B.A.R.C./I-259

1 < n

GOVERNMENT OF INDIA ATOMIC ENERGY COMMISSION

SAFETY ASPBCTS IN PLUTONIUM HANDLING by S. Janardhanan, T. N. Krishnamurthi, S. B. Dabhadkar and S. D. Soman Health Physics Division

BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA 1973 B.AJI.C./I-259

GOVERNMENT OF INDIA ATOMIC ENERGY COMISSION

SA5BTY ASPECTS IN FTAJTONIUM HANDLING by

S. Janardhanan, T.N. Krishnamurthl, S.D. Dabhadkar and S.D. Soman Health HiyBics Division

BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA 1973 COMPEMPS fege So. 1. INTRODUCTION 1

2. RADIOLOGICAL SAFETY 1

2.1 Formation of ELutoniun iBotopes 1

2.2 Nuclear Characteristics of FLutoniua Radionuclides 2

2.3 External Hazardo 2

2.4 Internal Hazards 23

3. ENGINEERING SAFETY 31

"5.1 Glove Box for Hutoniun Handling 31

3.2 Design Safety Consideration 32

3.3 Ventilation for Hutonlum Laboratory 35

3.4 Surfaoe Finish for Plutonium Laboratory 36

3.5 Ere -commie 8 ioning Checks for Plutonium Laboratory 56

4. aONTAMINVTION EVALUATION, CONTROL AND HffiVENTION 37

4.1 Special Technique for Pu-Alr Monitoring 38

4i2 Surface and Bars onnel Monitoring 39

4.3 Decontamination 39

5. PLUTONIUM FIRE SAFETY 48

5.1 Plutonium Metal FireB . 49

5.2 Protective Measures for Hutonium Fires 50 5.3 Fire Safety in Storage, Handling and Shipment of 51 HLutonium Metal

6. CRITICALITY SAFETY . 51

6.1 Oriticality BarameterB 52

6.2 MethodB of Criticality Control 54

6.3 Safety FactorB 55 Page No.

6.4 Criticelity Data 57

6.5 Soluble Poisons 61 f, .6 Solid Poisons 61 f) .7 Pipe Intersections 61

6.8 Storage of Pu Metal, Compounds and Solution 61

6.9 Transport of Plutonium Metal, Compounds and Solution 68

6.10 Magnitude of Critieality Accident 68

6.11 Administration of Nuclear Safety 72

REFERENCES 73

Appendix-Is Design Details of a Typical Inert Atmosphere 75 Glove Box for Plutonium Handling laboratory

Appendix-Hi Design Details of a Typical Normal Atmosphere 79 Glove Box and Pumehood SAFETY ASPECTS ITT PLUTONIUM HANDLING by S. Janardhanan, T.N. Krishnamurthi, S.B. Dabhadkar and S.D. Soman

1. ISTBODOCTION Jlutoniuv is being extensively used in the atoaio enorgt industry in varioua forms. Its radioactive and fissile properties •ad its biologioal behaviour present a number of basards a ad spaoial considezations mist be given to the safety problems arising in the baodling of plutooiua. Much experieooe tea been gained in this fiell at Iroabay and other laboxatories art this report compiles important safety aspeots in handling plutoniua inoluding critioaiity oonaiderationa. The safe handliqg problems could be broadly divided into the following groupst 1. Radiological safety 2. Engineering safety 5* Contamination, its evaluation,control and prevention 4* Decontamination 5* Fire hazards and control 6. Critioaiity Bafety. 2« RADIOLOGICAL SAFETY 2.1 ?o mat ion of ELutoniua Isotopes PlutonlUB is formed by neutron irradiation of Uraniu&~238. Formation of Plutonium isotopes and their daughters are shown in ohart-V '. Composition of plutonlum changes with uranium burn-up in the reactor* while at low buro-ups Pu-239 and Pu-240 are the significant plutonlum radionuolidea, at high burn-ups, plutonium isotopes from mass number 233 through 242 are generated) in quantities, significant enough to present severe hugarda during handling.

2.2 Suolear Characteristics of Plutonium Badionuolides Decay characteristics of plutonium isotopes and their daughters are shown in Table 2.1*

Amongst these isotopes of plutonium,Pu-239 ie important aa it 1B the major constituent isotope in plutonium and It has a fission cross section of 746 barns for thermal neutrons, •• greater than that for U«235 (580 barns). Pu-240 la a parasitio neutron absorber for low energy neutrons• Pu-241, though produced in smaller amounts, is a better fisaile material than even Pu-239 on account of its higher thermal neutron fission cross section of 1025 barns.

2.3 External HazardB The handling of large quantities of plutonlum requires appropriate administrative control to minimise personnel exposure. The external radiation dose rate varies with the isotopio composition of plutonium*

2.3*1 Alpha particles emitted by plutonium isotopes do not constitute an external hazard because of its short range in air and body tissue (for Pu-239 alphas, range in,air is 3.68 en and range in body tissue arri water is 40 u). r37: n.?n

(675 ft)

p-335D) ?38JL w24V .242 n. Ciay)

OCC163D) 242

CHART-1

FORMATION OF PLUTONIUM RADIONUCLIDES AND DAUGHTERS Iable-2.1 PAHIQACT3VE DECAY CHARACTERISTICS OP PKJTONIUM ISOTOPES ATO DAUGHTERS

Zaotopo Radiation Yield $ Energy Specifio (MeV) activity (V)

Pu-238 Alpha 100 5.49 66.4 Oamna 10-9 0.1b n 8x)0 5 0,10 2 11 3.8X10- 0.044 L-X-ray 0.017

Pu-239 Alpha too 5.U Gamma 2x10-5 0*038 11 3 0.052 2.436x10 yr» 0.062 0*12-0.20 3x10-5 0.38 L~X-raye 1.4 0.0136 2.2 0.0174 0.2 0.0205

Pu-240 Alpha 76 5.162 11 5.118 6.58x10 yre 0.23 1O~2 ' 0.044 Ir-X-raya 10 0.017 K-X-raya •tO 0.102-0.125

Pu-241 Alpha 3x10"5 4.9 13.0 yra 111.5 Beta 99.997k 0*02 Gamma 2x10"4 0*145 11 IO-5 0.10

Pu-242 Alpha 76 4.89 3.79x1O^yra 0.004 n 24 o 4.05 10-2 0.045 Ir-X-raya 10 0.017 (Tabl* 2.1 oontlmwd)

laotop* Radlfttian Yl«14g( Energy Half-1 if • Specific (May) aottritj (Cl/g)

U-Z37 100 .245 6.75 4aye 6.74x10 61 .059 H 35 .207 If 4 .334 Aa-241 Alpha 84 5.46 13*6 5.43 45S 3.13 37 •017 2.7 .026 it .05 •043 ti 37 .059 n .02 .099 2.2.2 Major fraction of the external dose rate originates from X-raya and gamma rays with energies below about 20 to 40 KeV. The dose due to plutooium X-rays could be significant but these weak X-rays are easily absorbed in the thinnest structural material ., The L or M X-ray activity in plutonium is quite high but these intensities can be reduced to almost zero by normal' rubber gloves. As Aaerieium-241 and uranium-237 build.up is the separated plutonium, higher energy gammas (> 40 K.eV) are emitted and gamma shielding may be required to maintain acceptable doso rates*

The surface doae rate contributions from daughters of Pu-241 are a function of the Pu-241 concentration and the time since purification of plutonium. For timea much less than 14 years, the surface dose rate from Am-141 in the mixture ' is given by

=» 0.2

where t is time in days since purification of plutonium and

P241 is tha weight fraction of Pu-241 in the mixture.

The surface dose rate contribution from U-237 in the mixture , for time a much less than 14 years is given by

RadsAr . 23 P2410- e""

where pg41 and t are, aa defined earlier.

The contributions from Am-241 and U-237 to the surface done rates are shown in Figure 2,1 against time since purification of Pu. U-237 contributes most to the gamma dose rate initially} however U-237 build-up reaches equilibrium in a few weeks. The 40 80 100 160 200 240 TIME IN DAYS SINCE PURIFICATION

U) FIG. 2.1- SURFACE DOSE RATES FROM ?U DAUGHTERS dose rate due to Am-241 continue a to increase and just equals the U-237 dose rate at 115 days after purification. The dose rate due to Ain-241 continues to increase for many years until the parent Pu-241 can no logger maintain the build-up of Am-241 •

The total X-ray and gamma radiation surface dose rate from massive plutonium can be estimated from the following equation!

980 p238 +'0.67 P259 + 14 p240 + 0.2

0 102 • 23 p2U (1 - e" ' *) where t is the time in days since purification of plutonlum

and p ia the weight fractions of various pu isotopest(denoted by the subscript)*

Tha surface dose rates, for massive plutonlumi from X and gamma rays with energies greater than 40 KeT? can be calculated from the equations

Bem/hr - 2 p23Q + 0.056 p2jg + 0.62 p2AQ + 0.1 P24l*

102 • 23 p241 (1 - e-0' *)

where p and t are, as defined earlier.

An adc'ltioml source of gamma radiation from Plutonium, could be fission product contamination of Pu» however, this depends on the degree of decontamination achieved in the reprooeesing of spent fuel* Amongst the fission products, Bu, Zr, and Nb are the difficult oneB to remove. The tolerable fission product contamination level appears to be between 5 to 1.0 pCi par gram of Plutonium.

2*3»3 Beutron Emission from Plutonium Appreciable neutron dose rates are also associated with plutonium. Neutrons emitted have a wide range of energies upto tO ifeV or more. Fast neutron yields from spontaneous fission of plutonium are given in Table 2.2.

For a spherical mass of plutonium, the dose rate due to fast neutron can be estimated from the equation*

{ 009 0<02 °» P240 * > 2 where M is the mass of Fu in grams y- la the distance in cm and p is the weight fraction of Fu isotope (denoted by subsoript)

Neutron emission from plutonium in contact with light elements; due to ( <*> , n) reactione, contributes significantly to the neutron dose rates. Beutron yields from the light element-plutonium compounds, plutoolum fluoride and plutonium oxide are given in Table 2.3. Neutron emission rate from Plutonium chloride is about 100 fold less than that from fluoride. Approximate neutron emission from ( «fi , n) reactions with Be and Al are given belowt

Element Heutrons/aeo per curie of alpha emitter Be 2 x 10 Al 10+ 10

SPONTANEOUS FISSION NEUTRON MISSION RATES

Pu radionuolida Neutron yield q/aao-g of Pa

Pu-238 3.4 x 103

Pu-239 2.0 x 10*2

Pu-240 1.0 x to3

Pu-241

Pu-242 1.7 X 103 11

Table-2.3 0)

PAST NEUTRON EMISSION RATES PRCW PuF4 AND PuO2

Fa oompound Neutron emission rate n/aac-g of Pu

2,1 x 106

4.3 x 103

1.6"x 104

25a Pu 0g 1,4 x 10*

4.5 x 101

24O24O 2 Pu O 2 1.7 x 1100

2.7 12

Neutron emi&sion rate from plutonium in the presence of light element Impurities, through(

n • E(A> + BN) where n • neutron emission rate, n/g-min E • light element concentration, ppm A1 - constant for Pu-239 + Pu-240 B - constant for Pu-238 S - concentration of Pu-238, mol 238/mol Pu

where <^c» alpha disintegration rate/g mixture A m constant for pure Pu-239 Constant A for Pu-239 may be used for Pu-240 also, because the alpha energy for Pu-240 is approximately the same as Pu-239• The constants A and B for various light elements are tabulated in Table-2.4. Tabl3=2,5 lists the maximum permissible concentra- tion of each light element (if it existed alone) to give a neutron dose of 1000 mr/40 hr week at the surface of a kllogxen sphere. The actual concentration should be less than that listed if Pu-240 content is greater and other Impurities also existed, to keep the neutron dose below tolerance.

2.3.4 Calculated Surface Dose Bates Table-2.6 gives the summary of surface dose rates of Plutonium isotopes due to X rays, gamma rays and neutrons. 15

Iable-2. if

CONSTANTS FOB LIGHT ELMENT IMPURITIES^

Light Pu-239 oonatant A Pu-238 constant B element (S) (u/g-mln/pp« B)

Li 2.92 1,O9x1O3 Be 43.2 1.9U1O4 B 12.6 6.38x103 C 7.3x10"2 38.4 N 6.6x1 O**4 5.46 0 2.9xt0"2 15.5 3.49 2.01x103 ]fa 0.266 1.53x1O2 2% . 0.139 1.095x102 Al 0.179 1.355x102 Si 0.359 5.29x102 01 4.78x1O*2 89.6 A 0.106 1.515x1O2 E 1.06x10"2 20.0 Ca 9.3x10"3 18.25 So 1.O6x1Or3 3.10 6.6x10"* 2.74 14

Table-2.5

MAXMtM PERMISSIBLE CONCENTRATIONS OP LIGHT ELEMENT MPURITIES IN PLUTONIUM

Maximum Permissible Concent xat ion (ppm) Light element Pure Pu-239 Pu-239 +13.4 Pu-239+1000 Pu-239+13.5 wt# Pu-240 ppm Pu-238 wt# PU-240+ 1000 ppm Pu-238

LI 11 7.3 6 5*3 Be 0.47 0.31 0.32 0.24 B 1.59 1.06 '<.1 0.79 C 2575 1716 1680 1265 K 50,000 33,500 6100 5800 0 6500 4330 4200 3175 F 53 3.83 3.7 2.76 Ka 98 66 62 47 Mg 1350 900 760 590 Al U5 97 83 64 Si 280 166 100 84 Cl 1100 733 580 328 A 212 142 88 73 £ 3200 2133 1100 945 Ga 5650 3765 1900 1630 Sc 28250 18850 7200 6400 Fe 250,000 165,000 55,000 50,000 15

Table-2.S(4)

SUMMARY OF DOSE RATES OF Pu ISOTOPES

Surface doee xateCreayfar) with no shielding Total (reat/hr) 190eOp* X-ray Gasma Beutroc Total through glovoa

640 8.3 15 660 560

Pu-239 0.59 6.6X10""2 ,.1X10-* 0.66 0.56

Pu-240 7.5 0.11 4.6 12*2 11*1

Pu-241 - 2.7 - 2.7 2.7

Fu»242 - - S.2 6*2 8.2

tI-237* 0.22 11.5 - 11.7 11.7

Aift-241**• 0.06 0.08 - 0.14 0.13-

* fiiuilibriua valuta at the surface of c nsata *• % day afttr chemical Initially pure Pu-241 • sepaieticn 16

Surface dose rates for massive plutonium at different bura-ups and at various periods after separation have been calculated and are given in Table-2.7. The effect of self- abaorption has been neglected in these calculations, although it nay be quite significant for Pu-240 X-rays and Am-241 gamma rays in massive plutonium. Table-2.8 gives thefi,i and X rays doae rate at the surface of a 1 kg sphere of plutonium as a function of the burn-up and time since separation of plutonium. Plutonium produoad from low burn up fuel contains mostly Pu-239 and Fu-240 whose contribution to surface dose rate is not significant) in recycled plutonium higher isotopes of plutonium are present in considerable quantities and make large contributions to the dose rate* Calculated values for L or K X-rays and other hard components from pure Pu-239 and Pu-240 metallic cylindrical shapes are given in Tables 2.9 and 2.10. Table-2,11 gives the fast neutron fluxes and dose rates due to spontaneous fission from a one kg Pu sphere, representing certain typical Pu isotoplc compositions* The table assumes 3 neutrons/apontaneoue fission and 2 MaV/neutron. In all cases, self absorption of neutrons is Ignored. The dose rate for 2 MeV neutrons is 7.5 mr/hr per 44 neutrons/cm .see

Table-2.12 gives the calculated surfaoe dose lates due to fast neutrons from a 1 kg spherical billet for different reactor irradiations. A multiplication factor of 1.5 has been considered in this calculation to take Into account further Missioning* 17

(5) Table-2.7

SURFACE DOSE- HATES FROM MASSIVE PLUTONIUM (FROM X-BAYS AND GAMMA HAYS WITH ENERGIES GREATS THAN 40 lrev)

Irradiation Pu lsotopic Time after Dose rate (MtfD/Te) composition separation R/nr (wt i (yra) •

- 0.002 0.1 0.43 Pu-239 - 90.2 1000 Pu-240 - a.5 0.2 0.49 Pu-241 - 1.27 1 Pu-242 - 0.028

Pu-238 - 0.02 0.1 1.74 Pu-239 - 75.5 0.2 1 98 3000 Pu-240 - 19.8 Pu-241 - 6.04 1.0 5.76 Pu-242 - 0.64

Pu-238 - 0.10 „ . , „ Pu-239 - 60.0 °s1 3t33 10,000 Pu-240 - 25.0 0.2 3.78 Pu-242 -2.90 18

( P ,V , X) DOSE RATE AT SURFACE OF 1 KG SPHERE OP Pu)

Isotope Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Ao-241 U-237 Total does rate •Dime of (Bad/hr) separation (years)

1000

0.1 0.015 0.97 0.62 0.16 0.19 0.60 2.56

0.2 0.35 2.73

1.0 1.78 4.16

7>000 MWD/Te

0.1 0.15 0*79 1.44 0.75 0.86 2,86 6.65

0.2 " nun ?>68

1.0 " " " " 8.43 14.42

*3aa Table-2.6 for Pu isotoplc composition. 19

APPROXIMATE RADIATIOH IHTEffSITIES FROM Pu-239 FOR METALLIC CYMHERICAL SHAPES

Badiua Height Badlation level (sB/nx)at a distance from cyliader(cn)of («*) (on) 0 5 10 30 100

Prom L X-zaya associated with 11!i of decays

1 8 1280 172 51 5.2 0.5 1 15 n 236 79 10.5 1.0 1 30 it 262 97 13.6 1.3 3 24 n 586 248 46.6 4.3 3 45 it 628 273 51.2 5.5 3 90 n 663 325 65.2 7.1

From K-X rays, raya and other hard components

1 8 20 2S? 0.73 0.16 0.008 1 15 « 3.7 1.2 0«17 0,015 1 30 H 4.4 1.5 0.21 o.ceo 3 24 It 17.3 7.3 1.4 0.13 3 45 It 18.6 7.9 1.5 0.16 3 90 n 19.6 9.7 2.0 0.21 20

Tabla-g.1 or '

APPROXIMATE RADIATION INTENSITIES ERCM Pu-240 KB METALLIC CYLINDRICAL SHAPES

Badius Height Radiation level(mR/hr) at a diatarce from oyliader(cn)of (om) (om) 0 5 10 30 100

Vroa L-X ray* aseooiated with 24$ of dtoay*

1 8 6800 911 272 27 2.7 1 15 it 1250 422 56 5.3 1 30 n 1500 517 72 6.3 3 24 f 3114 1319 248 22.8 3 45 n 3340 1448 272 29.3 3 90 n 3520 1727 346 37.6

2rom K X-raya, raya and other hard componenta

1 8 33 4.4 1.3 0.13 O.pi3 1 15 n 6.1 2.0 0.27 0.025 1 30 n 7.2 2.5 0.35 0.033 3 24 H 28.3 12.0 2.3 0.21 3 45 n 30.6 13.0 2o5 0.27 90 " 32.2 16.0 3.2 0.34 2i

Iable-e.1t(3)

SPONTANEOUS FISSION DOSE HATES AT SURFACE AND ONE METES EROM 1 KG SPHERE

Spontaneous fission Spontaneous flaalon neutron flux at neutron dose ie>te(aB/hr) Tutl surface .. - .. « (n/W- seo) At aurfaoe It 1 wter

Pu-239 (100$) 0.667 0.1U 3.1Z x 10

Pu-239 + 20^ Pu-240 5.6 x 10J 920 0.252

.06x 10 1610 0.496 Pu-242

Pu-2.59 + 1000 ppm Pu-238 75 13 3.56 x 10

Pu-233 + 2$ 5u-240 + 1,06 x 1On 1810 0,496 Vu-2A2 + 1000 pp» Pu-238

Equilibrium Fu 2.44 x 4160 1.04

Pu-238 - 0.55^ Pu-239 - 36.1555 Pu-240 - 14.71^ Pu-241 - 8.82^ ioi-242 - 40.31^ 22

Table-2.12 (5)

N£UTflCN iX)SE JtATB PRCM 1 KG SFHERTCAL BILLET FOR COTTTACP HANDLING

Irradiation Spontaneous Surface neutron Doee rate (lWD/Te) fission flux (n/sec) (n/onr seo) Bad/nr

1000 1.8x1l.8x1005 2.65x10-' 0.663

3000 4.42x1O5 6.54X1O3 1.635

10,000 S.JxiO3 9.32x1O3 2,33 for a sphsrieal «aaa of 1 kg PuP. surface neutron 5 2 doss rate is 3 Had/ur and neutron flux is 2 x 10 n/c* -see. 2,3.5 Self Haating in Pu-2gB Soureast'6' PU-23B fuels generate 0*456 Wg" as oaapired with a Talue of 0.002 Wg~ Pu-239. She temperature attained by Pu-238 materials depends upon the total power aid the rate of heat transfer to the surroundings. A 5 W-piece of Pu-238 is aaid to have a temperature of several hundred degrees. Temperatures of over 900*0 have been measured for oxide pellets of less than 5 W under conditions of poor heat transfer. Because of the high specific heat of Pu-238, contaat with conventional combustibles must be minimised, Pu-238 aetal sources Bust be handled in inert atmosphere syBtenw to prevent oxidation and ignition.

Also aging of Pu-238 on account of alpha part ids decay results in continuous production of Helium at a rate of O»74 cm yr" g~ at STP. Provision for accoaodating thie gas aust be svads in containers. Pu-238 daughter nuclidoe like Tl-<:oa increases photon emission rate with tine

2.4 Internal Hagard The radioactive properties and the biological of Plutonium rake it one of the moot toxic oeteriale when taken into the body. The long biological half-life, the high energy of the eaitted alpha particloa and the body's selscti-re localisation of plutoniua in the bone lead to low ifesirua Peraiasible Body Burdens (MPBB) for plutoniua radionuolidea. 24

2.4*1 Critical Organ* Critical organ for plutoniua varies with the solubility of the material and method of intake* Bon* is th« oritieal organ for inhaled, ideated or Injeoted plutoniua in soluble fora. Lung is the critical organ for inhaled plutoniua in the insoluble state. The gastrointestinal traot (lower large Intestine) is the critical organ if Insoluble Plutonium la swallowed. Under oertain conditions of oonta- Bination of scratches or wounds with plutonlum, the wound site nay be the oritical tissue.

2.4.2 MPBB and MFC for Plutonium Isotopes Maximum permissible body burden (MPBB) and Maximum Permissible Concentrations (MFC) in air and water for plutoniua radionuclldes are presented in Table-2.iy.

2.4.3 Boutea of Entry The four major routes of entry for plutoniua into the body are; (i) by ingestion and subsequent absorption from the gastro-intestioal traot (ii) by Inhalation and subsequent absorption from the lung (iii) by absorption through unbroken skin and (ir) by direct ingestion into body tissue or into the blood stream. 2.4.3.1 Jngestlom Extensive studies on chronio and acute intake of plutoniun showed an average absorption of 0,005% froa the 91 tract and the present MFC values for plutonlum in gabla-a.13 (7)

MPBB AND WPC IN AIH AND WATER 5DH Pu RAD10NUCLIDES (SOLUBLE) CRITICAL ORGAN-BONE

UPBB MPC(4O hr •xpoBU.Ta/Week) (po/cc) Badionuclide Air • Water

Pu-23Q 0.04 2.4x10*"3 2x10 10 -12 Ptt°239 0.04 0.6? 2x10 10 -12 Pu-240 0.04 0.16 2x10 10

3 .-3 0.9 8i2s10" 9x10 7x10

Pu-242 0.05 12,8 2x10 10"

2 -12 An-241 0.09 1.6X10* 6x10 10- -12 U-237 0,06 89 4x10 9x10*

"Baaic Safety Standards for Radiation Prot»otionw, USk 8af

Po-236 - 3x10'11 uOi/oe Pu-239 - 4x10-11 Pu-240 - 4x10.-11 Pu-241 - 4x10" -11 Pu-242 - 4x10 26 drinking water are calculated on the basis of this figure. However, high acidity, presence of oomplsxing agents and hexavalent plutonium will lead to greatly increased absorption. Any plutonium left unabsorbed in the intestinal tract appears to be rapidly eliminated.

2,4,3,2 Inhalation^ Lung retention data and tissue deposition studies show that ultimate fate of inhaled Plutonium depends on the chemical form. Experiments show that PuOo(insoluble) is quite tenaciously retained in the lung and only a few percent is absorbed in the blood stream over a long period. However, very small PuOg particles, with mass media* diameter less than one micron, are absorbed as readily as plutonium nitrate. If plutonium is in the fora of soluble compounds, 10 to 50 percent is quite rapidly absorbed frcm the lung. A fraotion of plutonium left in the lung is retained with an increasingly lengthening half-life, probably because of conversion to more insoluble forms* Long term studies indicate accumulation of plutonium in pulmonary lymph nodes* Deposition ocours mainly in the bronchial lymph nodes with muoh smaller amounts in other lymph nodes.

Autoradiograph studies Indioate rapid clearance of Plutonium, deposited on ciliated parts of the lung. Host of the plutonium retained for as long as 16 days was present in the parenohymal areas. Elutonium particles deposited in lymph nodes were located within maorophagea. Smallest particles arc retained in the parenchymal areas while larger particles are found in the lyaphold tissue. 2?

Plutonium absorbed from the lung is distributed throughout the body with deposition occurring mainly in skeleton end liver.

Batention of Plutonium in the human lurg depenia on particle size in addition to chemical form anj. other complex biological factors. Of particular interest in air samplii^j work ie the importance of particle size studies in order to establish the reapirable fraction of the Pu-aerosols. The Task group of the ICEP on Lwag Dynamics has revised the earlier long model used for computation of the duat deposition in and clearance from human respiratory system* The hazards evaluation is made by calculating deposition pattern of tha aerosols with the aid of curves which relate deposition in the respiratory tract to aerodynamic behaviour. The Health Physicist has to assess the size spectrum to evaluat e the hazard. The distribution pattern adopted by the IGRP is shown in Chart-2.

2.4.3*3 Absorption through the skin; Skin contamination could occur in plutonium processing operations. The intact skin is a very effective barrier to absorption of plutoniua. Plutonium absorbed throi^gh the skin is deposited mainly in bone with only a few percent appearing in liver. Data on absorption of plutonium through the intact skin of man(4 ') is given in Table-2.14.

2.4.3.4 -Injections Injection pOBes a significant hazard in the industrial bailing of plutonium and in laboratory* Absorption of plutonium after intraderaal or subdenaal injection appears to depend on the aize of the w«uua. Between 20 to 40$ of the plutonium absorbed f rota the cuts (or sub- cutaneous injection) gets deposited in tha liver, while less Chart-2(a) Total mass inhaled^) Expired air eoncentzation 47/C of £• Total deposition in the respiratory ayaten 535& of \

D2 "5 556 of D of t 8^ of D1 Baso-pharnyogeal Tracheo-bronchlal PaUMBoary deposition deposition deposition I I \&jL of D^ 245C of Dt clearance clearance GI tract 360 days (4 aio.half time) (10 min. half tiae) 24 hr half half tlae time 1St of 995* of D2 % of Dj 995b of Dj HLood

(4) Table-2.U

ABSORPTION OP Pu THROUSI INTACT SKIN OF MAN

conditions absorption

10 ug Pu 0.0002$ absorbed per hour 0.4 H nitrio add solution Balaar skin

2» 2-5 rac Pu/nQ. 0.00002J* absorbed of arount CC14-TBP solution Initially on the hand Hand wrist immersion

Total deposition before deccntaaiaation about 5 mo than 2$ of the plutonium entering through the intaot skin deposits in liver .

2.4*4 Retention in various organ systems*

2.4.4.1 .Circulatory systemt Behaviour of plutonium in the blood depends on the ahemical and physical form of the material and the route by which plutonium enters the blood* Between 15 to 50$ of intravenously injected plutonium is usually lost from the blood within a matter of minutes* jautonluro remaining in the blood is almost completely oombined with protein and is reported to be bound to beta-globulin* 5 to 25$ of the initially injected plutonium may Btill be present in the blood after 24 hours. Behaviour of plutonium entering the blood through other routes is not known. Maximum blood plutonium level a were found one week following inhalation*

2,4*4.2 Bonet Bone is the major site of plutonium deposition and ia the critical organ from the point of view of radiation damage. The half retention time of plutonium is estimated to be about 200 years and this figure is used by ICBP in calculating the permissible exposure limits. Hence very littl* plutonium is lost from bone during normal life span.

Plutonium deposits on all calcified bony surfaces to which the blood has access. The greatest deposition ooours on endosteal surfaces,lining the marrow cavity and ooverlng the trabecula. Deposition occurs to a lesser extent on the periosteum and on the linings of haverslan canals and vascular channels. Redistribution of plutonium In bone with time following initial deposition has been observed and is found to be sensitive to plutonium dose levels* With high 31

•lose levels, bone cells in the neighbourhood of the initially deposited Plutonium are destroyed and the Initial distribution remain*' almost the same with the passage of time. At lower dose levels extensive redistribution of plutoniua occurs in trabciular bone and to a lesser extent in cortical bone. The redistribution results in spreading the plutoniua over a larger volume of bone thereby reduoing the hazards.

2.4*4*3 Livert In addition to bone, liver has been found to be a principal site for plutonium deposition. In certain human exposures via inhalation, plutonium concentrations 5 to 20 times higher than in bone have been observed* The partition between bone and liver is a complex function of the chemical and physical state and the route of entry of plutonium.

2.4,4,4 Other organst Spleen, kidney* adrenal glands have sometimes shown a relatively high depoaition of plutonium.

3. .EfiGINEERIHG SAMTY

3.1 Glove Box for Plutonium Hapdlinft'*'9'

In view of the high toxiolty of plutonium, a high degree of containment is neoessary for glove-boxes* In glove box design the following points are of importancei

a) Good containment during normal operations to prevent plutonium from entering the operating area. Comfortable working conditions including good lighting. This can minimise the risk of accidents in the box which could affect the operating areas, b) Safety of transfer into and out of the boxes, c) facility for cleaning and naintalnance of the boxes* Glova boxes are cecommended for operations with more than 1 mCi plutoniuiD in liquid form or 10 uCi in solid fora. Baaed on ventilation patterns, two types of boxes are identified namely (i) jtyntuflio flow boxes and (li) Static atmosphere boxes. Dynamio flow boxsa are supplied with dry air; on the contrary etatio atmosphere boxea depend on inlealoage for supply of air, with the result that air renewal rate is poor. Both types of boxep are operated with a pressure differential of about 1" W.G. for comfortable glove operation. Sensitive faat acting differential pressure controls are required for satisfactory operation of these box«>s. Ventilation system with self controlling diaphragm valvea is now a common feature. If Pu metal or other potentially pyro- phorio material ie to be handled, dynamic flow boxes purged with inert gas like argon are recommended. Higher purity atmosphere can be obtained by utilisation of a gas purification and recirculatioa system* Fr«e standing glove boxea can be connected together by tunnele using ports on either side of the boxes to preclude the necessity of intermittent baggirg-out operations. Where a fire hazard exists, use of fibre glass boxes or viewing panels made of porspex: are not recommended.

3.2 Pgalfin Safety Consideration

Sioco penetrations into tjlove boxes pose problems of air laakagc, performance of glove box fitted with equipments should always b

3.2.1 Leakage fiate Specification for the Glove Boi^ '

for air filled glove boxes the leakage rate should be lees than O.25;i of the box volume per hour with the box at 4" WG ne^a-clve pressure and with all glove port bungs fitted in position. For the higher integrity inert gas-purged boxes, the leakage rato should be 0«05Ji of box volume/hour. 3.2.2 Choice of Inert Gaa In Glove Box Inert gas is required for handling Pu-uetal, carbide* and pyrophoria compounds. It should el so be moisture free. Generally nitrogen aod argon are used as filling gases. Trace quantities of oxygen is monitored by a paramagnetic analyser or a Herach conductivity meter and moisture content with a

P2O5 cell.

3.2.3 Leakage Test This is carried out by filling the box with compressed air to a positive pressure of about (+)iOn WO and then isolating the box. The box pressure is read with a magnehelic gauge pressure meter every hour or so and from this, leakage rate is worked out.

3.2.4 Open Port glow Measurement Glova box ia connected to the exhaust line and a shutter device is mounted on the open port. '.fills shutter is slowly opened eo that the negative pressure in the box and any adjoining box connected to it is not disturbed. The air flow is measured by e sensitive veloaeter. An ideal box should have an open port flow of 100-150 linear feat per minute.

3.2«5 Testing of Glove Leakage The gloves of glove boxes are tested for any possible leakage or void over it while installing it on glove box. Compressed air is rushed in the glove fitted on ^Lova port and then oloeed with glove bungs. Leakage of air from the glove Is tested by observing the time for which th« glove remains atralglrfc with air ineide. Another method would be to release ammonium chloride smoke and visual observation of leakage,

3,2,6 Qlovea for glove Boxea and Qlpve Changing Opeiation In general, neoprene gloves are used on all boxes* They are made in two., grades either 0.010 to 0.019 inches or 0.015 to 0.022 inches thick. It has been found that HNOj and kerosene could attack neoprene gloves in such a way as to make them very sticky. This effect can be reduced considerably by giliing an additional pair of PVO or latex wrist gloves inside the box. Dipping gloves in butyl rubber is said to afford reasonable protection without reducing the ease of operation. The glove must be inspected daily to ensure its perfeotneaa and absence of contamination. The(glove should be changed either on account of Its defect or contamination problem. However* experience shows that it is preferable to change them regularly after one year period. During glove changing operation, respirators should always be worn as a precautionary measure* First the old glove is taken inside the glove box and new one is put on the port without removing the old one. Putting the hand from an adjacent glove port, the old glove is carefully , loosened from its 0-ring taking enough care that the new glove is fitting properly. Masking tape ia put over it. The old glove is then removed slowly and carefully into the glove box and new one is pushed insIda and 0-band fixed. Contamination oheck for air as well as surface is irade to confirm that no activity is released into the vicinity if the glove box.

3.2.7 Ventilation for glove Box The atmosphere in a glove box will be highly contaminated} hence efficient filtration system is required 35 to reduce aotivity levels in the exhaust-air, discharged to the environment. High efficiency particulate filters, at least three in eerlea should be provided to reduce the aotivity in the exhausted air to permissible levels. She first filter should be installed inside the glove box or near to it to prevent significant contamination of the exhaust duct. Air for the ventilation of the glove box is drawn from the laboratory through a high eff icienoy inlet filter; this prevents spread of contamination if the pressure in the box becomes positive accidentally and also prolongs the life of the exhaust filters.

3.3 Ventilation for Plutonium Laboratory

She laboratory handling Pu should be designed as a highly flexible faoility capable of handling all types of Plutonium bearing oateriala like metal, oxidea, carbide eto. All high active areas are to be designed so as to have at all times air changes between 10-20. Store areas or vaults should have 20 airchangae at all times and maintained at a negative pressure of 0.08" WO*

The ventilation system should preferably be of once- through type. Flow of air should be from zones of lower activity to zones of higher activity, with each higher aotivity ssone maintained at a negative pressure with respeot to the preceding zone. The inlet air should be filtered to reduce the dust load. Exhausted air should be adequately filtered through high efficiency filters and be released at ground level or through a stack depending on the degree of filtration and toxicity of the materials released to the environment. The filters should be installed before the exhaust fans to prevent lea&tge of dust from the ducts and filter housing. The ducta and filters should b« acid resistant if acid fuses are likely to arias, filters and their frames should be fire resistant* In-aitu tasting of filters should be conducted with suitable aerosols to ensure their performance. Emergency ventilation system should be provided to taka care of accidental situations.

3.4 Surface fflniah for Plutonium Laboratory

0) Epoxy coated resin paint is to be used in cell and high aotiva area, (2) Acrylic emulsion fungus resistant wall paints are to be used) (3) Stainless steel or ceramic flooring is to be used in potential active sinka and floor areas in decontamination rooms, (4) General floor aroa should use removable and acid-proof PVC tiles.

3.5 Pre-coauniaaiqninj; Checks for Hutonlum Laboratory Before plutonium laboratory ia cleared for any active work the following pre-cominissioning checks are to be undertaken*

3.5.1 J3r.ai4.iua cold runs. Using uranium instead of plutonium the cold runs are simulated. Air samples collected on filter papers, swipes collected on glove box panels, off gaa exhaust lines and floor areas are subjected to fluorimotric analysis of uranium. Tho data obtained are extrapolated to plutonium in terms of weight 37 and this is compared to its permissible levels. Proa the performance of these cold runs and operatlocal confidence derived from it, Integrity for plutonium work is adjudged.

3.5.2 Checking of drainage system of glove boxes The liquid waste from the laboratory is usually collected in waste collection tanks. The connections to the drains and thus the whole drainage system should be checked for operation of valves, non-leakage end also for proper connections. Then rhodamin dye diluted with sufficient water is poured into water lines and test carried out to confirm whether dye mixture is collected in the respective waste collection tanks. This would also indicate any mechanical faults requiring rectification, colour of the dye serving for physical identity.

4. CONTAWNA'PION EVALUATION, COJ3THOL AHD PR£V£NT1OH The two basic contamination control measures are based on (i) the detection and measurement of airborne radioactivity la working environment (il) monitoring of surface and personnel for plutonium. Because of the small range of alpha particle in air, monitoring of surface for alpha emittiry contamination Breaen* greater probleas compared to beta-gamma contamination. For the purpose of evaluation and control of health hazards in a plutonium laboratory the following measures are adoptedt (i) Film badge service including extremity monitoring, (ii) Radiation surveys of the laboratory areas,

(iii)Zoning of the laboratory into different contamination-potential areas to avoid cross contaminations, (iv) Air (r) Bioassay and whole body counting.

In order to carry out measurement and assessment of radiation data, the instrumentation and equipment provided in a high active laboratory should include a Central Badi&tion Health Console, Central air sampling system, portable and mains operated radiation survey instruments, fin detection systems and also criticality monitors. 4.1 Special Technique for Pu~Air Monitoring °'

One of the problems of air sampling in Pu-areas is the Interference caused by the naturally occurring alpha-emitters, radon and thoron daughters, present in air* Gross amounts of Pu can be detected immediately by measuring the changes in beta/alpha ratio of the air sample, beta/alpha ratio fcr natural activity of atmosphere air being 2»3. Use of imps otor air sampler in made routinely for Pu- air sampling. This doss not collect the natural emitter? "ich are usually associated in very snail, sub micron particles and are discriminated against. More refined techniques have been used to obtain information on the particle size distribution. Use of continuous plutonium air monitor is made for plutonium monitoring in th« laboratory. 4.1.1 Annular Impactor Collection ia based on the principle that Pu-partioles are usually larger than 1 micron size and comparatively heavy, whereas natural activity is found mostly associated with particles of less than 0.03 micron in diameter. Heavier particles art coll eat«d on an impactor disc while lighter and small particles are not collected. An efficiency at 90$ separation is obtainable which is sufficient to allow normal concentration limit of Pu to be measured in less than 10 mts. of counting time. 4.2 Surfaoe and Personnel Monitoring Monitoring of surfaces In the laboratory is very essential for contamination control measures, Monitoring of surfaces fo? Pu-alpha emitting contamination presents far greater difficulties than for beta-gamma emitters* The maximum permissible levels of contamination on surfaces for Pu- are given in Table-4.1. Two methods of surface monitoring are adopted and they are:(i) direct monitoring with alpha sensitive probes and (2) smear checks« Thorough area monitoring is very essential to control the personal exposure in the laboratory. Swipe samples are normally taken on pieces of filter paper by wiping over a given area of surface, usually some thing like 100 eg•cm. This will only give information on removable contamination* One very useful and practical form of contamination check is the monitoring the solea of ahoes when people leave the active lab. at a work break.

Direct Monitoring Use of scintillation counting set up ia made for alpha* monitoring. Floor monitorir^ is done by using a floor monitor which covers a large floor area as it consists.of a laige area scintillation set up mounted on wheels so that it can be pushed around. Clothings of personnel leaving active work are monitored by usual radiation monitors, particularly for alpha. Hand monitors ^ Pu- wound monitors are available for checking alpha contamination on hands and contaminated wounds respectively. 4«3 Decontamination 4»3«1 KLutonium Excretion Extensive biological data on plutonium has been obtained from studies on plutonium excretion in urine and faeces. 40

Table-4.1 (11)

MAXIMUM PfijWISSIHLE LEVELS OP SURFACE CONTAMINATION

Alpha Beta S.No. •Hace of conta- mination 2 sr dpVom

1 • General aurfaoe 2.5 2.5 50 140

2. Handa Nil 200 Nil 400

3. Skin Nil 800 Nil 40 105

5 4. Personnel clothing Nil aoo Nil 40 io

5 5. Company olothing Nil 2800 Nil 140 10

6. Personnel shoes 1*0 1100 200 280 1.4x105

7. Company ehoea 2.5 2800 50 700 3.5x1 Of* 41

4>3<1<1 Excretion of Absorbed Plutonium: Human excretion data have been obtained by Wright Langham after intravenous Injection of Plutonium citrate. The percentage daily excretion of plutonium was found to be beat described by power functions. The cumulative total excretion data obtained by tengham is shown in Table-4^2. Extrapolation of hie data leads to an excretion of 17.6$ in 50 years. Moat Investigators, have found that the pattern of urinary excretion closely follows a power function of time. 4.3.1.2 Kjxoretlon of Inhaled Klutoniumi Studies on animals have established that plutonlum after inhalation is mostly excreted in faeces and this has been confirmed from the data obtained from humans after accidental intake* Substantial differences exist in both faecal and urinary excretion and thia is attributed to the particle else and chemical nature of the inhaled material. 'Excretion rate of the smallest Plutonium oxide particles falls off less rapidly than that of larger oxide particles. 4.3.1.2 Bioaaaays Measurement of plutonium content in urinary and faecal excretions is the only method nomally available to estimate the systematic burden of plutonium after accidental intakes aa wall as routine exposures. However, Information can also be obtained from analysis on blood, sputum and nasal swabs or nose blows. Usually routine exposures are controlled by air sampling and -environmental monitoring. The main purpose of urine and faecal analysis is to detect unnoticed accidents and to investigate their severity. If .Langham'a equation for urinary excretion is adopted for estimating the systemic burden of plutohiuo, the daily elimination rate after two months Is only 0«01$ of the Table-4«2

ACCUMULATED URINARY JLU'o FdBCAL EXCRETION OP PLUTONIUM-239 BY MAN'4)

Time after Accumulated excretion administration (% administered doaa)

1 day 0,5

10 days 2.6

50 days A.O

100 days 4.7 1 year 6.5

3 years 7,8

5 years 8.7

10 years 10.0

20 years 12O2

50 years 17.6 intake. Por an intake of 0.04 uc (i.e. 1 MEBB), this corresponds to 4 pc/day. A person could have a large body content of plutoniua without any significant excretion in urine, particularly when Plutonium le in insoluble form and is retained in the lung. Amount a excreted in faeces, however, may be very high compared with that in urias. "Assessment of dose from Internal Contamination", ICHP Publication 10, recommends the following figures for investigation lerelat (a) Transferable - Pu-239 -4x10 uc in Bond ?u-241 - 2x1 o"3 p in Bone (b) Non-transferable - Pu-239 -6x10 uc in Lung Pu-241 - 0.6 pa in Lung

The urinary exoretion correspondii% to (a) oay be calculated from the formula suggested by Xtangham*

DJJ - 500 Of74

where Dg •» systemic burden at the time of intake U « dally exoretion in urine T days after intake T • number of days between bloassay sample and intake.

Waits of detection for radiochemical analysis, on urine aret Pu-239 (alpha) - 6x1 o"8 uo/24 hours Pu-241 (beta) * 1x1 o"44 po/sample • 1x1 o""11 ucAitre of uxim if it la assumed that 1x10*** uc is in 1 ml. 44

4.3.1.4 In Vivo Measurementt L X-raya from uranium, the daughter product of plutonium and 60 kev gamma rays from Americium-241 can be detected outside the body following an inhalation ov injection through a wound aite. Wound monitoring: Small area scintillation crystals can be uaed with good sensitivity for measurement of plutonium In punctured wounds. Sensitivity falls off as the depth of Plutonium in the wound increases. It is possible to measure 2 nc of Pu-239 by its X-ray emission, at the surface of the wound. Wound monitoring Instruments should be capable of rapid assay of Pu in the wound and for final assessment of the amount remaining in the wound after decontamination and surgical operations. Plutonium wound contamination is often accompanied by akin contamination and this should be taken into account in wound monitoring. Lung burden estimates* Lung burden measurements have been mads with two types of detectors:- the thin Nal crystal and the gas filled proportional counter. Large area Nal crystals, as thin aa 0.25 cm, have an efficiency of 90^ or more at plutonium X-ray energies and comparable efficienoiee can bo achieved with proportional counters, having large area (detector) windows. Currently the limit of detection by those counters is in the range one, half to one maximum permissible lung burden»but for radiological protection purposes it la desirable to have a limit cf about one tenth of the maximum permissible burden. In cases where sufficient Am-241 is present in the plutoniun, then Pu-Am-241 ratio is known and the limits of detection can be lowered down to 0.5 - 2 no. 45

4.5.2 p

4.3.2.1 External Decontamination ' 't Stepwise adopt th* following methods?

i) Thorough, washing with detergents and water is the beet general method for decontamination of bands and other parts of the body, regardless of the contaminant.

Wash for about 2 to 3 ainuteo with detezgents with a good lather in tepid water, covering the whole of the affected area. Mild detergents like Teepol, Aoinol-ff, soap impregnated tissues etc. can be used. Speoial attention should be given to the areas between fingers and around the nail grooves, because the outer edges of the hands are readily contaminated and are often negleoted in washing* Binse and wash with pure water, monitor and repeat if necessary.

ii) Scrub with soft brush using detergent and tepid water* Brush should be applied lightly so that no scratches are produced in the skin. Scrubbing for about 5 minutes Is sufficient.

iii) Alternately, use a mixture of liquid soap and oorn-aeal.

iv) Soak the contaminated area with a 5% solution of ammonium citrate or citric acid* Do not use brush. A 2jL solution of 1t1 mixture by weigr< of tartario acid and citric acid may also be used. This will be useful in the oase of many alpha contaminations, particularly Piu

v) Washing with 10$ solution of ethylene diaaine tetra acetic acid (SDTA) mixed with detergent to give gcod lather gives a very high decontamination efficiency in most of the cases. Use EDTA-aoap mixture and wash with pure water. 46

vi) Mix equal quantities of saturated solution of potassium permanganate and 0.1 N sulphuric acid. Pour this over wet hands (for baud contamination only), rubbir^ the entire surface uuiqg a hand bruah. Treatment should be restricted to short tiros - 2 minutes at the most - because this application will remove u layer of akin if allowed to remain in contact for long. Flush v»ith warm water. To remove the atain apply freahly prepared !?$ solution of sodium bisulphite (NaHtfO ) in the same aemner as above. Thee wash with soap anl water. Apply a thin layer of hsnd cream. The treatment is quite effective in the case of plutonlunt contamination.

Tii) If eyas get contaminated wash with water Immediately. Syn lotion has to be applied and medical care has to be given to the person at the earliest.

viii) In case of minor wounda, the following first aid should

be biven:- Wash the wouM with large volun&s of running water iiwnediutely. Li^ht tourniquet action to stop venous return E»y be desirable to stimulate bleeding. Wash with soap and water. Monitor and check for contamination. Do not use complexity agents for washing I they make tha contaminant penetrate easily into, the body. In the case of major wounds, treatment should be given, by the radiological safety physician.

4.3.2.2 .Internal Decontaminatiom Internal decontamination ia the work of radiological safety physicians. Hence one should not attempt this without a thorough knowledge of the treatment and the toxicity of the reagents used. Mention of general principles of interml decontamination is made* •17 a) Treatment for Inhaled Hadioisotopeat

for the insoluble radionuclides retention in the body depends on the particle size of the inhaled substance. If the biological half life of contaminant ia considerable, absorption into blood takes place slowly. In such case and also for inhalation of soluble substances, the treatment with chelating agents, to increase the urirary excretion und to prevent bone deposition has been found effective. b) Treatment for Orally Ingested Badionuclides:

There are four approachec to this treatment:

(i) use of an emetic, (ii) use of selected laxatives (ill) use of precipitating agents (iv) isotopic dilution

Oral administration of synthetic ion exchange compounds has shown promise in the decontamination of QI tract.

c) Treatment for Absorbed Radioiaotopee:

Treatment with complexing agents ia now becoming increasingly important. The important complex ing agents experimentally examined are given below:

1. Intravenous administration of sodium citrate increases the urinary excvetion of pi titanium, strontium and thorium manifold. The difficulty is that the citrates undergo metabolism and thus loss effectiveness in the system with time.

2, One of the important chelating agents used is EDTA. This forms very stable ohelatea with polyvalent cfcvtions. For effective removal of deposited plutonium, •'njection of calcium salt of EDIA is carried out. Combined treatment with Vitamin A and Ca-EDTA ia effective in increasing the rate of excretion of Fu and other isotopes deposited in the body. 48

3. Anotbar widely used deeontamlnant known aa HTPA (Diethyleut triamine penta acetic acid) ia a most effective chelating agent, for increasing the elimination of Pu and many fission products.

4. Zirconium citrate is also used as a chelating agant. 95% of injected airoonium comes out in the urine, while about 5% ia retained In skin and muscle* Combined treatment with Ca El/TA and Zirconium citrate has been found more effective in decreasing the tissue and bone deposition of Pu. TTA and BAL are also some times used.

5. PLUTONIUM FIRS SAPiSTY^4* 3'

The most serious accident that can occur in a glove box is a solvent fire or gas explosion. Experiments have shown that glovea can be blown off within seconds at the ignition of only 100 ml of inflammable solvent and whole panels can be blown out if all the port bungs are in position.

Because of the possible serious spread of contamination in tha vent of a fire in the plutonium laboratory or store room, it la racoiBfiiendtd that all precautions practical ba taken to prevent fires. The use of fire resistive construction, metal furniture and equipment, fire detection and alarm systems, fire resisting walls and doors and emergency procedures including possible evacuation should be considered*

Although fire detection and fire fighting system will b« of ^raat help in controlling fires, it is necessary to take measures that will reduce the chances of any fire taking plaoo. Some measures for this purpose are mentioned below*

i) Good house keeping ii) Use of proper types of dosed containers for various hazardous and inflammable ohemicala, 49 ill) Provision of spark proof type electrical fittings with enclosed electrical wiring wherever a fire hazard exists, iv) Storage of incompatible materials in separated places. If this is not possible, some shielding should be interposed between the fire incompatible materials.

5»1 Plutonium Metal Fires Plutonium metal, because of its pyrophoricity, can ignite spontaneously in air at room temperature both in finely divided and massive form. Plutonium fires generally take place without visible flame and are characterised by comparatively alow combustion accompanied by local emission of intense heat and brilliant white li&ht. Though the white light may be partially masked by the oxide coating, the intense heat generated is often sufficient to causa the melting of stainless steel. The melting of stainless steel may be partially due to formation of eutectic alloys having a lower melting point and higher pyrophoricity than plutonium. Apart from the possible loss of material involved, plutonium fires are of concern because of the following considerations. l) Critical mass* When the Quantity of metallic plutonium involved in a fire exceedB "always, safe" quantities, a self sustaining nuclear chain reaction may result from ehangeB in metal configuration and/or as a consequence of neutron moderating effects of materials applied to fire. ii) Personal hazards* The extreme toxicity of plutonium oxides that may bs generated and dispersed during a plutonium fire tends to complicate fire control and extinguishment measures. ill) Decontamination* Jhe expense in time, money and man-power, necessary to decontaminate an area in which a plutonium fire has occured is extremely large, compared to clean up expense following a small fire from other metalB. 5,2 Protective Measures ff r Plutonium Fires

The following protective measures ahould be taken in enclosures where plutonium metal is handled• i) Plutonium metal should be handled in moisture free argon atmosphere. ll) The base of the enclosure, containing plutonium metal, ahould have graphite lining of about half cm. thickness, to prevent plutonium metal from coming in contact with stainless steel in case of fire. iii) Low melting euteotic salt mixtures (35 wt.$ sodium chloride + 40 wt.fS potassium chloride + 25 wt.?S barium chloride) in Bealed plastic bags should be kept ready in the enclosures. In case of metal fire, the bag ahould be placed directly on the fire to melt the plastic so that .the extinguishing agent flows over the burning metal.

IV) Stainless ateel container with loose fitting cover, lined ineide with graphite ani partially filled with eutectic chloride mixture, should be kept inside the plutonium metal handling enclosures, along with two stainless ateel spoons lined with graphite. In case of. plutonium fire it should be possible to transfer the burning metal into the stainless ateel container , with the help of acoopa. v) A dry powder extinguisher system, consistirig of euteotic salt ao extinguisher and argon as expellant, should be connected to the plutonium metal handling enclosure through a nozzle and a diffuser. In cage of fire, on pressing the trigger, argon gag ia released which in turn expels the powder through the diffuser. 51

5.3 Pira Safety ±u Storage, Handling; and Shipment of Plutonium Metal(in quantities larger than "always safe" anpunta}

Bach problem of thia type will require separate evaluation from a fire stand point of view but a few precautions can be mentioned as given below:

l) The amount of combustible material in the immediate vicinity of plutonium metal should be held to a minimum.

ii) Periodic inspection and thermistor fire detection devices should be provided in plutonium store.

iii) Non combustible physical barriers should be provided to limit the maximum quantity of plutonium that can conceivably get involved in a single fire, to a quantity which if lost, ia acceptable as a known calculated risk.

iv) In providing the physical barriers, consideration should be given to possible changes in metal geometry from fire,accumulation of plutonium oxide and the effeot of various fire extinguishers that might be used.

v) Where warranted, the possibility of providing autoiratic fire extinguishment through inert gaa or dry powder flooding should be studied.

6. GfilTICALITY SAFETY

When plutonium is handled in large quantities such &a in fuel processing or fuel fabrication plants, a potential criticality hazard exists. The most1 important aspect of criticality control is accident prevention because of the potentially serious consequences that can result. Kuclear safety must be evaluated not only for the normal operating conditions but also for those off-standard conditions, which are physically possible and which

might be conducive for a nuclear chain reaction. b>^ Criticallty Parameters

For a self sustaining nuclear chain reaction a certain minimum quantity oi" fissile ne-terial, refered to as the critical ma38 is required. The critical mass depends very strongly on a number of fact • • :.::Y,>V:

(i) leakage of iu-utrons froic I;: j.jtem which in turn depends on the size, shape and composition of the. 3yatem and on the neutron reflecting properties of surrounding materials,

(ii) presence of neutron moderating elementsj particularly hydrogen, mixed with the fissile material,

(iii) presence of any neutron absorbing materials like boron, cadmium etc. and

(iv) neutron interaction with other fissile material systems*

Por a given volurte, sphere h«3 the least surface area- to-volume ratio and hence a spherical system can go critical with a minimum m* JS r-i" fissile material. Nuclear Bafety can be ensured by limiting container dimensions, with a large surface area-to-volume ratio-, thereby giving sufficient leakage of

neutrons to maintain ths gvatem subcritical.

Secondly5 v- iv.,, VOJlection should be minimised as much aa possible, ik-naally reflection is provided to some extent by the contain-rvv wall itself. Provision of a cooling jacket to the vessel ov placement of the equipment near other process vessels or structural materials like concrete wall, will provide additional neutron reflection. Though materials like beryl>lium, heavy water etc. are very efficient reflectors, their use is quite uncommon in fuel processing or fabrication plants. Generally, in criticality evaluations, water reflection is taken into account, owing to the possibility of flooding in the fissile material handling areas. About three inch thickness of water provides the maximum reflection and tnere is little change in the critical nasa ii' water thickneBS is increased beyond this.

Critical mass is a very sensitive function of the hydrogen density in the mixture. Hydrogen concentration in a mixture ia generally expressed as the ratio of the number of atoms of hydrogen to the number of fissile atoms (e.g. H/Pu). This ratio may range from zero for metal or unhydrated salt to several thousand for very dilute aqueous solutions. Over the &bove concentration range the critical mass of plutonium varies from a few kilograms through a minimum of few hundred grams, to infinity for very dilute solutions, where neutron absorbtion by hydrogen makes chain reactions impossible. On the other hand, critical volume (or dimension) continuously increases with increasing hydrogen concent rat ion. For aqueous Pu solutions, minimum critical mass of plutonium is 510 gms and this occurs at H/Pu«' 750 (^ 33 gms PuAitre).

Nuclear safety of an equipment can be assured through the use of soluble or fixed (non-soluble) neutron absorbers or poisons. Soluble poisons are highly effective in that homogeneity can usually be assured and that snail amounts of poison are enough to ensure safety. Soluble poisons lite Gd nitrate* boric acid, sodium borate, should be added only under effective administrative control. Fixed non-soluble poiaons like borated Pyrex glass, raechig rings etc. are particularly advantageous, as no process step is required to remove the poison from the final product of the fissile material. Fixed poisons give rise to heterogeneous distribution of the neutron absorber and as 3uch are leas effective than soluble poisons. While Using these poi3ona the designer should ensure himaelf about the integrity of the poi3on against chemical attack or mechanical dialodgement, and particularly for soluble poisons, where some chemical reactions may selectively precipitate the poison.

The effect of neutron interaction between subcritical unit3 in an array must be considered to .-nsure the nucleBr safety of the array as a whole. Adequate spacing between the fissile units and the maintenance of the spacing under all conditions should be considered in the design of plant facilities and particularly in the storage and transportation of the units.

6.2 Methods of Criticality Control

The various methods of criticality control in the handling of fissile material are aa follows:

6.2.1 Mft33 Control;, The mass of fissile material handled in a single operation is restricted to a certain upper limit. Such a limit 3hould provide a suitable safety factor, which in most oases allow, for double batching.

When large quantity of fiasile materials are to be handled, mass control may be an unnecessary restriction and • it may be .essential to U3e other methods of controL. fc.2.2 Concentration Control! In operations involving aqueous solutions particularly those containing low concentrations of fissile material, Bafety is ensured by imposing a limit on the concentration or density of the fis3ile material in solution. 55"

6.2.3 _fa>lun»e Control; It ia possible to use vesselg of restricted volume . to ensure safety. For solutions, the minimum volume occur at high concentrations of fissile material, 5.2.4 geometry Control; iiliuipments can be Ae&*—' to be safe by virtue of their geometry, where the neutron leakage is so high aa to prevent criticality, regardless of the quantity of fisaile material contained therein. Examples of such systems are (i) (an infinite) long cylinder which cannot go critical, its diameter 1B Ie33 than a certain limiting value and (ii) (an infinite) slab, with the thickness of the slab less than a certain limiting value. 6.2.5 -Addition of Nuclear Poisons; Soluble or fixed neutron poisons may be used in some cases to increase mass limit or to make an equipment safe by geometry. 0,2.6 Spacing of Equipments; Restrictions are placed on the amount and form of TSterial permitted in a given storage array and on the distribution in space of the material contained within the array. 6»3 Safety Factors fiuclearly safe limit for an operation is obtained by the application of a suitable safety factor on the critical parameter of interest. The safety factor should provide allowances for uncertainties in the data used in the calculations, inhoffiogeneties in the medium due to varying concentrations, errors or inaccuracies in the sampling, batch doubling due to nsaloperation on the part of plant personnel or plant control mechanism and other unforeseen circumstances.

Safety factors recommended for criticality control in U.K. are listed helowr

Values of safety factors on mass on concen- on dimensions tration sphere Infinite infinite radius cylinder slab diameter thiokneaa

Initial safety 2/3 2/3 (2/3)1/3 (2/3)1/2 2/3 factor

Inhomogeneties 3/4 3/4 (3/4)V3 (3/4)V2 3/4

Errors in 3/4 3/4 (3/4)V3 (3/4)i/3 3/4 sampling

Tho safety factors are cumulative; when batch doubling is possible, the concentration or masses should be halved to take care of this. Safe operating mss limits adopted in USA include a safety factor of 2.3 aa a safeguard against double batching and volume limits include a safety factor of 1.33 and equivalent margins appear in dimensional limits.

With regard to shipment of fissile materials, IAEA recummends that where mass is the controlling factor, the permissible value in any single package must not exceed 80$ of the critical mass under the conditions of packing, with due consideration for built-in neutron absorbers and where geometry is the controlling factor, the permissible value for each controlling dimension muat incorporate a 10J£ Bafety factor. For shipment of fuel elements, the effective neutron multi- plication factor of the system shall not exceed 0.9, under all conditions.

6,4 Criticality Data

6.4.1 Pu Solutions and Metal Basic nuclear safety parameters for aqueous homogeneous Plutonium solution and plutonium metal, under water reflection are listed in Table-6.1* '. Safety factors employed in deriving the safe limits are those adopted in United States.

6.4.2 Plutonium Oxide Table-6,2 gives the critical and safe parameters for dry plutonium oxide^ ' (H/Pu » 0) (The safety factors employed differ slightly from those used in Table-6.1).

fe«4.3 Other Pu Compounds^ ' Table-6.3 givea the critical masses of other plutonium compounds for unreflected and water reflected spherical systems. (H) 6.4.4 ,AnniL}.ar Cylindrical Geometry* A method to increase the storage capacity for aqueous plutonium solutions (df any Pu concentration) is to use an annular cylindrical tank, with a built-in neutron absorber. The inner cylinder is lined with cadmium of 20 mil thickness (0.5 mm) and is filled with water or materials of equivalent hydrogen density. For plutonium solutions recommended aafe annular thickness, under water reflection external to the cylinder, is 2.1 inches, for any combination of inner and outer radii. If external reflection is provided only by the Table-6.1

CRITICAL AND SAFETY FARA'TETKKS FOR PLUTONIUM SOLUTION AMD METAL (UNDER V/WEK REFLECTION)

Control Parameter Minimum Becommended critical value value

Uaea (kgms) Solution 0.55 0.22 Metal 5.0 2.6 - oC phase 7.6 3.5 - S phasa Diameter of Infinite cylinder(inches) Solution 4.9 4.2 Metal 1o7 1*4 - <£ phase 2.1 1.8 - £ phase Thickness of Infinite slab(inches) Solution 1.3 0.9 Metal 0.24 0.18 - i£ phase 0.28 0.22 - S phase Solution Volume(litres) 4.5 3.4

Concentration in Aqueous 7.8 6.9 solutions(gma of Pu/litre) Table-6.2

CRITICAL AND SAFE PARAMETERS TOR 1)RY PLUTONIUM - (PuO,)( TINDER WATER INFLECTION)

Control parameter Critical value Safe value

Mass (kgma) ,(16) 5*3

Volume^15\litrea) 1,18 0.94

Dlaraeter.of5 infinite 8.3 7»6 cylinder^' ) (cms) (approximate) (approxiaBte)

Thicknesg of infinite 1.9 1.75 15)() 60

Table-6.3

CRITICAL MASSES^' OF PLUTONIUM COMPOUNDS (H/PUPO) (SPHERICAL SYSTEMS)

Compound Density (grams Critical of compound Bare water reflected per c.c)

PUO, 11.46 24.5 12.2

TuN 14.25 18.? 9.23 PuO 13.6 17.93 9.07 10.4 15.64 7.49

+ PUH3 9.61 12.69 6.32 12.7 19.72 9.96 ^0^ 11.47 24.95 12.05

PuCl3 5.7 167.05 61.59

PuP3 9.32 32.59 16.0 PuF, 7.0 4 56.29 25.29

Pu(0204)2 4.5 152.0 66.07 **(»,)„ 6.2 103.96 54.94

1. * H/'Pu = 2 2. + H/Pu = 3 61

vessel walls ( 1/8 in thickness), recommended safe annular thickness ia 3.0 inches. There ia no restriction on the height of the cylinder.

6.5 Soluble Poiaona^14'

When nuclear poiaons are mixed homogeneously in solution, Plutonium and cadmium or its nuclear equivalent should be present in equal molar quantities.

6.6 Solid Poisons^14^

When solid poisona are uged for plutonium solutions with concentration upto 25 g Pu/iitre, the neutron absorber must contain at leaat 4 weight percent boron or its nuclear equivalent and occupy 17.5 volumes percent of the vessel and be uniformly distributed throughout the volume -

Safe concentrations of aqueous plutonium 3olutiono- ' in tanka of unlimited aiae, packed with 1.5 inch diameter (O.D.) boroa11icate glass raschig ringa are presented in Table-6.4. Raschig rir\j ia a small hollow cylinder having approximately equ'3.1 length and diameter. These shall bo 11.8 to 13.8 wt 'fo BoO, in the glase and boron in the ylaaa shall contain a * •* 10 10 11 concentration of B isotope auch that the B to B atom ratio is not leas than 0.24. (14) 6«7 Pipe Intersections

Recommended safe 3izes for intaraecting pipes carrying plutonium solution are ^iven in Table~6.b. If » pipe has multiple intersections no two intersections Bhould occur within 18 inches,(axis-to-axia) from one another. (14) 6.8 Storage of Pu Metal, Compounds and Solutiona^ A typical storage array recommended for Pu metal, compounds and solutions, consists of spherical fisaile units 62

Table-6.4

SAFE CONCENTRATIONS OF AQUEOUS PLUTONIUM SOLUTIONS, POISONED WITH BOROSILICATE GLASS RASCIIIO HTNR3*

Pu-240 Minimum Maximum Minimum content glass content safe Pu H/Pu ft i* of vessel concentra- atom ratio (vol. $) tion (gA)

5-10 24 1B0 135

10-15 24 200 120

15-20 24 225 105

5-10 30 260 90

10-15 30 315 75

15-20 30 380 60

•Maximum outer diameter - 1.5 inches 63

Table-6.5

SAFE INSIDE PIPE DIAMETERS POH INTERSECTIONS CONTAINING PLUTONIUM SOLUTIONS (H/Pu 20) (UNDER WATER REELECTION)

Type of intersection Inside pipe diameter (inches)

BLL3 4.0

TBB3 3.6

CROSSES or WYES 3.4 64 having their maximum sizes as given in Table-6.6.

The permissible apacirgs for the maximum sized units, specified in Taole-6.6, are listed in Table-6.7, when the units are arranged in linear or plane arrays. Figure 6.1 specifies the allowable number of maximum sized units, in a cubic array, for various Bpacings between the plutonium units.

Arrays in which the units meet the spacing criteria of Table-6.7 and Figure 6.1 may be considered isolated, when separated by a layer of conorete or water at least 8 inches thick.

Two plane or cubic arrays may be considered isolated, if the surface-to -surface separation ia greater than larger of +he following quantities:

(i) the maximum dimension of either array (ii) 12 feet

Two linear arrays are isolated, regardless of length, if their separation is at least 12 feet.

Arrays which are not isolated, are considered associated if the minimum surface-to-surface spacing is at least 7.5 feet| if the spacing is less, they are to be regarded as a single array.

In case of storage of plutonium solution (any concentra- tion) in cylinders of diameter not exceeding 5 inches, the recommended minimum oentre-to-centre spacing of the cylinders in a linear arraj is 24 inches? and for two associated linear arrays, the surface-to-surface spacing between the cylinders in each array is 24 inches.

There is no restriction on tne storage of plutonium solution at concentrations not exceeding 6.9 gms Pu/Litre (of.Table-6.1). SAFETY FACTOR 2

MINIMUM OPEN SPACE BETWEEN UNITS (THE EQUIVALENT IN SMALLER UNITS MAY BE CONSIDERED A MAXIMUM UNIT}

NOMINAL REFLECTOR ABOUT ARRAY (AS IN LARGE ROOM )

CURVE A THICK, CLOSE-FITTING REFLECTOR ABOUT ARRAY

10 0.3 0.7 1 13 15 7 10 LATTICE VOLUME PER MAXIMUM UNIT, FT. 3/UHIT I I III 8 10 12 15 18 20 24 30 36 CENTRE-TO-CENTRE SPACING IN CUBIC ARRAY, fN.

FIG.6-1- ALLOWABLE NUMBER OF MAXIMUM SIZE UNITS (OF TABLE 6-6) IN A CUBIC ARRAY. 66

Table-6.6

MAXIMUM SIZES OP SPHERICAL UNIT.'3 TO WTITCH STORAGE LIMITS APPLY

Material type Maximum sized unit

Metal, compounds or mixtures . 0.5 t Mass limit 4.5

Metal, compound or mixtures 0.5 <1H/PLI 4 2 Mass limit 4.5

compounds or mixtures 2^H/Pu<20 Maaa limit 2.4 kg

Solutions or hydrogencous mixtures, 204H/Pu voli.mie limit .2.4 litres

*Thia limit is for Pu metal at =19.6 g/cm For the alloy at =15.8 g/cm , the limit is 6,0 kg 67

Table-6.7

LIMITS FOR STORAGS OF MAXIMUM SIZED UNITS

Minimum centre-to- Number of units Type of Array centre Bpaoing* of per array unite of maximum size (inches)

Isolated linear or 16 No limit plane array

Two or more associated 30 120/array; 240 total plane arrays 24 90/arra-yj 180 total

20 50/arrayj 100 total

* There must be at least 8 inches open apace between maximum units. 68

b.y Transport of Plutonium Metalt Compounds and Solution '

Table-6,8 gives specific recommendations for controlled shipment (i.e. no off-loading and reloading enroute is allowed and planned arrangement of the cargo will be maintained) of maximum sized plutonium unite defined in Table-6.6. The total amount of Pu in a single shipment ahall not exceed £±fty of the units. Integrity of the Bhipping cases or spacers should be assured in the case of an accident.

6.10 Magnitude of Criticality Accident

Magnitude of a oriticality accident is generally expressed as number of fissions taken place during an excursion. A review of the various accidents^ ' that have taken place in reactors, fuel processing plants and experimental laboratories indicates fiasion yields ranging from 3 i 10 to U x 1020. The course of events in an excursion depends on whether the system is a solution system or a solid one, and the excursion is limited or terminated by different mechanisms inherent to the system itself. The four natural limiting mechanisms are (l) thermal expansion (2) rise in neutron temperature (3) boiling and (4) raicro-rbubble foraiation from fission products.

In a solution system, with the initial burst, lasting a few mini-seconds, loicrobubbles are formed, which reduce the density of the system and thereby the reactivity of the system is ..educed. When the reaction dies down, the bubbles escape and a further pulse of activity may occur) after perhaps a number of such pulse3, loss of water moderator reduces the system to a subcritical state. In a liquid system, only a small amount of the total yield is associated with the spike (i.e. initial burst) and the rest with the plateau. While exposure of persons 69

Table-6.8

LIMITS K)R CONTROLLED SIIIMBJTS OP HOTONTM TOTITS DEFINED IN TAELE-6.6

Type of Material Maximum density Norn&l car load or established by truck load limit Spacers* (50 maximum units)

Metal,compound or 1 kg/ft 225 kg mixture H/*u ^ 2

Hydrogeneous compounds 0.5 kg/ft 120 kg 2

Solutions or hydrogeneoua 0.5 litre/ft 120 litres mixtures 20

"Spacers shall establish at least 8 inches open space between the units. to the radiation dose from the spike cannot be prevented, a large fraction of the total dose can be avoided by running away quickly.

In a solid system, the spike contributes nearly all the yield and there is little or no plateau. The shut down mechanism in a metal system is only thermal expansion. The excursion may be over in milliseconds and running away is not as effective as with a solution system. Due to spontaneous fission of plutonium, fission yield from a plutonium system is lesa by more than an order of magnitude than for a corresponding uranium system.

The ratio of neutron to yamma radiation dose also depends on the system. In a solid ey3tem, the ratio can be as hijjh aa 10 and for a solution system, it can be as low as 0.1.

Prompt neutron and gamma dose rate at exterior of an ordinary concrete shield , from a nuclear excursion of 10 fissions are presented in Table-6O9.

Provision of alarm Instrumentation, based on gamma or neutron radiations levels must be considered in all areas, where a potential criticality hazard exists. The instrument should be able to give alarms even for low yield excursions, resulting from very small additions of reactivity. At the same time, the alarm level should net be set so low., as to give false alarms during normal movement of sources within the laboratory.

The sensing device, in U3e at Trorabay, consists of an ion chamber, connected to a period amplifier. The amplifier gives an indication of the rate of rise of gamma field. The criteria for alarm settings of the system have been fixed as follows: 71

Tabls-6»9

NEUTRON AND GAMMA DOSE THHOUGH COHCRETE SHIELD mm A NUCLEAR Exorasiotf OP 10 FISSIONS

Concrete shield Dose at outer side of shield(Rem) thickness (ft) Metal system Solution system

1 88,000 5f2OO

3 317 23

4 17.0 1.9

5 0.960 O.U

6 0.059 0.012

Dose rates can be calculated for other fission yielde by direct proportion. (i) the system shall Bound a positive alarm if a criticality burst of 10 -* fissions occurs at a distance of 30 feet from the detector and delivers a prompt gamma dose in 100 milliseconds,

(ii) the system shall not give alarm as a result of handling 10 Ci Co-60 source at a distance of about 10 feet from the chamber.

The above criteria will be satisfied', if a change in radiation level of 6 decade3 (i.e. 10 mB/hr background to 10 R/lir), triggers the alarm, i'his has been taken as the alarm setting for the criticality monitoring system.

6.11 Administration of Nuclear Safety

In order to ensure that safe practices are followed in fisBlle material handling areas, it is necessary to establish administrative controls of nuclear safety. Safety aspects of all operation must be evaluated and approved by a criticality safety committee. Written procedures covering the various phases of an operation including the nuclear safety precautions to be followed, should be prepared and aj.1 operators should be familiar with those procedures that wncum them. It is desirable to have the operators sign, a check-off sheet on which, the special controls to ensure nuclear safety are listed. Responsibility for any operation must be clearly defined. Safety measures, however carefully conceived, are not absolutely fool-proof and hence planning is essential to meet any emergency and written procedures for emergency action should be prepared. 73

REFERENCES 1. Unruh CM. The Radiologioal Physics of Plutonium, Proceedings of a Symposium on Plutonium aa a Power Reactor Fuel, American Kuclear Society Topical Meeting, Richland, Washington. HW-75OO7, Sept.13-14,1962.

2. Appleton G.J. and Eunster H.J. Recommended Safe Fraetice in the Safe Handling of Plutonium in laboratories and Plants AHSB (RP); 1961.

3« Arnold E.D. Radiation Limitations on the Recycle of fuels. Geneva Conference on Peaceful Uses of Atomic Energy, P/1838, Vol.13 (1958).

*>* Lister B.A.J. Lecture Notes on Health Physics Aspects of Plutonium Handling. AEI2B-L-151, (1954).

5. Birchall.D. Hadiation Dose Hates, for Plutonium Isotopes AHSB(S) E-1O, 1960.

.6, Shupe M.W., MullinsL. J., Morgan A.N., OonsaleB A.L., Ogard A.E., Valentine A.M., and Leary J.A., Operating Experience on an Economical Plutonium-238 Processing facility. Proceedings of a symposium, on Radiation Safety Problems in the Design and Operation of Hot Facilities, Saclay, Erance (lAEA-SM-125/45) 13-17 Oct.1969.

7. Recommendations of the International Commission on Radiological Protection- Supplement No.6, British Journal of Radiology(1955).

8. Report of the Task Group on lung lynamics, Health Physics 12, 173 (1966).

9# Walton G.N. Clone Boxes and Shielded Cells for Handling Radioactive Materials, 1958. Butterworths Publications, Xondon.

10. Stevens D.C., Air Sampling with the Annular Impactor AERE/HP/M-IIO. 74

11. Manual for Radiation Protection in AEET. AEET-223, 1965, Issued by Health PhygicB Division,BARC.

12. Bhat I.S.f Kamath P.R. and Somasundaram S. Personnel Decontamination - A?iET/H.P./SM-1, (1958).

13. TAinster H.J. and Benuclick E.J. The Handling of Plutonium in laboratories: Precaution, Atomics Vol.6, Page 312 (1955).

14. Callihan A.E. et al, Nuclear safety Guide, TID-7O16 (Pev.1) (1961).

15. Vi'oodcock a.H. , Fundamentals of Criticality Control, Proceedings of a Symposium on Criticality Control in Chemical and Metallurgical Plants, Karlsruhe,(1961).

16. Hansen L. P. and Clayton R.D, , Criticality of Plutonium Compounds in the Undermoderated Range H/Pu 20. Nuclear Applications, Vol.3, NO.8,(1967).

17. Schuske a.L. ct al; Use of Borosili cate C.ass Raschig Rings as a Fixed Neutron Absorber in Solutions of Fissile T/laterial (Proposed Standard) Nuclear Engineering Bulletin, Vol.3, No.3,(1965).

18. Stratton >7.R. A Reviev/ of Criticality Accidents, Proceedings of a Symposium on Criticality Control in Chemical and Metallurgical Plants, Karlsruhe (1961).

19. WicholB J.P. Soluble Nautron Poison as a Primary Criticality Control in Shielded and Contained Radiochemical Facilities, ORNI-3309 (1962). APPENDIX-I

DESIGN DETAILS OP A TYPICAL INITtT AT,:0STHHiE GLOVE BOX FOR PLUTONIUM HANDLE 0

1. Construction details:(A) Dimensions: Single module box: 1094 mm length 1094 mm height 637 mm depth Larger Box: a) 2188 mm length 1094 nan height 637 mm depth b) 2188 mm length 1094 mm height 1094 mm depth (B) Materials used: Main frame and the floors S3 304 formed by pressing 1/8" thick sheets. panels, the ceiling.front and back: 10 mm thick laminated safety glass} preferable to commonly used transparent perspez being combustible from safety considerations. Side panels: 12.5 mm thick 38 Aluminium. Aluminium plates machined to 10 mm thickness all around periphery.

2» Glove port-rings* Sizes: Lower:- 8" dia for normal and regular used. Upper:- 5" dia mostly for filter changing operation. Material: Alloy of aluminium and \2^> silicon and moulded round U-section neoprene gaskets.

3. Glovea used: Size: 0.8 mm thick aai 800 ram long with headed cuff. Make: 'Milled1 neoprene gloves 4* Filter box; Housing: Aluminium frame work. JSaay to replace inlet and outlet filters from inside glove box. Single hand operation using upper glove port is possible. 5. Filtera(Inlet ft Outlet)t Efficiency and typet Absolute filters of the MSA honey comb type with an efficiency of 99.9"$ for 0.3 u particles. Single flange type with gaskets on both sides of the flange. Capacity: 50 ofm with initial pressure drop of 0.9" WO and temp.rating of 500 °F.

6. Transfer portt Funet iont 1) bag-out port In a free standing bo*. 2) for a train of boxes, as an inter- connecter for adjacent boxes through an aluminium tunnel(connecting link)* PVC-covered short aluminium tunnel with individual doors on either side for each one of the boxes. 7* Glove box services Electrical supply8 Through copper conductors moulded in a backelite disc which is sealed to outside panel by neoprene '0' ring. Glove boxes should always be earthed. Lighting: Fluorescent lights from top windows. Water supply, gas supply t Through four SS tubes welded to the box frame top. Self sealing couplings are used wherever required. Lines are marked for identification* 8. Air lock system: Each Glove Box at the beginning of each train has an airlock for transferring material into the train. Airlock chambers should also be evacuated and filled with argon. 9. Ventilation Glove Eox under argon atmosphere. Separate argon recirculating and purification eystem for each laboratory is required. 77

10» Pressure control system: 1) To maintain box pressure between (-} O«75" and (-) 1.0" WG.

a-ji Magnehelic gauge is used to monitor '•boA pressure • 3) Mechanical praasure controller will maintain glove box pressure constant during normal operation.

4) Purified gas entersthe glove box through an isolation valve, a rotanster, an Inlet butterfly valve located in the pressure controller, a diaphragm valve, an absolute filter and then?a large pipe for distribution. 5) The exhaust gaa will go out through absolute filters, a diaphragm valve, a three way solenoid valve, the outlet- butterfly valve located in the pressure controller aai isolation valve at the top.

6) Both butterfly valves are ltept open at a aet angle to allow desired gaa flow. The solenoid valve in the exhaust-line will be kept open by being energised.

7) Mechanical pressure controller will automatically stabilise the glove box pressure in cage of a pressure suiige due to glove movement or fluctuation in gas flow.

11 a Emergency HbLiauat System: 1 ) Design capacity of exhaust fan: Nearly 200 cfm at (-) 10" WG suction of the fan.

2) Design feature: An emergency control syatem will come into operation when the mechanical pressure controller becomes in effective to arrest the abnormal rise in box pressure due to either a)loes of integrity arising from accidental pulling out of a neoprene glove from the port ring or rupture of the gloves or development of cracks on the window panels or b) eudden vapourisation of some volatile fluids inside glove box or c) failure of the normal reciroulation system. 3) Negative pressure will be maintained in the emergency exhaust line by continuously running the exhaust fan located in the filter house(one standby exhaust fan is also provided).

4) Safety point: Any rise in box pressure will stop the normal argon inlet flow to the box by closure of inlet butterfly valve in the pressure controller which effectively isolates the box from argon recirculating system and helps in maintaining atmospheric purity in the rest of the boxes. An alarm in the form of flashing light, activated by the differential pressure switch and located on the top of glove box will alert personnel about emergency condition*

12. Glove Box Maintenance! For any maintenance work on glove box it will be wheeled out into the isolated room (having adaptor panels on both sides) and then the panel closure plate on the otherside will be opened by a frogman who then removes the box side panel and works on the box from the frogsuit area. A plastic tent is erected in the frogauit area against the adaptor panel for confinement of activity.

13* Glove chafing procedure! Steps 1. Removal of adhesive tape from glove-port. 2. Removal of O-ring 3. Transfer of glove from secoirl groove to first groove.

4. Putting of new glove on the 2nd groove over the old glove in proper position. 5. Fixing of adhesive tape. 6. Pulling out the old glove into the box with enough care that new glove is well fixed to the groove. 7. Putting of O-ring 8» Final putting of adhesive tape. 79

AFPMDIX -II

DESIGN DETAILS FOR A TYPICAL NORMAL ATM03IHERE CLOVE BOX AND FUMEHOOD

Glove Box:

1. Construction detailSJ (a) Dimensions: Single module box: 884 mm length 872 mm depth 960 .mm height Double module box: 1668 mm length 672 mm depth 960 mm height (b) Materials used: Main frame and the floor» welded 1/8" thick mild steel. The Bide panels are also of 1/8" mild steel, carry a square air lock transfer port (368 mm x 368 mm) on one side and two sets of filter boxes (212 mm x 212 mm) on the other side. Panels, the ceiling, front and backs 10 mm thick laminated safety .glaaa.

2. Glove port Sizes: Lower and Upper: 6" dia. Materials: Alloy of aluminium and 12$ of silicon and moulded round U-section neoprene gaskets*

Glovea used: Size: 0.8 mm thick and 800 mm long with headed cuff. L&ke: 'Milled' neoprene glovea. Fumehooda: (a) Dimensionst Pumehood type FH1J 4' wide x 2'9" deep x 5' height with a front opening of 31 height, which fitted with a laminate safety glass panel. Two holes of 6"dia for regular operating in the fumehood. 80

Fumehood type JTHidi They are of the same dimt-nsiona and design ag that of PH1, except Tor the additional provision of an integral drain board and sink on the hood flooring. The size of the sink ia 17"xH "x7".

(b) Materials used: Main frames 2-^" square pipe mild steel. Floor: jjr" thick mild street. Front panel: 10 mm thick laminated safety ^laaa. (o) Paint used: The entire fumehood and table struoture is painted with epoxy chemical resistant paint.