite and 5 - c m iron

tra in the presence of a high background of epithermal neutrons and gamma radiation as wel l as to the measurement of the moderation spectrum in heavi-ly absorbing media.

Two types of neutron s p e c t r o m e t e r s have been developed to check ex-perimental ly the calculation of the integral fast neutron fluxes in the reactor pressure v e s s e l of neutrons transmitted through a heavy-water ref lector and s t e e l thermal shie ld; one us ing s e m i c o n d u c t o r d e t e c t o r s , the other us ing pulse d i s c r i m i n a t i o n c i r c u i t s .

The bas ic faci l i ty for this experimental shielding programme will con-s i s t of a 4 0 - c m diam. s o u r c e plate of about 60 W which wi l l be mounted in the thermal column of the reac tor WWR-S. A R a - B e source of an inten-s i ty of 107 n / s was made for s o m e m e a s u r e m e n t s . It i s used with o thers in combination with a graphite p r i s m for detector cal ibrat ion and standard foi l production. It i s a l so used for personnel training.

The work described here i s being carried out principally in the Nuclear R e s e a r c h Institute of the Czechoslovak Academy of Sc iences , in the Skrda-Works R e s e a r c h Inst i tute in P l z e ñ , at T e c h n i c a l U n i v e r s i t y in Brno .

17

1

Fig. 15

Neutron fluxes in l amina ted shield of iron and borated graphite

£ a = 5 c m - 1 for 2200 m / s neutrons

* in m\ iron I I borated graphite

18

Fig. 16

Neutron a t tenuat ion in the shield mock- up (All lengths in cent imetres)

Fast neutrons: A phosphorus Д sulphur Epithermal neutrons: • РЫ foils The rma l neutrons: О Au foils x Dy 20 3 foils

Fig. 17

Channel m o c k - u p (All lengths in mi l l imetres)

19

Fig. 18

Fast neutron transmission through gap

Scale 1 : 1 . 5

R E F E R E N C E S

[1] AVERY, A. F. et al. , Methods of Calculations for Use in the Design of Shield for Power Reactors, Rpt AERE-3216 (1960).

[2] STOäES, В., Mnohogrupov^ vjipoïet neutronového stfnïnf г urstev grafitu a oceli, Rpt ÚJV-1032 (1964).

20

S M E L D I N G R E S E A R C H IN THE F E D E R A L R E P U B L I C O F GERMANY

D. BÜNEMAN

1. INTRODUCTION

Severa l fac i l i t i e s for shie lding r e s e a r c h are instal led at the R e s e a r c h Reactor Geesthacht (FRG). Extended studies are performed in co-operation with the Institut für Reaktorphysik at Geesthacht and industrial companies interested in this f ie ld. An independent experimented programme has been undertaken by the Arbei tsgruppe für bautechnischen Strahlenschutz of the Technische Hochschule Hannover in connection with the r e s e a r c h fac i l i t i e s at Geesthacht .

2. EXPERIMENTS WITH COLLIMATED BEAMS

2.1. ESTAKOS experiments

Experiments with col l imated beams are of special interest for shielding re search due to the following theorem:

The r e s p o n s e to an infinite monodirect ional beam through a plane shie ld can always be replaced by an integral over the response of a f i -nite monodirect ional beam.

This theorem i s true for both the neutron flux and the y -rad ia t ion and even m o r e genera l for a mixed radiation f ie ld . A few r e s t r i c t i o n s for the application of this theory mus t be mentioned:

(1) The assumption of an exactly coll imated beam i s not real ized. How-ever i t i s p o s s i b l e to extrapolate for an ideal co l l imat ion by changing the d iameter of the beam.

(2) Side e f f e c t s might disturb the r e s u l t s . Th i s point can be checked by changing the upper s i d e wal l and rep lac ing i t by the s a m e m a t e r i a l a s that of the exper imenta l s h i e l d arrangement s o as to ensure that even up to a cer ta in th ickness no m a j o r e r r o r i s introduced by th i s e f f e c t .

(3) The finite s i z e of the measuring device might introduce some error s ince mean va lues over a f inite reg ion are taken. This error can be e s t i -mated by compar ing the e x t e n s i o n of the f lux pattern with the s i z e of the m e a s u r i n g d e v i c e .

The exper imenta l tank of ESTAKOS (Fig. 19) has a length of 2.75 m para l l e l to the beam. Perpendicu lar to the beam it i s quadratic with a s ide length of 60 c m . The c r o s s - s e c t i o n of the co l l imat ing beam hole can be changed by flooding different concentric aluminium cyl inders yielding three di f ferent d i a m e t e r s (1 c m , 5 c m , 15 cm) . By these m e a n s it i s p o s s i b l e to e s t i m a t e the inf luence of the b e a m ' s d iameter on the m e a s u r e m e n t r e s u l t s .

The shielding plates can be placed in the experimental tank up to a thick-n e s s of 2 m. To avoid a transmitt ing gap between the wal ls and the plates ,

21

to to

Fig. 19

Elevations of the ESTAKOS fac i l i ty

the inner s ides of the walls have a profile with an alternating structure chang-ing each 6.5 cm. The mean s i ze of the plates i s 60 cm X 60 cm.

In v iew of the r e s t r i c t i o n s to these e x p e r i m e n t s , the r e s u l t s obtained are sat i s factory . The integration' over the plane perpendicular to the beam axis can be res tr ic ted to a relat ively smal l region around the beam axis as shown in Fig. 20. There, as an example, the integrand of

2тг frd (r, x) dr и '

i s plotted as a function of the radius r for four different distances along the beam axis , where d y ( r , x ) i s the measured y - d o s e rate .

S e v e r a l m e a s u r e m e n t s have been p e r f o r m e d for shie ld arrangements consist ing of water, polyethylene, iron and lead. The facility i s used mainly for optimalization studies.

Fig. 20 .

y - m e a s u r e m e n t in pure water

Beam d iamete r = 15 c m Ionization chamber vo lume = 100 c m 3

2.2: Experiments at the Technische Hochschule Hannover

Similar measurements are being performed by the group from Hanover who are test ing sh ie ld e f f i c i e n c i e s of concre te s and other mater ia l s . Some r e s u l t s can be found in [1].

23

3. BULK SHIELD EXPERIMENTS (ESTAGROP)

Another experimental facil ity has been built to test a real reactor shield in a mock-up experiment , s ince the measurements with a col l imated beam give re su l t s for a m o r e or l e s s monodirect ional source only. The applica-bi l i ty of s u c h an e x p e r i m e n t wi l l be l i m i t e d for the fo l lowing r e a s o n s :

(1) The spectrum of the swimming-pool reactor at Geesthacht may differ from that of the simulated reactor; and

(2) Only plane geometry of s labs with f inite s i z e can be used to imitate a radial sec t ion of a curved reactor shie ld .

In order to obviate the f i r s t point many of the e x p e r i m e n t a l f a c i l i t i e s have a uranium converter l ike the BSF at Oak Ridge which then acts as an ideal source direct ly comparable to theoretical calculations. We felt, how-ever , that there was need for a very strong radiation source which we could not achieve by a converter. Since the reactor was not only used for shielding experiments a complicated core geometry resulted.

Curved geometry i s s imulated by plane s labs . Fortunately plane geo-metry i s better access ib le to theory than any other geometry and it is thought that c o m p a r i s o n of m e a s u r e d and ca lcu lated quantit ies in plane g e o m e t r y would g ive a good cr i t er ion of the re l iabi l i ty of the theory to be applied to curved geometry, in order to extrapolate for the curved reactor shield. This t e s t of re l iab i l i ty wi l l check the ideas underlying opt imal izat ion of shields, which i s of spec ia l in teres t for nuclear ship reactor sh ie lds . By comparing the experiments with ESTAKOS and ESTAGROP for the same shield arrange-ment , it should even be p o s s i b l e to re la t e ESTAKOS r e s u l t s to r e a c t o r sh ie lds too.

One of the major e f for ts has been an attempt to get m o r e detai led in -format ion about the neutron- and y - s p e c t r a as functions of d i s tance in the shield. Therefore a broad programme for the development of special radi-ation detectors has been started.

In front of the reactor core a large frame has been instal led into which the s labs are suspended (Fig . 21). The base of the pool i s 7 m X 7 m and i s f i l led up to a height of 9 m with deminera l i zed water . Specia l care has to be taken to avoid contaminating the water . Since d e m i n e r a l i z e d water i s ex tremely aggres s ive to some shield mater ia l s , such as concrete or or-dinary iron, the s labs were canned or coated by a protective aluminium and epoxy c o v e r . The s i z e of the s labs i s rather large , 2 m X 2.5 m, to avoid s ide e f f e c t s .

The ESTAGROP II facility at the radiation channel is used for performing spec ia l exper iments outside the pool. Here all kinds of reactor mater ia l s can be used, avoiding diff icult ies with the pool water. The f i rs t experiment to be performed by the Hanover group wil l study a mock-up of a special r e -actor shield.

3.1. Comparison of theory and measurement

The f a s t neutron flux i s ca lcu lated by r e m o v a l theory approximat ing the reac tor by a sphere , fol lowing ANL 6000, with constant source distr i -

24

STEEL P L A T E

REACTOR CORE

- D E T E C T O R HOLDER

Fig. 21

Elevations of ESTAGROP I faci l i ty

bution of f i s s i o n n e u t r o n s . In a p r e l i m i n a r y a n a l y s i s s i x m e t h o d s w e r e c o m p a r e d :

(1) One-group r e m o v a l theory with r e m o v a l c r o s s - s e c t i o n s from ANL 5800;

(2) 17-group r e m o v a l theory; (3) Remova l theory with the Albert -Welton kernel; (4) Remova l theory with the Wick-Welton kernel; (5) As (3) but with fas t neutron bui ld-up fac tors f rom NDA 15-C-39; and (6) A s (4) but wi th f a s t neutron bu i ld -up f a c t o r s f r o m NDA 1 5 - C - 3 9 .

Compar i son of the r e s u l t s with fas t Дих m e a s u r e m e n t s in water showed the bes t a g r e e m e n t for methods (5) and (6), as was to be expected f rom the e x t e n s i v e t h e o r e t i c a l input c o n n e c t e d wi th t h e s e m e t h o d s . H o w e v e r , they cannot be used for l a y e r s with v e r y di f ferent build-up factors because of the break down of the bui ld-up concept . F o r the c a l c u l a t i o n s u s i n g 17 groups or the Wick-Wel ton or A l b e r t - W e l t o n k e r n e l s a c o r r e c t i o n for the i n c r e a s e of the F e r m i age with water th i cknes s i s n e c e s s a r y . As no computer code w a s a v a i l a b l e f o r t h i s , al l fur ther c a l c u l a t i o n s of the f a s t f lux w e r e done by a one-group remova l theory which gave poorer but s t i l l reasonable agree-ment. The thermal flux was calculated by the RASH code using f ive diffusion groups. The group constants w e r e taken from AERE 3216 and RSA-TMP-183.

The c o r e -y-dose r a t e s w e r e c a l c u l a t e d in ten groups f r o m 0 - 1 0 MeV by the bui ld-up theory us ing the s a m e approximat ion of the r e a c t o r c o r e as at the r e m o v a l f lux . The capture y - d o s e s w e r e c a l c u l a t e d s e p a r a t e l y by the GASH-code for inf in i te plane s l a b s . The e n e r g y s p e c t r u m of the c o r e y - r a d i a t i o n w a s c a l c u l a t é d f r o m the f i s s i o n 7 - s p e c t r u m and c o r r e c t e d for capture y ' s in the c o r e .

D i f f i c u l t i e s a r o s e in the i n t e r p r e t a t i o n of the e x p e r i m e n t a l r e s u l t s in a conf igurat ion containing an air tank. It was found that the deviat ion of the p r e d i c t e d t h e r m a l f lux v a l u e s i s due to the n e g l e c t e d outf low f r o m the a ir tank into the surrounding pool . Par t of t h e s e neutrons reappear in the tank wi th l o w e r e n e r g y m o d e r a t e d by the s u r r o u n d i n g pool w a t e r . In o r d e r to c o r r e c t for this e f f ec t , the m e a s u r e d t h e r m a l f lux was u s e d in the ca l cu la -tion of the capture y - d o s e rate . This gave better r e s u l t s for the total y - d o s e r a t e . S ide e f f e c t s w i l l appear in c y l i n d r i c a l s h i e l d s ana logous to th i s example . New theore t i ca l methods have to be developed to account for such e f f e c t s .

4. RADIATION DETECTORS

Only conventional methods have been used for the measurements carried out s o far . Since there i s need for more detai led m e a s u r e m e n t s of fast neu-tron and y - s p e c t r a , new measur ing dev ices are being developed, e. g. proton-r e c o i l s p e c t r o m e t e r and semiconductor dev i ce s for fas t neutrons, pair spec-t rometer for y -radiat ion .

26

5. MEASUREMENT BY MEANS OF OTHER RADIATION SOURCES

At Hanover 7 - rneasurements are performed at the betatron of the Tech-n i s c h e Hochschule to obtain the bui ld-up f a c t o r s for e n e r g i e s in the range of 4 - 1 5 MeV.

In o r d e r to check s h i e l d c a l c u l a t i o n s , m e a s u r e m e n t s have b e e n done at the s m a l l power r e a c t o r at Kahl.

6. THEORETICAL WORK

Para l l e l to these experiments severa l groups are performing theoretical studies . At the Univers i t i e s of Stuttgart and Hanover, at different industrial companies and at the FRG, extensive work i s being done in transport theory as well as in the conventional methods. A new code for optimalization studies i s being developed.

R E F E R E N C E S

[1] FUTTERMENGER, W. , GLUBRECHT, H. , NIEMANN, E.G. and SCHULTZ, H . , Bestimmung der Schwä -chungsfaktoren von Strahlenschutzmaterial mit Hilfe differentieller Streustrahlungsmessungen, Kerntechnik 4 (1962) 504.

27

SHIELDING RESEARCH IN FRANCE

P. LAFORE

1. INTRODUCTION

Shielding r e s e a r c h as an independent subject in France dates from 1956. The importance of these studies has been ref lected in the contribution which they have made to power reac tor design and in the resultant savings in ex-penditure for c iv i l eng ineer ing and m a c h i n e r y for the r e m o v a l of m o b i l e s h i e l d s .

2. ADMINISTRATION

In the F r e n c h Atomic Energy C o m m i s s i o n , a l l p r o b l e m s re la t ing to reactors are dealt with by a Directorate of Atomic Reactors, one department being concerned in particular with research (Reactor Research Department) and another with construction (Reactor Construction Department).

The Reactor Shielding Research Division forms part of the Reactor Re-search Department and participates in reactor design from the prel iminary planning s tage onwards. Owing to the great th ickness of the m a t e r i a l r e -quired, shielding plays a very important part in reactor design. An under-standing of the p r o b l e m s re la t ing to radiat ion during operat ion and shut -down, and to fue l and coolant fluid activity, i s imperat ive f rom the outset . Knowledge of the properties and use of materials i s equally necessary. Cal-culat ions obtained by s e m i - e m p i r i c a l methods enable s o m e opt imizat ion r e s e a r c h to be undertaken at this stage. During the des ign phase detai led calculat ions are made by the construction companies under the supervis ion of the Reactor Shielding Research Division. These relate to biological shiel-ding, heat production in the shie ld and predict ions concerning the high flux respons ib le for damage to mater ia l s .

When the reactor has been built and started up, it i s the task of the Re-actor Shielding R e s e a r c h Div is ion to a s c e r t a i n whether the shie lding p r o -vided i s e f fect ive and above all whether there i s agreement between experi -ment and calculation. The purpose of this check i s to get lower safety fac-tors in the future, if poss ib le , and to improve methods of calculation.

3. FACILITIES UTILIZED

The Reactor Shielding R e s e a r c h Div i s ion n u m b e r s approx imate ly 60 p e r s o n s and u s e s s e v e r a l e x p e r i m e n t a l f a c i l i t i e s . T h e s e include:

(1) NAÏADE I, in s ta l l ed near the ZOE r e a c t o r and operat ing with a natural uranium s lab 2 cm thick (an e f fec t ive d iameter of 60 cm i s the one most commonly used) .

(2) The TRITON s w i m m i n g - p o o l reac tor , main ly used in shie lding s tudies , inc ludes an a c t i v e - w a t e r loop, by m e a n s of which the s econdary sh ie lds required for l ight -water r e a c t o r s can be studied. In addition, the

28

removable c o r e p e r m i t s of m e a s u r e m e n t s being c a r r i e d out on l a r g e - s c a l e m o c k - u p s in a central chamber of the pool m e a s u r i n g 10 m X 6 m an l 10 m in depth.

(3) Another core , NÉREÏDE, which i s s i tuated near a 2 m X 2 m alu-minium window enables a large neutron source to be placed in a compartment without water in which l a r g e - s c a l e mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavit ies , and of reactor shielding of greater thickness than that in NAÏADE I. In this case an enriched uranium s lab i s s i tuated behind the graphite r e f l e c t o r and the fac i l i ty i s ca l l ed NAÏADE II.

(4) A SAMES 600 keV a c c e l e r a t o r i s used for monoenerge t i c neutron s tudies .

Instrumentation s tudies are an important part of our work, mainly in the measurement of fast neutrons and their spectra by activation detectors . Of late, our attention has been directed towards the use of (n, n1 ) (rhodium) reac t ions and of heavy de tec tors for l ow- f lux m e a s u r e m e n t s . The s i m u l -taneous use of a large number of detectors poses automation problems. With our instal lat ion we can count 16 detec tors s imultaneously . We use an ana-l y s i s programme in conjunction with a digital computer. Neutron spectrum studies are conducted with nuclear emuls ions and a l i thium-6 semiconductor s p e c t r o m e t e r . The key problem in these spec trum m e a s u r e m e n t s is, that of ana lys i s .

As to the m a t e r i a l s used, the r e s e a r c h carr ied out in France involves ch ie f ly graphite , i ron and concre te at v a r i o u s ' t e m p e r a t u r e s up to 800°C. Di f ferent compounds , borated and non-borated and of d e n s i t i e s up to between 1 and 9 are under consideration. Problems connected with applica-tions are also being studied.

4. NEUTRON ATTENUATION

4. 1. Removal cross-section

Although t h e o r e t i c a l s tud ie s have provided a s c i e n t i f i c b a s i s f o r the concept of r e m o v a l c r o s s - s e c t i o n for monoenerget ic neutrons and we have shown that the choice of removal c r o s s - s e c t i o n s in t e r m s of energy enables us to find the v a l u e s of fas t f l u x e s in water , we are continuing to u s e p r o -g r a m m e s based on Albert and Welton1 s concept of removal c r o s s - s e c t i o n , modi f ied to supply an answer in t e r m s of fas t neutrons. This method en-ables us to break the c o r e into point s o u r c e s and to add the contr ibut ions at v a r i o u s po ints of the sh ie ld ing . The subsequent u s e of a t h e r m a l d i f -fus ion group m a k e s it p o s s i b l e to f o r e c a s t the t h e r m a l f lux, and thus the captures , in the neighbourhood of the iron p la te s .

4. 2. Diffusion

In other c a s e s it i s preferable to use diffusion calculations. One group i s often enough for fast f lux (phosphorus). By m e a n s of a s e m i - a n a l y t i c a l method (the P E N E L O P E В p r o g r a m m e ) we can c o m p a r e t h e r m a l f lux

29

attenuation f r o m s o u r c e s of var ied geometry , us ing the f o r m a l i s m of a 2-group dif fusion theory. The method i s part icu lar ly e f fec t ive inasmuch as it a l l o w s of t r a n s i t i o n s be tween g e o m e t r i c a l a r r a n g e m e n t s (d i sc , p lane, sphere , cy l inder) .

4. 3. Cross-sections

Instead of us ing s e m i - e m p i r i c a l methods of propagation der ived f r o m mock-up shielding exper iments , we can cons ider us ing calculat ions based on a knowledge of m i c r o s c o p i c c r o s s - s e c t i o n s . The ex i s t ing d e v i c e s are able to m e m o r i z e a l i s t of c r o s s - s e c t i o n s in suff icient detail and these pri-mary constants can also be mechanically processed so as to yield secondary constants, averaged in accordance with some particular formal ism and some part icular energy band. We u s e c r o s s - s e c t i o n s sent to us on cards by the AWRE. With these cards we can get constants for numerical methods pro-grammed by other methods (multi-group, Sn , Monte Carlo). The accuracy which can be achieved in connection with c r o s s - s e c t i o n s and angular distr i -butions i s constantly improving and it may be hoped that in a few years from now such methods wil l allow us to make full use of computer developments .

At the present t ime the preparation of a multi-group code i s under con-sideration. The diff icult problem is to know how such a c o d e could be used and whether it would be equally accurate in all applications. The application of particular interest to us i s that of very fast (between 20 MeV and 10 keV) neutron propagation, so that we may d i s c o v e r the s p e c t r a r e s p o n s i b l e for damage to m a t e r i a l s and to human subjects .

If it w e r e at p r e s e n t p o s s i b l e to know the r e s p o n s e s of d i f ferent m a -t e r i a l s with suff ic ient accuracy, methods s i m i l a r to those used in relat ion to the bui ld-up fac tor might be cons idered h e r e too. A M o n t e - C a r l o c a l -culation which we made on the propagation of fast neutrons in graphite shows that such methods can apparently be designed.

The most important study at present appears to us to be that of neutron and gamma propagation in heterogenous media with cav i t ies and that of neutron diffusion in l a r g e - s c a l e cav i t i es . Our work at present mainly r e -vo lves around t h e s e two quest ions .

5. INTERNATIONAL COLLABORATION

We w e r e part icularly interested in the establ ishment of an information centre for shielding studies at Oak Ridge and the development in the United States of Amer ica of codes for shielding calculations.

Among the obs tac l e s in the way of international collaboration, mention must be made of the variety of equipment used and the need for private f i rms working on shielding to be equally wel l equipped with them. It i s a l so de-s irable that the language in which such codes are written should not change too much in the future. However , great p r o g r e s s can be expec ted in the working out of c o d e s e a s i l y in terchangeable b e t w e e n v a r i o u s nat ions .

30

B I B L I O G R A P H Y

BOURGEOIS, J. et LAFORE, P. , Dispositif expérimental pour l 'étude des protections, Rapport CEA 825 (1958).

BEAUGE, R., MILLOT, J. P. et RASTOIN, J. , Etude expérimentale de l 'atténuation des neutrons dans le béton

ordinaire, Rapport CEA 1186.

BEAUGE, R. , MILLOT, J. P. et RASTOIN, J. , Etude expér imentale de l 'a t ténuat ion des neutrons dans les

mélanges p lomb-eau, Rapport CEA 1187.

HENRY, B. et RASTOIN, J. , Age à la résonance de l ' indium des neutrons de la réaction T(p, n)He3, Rapport

CEA 1632.

DEVILLERS, C. , LAFORE, P. , LE DIEU DE VILLE, A. et RASTOIN J . , Protection de machines de chargement

et déchargement. Considérations d'ensemble, source de rayonnement, Note CEA 374.

MILLOT, J. P . , Etude de la diffusion des neutrons rapides. Section e f f i c ace de déplacement (Thèse de la

Facil i té des Sciences de Paris) Rapport CEA 2142.

DULIEU, P., Mesure des flux de neutrons intermédiaires et rapides au moyen de diodes au silicium, Note CEA

391.

Les Ingénieurs du Service d'Etudes de Protections de Piles, Formulaire sur le calcul de la protection des réacteurs,

Rapport CEA 2253.

CONDAT, M. J. , LAFORE, P. et RASTOIN, J . , Les bétons spéciaux dans la protection des piles, Rapport CEA

2312.

BRISBOIS, J . , KO, P. et RASTOIN, J . , Etude du détecteur de rhodium pour la dosimétrie des neutrons rapides,

Note CEA 410.

LETOURMY, F. , CONDAT, M. et PEPIN, P . , Note au sujet d'un béton réfractaire spécial de protection pour

les piles de puissance type EDF, Note CEA 423.

LOTT, M. , OCERAIES, Y. , PASSE, S. et RASTOIN, J . , Etude d'un dosimètre à neutrons rapides, Note CEA

445.

BRISBOIS, J . , « E t u d e de la section e f f icace de la réaction (n,y) de l 'o r : Calcul de l 'ac t ivi té d'un détecteur

d'or épa i s» . Neutron dosimetry, Vol. I, IAEA, Vienna (1963) 181.

PASSE, S . , « M e s u r e de spectres de neutrons rapides â l ' a ide d'émulsions nuc léa i r e s» , Neutron Dosimetry,

Vol. I, IAEA (1963) 481.

31

WORK AND P R O G R A M M E S C O N N E C T E D WITH SHIELDING P R O B L E M S IN I T A L Y

A.M. MONCASSOLI

1. INTRODUCTION

The work that has been done on shielding and that i s s t i l l going on can be briefly descr ibed under two main headings: the experimental part, which deeds with the m e a s u r e m e n t of neutron and g a m m a penetrat ion in laminar sh i e lds , and the part concerned with theoret ica l analys i s and development of re l iab le shie lding des ign methods . Part icular r e f e r e n c e i s made to the work being done under the Euratom-FIAT-Ansa ldo naval contract , with the participation of CNEN (FIAT reports: Nuclear Energy Div i s ion GN).

The work performed has served to es tabl i sh greater confidence in c e r -tain methods of calculat ion and to show the l i m i t s to which the solut ion of certain problems are subject. The work in which we are engaged deals with problems of neutron penetration in laminar water- lead shie lds and also with the deve lopment of methods of ca lculat ion , the i m p r o v e m e n t of m e a s u r e -ment techniques and the evaluat ion of exper imenta l e r r o r s .

Th i s s u m m a r y m a k e s no c l a i m to being a complete picture of the r e -s e a r c h on sh ie ld ing which i s being conducted in Italy. It m e r e l y r e f l e c t s the act iv i ty with which the author i s persona l ly acquainted.

2. SHIELDING EXPERIMENTS

2.1. Neutrons

A dev ice has been developed for m e a s u r i n g the penetrat ion of f i s s i o n neutrons in laminar shields (ETNA f i s s ion plate shielding facility of the Avo-gadro RSI reactor) . With this device measurements , based on the activation technique, are being made of the penetration of fast, epithermal and thermal neutrons in laminar shie lds of iron and water [1]. Use i s made of a number of threshold detectors ; Dy, Au, Au on Cd, Mn, In, S, Ni. A compar i son of the r e s u l t s obtained with v a r i o u s de tec tors i s now under way. S imi lar exper iments have been undertaken to study penetration in laminar H 2 0 - P b sh ie lds .

2.2. Gamma rays

Exper iments , based on the photographic emulsion technique, have been conducted to invest igate g a m m a penetration in laminar sh ie lds of lead and polyethylene [2]. The source used was the bremsstrahlung spectrum of the 31 MeV betatron (Molinette-Turin).

32

2.3. Specific activity of water

D o s e rate mapping was c a r r i e d out in the pump r o o m of the Saluggia Avogadro RSI reactor [3], where piping, heat exchanger and pumps aire highly radioact ive due to the spec i f i c activity of the pr imary coolant water; The purpose of that work was to t e s t theore t i ca l methods for predic t ing main gamma dose on the inside wal ls of the plant container of a water cooled and moderated reactor. Calculated dose rate distribution has been found to agree with experimental data within 10%.

3. DEVELOPMENT OF CALCULATION METHODS

CNEN, on i ts IBM-704 computer at the CNEN computer cen tre in Bologna, has developed the available p r o g r a m m e s for the calculation of shielding. The following p r o g r a m m e s are at present in operation [4]:

GRACE I (NAA-SR-3719) GRACE II (NAA-SR-MEMO-4649) S PIC I (WAPD-TM- 196) SPAN 2 (WAPD-TM- 176) MAC (HW-73381)

The following programmes have been worked out and are now in operation:

FOG-S: A method for three-group (one of them a removal group) calculation of neutron penetration in F e - H z O shie lds , in accordance with the method suggested inGNEC-187 [5] ;

GAMMONE: Monte-Carlo method for the calculation of gamma attenu-ation and build-up in laminar shie lds [6]. Exponential transformation i s used for speeding up the problem s o -lution. A typical problem can be run in one hour on the IBM-704.

The following programme is now being developed:

COFIAT: Calculation of the activity of f i s s i o n and corros ion prod-ucts in the pr imary and purif ication circuit of a water reactor [7].

4. COMPARISON BETWEEN THEORY AND EXPERIMENT

4.1. Neutrons

Three different methods of analysis have been used for comparison with experimental results:

(1) A multigroup one-dimensional diffusion equation code, AIM-6,.which handles a 16-group c r o s s - s e c t i o n l ibrary with f ive down scattering groups.

33

(2) A multigroup attenuation code MAC which evaluates neutron f luxes for 31 energy groups . A r e m o v a l theory i s used for the evaluation of un-collided flux in slab geometry. The flux generated in the shield by scattered neutron flux is evaluated by means of multigroup diffusion theory.

(3) Final ly a three-group diffusion theory code, FOG-S, with s e m i e m -pir ical water fast group constants .

Comparison between experimental resul t s and calculations are reported for a typical iron-water configuration in Figs . 22, 23 and 24 (fast, epithermal and thermal act ivat ions) . The bes t agreement i s obtained with the three-group method. MAC compar i son shows v e r y poor agreement for thermal flux.

id* s e i

2

10Э

8 6

i

2

6 i

2

101

8 6 «

2

10°

0 10 2 0 30 <0 5 0 6 0

Fig. 22

Fast ac t ivat ion (sulphur foils)

FOG 3 groups AIM 16 groups MAC 31 groups

® Experimental points

4.2. Gamma rays

Two different approaches are used for g a m m a - r a y attenuation evalua-tion, an exponential kernel corrected for build-up and a Monte-Carlo method.

1 . 4 5 - 1 0 4 1 1 1 1 I

* \

* i 1

1

V V

в s

*

H2 ) i • 20 •Í H2( Fe H ¡0 •я нр Fe

1 % i

34

Fig. 23

Epi thermal act ivat ion (Au /Cd foils)

FOG 3 groups AIM 1С groups MAC 31 groups

® Experimental points

10 2 0 3 0 4 0

Fig. 24

Thermal ac t iva t ion (Au and Au /Cd foils)

FOG 3 groups AIM 16 groups MAC 31 groups

oí Exper imental points

The build-up factor i s evaluated for a laminated shie ld by weighting infinite media build-up factors of each slab over its relaxation length. A code (GAMMONE) has been put into operation on the IBM-704 using Monte-Carlo technique for the evaluation of gamma penetration and build-up in laminar shie lds . Comparison with experiments shows quite good agreement for both the methods.

4.3. Specific activity of water

A comparison of calculation and m e a s u r e m e n t has shown that there i s a discrepancy by a factor of two in the calculation of the speci f ic activity of water in the case of the Avogadro RSI reactor .

R E F E R E N C E S

[1] CHINAGLIA, B. and MONCASSOLI, A . M . , Neutron attenuation experiments in slab type iron-water shieldings, Rpt GN-36 (Sep. 1963).

[2] Di G RAZIA. G. , Determinazione sperimentale dell ' efficienza di schermi compositi piombo-polietilene per at tenuazione di raggi gamma di elevata energía e confronto con la teoria, Rpt GN-35 (Apr. 1963).

[3] GALLO CURCIO, A . , MONCASSOLI, A .M. and TODISCO, S. , Radiation doses from a water reactor loop, Rpt GN-26 (lun. 1963).

[4] CENTRO DI CALCOLO DEL CNEN, Programmi nucleari in dotazione al Centro di Calcolo (Oct. 1963). [5] FOZZI, V. and MONCASSOLI, A .M. , FOG-S-Modifica del FOG-I per il calcolo di schermi a strati di

ferro e acqua, Rpt GN-31 (Feb. 1963). [6] CRISCUOLO, L. , GAMMONE-Programma Montecarlo per il calcolo del l 'a t tenuazione dei raggi gamma

attraverso schermi laminari compositi, Rpt GN-38-41 (Oct. , Nov. 1963). [7] POTESTÀ, A. , COFIAT-Programma per il calcolo dell 'attività dei prodotti di fissione e di corrosione,

àpt GN-34 (Feb. 1963). [8] MONCASSOLI, A.M. and PATERLINI, W. , Verifica teórica della dosi misurate durante il funzionamento ' a potenza del reattore BR-3, Rpt NTI-FN-22 (Feb. 1964).

36

STATUS R E P O R T O F SHIELDING INVESTIGATION IN J A P A N

M. SHINDO

1. INTRODUCTION

Ten y e a r s have passed s ince the peaceful use of nuclear energy started in Japan. The Japan Atomic Energy R e s e a r c h Institute (JAERI) w a s e s -tabl i shed in 1954, and i m m e d i a t e l y p r o c e e d e d with the c o n s t r u c t i o n of a research reactor. Operation was initiated in every field of nuclear research, but for a long t ime shielding studies were only a smal l part of health-physics ac t iv i t i e s , incorporating such a s m a l l group of r e s e a r c h workers that they could not properly organize their act iv i t ies . The f i rs t symposium in Japan on nuc lear energy was held in 1957. Most of the papers p r e s e n t e d in the field of reactor shielding were limited to shielding materials and their fabri-cation. In the f i r s t stage of our invest igat ions, our ef forts were devoted to pract ical des ign studies of reac tor shielding. A s a resul t of these s tudies , it was found that the formulae at hand for calculations were inadequate, but at that t ime no electronic computer was available in Japan nor were theoreti-cal calculations very actively undertaken.

P r o b l e m s on nuc lear ship shie lding had been inves t iga ted at the Ship Research Institute, the f o r m e r Transportation Technical Research Institute, s ince 1956 and many fruitful re su l t s had been obtained. About that t ime the Japan A t o m i c Industry F o r u m s tar ted a c t i v i t i e s and took the in i t ia t ive in organizing shielding research. Research workers in the shipbuilding industry in particular have been seriously studying shielding problems.

A few years after the f i rs t symposium, problems concerning more funda-mental s tudies were treated by many r e s e a r c h workers . Shielding exper i -ments us ing radioisotopes w e r e carr ied out and many fruitful r e su l t s w e r e obtained. They are descr ibed in the next sect ion. Medium s i z e e lectronic computers became available in Japan, permitt ing a theoret ical study group to make an active contribution. They (the SCG-group) produced some codes, sponsored by the Nuclear Powered Ship Research Assoc iat ion of Japan, and their re su l t s are a l so descr ibed in the following sect ions . This constituted the second stage of our invest igat ions .

A swimming-pool reactor, JRR-4 (Japan Research Reactor-4), has been under construction at JAERI s ince 1962 and wil l become cr i t ica l in autumn 1964. A f t e r c h a r a c t e r i s t i c t e s t s , it wi l l be a v e r y powerful tool f o r the sh ie ld ing i n v e s t i g a t i o n s . We are now enter ing the third s tage of our i n v e s t i g a t i o n s .

2. EXPERIMENTAL STUDIES

2.1. Gamma-ray shielding

2 . 1 . 1 . P e n e t r a t i o n and s c a t t e r i n g

The m o s t important data for g a m m a - r a y penetrat ion ca lculat ions are bui ld-up f a c t o r s . Miyasaka et al . obtained d o s e bui ld-up f a c t o r s of eight

37

/

mater ia l s (water, graphite, ordinary g las s , aluminium, ordinary concrete, heavy concrete , iron and lead) for Co60 plane monodirectional sources using a long ionization chamber and with the method developed by KIRN et al. [ l ] . The data were compared with the theoret ical values (except graphite, g lass and concrete) by the moments method using Berger ' s finite effect correction, and the r e s u l t s showed good a g r e e m e n t within 19% up to s e v e n mean f r e e paths for all these mater ia ls except water. For water, the discrepancy was about 37% for seven mean free paths. To trace the origin of this discrepancy, further Monte-Carlo calculat ions were performed and the resul t supported the ca lcu la ted v a l u e s . The s a m e exper iment has been made by T a m u r a et al. for Cs131 plane monodirectional sources , and Monte-Carlo calculations are also in progress .

Many experimental studies of build-up factors have been made for point i s o t r o p i c s o u r c e s and plane monodirec t iona l s o u r c e s , but there a r e few studies covering plane isotropic sources . Using the long ionization chamber and a point isotropic source, a gamma-ray penetration study has been made for plane distributed isotropic sources by Tsuruo and Toyota. The sources used were Au!98(0.412 MeV), Cs137 ( 0.6 6 2 MeV), Co«* (1.17 MeV and 1.33 MeV) and Na24 (1.37 MeV and 2.75 MeV). Aluminium, iron and lead were used as

THICKNESS

Fig. 25

Dose bu i ld -up factors of multi- layers for lead and ordinary concrete

- - - Ordinary concrete monolayer Lead monolayer

38

shielding mater ia l s , with se lec ted th i cknesse s up to eight mean free paths. The dose-rate distribution obtained experimentally was extrapolated by calcu-lation to the c a s e of infinite plane isotropic sources . The difference in value between l imited plane s o u r c e s and infinite plane s o u r c e s was l e s s than 3%. From these results , gamma-ray attenuation curves and dose build-up factors were obtained for infinite plane isotropic sources . x

A s a .part of the g a m m a - r a y penetrat ion s tud ies ment ioned above, Miyasaka et al. obtained dose build-up factors of mul t i - layers . Experiments w e r e p e r f o r m e d for the combinat ions of l e a d - i r o n , l e a d - a l u m i n i u m , and lead-ordinary concrete for Co60 plane monodirect ional s o u r c e s . The c o m -pos i t e sh ie ld ing was arranged in a var ie ty of s e q u e n c e s . Typica l r e s u l t s are shown in F ig . 25 for the e f fec t of var ious arrangements of heav ier and l ighter m a t e r i a l s .

Other inves t iga t ions on m u l t i - l a y e r s of water , i ron and lead f o r Co60

plane monodirect ional s o u r c e s were performed by MOCHIZUKI et al. [2-4] using the method of Kirn et al. In the investigations, an empirical formula was proposed. The formula enables quick and easy calculat ion of build-up factors for mul t i - layers composed of an arbitrary combination of mater ia ls , by us ing known build-up factors for finite m o n o - l a y e r s of the mater ia l s in-cluded in the compos i t ion .

Experimental r e s u l t s indicate that the build-up fac tors agree we l l with theoret ical values for mono- layers of iron and lead, and mult i - layers of the m a t e r i a l s invest igated in the c a s e of re lat ive ly thin lead thickness , but not for water and m u l t i - l a y e r s with thick lead l a y e r s .

Furthermore, experiments were performed for build-up factors of rela-tively thick lead mult i - layers , and comparison was made between the experi-menta l and theore t i ca l r e s u l t s ca lculated f rom the genera l i zed formula of KALOS [5] and from an empir ica l formula derived during the investigation. The r e s u l t s showed good a g r e e m e n t with theore t i ca l v a l u e s for i r o n - l e a d mul t i - layers , but not for l ead- irçn mul t i - layers with very thick lead layers .

Dose build-up factors of iron slabs for Ccfi0 and Cs1 3 7 infinite line sources were obtained experimental ly by Umeda. Furthermore , experiments were p e r f o r m e d to obtain dose build-up fac tors of s t e e l p ipes for Co60 and Cs 1 3 7

i so tropic l ine s o u r c e s placed on the centre axis of the pipe. In order to obtain the scattering of reactor g a m m a - r a y s , Nakata et al .

measured gamma-ray dose-rate distribution in water subjected to radiations from a beam hole of the reac tor JRR-1 . The re su l t s were compared with those from attenuation ca lculat ions us ing the energy spec trum obtained by SUGURI et al. [13] with a mul t ip le -crys ta l gamma-ray spectrometer; good agreement was obtained.

ISHIMATSU [14] obtained the energy flux and dose rate distributions of g a m m a - r a y s from a point source of Co60 in an infinite water medium. The experimental resul t s were compared with calculated resul ts by the moments method. A sa t i s fac tory agreement with ca lculat ions was obtained for the dose build-up factors given by the experimental resu l t s . On the other hand, a slight discrepancy seemed to exist between the experimental and the calcu-lated energy flux spectra.

In pool-type reac tors , it i s general ly considered that Na24 atoms in the water contribute most ly to the gamma-ray dose on the water surface of the r e a c t o r . To d e t e r m i n e th i s quanti tat ively f o r the c a s e of TRIGA Mark II

39

reactor (St. Paul 's University), Hattori et al. have measured the attenuation of gamma rays in the shielding water of the reactor tank. The results were compared with ca lcu la ted va lues and good a g r e e m e n t was obtained. The gamma-ray spectrum in the water had been obtained using nuclear emulsions soaked with heavy water.

2 . 1 . 2 . Reflection i

Nakata et al. have measured the energy distribution of back-sca t tered gamma rays from lead, iron and ordinary concrete slabs using Co60 and Cs137

sources . The resu l t s show that the backscattered rays from a concrete wall are made up from m u l t i - s c a t t e r e d rays , while backscat tered rays from a lead s lab c o n s i s t mainly of s ingly scat tered photons and charac ter i s t i c X - r a y s .

NAKATA [9] has a l so measured backscattered gamma rays with a lead plate p laced on a c o n c r e t e wal l and found that the lead plate markedly de-pres sed the ref lect ing photon flux. Dose rates in a model room of which the wal ls were covered with lead plates amounted to about one half of that of un-covered wal ls .

The backscattered gamma rays from sca t terers of paraffin, aluminium, iron, tin and lead of e f f ec t ive ly s e m i - i n f i n i t e th ickness were measured by HYODO [10] . Point i sotropic sources of Co60 and Cs1 3 7 were placed on the front face of the scat terer and a 3-in diameter by 3- inlong Nal(Tl) scintillator was moved in a c irc le centred on the position of the source . The energy and number albedos, the angular distribution of scattered energy and number of photons, and the energy distribution were obtained for each combination of gamma source and scat terer . Table I shows the value of the number and the energy albedo obtained in the experiments.

TABLE I

THE NUMBER A L B E D O A AND THE ENERGY A L B E D O AE

Scatterer Number a lbedo Energy a lbedo

Scat terer CÍ" Co60 Cs137 Co6 0

Paraffin 0 . 5 7 ± 0 . 0 2 0 . 4 6 ± 0. 02 0 . 1 9 ± 0. 01 0 . 1 1 2 ± 0 . 0 0 5

Alumin ium 0. 59 ± 0. 02 0. 52 ± 0. 02 0. 20 i 0. 01 0 . 1 3 6 ± 0. 005

Iron 0 . 4 2 ± 0. 02 0 . 4 2 ± 0. 02 0 . 1 6 ± 0. 01 0 .120 ± 0. 005

Tin 0. 22 * 0. 02 0. 25 i 0. 02 0 . 1 0 ± 0. 01 0. 088 ± 0. 005

Lead 0. 09 i 0. 02 0. 17 ± 0 . 0 2 0. 04 ± 0. 01 0. 059 ± 0. 005

Reflection dose build-up factors for Co60 plane monodirectional sources were obtained both experimentally and by Monte-Carlo calculations byTsuruo and Miyazaki. For aluminium and lead, the two resul t s showed good agree-ment and, furthermore , agreed a l so with data obtained by the US National

40

Bureau of Standards. In the study, re f l ec t ion dose bui ld-up f a c t o r s w e r e obtained for ordinary concrete , heavy concrete , copper s lag, and ordinary concrete coated with s e v e r a l kinds of paint containing oxides of lead, barium and tin.

The e f f ec t of sur face m a t e r i a l on backsca t t ered g a m m a - and X - r a y s w e r e a l so studied. F o r Cs1 3 7 gamma rays , backscat ter ing f rom concre te s e e m e d to be great ly d e c r e a s e d by cover ing with lead lamina, but l e s s so with iron. Backscattering from lead might be slightly decreased by covering with tin. The problem i s complicated for 70-keV X - r a y s . Generally, ma-t er ia l s whose atomic number i s about 22 and between 60 and 70 are usefu l for degradation of backscattered radiations; lead-iron-titanium lamina placed on thick concrete or wood slabs were found to be effective.

2 . 1 . 3 . Volume source shielding

Few exper iments have been made of g a m m a - r a y shielding for volume sources as yet . Furuta and Kanemori have studied gamma-ray penetration problems for cy l indr ica l s o u r c e s . The spec ia l feature of the exper iments was the use of a large number of plastic spheres each containing a Co60 pellet in the c e n t r e . Many f o r m s of cy l inder w e r e const i tuted by t h e s e p las t i c spheres and their effect was analogous to the cylindrical sources of Co60

aqueous solutions. Dose rates were also measured from a bare cyl indrical source and from the cy l indr ica l source covered with a cy l indr ica l she l l of iron.

2. 1 . 4 . Streaming and labyrinth

Takumi and Mitsui made exper iment s on t h e r m a l - n e u t r o n s t r e a m i n g through ducts. The ducts were formed with eight blocks of heavy concrete , 500 m m X 500 m m X 500 m m , and the ir d i m e n s i o n s could be changed cont inuously .

Radiation leakage f rom pipes in water has been invest igated in detai l . The pipes used were 16 ~ 200 mm in inner diameter, 0 ~ 1500 mm in length and made of vinyl chloride, iron, lead, and aluminium.

SAKUTA [11] has made a s e r i e s of measurements of gamma-ray distri-butions in the entrance ways of an irradiation faci l i ty and showed how bends contributed to decreas ing gamma-ray dose rate at the entrance.

2. 2. Neutron Shielding

Yamaki made exper iments on neutron attenuation in l a y e r s of iron and water with 1 4 - M e V neutrons produced by the D(T, n)He 4 r e a c t i o n with a Cockroft-Walton type acce l era tor . Dependence of removal c r o s s - s e c t i o n s on the thickness of iron was obtained.

Experiments have been made on the attenuation of fast neutrons, thermal neutrons and g a m m a rays in m u l t i - l a y e r s of i ron and water with neutrons from natural uranium converter , moreover , neutron attenuation in lead and m e s o n i t e w e r e m e a s u r e d in the b e a m hole of a w a t e r - b o i l e r type r e a c t o r (JRR-1) by Tomabechi and Ishi i .

41

2. 3. Instrumentation

Information on the gamma-ray spectrum penetrating shielding materials i s very important for the investigation of shielding problems. Many attempts have been made to measure such continuous gamma-ray spectra with a single crys ta l . ISHIMATSU [12] developed a s imple method of obtaining the cor -rected gamma-ray spectrum and made a response matrix for a 2- in long by 2 - i n d iameter Nal(Tl) c r y s t a l using that method. It i s applicable to broad paral le l gamma rays up to 1.5 MeV and i ts application to the incident gamma rays of uniform angular distribution has a l so been d i scussed . HYODO [13] has been studying the response matrix for a 3- in long by 3- in diameter Nal(Tl) crysta l for the purpose of obtaining response -correc ted spectra of scattered g a m m a - r a y s . The matr ix was made for axial incident gamma rays having energ i e s ranging from 0 to 1.440 MeV.

Another method of g a m m a - r a y s p e c t r o m e t r y has b e e n deve loped by Furuta. This i s a scinti l lat ion spec trometer based on the mul t ip le -crys ta l g a m m a - r a y s p e c t r o m e t e r by MAIENSCHEIN [14] . A t r ia l m o d e l of the spectrometer used f ive 1 - in long by 1- in diam. Nal(Tl) crysta ls and the use of 19 crysta ls (each 1- in long by 1- in diam. ) or ring type crystals for the purpose of increasing the eff iciency i s envisaged.

A g a m m a - r a y dos imeter with a response expres s ib l e in roentgens and high sens i t iv i ty with a smal l sens i t ive volume has been perfected by Furuta

mr/h

10' • 10' 10'

PULSE RATE

Fig. 26

Sensitivity of y - ray pulse dosimeter a , b , c and d show pulse rates obtained for the natural background radiations with respective

scintil lators at the same point

Scint i l la to; A: 5 m m x 5 m m d i a m . B: 10 m m x 10 mm d i a m . C: 10 m m x 25 m m d i a m . D: 25 m m x 25 m m d i a m .

42

and Kinbara. This d o s i m e t e r (the g a m m a - r a y pulse dos imeter ) employs a p las t i c s c i n t i l l a t o r and a s p e c i a l e l e c t r o n i c technique which i n v o l v e s no counting l o s s and, even with a 5 m m long by 5 m m d i a m e t e r s c in t i l l a tor , natural background radiations can be measured. Figure 26 shows the sens i -tivity of this d o s i m e t e r for s e v e r a l s i z e s of sc int i l la tors . The gamma-ray pulse dos imeter i s sensit ive to X- or gamma-rays with energies above 50keV, and has air-equivalent response for energ ies above 200 keV, with maximum response at 80 keV (+14%). Furthermore , a l inear temperature dependency i s obtained (0.6% per deg Q under 40°C.

As a neutron spectrometer for shielding experiments, the detector must have a high e f f i c i e n c y and be i n s e n s i t i v e to g a m m a f luxes . The threshold detector i s i n s e n s i t i v e to high g a m m a f l u x e s . The min imum neutron flux that can be measured by the threshold detector was calculated and a rev ised orthonormal expansion method, minimiz ing the mean square re lat ive devi-ation, was proposed by Fuse to obtain the fast neutron spectrum. Threshold de tec tors , however , are a rather rough dev ice for neutron s p e c t r o s c o p y . For prec i se measurements of the neutron spectrum, a proton recoi l neutron spectrometer i s being prepared by Furuta and F u s e . It wil l be used for the spectrum measurements of the JRR-4 core neutrons. Scintillation detectors usua l ly respond to both types of radiat ions f r o m a m i x e d g a m m a - r a y and neutron radiation f ield and so neutrons must be discriminated from gamma-rays in the neutron exper iment s on reactor shielding. FUSE [15] has tried to find the optimum conditions for the discrimination of neutrons from gamma-rays using a method s imi lar to that of OWEN [16] . Several types of photo-mult ip l iers were examined for this purpose and the resu l t s showed that the degree of discriminat ion depend on the type of photomultiplier tube. A clue to the mechanism of scintillation in anthracene i s also suggested in that there may be some dif ference between the spectra of the fast and slow components of the light of scintil lation, both of which are distributed around 4000 Â.

3. THEORETICAL STUDIES

3. 1. Transport and Monte-Carlo calculations

The theoretical study of shielding in Japan started with the design of the JRR-3 reactor . • In the shielding design calculation for this research reactor undertaken from 1956 to 1957, the gamma-ray build-up factors of magnetite concrete were calculated by the moments method [17] . An electronic com-puter suited for these calculat ions was not at hand in Japan at that t ime, so IBM-602A and hand calculat ions were applied.

The Monte-Carlo method has b e c o m e a matter of in teres t to many in-v e s t i g a t o r s s i n c e e l e c t r o n i c c o m p u t e r s b e c a m e ava i lab le . Us ing the Monte-Carlo calculation in the correlated sampling and expected value method for gamma-ray transmiss ion and reflection problems, TSURUO and SIBUYA [18] veri f ied BERGER's expectation [19] by showing that the ratio of gamma-ray t r a n s m i s s i o n at the c o r n e r of a s e m i - i n f i n i t e quadrant to the c o r r e -sponding penetrat ion in a s e m i - f i n i t e medium b e c a m e constant at a t rans -m i s s i o n d is tance of 2 - 3 mean f ree paths.

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Let the t r a n s m i s s i o n build-up factor of photon obliquely incident upon the corner of the quadrant be B(°°/4) and let the penetration build-up factor of photon obl iquely incident upon the s e m i - i n f i n i t e m e d i u m be B(°°/2), then

к = B(°°/4)- l B(w/2 ) - l ( 1 )

b e c o m e s constant. F igure 27 shows the value of к in the c a s e of a photon of energy 1.25 MeV when the incident angle i s 45° , the medium being a luminium.

0 1 2 3 4 5 6 JJ t

Fig. 27

Monte- Carlo ca lcu la t ion of the ratio of build- up factor for a corner of semi- inf ini te quadrant В (°°/4) to that for a s emi - in f in i t e medium В (°°/2), p lane monodirect ional source, incident angle 45°

к - В (« /4 ) - 1/B (°°/2) - 1

Number flux k j j for s emi - in f in i t e quadrant of Al : E„ = 1 . 2 5 MeV (Tsuruo) Energy flux kg for s emi - in f in i t e quadrant of Al : E0 = 1 . 2 5 MeV (Tsuruo)

д-O -

— x — Energy flux kg for slab of H 2 0 : E0 = 4 MeV (Berger) - - - E l - - - Energy flux kg for slab of Fe : E0 = 1 MeV (Berger)

TOYODA and KONDO [20] tried the bias sampling technique for calcu-lating the fast neutron penetration in water . Since 1962, codes for gamma-ray penetrations in s labs and multiple layers have been developed and com-pared with e x p e r i m e n t s of g a m m a - r a y bui ld-up f a c t o r s . Calcu lat ions of t r a n s m i s s i o n in a sphere and a cy l inder w e r e a l s o made by Umeda .

The conditional Monte-Carlo technique i s now a matter of interest and NAKANO [21] has tr ied to apply this technique to the g a m m a - r a y problem in double- layer penetration. Monte-Carlo calculations on complicated geo-metr i e s have been tried by severa l investigators [22], but meaningful results have not ye t been obtained, owing to computer speed l imi ta t ion . S e m i -analyt ic Monte -Car lo ca lcu lat ions on s t r e a m i n g through ducts are now planned [23] and good r e s u l t s are expec ted .

In the r e s p o n s e matr ix method, the three kinds of r e s p o n s e m a t r i c e s , i . e . the number current, the energy current and the energy flux, are evalu-ated at the s a m e t ime by a Monte-Carlo calculation. The examples of build-up f a c t o r s shown in F i g . 28 a r e those not of "energy flux" but of "energy current". The energy flux build-up factors computed by the response matrix method a g r e e w.ell with the r e s u l t s of other Monte-Carlo ca lcu la t ions and experiments . Those f igures wil l be i l lustrated in a paper at the forthcoming United Nations Geneva Conference [24] .

44

Fig. 28

Energy bu i ld -up factor for mul t ip le layer by the response mat r ix method Source energy E0 = 1 . 5 MeV

Angular distribution of the plane cone source 1 . 0 > cos 9 0 > 0.9

© H 2 0 - P b - H 2 0 mul t ip le layer X H 2 0 - Pb mul t ip le layer

The direct numer ica l integration of the transport equation has b e c o m e a matter of interest . A numerical integration code of the transport equation for slab geometry, EOS, has been developed by KATAOKA and TAKEUCHI [25] for the NEAC 2206 computer . Some of the re su l t s obtained by EOS-1 code, code for gamma-ray calculation,are shown in Fig. 29. Build-up factors deduced from these resu l t s coincide wel l with other calculations by moments method and Monte-Carlo method. EOS-2 i s the code for neutron calculation.

The numer ica l integrat ion code of Bol tzmann equation for spher ica l ly symmetr ica l geometry, NIOBE, originally devised for the IBM 704 and 7090, has been prepared by Kataoka and Takeuchi for the NEAC 2206 computer . Severa l points to be borne in mind in applying the NIOBE to prac t i ca l p r o b l e m s have been r e v e a l e d .

3.2. Design techniques

Scient i s t s and eng ineers have been interes ted in developing s imple and accurate e m p i r i c a l formulae for des ign ca lcu la t ions . Ono and the author have dev i sed ref ined approximate calculat ion methods for slab shielding to a volumetric source. These are based on the replacement of orthodox source g e o m e t r y by other su i table g e o m e t r y : A s p h e r e i s r e p l a c e d by a par t ia l spherical shell , and a cylinder by a partial cylindrical shel l . Thus the inte-gration i s s impl i f ied and reduces error .

The geometrical e f fects of shields were investigated by the author. For a spher i ca l or cy l indr ica l source , four kinds of sh ie lds , shown in F i g . 30, were invest igated. On unscattered flux ф at point P, the following re lat ion was obtained:

¿a * *Л a ^ ä <¿c (2)

45

OS -

C O •0.6 -0.4 -02 ОЛ 02 o.« 0.6 0.8 10

rig.^i

Calcu la ted y - r a y spectrum of energy flux in the finite slab shield of EOS-1 code

(a) Dif ferent ia l angular spectrum

water 60 cm plane monodirec t ional < 0 MeV 0 .9862 E = 2 . 0 MeV E = 0 . 5 MeV E = 0 . 2 5 MeV

Medium Thickness Source

E„

(b) Dif ferent ia l energy spectrum

Medium Thickness Source

E» %

iron 20 cm plane monodirec t ional < 0 MeV 0 .9662

(c) Different ia l energy spectrum

Medium iron + water Thickness 4 m . f . p . + 5 . 3 m . f . p . Source plane monodirect ional E0 1 . 0 MeV w„ 0 .9862

Fig. 30

Typica l shapes of shields for spherical source and cyl indr ica l source

(a) Slab shield (b) Shell shield, slab thickness t is smal ler than x0. the dis tance between source surface

and observation point (c) Shell shield, thickness t is infini tely thin and the a t tenuat ion coef f ic ien t fi is inf in i te ly

large so that /jt = p is f in i te constant (d) Shell shield, slab thickness t is equal to x 0 , the dis tance between source surface

and observation point , and o c Ob s o j < Oa for unscat tered flux at point P.

Formulae express ing фа, i^and фс were obtained, as was also their approxi-mate formulae . Ratios of ф3 to фь and фс, and geometr i ca l e f fect functions, were calculated for the spher ica l and cyl indrical s o u r c e s . Figure 31 shows the g e o m e t r i c a l e f fec t functions for spher ica l geometry , corresponding to F i g s . 31 (a) and (b) r e s p e c t i v e l y , S1(x0/R0,/Lit) and S2(x0 /R0 , /ut).

Kataoka has made a theoret ical study of the combination of the ef fect ive r e m o v a l theory and the di f fus ion equation for neutron t r a n s m i s s i o n ca l cu -la t ions . The t h r e e - g r o u p theory, in which the f i r s t group i s the r e m o v a l flux and the third i s the thermal flux, i s re lat ive ly simple and can be applied to multiple l a y e r s of water and metal . The upper l imit of the second group has been obtained by tr ia l and e r r o r so that the experimental resu l t s agree with the calculat ions . Kataoka obtained the upper l imit of the second group f r o m a t h e o r e t i c a l study of neutron di f fus ion in w a t e r c o m b i n e d with the exper imenta l r e s u l t s obtained on LIDO by COOPER [26] . Using s i m i l a r methods , a way has been found of obtaining the d i f fus ion length and other constants of the second group, in good agreement with experimental resu l t s .

47

Fig. 31

Geometrical e f fec t function for shell shield of spherical source

V l K ft) = V 4 = ®a/»a

and i i

S 1 ( k , p ) = / e x p ( - p secß)du/ f exp( -p /u )du uo uo

Sj (k ,p) = / exp( -PMj(к ,u ) /k )du / / exp( -p /u )du , uo uo

where

0a. 0C. ®d> s e c ß . U = cos 0 , p = jjt, к = x„/R0, are illustrated in Fig. 30 i

The express ion for the intensity of gamma radiation which i s generated in an infinite slab shield due to thermal neutron capture and e scapes a c r o s s the front and rear f a c e s of the mater ia l has been obtained by HOKKYO and KIT AZUME [27] for given fast and thermal neutron currents incident on the shield. The result i s based on the two-group diffusion approximation for the thermal neutron distribution and i s a natural extension of ILIFFE's resul t [28] which does not take the incident fast neutrons into account . Another express ion i s obtained for the case of a water shield based on a fast removal-group theory in connection "with a thermal-group diffusion equation, and the ca lcu lated i n t e n s i t i e s d i f fer by about 40% between the two e x p r e s s i o n s .

48

Approximate calculations of gamma-ray transmission in multiple layers have a l so been a subject of study. MOCHIZUKI and TANAKA [29] extended KALOS's formula [5] for build-up factors of water and lead lamination, taking into account the resul ts of their gamma-ray experiments on multiple layers . They proposed approximate formulae for build-up factors of double layers , other than water and lead laminas, and of four and six layers of double laminas of water and iron.

Comparing BRODER's formula [30] for build-up fac tors of mult iple layers for a point isotropic source with Kalos* s formula for vertically incident gamma rays on a plane and with the resu l t s of Mochizuki's experiment, a revised Broder's formula for vertically incident gamma rays on a plane was proposed by Kitazume:

В(miXi, u2x2, - - - -mnxn) = Вц ( ц Л + - - - -+nNxN) + { в ( ц Л . - - - - í V ^ . j )

- B^íÍMjX^ +ííN_1xN_1)j-exp(-ornNxN)f (3)

where Bf/^Xj—Mnxn) the build-up factor of N layers of slabs, the thick-nes s of the i-th slab being MiXJ, and В м ( ^ х х + — i s the build-up factor of the N-th slab. The build-up factor of N layers can be obtained by i t er -atively applying the resul ts for a single layer. Depending on the materials , a must be determined empirical ly.

TSURUO [31] has proposed a kind of matrix method for t ransmiss ion in multiple layers . The differential energy spectra of transmitted gamma ray from a slab of single material of thickness t are

I°(E)=I(E) exp (-n(E)t) E max

l\(E)=J I^(E')BE (E1, t)k(E', t) (//(E'-> E, t)dE' (4) E

1г(Е) = l f (E) + lf(E),"

where ID = unscattered gamma ray in the slab; Is = scattered gamma ray in the slab;

m(E) = attenuation coefficient; BE(E,,t) = energy build-up factor in infinite medium;

k(E, t) = Berger ' s finite slab effect coefficient; ^(E'-»E, t) = probability density of scattering in which the energy

of a photon changes from E' to E; and

J i ¡ j ( E ' - E , t ) d E = 1. о

If scattered gamma-rays are assumed to travel vertically from the slab af ter t r a n s m i s s i o n , Eq. (4) can be applied to mult iple l a y e r s . It i s con-

49

venient to e x p r e s s Eq. (4) in the form of a matrix. The incident differential energy flux i s e x p r e s s e d by a v e c t o r and the sh ie ld ing p r o p e r t i e s of each slab of the mult iple l a y e r are e x p r e s s e d by a s e r i e s of m a t r i c e s . T h e s e m a t r i c e s can be cons truc ted by a re la t ive ly s i m p l e procedure , us ing the data of penetration in an infinite medium, calculated by the moments method, and the data of the e f f e c t of a f in i te s lab, c a l c u l a t e d by the M o n t e - C a r l o method. The resu l t s of the method are compared with those of the Monte-Carlo calculat ion for Pb- H2O and НгО-РЬ strat i f ied s labs , and a l so with those of experiments for Pb, Fe and Al stratif ied s labs . The comparison shows that this method i s useful for all combinations of materials at high-incident energy and when low-energy g a m m a - r a y s enter the heavy layer followed by a light or heavy layer .

Some tra i l s were made on duct streaming. Kitazume derived a formula extending Simon-Cli f ford's to two-group theory. If the length of a duct L i s sufficiently long compared with its diameter Rn , the fast and thermal neutron fluxes are respect ively expressed by:

¿ r ^ U o b + ß i i + l t a n 1 Ж 21/ r i u ^ L V5IV

2LF Ф20 j1 + , M . , - U + Га. + ( M ) ß J f i + S i t a n - i ( 5 )

4^20/ L 5 R 0 j / '

where <j>i = fast neutron flux at the inlet of the duct; ф0 = thermal neutron flux at the inlet of the duct; Фх and ф2 are isotropic; ß-L = albedo of fast neutron; ß 2 = albedo of thermal neutron; ß 1 2 - albedo of fast neutron slowing down to thermal neutron.

Shindo et al. have systematical ly studied duct streaming. The radiation f luxes at the exit w e r e divided into that f r o m the in let d isc and that f r o m the annular part surrounding it. Further, unscattered, single scattering and mult i scat ter ing radiations w e r e separate ly cons idered and the contribution of each of them invest igated.

A s imple method was derived by Tagami and Kitazume to es t imate the e f f ec t s of g a m m a - r a y scattering in the infinite air medium around the local shielding. The mapping of gamma radiation dosage around local shielding can be readily obtained by this method, which a s s u m e s the point source and single scattering of gamma-ray in air but involves the correct ions for multi-sca t ter ing e f f e c t s by ut i l i z ing the bui ld-up fac tor . Compared with the measurements of gamma-ray distribution around the Ozenji Critical Facility, this method g i v e s a reasonable representat ion of the a ir scat ter ing of gamma r a y s .

These workers have a lso obtained express ions for the primary neutron and g a m m a - r a y attenuations in water of a swimming pool reactor by uti l izing a single removal-group theory in connection with a two-group diffusion equation for neutron attenuation and an e ight-group energy source for g a m m a - r a y s , and using a modif ied point source approximation for the reactor core . The

50

e x p r e s s i o n for the photoneutron distribution in water , generated by photo-reaction with deuterium contained in natural water, i s also obtained by intro-ducing the source t erm of photoneutron into the one-group neutron diffusion equation. Utilizing the above express ions for neutrjon flux distributions, the expression's for the captured g a m m a - r a y distribution generated by thermal neutron capture and thermal photoneutron capture are obtained. The calcu-lated neutron and g a m m a - r a y d is tr ibut ions in the pool w a t e r a r e iri good a g r e e m e n t with the exper imenta l r e su l t s f rom the HTR at Hitachi and the BSR at ORNL, to an a c c u r a c y of approximate ly 30% over al l pos i t ions in s e v e n m e t r e s of water .

3. 3. Optimization

Tanaka et a l . have inves t igated m a r i t i m e r e a c t o r shie ld opt imizat ion with r e s p e c t to weight . A formula on the weight of the m a r i t i m e r e a c t o r shield, cons i s t ing of the weight of the shield of the reactor i tse l f and auxil-l iary equipment, was derived with some assumpt ions . This i s a function of the d imens ions of the reactor , power, temperature , p r e s s u r e , enrichment of uranium e t c . , and the changes in the weight of the shie ld w e r e i n v e s t i -gated with changing p a r a m e t e r s . As the weight of the secondary shie ld i s important , i t s e f fec t on the pr imary shie ld was widely inves t igated .

Some tr ia l s have a l so been made for obtaining optimum condit ions for shie lding f a c i l i t i e s ut i l iz ing radiat ions outside the reac tor . SAKUTA [32] has investigated the relation between the position of the target and the volume of c o n c r e t e for the sh ie ld of a l inear a c c e l e r a t o r fac i l i ty where the target room i s above ground, part ly underground or wholly underground. Initial and running c o s t s of a ir conditioning was invest igated a l so .

Okajima has sy s t emat i ca l l y invest igated the optimum shielding for ac-ce lerators , taking economics and constructional methods into consideration.

4 . COMPUTER CODES

The development of computer codes for shielding calculations in Japan was s tarted in 1958 us ing a re lay type computer , FACOM 128. The use of s m a l l digital computers fo l lowed immediate ly , e . g . the IBM 650 insta l led at JAERI and the USSC-90, NEAC 2203 and Bendix-G15D.

In 1959 an act ive group of r e s e a r c h e r s and e n g i n e e r s f r o m r e s e a r c h inst i tutes , un ivers i t i e s and private organizat ions throughout Japan was e s -tablished with the object of mutual collaboration in the development of com-puter c o d e s for shie lding ca lcu la t ions . This group, the Shielding Codes Group (SCG), has completed th ir ty- f ive computer codes relat ing to a l m o s t all kinds of shielding design calculations [33] inc luding the following: neutron sh ie ld des ign; p r i m a r y g a m m a sh ie ld des ign; s e c o n d a r y g a m m a s h i e l d design; est imation of coolant activity source intensity; calculation of f i ss ion-product s o u r c e intensi ty; evaluat ion of g a m m a - r a y s s c a t t e r e d by a i r and walls; est imation of temperature distribution and thermal s t r e s s in a shield; evaluat ion of g a m m a - r a y dose outs ide a r e a c t o r conta iner in the event of f i s s ion-product r e l e a s e incident; and fundamental survey codes for compu-tation of gamma-ray streaming through a cylindrical straight duct. In addition

51

to SCG act iv i t i es , JAERI, the Ship R e s e a r c h Institute, the Universi ty of Tokyo and s e v e r a l private organizat ions have a l so been working on the de-ve lopment of shie ld computer c o d e s and have produced twenty-nine c o d e s between 1959 and 1963.

M e d i u m - s c a l e computers , such a s the IBM 7070, FACOM 222 and NEAC 2206, and three l a r g e - s c a l e computers , al l IBM 7090, have become available in recent y e a r s for shielding r e s e a r c h e r s and engineers . Among others , a NEAC 2206 computer was instal led in the Ship Research Institute in 1962 espec ia l ly to develop marine reactor shielding evaluation techniques and for m a r i n e r e a c t o r hazard a n a l y s i s . An IBM 7044 c a m e into u s e at JAERI in 1964 f o r g e n e r a l r e s e a r c h work.

Because so many codes have been produced, it i s not possible to explain them all here . Among the work done by SCG [33], the SCG-RAC code de-ve loped by Abe et a l . for the IBM 7090 i s one of the mos t d is t inguished shielding design codes in Japan. It can handle the following four categor ies of shie ld ing ca lculat ions for cy l indr ica l s o u r c e g e o m e t r y with cy l indr ica l she l l shie lds in a s ingle computer run in any sequence desired: REMOVAL computes energy-dependent removal -group neutron distributions in a shield up to eighteen energy groups; DIFFUSION i s a neutron diffusion code with up to s ix energy groups; PRY. GAMMA calculates the gamma-ray f luxes in energy groups up to ten, using a single build-up factor of each energy group def ined throughout the whole shie ld; SEC. GAMMA eva lua te s s e c o n d a r y gamma-ray flux distributions in the same manner as PRY. GAMMA but with source distribution determined by the thermal neutron flux of DIFFUSION. The SCG-RAC code a l so prepares the input data to the SCG-GH-N code by Fukugaki for the IBM 7090 which c o m p u t e s t e m p e r a t u r e dis tr ibut ion and t h e r m a l s t r e s s e s in the sh ie ld .

To e s t imate g a m m a - r a y source intensity by f i s s i o n products accumu-lated in the core , a s e r i e s of codes w e r e prepared for the IBM 650 and FACOM 222. They are SCG-FP-G, S C G - F P - B , SCG-FP-I and SCG-FP-B and I by Kikuchi et al . SCG-CA-P, SCG-CA-I and SCG-CA-F by Mori and Nakano for the NEAC 2203 evaluate the coolant activities caused respectively by the induced activity of the coolant i t s e l f , by the induced activity of i m -pur i t i e s in the coolant and by the radiation of f i s s i o n products which have leaked into the coolant from fuel e lements . Kitazume has developed SCG-SCA-1 and SCG-SCA-2 for the IBM 7090 which calculate the gamma-rays scattered from both large and s m a l l wal l s compared to the s i z e of the incident beam width.

Taking into consideration details of ship structure, tanks, reactor plant container, plant components and shielding structures, a code was completed for the NEAC 2206 computer by Kataoka et al . The main f ea tures of the code, MARINE-1, a r e a s fo l l ows . F i r s t , the radiation s o u r c e s can be treated as a sum of volume e lementary sources and/or surface e lementary sources up to two hundred. The angular distribution of source intensity can be e x p r e s s e d by the power s e r i e s of the cos ine of the angle to the direction which i s defined arbi trar i ly at each source point. If des i red , the angular distr ibution of each s o u r c e intens i ty can be l imi ted within the half space , i . e . no photon radiates in the half space chosen arbitrari ly. The secondary build-up factors of g a m m a - r a y s obliquely incident to finite s labs are evalu-ated with the response matrix method described in sect ion 3.

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B e s i d e s shielding des ign codes , there are many other codes prepared with a view to research studies as well as to design work. The EOS code, a numerical integration code of the transport equation for slab geometry, has been developed and i s de scribed in sect ion 3. There are s e v e r a l Monte-Carlo treatments of gamma-ray calculations. In the slab geometry, bothKAZAK-II by Tsuruo for the IBM 7090 and AMONTF by Nakano for the IBM 7090 treat gamma-ray multi layer problems. MC-MATRIX code by KATAOKA [24] for the NEAC 2206 has been applied to prepare the m a t r i c e s of each mater ia l for the r e s p o n s e m a t r i x method. NKM-1 and NKM-2 by U m e d a for the IBM 7070 compute g a m m a - r a y behaviours in the spher ica l and cyl indrical geometry shield.

MULTI-LAYER SYNTHESIS 2 by KATAOKA [24] for NEAC 2206 i s a code to calculate the g a m m a - r a y spectrum transmit ted and re f lec ted from mult i layer sh ie lds by means of the response matr ix method.

There are a s e r i e s of codes which evaluate g a m m a - r a y s treaming through straight cyl indrical ducts. Among them, SCG-LGE1 by Tsuruo for the IBM 7090 ca lculates the unscattered and singly scattered flux by means of a numer ica l integrat ion. SCG-SC-M1 by Shimamura for the IBM 7090 evaluates g a m m a - r a y flux and current us ing the Monte -Car lo method for the s a m e c a s e .

5. FACILITIES FOR SHIELDING EXPERIMENTS IN JAPAN

The Research Reactor -4 (JRR-4) at the Japan Atomic Energy Research Institute i s now under construction and will become cri t ical in October 1964. It i s an enriched-uranium fuelled, water moderated and cooled, swimming pool type reactor with a normal operating power of 1000 kW and a maximum power of 3000 kW. Figure 32 shows the vert ical c r o s s - s e c t i o n of the JRR-4 reactor .

5. 1. The pool and its facilities

5 . 1 . 1 . Pool

The pool i s separated by a gate into two parts, No. 1 pool and No. 2 pool. The f o r m e r i s for fundamental exper iments with re lat ive ly s m a l l s a m p l e s . The latter i s for ac tua l - s i z e experiments with large samples . No. 1 pool i s of r a i s e d type, 7 m X 7 m X 1 0 m deep, the water depth being 9.8 m. Al l ins ide f a c e s of the pool are lined with aluminium. There are a beam hole, a thermal column and an air rabbit tube in the three directions of the raised part . No. 2 pool i s 7 m X 9 m X 10.3 m deep, the water depth being 9.8 m . It i s a l so l ined with aluminium and has a we l l 3 m X 4 m X l m deep in the centre to provide a space free from the effects of wall scattering and induced act ivi ty , and a l s o a l l o w s for the adjustment of the s a m p l e height to a l ign the s a m p l e with the c o r e cen tre l i n e s .

The reactor i s operated at high power (above 200 kW) with core placed at the centre of the r a i s e d part of No. 1 pool and at a d is tance of 3 m f rom the c e n t r e of the gate in No. 2 pool . It i s operated at low p o w e r near the high power pos i t ion .

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Reactor room 8. 1 5 / 5 - 1 crane Dry shielding test room 9. Beam hole Measuring room 10. Shielding door N o . l pool 11. Gate N o . 2 pool 12. Instrument bridge Reactor bridge (dry shielding test room) Instrument bridge (pool) 13. 5-1 crane

Fig. 32

Ver t ica l section of JRR-4 reactor

5. 1 . 2 . Instrument br idge

The measur ing ins truments are hung f rom the instrument bridge that spans the pools in para l l e l to the r e a c t o r br idge . Two c a r t s t r a v e l inde-pendently on a pair of ra i l s on the bridge. A shaft, suspended from the cart, c a r r i e s at i t s end the measur ing instruments which can be posit ioned at any point in the pool. The instrument bridge i s operated from the control panel located at the corner of the bridge. The position of the instruments i s shown by synchros to an accuracy of 1 mm.

5. 2. Lid tank facility

5 . 2 . 1 . Lid tank

The lid tank i s 4 m X 4.5 m X 6.5 m deep, the water depth being 6.2 m. It i s a l s o l ined with a lumin ium and spanned by an i n s t r u m e n t b r i d g e .

5 . 2 . 2 . T h e r m a l c o l u m n

The thermal column i s placed between the lid tank and the reactor on the ra i sed part of N o . l pool. It i s 4.36 m long. The c r o s s - s e c t i o n s at the reactor s ide and the lid tank s ide are 50 cm X 50 cm and 1.2 m X 1.2 m re -s p e c t i v e l y . The t h e r m a l f lux i s expected to be above 108n cm"2 s"1 at the outer end of the co lumn.

5 . 2 . 3 . Conver ter

The c o n v e r t e r i s a d i sc , 60 c m in d iameter , of uranium enr iched to 20% in U 2 3 5 (3 kg of U 2 3 5 ) . The power i s expected to be about 10 W, though this wil l depend on the thermal flux from the thermal column.

5. 3. Dry shielding test facility

The dry sh ie ld ing t e s t fac i l i ty i s f o r sh i e ld ing e x p e r i m e n t s in a i r .

5 . 3 . 1 . Dry sh ie ld ing t e s t r o o m

The dry shielding test room is 14 m X 14.5 m X 9.4 m high. It i s a semi-underground re in forced building, adjacent to the r e a c t o r r o o m and has a beam hole, a gamma-ray source apparatus and an instrument bridge. When exper iments are carr ied out, personnel are not allowed in the room and so the m e a s u r i n g i n s t r u m e n t s must be contro l l ed f r o m the m e a s u r i n g r o o m adjacent to the test room and monitoring i s by c lo sed -c i rcu i t t e l ev i s ion and through a shielding window.

5 . 3 . 2 . Beam hole

The beam hole pene tra tes the wall , 3 m thick, between No. 1 pool and the dry shielding test room. The c r o s s - s e c t i o n of the beam hole on the pool s ide and the room sidë are 30 cm X 30 cm and 60 cm X 70 cm, respect ive ly .

55

A he l ium tank i s pos i t ioned over the p o o l - s i d e end of the b e a m hole and a shie lding door i s over the r o o m - s i d e end. The shielding door has a beam hole. The strength and c h a r a c t e r i s t i c s of radiation are changed by moving the hel ium tank up or down and insert ing f i l t e r s into the beam hole .

5 . 3 . 3 . G a m m a - r a y source apparatus

The g a m m a - r a y source apparatus i s composed of a device to insert and r e m o v e radioisotope capsu les , a s tore room of 12 radio i sotopes (maximum 50 с each), an activation device, a gamma-ray irradiation device and rabbit tubes connect ing t h e m . The g a m m a - r a y i r r a d i a t i o n d e v i c e i s in the dry shielding tes t room and the activation device i s near the core (thermal flux i s about 1010 n c m - 2 s"1; the o thers are in the r e a c t o r r o o m . The i rrad i -ation device i s provided with a co l l imator to let the g a m m a - r a y s subtend in the range of 0 - 45° . The centre l ine of the f lux can a l s o be changed up to 20° f rom the hor izonta l .

R E F E R E N C E S

KIRN, F. S . . KENNEDY. J. and WYCOFF, H. 0 . . Radiology 63 (1954) 94. MOCHIZUKI, H . , TANAKA, Y. , HIGASHIHARA, Y. , NAGATO. N. and YORIHISA, K. , J. Atomic Energy Soc. Japan 4^(1962) 448. MOCHIZUKI, H. , TANAKA, Y . , HIGASHIHARA. Y. and NAGATO, H. , J. Atomic Energy Soc. Japan

_4 (1962) 103. MOCHIZUKI, H. and TANAKA, Y . . Nucl. Engng _9 (1963) 54. KALOS, M . H . , Rpt NDA-5607 (1956). SUGURI, S . , FURUTA, Y. and TANAKA, Y. , Rpt. JAERI 5002 (1960). ISHIMATSU, K. , J. Atomic Energy Soc. Japan £(1962) 175. NAKATA, M . . FUSE, T. and TAKEUCHI, K. . J. Transport. Tech. Res. Inst. П_(1961). NAKATA, M . , J. non-destructive Inspection 10 (1961). HYODO, T . , Nucl. Sei. Engng 12 (1962) 178. SAKUTA, M . , Trans. Architect. Inst. Japan 8£(1963). ISHIMATSU, K. , J. Atomic Energy Soc. Japan £(1962) 24. HYODO, T. and MAKINO, F . , Memoirs Faculty Engng Kyoto Univ. 24_ 2 (1962). MAIENSCHEIN, F. С . , Rpt ORNL-1142 (1952). FUSE, T . , Rpt JAERI 1015 (1961). OWEN, R.B., Rpt AERE-EL/R-2712 (1958). ISHIKAWA, H. , ASAOKA, T. and SASAKURA, H. , Rpt JAERI 1002 (1960). TSURUO, A. and SIBUYA, M. , Proc. 3rd. Jap. Conf. Radioisotopes (1959). BERGER, M.J . and DOGGET, J , . J. Res. Nat. Bur. Standards 56 (1956) 89. TOYODA, Y. and KONDO, T . , 1st Nat. Nucl. Congr. (1960). NAKANO, Y . , Jap. Nucl. Ship Res. Assoc. 42(1963). ISHII, T . , SERINE, T. and ONÓ, K. , Codes for Reactor Computations, IAEA, Vienna (1961) 53.

TSURUO, A . , Jap. Nucl. Ship Res. Assoc. 33(1962).

KATAOKA, I . , to be presented at 3rd. UN Int. Conf. PUAE(1964).

KATAOKA. I. and TAKEUCHI, K. . to be presented at Ann. Meeting Atomic Energy Soc. Japan (1964).

COOPER, C. et a l . . Proc. 2nd. UN Int. Conf. PUAE 13 (1958) 21.

HOKKYO, N. and KITAZUME, M. , Rpt AEC-tr-5077 (1960).

ILIFFE, C . E . , J. Inst. Mech. Engrs, Nucl. Energy Symp. (1956).

MOCHIZUKI, H. and TANAKA, Y. , J. Atomic Energy Soc. Japan±( 1962) 448.

BRODER, D. L. et a l . , Atomnaja Energija 12 (1962) 30,

TSURUO, A . , J. Atomic Energy Soc. J a p a n £ 2 (1962).

SAKUTA, M . , MAEDA, H. and YAMAMOTO, M . , Nucl. Engng 8_ 8 (1962) 27.

Rpt Nucl. Ship Res. Assoc. Japan 21 (1961) ; 33(1962); 45(1963).

56

SUMMARY OF SHIELDING WORK IN NORWAY

U. TVETEN

1. INTRODUCTION

Work on shielding in Norway i s concentrated at IFA (Institut for Atom-energi) and it has mos t ly cons i s ted of design of sh ie lds for our current na-tional reac tor projec ts . Some r e s e a r c h has been per formed in connection with the reactor projects , but most ly we try to make use of re su l t s and ex-perience obtained in other countries .

Our shielding group was founded in 1959 in connection with the REDERI-ATOM-Projec t , which was the pre l iminary study for our current ship r e -actor project. Considerable effort was spent on the design of shie lds . The design of a shield for a ship reactor , for which it was decided to use a lami-nated water- iron shield was the f irs t task. The possibility of utilizing boron-containing m a t e r i a l s in this shie ld was invest igated. The next task was to design the shield for a 2-MW r e s e a r c h reactor JEEP II. Here a design cr i -t er ion was to find the c h e a p e s t p o s s i b l e sh ie ld and i t was dec ided to u s e a type of heavy c o n c r e t e .

The main e f for t has been concentrated upon the two t a s k s ment ioned above, but there has a l so been des igned a shie ld for a z e r o - p o w e r reac tor NORA and a shie ld for the boiling heavy-water reac tor Halden, as we l l as gamma shie lds for isotope production plants and uranium purification plants both in Norway and abroad.

At present work i s concentrated upon the des ign of the shie ld for our new ship reactor project . We st i l l have bas ica l ly a w a t e r - i r o n shie ld, but some regions may contain lead and we are now trying theoretical ly to decide upon the m o s t advantageous posi t ions for these . Another paral le l aproach i s the usé of boron water between s t e e l s l abs . Both these types of sh ie lds are opt imal ized using a Ferrant i Mercury computer and the shie lding pro-g r a m m e s LIDO, RASH B3 and GASH B2.

The r e a c t o r J E E P II i s under cons truct ion , and additional sh i e ld ing p r o b l e m s are t rea ted a s they a r i s e .

2. THE SHIELDING P R O P E R T I E S OF IRON AND BORON PLASTIC [1]

T h e s e exper iment s w e r e per formed in the sw imming pool reac tor SAPHIR at Reactor A . G . Würenl ingen in Swi tzer land in 1959, and w e r e performed in connection with the REDERIATOM-Project . In the f i r s t part of the exper iment the sh ie ld m e a s u r e d was composed of i ron s l a b s of dif -ferent t h i c k n e s s e s p laced tightly together so as to form one compact s lab. In the second part of the exper iment shee t s of boron plast ic were placed at different posit ions between the s labs . The plast ic contained 400 mg B / c m 2

and the thickness was 1 cm. The shield was fully i m m e r s e d in light water and the inner surface of the shield was 25 cm distant from the core surface.

The c o n c l u s i o n s drawn f r o m the e x p e r i m e n t s w e r e that one s h e e t of boron p las t i c p laced d irec t ly in front of the iron sh i e ld had no e f f ec t upon

57

the total g a m m a dose rate measured 3 c m outside the outer sur face of the shield. The s a m e boron plast ic sheet placed at the outer surface of the shie ld gave a reduct ion of the g a m m a dose rate by a factor of two. More boron plast ic s h e e t s placed between the iron s labs in the outer parts of the shie ld gave further reduct ions . The exper iments w e r e compared to ca lcu-lations performed by a s imple method (three groups) which was used at that t ime . The agreement outside the slab was rather good, but ins ide the s lab the e x p e r i m e n t a l r e s u l t s for the t h e r m a l f lux w e r e m u c h higher than the theore t i ca l . This was probably due to s m a l l water gaps between the i r o n s l a b s .

3. SHIELDING PROPERTIES OF MAGNETITE HEAVY CONCRETE

T h e s e e x p e r i m e n t s w e r e per formed by K. B e r g e as a t h e s i s work at the R 2 - 0 reac tor belonging to AB Atomenergi , Sweden in 1960. R 2 - 0 i s a s w i m m i n g pool r e a c t o r , nominal power 100 kW. The concre te s lab to be invest igated was placed outside a 2 m X 2 m window in the shield separated f r o m the r e a c t o r c o r e by 15 c m light water and the window c o n s i s t i n g of 2 c m a luminium and 1 c m P l e x i g l a s s .

The heavy concre te was chosen so as to correspond to the heavy con-crete meant to be used in the shield of our reactor JEEP II and the exper i -ment was performed in connection with this project. An analysis of the con-cre t e i s g iven in Table II. Th ickness of the s l ab was 90 c m and den-s i ty 3.74 g / c m a .

TABLE и

HEAVY CONCRETE, E L E M E N T A L COMPOSITION

Element Concre te ( g / c m s )

0 1 .073

Fe 2. 046

Mn 0. 007

Ca 0 .246

Mg 0. 049

Al 0. 041

Si 0. 168

Ti 0. 004

V 0. 004

P 0. 029

н2о 0. 073

Tota l 3. 740

58

The fluxes m e a s u r e d w e r e af terwards compared to theore t i ca l va lues computed on a F e r r a n t i computer us ing the p r o g r a m m e s LIDO, RASH B , and GASH B. The r e s u l t s are g iven in F i g s . 3 3 - 3 6 .

=

= j

ч

•

10 20 30 40 SO GO 70 80 90 100 110 120 130 U 0 150

DISTANCE FROM CORE/REFLECTOR BOUNDARY С c m )

Fig. 33

Fast neutron flux: group 1 of RASH В (2 - 0 . 3 MeV)

computed f lux-ф-measured flux

4. NEUTRON AGE MEASUREMENTS IN CONCRETE [2]

Some m e a s u r e m e n t s of neutron age in concre te have been p e r f o r m e d at Oslo Univers i ty us ing a neutron generator . The age m e a s u r e d was the age from 2.5 MeV ((D-D)-neutron production) to 1.44 eV (indium resonance). The distribution of 1.44-eV neutrons was determined using cadmium-covered indium fo i l s . The concrete density was approximately 2.05 g / c m 3 . There was s o m e di f f icul ty in deciding the hydrogen content and it i s only known that it w a s s o m e w h e r e between 0 .010 and 0.007 g / c m 3 at the t i m e of the exper iment .

The m e a s u r e d F e r m i age was 444 c m 2 and ca lcu la ted v a l u e s for the two border values of hydrogen content were 224 cm 2 and 297 cm 2 respect ive-ly.' For the resul t s of the measurements s e e Fig. 37.

5. SHIELDING COMPUTER PROGRAMMES

Mostly we have depended upon computer programmes developed at other inst i tutes , but in connection with our shie ld des igns we have developed two

59

DISTANCE FROM CORE/REFLECTOR BOUNDARY ( c m )

Fig. 34

Epithermal neutron flux: group 5 of RASH В (100 - 0 .07 eV)

computed flux -f*- measured flux

•ISTANCE FROM CORE/REFLECTOR BOUNDARY С c m )

Fig. 35

Thermal neutron flux: group l> of RASH В

computed flux-ф- measured flux

60

L Г L Г

«fas "-••'if i -rb TO 20 30 40 SO 60 70 вО 90 100 ПО 120 130 UO ISO

DISTANCE FROM CORE/REFLECTOR BOUNDARY ( c m )

Fig.3G

Comparison between exper imenta l ly and theoret ical ly obtained g a m m a dose rates

computed flux ф measured flux

OISTANCE, r Ccm)

Fig.37

Results of neutron age measurements

61

programmes for calculation of the gamma flux from a homogeneous cylinder. This allows a number of shields to be placed between the source and the dose point and computes the fluxes throughout the shields.

In connection with the design of component sh ie lds for our present ship r e a c t o r , a p r o g r a m m e for calculat ion of g a m m a fluxes f rom a number of arb i trar i ly p laced s o u r c e s i s under deve lopment . Th i s p r o g r a m m e wi l l allow a number of plane shie lds and cyl indrical shie lds to be placed between and around the s o u r c e s .

6. WATER TANK OF REACTOR JEEP II

The r e a c t o r J E E P II, a 2-MW r e s e a r c h r e a c t o r , which i s now under construct ion at K j e l l e r , a l so includes a water tank which wi l l be used for bulk shielding s tudies . The water tank i s separated f rom the reac tor tank by an intertank which may be empty or f i l l ed with heavy or light water . A lead shutter may a l so be lowered into the intertank.

The f i r s t s e t of e x p e r i m e n t s to be p e r f o r m e d w i l l probably be an in-ves t iga t ion of the ship r e a c t o r shie ld .

JEEP II i s scheduled for completion in June 1965. For dimensions s e e Fig. 38.

CORE H 2 0 OP

AIR

52.5

v. • .

.' - « ' ?

H20 or AIR

- _\j 70 cm

400 cm ж 250 cm

WATER TANK H 2 0

60 m '

400 cm

HEAVY CONCRETE

ч - .

' ' 0 ' . •

. 1 . * 0 : p . '

\ 30 cm /

Fig. 38

Diagram of water tank, inter tank and lead shutter

R E F E R E N C E S [1] REDERIATOM Rpt 15. [2] GRIMELAND, B. and D0NVOLD, S . , Age to 1.44 eV for (D, D) neutrons in concrete, to be published.

62

A SUMMARY OF SHIELDING RESEARCH IN SWEDEN

E. AALTO AND J. BRAUN*

1. INTRODUCTION

In a smal l country like Sweden the yearly expenditure on nuclear energy is smal ler than that of the biggest companies in big countries. The shielding work is therefore necessar i ly concentrated on practical or applied shielding. Thus, shielding research has been directed towards some gaps in knowledge, the fi l l ing of which was felt to be of the greatest economic benefit in the near future.

In this paper a description of the Swedish studies is presented. At some points a s u m m a r y of unpublished r e s u l t s i s given. D e t a i l s are omit ted , and we r e f e r to c o m p l e t e r e p o r t s p r e s e n t i n g the work introduced h e r e .

As the main part of our r e s o u r c e s has been directed toward the design of shie lds for the national power reactor programme, the publication of r e -ports has been delayed. We hope to improve on this, and our re ference l i s t contains r e f e r e n c e s to reports to be issued during this year (1964).

The shielding r e s e a r c h for civi l applications i s concentrated in the Swedish Atomic Energy Company (AE) which i s partly owned by the govern-ment. The mi l i tary r e s e a r c h i s done by the Research Institute for National Defence (FOA). When the nucleus of our present shie ld ing group was s e t up in 1956 at AE, it was felt that a thorough understanding of the penetration of radiation through laminated m a s s i v e shields in s imple geometr ies should be the backbone of all subsequent shielding r e s e a r c h . After this , one may try to attack m o r e compl i ca ted p r o b l e m s and g e o m e t r i e s . We have thus started with a thorough experimental and theoretical study of m a s s i v e lami-nated shie lds .

One of the most difficult problems in reactor shield design is to estimate the contribution to the dose f r o m var ious kinds and shapes of ducts p e n e -trating the shield. Therefore , the study of 'vo ids and ducts' was the next logical step. Our r e s e a r c h in this area has of neces s i ty touched only those problems which are current in the Swedish DzO reactor programme.

In our exper ience a very c lose co-operat ion between development (both exper imenta l and theoret ical ) and the shield des igner i s paramount for ob-taining good r e s u l t s and for d i rec t ing the r e s e a r c h in frui t ful d i r e c t i o n s . B e s i d e s , f r o m the u s e r ' s point of v iew, a crude method with known e r r o r l i m i t s i s bet ter than a t h e o r e t i c a l l y comple te m o d e l untes ted in p r a c t i c e .

In the following, the r e s e a r c h fac i l i t i e s wil l f i r s t be introduced. Next c o m e s the s u m m a r y of the current f in i shed work in bu lk - sh i e ld ing and of the studies of i rregular i t ies . Finally, some future plans and problems will be d i scussed .

2. THE EXPERIMENTAL FACILITIES

The main exper imenta l f a c i l i t e s are s ituated at the Studsvik r e s e a r c h centre of the Swedish Atomic Energy Company (for exceptions s e e sec t ions

* Presented by J. Braun.

63

3. 1 .4 . and 3. 1. 5). The radiation source i s the R2-0 reactor, 100 kW (max) with natural c ircu lat ion cool ing. P lans are be ing made to run it at 1 MW with forced circulation. These have, however, not been initiated for reasons having to do with shielding research.

The shie lding f a c i l i t i e s avai lable are p r e s e n t e d in F ig . 39. The m o s t important faci l i ty used has been the ~2 m X2 m window N1. The smal l win-dows ( 0 . 5 m X 0 . 5 m) are usab le for penetrat ion m e a s u r e m e n t s up t o ~ 5 0 - 7 0 cm depth in m a s s i v e mater ia l s . At greater depths, the s t reaming in the s lo t s around the experimental plug begins to dominate.

Fig. 39

The R2-0 shielding fac i l i t i es , p lane view

The reac tor i s movable on a tro l ley along the two horizontal axes and can be turned toward any side of the pool wall by a simple lift with the over-head crane. This operation takes approximately 15 minutes in practice. The c l o s e s t distance between the reactor face and the pool wall i s 10 cm which avoids the poss ib i l i ty of unexpected changes in react iv i ty . The advantages of the R2-0 N l - fac i l i t y are: (1) Dry faci l i ty; (2) Penetrat ion depths up to ~ 2 m can be studied; (3) Exper imenta l s e t - u p s up to the total weight of ~ 4 0 t are moved by the

driving machinery of the normal shie lding plug; sind (4) Set -ups are a c c e s s i b l e through a s y s t e m of tubes (Fig. 39) during e x -

posure; The greatest disadvantage i s that the stiffening structure of the window con-s i s t s of two I - b e a m s , and thus g i v e s an unavoidable a ir gap about 30 c m thick. Because none of the one-dimensional shielding codes can accurately take care of the leakage in this kind of a s lot , it pre sen t s s o m e di f f icul t ies in the evaluation of the experimental resu l t s . We may compare this window with the window at the U. К. A. E. A. 1 s pool-type reactor LIDO. There, this diff iculty has been avoided by f i l l ing the window, which i s s t i f fened with a sort of sandwich construct ion, with water .

64

The neutron detection i s based on foi l techniques. The counting of fo i l s and the subsequent data p r o c e s s i n g has been c o m p l e t e l y automated. The principle has been to keep the e l ec tron ics stat ionary and out of the reactor hall .

The fo i l s are counted in a sample changer which punches the data di-rect ly on tape. This tape i s fed without modif icat ion into the computer (Ferrant i -Mercury) and reduced into flux va lues .

Gamma m e a s u r e m e n t s are based on ordinary ionizat ion chamber and f i l m techniques .

3. BULK-SHIELDING STUDIES

3. 1. Experimental work

3 . 1 . 1 . His tory

B e f o r e 1958 very l i t t le published data w e r e available on the re l iabi l i ty of the ex i s t ing des ign methods for b io log ica l sh i e lds . Cons ider ing the f i -nancial gain obtainable f r o m better and m o r e accurate des ign methods , a study of the bas i c , m a s s i v e b io log ica l sh ie lds was initiated.

The e x p e r i m e n t a l work got s tar ted during the fa l l of 1960, a f t er the R2-0 reactor with its fac i l i t i e s was completed. Prel iminary resul t s , com-pared to the Engl i sh mul t igroup-method [2], whiçh w a s new at that t i m e , were presented at the EAES Symposium on Nuclear Reactor Shielding Theory and E x p e r i m e n t s [3]. Af t er this report the e x p e r i m e n t s went on until the s u m m e r 1962. The bulk sh ie ld ing s tudies w e r e t e r m i n a t e d by a study of laminatèd sh ie lds of Fe and DzO and included a compar i son of the r e s u l t s with calculations. This work was completed during the fall of 1962 and wil l be reported e l s e w h e r e [4]. At that t ime it was cons idered that the amount of exper imenta l data was enough' for evaluation of the re l iab i l i ty of the exist ing design methods and the main effort of the smal l experimental group was d i rec ted towards the study of vo ids and ducts ( sec t ion 4 . 2 ) . A study of neutron s t r e a m i n g in D 2 0 pipe's was the f i r s t problem of this type to be completed ( sect ion 4. 2. 1).

3. 1. 2. Some notes on the s tudies of m a s s i v e concre te sh ie lds (AE)

The economic point of view normally determines the material to be used in b io log ica l s h i e l d s . The cheapest m a t e r i a l i s ordinary c o n c r e t e f r o m loca l m a t e r i a l s . If there i s a lack of space , however , s o m e h e a v i e r m a -t e r i a l i s used and in s p e c i a l c a s e s s o m e expens ive and unusual m a t e r i a l s have been used. The trend c l ear ly goes toward the use of the mos t econo-m i c a l mater ia l . That i s the type of heavy mater ia l which can be obtained loca l ly .

In ant ic ipat ion of th i s , the m a i n m a t e r i a l s s tudied in our work w e r e ordinary and magnet i te c o n c r e t e s of local m a t e r i a l s . The conf igurat ions studied are presented in F ig . 40.

In order to invest igate the accuracy and re l iabi l i ty of the method d e s -cr ibed in s e c t i o n 3 . 2 . 1 and s o m e other m e t h o d s , c a l c u l a t i o n s have b e e n

65

- R E A C T O R F A C E SO 100 150

< c m > -200 250

2 r A l 2 c m H A J C I T E 1 c m

П ¡ J

A l 1 c m

i 1 L

20

I 5 0

Fig. 40

Configurations for the bulk shielding study

3 water

МП и h \ ordinary concre te IV V VI magne t i t e concre te

I i I I I barytes concre te I I air

• lead (5 cm) + borated Lucite

performed in the s a m e configurations using the best available input values . Calculations have been performed a l so with:

(a) An ear l i er 18-group removal method combined with a build-up fac-tor treatment of gamma transport [2];

(b) With a method calculating neutron f luxes by numerica l integration of the Boltzmann equation (NIOBE); and

(c) With a neutron Monte-Carlo code (sect ion 3. 2. 3). All of the methods give rather good agreement with the measured values,

n o r m a l l y within a fac tor of 2 to 3 af ter ~ 2 - m penetrat ion. A typ ica l s e t of r e s u l t s i s g iven in F ig . 41 .

An e x a m p l e of the neutron s p e c t r u m in c o n c r e t e that w a s ca l cu la ted using our new code ( 3 . 2 . 1 ) i s presented in Fig. 42. We have also made m e a s u r e m e n t s with thresho ld de tec tors , to compare the ca lculated and m e a s u r e d fas t s p e c t r u m s [5].

In the comparison of the measured and calculated attenuation, we have made as accurate an e r r o r analys i s as poss ib le . To elucidate the i m p o r -tance of a s ingle i solated factor, Fig. 43 gives the dependence of the resul t s on the variations in the density of the concrete.

66

F i g . «

Measured and ca lcu la ted the rma l neutron fluxes ( re la t ive) for configurat ion 6

^ = Oth measured /0 t j ] by RASH В X = 0 t l l by NRN/Oth by RASH В ф = measured /0 t [ j by NRN

П П П •J UL

LI PI

Iff* 10"5 10"' 10"3 10"2 10"' 10° 101

E (MeV) Fig. 42

Neutron spectrum by the NRN method af ter penetrat ing 20 cm H 2 0 and 35 c m magne t i t e concre te

,67

Fig. 43

Relative neutron at tenuat ion as a funct ion of the concre te properties

Regarding the sa fe ty f a c t o r s to be used when des igning m a s s i v e b io -logical shie lds , we draw the conclusion that the accuracy of the best avail-able methods i s so good that the uncertainty originating from other factors (e. g. radiat ion s o u r c e s or c o n c r e t e density) produces e r r o r s at l e a s t of the s a m e order of magnitude as the ca lculat ional methods .

T h e s e bulk sh ie ld ing s tud ies w i l l be c o m p l e t e l y repor ted in [1], and a s u m m a r y wi l l be p r e s e n t e d at the Geneva c o n f e r e n c e in 1964.

3 . 1 . 3 . Studies of neutron penetrat ion in laminated F e - D ? 0 (AE) [4]

The t h e r m a l sh i e lds and the v e s s e l wal l are c r i t i c a l points f r o m the point of view of the shield designer and material technologist. Both the rate of heating and radiation damage f l u x e s must be predicted with rather high accuracy. This applies even more to the D 2 0-reactors , where, for reasons of cost , the re f l e c tor i s kept at minimum thickness , i. e. ~ 3 0 cm. On the other hand there i s a region where mult igroup diffusion theory has s e v e r e l imi ta t ions . Thin r e g i o n s of the v a s t l y d i f ferent m a t e r i a l s , DgO and Fe , alternate rapidly, violating the r u l e s for the proper use of diffusion theory which require the absence of strong flux gradients as well as distance from sources and boundaries.

In order to study the re l iab i l i ty of the ex i s t ing des ign methods in th i s c a s e , a s p e c i a l s e t - u p w a s cons truc ted for the e x p e r i m e n t a l window N 1. Using a U - f r a m e and lock (Fig. 44), the e f fect ive c r o s s - s e c t i o n of the win-dow was d e c r e a s e d to 1.5 m X 1 . 5 m. In this f r a m e l a y e r s of Fe and DzO (in aluminium tanks) can be arranged in varying configurat ions . Thus we had the advantages of a dry fac i l i ty .

68

Fig .44

Set -up for the study of l amina ted F e - D 2 0 shields

Measurements have been performed in var ious configurations of m a s -s i v e l a y e r s and in l a y e r s penetrated by (1) continous channels of D 2 0 and (2) continous channels of air. The r e s u l t s are compared to those obtained by various calculating methods and the agreement i s surpris ingly good. An example of the r e s u l t s for the sulphur-act ivat ion r a t e s i s given in Fig. 45.

3. 1 . 4 . Exper imenta l fa s t neutron spec tra in Al and Fe (FOA)

In this study [6] the spectrum of fast neutrons produced by D-D (3 MeV) or D - T (14 MeV) r e a c t i o n s w a s studied in and behind s h i e l d s of d i f ferent t h i c k n e s s e s of F e ( 5 - 2 0 cm) and of Al (20 cm) . The ins trument u s e d for de tec t ion was a s t i lbene s p e c t r o m e t e r . To obtain in format ion about the neutron s p e c t r u m be low 0.5 MeV, m e a s u r e m e n t s with a Li6I s c i n t i l l a t o r were performed. The experimental results are compared with Monte-Carlo calculations.

3. 1. 5. Model experiments for shielding studies (University of Lund)

The possibi l i ty of using model experiments to study gamma-penetration problems has been studied at the University of Lund. The resu l t s are des -cribed in [7].

69

FROM REACTOR (cm! •

Fig. 45

Measured and ca lcu la ted fast fluxes in l amina ted Fe-DjO

NRN method О observed

3. 2. Theoretical work on bulk shields

3 . 2 . 1 . A bulk shield design method (AE)-

Our mos t important theoret ica l work in this s e c t o r i s the development of a computer method [8] (for Ferrant i -Mercury) for the design of laminated shie lds . The attenuation and slowing-down of neutrons i s treated by a com-bination of f i r s t - f l i gh t removal and mult igroup-dif fusion theor ie s [9]. The energy-dependent flux of such f i s s ion neutrons which are uncollided or which have suffered only smal l -angle co l l i s ions i s calculated throughout the shield; this flux provides source t erms in a s e r i e s of multigroup-diffusion equations. Slowing-down between the diffusion groups i s treated by a d iscrete col l i s ion f o r m a l i s m in which the c o r r e c t average energy l o s s of a neutron in a s p e -cif ic energy group i s taken into account. Inelastic scattering and moderation by l ight nucle i c o m e in natural ly , as s lowing-down be tween all groups i s permi t t ed .

The principal dif ference between this and the well-known English multi-group code [2] i s in the connection between removal group and diffusion part and in the t r a n s f e r be tween di f ferent dif fusion groups . Th i s i s d e s c r i b e d in Fig. 46. The p r o g r a m m e includes an initial part which provides all the requ ired neutron c r o s s - s e c t i o n data. The main advantage of the method is that it requires only a modest amount of computer t ime and capacity, com-pared to other s y s t e m s giving the same quality and accuracy of results , e. g.

70

R A S H В " " R A S H D " " N R N "

Fig. 46

Principles of some removal- mult igroup design methods

Fixed energy boundary Boundary of f ree cho ice

R = Removal groups D = Diffusion groups

the spectrum calculated agrees well with the one calculated by NIOBE. Still, the running t i m e s on IBM 7090 for these two codes would have a rat io ~ 1 : 1 0 0 .

The t ranspor t of g a m m a radiat ion i s ca l cu la ted with a M o n t e - C a r l o scheme employing various reducing techniques, e. g. the exponential t rans -formation, whereby the outward directed motion of photons i s enhanced and the s lope of the source distribution i s modif ied so as to be more favourable for Monte-Carlo treatment. The running t ime on Mercury i s ~ 1 to 3 s per case h is tory , and a typical reactor shield (with one main energy group) can be calculated within ~ 1 h. The main advantage of this code i s the e l i m i -nation of the experimental (or Monte-Carlo calculated) build-up factors , and yet it r e q u i r e s e s s e n t i a l l y no m o r e computing t i m e than do the analyt ica l methods. This part of the shie lding programme has been descr ibed in [10].

3 . 2 . 2 . Gamma-ray backscattering (AE)

Us ing M o n t e - C a r l o methods , e x t e n s i v e s tud ie s of th i s p r o b l e m have been made . The r e s u l t s a r e publ ished in [11].

. 3 . 2 . 3 . The use of Monte-Carlo methods in various shielding problems (FOA)

Studies of the use of Monte-Carlo methods for solving gamma transport problems are described in [12].

The application of Monte-Carlo to neutron transport i s being worked on. Neutron ref lect ion from H, H 2 0, AI, Fe, U238, and from concrete i s studied in [13]. A Monte-Carlo programme for calculating the deep penetration of neutrons in he terogenous s h i e l d s has been deve loped [14]. T h e s e c o d e s are for the IBM 7090 computer .

71

4. STUDIES OF DUCTS AND VOIDS

4. 1. Introduction

One of the most difficult problems in present-day reactor shield design i s how to e s t i m a t e the contribution to the dose on the outside or in s p e c i a l parts of a bulk shield arising from radiation streaming along ducts of various kinds and shapes penetrating the shield. As yet no generally accepted theory exis ts which can be applied to the great variety of problems of this type. However, many s p e c i f i c prob lems have been invest igated m o r e or l e s s s u c c e s s f u l l y .

4. 2. Experimental and theoretical work performed

4. 2. 1. Neutron s t r e a m i n g in D 2 0 pipes (AE) [15]

This i s an investigation of the attenuation of neutrons inside D^O-filled pipes penetrating a concrete shield. Since the purpose has been to simulate the conditions around a heavy-water power reactor , even pipes surrounded by an annular air gap have been considered.

It i s shown that the thermal flux distribution can be predicted theoret i -cal ly by assuming it to be composed of three components, originating from a fast exponential volume source and two surface s o u r c e s at the origin, a 1 /E-d i s tr ibuted source and a thermal source , of which the latter i s proved to be negl igible . The source t e r m s can be calculated with a bulk shielding code. The f a s t - f l u x dis tr ibut ion i s approximated by a s i n g l e exponent ia l e x p r e s s i o n f o r the conf igurat ion with no annulus and with the s u m of two exponent ia l s when an annulus i s p r e s e n t .

The f i r s t part of [15] descr ibes the exper iments carr ied out with pipes of t h r e e d i f ferent d i a m e t e r s (15, 22 and 28 cm) and with annular a ir gaps ranging f r o m 3. 2 to 9. 7 cm in width. Then fo l lows a brief presentat ion of the theoret ica l approach used to predict the thermal f lux distribution and a sec t ion on how to evaluate the different p a r a m e t e r s needed to account for dif ferent pipe g e o m e t r i e s (a "removal" c r o s s - s e c t i o n and an extrapolated radius). Finally the theoret ical calculations are compared with the experi -ments descr ibed in the f i r s t part. F i g u r e 4 7 presents as an example the r e -sul ts for the 1 5 - c m diam. pipe. The thermal f lux components originating f r o m the fa s t and in termedia te f l u x e s are marked by 0[ h and r e s p e c -t ively. It i s a lso shown that the annular air gap can be handled by a s imple c o r r e c t i o n in the attenuation c o e f f i c i e n t for the fa s t s o u r c e . See F ig . 48.

4. 2. 2. The t r a n s m i s s i o n of thermal and fast neutrons in a i r - f i l l ed annular ducts through laminated Fe-D2Û shie lds (AE) [16]

The s e t - u p d e s c r i b e d in s e c t i o n 3 . 1 . 3 and Fig . 44 has b e e n u s e d for these m e a s u r e m e n t s . Iron plugs of various diameters were inserted in the centre hole of the 3X3 grid and supported in order to f o r m annular ducts . Cadmium s h e e t s w e r e used to separate the a ir -gap thermal f lux into c o m -

72

—»DISTANCE FROM FRONT END OF PIPE (cm)

Fig. 47

Neutron s t reaming in DzO pipes Theore t i ca l and exper imenta l thermal and fast flux along axis of 1 5 - c m d i a m . pipe

surrounded by concre te

О exper imen ta l points 0 ^ • exper imen ta l points 0^

theo f epi 0 , = 0 , + 0 , f th th th

ponents suitable for a theoret ical analys is . As shown in Fig. 49 the thermal flux at a f ield point in an annular duct is resolved into three components [17]: (a) Streaming component (S) which c o n s i s t s of t h e r m a l neutrons enter ing

through the annular duct entrance and attenuated geometr i ca l l y ; (b) Albedo component (A) a r i s i n g f r o m neutrons (of al l e n e r g i e s ) which,

having entered the annular duct mouth, travel into the duct wal ls before diffusing back into the duct and reaching the f ie ld point as thermal neutrons; and

(c) Leakage component (L) c o m p o s e d of neutrons (of al l e n e r g i e s ) which penetrate to the duct wall and arr ive at the f ie ld point as thermal neu-trons without having pas sed the annular duct. The leakage i s proved to be proportional to the f lux in a bulk shield of

the same configuration but without duct. The fast flux i s r e so lved into two

73

0.01 r* Ar —

Fig. 48

Neutron scanning in D 2 0 pipes 'Removal ' cross-section as funct ion of a i r -gap width

y / / / / х / / / / / / / / / A / . \ s z z / / / / / / / / / / / / / л y z z z z / / / / z z z z / / \

Fig. 49

Annular duct , flux components

components analogous to (a) and (c) above. However, the Albedo component (b) which involves an energy degradation does not contribute to the fast flux.

The flux at the duct mouth must be calculated to obtain sources for com-ponent (a) and (b) . The s o u r c e of component (c) i s the f lux in the unper-turbed shield. Albedo and leakage coef f i c i ents taken from the exper iments have fortunately proved to be nearly constant for the experimental conditions so far treated (outer diameter 15 cm, gap width 1 -2 cm, gap length 50 cm). The method i s promis ing for ,ca lcu lat ing the engineering t o l e r a n c e s in dif-ferent configurations of a reac tor top. These to l erances are of great e c o -nomic importance .

4 . 2 . 3 . Fas t neutron dose in material"surrounding a long duct (AE)

The shielding of the steam outlet ducts i s a severe problem in all boiling reac tors . T h e s e outlet ducts are e spec ia l ly cr i t ica l in D 2 0-moderated re -

74

actors , because the pres sure v e s s e l volume has to be kept as smal l as pos-s ible . Thus, there i s l i t t le or no place for shie lds or labyrinths inside the v e s s e l . T h e r e f o r e the compl icated construct ion e l e m e n t s in t h e s e outlets e a s i l y get r a d i a t i o n - d a m a g e d o s e s equal to or g r e a t e r than the d o s e s r e -ceived by the v e s s e l wal ls . The s imples t form of shielding, decreas ing the channel diameter, can be very expensive because of p r e s s u r e l o s s e s .

Many e x p e r i m e n t a l r e s u l t s on radiat ion attenuation ins ide a duct are available but to our knowledge nothing has been published about f luxes in the materia l surrounding a duct when the latter i s f i l led with a low density scat-t erer (e. g. steam).

A m o c k - u p for m e a s u r e m e n t s on th i s type of p r o b l e m has been built , based on the set -up described in sect ion 3 . 1 . 3 . The channel length is 1.3 m and the diameter can be varied with in ser t s of various d iameters (Fig. 50a). M e a s u r e m e n t s wi l l be s tarted during the f i r s t quarter of 1964. In th i s ex -periment as wel l as in that described in sect ion 4 . 2 . 1 the greatest difficulty i s the t rans la t ion of the angular and s p a c e d i s tr ibut ions of the s o u r c e funct ions b e t w e e n the m o c k - u p and the r e a l r e a c t o r channel (F ig . 50b).

POWER REACTOR' П 4 NGULAR F U E L ELEMENT-1 DISTRIBUTION

* P

" F A N C Y " - P R O i

a )

b )

• N E W S " Ц \ d )

Fig. SO

Neutron dose in m a t e r i a l around a duct

To study this problem theoret ical ly , two s imple computer programmes for Ferrant i -Mercury are being tested: FANCY = FAst Neutron scat ter ing in a c y l i n d r i c a l g e o m e t r y (Fig . 50c) and NEWS = NEutron Wall Scat ter ing (Fig . 50d). Both are f i r s t - c o l l i s i o n p r o g r a m m e s which c a l c u l a t e the un-co l l ided and o n c e - c o l l i d e d f l u x e s at a g iven d o s e point P .

It i s c l ear that these s imple models cannot be expected to give true values for the fast fluxes in all c a s e s . Therefore their range of validity has to be inves t igated care fu l ly . F o r th i s r e a s o n a M o n t e - C a r l o p r o g r a m m e has been init iated. Th i s p r o g r a m m e , however , i s too expens ive for pro-duction runs. P r e s e n t es t imat ions give a running t ime of 1 -2 h on the IBM 7090 per dose point. This compares to ~ 0. 5 h on the Mercury using NEWS and FANCY.

75

5. FURTHER PLANS AND CONCLUSIONS (AE)

5. 1. Check measurements in existing constructions

We Relieve that m o r e attention should be given to this point. Actually, cons ider ing the great number of nuc lear power plants in operat ion, s u r -prisingly little has been published about how well predicted dose rates agreed with m e a s u r e m e n t s made on them. The r e a s o n may be that the r e s o u r c e s in shielding have always been directed towards developing new methods, as this i s perhaps more sat isfying and more creditable from the academic point of view.

We have s t a r t e d m e a s u r e m e n t s on the b i o l o g i c a l sh ie ld ing of the R3 (Âgesta) reactor near Stockholm. This reactor wil l be brought up to power during the f i r s t half of 1964.

B e s i d e s reac tor sh ie lds , we be l ieve that a check on the other types of shie lds , e . g . c a s k s and transport shie lds , should be profitable. This check can give valuable data about the quality of the s tandard methods u s e d for routine calculat ions and for to lerances to be expected in manufacturing and construction. For example, we were once cal led upon to check a lead cask giving'double the dose predicted. It was found that the lead th ickness was 24 cm instead of the p r e s c r i b e d 25 cm.

5. 2. Bulk shielding

It s e e m s that the calculat ion methods for bulk sh i e lds have reached a state where an improvement in accuracy may be expected to be proportional to the logar i thm of the r e s e a r c h and development e f for t . This appl ies at l eas t to the stat ionary plants. E r r o r s f r o m the uncerta int ies in bas ic ra -diation s o u r c e s , f r o m b a s i c c r o s s - s e c t i o n s and f r o m m a t e r i a l s data are at l eas t of the same order of magnitude as the uncertainty in the penetration predicted by the bes t available methods.

One region where we fee l that more work is needed, i s in the prediction of neutron spectrum and radiation damage rates at the p r e s s u r e v e s s e l wall c l o s e to the c o r e . High accuracy mus t be sought b e c a u s e a use fu l l i f e of t h r e e or even 10 y e a r s ins tead of the planned 30 y e a r s would be ra ther a disappointment. We be l i eve that in connect ion with th i s prob lem one has to start studying the local variations of the dose at the pressure v e s s e l walls. The homogenizat ion of the c o r e may not be al lowed in t h e s e prob lems . It should be kept in mind, however, that the main problem in this connection i s the hitherto unsat is factori ly explained relat ionship between dose and ra -diation damage, which i s outside the scope of shielding research .

5. 3. Ducts and voids

The poss ib l e number of var iat ions i s pract ica l ly unlimited, there fore the direction of r e s e a r c h must be dictated by problems aris ing in pract ical design work. Monte-Carlo calculations s e e m to offer a useful general , though expensive, tool for the solution of these problems. Classes of prob-l e m s of general in teres t wi l l be s e l ec ted and the Monte-Carlo calculat ions

76

will be used for developing and checking s impler methods for daily use. The current trend along this line has been given in section 4. 2 for straight ducts and for the ir surroundings . Other c l a s s e s of i m p o r t a n c e a r e ducts with i n s e r t s of v a r i o u s f o r m s , e . g. h e l i x e s . E v e n mul t ip l e bends m a y be of i n t e r e s t .

5. 4. Engineering manual

"The art of shielding" has advanced during the last decade from the state of really being an art, towards being a more complete sc ience of engineering This has , at the s a m e t i m e , led to a t i m e lag be tween the m o r e b a s i c r e -s e a r c h work and appl icat ions . The r e a s o n for th i s i s that it was found to be dif f icult , for the r e a s o n s touched upon in sec t ion 5. 1, to get p h y s i c i s t s in teres ted in the eng ineer ing appl icat ions . T h e r e f o r e the t rans format ion of the re search resu l t s into a form useful to technicians has been neglected. We have made e f for t s to mee t our own needs , but we f e e l that much m o r e should be done. A good l i s t of r e f e r e n c e s with d i f ferent p r o b l e m s taken as subject categories would be much more useful than l i s t s of data and equa-t ions . Of course , the next step i s the condensing of t h e s e r e f e r e n c e s into a s ingle vo lume.

T h e r e f o r e , the in i t ia t ive of the IAEA in edi t ing a p r a c t i c a l handbook i s m o s t grateful ly acknowledged. We may a s s u m e that it i s going to take another 20 y e a r s b e f o r e a standard handbook in sh i e ld ing c o m p a r a b l e to the handbooks for e l ec t r i ca l engineers , boi ler engineers etc. can be edited. The IAEA1 s effort i s going to lay a good foundation for this book, especial ly if subsequent new editions can be expected from time to t ime. In the mean-t ime it i s s t i l l n e c e s s a r y to make a gradually decreas ing local effort for the transfer of information on physical methods to the technicians.

5. 5. Overall station planning

The greatest part of overal l station planning i s conventional power en-gineering and the work is therefore administered by 'conventional' engineers. The new extra i tems result ing from nuclear power may from the conventional point of v iew be roughly divided into: reac tor phys ic s , mater ia l problems and shie lding and health p h y s i c s . The f i r s t two are the m o s t dramatic in the e y e s of the conventional eng ineers . Health p h y s i c s i s for obvious r e a s o n s a subject in which any controvers ia l r e m a r k s by one who i s not a health physicist may be almost considered as a near sabotage. One nowadays often f e e l s that the sh ie ld des igner i s cons idered as a fe l low who s p l a s h e s around chunks of concrete and guarantees that everything i s operable after the others have got the station pract ica l ly ready.

Reactor d e s i g n e r s have only recent ly l earned to accept that a proper approach to shielding p r o b l e m s i s important to station operation and to the f inal construct ion c o s t s . This demands that the sh ie ld d e s i g n e r take part in all s tages of station planning. In order to achieve this, continuous training in shielding of a suff ic ient number of people with a sound engineering back-ground i s n e c e s s a r y . This wil l in the long run contribute to better station planning.

77

R E F E R E N C E S

[1] AALTO, E. and NILSSON, R., The Radiation Attenuation in Massive Biological Shields and a Study of

the Accuracy of Some Design Methods, to be-published.

[2] AVERY et a l . , Methods of Calculation for Usa in the Design of Shields fot Powe! Reactors, Rpt AERE-R-3216

(Feb. 1960).

[3] AALTO, E. and NILSSON, R., Measurements of Neutron and Gamma Fluxes through Thick Shields of

Magnetite and Barytes Concretes. A Comparison with Calculation, EAES Symp. Nucl. Reactor Shielding,

Studsvik (Mar. 1961).

[4] AALTO, E., to be published.

[5] NILSSON, R. and AALTO, E. .Tests of neutron spectrum calculations by means of foi l measurements in a

DzO- and an H20-moderated reactor and in reactor shields of concrete and iron, EAES-Symp. Fast and

Epithermal Neutron Spectra in Reactors, Harwell (Dec. 1963).

[6] DURING, G„ JANSSON, R. and STARFELT, N. , Experimental Fast Neutron Spectra in Al and Fe, to be

published in Arkiv for Fysik.

[7] JOHANSSON, S. A. E., On the possibility of using model experiments to study shielding problems, Nucl.

Sei. Engng 14 2 (1962) 196.

[8] FORSBERG, L. e t a l . , The User's Manual for the NRN-Shield Design Method, to be issued (1964).

[9] FORSBERG, L. and LEIMDÖRFER, M. , A New Method for Calculat ing the Penetration of Neutrons in

Reactor Shields, to be published.

[10] LEIMDÖRFER, M. , A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed

Sources in Laminated Shields. Nukleonik 6 (1964) ; On the Transformation of the Transport Equation for

Solving Deep Penetration Problems by the Monte Carlo Method, to be published in Trans. Chalmers Univ.

Tech. (Mar. 1964) ; A Monte Carlo Method for Calculating the Penetration and Energy Distribution of

Gamma Radiation from Distributed Sources in Laminated Shields, Trans. Amer. Nucl. Soc. 6 2 (1963).

[11] LEIMDÖRFER, M . , The Backscattering of Gamma Radiation from Plane Concre te Walls, Nucl. Sei.

Engng 17 (1963) 345; The Backscattering of Gamma Radiation from Spherical Concrete Walls. Nucl.

Sei. Engng 17 (1963) 352; Multiple Reflection of Gamma Radiation in a Spherical Concrete Wall Room.

Nucl. Sei. Engng 17 (1963) 357.

[12] LEIMDÔRFER, M. u s e Monte Carlo methods for solving gamma radiation transport problems,

Nukleonik £(1964) .

[13] LEIMDÔRFER, M. , The Backscattering of Fast Neutrons from Plane and Spherical Reflectors, to be published

in Trans. Chalmers Univ. of Tech. (Mar. 1964).

[14] LEIMDÔRFER, M. , On the use of Monte Carlo techniques for calculating the deep penetration of neutrons

in shields, to be published in Trans. Chalmers Univ. Tech. (Mar. 1964).

[15] BRAUN, J. and RANDEN, К . , Neutron Streaming in DzO Pipes, Rpt AE-98 (1962) ; Trans. Amer. Nucl. Soc. 5 2 (Nov. 1962) 397.

[16] NILSSON, J. and SANDLIN, R., The Transmission of Thermal and Fast Neutrons in Air Filled Annular Ducts through Slabs of Iron and Heavy Water, to be published.

[16] NILSSON, J. and SANDLIN, R. , The Transmission of Thermal and Fast Neutrons in Air Filled Annular Ducts through Slabs of Iron and Heavy Water, to be published.

[17] PIERCY, D. C . , The transmission of thermal neutrons along air filled ducts in water, Rpt AEEW-70 (1962).

78

STATUS R E P O R T ON R E A C T O R SHIELDING R E S E A R C H IN S W I T Z E R L A N D

J..M. PICTET AND A. ETEMAD*

1. PAST HISTORY AND FUTURE PLANS

1.1. Past history

R e s e a r c h on r e a c t o r sh ie ld ing s tar ted in 1958 in order to provide in -formation needed for our national reactor development programme. F r o m the var ious s tudies and ca lculat ions per formed up to now, we have c h o s e n s o m e invest igat ions of more general interest to be descr ibed brief ly in this paper.

1.1.1. Bulk shielding studies at the reactor SAPHIR

An experimental shielding facility has been set up in the pool of the 1-M\y swimming pool reactor SAPHIR located at the EIR in Wiirenlingen. The core of the reactor i s used as the radiation source . Bulk shie lds such as water , ordinary concrete (p = 2.4 g / c m 3 ) , barytes concrete (p = 3.6 g / c m 3 ) or iron-water laminated s h i e l d s (with and without boron i n s e r t s ) w e r e s tudied by means of this faci l i ty . Concrete and iron plates of 100 cm X 100 cm and of different thicknesses were used to mock up concrete shields of 120-cm thick-ness and iron-water shields containing iron thicknesses of up to 60 cm.

The e x p e r i m e n t s c o n s i s t e d of m e a s u r i n g the d i s tr ibut ion of f a s t (E > 0.5 MeV), intermediate (0.5 MeV > E > 0.125 eV) and thermal neutron fluxes and gamma dose rates through the shie lds . Measurements were per-formed by means of activation detec tors , such as indium, gold and dyspro-s i u m for i n t e r m e d i a t e and t h e r m a l neutrons and phosphorus , a l u m i n i u m and indium for fast neutrons. Whenever poss ible and des irable , BF3 -counters w e r e u s e d for thermal neutron flux m e a s u r e m e n t s . F o r g a m m a dose rate m e a s u r e m e n t s we u s e d graphite ionizat ion c h a m b e r s .

The r e m o v a l flux d i s tr ibut ions through the s h i e l d s w e r e d e t e r m i n e d by using a code (named FLUSS) which performs an integration of the Albert-Welton kernel over the reactor core numerical ly . This calculation enabled us to de termine the rat io of the thermal flux to the r e m o v a l flux at equi l i -br ium in water , ordinary concre te and barytes concrete .

A three -group (removal , intermediate and thermal) model was applied to ca lculate the thermal neutron flux in i r o n - w a t e r laminated sh i e lds . R e -s u l t s w e r e found to be in good a g r e e m e n t with the m e a s u r e d dis tr ibut ion.

The bulk shielding s tudies mentioned here are reported in detail in [1].

1.1.2. Streaming of neutrons in ducts

The s t r e a m i n g of neutrons ( e s p e c i a l l y t h e r m a l neutrons ) in s t ra igh t cyl indrical ducts through water was studied experimental ly by means of our

* Presented by J. M. Pictet .

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shie ld ing fac i l i t y at the r e a c t o r SAPHIR. M e a s u r e m e n t s of neutron f lux distributions were performed in two ducts of different dimensions (one 250 cm in length by 12.5 cm in diameter and the other 170 cm in length by 5.9 cm in diam. ). The thermal neutron flux in water around the ducts was also mea-sured. Activation detectors (phosphorus pe l le ts , gold and dysprosium foi l s , dyspros ium and copper bands) as wel l as BF 3 - counters were used for these m e a s u r e m e n t s .

A deta i led method was deve loped for ca lcu lat ing the t h e r m a l neutron flux in ducts through hydrogenous media . This method treats the distribu-tion of neutrons by means of transport theory inside the duct and by means of two-group dif fusion theory in the medium, surrounding the duct. Appli-cat ion to ducts through water gave r e s u l t s in e x c e l l e n t a g r e e m e n t with m e a s u r e m e n t s .

A code named FANET i s being written for calculating the fast neutron flux in ducts. This code per forms a numerical integration of a kernel , de-rived from the resu l t s of moments method calculations, over the core of the reactor . Pre l iminary resu l t s obtained from this code indicate good agree-ment with m e a s u r e m e n t s .

D e t a i l s of work on neutron s treaming in ducts (except for calculat ions of fast neutrons) are reported in [2].

1.1.3. G a m m a - r a y backscatter ing in s lab g e o m e t r y

A build-up corrected, s ing le -co l l i s ion kernel was developed to calculate the backscattered energy flux at a speci f ied location result ing from a mono-e n e r g e t i c , m o n o - d i r e c t i o n a l g a m m a - b e a m incident at a g iven angle on a s c a t t e r e r of s p e c i f i e d th i ckness . The formulat ion i s supported by pheno-menolog ica l analys is and by comparing the total energy albedos, as calcu-lated by a s ing le -co l l i s ion kernel, with Monte-Carlo data. The angular dis-tribution of the r e f l e c t e d energy i s computed by m e a n s of a bui ld-up c o r -rected , s ing l e - co l l i s i on kernel for 1-MeV gammas backscattered from con-c r e t e . The r e s u l t s are shown to be in good a g r e e m e n t with Monte -Car lo data. The method i s also applied to the case of slant incidence transmiss ion p r o b l e m s . A thorough d e s c r i p t i o n of this work i s g i v e n by LOWEN [3].

1 . 1 . 4 . Shie ld ing of the DIORIT r e a c t o r

DIORIT i s a 20-MW, natura l -uran ium, h e a v y - w a t e r - m o d e r a t e d r e -search reactor located at the EIR Würenlingen. Calculations of the shielding w e r e p e r f o r m e d in 1955 by the application of a two-group di f fus ion theory for neutrons and of a build-up corrected , uncollided flux model for gamma rays .

The e f f e c t i v e n e s s of the sh ie ld has been t e s t e d at ful l r e a c t o r power and the average dose ra te s w e r e found to be of the order of 0.1 m r e m / h as compared to an expected value of the order of 1 m r e m / h . About half of the m e a s u r e d dose rate, was due to neutrons, the other half to g a m m a s . Most of these gamma radiations were found to originate in the radioact ive argon expel led through the reactor chimney. Dose rates above the average value

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were measured at certain locat ions on the surface of the shie ld correspond-ing to the exper imenta l or operational f a c i l i t i e s . T h e s e dose r a t e s , how-ever , do not affect the physical experiments because of the s m a l l extension of the area where they occur . This work i s reported [4].

The reactor contains, in the upper part of the shield, a chamber in which the radiat ion l e v e l s both during operat ion and af ter shut -down have to be kept as low as poss ib le . The part of the upper shie ld located between this chamber and the core cons is ts of iron, whereas the shielding above the cham-ber cons i s t s of concrete . The problem was to find a suitable shield for the bottom of the chamber above the i ron sh ie ld . Shielding e x p e r i m e n t s ( re -ported in [5]) per formed with appropriate m o c k - u p s at the reac tor SAPHIR showed that moderat ing m a t e r i a l containing boron should be used . F o r th i s r e a s o n a tank containing diphenyl and bora l i n s e r t s w a s i n s t a l l e d .

1 .1 .5 . Shie lding of the LUCENS r e a c t o r

The r e a c t o r , now under construct ion near Lucens by T h e r m a t o m AG, Zürich, i s án experimental nuclear power plant of 30 MW (8.5 MW(e)) using s l i g h t l y - e n r i c h e d uranium as fue l , heavy water as modera tor and C 0 2 a s p r i m a r y coolant.

Calculat ion of the shie ld was per formed by the sh ie ld ing group of Thermatom AG, Zürich , by us ing removal theory for fast neutrons anda two-group diffusion theory for intermediate and thermal neutrons. Gamma-ray attenuations were determined by using a build-up corrected, uncollided flux model . The main diff iculty cons i s ted in evaluating the radiation s treaming through the paral le l channels located in the upper shield above the fuel e l e -ments . In order to so lve this problem, experiments were performed on the penetration of neutrons and gamma rays through an iron shield (80 cm thick) containing paral le l cylindrical channels f i l led with air or water.

Theore t i ca l interpretat ion of the r e s u l t s provided s a t i s f a c t o r y a g r e e -ment . In this t reatment the channels w e r e f i r s t a s s u m e d to be empty and the t r a n s m i s s i o n of radiations was calculated by transport theory. The r e -su l t s w e r e then c o r r e c t e d by taking account of the e f fec t of m a t e r i a l s p r e -sent in the channe l s . Th i s w a s done by a s s u m i n g each c o l l i s i o n in t h e s e mater ia l s to resul t in the el imination of the incident particle from the beam.

Shielding of the auxil iary c i rcu i t s in the plant was determined by con-ventional methods of calculat ing g a m m a - r a y attenuation.

1.1.6. Shielding of the EIR hot- laboratory

The hot- laboratory of the EIR i s equipped with four 1000-c c e l l s and one 100 ООО с c e l l . Calculat ion of the g a m m a - r a y attenuation in the c o n c r e t e w a l l s (90 c m ) w a s p e r f o r m e d by means of the usual method of c o r r e c t i n g the uncol l ided flux with bui ld-up f a c t o r s . The e f f i c i e n c y of the sh ie ld ing wal ls was tested using a 10 000-c cobalt source and the measured dose rates w e r e found to be l ower than the calculated o n e s . Deta i l s of this work are reported in [6].

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1.2. Future plans

Studies for the design of a fu l l - sca le nuclear power plant are in progress in Switzerland. The design of this reactor will certainly bring new incentives for further developments in the f ield of reactor shielding. Future work will then partly depend upon the shielding requirements of this new reactor .

We have, however, a re search programme going on independently of the new reactor project. Some points of more immediate interest are discussed below. Two shie lding codes are in preparation and wi l l be ready in a few months . The f i r s t i s a mult igroup diffusion code with a r e m o v a l s o u r c e -group for neutron penetration calculations. The second is a code for gamma shielding calculations based on the usual build-up corrected, uncollided flux mode l . In these codes the problem i s treated analyt ical ly , as fa,r as pos -s ib le , in order to reduce computation t ime .

Fast-neutron spectrum studies in the shield i s another problem on which we are planning to work. One of the main purposes of these s tudies i s to find a proper way to connect a multigroup diffusion theory for intermediate and thermal neutrons with a f i r s t flight r e m o v a l theory for fast neutrons . Actually, work in this f ie ld started las t year with studies on the feas ibi l i ty of a proton reco i l , liquid sc int i l lator , fast -neutron spectrometer with high e f f i c iency . No sa t i s fac tory solution i s yet in sight.

Neutron and gamma-ray streaming in ducts needs further investigation. It i s important to study the radiation l eve l s not only inside the ducts but also in the media surrounding them. We plan to continue the work on this sub-jec t by p e r f o r m i n g new s tudies on s traight and stepped ducts through di f ferent m a t e r i a l s . The important c a s e of a matr ix of channels through m a t e r i a l s i s a l so being cons idered .

2. COMMENTS ON SOME SHIELDING PROBLEMS

2.1. Maximum permissible irradiation levels

The mean yearly accumulated dose for an occupationally exposed person should not exceed 5 r e m , as recommended by the ICRP. This i s the l imi t -ing value which should be cons idered here . F u r t h e r m o r e , it i s a l so r e -commended that the dose sha l l not exceed 3 r e m over any period of three consecutive months. The dose rates at various locations in a reactor plant are , in contras t , not submitted to l imi tat ions as long as the a c c e s s i b i l i t y to these locat ions i s such that the accumulated d o s e s mentioned above are not exceeded .

The radiation dose rece ived by the personnel of a reactor plant depends on the sh ie ld ing d e s i g n of the r e a c t o r and i t s aux i l iary s y s t e m s , as w e l l as on the operational work required from the staff. These two aspects should be considered together. The design of the shields, the amount of operational work requ ired and the number of p e r s o n s to be attached to the s ta f f , Eire f a c t o r s which wi l l be opt imized for f inancia l r e a s o n s . P r o p e r a l lowance must be made in the e s t imates to allow for uncertainty and securi ty factors .

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2.2. Uncertainty and security factors

It i s useful to dist inguish between uncertainty and security factors . By uncertainty factor we mean the maximum discrepancy which can reasonably be expected to ex i s t between an e s t imated value and the corresponding e f -fec t ive va lue obtained when the reac tor i s operat ing. Uncertainty f a c t o r s are inherent to both sh ie ld ing ca lcu lat ions and evaluat ions of operat iona l condit ions .

By s e c u r i t y factor we m e a n a factor which i s de l ibera te ly appl ied to the finEil re su l t of an e s t imate , where the uncertainty fac tors have already been accounted for. It might be required,for instance, that the accumulated dose per year, and per person should not exceed 2 r e m compared to a maxi-mum permiss ib le l eve l of 5 r e m . The securi ty factors are rather arbitrary and it i s doubtful, even after long exper ience in reac tor construction, that general ru les wil l be found such as those existing in other engineering branches.

In contras t to th i s , one can usual ly make a rough e s t i m a t e of the un-certa inty f a c t o r s with s o m e degree of a s s u r a n c e . Both the sh i e ld d e s i g n and the operat ional condit ions mus t be c o n s i d e r e d in that r e s p e c t .

2 .2 .1 . Uncerta inty f a c t o r s r e l a t e d to the sh i e ld d e s i g n

The theor ies available at present for calculating dose rates at the sur-face of shie lds surrounding a reactor core do not yield very prec i se resu l t s . Their prec i s ion i n c r e a s e s with the degree of ref inement of the calculational method and as the uncerta inty fac tors b e c o m e s m a l l e r . The d e c i s i o n whether to u s e an e laborate method with a s m a l l uncertainty fac tor or to compensate a larger uncertainty factor by a s l ight ly thicker shield i s often based pr imari ly on financial considerat ions . Aside from the theory chosen for the calculat ion, the uncertainty factors a l so depend v e r y int imate ly on the t h i c k n e s s , the compos i t ion and the g e o m e t r y of the reac tor sh ie ld , so that it i s diff icult to attribute a given value to a given theory.

2 .2 .2 . Uncertainty fac tors re la ted to the operat ional conditions

The d o s e s accumulated by the m e m b e r s of the reactor staff depend on the length of t ime they spend on routine controls , readings, serv ic ings , etc. at var ious locations subjected to radiation. It i s c lear that such periods are difficult to est imate and are subject to uncertainty factors. Some unexpected repa ir s or manipulations may have to be performed, perhaps in a more in-t e n s i v e radiat ion f i e ld , adding to such uncerta inty f a c t o r s . A s p e c i a l i s t may be required in such a situation and the planning of the accumulated doses i s upset s ince one man may absorb, within a re la t ive ly short t i m e , an ap-p r e c i a b l e part of his a l lowed y e a r l y dose . If the absorbed d o s e amounts to 3 r e m , for instance, the person in question would not be allowed to par-ticipate in operational work for at l eas t three months.

D o s e s rece ived during shut-down periods should not be underest imated s ince they are often found to be higher than those absorbed when the reactor i s running. A further uncertainty factor i s introduced by additional doses f rom interned irradiat ion, part icular ly f rom inhalation. F ina l ly , the un-

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certainty factor related to the precis ion of the dosimetry should not be over-looked, even though it i s smal l compared to others.

It i s obvious that the uncertainties mentioned depend on the plant under considerat ion and on the care which i s taken in the choice and construction of the equipment.

2.3. Experimental facilities for shielding research

Dif ferent types of exper imenta l fac i l i t i e s ex i s t for studying the pene-tration of reactor neutrons and gamma rays . Advantages and disadvantages of these fac i l i t i e s , which may be s imple or complex , have often been d i s -c u s s e d . They all suf fer f rom s i m i l a r d i f f i cu l t ies inherent in the methods of m e a s u r e m e n t .

Comparison of theoret ical f luxes , particularly for neutrons, with mea-sured ones i s usually found to be a very delicate matter . The removal flux, for ins tance , which i s used v e r y often in shie lding ca lcu la t ions , can only be compared qualitatively with the fast neutron flux measured by means of threshold detectors , proton recoi l instruments, etc. The intermediate neu-tron flux i s also very difficult to measure in a way that will allow comparison with theory. This flux can be important when dealing with reac tors which are not wel l moderated and for which it const i tutes the origin of an appre-ciable part of the capture gamma rays emitted from the core .

Only the thermal neutron flux offers the possibility of a valid comparison between theory and experiment. It must, however, be emphasized that good a g r e e m e n t in the t h e r m a l energy r e g i o n does not n e c e s s a r i l y m e a n good a g r e e m e n t in other e n e r g y r e g i o n s .

Measurements on gamma rays are performed by means of instruments which indicate pulse rates or physical doses . In the f irs t case it i s possible to use a pulse height analyser , but the convers ion into true spectra and ab-solute flux va lues i s a very del icate matter; -in both c a s e s the influence of neutrons i s diff icult to account for .

A further problem is to determine the energy spectrum inside the shield under study. Except for activation detectors for neutron flux measurements, the detect ion ins truments for spec tra l analys i s are too large and can only be placed outside the shield. The shield thickness may then be varied, but the measurements wil l always suffer from boundary ef fects .

A di f f icul ty inherent in sh ie ld ing f a c i l i t i e s us ing the r e a c t o r c o r e as radiation s o u r c e i s that the spec trum and the intens i ty distribution ins ide the source are uncertain. The fac i l i t i e s using a converter plate are m o r e suitable from this point of view but, for deep penetration studies , they are usual ly handicapped by a low source intensity.

2.4. Radiation sources

The neutron s p e c t r u m emit ted after f i s s i o n i s suf f i c i ent ly we l l known for energ ies up to the order of 10 MeV, but l e s s certain for higher energies which can s t i l l be important for shielding calculat ions . Neutron c r o s s -s ec t ion data for fa s t neutrons are s t i l l f ragmentary and insuf f ic ient . The prompt f i s s i o n g a m m a - r a y s p e c t r u m and the f i s s i o n product g a m m a - r a y

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spectra are probably wel l enough known. For capture gamma rays the ener-gy spectra have been studied suff iciently in the case of thermal neutron cap-ture, whereas data on gamma spectra following epithermal neutron captures are quite poor.

In problems re la t ing to the shielding of p r i m a r y cool ing s y s t e m s , the determination of coolant activation probably r e p r e s e n t s the main source of uncerta inty . It i s of ten d i f f icu l t to e s t i m a t e the neutron f lux r e s p o n s i b l e for a particular reaction, e . g . Oi6(n, p)Ni6, from the flux values which have been calculated for other purposes .

There does not s e e m to be any problem with the activation of impurit ies carried along with the primary coolant as long as the reactor i s in operation, s ince the act ivit ies in question are smal l compared to the intrinsic activit ies of the coolant i t s e l f . After shut-down the s i tuation may be r e v e r s e d s ince the-coolant act iv i t i es usual ly have short h a l f - l i v e s and the act iv i t ies due to impurit ies may be a ser ious source of trouble for some manipulations. The amount of impur i t i e s depends on mater ia l and construct ional f ea tures and on c o r r o s i o n and e r o s i o n and i s v e r y dif f icult to e s t i m a t e . F u r t h e r m o r e , the se impur i t i e s may accumulate at certa in loca t ions , diff icult to f o r e s e e , where they const i tute an intense radiation s o u r c e .

2.5. Methods for calculating the attenuation of neutrons and gamma rays

In genera l , methods for calculat ing neutron attenuation in sh ie lds are not yet s a t i s f a c t o r y s ince there ex i s t l arge d i s c r e p a n c i e s between r e s u l t s obtained from different methods . Even where elaborate theor ies are used, d i screpancies s t i l l s e e m to be appreciable. As mentioned in sect ion 2 . 2 . 1 . , the choice of an appropriate calculational method i s mainly a financial ques-tion. It s e e m s at present that the best solution, for reactor plants not con-cerned with propulsion, cons i s t s of using a method of medium accuracy and compensat ing the uncerta inty factor by s o m e additional sh i e ld t h i c k n e s s .

A weakness in the use of multigroup diffusion theory i s that the diffusion flux i s diff icult to connect with the remova l f lux. The age of in termediate neutrons, considered as belonging to a single group, i s difficult to establish, but the s i tuat ion i m p r o v e s when the number of groups in the in termedia te energy range i s i n c r e a s e d . F o r thermal neutrons a prob lem a r i s e s when the energy dis tr ibut ion i s expec ted to depart apprec iably f r o m a Maxwel l distribution. In this c a s e the average thermal -neutron capture c r o s s -s ec t ions should be correc ted .

The s i tuat ion with r e s p e c t to g a m m a - r a y attenuation ca l cu la t ions i s m o r e favourable. Build-up correc ted , uncoll ided flux mode l s can be used, but d i f f icul t ies often a r i s e due to the fact that build-up fac tors for hetero-geneous sh ie lds and for f inite g e o m e t r i e s are m i s s i n g .

2.6. Heat generation in shields

The heat produced by neutrons i s appreciable in the reac tor core , but s o o n b e c o m e s neg l ig ib l e in the sh i e ld c o m p a r e d to the heat g e n e r a t e d by gamma rays . The problems related to heat generation in the shield are then the s a m e as those for g a m m a - r a y attenuation, but m o r e p r e c i s i o n wi l l be required here .

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The build-up factors for energy absorption are not known for sufficient mater ia l s and g e o m e t r i e s . Maximum p e r m i s s i b l e t emperatures and tem-perature gradients in c o n c r e t e should be bet ter def ined, so that d a n g e r s re lated to crack formation and e x c e s s i v e water evaporation could be m o r e e a s i l y es t imated .

2. 7. Effects of ducts in shields

The problem of neutron and gamma-ray ducting remains one of the major diff icult ies in the f ield of shielding. The most elaborate attenuation theories l o s e a great deal of their val idity when the shie ld contains many channels . In order to de termine the radiation attenuation through ducts, the angular distributions and the flux distributions at the entrance and along the wal l s of the ducts should be known. This constitutes the mos t delicate part of the treatment . The problem i s s l ight ly s i m p l e r for gamma rays than for neu-t r o n s , s i n c e the contribution to the d o s e ra te of the component r e f l e c t e d -from the wal l s i s l e s s important. For neutrons the m o s t important group to cons ider i s the fa s t group, s i n c e it prov ides the m a j o r contribution to the dose at the channel outlet .

A C K N O W L E D G E M E N T S

We should l ike to thank Dr. F. Alder of the EIR in Würenlingen, Mr. J. Marti of Therm-Atom AG in Zürich, Mr. Binggeli and Mr. Vermeil le of the Bureau Bonnard et Gardel in Lausanne for the ir helpful a s s i s t a n c e in d i s c u s s i o n s re la t ing to this paper.

R E F E R E N C E S

[1] PICTET, I i i . and ETEMAD, A . , Bulk shielding studies at the reactor SAPHIR, EIR-Bericht 27 (1962). [2] ETEMAD, A. , Etude de la pénétration des neutrons thermiques dans des canaux cylindriques vides tra-

versant un milieu hydrogéné, EIR-Bericht 60 (1963). [3] LÖWEN, W. , A bui ld-up corrected single collision model for gamma ray-backscattering, EIR-Bericht

53 (1963). [43 PICTET, J . M . , Efficacité des blindages de DIORIT, EIR-Bericht 36 (1962); Nouvelles Techniques 9(1962> [5] PICTET, J . M . , Etude des écrans protecteurs, Nouvelles Techniques 7.(1959); 1(1960). [6] PICTET, J.M. , Calculs des protections pour les cellules chaudes, Nouvelles Techniques 9 (1963).

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SHIELDING RESEARCH IN THE UK

A summary of the history and current programme of the U. К. A. E. A: Shielding Group with an

indication of future trends

J. BUTLER

1. INTRODUCTION

R e s e a r c h into shield des ign techniques in the UK had i t s or ig ins in the construct ion of the f i r s t Harwel l reactor, GLEEP, in 1946. But it was not until the advent of the Magnox power reac tor , the construct ion of the fu l l -s ca l e prototype at Calder Hall was started in 1953, and i t s subsequent de-ve lopment in the UK Civi l P o w e r p r o g r a m m e , that the need a r o s e for the accurate methods of calculation to optimize sh ie lds with a minimum capital cos t . F r o m the outset it was c l ear that substantial economic savings could be ach ieved with i m p r o v e m e n t s in the d e s i g n methods s i n c e the sh ie ld ing requirements on a large nuclear power station have a profound influence on the des ign of the reactor i t se l f , the buildings and layout of the plant.

The methods of calculation which were available at the time were clearly inadequate to meet the needs of the c ivi l power programme which called for economy in computing effort and standardization of methods in addition to the essent ia l requirements of improved accuracy with a "safe" estimate of shield performance under all condit ions. Moreover, it was e s sent ia l that the re -commended methods of calculation should, wherever possible, avoid empiri-ca l parameters to faci l i tate their use by shield des igners without r e c o u r s e to mock-up studies for each new type of shie ld configuration.

The bas ic requirement was c l ear ly for a method of predict ing neutron penetration since the gamma-ray problem which, in the nature of the designs was c o n c e r n e d p r i m a r i l y with penetrat ion in a s ing le mater ia l , had been largely so lved by the moments calculation of Goldstein and his c o - w o r k e r s . Their resu l t s for neutrons, on the other hand, were not applicable to lami-nated shie lds containing both s tee l and moderating material such as graphite and concrete .

Against this background the concept of the energy-dependent r e m o v a l c r o s s - s e c t i o n was pioneered by the late К. T. Spinney and introduced in the recommended manual of shield design techniques* in the form of the removal age-diffusion code RASH В programmed for the Ferranti Mercury computer. The r e m a r k a b l e s u c c e s s of this method in predict ing the t h e r m a l neutron flux dis tr ibut ion in laminated sh ie lds which, by v irtue of the s e c o n d a r y g a m m a - r a y production i s the key to the b io log ica l shie ld des ign p r o b l e m , led to i t s development and application to a variety of other shielding problems on graphite-moderated power reac tors . F o r this purpose the U . K . A . E . A . Shielding Group w a s se t up at Harwel l u t i l i z ing the f a c i l i t i e s of the LIDO swimming-poo l reactor with a substantial proportion of the graduate effort

* Rpt AERE/R-3216.

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provided by staff attached from the Industrial Power Consortia. Animportant consequence of this arrangement has been the c l o s e col laborat ion between the Shielding Group in developing methods of calculat ion and the Consortia t e a m s applying t h e s e methods to pract i ca l d e s i g n s . This a s s o c i a t i o n has continued during the construct ion of the f i r s t c iv i l power stat ions and a programme of shielding measurements has been completed during the com-miss ion ing of the Berke ley and Bradwel l reac tors in col laboration with the Consort ia and Generat ing B o a r d s . T h e s e m e a s u r e m e n t s have enabled a detai led corre la t ion to be made between theory and experiment [1] and, at the same t ime, afford additional information on those aspects of the design which cannot by t h e i r nature be proper ly i n v e s t i g a t e d in the l abora tory .

The Shielding Group i s now part of the Water Reactor Phys ics Division of the Atomic Energy Establishment at Winfeyth but continues to be stationed at Harwel l in order to ut i l i ze the fac i l i t i e s on LIDO and the a c c e l e r a t o r s . The current p r o g r a m m e inc ludes the study of sh ie ld ing p r o b l e m s on both land-based and m a r i n e propuls ion r e a c t o r s including h igh- t empe rature graphite, l i gh t -water and DzO s y s t e m s , and s o d i u m - c o o l e d fast r e a c t o r s . The primary object ives remain the establishment of rel iable design methods and the Group continues to work in c lose collaboration with both the Authority design teams at the Ris ley Establishment of the U.K. A. E. A . , who are con-cerned with design of shields for prototype reactors , and the Nuclear Power Consortia. This sepàrat ion of the functions of r e s e a r c h into methods, and their application in des ign work i s an e s s e n t i a l requirement for the future development in the f ield. Detailed consideration i s now being given to trans-port codes including Monte-Carlo techniques, but in i t s la tes t form [2] the r e m o v a l a g e - d i f f u s i o n method r e m a i n s the c o r n e r s t o n e of power r e a c t o r shield des ign and i s l ikely to do so for some y e a r s with increas ing support from two-dimensional Monte-Carlo codes as computing faci l i t ies develop in the UK.

2. GENERAL APPROACH TO THE NEUTRON ATTENUATION PROBLEM

During the last few years three important developments have taken place in the f ield of power reactor shield design. Firs t , the advent of high-speed digital computers of large capacity has enabled "exact" numerical solutions to be accomplished for laminated sys tems in one dimension. Secondly, with the exploitation of importance sampling techniques, Monte Carlo has emerged as a method which i s capable of dealing with deep penetrat ion problems in complex g e o m e t r y . F inal ly , there i s the remarkable accuracy which has been achieved by the s emi -empir i ca l removal age-diffusion method, developed orig inal ly by Spinney, in deep penetration p r o b l e m s . In the light of t h e s e events it i s c l e a r that sh ie ld ing has e m e r g e d f r o m the s tatus of an art in which a we l l -des igned shield was largely a ref lect ion of the skil l of the d e s i g n e r in s ca l ing r e s u l t s obtained f r o m sh ie ld m o c k - u p s tud ies , to the present situation where a variety of numerical methods are available, some of which are potential ly capable of solving all r e a c t o r shielding prob lems .

In pract ice , it i s important to recogn i se that the principal fac tors af-fecting the choice of a design method for a power reactor shield are, in order of importance;

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(a) The ability to deal with large s tructures in complex geometry; (b) E r r o r s ar i s ing from uncertaint ies in the fundamental nuc lear data

and the composi t ion of shield mater ia l s ; and (c) E a s e of application and economy in computing c o s t s .

Judged by these standards it i s c lear the neither of the two iterative methods which have been evaluated in the UK, the Carlson Sn code SAINT [3] written at AWRE, Aldermaston, and the United Nuclear code NIOBE [4], i s suitable for design work. The former which i s a group method in which the angular flux i s expanded as the sum of n l inear functions of ц (the cos ine of the angle between the axis and neutrons' direct ion of motion) i s available in plane or spher ica l geometry w h e r e a s the latter , which e m p l o y s a Legendre p o l y -nomia l expans ion of the angular flux, i s r e s t r i c t e d to s p h e r i c a l s y s t e m s . In p r a c t i c e both s u f f e r f r o m capaci ty l imi ta t ions , SAINT i s the m o r e r e -s tr ic ted and l imitat ions are imposed on the o r d e r n of the angular f lux ex-pansion, the number of groups and the number of rad ia l (or axia l ) m e s h points. The present vers ion of fers the alternatives of either a large number of groups (to give proper coverage in the MeV region) at smal l penetrations or fewer groups with a corresponding i n c r e a s e in penetration. It i s there-fore suitable for p r e s s u r e - v e s s e l damage spectrum calculations, subject to e r r o r s ar i s ing in the g e o m e t r i c a l approximation, where the lower energy l imit of interest i s about 0.1 MeV and only moderate penetrations appropriate to the pressure v e s s e l walls are required. Such a calculation will take about one hour on the IBM 7090. NIOBE on the other hand, can deal with the deep penetrat ion problem but the computation t ime b e c o m e s prohibit ively large (of the order of hours) when the spectrum i s carried down to thermal energies . In each c a s e the t e s t s undertaken at H a r w e l l have shown that the e r r o r s ar i s ing from the approximate representat ion of a power reactor shie ld in a one-dimensional code can exceed a factor of two in regions which materially affect the performance of the design.

These geometric l imitations are to some extent overcome in the Harwell Monte-Carlo programme McNID [5] which uses splitting and Russian Roulette. It i s written in two dimensions and calculates the flux distribution in a system composed of regions bounded by со-axia l cyl inders and planes perpendicular to the a x i s . McNID can there fore be used for bulk shield calculat ions and it wi l l predict neutron f luxes s treaming along cy l indr ica l or annular voids containing s t eps . Prov i s ion i s a l so made for the s imulation of a matrix of such voids by reflection p r o c e s s e s . The specif ication of splitting boundaries i s at present left to the u s e r and the optimum choice in a complex duct con-figuration ca l l s for a considerable amount of expert i se . This difficulty can be overcome, in s imple geometr ies at least , by using an analytic importance function, but the computing t ime requirements for a typical Magnox shie ld conf igurat ion c a r r i e d down to low e n e r g i e s would s t i l l appear to be of the order of one hour on the IBM 7090. Nevertheless , codes such as McNID, and the United Nuc lear code ADONIS [6] which i s wri t ten in t w o - d i m e n s i o n a l rectangular geometry presumably comes into this category, can in principle reduce the g e o m e t r i c a l e r r o r s in many p r a c t i c a l p r o b l e m s to neg l ig ib l e proportions. In these c i rcumstances the factors which should be considered in a s s e s s i n g the place of Monte Carlo in practical shield design are:

(a) Success fu l operation ca l l s for a cons iderable degree of ski l l in the

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u s e of computers , the choice of splitting boundaries and the proper interpretation of resu l t s .

(b) The capacity of the programmes in t erms of the number of spatial reg ions and spec tra l coverage i s at present l imi ted for prac t i ca l p r o b l e m s on large r e a c t o r s .

(c)~ The computing t ime requirements for c a s e s which are carried down to low e n e r g i e s in non-hydrogenous media wi l l be of the order of hours on the IBM 7090 which i s excess ive for routine design studies.

(d) Currently available codes [5-7] do not calculate the thermal flux, which i s usually the most important factor in the design. Monte Carlo i s not, of course , an eff ic ient method for thermal flux calculations in sh ie lds even when the approximation of a s ingle thermal group i s employed in which al lowance i s made for sca t ter ing with z e r o energy l o s s .

(e) The accuracy at deep penetrations, even when the geometry i s cor-rect ly treated, i s s t i l l l imited by e r r o r s in the nuclear data and by uncertaint ies , which are probably the more ser ious in practice, in the composit ion of the shield material .

The impor tance of the se f a c t o r s i s di f f icult to a s s e s s and w i l l depend to some extent on the nature of the problem and the computing faci l i t ies avail-able to individual organizations. Thus a substantial improvement with regard to i t ems (b) and (c) above will be possible in future codes written for the new computers of the ATLAS generation although in the UK, for economic reasons, the full potential of Monte Carlo i s unlikely to be real ized for several years .

The problem of c r o s s - s e c t i o n e r r o r s which i s inherent in all r igorous techniques i s part icularly important in penetration calculat ions . Thus in a typical power reac tor shie ld and overa l l attenuation of radiation dose rate from source to e m e r g e n c e i s of the order of е~зо from which it fo l lows that an error of 5% in the basic c r o s s - s e c t i o n s will result in an overall uncertainty of factor 4. 5 in the calculated external dose rates . In practice the accuracy of c r o s s - s e c t i o n s in the MeV region, which determine the penetration, i s nearer to 10% for the common mater ia l s and in bio logical shie ld concre te s uncertainties in the composition will enhance the errors in the average cross -sec t ions computed for the .mixture. The migrat ion of water in shield con-c r e t e s may a l s o be a s igni f icant factor during the operat ional l i fe of the v

reactor and a change of the hydrogen content by 0.1% can affect the thermal flux by as much as a factor of two. Resonance activation predictions afford another example where again both nuclear data e r r o r s and uncertaint ies in compos i t ion are important.

Under t h e s e condi t ions an e laborate ca lcu la t ion of high p r e c i s i o n i s frequent ly not jus t i f i ed and the r e m o v a l age d i f fus ion method b e c o m e s an indispensable tool for rapid calculations with an accuracy which i s adequate for m o s t p r a c t i c a l p u r p o s e s . A detai led evaluat ion of th i s technique e m -bodied in the la tes t v e r s i o n of the RASH codes , RASH E, i s presented in the accompanying paper to this panel [2] where it i s shown that in spite of i ts s e m i - e m p i r i c a l nature the scope and accuracy of the model can be under-stood in qual i tat ive t e r m s by cons ider ing the b a s i s of r e m o v a l f l ight c o r -rect ions to age theory. The method can be s imply adapted to complex geo-metry by virtue of the l ine of sight removal source term which may be com-puted by numerical integration over the core; the subsequent build-up of the

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low energy fluxes being re la t ive ly independent of g e o m e t r y . In s i tuat ions where the m o d e l i s va l id , the o v e r a l l e r r o r s in good g e o m e t r y a r e about a factor of two or three and in more complex g e o m e t r i e s they do not general ly exceed a factor of f ive .

It i& of interest to note a further advantage of the removal age-diffusion model which has not yet been fully real ized. The energy-dependent removal c r o s s - s e c t i o n s in water have been m e a s u r e d [8] to within an a c c u r a c y of better than 1%, which i s significantly higher than that assoc iated with "barn book" data in the MeV region. Considering for s impl ic i ty a plane paral le l beam of monoenergetic fast neutrons incident on a homogenous water shield; the s lowing-down densi ty of "diffusing" neutrons of age т given by the r e -moval age -d i f fus ion mode l i s

q(TZ)=£ r exp-(^T-Ef .Z) , (1)

where Er i s the removal c r o s s - s e c t i o n at the source energy. A s i m p l e ana lys i s of the e r r o r s wi l l su f f i ce to show that the example

quoted above of a 5% error in the fundamental c r o s s - s e c t i o n data resu l t s in an o v e r a l l e r r o r of the order of 40% in the s lowing-down densi ty at the thermal cut-off when Er i s determined to within 1%. This e s t imate should be c o m p a r e d with the fac tor of 4. 5 above for the corresponding c a s e of a Monte Carlo or other r igorous methods. Thus on the l eve l of an empir ica l technique in which the fundamental quantity n a m e l y , the r e m o v a l c r o s s -sect ion , i s determined by experiment , the remova l age -d i f fus ion mode l i s in tr ins ica l ly capable of grea ter accuracy at deep penetrat ions . Cons ider -ation i s accordingly being g iven at Harwel l to the m e a s u r e m e n t of energy dependent r e m o v a l c r o s s - s e c t i o n s for non-hydrogenous m a t e r i a l s .

To summarize therefore, the approach which i s at present recommended in the UK for the design of power reactor shie lds i s based on the code RASH E with l ine of s ight c o r r e c t i o n s for s t r e a m i n g in ducts . The m o r e r i g o r o u s techniques such as NIOBE and Carlson , however , play an important part in establ ishing the validity of the s imple methods in regions of the spectrum which are not readi ly a c c e s s i b l e to exper iment . In the light of the recent resu l t s outlined in [2] it has been poss ib le to spel l out c lear ly the l imitation of the RASH method and when these are encountered in practice they can be o v e r c o m e e i ther by i n c r e a s i n g the degree of e m p i r i c i s m with r e s t r i c t i o n s on the energy group widths, or by having recourse to a more rigorous tech-nique. Since t h e s e s i tuat ions invar iably o c c u r in c o m p l e x g e o m e t r y th i s technique must c learly be a two-dimensional Monte Carlo. The present code McNID i s too complicated for use in design problems, i t i s writtenin machine code f o r the IBM 7090 and u t i l i z e s an ear ly v e r s i o n of the AWRE c r o s s -sec t ion data l ibrary . It i s there fore being rewri t ten in FORTRAN for the ATLAS computer and the project i s scheduled for complet ion during 1964.

In the author's opinion these recommendations wil l continue to meet the des igner ' s requirements in the UK for some years s ince RASH E affords an order of magnitude sav ing in computing c o s t s c o m p a r e d with the m o s t e f -f ic ient one -d imens iona l transport and Monte-Carlo codes avai lable*. The role of two-d imens ional Monte Carlo wi l l of course i n c r e a s e in importance

* A typical power reactor shield configuration takes between 5 and 10 min on the IBM 7090.

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when the ATLAS comput ing f a c i l i t i e s b e c o m e ava i lab le for d e s i g n work. The two methods are to some extent complementary and obviate the need for the v a r i o u s "middle of the road" techniques , such as Carlson Sn, NIOBE and the Pn multigroup approach for example, which offer neither the virtues of economy and versat i l i ty on the one hand, nor those of a proper treatment of complex geometry on the other.

3. MEASUREMENT OF SHIELD PERFORMANCE

In spite of the recent developments which have taken place in penetration codes , a bulk shielding faci l i ty i s st i l l an e s sen t ia l requirement for test ing the methods of ca lculat ion under control led condit ions . The nature of the des ign i s large ly determined by the current l imitat ion of radiation detector sens i t iv i ty and, moreover , it i s difficult to mee t the dual requirements for the study of both neutrons and gamma rays in one fac i l i ty . In neutron at-tenuation problems there are two cr i ter ia to cons ider; f i r s t , in bulk shield studies it i s only n e c e s s a r y to measure out to penetration distances at which the asymptotic equil ibrium conditions are achieved and, s imi lar ly , there i s no requirement to m e a s u r e in duct configurations beyond the point at which the geometr ic inverse square attenuation has been established. Experience at Harwel l has shown that these conditions are general ly fu l f i l l ed in an at-tenuation range of the order of 107. Activation detectors are essent ia l for at-tenuation studies in solid shield materials and the minimum "fast" flux which can be m e a s u r e d with irradiat ion per iods of a few hours duration i s about 103ncm~2 s"1 using the S^fn, pJP32 reaction; the fast flux incident on the fa-ci l i ty should therefore exceed 101 0n cm" 2 s _ 1 .

T h e s e requirements rule out the conventional f i s s ion plate but they can be achieved in a panel type of facility in which the neutron source i s provided direct ly by the fuel e l ements in a swimming pool reactor core . In the LIDO fac i l i t i e s [9] , for example , the emergent fast flux at the panel i s about 5 X 10 1 0 n cm"2 s"i during operat ion at 100 kW. This s o u r c e s trength i s su f -f ic ient to permit the use of var ious f i l t e r s composed of s t ee l , lead, water and graphite to modify the spectrum incident on the shield, and in the recent studies of the penetration of neutrons in the keV range [2], s labs of s teel up to 2 ft thick have been used to produce a spec trum in which the proportion of neutrons above 1 MeV i s negl igibly s m a l l .

The main disadvantages of this type of fac i l i ty are fourfold: (a) The absolute accuracy of the f i s s ion source strength i s rather lower

than that of a f i s s i o n plate but certa inly , with carefu l cal ibration, better than 10%. The variation with power l eve l due to control rod posit ion in an enriched l ight-water-moderated system of this type i s however, negligibly smal l .

(b) The geometry i s somewhat i l l -def ined although the resul t s obtained with NIOBE show agreement to within experimental e r r o r for fast neutron react ion rate predict ions beyond about 50 cm in water.

(c) The source i s l oca l i zed over a s m a l l volume, but this of course i s a l so true of mos t conventional f i s s i o n plates .

(d) The gamma background, principally due to prompt f i s s ion in the fuel and thermal capture in the aluminium core support structure, i s high

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and in some shield configurations precludes the use of ionization dev ices and nuclear emuls ions .

In p r a c t i c e these l imi ta t ions are not s e r i o u s , the a c c u r a c y of 10% i s quite acceptable for shielding studies; indeed, it i s doubtful whether agree-ment to bet ter than this i s meaningful at penetrations in e x c e s s of ten mean f r e e paths . The treatment of the core geometry p r e s e n t s no diff iculty for removal age-d i f fus ion codes and in comparisons with transport calculations in spher i ca l geometry the e r r o r s can be reduced by withdrawing the c o r e away from the panel into the pool. F o r problems in which a l arger source area i s required, such as a matrix of parallel ducts for example, the lateral and v e r t i c a l "bucklings" can a l s o be reduced in this way for l o w - e n e r g y neutrons; alternatively s labs of graphite can be interposed between the panel and the shie ld . This technique i s not suitable for high-energy neutrons al-though the p o s s i b i l i t i e s of us ing a s t e e l "lens" s y s t e m are being examined at Harwel l .

The Pane l С faci l i ty on LIDO has recent ly been modi f ied to take dry shield combinations with a total th ickness in the axial direct ion up to 12 ft. The l a t e r a l d i m e n s i o n s of the e x p e r i m e n t a l s h i e l d a r e 6-ft square which, together with anti -streaming steps, i s sufficient to ensure that the axial flux at deep penetration i s not perturbed by leakage from the s ides of the facil ity. Special precautions are taken to reduce activation of the steel work and shield slabs, and activation detectors are used with ha l f - l i ves which are long com-pared with the dominant act iv i t ies generated in the structure to permit their withdrawal after suitable decay periods .

In o r d e r to t e s t a l l the v a r i o u s a s p e c t s of the d e s i g n methods it i s at p r e s e n t n e c e s s a r y to have r e c o u r s e to a l t ernat ive s o u r c e s of data in the fo l lowing c a s e s :

(a) Neutron s p e c t r a in the range b e t w e e n 300 e V and 0.5 MeV; (b) F a s t neutron penetrat ion in duct m a t r i c e s ; and (c) The attenuation of neutrons in large g a s - f i l l e d vo ids and m a s s i v e

s t e e l s t r u c t u r e s . F o r i t e m s (a) and (b) above c o m p a r i s o n s a r e made at p r e s e n t with

spectra computed by transport codes and Monte Carlo in s imple geometry , and reaction rate predict ions are compared direct ly with integral measure -ments made in the faci l i ty . The problems l i s ted under i tem (c) can only be proper ly a s s e s s e d by m e a s u r i n g on a prototype r e a c t o r during the c o m -miss ioning stage; reference has already been made to work at Bradwell and Berke ley , and m e a s u r e m e n t s on the Authority r e a c t o r s at Calder Hall and Chape lcross are d i s c u s s e d in [2] .

The panel fac i l i t i e s are not, however, suitable for determining the ab-solute accuracy of r igorous methods at deep penetrat ions . This c a l l s for a source with c lean geometry and a wel l -def ined spectrum which i s real ized in t i m e - o f - f l i g h t m e a s u r e m e n t s with a l i n e a r e l e c t r o n a c c e l e r a t o r [10] . S imi lar ly they are not, in the author's opinion, u se fu l for g a m m a - r a y penetrat ion s tudies b e c a u s e of the complex s o u r c e distribution of g a m m a s f r o m f i s s i o n and capture in the c o r e and capture in the sh i e ld m a t e r i a l .

4 . OUTLINE OF THE RESEARCH PROGRAMME

The general programme of r e s e a r c h in shielding techniques at Harwell can be summarized under the following headings:

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(a) Monte-Carlo and transport codes for neutron penetration; (b) Evaluat ion of d e s i g n methods b a s e d on the r e m o v a l a g e - d i f f u s i o n

model ; (c) G a m m a - r a y scat ter ing , bui ld-up and oblique penetration; (d) Act ivat ion of coolant and shie ld components ; and (e) Measur ing techniques . The main i t e m s which fall into these ca tegor ies are outlined below and

give s o m e indication of the scope of the current programme and the trends envisaged during the next few y e a r s .

4.1. Monte-Carlo and transport codes

The place of both Monte-Carlo and numerical solutions to the transport equation have been d i s c u s s e d above in sec t ion 2. The main tasks with Monte Car lo apart f rom the preparat ion of the new v e r s i o n of McNID in FORTRAN will be the comparison of predictions with experiment to deter-mine the accuracy which can be achieved in a pract i ca l des ign and the in-ves t igat ion of importance sampling techniques .

Studies are currently in progress to evaluate the Carlson Sn method for shielding, and s imi lar investigations will also be made of Pn codes. In each case studies are required of the importance of the order of scattering treat-ment and the expansion of the angular flux. The effect of changing the various input parameters of NIOBE is also being investigated and parametric studies are planned to d e t e r m i n e the e f fec t in t h e s e v a r i o u s methods of d i f ferent kinds of errors in the fundamental nuclear data in order to define more clearly the users ' nuclear data requirements.

4.2. Evaluation of design methods based on the removal age-diffusion techniques

The v a r i o u s d e s i g n methods which are current ly being developed are discussed in [2], and the general programme which has been adopted to over-come limitations in measuring techniques and LIDO facil it ies is summarized in Table III.

4. 3. Gamma-ray scattering and build-up problems

It i s apparent f rom the e m p h a s i s in this d i s c u s s i o n that hitherto, at-tention has been devoted mainly to neutron attenuation problems. The status of the latter, however, i s now such that in some c a s e s the overriding factor affect ing the accuracy of the des ign i s the inadequacy of g a m m a - r a y tech-niques. Problems in this category are associated with build-up and oblique penetrat ions in laminated shie lds , e spec ia l ly at low energ i e s in the des ign of transport coff ins and nuclear heating calculations where the conventional technique of the s ingle equivalent build-up factor can lead to ser ious o v e r -e s t i m a t e s . The s ca t t er ing of g a m m a r a y s in c o n c r e t e - l i n e d ducts a l s o presents di f f icul t ies for s imple design methods based on the albedo concept. Both t h e s e a s p e c t s can be treated by Monte Carlo but there i s a need for contro l led e x p e r i m e n t a l data to de termine the a c c u r a c y .

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TABLE III

THE COMPILATION OF E X P E R I M E N T A L D A T A FOR T H E EVALUATION OF THE RASII DESIGN METHODS

Source of data

Shield conf igurat ion

Integral measurements in the LIDO fac i l i t ies

Simple flux measurements m a d e during commissioning and

operation of prototype power reactors

Spectral comparisons with transport codes

Bulk shields Laminated slab shields including water , D 2 0 , graphi te , s teel , lead, sodium and boron containing mater ia ls

Measurements in holes drilled in b iological shields

NIOBE, Sn and Monte Car lo in s imple geometry

Shields pierced by a single duct system

(i) Straight cyl indr ical ducts (ii) Straight and stepped annular ducts

(iii) Bent ducts (iv) Stepped and bent slots (V) Ducts and slots f i l led with solid

mate r ia l or liquids

Measurements in coolant ducts, instrument locat ion holes and beam tubes e tc .

ч

M с NID - Two-dimensional Monte Carlo

Mult ip le duct systems The study of various mat r ices as a funct ion of pitch and d iamete r with both air f i l led ducts and ducts conta ining solid mater ia l s and liquids

Measurements in reflectors and ax ia l b iological shields

McNID — Using re f lec t ion processes to s imula te interact ion ef fec ts

Complex geomet ry Measurements with LIDO core for smal l water reactor systems

Scans round periphery of ref lector and pressure vessel and across air gaps

McNID for smal l cores

G a m m a - r a y s o u r c e e s t i m a t e s f r o m f i s s i o n products of plutonium in irradiated fuel are at present subject to uncertaint ies due to a lack of data, and a p r o g r a m m e has been writ ten to compute s p e c t r a for t h e s e i s o t o p e s along the l i n e s of the c l a s s i c compi la t ion of PERKINS and KING [11] .

4. 4. Activation of coolant and shield components

Considerable attention has been devoted in the UK to the predict ion of coolant actiyity in CO2 gas-cooled power reactors and the principal activities due to N1 6 and A r 4 1 can now be calculated f rom des ign data to within about 50%. S imi lar s tudies are being undertaken for light and heavy water, and sodium coolant ac t iv i t i e s .

Cons iderat ion has a l s o been given to ep i thermal act ivat ion and work i s in p r o g r e s s to der ive average c r o s s - s e c t i o n s for RASH energy groups for m a t e r i a l s in the ep i thermal range.

4. 5. Measuring techniques

The development of measur ing techniques i s an important facet of shielding re search where there i s often a requirement for the determination of quantities which are not of general interest in reactor design.

The special problem in penetration studies i s to measure reaction rates and spec tra over an attenuation range of the order of 101 in the presence of high gamma and thermal neutron backgrounds. Considerable effort has ac-cordingly been devoted to activation and permanent recording techniques for the m e a s u r e m e n t of fas t neutrons including sulphur burning [12] , the r e -action Rh103 (n, n') Rh103m [13] which has a threshold at about 0.3 MeV and the measurement of f i s s i o n rates in U235 under boron by etching damage tracks in a polycarbonate re s in .

Fol lowing the recent measurements of the fast neutron spectrum in the BSR at Oak Ridge by VERBINSKI et al . [14] attempts wi l l be made during 1964 to u s e s i l i con b a r r i e r s p e c t r o m e t e r s in the LIDO panel f a c i l i t i e s in addition to nuc lear e m u l s i o n s . Spectra up to 300 eV have already been m e a s u r e d with f ive resonance sandwich de tec tors [15] in a var ie ty of s lab shield arrays including graphite, sodium and iron, and consideration i s being g iven to other r e s o n a n c e r e a c t i o n s in an attempt to extend th i s technique up to the keV reg ion .

Various methods for measuring the angular distribution of thermal and r e s o n a n c e neutrons in ducts a r e current ly be ing e x a m i n e d . T h e s e t e c h -niques will be used to investigate the ef fects of scattering from the walls and the anomalous behaviour reported by PIERCEY [16] who o b s e r v e d an asymptot ic - 2.3 power law for thermal neutrons in long ducts in the LIDO water f a c i l i t i e s .

Fur ther inves t iga t ions of techniques for m e a s u r i n g energy-dependent removal c r o s s - s e c t i o n s in non-hydrogenous mater ia l s are planned for 1964 in the energy range from about 1 to 14 MeV using the Harwell VandeGraaff and Tandem generators . The object of this work will be to determine whether the unique advantages of the s e m i - e m p i r i c a l removal model mentioned above in sect ion 2 can be rea l ized in practice for laminated shields.

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R E F E R E N C E S

[1] YOUELL, F. P. , McCRACKEN, A. K. and OLIVER, P., Bradwell Nuclear Power Station - Comparison of Shielding Performance with Design Calculations, Paper to be presented at the 3rd. 'UN Int. Conf. PUAE (1964). «

[2] BUTLER, J. , The status of theoretical methods for reactor shield design. Supporting paper presented at this Panel, Abstract in these Proceedings.

[3] WADE, R. D. (Ed. ) Solution of the One Dimensional Multigroup Stationary Neutron Transport Equation on the IBM 7030 (STRETCH) Computer, Rpt. AWRE 0-12/63.

[4] YETMAN, D. . EISENMAN, B. and RABINOWITZ, G . , Description of Input Preparation and Operating Procedures for 9-NIOBE, an IBM 7090 Code, Rpt. NDA-2143-18.

[5] BENDALL, D. E . , Mc-NID - A Monte Carlo Programme for Calculating the Penetration of Neutrons in Ducts with Cylindrical Symmetry, to be published.

[6] EISENMAN, B. and HENNESSY, E . , Advanced Shield Calcula t ional Techniques V, UNUCOR-635. [7] GOBER. W. and SHAPIRO, M . , Advanced Shield Calcula t ional Techniques III, UNUCOR-633. [8] AVERY, A. F . , BUTLER, J . , McCRACKEN, A. K. and PACKWOOD, A . , Trans. Amer. Nucl. Soc. 5

(1962) 2. [9] HYDER, H.R. McK. , WEALE, J. W., GREEN, A . , JONES, E. D. , KENWOOD, C.J . and ORAM, P. J . ,

LIDO - The Engineering and Physics Design, Rpt. AERE R/R-2340 (1957). [10] VERBINSKI, V. V. , KIRKBRIDE, J . , PHELPS, P. and COUNTENAY, J. C . , Measurements of Fast Neutron

Transport in Water by Time of Flight Techniques, Rpt. ORNL-3499 1 (1963). [11] PERKINS, J. F. and KING, R. W . , Energy Release from Decay.of Fission Products, Nucl. Sei. Engng 3

(1958) 726. [12] PARKER, B. H. , The Measurement of Fast Neutron Flux using the S ^ n , p)P32 Reaction, Rpt. AERE/R - 3443

(1960). [13] HOWEY, K. E. , The Measurement of Fast Neutrons with the Rhl03(n, n'JRh1031^ Reaction, to be published. [14] VERBINSKI, V. V. , BOKHARI, M. S. and KINGTON, J. D . , Measurements of Fast Neutron Spectra within

and at Various Distances from the Bulk Shielding Reactor 1, Rpt. ORNL-3499 1. [15] McCRACKEN, A. K. and GAMMON, R. B., The Measurement of Epithermal Neutron Spectra by means

of the Resonance Sandwich Technique, to be published. [16] PIERCEY, D. C . , The Transmission of Thermal. Neutrons along Air Filled Cylindrical Ducts, Rpt. AEEW-R70.

97

REACTOR SHIELDING IN THE UNITED STATES OF AMERICA

M. GROTENHUIS, P. S. MITTLEMAN AND E. P. BUZARD*

1. INTRODUCTION

The f ield of reactor shielding had i ts inception in 1942 with the start of the f i r s t reactor , C P - 1 . Shield design was recogn ized a s a n e c e s s a r y ad-junct to reactor des ign. The following i s e ssent ia l ly the s ta te -o f - the art in the United States twenty-one y e a r s after the f i r s t r eac tor shie ld was built. The very fact that the few basic re ferences can s e r v e so wel l indicates that the f ie ld i s maturing. That there i s s t i l l development work i s evident from late r e f e r e n c e s .

2. SOURCES OF RADIATION

The s o u r c e s of radiation which are important in reac tor shie ld des ign are those which produce neutrons or gamma rays. These sources have been studied extensively and except for ref inements [1] are unchanged in the past few y e a r s . This informat ion has been s u m m a r i z e d [2, 3] and only the new information or a r e a s that may s t i l l require e f for t s wi l l be d i s c u s s e d here . In general , the accuracy of the shield design i s not l imited by the knowledge of the primary s o u r c e s .

The secondary s o u r c e s of radiation represen t the l a r g e s t uncertainty in the descript ion of s o u r c e s and that may be ascr ibed in part to the l imited knowledge of the a s s o c i a t e d neutron c r o s s - s e c t i o n s . The product ion of neutrons by the decay of N 1 7 has had a rather l arge uncertainty , although recent resul ts [4] have been encouraging. Gamma rays from thermal neutron capture, or , neutron radiat ive capture g a m m a r a y s , a r e wel l -known with the exception of a few mis s ing e lements and a few discrepancies . The note-worthy exceptions are U235 and Pu239 among the f i ss ionable i sotopes , and Ge among the r a r e earth e l e m e n t s . D i s c r e p a n c i e s a r e unfortunately in the reg ion of 3 to 4 MeV, which i s often the energy of g r e a t e s t penetrat ion in shielding problems. In addition, a number of e lements have spectra measured above 3 MeV but not below that energy. At a recent international conference [5], two full s e s s i o n s were devoted to recent work on capture gamma rays . It should be noted that the capture g a m m a - r a y spec tra for the concre te s in the above r e f e r e n c e s [2, 3] have been r e v i s e d [6] to accommodate the data of TROUBETZKOY and GOLDSTEIN [7] . E f for t s a r e now needed to de-termine a s imi lar set of data for the epithermal neutron capture gamma rays [8] . Recent data indicate fairly important changes in gamma spectrum with the energy of the neutron being captured [9] .

A large part of this problem i s the accurate determination of the c r o s s -sect ions involved. Here c r o s s - s e c t i o n s become rather smal l above 10 keV and a r e of ten neg lec ted , but in non-hydrogenous s h i e l d s the g a m m a r a y s

Presented by M. Grotenhuis.

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from neutron absorptions at these energies assume a greater importance and may even dominate. At the international conference [5J mentioned above, a s e s s i o n was a l so devoted to c r o s s - s e c t i o n s . This report, the latest report to the AEC Nuclear Cross Sections Advisory Group [10] and the latest report f rom the Brookhaven Sigma Center [11] r epresen t the mos t recent c o m p i -lat ions of l i terature on c r o s s - s e c t i o n s . A recent Euratom report [12] and the l a t e s t Nuc lear C r o s s Sect ion Advi sory Group compi la t ion [13] de f ines the situation in regard to requests for future measurements of c r o s s - s e c t i o n s . Compilations of useful c r o s s - s e c t i o n s ' s e t s appear regularly in the l iterature [14, 15 ] , each se t r e f l e c t i n g the i m p r o v e m e n t s up to that date. A c r o s s -sect ion set for fast r eac tors i s present ly in preparation at Argonne. It wil l be presented at the 1964 Geneva Conference.

The situation in regard to the gamma rays from the inelast ic scat ter of neutrons i s , of c o u r s e , much m o r e c o m p l e x . It i s a l s o m o r e d i f f icu l t to make the required m e a s u r e m e n t s . Continuing work [ 1 6 - 1 9 ] i s gradual ly increas ing the s tore of available data [14, 15, 20] and, although it wil l never be a s s imple a s the data for thermal neutrons , i t i s gradually approaching the min imum n e c e s s a r y amount. To date l i t t le has been done to m e a s u r e or calculate the angular distribution of the inelastic neutrons or gamma rays, although recent data [21 ,22 ] indicate that s ign i f i cant a n i s o t r o p i e s e x i s t .

3. METHODS FOR CALCULATING RADIATION ATTENUATION

The impact of h i g h - s p e e d comput ing m a c h i n e s h a s m a d e a profound change in the computation of radiation attenuation. There are many com-p l e x i t i e s in the format ion, computation and interpretat ion of the p r o b l e m . T h e s e a r i s e f r o m the c h a r a c t e r i s t i c s of the mach ine , the m a t h e m a t i c a l methods and/or the attenuation methods. The result i s that the computation of radiation attenuation s t i l l requires a highly competent individual to apply al l the skills' proper ly . Thus, the sophist icated methods are avai lable but not in the form of an engineer ing handbook.

The many methods of calculating the attenuation of neutrons and gamma r a y s have been adequate ly d i s c u s s e d [3, 23] and, t h e r e f o r e , only r e c e n t d e v e l o p m e n t s wi l l be ment ioned h e r e .

An interesting comparison of neutron attenuation methods [24] was made at the 1963 meeting of the American Nuclear Society Shielding Division. The r e s u l t s a r e expec ted to be publ i shed to enable m o r e deta i l ed c o m p a r i s o n before the annual meet ing in June 1964, and until then it i s difficult to draw detailed conclusions. Ingeneral , the "exact" methods, Moments, Monte Carlo, NIOBE, Transmiss ion Matrix, agreed fairly well, though nc; quite as well as expected . The combination remova l -d i f fus ion theory methods appeared to be about as expected, a reasonable overest imate . Diffusion theory solutions w e r e included for comparat ive purposes and these solut ions had fluxes that w e r e low compared to the o thers at deep penetrat ion. The r e s u l t s f o r al l methods were bracketed by a factor of ten to one hundred at deep penetration except for one problem with a LiH shield where the range of values varied by 105 . Since the la t ter problem i s not fundamental ly m o r e diff icult than those involving water shie lds and has in fact been accurately solved by point kerne l , Monte Carlo and NIOBE p r o g r a m m e s , i t i s l ike ly that s i m p l e

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numerical e r r o r s and di f ferences in normalization were responsible for the large d i screpanc ies .

The technique for solving neutron and gamma-ray attenuation problems which has expanded in importance m o s t rapidly in the recent past i s that of Monte Carlo, or the method of random sampling. In the recent past it has enjoyed an advantage in prob lems with re la t ive ly few m e a n - f r e e - p a t h s but where geometry or scattering i s too complex for other methods. Examples of such problems are slant penetration, duct and void penetration, stratif ied s lab conf igurat ions and a i r s c a t t e r i n g . More r e c e n t l y , f a s t e r m a c h i n e s coupled with modi f i ca t ions such a s spl i t t ing, b i a s e d sampl ing , e t c . have made Monte Carlo a practical tool for deeper penetration problems although it i s not a s imple one to operate properly and interpret. Varying confidence i s d i sp layed in the u s e of Monte Carlo . Some organizat ions depend on it for al l prob lems and have both a trust in i t s capabi l i t ies and a re spec t for the a s s o c i a t e d p i t fa l l s . Others r e g a r d the mod i f i ca t i ons with s u s p i c i o n although s t i l l employing it a s adequate for the l imi ted accuracy required . Others do not use it at a l l because they e i ther cannot afford to develop the techniques or the capabil i ty to u s e the techniques , or e l s e are working on other techniques which appear promising. It would appear that Monte Carlo i s potentially the ultimate technique for all shielding computations. Certainly, it would sat isfy the requirements for the more demanding gamma-ray problems, such a s those of heat generation. Excel lent Monte-Carlo gamma-transport p r o g r a m m e s ex i s t l ike Tr igr P for s labs [25] , SAGE [26] for spheres and ADONIS [27] for combinat ions of b o x e s , s p h e r e s , c y l i n d e r s and w e d g e s . The economics of the situation i s likely to make the changeover much slower than the advance in technology would allow and many of the other methods, both good and bad, will pers i s t for some t ime. In many instances, however, the economy of an approximate technique i s i l lusory and accurate a n s w e r s are avai lable at about the s a m e c o s t .

The greater energy range and larger uncertainty of the c r o s s - s e c t i o n s make the neutron attenuation resul t s l e s s rel iable than those of gamma-ray attenuation, although in principle the Monte-Carlo procedures apply equally wel l to either. Among the instal lat ions making common use of Monte Carlo are General Dynamics , For t Worth [28, 2 9 ] , United Nuclear Corporation [26, 2 7 ] , Genera l E l e c t r i c N u c l e a r M a t e r i a l s and P r o p u l s i o n Operat ion [30 , 31 ] , Oak Ridge National Laboratory [32] , the B a l l i s t i c R e s e a r c h Laboratory and TRG Incorporated [25] . It will be noticed that in the above r e f e r e n c e s , which a r e only a s m a l l s a m p l e of the l i t e r a t u r e , t h e r e a r e Monte-Carlo p r o g r a m m e s to so lve virtual ly any of the shielding prob lems . A note of caution should be i n s e r t e d a f t er t h e s e r e f e r e n c e s . Due to the rapidly changing nature of the Monte-Carlo procedure, it would be advisable to contact responsible individuals before attempting to adopt any of the above procedures .

A method which has had a very strong influence in reactor shield design a s we l l a s in other a r e a s of shie lding [33] i s the moments method. It has been the source of gamma-ray build-up factors , differential energy spectra and fast-neutron attenuation curves . This data has served to make the rela-t ively s imple point kernel technique one of the most used techniques in the past. In spite of the serious limitation, that it applies only to a single homo-geneous region, the data has been and st i l l i s useful [32, 34-39] in practical

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shield design problems. Recent efforts have included attempts to expand and check the original gamma-ray data and to fit additional data to the exponential form [40] . Of course , the above methods do not employ the moments tech-nique d irec t ly . They employ point kerne l t echniques us ing the b a s i c data from m o m e n t s ca lcu la t ions , such a s the g a m m a - r a y bui ld-up f a c t o r s and d i f f erent ia l energy s p e c t r a , and the neutron d o s e d is tr ibut ion. The un-certa int ies in this procedure are all based on the fact that the moments data are l imited to infinite homogeneous media and pract ica l sh ie lds have more than one f inite reg ion . Any of the ways of e s t imat ing the e f f e c t s of u s ing infinite medium data on finite shield regions, Monte Carlo near boundaries, equivalent material , e t c . , are usable in some c ircumstances , but very un-s a t i s f a c t o r y in o t h e r s . While th i s procedure has been the w o r k h o r s e of shielding work in the past, most workers are looking for a way to calculate the re su l t s for laminated shie lds directly.

In t e r m s of u s e , if not importance , the next method d i s c u s s e d should be that s o m e t i m e s ca l l ed the "combination" method. The "combinat ion" r e f e r r e d to i s that of remova l theory calculat ions of h igh-energy neutrons by the point kernel technique combined with diffusion theory calculat ions of intermediate and thermal energy neutrons. The removal theory calculation i s p r i m a r i l y intended for the ca lculat ion of neutron d o s e s in hydrogenous media but, in this c a s e , a l s o s e r v e s as a source t erm for the intermediate and thermal energy neutrons that penetrate deeply in the shield. Although of doubtful accuracy, it i s general ly considered a reasonable overest imate for capture g a m m a - r a y s o u r c e s . It has found great u s e , perhaps m o r e than warranted, primarily because it was simple, it was available and comparison with experiments showed it to be in reasonably good agreement [41] . Recent appl icat ions [42, 43] have tended to ut i l ize the energy-dependent r e m o v a l c r o s s - s e c t i o n s with good resul ts in thick iron slabs as well as concrete. The "two-component" method [44] has many s i m i l a r i t i e s to the combinat ion method in that they each have a direct component and a diffuse component.

Two approaches for solving the problem of radiation transport through matter in a re lat ive ly elegant fashion, but which are not in current use for design work, are the numerica l integration of the Boltzmann equation (NIOBE) [35, 45] and the t r a n s m i s s i o n matr ix [46] . Both are applicable to neutron attenuation problems in non-hydrogenous mater ia l s , although neither have been applied to any great extent. NIOBE has a current disadvantage in being re s t r i c t ed to large core reg ions so that the angular separat ions are not too s m a l l . Angular flux at a l l e n e r g i e s at the outs ide of the shie ld i s w e l l r e -presented by the Legendre polynomial expansion used. (The programme has provis ion for handling expansion up to P3 2 . ) The programme has been used for design studies in connection with mobile reactors . Its current applications are l imited to the evaluation of experiments. The transmiss ion matrix tech-nique has been t e s t e d for g a m m a - r a y penetrat ion in s lab g e o m e t r i e s . It should be useful for both gamma-ray and neutron problems in slabs, spheres and c y l i n d e r s . No appl icat ions of the method a r e current ly be ing m a d e .

Workers who must ca lculate neutron attenuation for a pract ical shield des ign a r e in somewhat of a quandary. There i s s t i l l s o m e r e s e r v e o v e r the Monte-Carlo technique; the moments method has its advantages but these are lost in problems involving multi layer shields; NIOBE and transmiss ion matrix are not yet perfected. As a result, the combination removal-diffusion,

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or two-component method wi l l s t i l l find favour for neutron attenuation in hydrogenous m a t e r i a l s . A method which appears to be gaining in favour, part icular ly for appl icat ion to meta l -hydrogen sh ie ld s y s t e m s , and which appears to e l iminate the empir i ca l nature of the combination remova l -diffusion i s the one -d imens iona l multigroup Pt approach. The Pj MG had rather ser ious l imitat ions, but the P3MG [47, 48] predicts thermal neutron •flux and fast neutron spectra in good agreement with experiments and other calculations. No correct ion i s needed for poor penetration, but a geometri-cal factor i s n e c e s s a r y to convert the one-d imens iona l re su l t s to those ap-pl icable to more rea l i s t i c source distributions.

Calculat ions of neutron dis tr ibut ions in non-hydrogenous s h i e l d s a r e genera l l y done with Monte -Car lo or with point k e r n e l p r o g r a m m e s us ing s p e c i a l k e r n e l s evaluated f rom moments ca lcu la t ions . Probably the b e s t studied non-hydrogenou s medium i s air , for which numerous Monte-Carlo ca lculat ions are avai lable [49] .

G a m m a - r a y attenuation problems including heat generat ion have t r a -ditionally been solved by point kernel techniques which employ the data from the inf ini te homogeneous reg ion m o m e n t s ca lcu la t ions in an e m p i r i c a l fashion. It i s probably just a question of t ime before this procedure can be .discarded. The feas ib i l i ty of using the spher ica l harmonics technique has been invest igated [50] . It i s found to agree wel l with moments ' resu l t s for situations where it may be compared and has been extended to y ie ld energy depos i t ion and s p e c t r a in m u l t i - s l a b s h i e l d s . D i f fus ion theory [51] i s a p r o m i s i n g technique f o r g a m m a - r a y heat ing i n s i d e the s o u r c e r e g i o n .

Neutron and g a m m a - r a y s t r e a m i n g through ducts i s be ing computed pr imar i ly by Monte-Carlo methods, although there wi l l be occas iona l u s e of analyt ical f o r m s employing the albedo concept . This latter concept has not been uti l ized to a great extent in reac tor shielding prob lems . It would s e e m that the t ime has come for the application of Monte-Carlo methods to a family of ducts which cover the range of duct problems occurring in reactor shield design. The proper choice of problems and a neat system of presenting the resu l t s could make the information suitable for an engineering handbook. This would not be s imple but it i s possible and it would be useful .

A subject which has b e e n a re la t ive s t r a n g e r to the f i e l d of r e a c t o r shielding i s that of optimization or the shield synthes i s technique. Limited r e f e r e n c e i s found in the l i t e ra ture of the past , but r e c e n t l y work on the opt imizat ion of s h i e l d s with r e s p e c t to weight has been reported [52] and consideration of the shield in an integral part of the reactor design has been d i s c u s s e d [53] . The minimum weight shield synthes i s procedure has been coded for use in shield design work under the names GOBLIN and DWARF. It i s be l i eved that th i s approach to reac tor sh ie lds wi l l b e c o m e e x t r e m e l y important in the future, e spec ia l l y in connection with the des ign of mobi le reac tor s y s t e m s .

4. MEASUREMENTS OF SHIELD PERFORMANCE

A s i s apparent f r o m the d i s c u s s i o n on methods of ca l cu la t ions , the necess i ty for measuring removal c r o s s - s e c t i o n s has decreased to the extent that the Oak Ridge l id tank i s not a n e c e s s a r y tool f o r sh ie ld d e s i g n any

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longer. Even the energy-dependent removal c r o s s - s e c t i o n s can be accumu-lated by other means , m e a s u r e d on a c c e l e r a t o r fac i l i t i e s or calculated, if ex tens ions of the present v a l u e s are deemed n e c e s s a r y . A major fract ion of the data required for shie ld ana lys i s work wi l l be done by exper imenta l mach ines which are not cons idered to be pr imar i ly exper imenta l shielding fac i l i t i e s . Capture and inelast ic c r o s s - s e c t i o n measurements , for example, do not occur at what are cons idered shielding f a c i l i t i e s per s e . Thus, the greates t use for shield test fac i l i t i e s in the future, and these wil l not al l be r e a c t o r s , wi l l m o r e l ikely be to m e a s u r e fundamental quantit ies . In s o m e c a s e s , it wi l l be n e c e s s a r y to study the attenuating propert ies of shie lds or portions of sh ie lds which are very complex and which have very demanding speci f icat ions . Mobile reactors , particularly for space application, are the present example of such experimental work.

At the Oak Ridge Bulk Shielding Faci l i ty and Tower Shielding Faci l i ty , the exper imenta l e f for t s are devoted to such work as neutron spec tra l m e a s u r e m e n t s , g a m m a - r a y spec tra l m e a s u r e m e n t s , neutron transport in water and comparison of neutron dose measurements with calculations [54] . The re su l t s of neutron scat ter ing exper iments on the Oak Ridge Tower fa-ci l i ty and Monte-Carlo calculations [55] for SNAP reactor sys tems indicate good agreement between theory and experiment.

The Aerospace Shield Tes t Reactor (ASTR) [56] i s capable of studying the shie lding p r o p e r t i e s of s lab sh i e lds a s a function óf angle of inc idence and angle of emergence , a ir and ground scattering phenomena, and heating m e a s u r e m e n t s as wel l as the usual bulk shield exper iments .

The Hanford shield test facility [57] has been utilized to measure attenu-ation of neutrons and g a m m a rays in concre te , in part icular the e f fec t that heating the concrete has on the fast -neutron attenuation charac ter i s t i c s . A fairly representat ive spectrum of concre tes has been studied, but there are s t i l l s o m e ques t ions remain ing in regard to the attenuation of neutrons in concre te . The most recent exper imenta l r e su l t s f rom the Hanford faci l i ty [57] w e r e compared to a combination remova l diffusion theory [43] ca lcu-lat ion with good a g r e e m e n t . The s a m e kind of ca lculat ion a l s o compared favourably with measurements in thick iron s labs [42] . This faci l i ty i s not now in operation but probably wil l be available for future use .

Another shield test fac i l i ty in current demand i s the SNAP shield test facil ity [58] . This, of course , resul ts from the des ire to put nuclear power plants into s p a c e v e h i c l e s and, there fore , has the weight -sav ing r e q u i r e -m e n t s of a mobi l e projec t .

A faci l i ty to measure secondary g a m m a - r a y production has been set up at United Nuc lear Corporation [8] . It i s a re la t ive ly s imple arrangement containing a P o - B e source of neutrons. Measurements have been compared to Monte -Car lo ca lcu lat ions with the ADONIS code [27] and they a g r e e to within 30 per cent for the iron s a m p l e s .

United Nuclear Corporation has plans completed for a shie ld mock-up reactor which will be used to evaluate shield mater ia ls , their most effect ive distribution around the r e a c t o r and the e f fect of -various des ign changes on the shielding and on the nuclear c h a r a c t e r i s t i c s of the reac tor .

103

5. SHIELDING MATERIAL

An advancing technology usual ly r e q u i r e s a great e f fort to expand the front iers of mater ia l s research . As reactors came into being,materials had to be found to sa t i s fy the new shielding requirements . This was a particu-l a r l y c r i t i c a l p r o b l e m for mob i l e r e a c t o r s . Consequent ly , t h e r e w a s a great deal of r e s e a r c h on the m o r e e f f i c i ent sh ie ld m a t e r i a l s , which a l s o oftën turned out to be the most expensive. Most of these studies have ceased o r at l e a s t have d e c r e a s e d in urgency . T h e r e was . c o r r e s p o n d i n g l y l e s s r e s e a r c h on the m a t e r i a l s for stationary plants . Al l the work done in the pas t i s brought toge ther in the r e v i s e d R e a c t o r Handbook [59] . Another summary of use to engineers i s that of KOMAROVSKII [60] . A compilation of papers on concrete i s a l so available [61] .

Current re search on mater ia l s for shielding reactors has been at a much reduced pace . A m a t e r i a l s d iv i s ion of the A m e r i c a n Nuc lear Society has been organized and at the November 1963 meeting [62] there were a number of s e s s i o n s devoted to reactor materia ls . These will s erve as an indication of present act iv i t ies in mater ia l s research .

The cho ice of m a t e r i a l s f o r a nuclear r e a c t o r shie ld i s quite s trongly a function of the type of project , i . e . mobile or stationary. The mater ia l s used for stationary shie lds have stabilized for the most part to water, steel , concre te , l ead and a boron-conta in ing m a t e r i a l . The m a t e r i a l c h o i c e s remaining to the shield engineer are the aggregate for the concrete, the type of s tee l , whether to include lead and the s e l e c t i o n of the borated mater ia l , if any. The cho ice of aggregate used, if other than ordinary, i s guided so strongly by economic cons iderat ions that the nearness of the supply wi l l be a strong factor. However, ferrophosphorus aggregate in quantities as large a s 10 000 t has been found economical ly impract ica l in a speci f ic instance at Argonne relative to the more common forms of heavy aggregates of magnetics and barytes , even if the latter must be shipped over long distances.

R e s e a r c h e f for t s on borated m a t e r i a l s for stationary plants have been quite l imi ted in the recent past . The only s ignif icant r e s u l t s reported [63] are for borated mater ia l s to be used in the Fermi reactor shield. This work a l so included the development of high-temperature concrete. Such materials a s borated s t e e l are now re la t ive ly easy to fabricate but s t i l l c o s t a great deal more than ordinary steel , as much as ten t imes more in some instances.

There has been a great deal of r e s e a r c h done on m a t e r i a l s for mobi le r e a c t o r sh ie lds [67] and s o m e of this i s continuing. T h e s e sh ie lds a r e g e n e r a l l y made of m o r e exot i c m a t e r i a l s [ 6 5 - 6 7 ] that conta in hydrogen , boron, l i thium, e tc .

An important cons iderat ion when m a t e r i a l s are be ing s e l e c t e d i s that of radiat ion e f f e c t s . A g e n e r a l s u m m a r y i s g iven in the r e v i s e d R e a c t o r Handbook [68, 69] along with another chapter [70] on spec i f i c information. Recent efforts have been directed toward co-ordinating procedures and infor-mation and in trying to understand the m e c h a n i s m involved [71] . Much of the past i s u s e l e s s because'of a lack of understanding and poor co-ordination of effort. The most recent reports on the subject were given at the American Nuclear Society Meeting in New York [72] .

104

Among the problems of the future i s radiation damage to s ta inless s tee ls at elevated temperatures . The temperature range 350 to 650°C is of growing importance and is not known well enough for present day applications.

A C K N O W L E D G E M E N T S

Information has b e e n included in th i s paper a s a d i rec t re su l t of a s -s i s tance from K. Shure, Westinghouse Electr ic Corporation, Bett is Atomic Power Laboratory, Pittsburgh, Pennsylvania; Frank Allen, B a l l i s t i c s Re-s e a r c h L a b o r a t o r i e s , A b e r d e e n Prov ing Ground, Maryland; E . C . Kidd, General Dynamics Corporation, Ft. Worth, Texas; W.E. Edwards, Nuclear S y s t e m s Eng ineer ing Operat ion, Cincinnati , Ohio; A . E . McArthy and H . F . Reed, Argonne Nat ional Laboratory , Argonne, I l l ino i s .

R E F E R E N C E S

[1] PERKINS, J. F . , Decay of U235 Fission Products, Rpt AD-415052; UC-34 (25 July 1963). [2] UNITED STATES ATOMIC ENERGY COMMISSION, Reactor Physics Constants. Rep ANL-5800 (2nd. ed. )

(July 1963). [31 "Radiation Shielding". Reactor Handbook, III B, (BLIZARD, E. P. and ABBOTT, L. S . , Eds. ) Interscience,

New York (1962). [4] AMIEL, S. and GILAT, J . , Reactions 0 I 7 ( n , p) N17 and O l 8 (n , d) N " with Reactor Neutrons, Nucl. Sei.

Engng 18 1 (1964) 105. [5] THROW, F.E. et a l . , International Conference on Nuclear Physics with Reactor Neutrons, Rpt ANL-6797

(15-17 Oct. 1963). [6] WALKER, R. L. and GROTENHUIS, M, A Summary of Shielding Constants for Concrete, Rpt ANL-6443

(Nov. 1961). [7] TROUBETZKOY, E. and GOLDSTEIN, H. , A Compilation of Information on Gamma-Ray Spectra Resulting

from Thermal Neutron Capture, Rpt ORNL-2904 (1961); Nucleonics 18 11 (1960) 171. [8] SCHAMBERGER, R. et a l . , Experimental Measurements of Secondary-Gamma-Ray Production in Shielding

Materials, Trans. Amer. Nucl. Soc. 6 2 (1963) 48. [9] BOLLINGER, L. M. et al. , Fluctuations in Partial Radiation Widths, Phys. Rev. Lett. 3 8 (1959) 376.

[10] SMITH, A. B. , Report to the Nuclear Cross Sections Group, Rpt WASH-1046 (1964). [11] GOLDBERG, M. D. et al. , Angular Distributions in Neutron Induced Reactions, Rpt BNL-400 (Oct. 1962). [12] SPAEPEN, J. , Compilat ion of Requests for Nuclear Cross Section Measurements from Euratom, Rpt

EANDC-E-43-L; UC-34 (Jan. 1963). [13] SMITH, A. B. , Compilation of Requests for Nuclear Cross Section Measurements, to be published as an

APR WASH document (1964). [14] CONNOLLY, L. D. et a l . , Los Alamos Group-Averaged Cross Sections, Rpt LASM-2941 (July 1963). [15] SCHMIDT, J. J . , Neutron Cross Sections for Fast Reactor Materials, Part II, Tables, Rpt KFK-120;

EANDC-E-35-U (Dec. 1962). [16] TROUBETZKOY, E. S. , Continue Theory of G a m m a Ray Spectra Following Inelast ic Scat ter ing, Rpt

NDA 2111-3, Vol. В (1 Nov. 1959). [17] TROUBETZKOY, E. S. , Fast Neutron Cross Sections of Iron, Silicon, Aluminum, and Oxygen, Rpt

NDA 2111-3, Vol. С (1 Nov. 1959). [18] TROUBETZKOY, E. S. et al. , Fast Neutron Cross Sections of Manganese, Calcium, Sulfur and Sodium

(Final Report), Rpt NDA 2133-4 (31 Jan. 1961). [19] TRALLI, N. et al. , Neutron Cross Sections for Titanium, Potassium, Magnesium, Nitrogen, Aluminium,

Silicon, Oxygen and Manganese, Rpt UN С 5002 (Jan. 1962). [20] GOLDBERG, M. D. et al. .. Angular Distribution in Neutron Induced Reactions. Rot BNL-400 (Oct. 1962). [21] CRANBERG, L. et a l . , Identif icat ion of Zero-Spin States by Inelastic Neutron Scattering, Phys. Rev.

Lett. 11 7 (1963) 341. [22] BORNING, J. W. and McELLISTREM, M. T. , Differential Cross Sections for (n,n\y) Reactions in Several

Nuclei, Phys. Rev. ^ 2 4 (1961) 1531.

105

GOLDSTEIN. H . . Fundamental Asnects of Reactor Shielding. Addison Weslev Pijbl. Co. (Feb. 1959). FODERARO, A . , Neutron Attenuation in Optically Thick Shields, Trans. Amer. Nucl. Soc. 6 1 (1963) 439. STEENBERG, H . , TR1GR-S and TRIGR-P, Nucl. Sei. Engng 12 (1962) 554. GUBER, W. and SHAPIRO, M . , A Description of the SANE and SAGE Programs, Rpt UNUCOR-633 (Mar. 1963). EISENMAN, B. and HENNESSY, E. , ADONIS and IBM Monte Carlo Shielding Code, Rpt UNUCOR-635 (Mar. 1963). COLLINS. D. G. , A Monte Carlo Multibend Duct Procedure, GD/FW Rpt MR-N-286 (Sep. 1962). COLLINS, D. G. , A Monte Carlo Procedure for Calculat ing Penetration of Neutrons through Straight Cylindrical Ducts, Rpt NARF-61-33T (24 Nov. 1961).

LOECHLER, J .J . and MacDONALD, J. E. , Flexible Monte Carlo Programs FMC-N and FMC-G. Rpt APEX-706 (July 1961). MacDONALD, J. E. et a l . , Specialized Reactor-Shield Monte Carlo Program 18-0, Rpt XDC 61-1-91; TID-11681 (26 Jan. 1961), Neutron Physics Division Annual Progress Report for Period Ending September 1, 1962, Rpt ORNL-3360 (Jan. 1963) 90. SPENCER, L. V. , Structure Shielding against Fallout Radiation from Nuclear Weapons, Rpt NSB-Mon-42 (1 June 1962). PETERSON, D. C. , Shield Penetration Programs C-17 and L-63, Rpt NARF-61-39T; FZK-9-170 (29 Dec. 1961). Neutron Physics Division Annual Progress Report for Period Ending September 1, 1962, Rpt ORNL-3360 (Jan. 1963) 94. CERTAINE, J. E. et a l . , 9-RENUPAK, Nucl. Sei. Engng 12_ (1962) 446. SHURE, K. , SPAN-3, personal communication. DUNCAN, D. S. and SPEIR, A. B. , GRACE I - An IBM 704-709 Program Designed for Computing Gamma-Ray Attenuation and Heating in Reactor Shields, Rpt NAA-SR-3719 (30 June 1959). DUNCAN, D. S. and SPEIR, A. B., GRACE II - An IBM 709 Program for Computing Gamma-Ray Attenu-ation and Heating in Cylindrical and Spherical Geometries, Rpt NAA-SR-Memo-4649 (12 Nov. 1959). STROBEL, G. L. , Additional Exponential Representations of Gamma-Ray Buildup Factors, Nucl. Sei. Engng 11 4 (1961) 450. AVERY, A. F. et al. , Methods of Calculation for Use in the Design of Shields for Power Reactors, Rpt AERE-R-3216 (Feb. 1960). PETERSON, E. G . , Comparison of Calculations with Experimental Data from the Hanford Shield Facility, Trans. Amer. Nucl. Soc. 4 2 (1961) 265. PETERSON, E. G . , MAC-A Bulk Shielding Code, Rpt HW-73381, (Apr. 1962). MOTEFF, J. and OSGOOD, H. W . , Two-Component Method of Neutron Shield Analysis for Space Ap-plication, Trans. Amer. Nucl. Soc. , 5 2 (1962) 403. PREISER, S . , 9-NIOBE (UNC-90-2), Nucl. Sei. Engng 12 3 (1962) 447. YARMUSH, D. et a l . , The Transmission Matrix Method for Penetration Problems, Trans. Amer. Nucl. Soc. _3 2 (1960) 352. SHURE, K. , P-3 Multigroup Calculations of Neutron Attenuation, Trans. Amer. Nucl. Soc. 6 1 (1963) 190. SHURE, K . , P-3 Multigroup Calculations of Neutron Attenuation, to be published in Nucl. Sei. Engng. PENNY, S. K. et a l . , Cumulat ive Bibliography of Literature Examined by the Radiation Shielding In-formation Center, Rpt ORNL-RSIC-2 (Sep. 1963) 63. LANNING, W. D. , Application of the Spherical Harmonics Technique to Problems in Gamma Transport, Nucl. Sei. Engng 15 (1963) 259-67. GREENHOW, C. R. et al. , Use of Diffusion Theory Codes to Predict Gamma Heating, Rpt KAPL-M-DNA-5 (1 Jan. 1963). TROUBETZKOY, E. S. , Minimum Weight Shield Synthesis, Rpt UNC-5017 (Part A) (15 Oct. 1962). PAWLICKI, S .S. , Optimum reactor shielding, Paper 19, Int. Cont. Canad. Nucl. Assoc. (27-29Mayl963). BOKHARI, M. S. et a l . , Neutron Physics Division Annual Report for Period Ending August 1963, Rpt ORNL-3499 1 (Dec. 1963) 98.

[55] MUCKENTHALER, F. J, et a l , , Measurements of scattered neutron intensities from cylinders of C, Al, Fe and Be in a col l imated beam of neutrons from the TSR-П, Trans. Amer. Nucl. Soc. 6 2 (1963) 424;

106

KAM, F. В. К . , HUBNER, R. S . , CLARK, F. H. and LaTORRE, J. G. , Monte Carlo calculations of neutron scattering from cylinders of Be, C, Al and Fe, Trans. Amer. Nucl. Soc. 6 2 (1963) 424.

[56] GABRO, A. N. and MOONEY, L. G . , ASTR fast-neutron and g a m m a - r a y spectral measurements , Rpt NARF-6J-21 T ; MR-N-262 (15 Jan. 1961).

[57] PETERSON, E . G . , Shielding Properties of Iron Serpentine Concrete , HW-73255 (4 Apr. 1962). [58] TOMLINSON, R. L. e t a l . , SNAP Shield Test Experiment Reactor Physics Tests, Rpt NAA-SR-7368

(15 July 1962). [59] CALKINS, V. P. , Shielding Materials, Reactor Handbook 1 G (TIPTON, C. R., Ed. ) Interscience (1960)

1023. [60] KOMAROVSKII, A . , Construction of Nuclear Installations, Rpt AEC-TR-5640. [61] AMERICAN CONCRETE INSTITUTE, Concrete for Radiation Shielding, Compila t ion No. 1, 2nd. ed.

(1962). [62] Trans. Amer. Nucl. Soc. 6_ 2 (1963). [63] HUNGERFORD, H. E. , New Shielding Materials for High Temperature Application, Nucl. Sei. Engng 6

(1959) 396. [64] CALKINS, V. P. and BLIZARD, E. P. Relationship of Materials to Shielding, Reactor Handbook, 1

^ (TIPTON, C.R. Ed.) Interscience (1960) 1023. [65] HAMILL, C. W. et a l . . Me ta l Hydrides for Shielding Applications, Rpt Y-1366 (6 Oct. 1961). [66] MANNING, L. and McKEE, D. J . , Investigation of the Effective Thermal Conductivity of Cast Lithium

Hydride, Rpt APEX-744 (5 June 1961). [67] YANKO, W. H. and GINN, M. E. , Development of a Light-Weight Plastic Neutron Shielding Material

Based upon Boron Compounds with High Hydrogen Content, Rpt AD-267561L (5 Oct. 1961). [68] BOWEN, D. B. et a l . , Generali t ies of Radiation Damaee , Reactor Handbook 1 (1960) 40. [69] CALKINS, G. D. and SCHALL, P. , Radiation Damage - Miscellaneous Materials, Reactor Handbook 1

(1960) 74. [70] STAPP, W. J. and TETENBAUM, M . , Radiation Damage, Reactor Handbook 1 (1960) 1102. [71] ROSSIN, A. D. , Significance of the Neutron Spectrum in Radiation Effects Studies, Amer. Soc. Testing

Materials Meeting, Los Angeles, Calif. (Oct. 1962). [72] Trans. Amer. Nucl. Soc. 6 2 (Nov. 1963) 383-90.

107

SOME PROBLEMS OF BIOLOGICAL SHIELDING IN REACTORS

Yu.A. EGOROV

1. INTRODUCTION

A s wel l known, the volume and weight of the bio logical shield in most nuc lear power ins ta l la t ions are much g r e a t e r than the vo lume and weight of the reactor core . Thus, the cost of constructing the shield generally r e -p r e s e n t s a cons iderable portion of the cost of the ent ire instal lat ion. It i s therefore only natural that there should be rapid development and improve-ment of methods f o r calculat ing shie lding and the p e r f o r m a n c e of a large number of exper iments , the data from which, when used in conjunction with the resul ts of calculations, enable us to reduce the s i ze and weight of shield-ing and also its cost .

Of all the components of a nuclear power installation, the most powerful source of radiation (gammas and neutrons) i s the reac tor core . By means of core calculat ions, we can ascertain the values of the neutron and gamma flux on the core surface. These values are such that they must be attenuated 1010 - i o n t i m e s in order to ensure safe working conditions in the reactor . This attenuation f a c t o r i s n e c e s s a r y if we are to have behind the r e a c t o r shield the maximum permis s ib l e dose rate of 0.8 /urem/s (i. e. 0.1 rem per 36-h working week) which has been adopted in the USSR in accordance with the recommendat ions of the International Commiss ion on Radiological P r o -tect ion (ICRP). Another fac tor contributing to the s i z e of the d o s e behind the shield i s the strength of the dose formed by secondary gamma radiation, which deve lops in sh ie ld ing m a t e r i a l s when they a b s o r b s l o w neutrons .

2. RADIATION S P E C T R A PROM THE REACTOR CORE

ï To calculate the shield'and se lec t the proper mater ia l for i t , we must

know not only the gamma radiation and neutron f luxes on the surface of the reac tor core but a l s o their energy distributions.

The ga*mma-quanta and neutron spec tra can be studied by ca lculat ion and by direct measurement . Anticipating a bi t , we may note that the energy distribution of the neutrons is generally calculated by the multigroup method, a technique which itself requires experimental data. For this reason, direct experimental study of neutron spectra f rom the reactor core i s of great i m -portance. We know that in the case of a neutron energy of more than a few MeV, the s p e c t r u m should coincide with the f i s s i o n neutron spec trum and not be dependent on the composi t ion of the reac tor core . However, in the l ower -energy region the distribution of neutrons in the spectrum i s already dependent on the composit ion of the core and on the presence of a ref lector , a s we l l a s on the m a t e r i a l and th i cknes s of the r e f l e c t o r . S ince the c o r e always contains m a t e r i a l s causing ine las t i c scat ter ing of neutrons, the spectrum due to inelast ic scattering must become softened. However, there

108

are a lways mater ia l s in the core which cause e la s t i c scattering of neutrons and it i s therefore n e c e s s a r y to take into account the fact that the spectrum must become hardened. The resulting effect of e last ic and inelastic scat ter-ing of neutrons on the type of spec trum depends on the re la t ive amount of the one m a t e r i a l or the other in the core . A s yet , there i s no set of e x -per imenta l data on reac tor neutron spectra showing the different ra t io s of e l a s t i c a l l y and i n e l a s t i c a l l y s c a t t e r i n g m a t e r i a l s in the c o r e . H o w e v e r , s p e c t r a s t u d i e s a r e being c a r r i e d on and m a y produce s o m e data.

By m e a n s of photographic p la tes with a hydrogenous e m u l s i o n ( thick-n e s s 200 /um), m e a s u r e m e n t s have been made of a s p e c t r u m (Fig . 51) [1] e m e r g i n g f r o m the B R - 5 r e a c t o r core [2]. The s p e c t r u m was determined on the b a s i s of an ana lys i s of m o r e than 2500 t r a c k s of r e c o i l protons ad-vancing within the l imi t s of an angle of ± 15°, making due al lowance for the relat ionship crH = aH(En), the background and the f inal th ickness of the emuls ion l a y e r . A s wi l l be seen , where En > 3 MeV, the s p e c t r u m f r o m the B R - 5 reactor core coincides with the f i s s i o n neutron spectrum, but in the region of l ower e n e r g i e s s o m e softening of the spec trum i s observed , so that the neutron flux with En = 1 MeV in the measured spec trum i s about four t i m e s g r e a t e r than in the neutron f i s s i o n spec trum. Somewhat l e s s softening of the spectrum (Fig. 51) [3] i s found in measurements on a water-water reactor

i 10

5

10"'

2,

0 1 2 3 4 5 6 7 8 9 10 11 12

N E U T R O N E N E R G Y ( M e V )

Fig. 51

Fission neutron spectrum (1), spectrum of neutrons from the core of the BSR reactor (2 , <£), spectrum of neutrons from the core of the water- water reactor (1 , and the spectrum of neutrons

behind the beryl l ium layer in the same reactor (3 , • )

4 \ V л V 1 к \ \ VI о

V в \ 4 ч \ о о

\ \ \ s . > \ 4 \

109

made with a s i n g l e - c r y s t a l scinti l lation spec trometer for fast neutrons. In this exper iment , the neutron flux was directed f r o m the surface of the r e -actor core (area 500 mmX 500 mm) into the detection element of the spectro-m e t e r , which was mounted in a wide -ang le co l l imator . A layer of lead 100 m m thick was placed on the surface of the core to reduce gamma back-ground. In thé s p e c t r u m shown in F i g . 51, a, sui table correc t ion has been made for the deformat ion of the spectrum by the l a y e r of lead. The c o m -posi t ion of the reac tor core on which the m e a s u r e m e n t s were made was as fo l lows: U 0 2 28.7%, Mg 6.3%, Al 26.5%, H 2 0 38.2% (by weight).

F o r the purpose of comparison, Fig . 51 a l so shows the spectrum of fast neutrons emerging f rom the core of the BSR reactor at Oak Ridge [17] which co inc ides , within the l i m i t s of error , with the f i s s i o n neutron spectrum in the reg ion of e n e r g i e s greater than 2 to 3 MeV.

Compared with the f i s s i o n neutron spectrum (Fig. 52) [4], the spectrum of the VVR-M r e a c t o r was cons iderably a l tered in appearance by a bery l l ium ref lec tor 10 to 12 c m thick. There was a sharp decrease in the flux at neu-trons having an energy greater than 2 MeV, but the peak of the spectrum shifted into a reg ion of higher energ ies by comparison with the posi t ion of the peak in the f i s s i o n neutron spectrum. The spectrum measurements on the VVR-M reactor were performed by the method based on the use of photo-graphic p lates and a hydrogenous emulsion. A f i l t e r cons is t ing of paraffin with boron" carbide and a layer of lead 3 c m thick was used to reduce background.

0 1 2 3 4 5 6 7 8 9 10

NEUTRON ENERGY(MeV)

Fig. 52

Spectrum of fission neutrons ( 1) and neutrons from the core of the VVR- M reactor behind the layer of beryll ium (2) [4]

Simi lar m e a s u r e m e n t s were made with a sc int i l lat ion spec trometer in a reac tor with the above-ment ioned core composi t ion. In these m e a s u r e -ments , t h e r e was a l a y e r of water 2.5 c m thick between the core sur face and the bery l l ium r e f l e c t o r (8 .5 c m thick). No pronounced deformation of

Ш 0

the spectrum was observed (Fig. 51). As in the case of the spectrum of the VVR-M reactor in the 4 to 6 MeV energy interval, an insignificant increase in neutron flux i s to be observed, apparently due to the energy dependence of the bery l l ium scat ter ing c r o s s - s e c t i o n .

The gamma-radia t ion spec trum f r o m the core surface of an operating nuclear reactor can be calculated, just like the fast neutron spectrum. How-ever, to make the calculation we must have experimental data on the spectra of the prompt g a m m a radiat ion and the g a m m a irradia t ion a c c o m p a n y i n g the s ca t t er ing and capture of neutrons by c o r e m a t e r i a l s and a l s o on the spectrum of gamma radiation f r o m f i s s i o n products . In addition, we must know the absorpt ion p r o p e r t i e s of the c o r e m a t e r i a l s . Calculat ion of the gamma-radiation spectrum is very complicated and i ts accuracy can hardly be expected to be sa t i s fac tory . Highly accurate r e s u l t s can be obtained in exper iments involving the direct m e a s u r e m e n t of gamma spectra f r o m the core and these can be used in shie lding calculat ions .

s

2

ю'

s

2

10°

2

S

2

0 1 2 3 < 5 6 7 8 S

G A M M A E N E R G Y ( M e V )

Fig. 53

Spectra of g a m m a radiat ion from reactor cores

. (1) da ta from [ 6 ] (2) data from [14] (3) spectrum of prompt g a m m a rad ia t ion [ 17] (4) data f rom [ 1 5 ]

t 1 ; ¡Л w f\

—

—1 --2 - 3

\ ч\ [Л \ —4

\ \ \ v.'.

4 A

л \> 4 bL 1 л

\ \ }

L \ V

\ \ 4«

• \

ц

111

Let us now consider the data obtained f rom some experimental studies. In [6], the gamma spec trum of a w a t e r - w a t e r reac tor was m e a s u r e d by a pair scinti l lat ion spec trometer [5]. The gamma quanta were directed from the reactor core into the detection e lement of the spec trometer by way of a la tera l experimental hole and a s y s t e m of co l l imators . Since the spec tro -meter i s sl ightly sensi t ive to neutrons, additional measurements were made and the spec trum was duly corrected [7]. The re su l t s of the measurement are shown in Fig . 53 and Table IV.

The principal gamma l ines of the mater ia l s (Al, Fe, H, U) compris ing the reactor core are separated in the spectrum. The gamma l ines of Ni and Cr w e r e a l s o found, these having been formed as the resul t of the capture of neutrons by the s tee l tube of the experimental hole. The measured gamma s p e c t r u m w a s accompanied by a cons iderab le continuous d is tr ibut ion due to the gamma radiat ion which i s produced in the f i s s i o n of U 2 3 5 and to the gamma radiation which undergoes multiple scattering in the core. Reference [8] mentions the gamma spectrum of the IRT reactor [9], measured by means of a m a g n e t i c Compton s p e c t r o m e t e r [10] . The g a m m a quanta f r o m the reactor core p a s s through a 1 0 - c m layer of graphite and, a s a result of the absorpt ion there , the soft part (E y < 0.7 MeV) of the s p e c t r u m i s great ly reduced (Fig . 54). The core of the IRT reactor contains, b e s i d e s uranium,

Fig. 54

Spectrum of g a m m a - radiat ion emerg ing from the core of the IRT reactor [8] The geometry of the exper iment is shown in the upper right

water and aluminium. Al l the l ines of capture gamma radiation which are c h a r a c t e r i s t i c of t h e s e m a t e r i a l s appear in the m e a s u r e d spectrum. The maximum, gamma-quanta energy in the 7.72 MeV spectrum i s that of capture radiation f r o m aluminium. As in the case of the spectrum considered above

112

TABLE IV

GAMMA SPECTRUM OF THE RESEARCH REACTOR

No. Energy of quanta

(MeV) Element, source of gamma radiation

N0. Energy of quanta

(MeV) Element, source of gamma radiation

1 1 .65 Al2 8-> Si28, Fe J 13 5.14 AL *

2 2 .18 H(n,y)D 14 5.38 AI, Cr

3 2 . 5 3 - 15 5.59 Cr, Fe

4 3 .16 Al, Cr?, Ni ! 16 5.86 Fe, AI

5 3 .54 * Al, Fe 17 6 .2 D(n,y)T, AI

6 3 .83 Fe, AI 18 6 .35 AI

7 4. 05 U2™, Fe, Ni 19 6 .58 -

S 4 . 2 1 Fe, Ni, AI 20 6 .8 Al, Ni

9 4. 37 Fe 21 7 .05 -

10 4 .52 AI? 22 7 .4 Fe

11 4. 75 AI, Cr, Fe 23 7 .6 Fe

12 4 . 9 AI, Cr 24 7 .73 AI

(Fig. 53), the gamma l ines of the capture radiation appear against a back-ground of continuous distribution, the origin of which i s the same as in the spectrum shown in F ig . 53. A substantial contribution to . the continuous distribution a l so appears to be made by gamma quanta scattered by the gra-phite ref lector . The gamma-radiation spectrum from the TVR reactor [11], measured by a magnet ic spec trometer of the "Elotron" type [12] , r e a c h e s up to 9.33 MeV (Fig. 55) [13] . The gamma quanta of maximum energy are the capture radiation f r o m iron. In contrast to the reac tors cons idered above, the TVR core, has s t a i n l e s s s t e e l a s i t s construct ion mater ia l , and for this reason h igh-energy radiation was found in the measured spectrum. The gamma l ines which are character i s t ic of all the e lements entering into the compos i t ion of the core are separated in the spec trum. Since the e x -perimental hole in the measurement leads almost up to the fuel element, the continuous distribution in the measured spectrum is relatively small .

It i s interest ing to compare the above-ment ioned data with the r e s u l t s of a study [14], in which reference i s made to the gamma-radiation spectrum from the core of the BSR reactor, consisting of H2O, Al and U. The measure -m e n t s w e r e made with a t h r e e - c r y s t a l sc int i l lat ion s p e c t r o m e t e r . The spectrum was found to contain gamma l ines due to neutron capture by hydro,-gen (2.2 MeV) and aluminium (in the 7 to 8 MeV energy range). Since there i s no s t e e l in the r e a c t o r core , no gamma quanta having e n e r g i e s g r e a t e r than ~ 8 MeV are found in the spec trum (Fig . 53). A s in the c a s e of the spectra considered above, gamma l ines of capture radiation appear against a background of continuous distribution.

113

Fig. 55

Equipment spectrum of g a m m a - r a d i a t i o n from a la te ra l exper imenta l ho le of the TVR reactor [11]

The prompt gamma-radiat ion spectrum (Fig. 53) [17] i s softer than the spectra from the cores of operating reactors owing to the absence of capture gamma radiation.

F r o m a comparison of the gamma-radiation spectra of various reactors, we may conclude that in the medium energy region (2-5 MeV), the yield and energy of the gamma quanta are determined mainly by the gamma radiation accompanying the f i s s i o n p r o c e s s and part ly by m u l t i s c a t t e r e d gamma radiation. In the h igher energy region (5-6 MeV), however , they are d e -t ermined by capture g a m m a radiat ion f r o m the construct ion m a t e r i a l s of the core . A s a r e s u l t , the m a x i m u m e n e r g y of the g a m m a quanta in the spectrum is a l so determined by the construction mater ia l s . In reac tors where aluminium i s used a s the construction mater ia l in the core, Eymax ^ 8 MeV; in those where there i s s t e e l in the core , Ey m ax 10 MeV. The soft part of the spec trum i s due mainly to mul t i scat tered gamma radiation and i s dependent on the compos i t ion of the core .

The quest ion of the y ie ld and the spectra of gamma radiation f r o m the c o r e of a r e a c t o r which i s shut down has been adequately studied [17, 18] and the known data a r e being u s e d in ca lcu la t ions .

3. SECONDARY GAMMA RADIATION FROM NUCLEAR REACTOR SHIELDING MATERIALS

Secondary gamma radiation i s produced when neutrons pass through the shield as a result of s low neutron capture. In the case of certain mater ia l s used in the biological shielding of reactors , it i s p r e c i s e l y this capture

114

g a m m a radiation which d e t e r m i n e s the s i z e of the dose behind the sh ie ld . In the calculation of the shield, it i s therefore neces sary to know the c r o s s -sec t ion for the format ion of capture gamma radiation <та(Еп), the Ny(Ev) spectrum and i t s yield a for al l the shielding mater ia ls . Many studies have been devoted to the investigation of capture radiation and the resul ts of most of them have been genera l ized in the handbooks [18, 19] . Using these data and knowing the value <pT(x) for thermal neutron flux, we can calculate the dose produced by capture gamma radiation behind the shielding, in accordance with the formula:

Py = \tT,JJ стс. (En)n¡ a i ¥> r(x)N i(E r)exp(-M ix)B(^ ix)dE rdx, X Ey

where summing i s p e r f o r m e d over a l l the component sh ie ld ing m a t e r i a l s and integration over the entire th i ckness of the shie ld and al l the e n e r g i e s of secondary gamma radiation, (¡л i s the l inear coefficient of the attenuation of gamma radiation, B(ai¡x) i s a suitably chosen build-up coeff ic ient for scat tered gamma rays . )

The energy distribution of capture gamma radiation i s dependent on the energy of the absorbed neutrons . Thus, the spec tra of the g a m m a quanta accompanying the capture of neutrons of 0.1 MeV energy by Ni, Au and Ag nuclei are considerably sof ter than the spectra of the gamma radiation pro-duced in the capture of thermal neutrons by the same nuclei [20]. Similarly, a d i f f erence has been found in the s p e c t r a of s e c o n d a r y g a m m a radiat ion f r o m Mn, Co, F e , Ni and Cu for the capture of t h e r m a l neutrons and that of neutrons having an energy greater than ~ 1 eV [21]. The resul ts of these studies should be taken into account in calculations of the dose due to capture gamma radiation, but for the t ime being the data on the influence of neutron energy on the type of gamma s p e c t r u m are inadequate and r e s e a r c h along this l ine must be continued.

The measurement of gamma-radiat ion spectra direct ly f rom the shield i s doubtless of pract i ca l importance, the resu l t s of such exper iments p e r -mitting account to be taken of the deformation of the spectrum during passage through the shie ld and s u g g e s t i o n s to be made a s to the c o r r e c t cho i ce of shielding mater ia ls and layout. As i s wel l known, lead-water heterogeneous m i x t u r e s are often used for reac tor shie lding [22] . However, it was p r e -c i se ly data about the gamma spectrum (Fig. 56) behind a shield of water and lead which made it p o s s i b l e to e s t a b l i s h that the t h i c k n e s s of the l a y e r of lead in water should not exceed 15 to 20 cm, since an increase in thickness causes the external surface of the layer of lead to become a source of capture g a m m a radiat ion m o r e powerful than a l l the p r e c e d i n g ones* . T h i s c o n -c lus ion i s c l ear ly i l lus trated by F ig . 56a. Actually, when the th ickness of the lead layer i s increased, there i s at f i r s t a decrease in the flux of gamma quanta, but a further increase in thickness leads to an ever smal ler relative decrease in flux and at a thickness greater than ~ 15 cm there i s practical ly no change in the f lux of gamma quanta. Thus far , not many e x p e r i m e n t s

* Data of E. Voskresensky, Yu. Egorov and Yu. Orlov.

115

G A M M A E N E R G Y l M « V )

Fig. 56

Spectrum of g a m m a - r a d i a t i o n from lead-po lye thy lene shield ( lead layer 15 c m thick)

(1) spectrum behind the layer of polyethylene (2) lead in a s teel jacket (3) lead without jacket

of th i s kind have b e e n p e r f o r m e d but they .would appear to m e r i t g r e a t e r attention.

4. CHOICE OF SHIELDING MATERIALS AND LAYOUT

In choosing mater ia l s for the biological shielding of a reactor the f i rs t point to bear in mind i s that a reactor i s a source of two types of radiation, the nature of whose interaction varies from substance to substance. In most c a s e s any particular materia l wil l always be most ef fect ive for the purpose of shie lding against one type of radiation. It i s c l e a r , however , that the most economic type of shielding will be made of a materia l which attenuates neutron f luxes (at a l l energy l eve l s ) and gamma-quanta f luxes in equal d e g r e e , i . e. a m a t e r i a l for which Xn = Xv for a l l energy l e v e l s (Xnand Xy being the relaxat ion lengths of neutrons and gamma rays respect ive ly) . In t e r m s of i t s compos i t ion such a mater ia l must include the nuclei of heavy

116

T H I C K N E S S O F L E A D L A Y E R ( c m )

Fig .56a

Decrease in yield of capture g a m m a - r a d i a t i o n f rom l ead -wa te r shield when thickness of layer of lead is increased

(1) layer of lead with s teel j acke t (2) layer of lead without j acke t

e lements , which are efficient absorbers of gamma-ray quanta and have large c r o s s - s e c t i o n s for the inelastic scattering of fast neutrons, and the nuclei of l ight e l e m e n t s with a large c r o s s - s e c t i o n f o r the e l a s t i c scat ter ing of neutrons . It i s a l s o useful to include e l e m e n t s which s e r v e to prevent the o c c u r r e n c e of capture gamma radiation. T h e s e r e q u i r e m e n t s are to a c e r t a i n d e g r e e m e t by heavy c o n c r e t e s with w a t e r - and boron-conta in ing additives. The neutron and gamma relaxation lengths even in such concretes are re la t ive ly high (of the order of 10 cm [24] ) and it i s not poss ible to ob-tain shielding of s m a l l d imensions . The Xn = Xv requirement i s a l so met by heterogeneous m i x t u r e s of iron and water [24, 25] and lead and water [22] . Obviously other m a t e r i a l s or mixtures of m a t e r i a l s can be found which are equally e f fect ive a s shielding against neutrons and gamma quanta. The fact that they must a lways contain both light and heavy e l ements makes it of interest to cons ider the ratio of the contributions f rom neutrons and gamma rays to the total dose behind the shield in the event, for instance, of its being neces sary for the shield to be as light as poss ible .

Any change in the thickness (or volume density) of any particular c o m -ponent of the shie lding mater ia l that i s des igned pr imar i ly to attenuate the f lux of one type of radiat ion wi l l resu l t in a change in the d o s e due to the other type of radiation. It i s therefore poss ible to find an extreme condition in which the shielding wil l be as light as poss ib le . If a change in the weight of the sh ie ld per unit area g = pX c a u s e s an e - f o l d change in the dose f r o m one type of radiation (p being the density of the material ) , then the formula

117

determin ing the dependence of the dose on the weight of the shie ld can be wr i t ten a s f o l l o w s [26]*:

р = р 1 е х Р Г - ( ' ^ + ^ ) ] + р е х р Г - Г ^ M 1 L \ q u 4 l 2 / J 2 L Vq21 q 2 2 y j '

where p1 and p2 are the doses del ivered by the two types of radiation behind the shield for a given arbitrary ratio between the amounts of the components and X] and x? a r e the changes in weight of the components . The e x t r e m e condit ions'are represented by x-, + x 2 = 0. The problem of finding the opti-mum ratio i s solved by the Lagrange factor method and it can be shown that in the optimum case the following relationship should apply:

By way of example the c a s e of l e a d / w a t e r sh ie ld ing f o r a r e a c t o r i s examined in [26] . It i s shown that the weight of such shielding wi l l be at a min imum if the part ia l neutron dose behind the shie ld i s f i v e to s i x t i m e s l e s s than the part ia l gamma dose .

Another important factor affect ing the weight and dimensions of biolo-gical shielding i s the exis tence of gamina-ray sources inside the shield (capture radiation, radiation emitted by the components and equipment of the coolant loops, etc. ). It has been sho&rn [26] that the weight of the shield will be at a min imum when the various s o u r c e s of this kind make an equal con-tribution to the total gamma dose behind the shield*.

In the light of what has been said above, it i s according ly poss ib l e in each spec i f i c case to se lec t the mater ia l s for shielding a particular reactor and their layout in the shield in such a way that i ts weight or dimensions will be as smal l as poss ible .

5. SHIELDING CALCULATIONS

In order to calculate the biological shielding required for a reactor; it i s n e c e s s a r y to de termine the spat ia l and energy dis tr ibut ion of neutrons and g a m m a r a y s in media containing heavy and light nuc le i . The spat ia l and energy distribution of both neutrons and gamma rays are of course de-scr ibed by the Bol tzmann kinetic equation. However, the complex nature of the shielding medium makes it imposs ib le to integrate the Boltzmann equation and a number of methods have accordingly been developed which make it poss ib le by one means or another, on the bas i s of this equation, to calculate both the spatial and the energy distribution of radiation in a shield. T h e s e methods include the m o m e n t s method, the asymptot i c method, the

* For simplicity's sake the shield is assumed to comprise two components. + If the shield is in the form of a cylinder or sphere, the contributions made by the two types of

radiation to the total dose must be determined, taking into account the surface area of the source.

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polynomial method and others which are d i s c u s s e d at length in [17, 27, 28] and a l so the method for solving s ing le -ve loc i ty problems in the P 3 approxi-mat ion [29] . The f o r m e r methods entai l formidable ca lcu la t ions and a r e therefore unsuitable for pract ical shielding calculations, whereas the l a s t -named can be used only for calculating the passage of neutrons through media consist ing of heavy nuclei, s ince it d i sregards the slowing-down of neutrons a s a re su l t of e l a s t i c sca t ter ing . The random te s t method [17, 28, 30, 31] o f f ers more promis ing prospects for shielding calculations but i s at present of l i t t le use for actual ca lculat ions owing to the laborious work it invo lves and the fact that certa in exper imenta l data a r e not yet avai lable .

F o r the p u r p o s e s of prac t i ca l neutron shie lding ca lcu lat ions the m o s t fruit ful methods avai lable are the mult igroup methods in the di f fus ion and the age -d i f fus ion approximations . Although there are s o m e f a c t o r s which would appear to mil i tate against the use of these methods (e. g. the presence of thin l a y e r s of m a t e r i a l in the shield and the non- i so trop ic charac ter of the scattering) , highly sat i s factory resu l t s can be obtained by choosing suitable constants and correcting them in the light of experimental data. Na-turally the accuracy of the resu l t s depends largely on the number of groups employed, but the resul t of increas ing the number of groups i s a l s o to i n -c r e a s e the amount of calculat ions required, so that it i s more appropriate to envisage use of a method involving f ive to ten groups with constants s e -lected accordingly. Reference [32] contains an experimental confirmation of the use of 7- and 10-group methods (in diffusion and age-diffusion approxi-mat ions r e s p e c t i v e l y ) for ca lculat ing sh ie ld ing containing heavy and l ight nuc le i ( including hydrogen) , and a l s o the nucle i of s t rong ly absorbent e l e m e n t s (boron).

For calculations by the 10-group method in the energy region En > 0.7 MeV four groups were chosen, the purpose being to ensure that the cross - sec t ions for neutron scattering by nuclei of the e lements most frequently encountered in shielding (H, Pb, Fe , Cr, Ni) remain re lat ive ly constant f r o m one end of the region to the other. The lower boundary of the fas t -neutron energy region more or l e s s coincides with the energy corresponding to the threshold for inelastic scattering of these e lements . In the energy range from 0.7 MeV down to 1 keV three groups were chosen; for En < 1 keV two groups of r e -sonance neutrons and one group of thermal neutrons were chosen. The di f -fus ion equation for neutrons of group j was written in the form:

D j A V j - E j V j + S f + Sj. = o,

w h e r e <pj i s the aggregate f lux f o r neutrons of group j, Dj i s the d i f fus ion coef f ic ient , Ej i s the c r o s s - s e c t i o n for the r e m o v a l of neutrons of group j and Sf and S f j are the neutron s o u r c e s result ing f r o m slowing-down and f i s s ion respect ively . The group constants were averaged over spectra taken from [33] and [34] . ,

In calculations by the 7-group method in the energy region En > 1.5 MeV one group was chosen and the spatial distribution of neutrons in this group was determined by the removal c r o s s - s e c t i o n method. F o r En in the range between 3 keV and 1.5 MeV three groups w e r e chosen, two f o r r e s o n a n c e and one for thermal neutrons. The age -d i f fus ion equation for group j took the fol lowing form:

119

D i v V E i V - s > '

where

« i. II 1

q ( U j , r ) M j j = 2 ,

® T . e_i. II 7

Di =

l / 3 E n l

ui

ui-i

du (ÇES+TEC)

du

J ÏZs + yZc ui-l

L 1 /ЗЕцх

3 = J j I ®e'/Me

г E remova l j

_du P.(uj) ?E s + t E c Pj (u) J

-1

uj-i ;cT

j = l

2=2 6

3=7 j= 2 , . . . , 6 j = 7

j = 2, . . . , 6

3=7

Pj (u) = exp du-ui-l

? E S + T E C

Comparison of the calculated re su l t s for the spatial distribution of neutrons with values obtained exper imental ly by means of de tec tors shows that the 7 - and 10-group methods can both be used f o r ca lcu la t ing m u l t i -layer composite shielding consisting of materials with different nuclear prop-e r t i e s , or of boron-containing m a t e r i a l s . In the latter c a s e there may be an e r r o r of approx imate ly 20% owing to the fact that the p r o b l e m i s not a diffusion problem and the actual spectra diverge from the constants obtained by averaging.

. F o r calculating the spatial distribution of fast neutrons in shielding ma-t e r i a l containing hydrogen and other e l e m e n t s the r e m o v a l c r o s s - s e c t i o n method i s widely used. If we. know the value of the removal c r o s s - s e c t i o n for any g iven mater ia l £ b in a hydrogen-containing medium such a s water, we can f ind the spat ia l d i s tr ibut ion of f a s t neutrons f r o m the f o r m u l a :

N(R,T) = N H (R)exp(-E b T),

where N(R, T) i s the neutron flux at a distance R f rom the source , NH(R) i s the neutron flux in the hydrogen-containing material in the absence of a layer of removing m a t e r i a l and T i s the t h i c k n e s s of the l a y e r of r emov ing

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m a t e r i a l . A s i m i l a r f o r m u l a can be wri t ten f o r a h o m o g e n e o u s mix ture of hydrogen-containing mater ia l and any other mater ia l . F o r calculations using the r e m o v a l c r o s s - s e c t i o n method it i s a c c o r d i n g l y n e c e s s a r y to know the Spat ia l d i s t r i b u t i o n of n e u t r o n s in the h y d r o g e n - c o n t a i n i n g m a t e r i a l o r in hydrogen and the value E b . The pattern of NH (R) can a l w a y s be taken f r o m [17] or [35] and E b f r o m the r e s u l t s of e x p e r i m e n t s . Thanks to i t s s i m p l i -c i ty , a g r e a t d e a l of w o r k h a s r e c e n t l y b e e n done on the r e m o v a l c r o s s -s e c t i o n method and v a r i o u s p a p e r s have been wri t ten in which it i s ref ined, e x t e n d e d or app l i ed m o r e g e n e r a l l y . In [ 3 6 ] , f o r e x a m p l e , on the b a s i s of an a n a l y s i s of a l arge quantity of e x p e r i m e n t a l data, it i s shown that the r e m o v a l c r o s s - s e c t i o n in the e n e r g y r e g i o n f r o m 3 to 15 MeV f o r i n t e r m e -diate and heavy n u c l e i i s only s l i gh t ly e n e r g y - d e p e n d e n t and p r o v i d e d the s p e c t r u m f o r n e u t r o n s e m i t t e d f r o m the r e a c t o r in th i s e n e r g y r e g i o n i s roughly s i m i l a r to the s p e c t r u m f o r f i s s i o n neutrons , the £ b va lue a s d e -t e r m i n e d in [37] can be u s e d f o r the ca l cu la t ion . T h i s i s not h o w e v e r the case with the light nuclei Be, В and C, the remova l c r o s s - s e c t i o n s for which i n c r e a s e c o n s i d e r a b l y when E n > 3 MeV. The r e m o v a l c r o s s - s e c t i o n s f o r h o m o g e n e o u s m e d i a d i f f er by only 5 to 10% f r o m t h o s e d e t e r m i n e d f o r a l a y e r , and by u s i n g the known v a l u e s of Eb w e can c a l c u l a t e the s p a t i a l distr ibut ion of fas t neutrons in a homogeneous mix ture at any d i s tance f r o m the s o u r c e , subject to a minor e r r o r ( 2 5 - 30%) at a d i s tance of two to three neutron re laxat ion lengths f r o m the s o u r c e . It has a l s o been shown that the data contained in [37] can a l s o be used in mult igroup shie ld ing ca lcu lat ions for de termin ing the spat ia l distr ibut ion of neutrons in the uppermost energy group by a s s i g n i n g t h e m an e f f e c t i v e e n e r g y t h r e s h o l d of 3 MeV; u s e can a l s o be made of r e m o v a l c r o s s - s e c t i o n s d e t e r m i n e d a s i n v e r s e r e l a x a t i o n l engths f o r E n = 3 MeV.

Owing to i t s s i m p l i c i t y it w a s natural that a t t empts w e r e ma de to f ind out w h e t h e r it w a s p o s s i b l e to u s e the r e m o v a l c r o s s - s e c t i o n m e t h o d f o r s h i e l d i n g c a l c u l a t i o n s in c a s e s w h e r e no h y d r o g e n w a s p r e s e n t . The e x -p e r i m e n t s that h a v e b e e n c a r r i e d out have shown that t h i s m e t h o d can be u s e d in the c a s e of sh ie ld ing in which the m o d e r a t i n g e f f e c t i s provided by nuc le i h e a v i e r than hydrogen, up to a l u m i n i u m [38] .

The r e m o v a l c r o s s - s e c t i o n v a r i e s only s l ight ly with the moderator m e -d i u m that i s c h o s e n and, e v e n if t h e r e i s a d i f f e r e n c e , it d o e s not e x c e e d e x p e r i m e n t a l e r r o r . It h a s b e e n shown that the r e m o v a l c r o s s - s e c t i o n method can a l s o be used in pr inc ip le in the c a s e of v e r y heavy m o d e r a t o r s , but in th i s event £ b i s v e r y dependent on the a t o m i c number of the m e d i u m and on the e f f e c t i v e thresho ld of the de tec tor u s e d to m e a s u r e the r e m o v a l c r o s s - s e c t i o n . In mul t igroup c a l c u l a t i o n s of h y d r o g e n - f r e e s h i e l d i n g the r e m o v a l c r o s s - s e c t i o n can a l s o be used to inves t igate the uppermost energy group. F o r this purpose , however , it i s e s s e n t i a l to choose the lower boun-dary of the group in s u c h a way that Eb d o e s not depend on the c o m p o s i t i o n of the m e d i u m i m m e d i a t e l y beh ind the l a y e r of r e m o v i n g m a t e r i a l .

Until recent ly there w e r e s e r i o u s d i f f i cu l t i e s in calculat ing the p a s s a g e of g a m m a r a y s through a mul t i layer r e a c t o r shie ld; in part icular there was s t i l l doubt r e g a r d i n g the bu i ld -up c o e f f i c i e n t f o r s c a t t e r e d g a m m a r a y s in such a sh ie ld . Certa in r e c o m m e n d a t i o n s that had been ma de regarding the cho ice of the magnitude B(/jx) [17] did not a lways y ie ld s a t i s f a c t o r y r e s u l t s . E x p e r i m e n t s w e r e accord ing ly c a r r i e d out [39] which y i e l d e d an e m p i r i c a l

121

formula that can be used for calculating the build-up coef f ic ients for gamma rays of various energy l e v e l s (the re su l t s were checked experimental ly for Ey = 1 - 6.4 MeV) f o r shield th icknesses up to 15 f ree path lengths.

The values calculated according to this formula were, for example, with-in 3% of the exper imenta l r e su l t s for a l ead /water shield. The formula in quest ion i s a s fo l lows:

N n N n - l

n where N equals the total number of l ayers in the shield and Bn(E/u¡ x¡) i s the

build-up coeff icient for one of the shield mater ia l s (n'th layer), for the c o r -responding thickness (in f ree path lengths).

N u m e r o u s ca lculat ions per formed on actual sh i e lds according to th i s formula , and then ver i f i ed experimental ly , bore out the conclus ion that the r e s u l t s show good convergence [39] .

Thus we now have methods for shielding calculations which make it pos -s ible to obtain resu l t s that are perfect ly sat i s factory f rom the point of view of a c c u r a c y . F u r t h e r deve lopment of t h e s e methods wi l l probably be d irec ted mainly towards supplementing them with exper imenta l data and, m o r e e spec ia l ly , using such data in order to make them l e s s complicated.

6. MATERIALS FOR BIOLOGICAL SHIELDING AND STUDY OF THEIR PROPERTIES

In order to supplement and c lar i fy the experimental data, a number of experiments have been carried out in the USSR to investigate the passage of radiation through var ious m a t e r i a l s used in reac tor shielding [40-42] and a l s o through c o m p o s i t e sh i e lds f o r m e d of the se m a t e r i a l s . A s ear ly a s 1957 the optimum rat ios of water and iron in an iron/water shield were de-termined [24,25] and the spatial distribution of neutrons in mixtures of boron carbide with i ron and lead was studied [ 4 3 ] . It i s of c o u r s e not p o s s i b l e to g ive the r e s u l t s of a l l t h e s e e x p e r i m e n t s h e r e .

The fact that many m a t e r i a l s have a lready been used in shielding for a long t ime has not prevented their shielding propert ies f rom receiving con-tinued attention. For example, AVAEV et al. [44] report on the relaxation lengths of neutrons with an energy exceeding approximately 1.5 MeV in iron (Táble V); this information had not previously been reported in the l i terature. The m e a s u r e m e n t s w e r e done by means of an indium detector activated by the In(n, n1) Inm react ion . A s can be s e e n f r o m the Table , the re laxat ion length for neutrons with energy exceeding 1.5 MeV i s cons iderably higher than the known relaxation lengths of neutrons of higher energy [45], owing to the fact that the c r o s s - s e c t i o n for the ine las t i c scattering of neutrons by iron nuclei d e c r e a s e s a s the neutron energy d e c r e a s e s . An increase in the th i cknes s of the i ron layer r e s u l t s in an i n c r e a s e in the re laxat ion length, which can be explained by the fact that neutrons with an energy c lose to the

122

TABLE V

RELAXATION LENGTHS OF NEUTRONS WITH ENERGY EXCEEDING 1.5 MeV IN IRON

Thickness of layer (cm) 0 - 1 0 1 0 - 2 0 2 0 - 4 5 4 5 - 9 0

X (cm) 8 .35 9 .85 11 .1 13.5

energy threshold of ine las t i c scat ter ing of neutrons by iron nuclei build-up in the iron.

F o r the purpose of making a judicious choice of shielding mater ia l s and averaging the constants used in multigroup shielding calculations, it i s u s e -ful to study the way in which the spectrum of neutrons emitted by a reactor i s distorted in the event of their passage through a shield. Such m e a s u r e -ments have been made f o r graphite, iron, lead and polyethylene in [6, 4 6 ] . The r e s u l t s obtained a r e shown i n F i g s . 5 7 t o 6 0 . P a s s a g e of the neutrons through polyethylene or graphite c a u s e s the s p e c t r u m to harden: there i s a decrease in the proportion of neutrons with an energy of < 3 - 4 MeV. The neutron s p e c t r u m obtained by m e a s u r e m e n t s behind l a y e r s of polyethylene was compared with the spectrum as calculated by the moments method [17] . F o r al l th icknesses of polyethylene where E n > 3 MeV the measured and c a l -culated spectra coincide; s o m e d ivergence not iceable at l ower e n e r g i e s i s due to the d i f f e r e n c e be tween the ini t ia l s p e c t r a , but the tendency f o r the spectrum to vary at E n < 3 MeV with an increase in the thickness of the poly-ethylene l a y e r i s ident ical a s between the calculated and the exper imenta l r e s u l t s . Data obtained in the e x p e r i m e n t s with graphite w e r e a l s o c o m -pared with the resu l t s of calculations using the moments method [17]. Where the thickness of the graphite layer was 22.5 cm or 4 5 c m and where En > 5 MeV the calculated spectra coincided with the measured spectra within the l imi t s of experimental error , whereas at lower energies the f luxes in the calculated spectra are two to three t imes greater than in the spectra a s measured. The exper iment s revea led i r r e g u l a r i t i e s in the spec trum (at En = 3 - 7 MeV and with a l ayer of t h i c k n e s s e x c e e d i n g 45 cm) , due to the peaks and t roughs that occur in the energy dependence of the total c r o s s - s e c t i o n for the in ter -action of neutrons with hydrogen nuclei . These i rregular i t i e s in the spectrum appear m o r e sharp ly a s the t h i c k n e s s of the graphite l a y e r i s i n c r e a s e d .

The f o r m of the r e a c t o r f a s t neutron s p e c t r u m i s changed in quite a different way upon pass ing through l a y e r s of lead and iron (F igs . 59 and 60). At neutron e n e r g i e s g r e a t e r than about 4 MeV, the f o r m of the s p e c t r u m and i t s s lope do not change when the shielding th ickness i n c r e a s e s and r e -m a i n about the s a m e a s in the in i t ia l s p e c t r u m . In l o w e r energy r a n g e s , however, a distinct softening of the spectrum i s observed; a relative build-up of the number of 0.7 - 3 MeV neutrons o c c u r s . The change in the f o r m of the spectrum i s governed by the energy-dependence of the c r o s s - s e c t i o n f o r the i n e l a s t i c s ca t t er ing of neutrons by the iron and lead nuclei: at EH$> 3 MeV the ine las t i c scat ter ing c r o s s - s e c t i o n shows pract ica l ly no d e -pendence on energy, but at E n < 3 MeV it rapidly drops with decreasing ener-

123

/

Fig. 57

Distortion of reactor- neutron spectrum on passage through polyethylene

(1) without polyethylene (2) thickness of polyethylene layer 10 g / c m 2

(3) thickness of polyethylene layer 20 g / c m 2

(4) thickness of polyethylene layer 30 g / c m 2

(5) thickness of polyethylene layer 60 g / c m 2

(6) thickness of polyethylene layer 80 g / c m 2

124

NEUTRON ENERGY ( M e V )

Fig. 58

Distortion of reactor- neutron spectrum on passage through graphite

(1) without graphite (2) thickness of graphi te layer 2 2 . 5 cm (3) thickness of graphi te layer 45 c m (4) thickness of graphi te layer 9 2 . 5 cm (5) thickness of graphi te layer 135 cm

The dotted l ine shows the results of ca lcu la t ions by the moment s method [17] The results have been standardized for a neutron energy of 5 . 4 5 MeV and a layer thickness of 2 2 . 5 c m

125

N E U T R O N E N E R G Y I M » V )

Fig. 59

Distortion of reac tor-neut ron spectrum on passage through iron

(1) without iron (2) thickness of iron layer 1 0 . 4 cm (3) thickness of iron layer 2 0 . 4 c m (4) thickness of iron layer 40 c m (5) thickness of iron layer 65 c m

126

N E U T R O N ENERGY (MeV)

Fig. 60

Distortion of reac tor -neut ron spectrum on passage through lead

(1) without lead (2) thickness of lead layer 12 cm (3) thickness of lead layer 24 cm (4) thickness of lead layer 49 cm (5) thickness of lead layer 74 cm

gy [47] . The re su l t s obtained for iron are in good agreement with the data in [44] , r e f e r r e d to above.

On the b a s i s of the energy distr ibut ions m e a s u r e d in these m a t e r i a l s , attenuation functions were constructed for the individual neutron energy groups, and f r o m these the re laxat ion lengths and removal c r o s s - s e c t i o n s (as rec iprocals of the relaxation lengths),for neutrons in these energy groups

127

were found. The divis ion of the entire energy range under study into groups was carr i ed out in accordance with the energy ranges chosen in connection with the 10-group method of calculating shielding. The data obtained f rom th is treatment of the exper iment can be used d irec t ly for shie lding c a l c u -lations. In addition, ranges were selected for control purposes within which the fast neutron f luxes were recorded by threshold detectors . The data y ie lded by this treatment of the spec trum were in good agreement with the re laxat ion lengths quoted in the l i terature .

The shielding qual i t ies of graphite are a l so invest igated in [52] . Fas t neutron f l u x e s w e r e recorded by phosphorus and a luminium threshold d e -tectors . It was found that with a shielding thickness of l e s s than 160 cm X= 15.7 cm and that with a greater thickness the relaxation length was greater: 18. 2 cm.

As a mater ia l with a high hydrogen nucleus concentration (7 .92X10 2 2 I / c m 3

with p= 0.92 g /cm 3 ) , polyethylene is of definite in teres t for shielding purposes and a number of s tudies have been devoted to the spatial distribution of neutrons therein: thus, invest igations have been made of the attenuation of monoenergetic neutrons (En = 4 MeV and En = 14.9 MeV) [48], and of reactor core neutrons [49] . The relaxation lengths (Table VI) obtained for reactor fast neutrons are 1 2 - 1 7 % l e s s in polyethylene than in water.

TABLE VI

FAST NEUTRON RELAXATION LENGTHS IN POLYETHYLENE

Thickness of layer >7 MeV > 5 MeV >3 MeV (cm) (cm) (cm) (cm)

0 - 3 0 7 .8 7 .8 6 .6

3 0 - 6 0 9 .7 8 . 9 8 .2

6 0 - 9 0 - 9 . 5 -

The drawback of polyethylene as a shielding material l ies in its relative-ly low heat r e s i s t a n c e . A hydrogeneous mater ia l capable of being used for long per iods at a t e m p e r a t u r e of 450°С without hydrogen l o s s i s concre te prepared f rom a serpentine aggregate [50] . The water content of such con-crete (p= 2.3 t / m 3 ) attains 10 to 12% by weight and, s ince its composit ion a lso includes e lements possess ing a large neutron inelastic scattering c r o s s -sect ion (Mg, Fe) , i ts shielding qualit ies could be expected to be quite high. In fact , e x p e r i m e n t s which have been carr ied out [51] have shown that the relaxation lengths of neutrons in serpent ine-aggregate concrete (Table VII) are not more than 12 cm and the gamma-radiatiôn lengths not more than 15 cm. Studies on the deformation of the fast neutron spectrum on pass ing through the concre te (Fig . 61) indicated that the f o r m of the s p e c t r u m i s modi f i ed mainly in the 3 to 8 MeV energy range: here the spec trum i s considerably f lattened owing to i rregular i t i e s in the total macroscop ic c r o s s - s e c t i o n for the interaction of neutrons with the concrete. Just as in the case of the ma-terials investigated previously, studies of the energy distribution of neutrons beyond the l a y e r s of the serpent ine-aggregate concrete yie lded f i gures for

128

TABLE VII

-ч FAST NEUTRON RELAXATION LENGTHS IN

SERPENTINE-AGGREGATE CONCRETE

Neutron energy (MeV) > 1 . 5 > 3 > 5 > 7

X (cm) 10.9 i 0 . 3 10. 9 i 0. 3 11.2 i 0 .35 11.6 ± 0 .4

N E U T R O N E N E R G Y ( M e V ) '

Fig. 61

Deformat ion of reactor neutron spectrum on passing through serpent ine concre te

( T h e figures on the curves represent the thickness of the concre te layer in cm)

129

the relaxat ion lengths for neutrons of four energy groups se l ec ted by reference t o the 10-group method of shie ld ing calculat ion. S i m i l a r m e a s u r e m e n t s w e r e made f o r f i l l e r of serpentine sand. The bulk weight of this shielding m a t e r i a l i s 1.62 t / m 3 . Table VIII g ives the neutron re laxat ion lengths obtained in t h e s e exper iments , and Table IX shows the re laxat ion lengths in graphite, iron and other mater ia l s , as determined from data on neutron energy distribution.

TABLE VIII

FAST NEUTRON RELAXATION LENGTHS IN SERPENTINE SAND

Neutron energy (MeV) > 1 . 5 > 3 > 5 > 7

X (cm) 15.2 15.7 17 17.7

TABLE IX '

RELAXATION LENGTHS FOR FAST NEUTRONS OF FOUR ENERGY GROUPS

Energy range (MeV) 0.7 - 1 .5 1 . 5 - 2 . 5 2 . 5 - 4 4 - 10

Material Relaxation length

(cm)

Graphite 11.9 - 15 .1 11.2 - 15.2 10 .1 - 14.7 11.5 - 16.7

Serpentine concrete 11.8 11.5 10.2 10.8

Serpentine sànd ,14.9 15.2 14.9 15.7

Iron 8 . 5 7 .8 6 .8 6 . 1

Lead 15.5 12.7 10.7 9 .7

The resul t s .of only a few experiments, carried out mainly with reactors are quoted here . Numerous experiments have been performéd with neutron and g a m m a s o u r c e s having s p e c t r a d i f ferent f r o m those of a r e a c t o r ; the purpose of the experiments was to investigate both the spatial and the energy distribution of radiat ions in various mater ia l s . 'Of interes t , for example , i s that reported in [53] on the subject of the deformat ion in water of the neutron spectrum of a P o + Be source; it was shown that spectra measured under condit ions of "barr ier" and of "infinite" g e o m e t r y d i f f er somewhat and, whenthe layer of water i s increased , the spec tra l f o r m i s determined m o r e and m o r e by the energy dependence of the total i n t e r a c t i o n c r o s s -sect ion of neutrons with oxygen nuclei . Of course , the resul t s of these and other s tudies with radioact ive s o u r c e s cannot e a s i l y be applied direct ly to

130

reac tor shie lding calculat ions , but they do help to d is t inguish a number of genera l pr inc ip les character iz ing the p a s s a g e of radiation through matter .

7. REDUCTION OF CAPTURE GAMMA OUTPUT

It has already been mentioned that in many c a s e s the gamma dose rate behind the shielding, and hence the d imens ions and weight, of the shielding, are determined by the capture gamma radiation aris ing in the reactor v e s s e l and in the f i r s t l ayers of the shielding. This being so, it i s of extreme i m -portance to seek ways of reducing the capture gamma output. Secondary g a m m a radiat ion o c c u r s during the radiat ive capture of1 t h e r m a l and l o w -energy neutrons and hence absorption of these neutrons by boron cons ider -ably reduces their capture in structural materials , particularly steel , which represents a high-energy gamma-quantum source. Neutron capture by boron nuclei leads to formation of gamma quanta with ET = 0.5 MeV, against which shie lding i s easy . Boron can be introduced into the sh ie ld ing in d i f ferent ways: it can be added to the light or to the heavy component of the shielding or to s tructural m a t e r i a l s , and can a l so be d isposed in a l ayer in graphite between the light and heavy components. The effort achieved i s different in each case .

If a 20-mm thick layer of boron carbide (density 1.1 g/cm3) i s placed at the boundary between a l ayer of s tee l (St-3) and a layer of water, the 7.6 MeV gamma-quantum flux i s reduced 13.4 t i m e s [54] , and when the St-3 s t ee l i s replaced by a layer of IX18H19T s t a i n l e s s s t ee l of the s a m e th ickness , the reduction i s 7.8 t i m e s . A study has been made [55] of the effect on capture gamma output of distributing boron uniformly in s tee l , accompanied by de -terminat ion of p a r a m e t e r s a and ß, governing the output-t)f capture gamma quanta in relat ion to the neutron flux entering and leaving the shielding. . It was shown that the introduction of smal l quantities of boron into s t ee l (1.5 -3 wt. %) effect ively reduces the capture gamma output; the s ize of parameter a depends on the thickness of the s tee l pr ism: a ~ e x p ( - a / 2 6 ) for s tee l with-out boron, and a ~ exp( - a / 1 1 ) for s t e e l with 3% boron (a= th ickness of the s t e e l pr i sm); the s i z e of ß i s not dependent on the th i cknes s of the s t e e l p r i s m . When the t h i c k n e s s of the l a y e r of boron s t e e l (3% boron) i s i n -c r e a s e d , the capture g a m m a f lux s inks proport iona l ly to e x p ( - a / 2 5 ) . A water ref lector i n c r e a s e s the capture gamma-quanta flux 7 to 10 t imes , and to reduce this a layer of mater ia l containing boron must be placed between the r e f l e c t o r and the s tee l .

Of value in reducing the t h e r m a l neutron f lux, and hence the capture gamma output, i s the introduction of s m a l l amounts of boron into concrete [56] .

8. INSTRUMENTATION AND METHODS FOR STUDYING THE SHIELDING PROPERTIES OF MATERIALS

Recently, a number of new instruments and methods have been developed for making a detai led invest igat ion of the spat ial and energy dis tr ibut ions of neutrons and gamma radiation in shielding mater ia l s . Most of these

131

p o s s e s s wide sens i t iv i ty ranges and high se lect iv i ty . Scintillation counters and s p e c t r o m e t e r s a r e finding ever wider application in shie lding e x p e r i -ments . As before, radioactive detectors are continuing to be used, and the m e t h o d s of de t ermin ing the i r ac t iv i ty have b e e n c o n s i d e r a b l y p e r f e c t e d .

In [57] and [58] detai led cons iderat ion i s g iven to the f e a t u r e s of the instrumentation used in shielding experiments . The instruments concerned make it poss ib le to measure the attenuation functions of the flux of fast (ZnS(Ag) sc int i l lat ion counter, e f fec t ive threshold 3 MeV) and thermal and e p i t h e r m a l ( L i - g l a s s sc int i l la t ion counter , B F 3 - c o u n t e r , e tc . ) neutrons . The use of sc in t i l la tors and photomult ipl iers of var ious s i z e s makes it p o s s i b l e to evolve ins truments f o r exper iments under var ious g e o m e t r i e s . Re ference [59] d e s c r i b e s a gamma dos imeter "without energy dependence" for measurements of dose rate; this ratemeter uses a combined scinti l lator of organic p las t ic and a CsI(Tl) crys ta l , and can m e a s u r e dose ra te s f r o m 3X 10-9 to 3X 10-3 r / s .

F o r determining the induced act ivi ty of threshold, resonance and 1 / v detectors scintil lation spectrometers are used, the activity being determined by the gamma l ine or group of gamma l ines charac ter i s t i c of the detector concerned [60, 61 ] . This method both permi t s a relaxat ion of the require -m e n t s regarding purity of the d e t e c t o r m a t e r i a l and the background l e v e l and - this i s part icularly important - makes it pos s ib l e to use detectors s e n s i t i v e to neutrons of var ious energy groups, for the purpose of d e t e r -mining the neutron f l u x e s of one energy group against the background of an intense flux of another.

Scint i l lat ion s p e c t r o m e t e r s [62] are the i n s t r u m e n t s mainly u s e d f o r studying the energy distribution of radiation emerg ing f r o m the shie lding. F o r measuring gamma spectra, spec trometers (pair, total absorption) have been developed, and also special methods as described in [5] and [7]. Fast neutron energy distribution i s determined from the results of measurements carr i ed out by a s p e c i a l procedure [3] involving a s ingle c r y s t a l s p e c t r o -m e t e r [63] .

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[57] DULIN, V. A. e t al . , "Experimental methods in shielding research (Radiation detectors)" . Questions of reactor shielding physics, Gosatomizdat , Moscow (1963).

[58] AVAEV, V . N . e t a l . , "Counters and dosimeters for studying shielding and the shielding properties of mater ia ls" . Questions of reactor shielding physics, Gosatomizdat , Moscow (1963).

[59] EGOROV. Yu.A. and PANOV, E . A . , Prib. Teh. Eksp. 4 (1961) 57. [60] AVAEV, V.N. et a l . , "The use of radiotracers in the study of shielding", Questions of reactor shielding

physics, Gosatomizdat, Moscow (1963). [61] MASHKOVICH, V. P . , "The measurement of fast neutron fluxes in shielding against a background of

intense fluxes of intermediate neutrons". Questions of reactor shielding physics, Gosatomizdat, Moscow (1963).

[62] EGOROV, Yu .A. , The scintillation method of gamma-ray and fast-neutron spectrometry, Gosatomizdat, Moscow (1963).

[63] EGOROV, Yu.A. and PANKRATEV, Yu .V. , "A single-crystal fast-neutron spectrometer for the measure-ment of continuous spectra" , Questions of reactor shielding physics, Gosatomizdat , Moscow (1963).

134

STATUS R E P O R T - E N E A

H. В. Smets

The European Nuclear Energy Agency (ENEA) of the Organisat ion for Economic Co-operat ion and Development takes part in the work on reactor shielding by means of i t s Joint Undertakings, such as the Dragon Project at Winfrith (United Kingdom). Shielding calculations have been carried out for this high-temperature, graphite-moderated, gas -coo led reactor .

ENEA has recently created two new international centres with the par-ticipation of 11 Western European countries . One i s a l ibrary of numerical computing c o d e s , which wi l l co l l ec t checked codes , e s p e c i a l l y those used for shielding calculations, and supply them to participating countries . This l ibrary, located at Ispra (Italy), Will come into serv ice during 1964.

The other e s tab l i shment i s the Centre for the compi la t ion of nuc l ear data, set up to col lect al l the resul t s of measurements of nuclear constants. This Centre, which was established at Saclay in 1964, will therefore bé able to supply a l l the ' in format ion ava i lab le on the c r o s s - s e c t i o n s n e e d e d f o r shielding calculat ions. The two new c e n t r e s wil l work in c lo se l ia ison with s i m i l a r centres in the United States of Amer ica .

More general ly , the microscop ic and macroscopic aspects of shielding problems are the concern of the European-American Nuclear Data Committee (EANDC) and the European-American Committee on Reactor Physics (EACRP), set up within ENEA in 1959 and 1962 respectively.

Shielding experts re fer to EANDC requests for measurements of c r o s s -sect ions for which they cannot find published va lues . F r o m the roughly 60 participating laborator ies , EANDC then s e l e c t s the one bes t able to carry out the measurement needed and a s s i s t s in providing the necessary materials.

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STATUS OF SHIELDING RESEARCH AT EURATOM ISPRA

H. Penkuhn

X. THEORETICAL DEVELOPMENT

1.1. Neutron transport theory

An attempt has been made to so lve the e n e r g y - d e p e n d e n t Bo l t zmann equation in plane geometry and for e last ic scatter ing. The c r o s s - s e c t i o n s were assumed constant, but the calculations can be general ized. As in [1], the angular neutron flux was expanded in eigenfunctions of the energy trans-f e r operator (powers of the energy); the i r c o e f f i c i e n t s obey o n e - v e l o c i t y transport equations, which can be solved, for instance , with the theory of Van KAMPEN [2], М Щ А [ 3 ] , CASE [4] and others . For energy-dependent c r o s s - s e c t i o n s the equat ions a r e coupled, for constant o n e s they a r e in-dependent. A p r o g r a m m e applying this theory to s o m e standard c a s e s i s prepared .

1.2.. Gamma transport theory

The Boltzman equation for gammas in plane geometry was written a s a pair of coupled integral equations [5] . After eliminating the Dirac-6-function by an az imuthal integrat ion, a 3 - d i m e n s i o n a l m e s h was c h o s e n in angle , space and energy, and a programme was written for the IBM 7090 to so lve the equations. Starting with the unscattered flux, the angular fluxes at lower energies were computed success ive ly without iteration. The build-up factors ca lculated in th i s way di f fer f r o m those of [6] and [7] by about 5% or l e s s .

2. EXPERIMENTAL DEVELOPMENT

The 5.5 MeV Van de Graaff acce lera tor at Padova was used to produce (in a g iven d irect ion) m o n o e n e r g e t i c neutrons be tween 1 and 8 MeV (with Li and D t a r g e t s ) .

F a s t ep i thermal and t h e r m a l neutron f l u x e s w e r e m e a s u r e d ; in the interpretation of fast neutron measurements a l inear build-up was taken into account . The v a l u e s of the energy-dependent r e m o v a l c r o s s - s e c t i o n s of water, measured by S32 and then by In45, showed di f ferences up to 7%. F o r a s p e c i a l energy a Monte -Car lo resu l t lay just between the two m e a s u r e d v a l u e s [8] . A s s u m i n g a l l in teract ions with hydrogen equivalent to an ab-sorption, the removal c r o s s - s e c t i o n of oxygen was about 60% of its scattering c r o s s - s e c t i o n . Measurements of the phosphorus activity at the Geesthacht swimming-pool reactor were compared with a calculation using these values; in a range of ~ 1 m, the g r e a t e s t e r r o r was 7%. At Saluggia, the E T N A converter plate was used to measure the neutron fluxes in straight cylindri-cal a i r - f i l l ed aluminium tubes in a water tank. Indium, neptunium, thorium, uranium, aluminium and nickel were used for the fast flux, gold, dysprosium

136

and plutonium for the thermal one. F o r a distance Z of 25 tube radii , the activit ies decreased like Z~2-15 for equal source and tube diameters (= 10cm),' but like Z-2.34 f o r ф source = 3 0 c m = 3 </>tube (the s lopes depended somewhat on the detector material ) ; the unscattered intens i t i es should theoret ica l ly be-have l ike Z'2. Other runs were made with tubes of iron or with tubes f i l led with iron shot.

The gamma streaming round rectangular corners in concrete was studied by m e a n s of a Co60 s o u r c e for d i f ferent square ducts (10 c m X 10 c m and 20 cm X 20 cm, 130-160 cm long) with a Victoreen ionization chamber. Be-hind a c o n c r e t e sh ie ld with a s traight duct, m e a s u r e m e n t s w e r e done in planes paral le l to the shield surface in order to find the lateral distribution of the dose (shield th ickness 4 0 - 8 0 cm, tube c r o s s - s e c t i o n 10 cm X 10 cm) with a chamber Mx 32.

In order to have an intense neutron source in a wel l -def ined geometry, it i s planned to construct a converter plate IRIS (Irradiation ISPRA) of highly enriched uranium at the thermal column of the reactor ISPRA I. The plate i s located about one metre from the core .

T h i s m e a n s a high p o w e r for a c o n v e r t e r p late — s e v e r a l k i lowat t s instead of s e v e r a l watts — and it must be cooled with a i r (not with water , which could eventually make the assembly crit ical in the case of an accident or distort the f i s s ion spectrum). Measuring shield models with such a device might become interes t ing for mobile r eac tors (ships, rockets) , because in this context optimization with respect to weight i s important, and it i s better to have an experimental check for the calculations.

Project work was a l so done for the fast reactor SORA (source rapide). Well-known programmes like GRACE I and II, RE-34 and MAC were used. In the case of MAC, some inconsis tencies had to be eliminated in co-operation with the AEG Frankfurt; the new v e r s i o n i s MAC-RAD ( r e v i s e d and de-veloped). Other p r o g r a m m e s like BARNS II or NIOBE have been prepared for production, and s m a l l e r auxi l iary p r o g r a m m e s have b e e n wr i t ten .

Some of the se r e s u l t s wil l be publ ished la t er in an externa l E u r a t o m report .

R E F E R E N C E S

[1] FERZIGER, J. H. and LEONARD, A.., Ann. Phys. 22 (1963) 192-209. [2] Van KAMPEN, N. G . , Physica 21 (1955) 949. [3] CASE, K. M . , Ann. Phys. 9 (1960) 1 -23 . [4] MIKA, J. R. , Nucl. Sei. Engng U (1961) 415-27. [5] WEINBERG, A .M. and WIGNER, E. P . , The Physical Theory of Neutron Chain Reactors, Univ. Chicago

(1958). [6] GOLDSTEIN, H. Fundamental Aspects of Reactor Shielding, Pergamon Press, London (1959). [7] GOLDSTEIN, H. and WILKINS, J. F. J r . , Rpt NYO-3075 (1954). [8] MATTHES, W., Internal Rpt EURATOM-ISPRA 430 (1963).

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SUMMARIES O F DISCUSSIONS

SOURCES O F RADIATION IN R E A C T O R S

M. GROTENHUIS*, H. DOPCHIE, H. SMETS AND U. TVETEN

1. PRIMARY SOURCES

F i s s i o n neutron s o u r c e s genera l l y have been determined for reac tor phys ics uses . As a result the high-energy component of the spectrum, 8 to 18 MeV, g e n e r a l l y does not r e c e i v e enough attention to s a t i s f y sh ie ld ing i n t e r e s t s . We do find that the s p e c t r u m from t h e r m a l neutron f i s s i o n in U235 i s known adequately for shielding purposes . While the spectrum from non-thermal neutron f i s s ion in U235 i s not known, the differences in the high-e n e r g y region, which i s of i n t e r e s t in sh i e ld des ign , are b e l i e v e d to be neg l ig ib le [ 1 ] .

The f iss ion-neutron spectrum from Pu239 in fast-neutron reactor sys tems i s known up to 9 MeV [2] . E s t i m a t e s at l eas t should be made for the spectrurn above this energy. Neutron spectra up to 18 MeV from the f i s s ion of Pu241, U233 and Cf2 5 2 wi l l be needed in the future. It should be mentioned that we need to know these spectra in the high-energy region for the purpose of deal ing with the attenuation of h i g h - e n e r g y neutrons and a l s o for the purpose of calculat ing r a t e s of h i g h - t h r e s h o l d neutron r e a c t i o n s such as N17 , N16 , A (n, p)Na, e t c . , which l e a d to s p e c i a l sh ie ld ing p r o b l e m s .

The p r o m p t - f i s s i o n g a m m a - r a y s p e c t r u m and the f i s s i o n product g a m m a - r a y spectrum from the f i s s i o n of U2 3 5 by thermal neutrons i s ade-quately known for shielding purposes . For U233 and Pu239 these spectra are presumed to be essent ia l ly the same as for U235. There i s no data for f iss ion caused by non-thermal neutrons but this is not l ikely to be radically different.

2. SECONDARY SOURCES

The secondary sources of radiation are the least well known of radiation s o u r c e s of importance for shielding. This uncertainty may be ascr ibed in part to the l imited knowledge of the associated neutron c r o s s - s e c t i o n s . The production of neutrons by the decay of N 1 1 hàs had a rather large uncertainty, although recent r e s u l t s have been encouraging [3]. G a m m a r a y s f rom t h e r m a l neutron capture , or neutron radiat ive capture g a m m a r a y s , are we l l known with the exception of a few m i s s i n g e l ements and a few d i s c r e -p a n c i e s . The noteworthy except ions are U2 3 5 and Pu 2 3 9 and a l s o s o m e of the rare earth e l e m e n t s . D i s c r e p a n c i e s are unfortunately in the region of 3 to 4 MeV, which i s often the energy of g r e a t e s t penetrat ion in shie lding problems. In addition, a number of e lements have spectra measured above 3 MeV but not below that energy. At a recent international conference [4] , two full s e s s i o n s were devoted to recent work on capture g a m m a rays . It should be noted that the capture g a m m a - r a y spectra for the concretes [1, 5]

* Sub-Panel Chairman

139

have been r e v i s e d [6] to a c c o m m o d a t e the data of TROU BET ZKO Y and GOLDSTEIN [7] . Efforts are now needed to determine a s imi lar set of data for the ep i thermal neutron capture g a m m a rays [8]. Recent data indicate fa ir ly important changes in g a m m a s p e c t r a with the energy of the neutron being captured [9] .

A large part of the above problem i s the accurate determination of the c r o s s - s e c t i o n s involved. These c r o s s - s e c t i o n s become rather smal l above 10 keV and are often neglected, but in non-hydrogenous shie lds the gamma rays from neutron absorpt ions at these e n e r g i e s a s s u m e a greater impor-tance and m a y e v e n dominate . At the s a m e internat iona l c o n f e r e n c e ment ioned above [4] , a s e s s i o n was a l so devoted to c r o s s - s e c t i o n s . This report , the la tes t report to the USAEC Nuc lear C r o s s Sect ions Advi sory Group [10] and the la te s t report from the Brookhaven S igma Center [11] , represent the most recent compilations of l i terature on c r o s s - s e c t i o n s . A recent Euratom report [12] and the la tes t Nuclear C r o s s Section Advisory Group compilat ion [13] define the situation in regard to requests for future measurements of c r o s s sect ions . Compilations of useful c r o s s - s e c t i o n se t s appear regularly in the l i terature [14, 15], each set ref lect ing the improve-m e n t s up to that date. A c r o s s - s e c t i o n se t for fast r e a c t o r s i s p r e s e n t l y in preparat ion at Argonne. It w i l l be p r e s e n t e d at the 1964 Geneva Conference .

The situation in regard to the gamma rays from inelast ic scattering of neutrons i s , of c o u r s e , much m o r e c o m p l e x . It i s a l so m o r e di f f icult to make the required m e a s u r e m e n t s . Continuing work [16 - 19] i s gradually increasing the s tore of available data [14, 15, 20] and, although it wil l never be as s imple as the data for thermal neutrons, it is gradually approaching the min imum n e c e s s a r y amount. To date l i t t le has been done to m e a s u r e or ca lcu la te the angular distribution of the ine las t i c neutrons or g a m m a -r a y s , although r e c e n t data [21, 22] indicate that s ign i f i cant a n i s o t r o p i c s e x i s t .

3. RECOMMENDATIONS

The fo l lowing addit ional data i s d e s i r e d : (1) The m e a s u r e d Pu 2 3 9 f i s s i o n neutron s p e c t r u m up to 18 MeV. (2) The f i s s ion neutron spe.ctrum of Pu241, U2 3 3 and Cf25Z, if appreciably

different from the f i s s ion neutron spectrum of U235. (3) The capture g a m m a - r a y s p e c t r a f rom capture of n o n - t h e r m a l

neutrons in the e l e m e n t s . (4) The n o n - t h e r m a l neutron capture c r o s s - s e c t i o n s . (5) The g a m m a - r a y spec tra from the inelast ic scattering of neutrons,

e spec ia l l y for neutrons from 5 to 15 MeV. (6) The neutron ine las t ic scatter ing c r o s s - s e c t i o n s . The Pane l sugges t s that the above recommendat ions be transmitted for

considerat ion to the national or international bodies actively engaged in the consideration of measurements in the f ields of nuclear constants and reactor p h y s i c s , such as the European A m e r i c a n Nuc lear Data Commit tee or the European A m e r i c a n Commit tee on Reactor P h y s i c s .

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R E F E R E N C E S

[1] "RadiationShielding", Reactor Handbook, III В (BLIZARD, E.P. and ABBOTT, L .S . , Eds) Interscience, New York (1962).

И STEWART, L. , Nucl. Sei. Engng 8 (1960) 595. [3] AMIEL, S. and GILAT, J . , Reactions (n, p)Nrl and О18 (n, d)N17 with Reactor Neutrons, Nucl .Se i .

Engng 18 1 (1964) 105. [4] THROW, F. E. et a l . , International Conference on Nuclear Physics with Reactor Neutrons, Rpt ANL-6797

(15-17 Oct. 1963). [5] UNITED STATES ATOMIC ENERGY COMMISSION, Reactor Physics Constants, Rpt ANL-5800, 2nd. ed.

(July 1963). [6] WALKER, R.L. and GROTENHUIS, M. , "A Summary of Shielding Constants for Concrete", Rpt. ANL-6443

(Nov. 1961). [7] TROUBETZKOY, E. and GOLDSTEIN, H . , "A Compi la t ion of Information on Gamma-Ray Spectra

Resulting f rom Thermal Neutron Capture" , Rpt ORNL-2904 (1961); Nucleonics, 18 11 (1960) 171. [8] SCHAMBERGER, R. et a l . , "Experimental Measurements of Secondary-Gamma-Ray Production in Shielding

Materials", Trans. Amer. Nucl. Soc. 6 2 (1963) 48. [9] BOLLINGER, L. M. et a l . , "Fluctuations in Partial Radiation Widths", Phys. Rev. Lett. 3 8 (1959) 376.

[10] SMITH, A.B. , "Report to the Nuclear Cross Sections Group", Rpt. WASH-1046 (1964). [11] GOLDBERG, M.D. et a l . , "Angular Distributions in Neutron Induced Reactions", Rpt BNL-400 (Oct. 1962). [12] SPAEPEN, J . , "Compilat ion of Requests for Nuclear Cross Section Measurements f rom Euratom", Rpt.

EANDC-E-43-L; UC-34 (Jan. 1963). [13] SMITH, A. B., "Compilation of Requests for Nuclear Cross Section Measurements", to be published as

APR WASH Rpt. (1964). [14] CONNOLLY, Lucie D. e t a l . , "Los Alamos Group-Averaged Cross Sections", Rpt. LASM-2941 (July 1963). [15] SCHMIDT, J . J . , "Neutron Cross Sections for Fast Reactor Materials, Part II, Tables", Rpt KFK-120¡

EANDC-E-35-U (Dec. 1962).

[16] TROUBETZKOY, E .S . , "Continue Theory of Gamma Ray Spectra Following Inelastic Scattering", Rpt

NDA 2111-3 В (1 Nov. 1959). [17] TROUBETZKOY, E . S . , "Fast Neutron Cross Sections of Iron, Silicon, Aluminium, and Oxygen", Rpt

NDA 2111-3, С (1 Nov. 1959). [18] TROUBETZKOY, E.S. et a l . , "Fast Neutron Cross Sections of Manganese, Calcium, Sulfur and Sodium

(Final Report)" Rpt NDA 2133-4 (31 Jan. 1961). [19] TRALLI, N. et a l . , "Neutron Cross Sections for Titanium, Potassium, Magnesium, Nitrogen, Aluminum,

Silicon, Oxygen and Manganese", Rpt UNC 5002 (Jan. 1962). [20] GOLDBERG, M. D. et a l . , "Angular Distribution in Neutron Induced Reactions", Rpt BNL-400 (Oct. 1962). [21] CRANBERG, L. et a l . , "Identification of Zero-Spin States by Inelastic Neutron Scattering", Phys.Rev.

Lett. 11 7 (1963) 341. [22] BORNING, J .W. and McELLISTREM, M. T . , "Differential Cross Sections for (n, n',y ) Reactions in Several

Nuclei", Phys. Rev. 124 (1961) 1531.

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E X P E R I M E N T A L FACILITIES AND E X P E R I M E N T A T I O N

, J . BRAUN* E. AALTO, G. RICHTER AND A . C . WHITTIER I

1. SHIELDING FACILITIES AND EXPERIMENTS

A l i s t of the main shie lding f a c i l i t i e s and p r o b l e m s studied has been extracted from the status reports submitted by the participants. This shows that the f a c i l i t i e s which are operat ing now thoroughly c o v e r the range of tradit ional neutron shielding inves t igat ions , i . e . s tudies of penetrat ion through s lab s h i e l d s made up of c o m m o n m a t e r i a l s .

A l a r g e number of s tud ie s concern ing ducts through convent ional mater ia l s have been performed. Many of these, especial ly the earl ier ones, may be cons idered as belonging to the area of fundamental invest igat ions . Recent ly , m o r e e m p h a s i s has been put on c l a s s e s of p r o b l e m s of s p e c i a l in teres t for s p e c i f i c r e a c t o r p r o j e c t s .

High-per formance sh ie lds or sh ie lds for fast r e a c t o r s require infor-mat ion of l e s s s ign i f i cance and thus have been left unexplored in shie lding studies for large thermal power reactors . Secondary gamma production and heat s o u r c e s in s h i e l d s are two of the f i e l d s in which i m p r o v e d a c c u r a c y i s n e c e s s a r y . Studies such as t h e s e could be p e r f o r m e d щ convent ional sh ie ld ing f a c i l i t i e s .

Further shie lding r e s e a r c h in each country should be f r a m e d to lend m o r e as s i s tance to the corresponding national reactor programme. It may be expected that in the future much of the bas i c r e s e a r c h wi l l be just i f i ed by its pertinence to specif ic reactor projects, such as fast reactors or those for mobi le appl icat ions . Space shie lding can be cons idered as a new dis -c ipl ine . It requ ires a broad range of fundamental s tudies and is not dis-c u s s e d in th i s report . Al l countr i e s having a r e a c t o r p r o g r a m m e of any kind should have a c c e s s to a shielding facility to support their project work.

The development of s tochast ic methods of calculation has reached the s tage where in many c a s e s mathemat i ca l e x p e r i m e n t s ( i . e . Monte Carlo) can considerably reduce the number of experimental observations necessary for any investigation. Since these methods require a large volume of c r o s s -sect ions which are not known suff ic iently prec i se ly , they must be supported by shielding measurements and experiments should be chosen with this pur-pose in mind.

More work should be done to invest igate the accuracy of calculat ional methods by m e a s u r e m e n t s on operating instal lat ions . This should include not only reactors , but also smal ler structures such as transport casks, spent f u e l conta iners , e t c . In addition, an i n c r e a s e d e f for t on co l l e c t ing data relating to different engineering problems should be made on existing nuclear instal lat ions . The type and amount of active corros ion products and actual l e v e l s of radiation for different types of irradiat ion conditions in reac tors are e x a m p l e s .

It m u s t be e m p h a s i z e d that only an i n c r e a s e d v o l u m e of th i s type of engineering data wil l make it poss ib le to uti l ize the full potential of ref ined

* Sub-Panel Chairman

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ca lculat ional m e t h o d s , Comparisons"between ca lcu la t ions and m e a s u r e -ments are important in .order to test the quality of the methods of calculation. To do this , a c c e s s to al l data about the e x p e r i m e n t a l fac i l i ty that1 i s u s e d i s n e c e s s a r y . Data g iven in published reports are very often insuff ic iently complete to al low others to per form comparat ive calculat ions . Therefore it i s r e c o m m e n d e d that a report should be put toge ther giving up^to-date data on the main experimental fac i l i t i e s of member countries . This report should be a helpful re ference for subsequent papers .

2. INSTRUMENTS FOR MEASUREMENTS OF SHIELD PERFORMANCE

We consider it unnecessary to d i s cus s the detection methods general ly u s e d in r e a c t o r p h y s i c s in this s e c t i o n as t h e s e a r e subjec t s f o r s p e c i a l pane l s in ins trumentat ion .

The detect ion of fluxes in the resonance of thermal , lower epi thermal and thermal energy reg ions are commonly based on the ordinary fo i l a c t i -vation and f i s s i o n chamber techniques . The accurate detection of detailed flux distributions can be made by these methods and we believe that no more spec ia l development work is needed in this area.

In fast neutron detection there are two diff icult ies not yet solved sa t i s -factori ly: sensit ivi ty is low and spectra are difficult to obtain. The sens i t i -v i ty of the thresho ld de tec tors used , p e r unit of detector v o l u m e , i s low. Thus the u s e of t h e s e de tec tors i s n o r m a l l y l imi ted to the f i r s t 1 to 1.5 m of the shie ld . Only s p e c i a l t echniques , e. g . the sulphur burning method, enable measurements further out to be made. No generally usable, practical spec trometer e x i s t s for the comparison of calculated and measured neutron spectra as yet. Neutron spectra would be needed, for example, to correlate radiation damage calculations with observations and also to check the avail-able sh ie ld des ign m e t h o d s . E x p e r i m e n t a l methods of s p e c t r o m e t r y r e -quiring extraction of beams from the point to be studied introduce consider-able e x p e r i m e n t a l d i f f i cu l t i e s . The s h i e l d i s disturbed and the s p e c t r u m to be m e a s u r e d i s changed. The g e o m e t r i c a l attenuation in the beam tube r e d u c e s s e n s i t i v i t y . In sp i te of t h e s e d i f f i cu l t i e s t i m e - o f - f l i g h t m e t h o d s s e e m to offèr a possibi l i ty of obtaining spectral information which i s hard to obtain in other ways .

In the area of g a m m a detection the methods for measurement in mixed n e u t r o n - g a m m a radiat ion f i e ld s should be studied m o r e . N o r m a l l y the g a m m a dose has been g i v e n as the to ta l ionizat ion m e a s u r e d wi th an ion chamber. In some studies, however, it has been indicated that the detectors u s e d m a y have rather high s e n s i t i v i t i e s for fast neutrons. In accurate m e a s u r e m e n t s this m a y g ive an appreciable e r r o r (10%).

2. 1. Trends in instrumentation

Semiconductor detectors for neutron spectrometry are being developed in many p l a c e s . They offer , at l eas t in theory, the poss ib i l i ty of a s m a l l -s i z e d detector for use in shie lds . No general ly accepted, working type has yet been developed. Secondary react ions and g a m m a pi le -up are the main

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s o u r c e s of diff iculty. Ef forts to overcome the second group of di f f icul t ies e l ec tron ica l ly (pulse shape, discriminat ion, e tc . ) are being made.

B e s i d e s the w e l l - e s t a b l i s h e d threshold de tec tors working in the MeV-reg ion , s o m e e f f o r t s have been reported in the u s e of ine las t i c r eac t ions leading to m e t a s t a b l e s ta tes ; In (n, n') and Rh (n, n') can be ment ioned as examples . The difficulty of detecting the resulting activity makes it unlikely that a consistent set of foi ls for convenient measuring of spectra in the keV-region will be developed.

3. LIST OF MAIN SHIELDING FACILITIES AND PROBLEMS STUDIED*

Belg ium

(a) BR- 1, B R - 2 and BR-02 can be u s e d (b) BR-3 v e s s e l i rradiat ion to be s tudied in the VENUS fac i l i ty ( c ) -

Canada (a) P lugs built into r e a c t o r sh i e lds (b) Extens ive m e a s u r e m e n t s through the main biologici i l sh i e lds of NRX,

NRU and N P D (c) NRX plug now r e m o v e d

Euratom (a) Lid Tank Fac i l i ty planned at reac tor . ISPRA-1 (Iris) (b) Laminated sh i e lds ( c ) -

CSSR (a) Thermal column in WWR-S (b) Laminated i ron-graphi te including channels (c) Plans on install ing f i s s ion plates; Ra-Be source 1 0 7 n / s at Skoda Works

Research Institute

Federa l Republic of Germany (a) Research reactor Geesthacht (FRG); collimated beam facility (ESTAKOS)

(since 1961); mock-up faci l i ty (ESTAGROP I) (1961): radiation window (ESTAGROP II) (1964)

(b) Laminated sh ie lds (c) Plans on studying shie lds for ship reactors; inserts in ducts; measure-

ments in shie lds of working reactors (KAHL)

France (a) Two f i s s ion plate fac i l i t i es (NAÏADE I in reactor ZOE 1957, N41 ADE II

in r e a c t o r NÉRÉIDE 1960); two r e a c t o r s (TRITON 5 MW, NÉRÉIDE 100 kW)

(b) Laminated s h i e l d s , s t reaming on ducts (c) Various m e a s u r e m e n t s by 600 keV a c c e l e r a t o r

* (a) Main shielding facilities; (b) main problems studied; (c) remarks.

144

- Italy (a) A lid tank facility (ETNA) at the reactor Avogadro at Sorin (1962) (fission

plate in thermal column) (b) Laminated shields; streaming in ducts ( c ) -

Japan (a) Under construct ion: JRR-4, wi l l have p o o l - l i d tank with f i s s i o n plate

and dry shie lding f a c i l i t i e s ( b ) -

(c) Cr i t i ca l end of 1964; ship shie lding s tudies planned

Norway . (a) Under construct ion: JEEP-II , has a water tank for sh ie ld ing s tud ie s (b ) -

(c) C r i t i c a l middle of 1965; ship shie lding s tud ies planned

Sweden (a) R 2 - 0 reac tor with one large ( 2 m X 2m) and four s m a l l ( 0 . 5 m X 0. 5 m )

windows (1960) (b) Laminated sh i e lds : s t reaming in ducts (c) Semiconductor detectors developed; measurements outside the sh ie lds

of R-3 power reac tor Switzerland (a) SAPHIR (b) Laminated shie lds; s treaming in ducts (c) Measurements per formed around the shie ld of the DIORIT heavy-water

(20 MW) r e s e a r c h reactor

United Kingdom (a) LIDO with i ts fac i l i t i e s (big window 6 ft X 6 ft X 12 ft) (1958) (b) Laminated shie lds; s treaming in ducts

(c) Measurements in exist ing shie lds of Magnox-type reac tors

USA (a) Oak Ridge lid tank (closed): Oak Ridge bulk shielding facility: Oak Ridge

tower shielding faci l i ty; Hanford shie ld tes t faci l i ty; Aerospace Shield Tes t Reactor (ASTR); SNAP shie ld test faci l i ty

(b) Too numerous to be l i s t ed here

(c) United Nuclear Corporation plans a shield mock-up reactor

USSR (a) "B-2" fac i l i ty at the fast r e a c t o r B R - 5 (Moscow) ( co l l imated b e a m ,

d i a m e t e r 25 c m ) (1958) (b) Laminated s h i e l d s (c) Scint i l lat ion methods for m e a s u r i n g fluxes, dose and spectra; s e m i -

conductor d e t e c t o r s

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METHODS F O R C A L C U L A T I N G RADIATION A T T E N U A T I O N IN SHIELDS

J.BUTLER*, D. BÜNEMAN, A. ETEMAD, P. LAFORE, A .M. MONCASSOLI, H. PENKUHN, M. SHINDO AND B. STOCES

1. INTRODUCTION

In recent years the development of high-speed digital computers of large capacity has revolutionized the f ield of reactor shie ld design. For compact s p e c i a l - p u r p o s e r e a c t o r s h i e l d s , M o n t e - C a r l o c o d e s in two- and t h r e e -dimensional g e o m e t r i e s are now available for the proper treatment of both the neutron and g a m m a - r a y problems. Furthermore, techniques are being developed for the t h e o r e t i c a l opt imizat ion of min imum-we ight sh ie ld con-f igurat ions for th i s type of r e a c t o r s y s t e m . In the d e s i g n of l a n d - b a s e d power reactors , on the other hand, there i s a strong incentive to reduce the capital cost of the plant, and economic cons iderat ions are a l so relevant to r e a c t o r s des igned for m e r c h a n t ship propul s ion . In th i s context s i m p l e methods are needed which are economic in the i r data input and computing t ime requirements and which, at the s a m e t ime , are suf f ic ient ly accurate for des ign work. In g e n e r a l the computing t i m e required for Monte-Carlo ca l cu la t ions in c o m p l e x g e o m e t r y i s e x c e s s i v e for routine d e s i g n c a l c u -lat ions and the capaci ty of the p r e s e n t c o d e s i s inadequate for the proper treatment of large reac tor sh i e ld s y s t e m s in t h r e e d imens ions . In t h e s e c i r c u m s t a n c e s a wide range of s i m p l e r techniques are current ly being employed for des ign ca lcu lat ions .

2. NEUTRON PENETRATION CALCULATIONS

The methods of calculation for neutrons in reactor shields fall naturally into four categories:

(1) Multigroup diffusion theory; (2) Multigroup diffusion with removal sources; (3) Transport codes; and (4) Monte-Carlo methods.

Owing to the difficulty of meàsuring neutron spectra, methods of calculation are usual ly t e s t e d against integra l m e a s u r e m e n t s of thermal , ep i thermal and fast f luxes made in bulk shielding f a c i l i t i e s . Calculated s p e c t r a m a y a l s o be c o m p a r e d with the r e s u l t s of the m o r e r igorous t ranspor t c o d e s , such as NIOBE, in s i m p l e g e o m e t r i e s .

In order to use s imple diffusion theory for neutron attenuation problems, empir ical correct ions must be applied to compensate for the underestimation of the penetrating f luxes . Thus the diffusion parameter in the French two-group method PENELOPE can be fitted to obtain an accuracy in the calcu-lated t h e r m a l f lux of better than 30% in the inner reg ions of laminated graphi te , i ron and c o n c r e t e s h i e l d s . This f lux i s , of c o u r s e , important

* Sub-Panel Chairman

146

b e c a u s e of the a s s o c i a t e d capture g a m m a r a y s . S i m i l a r m e t h o d s can be used to predict sulphur "fluxes" but they do not, of course, give the fast neutron spectrum. In genera l these methods are s imple and accurate but they r e -quire a large vo lume of exper imenta l data and are n e c e s s a r i l y r e s t r i c t e d in their application. Attempts have a lso been made to apply various mult i -group codes writ ten for core des ign ca lcu lat ions . E x a m p l e s in this ca t e -gory include the AIM [1] and the PIMG c o d e s [2] and they a r e probably adequate as they stand for predicting f luxes in r e f l e c t o r s but they fail con-spicuous ly at large d is tances from the core and var ious e m p i r i c a l fac tors are employed to compensate for their underestimation of the penetrating flux.

A more rewarding approach using diffusion theory in shields i s to adopt the s e m i - e m p i r i c a l removal source technique. Current methods are based on e i ther the Albert -Welton kerne l which ca l cu la te s the attenuation of the penetrating or "removal" flux in hydrogenous shields, or SPINNEY'S concept of the energy-dependent removal c r o s s - s e c t i o n [3] . In the former, empiri -ca l diffusion p a r a m e t e r s are again employed with two or three groups and the accuracy of the t h e r m a l f lux p r e d i c t i o n s in i r o n - w a t e r s h i e l d s i s g e n e r a l l y bet ter than 50%.

There are at present three dist inct methods based on the energy -dependent r e m o v a l concept . In the RASH s e r i e s of c o d e s [4] the Spinney removal c r o s s - s e c t i o n s are used, which are in fact transport cross - sec t ions calculated by Feshbach and Weiskopf for the hard sphere model of the nucleus, and the continuous slowing-down model i s employed in the multigroup scheme. In the Hanford code MAC [5] and the NRN [6] code deve loped in Sweden, a l lowance i s m a d e for d i s c r e t e e n e r g y l o s s e s in the mul t igroup s c h e m e . Under these conditions it appears to be n e c e s s a r y to redef ine the r e m o v a l cross^sect ion . The Swedish technique i s to adjust the portion of the e last ic s ca t t er ing c r o s s ^ s ec t ion which i s inc luded in the r e m o v a l c r o s s - s e c t i o n by comparison with bulk shielding exper iments . All three methods enable damage monitor react ion rates to be predicted to within about 50% at pene-trat ion d i s tances in light water , D 2 0 and graphite appropriate to r e a c t o r p r e s s u r e v e s s e l s in s l ab g e o m e t r y . The t h e r m a l f lux p r e d i c t i o n of the RASH E code i s better than about 50% at these penetrations and se ldom ex-ceeds a factor of two or three for attenuation of the order of e"30. The NRN code should a l s o be capable of th i s o r d e r of a c c u r a c y but it has not b e e n tested in such a wide range of mater ia l s as RASH E. The MAC code which has retained the Spinney c r o s s - s e c t i o n s with the d i scre te energy l o s s for -m a l i s m appears to be somewhat l e s s accurate for thermal f lux predict ions in hydrogenous shie lds .

Of the various multigroup techniques which are in use , it appears that the s impler methods based on empir ica l parameters are favoured by groups which are re spons ib l e for both exper iment and sh ie ld des ign ca lcu la t ions w h e r e r e c o u r s e cän be had to sh i e ld m o c k - u p s tud ie s for new s y s t e m s . Organizat ions which do not have d irec t a c c e s s to e x p e r i m e n t a l f a c i l i t i e s pre fer the l e s s empir i ca l methods based on the energy-dependent remova l concept in which all the parameters , with the exception of the removal cros s -sec t ion i t se l f , are ca lculated from the fundamental nuclear data. Finally, it i s of i n t e r e s t to note that al l the mul t igroup m e t h o d s d i s c u s s e d by the P a n e l fa i l to predic t the t h e r m a l f lux in thick s l a b s of i ron. This s h o r t -coming has been attributed to the approximation whereby thermal neutrons

147

are assigned to a. s ingle group, which i s c learly not adequate for an absorbing medium with poor moderating propert ies . In pract ice , however, the error i s not significant because the thermal flux in the inner regions of the s labs , which contribute the predominant source of capture gamma rays , i s due to neutrons which are t h e r m a l i z e d in the adjacent hydrogenous med ia and i s genera l ly predic ted to within about 50%.

The transport codes which are s p e c i f i c a l l y intended for sh ie ld ing problems have been written in one dimension and varying degrees of mathe-m a t i c a l approximation are employed ranging from the low order pn mul t i -group approach - recent t e s t s of the P3MG code [7] have shown agreement between ca lcu lated and m e a s u r e d t h e r m a l f l u x e s throughout i r o n - w a t e r shie lds to within 50% - to the f u l l - s c a l e numerical integration of the trans-port equation performed by NIOBE. The latter i s written in spherical geo-metry and subject only to truncation e r r o r s and uncertaint ies in the funda-menta l nuc lear data. It i s not however su i table for s m a l l s o u r c e a r r a y s and the computing t i m e s required for typical shielding problems are of the order of a few hours on the IBM 7090 computer. The Carlson Sn code SAINT [8] a l so affords a proper treatment of anisotropic scatter ing and it i s writ ten in both plane and s p h e r i c a l g e o m e t r y . Car l son codes are a l s o avai lable in two d imens ions although they do not appear to have been e x -ploi ted for r e a c t o r sh i e ld des ign p r o b l e m s .

Few of these codes have been tested directly against experiment because of their res tr ic t ion to ideal ized g e o m e t r i e s . In pract ice the approximation of a power reactor c o r e and shie ld geometry in one-d imens ional codes can introduce errors of the order of a factor of two in regions of the shield where the accuracy of the flux prediction material ly affects the shield design. Thus the NIOBE prediction of threshold detector react ion rates in the water tank of a swimming pool reac tor a g r e e s within exper imenta l e r r o r at d is tances in e x c e s s of ~ 50 cm where the required assumption that the core is spheri-ca l in shape i s not a disadvantage. The_ Carlson code SAINT in plane g e o -m e t r y has been u s e d to predict damage f luxes at the p r e s s u r e v e s s e l of a l arge Calder Hall reac tor to within ~30%. Some t i m e - o f - f l i g h t m e a s u r e -ments have been reported in lithium hydride and a comparison with the NIOBE spectrum at penetrations of the order of ten mean free paths revealed s igni-f icant e r r o r s . In tercompar i sons have a l so been made of transport codes using the United Nuclear Corporation data l ibrary at penetration d i s tances of the order of 15 mean free paths have revealed d i screpanc ies amounting at some energ ie s to a factor of two.

Severa l o n e - d i m e n s i o n a l Monte -Car lo codes are now avai lable for shielding calculation and the most eff icient i s undoubtedly SANE II [9] which employs an analytic importance function and can so lve a typical bulk shie ld problem in spherical geometry in about an hour on the IBM 7090. This may be compared with the est imate of 10 min required for the multigroup removal s o u r c e c o d e s RASH [4] and NRN [6] . T w o - d i m e n s i o n a l c o d e s have a l s o been written for shielding problems with cyl indrical symmetry and the most versat i l e of the currently available codes i s ADONIS [10] which treats combi-nat ions of b o x e s , s p h e r e s , c y l i n d e r s and w e d g e s in t h r e e - d i m e n s i o n a l g e o m e t r y .

The g e n e r a l conc lus ion and r e c o m m e n d a t i o n s of the P a n e l can be s u m m a r i z e d a s fo l lows:

148

(1) The bulk shielding problem in slab geometry i s largely so lved and a v a r i e t y of methods i s now avai lable to suit the r e q u i r e m e n t s of p o w e r reactor shield design during the next few years . There is , however, a need for some clean geometry measurements of high prec is ion with wel l -def ined source spectra to e s tab l i sh the validity of the transport and Monte -Car lo codes .

(2) The application of m o r e r igorous techniques to des ign p r o b l e m s wil l reveal the e f fects of nuclear data e r r o r s which are currently obscured in s e m i - e m p i r i c a l methods . The major uncerta int ies which affect the accuracy of shie ld des ign at present are epi thermal capture and ine las t ic scatter ing data.

(3) Recent attempts to elaborate the multigroup methods using energy dependent remova l s o u r c e s have not s igni f icant ly improved the accuracy of flux prediction. To rea l i ze the full potential of the method, it would be necessary to establish a technique for measuring energy-dependent removal cross - sec t ions to within an accuracy of ~1% which could be need with multi-group parameters calculated entirely from the nuclear data.

(4) From the large vo lume of exper imenta l data from bulk shielding faci l i t ies published in the l iterature there i s evidence of considerable dupli-cation of effort in this f ield especia l ly in'European countries. (It was, how-ever agreed that countries engaged in shield design should have experimental fac i l i t ies to a s s i s t in this work, and it i s l ikely that duplication in standard m e a s u r e m e n t s wi l l occur in pract ice and training. ) The m e m b e r s of the Panel s t r e s s e d the importance of including full detai ls about the geometry , source strengths and absolute calibrations in reports to enable the published data to be used by other organizations to test theoretical methods.

(5) There are comparatively few published accounts of c a s e s in which design methods have been checked in detail against measurements made on operating power r e a c t o r s during the c o m m i s s i o n i n g s tage . A higher accuracy i s required for this purpose than i s usual ly n e c e s s a r y in routine health-physics surveys, but investigations of this type are essential to a s s e s s the overall accuracy of design methods in practical situations.

(6) Future prob lems in reactor shielding wi l l be concerned with the treatment of complexgeometry, particularly for the computation of secondary gamma sources for nuclear heating. Monte Carlo i s potentially capable of solving all problems of this type but for economic reasons the full potential wil l not be real ized for severa l years .

3. GAMMA-RAY PENETRATION

Since the publication of the c la s s i c compilation of the moments method build-up factors by Goldstein and Wilkins in 1957, European e f for ts in shielding have been concentrated on the neutron problem. More recent ly however , Monte-Carlo c o d e s have been wri t ten to s o l v e the m u l t i l a y e r problem [11 - 13] . They are quite fast and can be u s e d for des ign work. Nevertheless considerable interest has been shown in rec ipes for combining infinite medium build-up fac tors us ing both Monte-Carlo r e s u l t s and e x -perimental data, and again, there i s a need for clean geometry experiments at high eriergies. A cons iderable amount of work has been done with low

149

energy sources at Universit ies and academic research institutions in Russia, Japan and the United States of America. In order to establ ish the accuracy of Monte Carlo and other methods at high energ i e s it wi l l be n e c e s s a r y to develop monoenêrgetic or discrete line sources of high intensity and for this purpose l inear a c c e l e r a t o r s and t h e r m a l neutron capture techniques are current ly being employed.

In spite of the numerous Monte-Carlo techniques which are avai lable for penetration and back scattering, ser ious problems are st i l l encountered in practice with the scattering of gamma rays from walls of buildings which contain crit ical faci l i t ies and also concrete-l ined discharge shafts containing irradiated fuel e l ements . The considerable volume of data in the unc lass i -fied literature on the solution of problems of this type in civi l defence work appears not to have been evaluated for reactor shield design.

4. THE TRANSMISSION OF NEUTRONS AND GAMMA RAYS IN DUCTS

In genera l there are three types of duct geometry to be cons idered in reactor design:

(1) Single ducts penetrating through a bulk shield; (2) Multiple duct s y s t e m s ; and (3) Large g a s - f i l l e d voids .

The streaming of neutrons and gamma rays in s y s t e m s of this type can only be properly tackled by the two- and three -d imens iona l Monte-Carlo codes d i s c u s s e d above for neutrons, mos t of which can a l s o handle g a m m a - r a y problems. In European countries however, effort has been mainly devoted towards the treatment of neutron streaming in both s ingle and multiple duct s y s t e m s us ing s imple l i n e - o f - s i g h t and homogeniz ing techniques , with diffusion theory to calculate the build-up t h e r m a l neutron f lux in the s u r -rounding m e d i a . The good agreement which has been obtained in s i m p l e configurations of air and liquid f i l led cyl indrical ducts in graphite, D2O and l ight-water s y s t e m s sugges t s that the major unsolved problems of complex ducts in large power reactor designs wil l in future be tackled by methods of this type which are checked against Monte Carlo and experiment.

5.' RADIATION HEATING IN SHIELDS

Both pr imary and secondary radiations gave r i s e to heat deposition in shie lds and if i s usually the temperature gradients rather than the absolute temperature r i se which constitute a design problem by virtue of the thermal s t r e s s e s which are induced in the shield members .

The deposit ion of the fast neutron kinetic energy contr ibutes usua l ly but a smal l fraction of the total heating. This heat distribution can be es t i -mated from the neutron flux distributions, the scattering cross - sec t ions and the associated energy loss . The magnitude and distribution of the heat depo-sited by particulate radiations induced by neutron absorption can be evaluated in a s imilar manner. The most important problem is the heating by primary and secondary g a m m a rays . The methods of a calculat ion for secondary g a m m a s o u r c e s , mainly due to thermal capture in power reac tor sh ie lds ,

150 -

have been discussed above and again the uncertainties in the primary gamma source calculations are mainly due to complex geometry . For more rapid evaluation, the point - to-point attenuation kerne l method i s often u s e d but the energy build-up fac tors are not known for all the m a t e r i a l s which are u s e d in for compos i t e sh ie lds .

The ability to ca lculate the overa l l sh ie ld attenuation for g a m m a - r a y dose rate does not imply automatically that shield heating can be calculated to adequate a c c u r a c y . Not only are the important g a m m a - r a y e n e r g i e s di f ferent - l o w - e n e r g y photons y ie ld heat concentra ted n e a r the s o u r c e , but do not é f f èc t the dose at a ful l Shield t h i c k n e s s - but a l s o the c o n s e -quences of e r r o r are g r e a t e r . Thus uncertainty of a factor of two in dosé can be compensated by a thickness increase of a few per cent, whereas the s a m e uncertainty in heating would require a doubling of thè sh i e ld cool ing s y s t e m .

R E F E R E N C E S

[1] FLATT, H.P. and BULLER, D . C . , AIM - A Multigroup One Dimensional Diffusion Code, Rpt NAA-SR-

4694 (Mar. 1960).

[2] ANDERSON, D . C . and SHURE, K . , Thermal Neutron Flux Distribution in Metal Hydrogenous Shields,

Nucl.Sei. Engng 8 (Sep.1960) 260.

[3] AVERY, A. F. , BENDALL, D. E., BUTLER, J. and SPINNEY, K. T . , Methods of Calculation for Use in the

Design of Shields for Power Reactors, Rpt AERE/R. 3216.

[4] BUTLER, J . , The status of theoretical methods for reactor shield design, Abstr. these Proceedings; AEEW/

R-361.

[5] PETERSON, E .G. , MAC - A Bulk Shielding Code, Rpt HW-73381.

[6] AALTO, E. and BRAUN, J . , A summary of shielding research in Sweden, these Proceedings.

[7] SHURE, K. , P-3 Multigroup Calculations of Neutron Attenuation, presented at Amer. Nucl. Soc. Meeting

(June 1963).

[8] Solution of the One Dimensional Multigroup Stationary Neutron Transport Equation on the IBM 7030

(STRETCH) Computer (WADE, R. D . , Ed. ) Rpt AWRE O-12 /63 .

[9] G UBER,' W. and SHAPIRO, M . , A Description of the SANE and SAGE Programmes, Rpt UNUCOR-633

(Mar. 1963).

[10] EISENMAN, B. and HENNESSY, E. , ADONIS an IBM Monte Carlo Shielding Code, Rpt UNUCOR-635

(Mar. 1963). -

[11] LEIMDORFER, M . , A Monte Carlo Method for the Analysis of G a m m a Ray Transport for Distributed

Sources in Laminated Shields, FOA 4 RAPPORT A4354-411 (Feb. 1964).

[12] CRISCUOLO, L. , GAMMONE - A Monte Carlo Programme for Calculating the Attenuation of Gamma

Rays in Laminated Shields, Rpt GN-38-41 OH 63 (Nov! 1963).

[13] STEINBERG, H . , TRIGR-S and TRIGR-P,, Nucl."Sei.Engng 12 (1962) 554.

151

ENGINEERING DESIGN PROBLEMS

J .M. PICTET*. C. CORTISSONÉ, Yu. A. EGOROV, P.S. MITTELMAN, I . RASTOIN AND A. ZBYTOVSKt

1. GENERAL REMARKS

It i s not poss ib l e to g ive recommendat ions which are valid in all c i r -cumstances regarding the specif ic choice of materials; since this will depend to a l arge extent on the part i cu lar r e a c t o r . In what f o l l o w s we are c o n -cerned essent ia l ly with mixed shielding only ( i . e . shielding that attenuates both gamma radiation and neutrons) as used for shie lding a reac tor c o r e . A very large number of materials can be used for this purpose, and we shall not deal with them separate ly . Our purpose i s to bring out certain general charac ter i s t i c s by means of a few chosen examples and to mention certa in mater ia l s which are of m o r e immediate in teres t .

2. CHOICE OF SHIELD MATERIAL

2. 1. Low-cost materials

The least expensive types of biological shields are st i l l made up of o r -dinary concrete or water . Sand can a l so be used, e s p e c i a l l y in countr ies where a high-density mater ia l - containing, for example, a high proportion of iron and m a g n e s i u m - can be obtained cheaply . F o r the absorpt ion of gamma rays, e . g . in a thermal shield, s tee l i s s t i l l the most advantageous mater ia l .

If it i s des ired to increase the ef f ic iency of a biological shield made of concrete, recourse will be had to components such as colemanite or pander-mite which contain boron; this wi l l enable the contribution due to capture gamma rays to .be reduced. However, this r a i s e s a problem of c o m p a t i -bility between the boron compounds and the cement, as there may be a c h e m i c a l react ion between them; it would be des i rab le to develop a boron compound by means of which such reactions could be avoided. Use can a lso be made of concretes containing rare-earth e lements .

A mater ia l which i s often used for the absorption of thermal neutrons i s boral . Boron carbide can a lso be used, but it i s expensive. It would, in general , be des irable to develop boron compounds of this type which could be obtained m o r e cheaply . Borax might p o s s i b l y o f f e r a good so lut ion .

2. 2. Materials for high-performance shields \

The most ef fect ive combinations are naturally the most expensive, and we have a whole range of mater ia l s available for which c o s t s wil l increase With e f f e c t i v e n e s s .

* Sub-Panel Chairman

152

We may start with the heavy concretes, for which densit ies up to 9 g / c m 3

(by including lead shot) can be obtained. Shie lds of this type may have the s a m e attenuating proper t i e s as alternating sh ie lds of light and heavy e l e -ments; their cost i s l e s s and they are e a s i e r to instal l , but their d i s -advantage i s that the i r total weight i s in g e n e r a l s l i ght ly h i g h e r .

Alternating arrangements of iron and water or lead and water are more eff ic ient in relat ion to the weight, and the ratio of the two components - the heavy e l e m e n t s to absorb the gamma rays and the water for attenuation of neutrons - can be optimized so as to obtain the best possible total absorption. Po lye thy lene can a l s o be used for the attenuation of neutrons and the fact that the shie ld can be constructed of b locks which can subsequently be r e -moved i s an advantage.

The bes t neutron a b s o r b e r s are m e t a l hydr ides , such as l i th ium h y -dride, this mater ia l makes it poss ib le to d e c r e a s e the production of s e c o n -dary gamma radiat ion. On the other hand, meta l hydrides are e x t r e m e l y expens ive , ma in ly owing to the high d e g r e e of c h e m i c a l puri ty required; their chemical reactivity i s also very high. The best elements for absorbing gamma rays are those with the highest density. Tungsten, for example, can be used in the form of var ious compounds; natural uranium and, m o r e e s -pecial ly , depleted uranium afford a l so poss ib i l i t i e s .

In general, for the purpose of providing mixed protection against gamma rays and neutrons, the most efficient materials , relative to the weight of the shielding, are pure homogeneous mixtures of hydrogen and heavy e l ements such as hydrides .

Arrangements such as those just referred to are of primary importance for the protect ion of mobi le r e a c t o r s , whose weight must be kept to a min imum.

2. 3. High-temperature materials

Shielding mater ia l s which can be used at high temperatures include s e r -pentine and other a g g r e g a t e s u s e d in c o n c r e t e , m i x t u r e s of graphite and iron - boron can be mixed with the graphite , at the cost of cer ta in m a n u -facturing diff icult ies - boron carbide, tungsten, lithium hydrides and finally, uranium; mixtures of lead and calc ium can be used at higher temperatures than lead by i t s e l f .

2. 4. General problems relating to shielding materials

A s we have s e e n , c o n c r e t e can be used in a wide v a r i e t y of p o s s i b l e combinat ions . More considerat ion has s t i l l to be given to the quest ions of maximum permiss ib le temperatures and temperature gradients, water l o s s e s and radiation damage . It has usual ly been supposed that radiation "damage in concrete was negligible in comparison to damage result ing from thermal s t r e s s e s ; this supposition needs careful re-examination when we are dealing with h i g h - p e r f o r m a n c e w a t e r - c o o l e d sh i e lds of the p r e s s u r e - v e s s e l type .

Under the e f f ec t of radiation, water d e c o m p o s e s at a rate which it i s n e c e s s a r y to d e t e r m i n e . In genera l , the damage caused in m a t e r i a l s by radiat ion should be be t ter known.

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The extent of c o r r o s i o n phenomena i s s t i l l d i f f icult to d e t e r m i n e . Finally, it should be made poss ib le to calculate the degree of activation of the mater ia l s m o r e accurately; in many c a s e s c r o s s - s e c t i o n s are not a c -curately known.

3. OPTIMIZATION OF DESIGN

3. 1. Optimization with respect to weight or volume

The problem of designing reactor shielding with a view to obtaining the minimum weight or volume i s a complex one and involves choosing the most eff icient absorbent mater ia l s poss ible and arranging them in the most su i t -able manner poss ib le .

Heavy e lements are n e c e s s a r y to absorb the gamma radiation from the core and from neutron interactions; e lements with a high inelast ic-scattering c r o s s - s e c t i o n for fast neutrons wil l be the mos t suitable . Light e l e m è n t s are needed to slow the neutrons down to thermal energ ies and for this pur-pose the choice wil l be e lements with a high c r o s s - s e c t i o n for e last ic s c a t -tering. In this case, it i s very desirable to add a component which will make it poss ib le for the s lowed-down neutrons to be rapidly absorbed without the e m i s s i o n of secondary gamma rays . , •

The kinds of e l e m e n t that have good absorpt ion ( c h a r a c t e r i s t i c s for gamma rays and for neutrons of d i f ferent e n e r g i e s are t h e r e f o r e not the s a m e . A change in the re lat ive proportions of l ight and heavy e l ements in the mater ia l s used in a shield wi l l resu l t in a corresponding change in the partial d o s e s due to gamma rays and. neutrons outside the shield. The problem is to determine the optimum combination of materials , with respect to either weight or volume, which wil l yield the des ired partial doses in the desired rat ios . These optimum combinations can be determined by exper i -ment and by calculation. They may consist of either homogeneous mixtures, such as heavy concrete with a high hydrogen content, or alternating arrange-ments of lead and water, iron and water, iron and graphite, etc.

For many years it was the practice to calculate the optimum thicknesses, composit ions and geometr ical arrangements by iteration methods. Recently more ef f ic ient methods have been developed for optimizing with respect to weight, volume or cost; these methods re late to spher ica l sh ie lds and are based on the ca lcu lus of var ia t ions . This new technique, which i s ca l led "shield synthes is" , often l eads to a choice of mater ia l s and arrangements which could not have been predicted intuitively. A s the technique i s based on simple exponential laws of attenuation for neutrons and gamma rays, it i s usual ly n e c e s s a r y to make certa in correc t ions in the r e s u l t s thus obtained before they can be regarded as f inal. The total weight of the shielding can of ten be reduced s t i l l further if it i s des igned in such a way that the l e s s important r e g i o n s outs ide the shie ld r e c e i v e l e s s protect ion .

3. 2. Optimization with respect to cost

The problem of optimization with respect to cost i s c lose ly linked with that of optimization with respect to weight or volume. The methods of c a l -

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culation are s imi lar , account being taken of the cost of the different materials apart from their attenuating properties .

For the purpose of optimization with respect to cost, account must a lso be taken of certain factors that are not direct ly related to the shield as such but which, neverthe less , play a part in determining the total cost o f / sh ie ld-ing; it i s often difficult to define prec ise ly what part of a reactor plantshould be regarded as belonging to the sh ie ld ing for c o s t i n g p u r p o s e s . F o r 'example, the s y s t e m used for unloading fuel e l ements may make it n e c e s -sary that part of the shielding should be to a special design; there are many s imi lar c a s e s for which the requirements do not permit optimization of the shield with respect to cost . Finally, it should be borne in mind that the transport of shie ld ing m a t e r i a l s f r o m extract ion s i t e to p lace of u s e a l s o af fects the f inal cos t of shie lding.

4. DESIGN CRITERIA

4.1. Maximum permissible levels of radiation

Maximum p e r m i s s i b l e l e v e l s per person and per year are laid down in international regulations for the various categor ies of population. Such r e -gulations are e s s e n t i a l if the e x p e r t s a r e to be able to p e r f o r m shie ld ing ca lculat ions proper ly .

4. 2. Safety factors

It i s a l m o s t i m p o s s i b l e to lay down s a f e t y f a c t o r s that are genera l l y valid. These factors depend on the type of reactor in question and wil ldiffer, for example, depending on whether it i s an experimental reactor or a power reactor; moreover , the manner of est imating the n e c e s s a r y safety factors will often vary from one country to another. It i s easy to imagine that safety factors might vary from 1 to 10.

4.3. High-dose-rate areas and "hot spots"

Reactor f a c i l i t i e s often c o m p r i s e s m a l l a r e a s with a r e l a t i v e l y high dose rate; such "hot spots" can be found for example in exper imenta l r e -search areas or on the outside of a piece of equipment. In view of the fact that it i s only the accumulated dose which i s res tr ic ted , it i s often difficult to know what m e a s u r e s should be taken regarding these "hot spots".

It would be v e r y des i rab le if international organizat ions w e r e to g ive s o m e recommendat ions on this subject .

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G E N E R A L CONCLUSIONS

E. P. BLIZARD

The Panel , meet ing in plenary s e s s i o n on the las t day, e x p r e s s e d the strong opinion that this mee t ing had been v e r y t i m e l y . They noted that a number of countr ies , such as Germany and Japan, w e r e just beginning shielding research programmes . The hope was expressed that this bringing together of those new in the f ie ld and those with many y e a r s of exper ience , would lend wisdom to the f o r m e r and inspiration to the latter . The beg in-nings of the se in terchanges w e r e a c c o m p l i s h e d at the P a n e l m e e t i n g . A larger, more open meeting would sure ly extend the good that was done, and the Panel was much in favour of one. They hoped that the IAEA would be able to make the n e c e s s a r y arrangements within the next y e a r .

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ABSTRACTS O F S U P P O R T I N G P A P E R S *

N P D BULK SHIELDING

A.C. WHITTIER AND, J .W. HILBORN

During construction of the 20 Mff(e) NPD reactor, radiation measuring f a c i l i t i e s w e r e built into the bulk sh ie ld . Fas t neutron, t h e r m a l neutron, and g a m m a - r a y d is tr ibut ions have b e e n m e a s u r e d through the sh ie ld . Surrounding the reactor i s a 33-cm thick H2O reflector-shield. Measurements were made with and without this H2O in place.

SHIELDING EXPERIMENTS AND CALCULATIONS -AT THE RESEARCH REACTOR GEESTHACHT

D. BÜNEMANN, H .G . PENDLER, R. F1EB1G, F. FR1S1US, O.HOFFMANN,

H. KOTSCHI, H. PATZELT, G. RICHTER AND G. THURO

M e a s u r e m e n t s at two shie lding f a c i l i t i e s and the ir t h e o r e t i c a l in ter -pretation are treated in'this paper. Collimated beam experiments have been performed at the f irst facility ESTAKOS. By testing relatively smal lprobes of shielding mater ia l s it is poss ible to derive some cr i ter ia for the shielding e f f i c i ency of different arrangements . The r e s u l t s can be re la ted to plane g e o m e t r y by m e a n s of a t h e o r e m of r e c i p r o c i t y . In the s e c o n d f a c i l i t y ESTAGROP m o c k - u p e x p e r i m e n t s of p o s s i b l e r e a c t o r s h i e l d s have been per formed . In order to avoid s ide e f f e c t s large s labs of 2 m X 2. 5 m are p laced d irect ly in front of the c o r e in the r e a c t o r pool up to a m a x i m u m thickness of about 2 m. The calculations of neutron f luxes are based on the current methods of removal and multigroup diffusion theory. For the calcu-lation of the 7 - d o s e the build-up concept i s used. In spite of the large slab s i z e , strong correct ions have to be introduced at certain points in order to fit the ca lculat ions to the m e a s u r e d data. A s imple explanation for these ef fects i s given. Some difficulties result from the complicated core geometry.

The shielding programme at the R e s e a r c h Reactor Geesthacht i s being carr ied out in co-operat ion with EURATOM.

EXPERIMENTS ON SHIELDING IN GAS-COOLED GRAPHITE REACTORS; COMPARISON BETWEEN

EXPERIMENTAL AND COMPUTED DATA

P. LAFORE

Exper iments des igned to y ie ld information on la tera l shielding and in particular on the sources of capture-gamma rays in graphite reactors were carr i ed out in NAÏADE I, which u s e s a s source a natural -uranium plate

* Enquiries about the full texts of these supporting papers should be directed to the authors concerned.

159

emitt ing 4 X 10 ; n cm'2 s"x. The experimental r e su l t s obtained in different graphite, iron and concrete mock-ups were used to work out methods of cal-culation applicable to reac tors . These methods w e r e checked by means of m e a s u r e m e n t s made in the la tera l shielding of the E D F - 1 reactor .

In order to a s s e s s the damage l iable to occur in p r e s s u r e v e s s e l s , we related in a graphite-reactor spectrum the changes occurring in the mechani-ca l c h a r a c t e r i s t i c s of s t e e l to the in tegrated f lux m e a s u r e d by m e a n s of phosphorus detectors; the attenuation in the graphite of the flux responsible for the damage and the flux m e a s u r e d with the phosphorus de tec tors w e r e compared.

In reactors fitted with exchangers where the inside of the pressure v e s s e l general ly has to be access ib le during shut-down, the activation of structural mater ia l s brought about by neutrons from perforated re f l ec tors i s a particu-lar ly important problem. A number of prel iminary investigations have been carr ied out in the MARIUS assembly . These have mainly involved the study of fas t -neutron s o u r c e s ins ide the core . This work i s due to be continued in a s u b - c r i t i c a l mock-up being built at F o n t e n a y - a u x - R o s e s .

STUDY O F SOME NEUTRON SOURCES

J . RASTOIN

Natural uranium plate used for shielding studies

The paper descr ibes a study carried out to investigate the characterist ics of a fast-neutron source, viz . the natural-uranium plate in NAÏADE I. This plate, which i s 2 cm thick, i s fed by a s t r e a m of thermal neutrons leaving the re f l ec tor of E L - 1 . The intensity, spectrum and angular distribution of the neutrons leaving the plate were calculated by m e a n s of a Monte-Carlo method and by analyt ica l techniques . The computed data w e r e compared, with e x p e r i m e n t a l data obtained with act ivat ion d e t e c t o r s .

Natural uranium,rod in a graphite reactor

On the bas i s of the f i s s i o n distribution obtained by means of m e a s u r e -m e n t s of fine s tructure in thermal neutrons, ca lculat ions w e r e made with a Monte -Car lo method and with analyt ical techniques of the spec trum and intensi ty of the fast neutrons leaving the rod.

With this primary source calculations were then made of neutron attenu-ation in the graphite , f ine s tructure in fast neutrons in a r e a c t o r c e l l and the source corresponding to the lateral walls of the channel, which is partly responsible for neutron leakage in perforated re f l ec tors .

160

SYSTEMATIC APPROACH IN THE DUCT STREAMING CALCULATIONS

A. TSURUO, M. SHINDO AND M. KAWABATA

A sys temat ic calculational programme for the distribution of the radi-ation f lux at the centre and near the exit of a c ircu lar duct in the shielding barr i er i s descr ibed .

An infinite plane source at the entrance of the duct has an i sotropic or c o s n 0 angular distribution. To obtain the flux distribution at the exit of the duct, the following are calculated separately;

(1) The flux from the disc source at the entrance of the duct; (2) The f lux from the annular s o u r c e surrounding the entrance of the

duct; (3) The unscat tered flux, (4) Singly s c a t t e r e d flux, (5) Mult i scat tered flux.

Each flux i s to be compared with that in the case of infinite slab shield with-out ducts.

Unscattered flux i s r e p r e s e n t e d by e l e m e n t a r y functions or s ing le or double integra ls , and s ingly s c a t t e r e d f lux by 2- to 5 - fo ld in tegra l s . For the purpose of obtaining mul t i scat tered flux, we propose a semi -ana ly t i ca l technique using a Monte-Carlo method.

We are now concentrating on obtaining the gamma-ray flux at the centre of the exit of a duct. Contributions from the disc source and annular source to the u n s c a t t e r e d f lux are being obtained, the s ing ly s c a t t e r e d f lux f r o m the d i sc s o u r c e i s being ca lcu la ted and the s e m i - e m p i r i c a l approx imate formulae are being introduced.

This method g i v e s good ins ight into duct s t r e a m i n g p r o b l e m s .

COMMENTS ON SOME SHIELDING P R O B L E M S

J . M . PICTET AND A. ETEMAD

Some of the points ment ioned in the p r o v i s i o n a l agenda for the pane l d i scuss ion are br ie f ly d i scussed . Comments on the maximum p e r m i s s i b l e irradiation l eve l s , the uncertainty and securi ty factors related to shielding calculations and to operational work of the staff in a reactor plant are given. Problems concerning experimental fac i l i t ies for shielding research, radiation s o u r c e s , neutron and g a m m a ray penetrat ion in s h i e l d s and radiat ion s t r e a m i n g in ducts are d i s c u s s e d .

THE STATUS O F T H E O R E T I C A L METHODS FOR R E A C T O R SHIELD DESIGN

J . BUTLER

The current techniques employed for the d e s i g n of sh i e lds of p o w e r r e a c t o r s in the UK are r e v i e w e d . The g e n e r a l approach i s b a s e d on the

161

new RASH E removal age-diffusion code in conjunction with line of sight c o r -rect ions for s treaming. The validity of this technique has been established by c o m p a r i s o n with t h e o r e t i c a l s p e c t r a computed by t ranspor t c o d e s in s imple g e o m e t r i e s and with integral exper imenta l data f rom a wide range of bulk shield and duct configurations in the LIDO panel facilities, at Harwell. As an example of i t s application, the theoret ica l predict ions are compared with measurements made during the commissioning and operation of the gas -cooled Magnox reactors at Çalder Hall and Chapelcross.

162

LIST O F P A R T I C I P A N T S

P A N E L MEMBERS

E . P . BLIZARD (Chairman)

J. BRAUN

D. BÜNEMANN

J. BUTLER

H. DOPCHIE

Y u . A . EGOROV

M. GROTENHUIS

P . LAFORE

A . M . MONCASSOLI

J. M. P I C T E T

Oak Ridge National Laboratory Oak Ridge, T e n n e s s e e , USA

AB Atomenerg i Studsvyk, Nykoping, Sweden

G e s e l l s c h a f t für Kernenerg ieverwertung in Schiffbau und Schiffahrt mbH Große Re ichens traße 2, Hamburg, FRG

United Kingdom Atomic Energy Authority Building 510 T Atomic Energy R e s e a r c h Establishment-Harwel l , Didcot, B e r k s h i r e , England

Centre d'Etudes de l ' E n e r g i e Nuc léa i re Mol-Donk, Belgique

State Commit t ee for Atomic Energy Moscow, USSR

Argonne National Laboratory Argonne, I l l ino is , USA

Département des Etudes de P i l e C o m m i s s a r i a t à l ' E n e r g i e Atomique Centre d'Etudes N u c l é a i r e s de F o n t e n a y - a u x - R o s e s (Seine) France

Sez ione e n e r g i a nuc leare Fiat , Via Set tembrini 23 5, Turin, Italy

Institut f é d é r a l de Recherche en Mat i ère de R é a c t e u r s , Würenlingen Switzer land

M. SHINDO Japan Atomic Energy R e s e a r c h Institute Tokai R e s e a r c h Es tab l i shment T o k a i - m u r a , Naka-gun, Ibaraki-ken, Japan

В. STOCES Institute of Nuc lear R e s e a r c h Re§, Czechos lovak Soc ia l i s t Republic

163

A . C . WHITTIER

OBSERVERS

E . AALTO

С. CORTISSONE

A. E T E M A D

P . S . MITTELMAN

M . J . RASTOIN

G. RICHTER

U. T V E T E N

A. ZBYTOVSKY

Canadian Genera l E l e c t r i c Company 107 Park Street North Peterborough, Ontario, Canada

A B Atomenerg i Box 9042, Stockholm, Sweden

Comitato Naz iona le p e r l ' E n e r g i a Nuc leare Via B e l i s a r i o 15, Rome, Italy

Institut f é d é r a l de R e c h e r c h e s en Mat ière de Réac teurs , Würenlingen, Switzerland

United Nuc lear Corporation, 5 New Street , White P la ins , NY, USA

Département des Etudes de P i l e C o m m i s s a r i a t à l ' E n e r g i e atomique Centre d 'Etudes N u c l é a i r e s de F o n t e n a y - a u x - R o s e s (Seine), F r a n c e

G e s e l l s c h a f t für Kernenerg ieverwertung in Schiffbau und Schiffahrt mbH Institut für Reaktorphysik Große R e i c h e n s t r a s s e 2, Hamburg, FRG

Abbediengrn 11, Oslo 2, Norway

Institute of R e s e a r c h at Skoda Works P l z e ñ , Czechos lovak Soc ia l i s t Republic

REPRESENTATIVES

H. В. SMETS

H. PENKUHN

ENEA, 38 boulevard Suchet, P a r i s XVI, France

Centre Commun de Recherche d'EURATOM à Ispra V á r e s e , Ital ie

SECRETARIAT

Sc ient i f i c S e c r e t a r y : Edi tor :

A. MERTON Div i s ion of Reac tors , IAEA С. N. WELSH Div i s ion of Sc ient i f ic and

Technica l Information, IAEA

164

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INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1964

PRICE: USA and Canada: US $3.50 Austria and elsewhere: S 73,50 (£1.10; F.Fr.14,-; DM 12,25)