Development of Polymer Concrete Radioactive Waste Management Containers

KR0000117 KAERI/RR-1945/98

3 1/31 DISCLAIMER

Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. KAERI/RR-1945/98

Development of Polymer Concrete Radioactive Waste Management Containers -1 - - 3 - IV. ^

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- 4 SUMMARY

I . Project Title

Development of Polymer Concrete Radioactive Waste Management Containers n. Objectives and Importance of the Project

Nuclear power production and the use of radioactive materials and inoizing radiation in industry, agriculture and medicine generate radioactive wastes. These wastes must be safely managed and isolated at all stages prior to and including ultimate safe disposal. Therefore, the purpose of this research is to develop a high-integrity container using polymer resin concrete.

IE. Scope and Contents of the Projects

Experimental apparatus were designed and prepared for hydrothermal experiments and sample preparation. Compositions of polymer concrete samples were estimated using a packing model of spherical particles. Five types of ceramic filler were used'- Calcium carbonate, silica, tabular alumina, calcined alumina, and chamotte. Polymer concrete samples were prepared with ceramic filler, polymer resin, and fine aggregate such as standard sand or silicon carbide grain. Effect of ceramic fillers and ageing atmosphere on the mechanical properties of polymer concrete samples has been analyzed with measurement data of bending strength and elastic modulus. Effect of r -radiation and acid or base treatment on the physico-chemical properties of polymer concrete also has been investigated. Polarized microscope and scanning electron microscope were

- 5 - used to correlate microstructure with the mechanical properties of polymer concrete. Furthermore, free drop tests and biodegradation tests were carried out to verity the integrity of the polymer concrete radioactive waste container.

IV. Results

A high-integrity radioactive waste container has been developed to immobilize the spent resin wastes from nuclear power plants, protect possible future, inadvertent intruders from damaging radiation. The polymer concrete container is designed to ensure safe and reliable disposal of the radioactive waste for a minimum period of 300 years. A built-in vent system for each container will permit the release of gas. An experimental evaluation of the mechanical, chemical, and biological tests of the container was carried out. The tests showed that the polymer concrete container is adequate for safe disposal of the radioactive wastes.

V. Proposal for Applications

The polymer concrete radioactive waste containers developed by this research are to be applied to the storage of radioactive spent resin waste from power plants.

- 6 - CONTENTS

1. Introduction 19

2. State of the Arts 21

3. Theoretical and Experimental 25 3-1. Related Laws for the Radioactive Waste Container 25 3-2. Polymer Concrete Technology 26 3-3. Requirement for Radioactive Waste Containers 44 3-4. Manufacture and Test of Polymer Concrete Specimen -45 3-5. Test of Polymer Concrete Standard Specimen for the -87 Evaluation of Environmental Resistance , 3-6. Manufacture and Test of a Polymer Concrete 108 Container 3-7. Conclusion 111

4. Achievements and Industrial Contribution 113

5. Proposal for Application 115

6. References 119

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NEXT PAGE(S) - 11 - left BLANK -2-3. Silane-§- =L ^3-2-4. Polymer modified concreted

PICS] ^|£ ^J. -4-1. Binary packing model of spherical particles. 3-4-2. Particle size distribution of ceramic fillers : (A) CaCO3, (B) calcined alumina (C) SiO2, (D) tabular alumina (E) Chamotte. -4-3. Morphology of ceramic fillers: CaCOs (A) powder, (B) surface of a CaCO3 particle, (C) powder and (D) surface of a SiQj particle. 3-4-4. Autoclave used for hydro-thermal ageing experiment of polymer concrete. -4-5. Bending strength of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates. ^2-^3-4-6. Elastic modulus of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates. ^3.123-4-7. Bending strength versus bulk density of PC samples. ^.^3-4-8. Bending strength versus polymer resin contents of PC samples. •3.^13-4-9. Elastic modulus versus bulk density of PC samples. -4-10. Elastic modulus versus polymer resin contents of PC samples. -4-ll. Bending strength versus elastic modulus of PC samples. 3-4-12. Effect of polymer resin and coupling agent on the bending strength of PC samples. Effect of polymer resin and coupling agent on the elastic modulus of PC samples. Bending strength of PC samples with and without cyclic freezing treatment. ^3,^13-4-15. Elastic modulus of PC samples with and without cyclic freezing treatment.

- 13 - -4-16. Effect of polymer resin content on bending strength of PC samples prepared with silica filler. -4-17. Effect of polymer resin content on bending strength of PC samples prepared with CaCOs filler. -4-18. Elastic modulus versus polymer resin content of PC samples prepared with silica filler. -4-19. Elastic modulus versus polymer resin content of PC samples prepared with CaC03 filler. -4-20. Effect of intensity of r -radiation on bending strength of PC samples prepared with two different polymer resin. ? 3-4-21. Effect of intensity of 7 -radiation on elastic modulus of PC samples prepared with two different polymer resin. -4-22. Bending strength versus acid treatment time for PC samples prepared with two different ceramic fillers. ^•^3-4-23. Elastic modulus versus acid treatment time for PC samples prepared with two different ceramic fillers. ^•^3-4-24. Effect of ageing atmosphere on the bending strength of PC samples prepared with silica filler. -4-25. Effect of ageing atmosphere on the elastic modulus of PC samples prepared with silica filler. 3-4-26. Effect of ageing atmosphere on the bending strength of PC samples prepared with CaCO3 filler. -4-27. Effect of ageing atmosphere on the elastic modulus of PC samples prepared with CaCO3 filler. ^2-113-4-28. Effect of ageing time on the bending strength of PC sample immersed in water (autoclave, at 80 ^C). ^.f) 3-4-29. Effect of ageing time on the elastic modulus of PC sample immersed in water(autoclave, at 80*0. ^.^3-4-30. Micrographs of a PC sample prepared with CaCO3 filler : (A) loose packing of CaCCb particles : (B) dense packing of CaCO3 particles- (C) Magnified image of Fig.30 (A): (D) Magnified image of Fig.30(B).

- 14 - ZL^3-4-31. Micrographs of a PC sample prepared with SiO2 filler. The unsaturated polyester resin 120% of theoretical resin loading (A),(B) and the unsaturated polyester resin contents was 100% of theoretical resin loading (C), (D). ^•^3-4-32. Micrographs showing the polished surface of PC samples '• (A) with coupling agent • (B) without coupling agent 3-4-33. Micrographs showing the fracture surface of PC samples prepared with CaCC>3 filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading. 3-4-34. HNO3-treated fracture surface of PC samples prepared with CaCO3 filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading. -35. NaOH-treated fracture surface of PC samples prepared with CaCC>3 filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading.

a. ^3-5-2. ^r^ A]^6fl vs^s. 7cj-£$£J-3} fitting curve. ^.^3-5-3. -&S.*\] <4-g- tl Zi^3-5-4. n^3-5-5. ^.^3-5-6. Specimens for Testing Resistance to Fungi. Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar(Aspergillus Niger). Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar(GHocladium Virens). -5-9. Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar (Aureobasidium Pullulans). 0. Photograph of Specimen Surface before treated with Aspergillus Niger.lOOO X. -11. Photograph of Specimen Surface before treated with Gliocladium Virens,1000. 3-5-12. Photograph of Specimen Surface before treated with Aureobasidium Pullulans,1000 X.

- 15 - ^133-5-13. Photograph of Specimen after 8 days' Experiment on the Surface of PotatQ-dextrose-agar(AspergiUus Niger). 14. Photograph of Specimen after 8 days' Experiment on the Surface of Potato-dextrose-agar(Gliocladium Virens). -5-15. Photograph of Specimen after 8 days' Experiment on the Surface of Potato-dextrose-agar(Aureobasidium Pullulans). 3-5-16. Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agar(Aspergillus Niger). -5-17. Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agarCGliocladium Virens). 3-5-18. Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agaKAureobasidium pullulans). 3-5-19. Photograph of Specimen Surface after 20 days' Experiment with Aspergillus Niger.lOOO X ^•^3-5-20. Photograph of Specimen Surface after 20 days' Experiment with Gliocladium Virens,1000 X. ^3.^3-5-21. Photograph of Specimen Surface after 20 days' Experiment with Aureobasidium Pullulans, 1000 X. ZL ^3-5-22. Specimens after Testing Resistance to Fungi.

- 16 - S.3-2-1. 23-2-2. 23-2-3. ^SM3 ^M^ JM$7MI3 Sf * 5.3-4-1. Properties of synthetic resins used for polymer concretes. 3.3-4-2. Particle size distribution of standard sand used as fine aggregated for polymer concrete samples. 3.3-4-3. Properties of two different fine aggregates. 23-4-4. Properties of CaC03 filler. 33-4-5. Properties of silica filler. 23-4-6. Properties of fine aggregates and ceramic fillers. 23-4-7. Estimated mixing ratios of fine aggregates and ceramic fillers for maximum packing density of powder mixtures. 23-4-8. Estimated compositions for polymer concrete samples with maximum packing density 23-4-9. Properties of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates. 23-4-10. Effect of polymer resin and coupling agent on the bending strengthand elastic modulus of PC samples. 23-4-11. Bending strength and elastic modulus of PC samples with and without the cyclic freezing treatment. 23-4-12. Bending strength and elastic modulus of PC samples prepared with different polymer resin content. 23-4-13. Effect of r -radiation on the bending strength and elastic modulus of PC samples prepared with two different polymer resin. 23-4-14 Effect of acid treatment on the bending strength and elastic modulus of PC samples prepared with two different ceramic fillers. 23-4-15. Effect of high temperature ageing atmosphere on the bending strength and elastic modulus of PC samples prepared with two different ceramic fillers. 23-5-1. 1*1 eH n}-& ^g-X\^ 23-5-2. -&^*\] 4-g- tl ^ t2

- 17 - - 18 - NEXT PAGE(S) left BLANK - 21 - - 22 - - 23 - - 24 - - 25 - - 26 - - 27 - - 28 - - 29 - - 30 - - 31 - - 32 - - 33 - $4. calcium carbonate^-

i. Polymer-modified (or cement) mortar (PCM) and concrete(PCC) -. Polymer mortar (PM) and concrete (PC) . Polymer-impregnated mortar (PIM) and concrete (PIC) PMM, PCC, PM ^ PCfe ^5] A]~g-3;2. $JA4 PIM^]4 PIC

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- 34 - (1)

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- 35 - I Uniform film

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- 39 - «i

(4) |fe if- silane coupling^ coupling^S.

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- 40 - (6)

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- 41 - 4. Polymer-Impregnated mortar and concrete (PIMJPIC)

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- 45 - 2.60 g/crf, 1.54 g/af, 1.69°1&S-*I. 2.46 g/crf, 0.86 g/cnf, 2.86^:

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COMPOSITION r - 0 O(+y= I)

URGE SPHERES SMALL SPHERES

Fig. 1. Binary packing model of spherical particles.

C, M, x, yfe AA -51 ell3 7}

M 2.86 -0.81 C - 1 + M 1.69 -1 + 2.86 y = I - x= 1 - 0.81 = 0.19

fe 81:197}- °1 1.1 ?1 71:16:135.

- 46 - 4-

(1)

7V Table 1. Properties of synthetic resins used for polymer concretes.

E£oxy Polyurethane Polyester PMMA resistance to acids G G G G/M resistance to alkali G M P P resistance to solvents G G G P variable degree of M G M M/G flexibility UV light resistance P G G G

adhesion G G/M G/M G/M

Possible hardening or polymerization at low P P P G temperatures down to zero : Good , M: Medium , P: Poor

, AA» 55% MEKPO(45% 1.12±0.02«]4. PC 4

. Table

(2) PC5]

- 47 - PC ±= KS L51(XH Table 2«H je.?SH

. PC Table

Table 2. Particle size distribution of standard sand used as fine aggregated for polymer concrete samples.

^^ -g-3 840/an 590^1 297jrai

a^A> - 1.0 61«|- 95.0 <>1# 0.4 »l€r 1.53-1.60

Table 3. Properties of two different fine aggregates.

(g/cc) (g/cc)

2.6 1.54 300

3.17 1.74 300-425

(3)

calcined 1HHM-, tabular *3-€- PC %$.4$= PC

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- 48 - DflTH N0.« SftMPLEl 1 1/-O7-J bfeTR Ha. 3 SRtfPLE* CR • •• • 71 • •'::::: o» •jr lo, ^» <*> ... <>:> v. PARTICLE SIZE J(<^^>> <«

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PflTfl Mo. 1 S«f1PLE> S1D2 DflTfl No.2 SRMPLE:TO ' • ' '•"": •: • V —: J : : : : <*> : • ' = = .... ft : : r- • le v / T "t : y -~.™C^rt ... wJlx|.xi....ri t? ( « 1 • r r t^ try::::::'A* :• pftRTICLE SIZE PRRTJCLE SIZE X<>n>

(E) Fig. 2. Particle size distribution of ceramic fillers : (A) CaCCfe, (B) calcined alumina (C) SiO2, (D) tabular alumina (E) Chamotte.

- 49 - <>1 Table Table . Fig. 3

SL,

Table 4. Properties of CaCOs filler.

- 93.4 KSL5113-87 CaO % 53.6 MASTER PARTICLE Fe2O3 % 0.13 SIZER M3 325mesh ^^ % 0.1 KSA5302-91 - 2.46 KSE3071-87 - 0.74 KSM0004-88 % 0.24 KSM0011-87 10.10 KSM0010-92 Ph(26.4C10%SOL) /an 7.4 LCD ^7} #7V^^r % 0.10 KSM6555-903 149 ^^r^r % 0 44 a||*r€- % 0.20

Table 5. Properties of silica filler.

AVERAGE 13.0±2.0 12.57 KMC-TM-001 PARTICLE SIZE SIEVE ANALYSIS 80 MESH ON 0 0 KMC-TM-002 120 MESH ON MAX. 0.7 I 0.16 VOLATILE MAX. 0.15 0.11 KMC-TM-003 CONTENT % PH 5.3-7.6 - 7.2 KMC-TM-005 SPECIFIC GRAVITY 2.65±0.10 - 2064 KMC-TM-006

- 50 - •Table 6. Properties of Fine aggregates and ceramic fillers.

(g/cc) (g/ce) 2.60 1.54 1.69 297

3.17 1.74 1.82 300-425

246 0.86 2.86 6B |pfp Isli- mi BD mi pifl •11 ill loill!; 1111 Hi;:;: ill!! ill Wm HI IWiil Hi Hi MM nil

Fig. 3. Morphok . I1. i -1 .- • CaCCb particle, (C) powder and (D) surface of a SiOz particle PC 120TC sjfe £S7]61H 24

bar 60*0 ^fe

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- 53 - (7)

. PC 44 autoclaveCFig. 4)» 4-§-«r^ ##4

Fig. 4. Autoclave used for hydro-thermal ageing experiment of polymer concrete.

(8) PC

- 54 - 2.

Table 7^r S zh PC

Table Table 8£r Table 7<=fl ^Ef\i PC

44 62~71wt%, 16~29wt%, 9~13wt% . Table 8^1 M-EKJ PC A]^ 3^3 , ^lfe, #^^1S^ tabular alumina* 29wt% A>-g-s|-7iq- 25wt% A)~g-is}-o} ^14^- pc A] #, calcined alumina, 4SI) ^^:## ^^^HS. 16wt% A>-§-sj-^u|- calcined alumina» 21wt% A)~g- , PC A]^ *fl4ofl Af-g-slfe ^^pl^^] f-^^. i3wt%S^ 7}% fe7fl 1+

Table 8£r S.^^3L ^4. 3£^f PC

tabular alumina4 ^?\7\ ^^-^r ^-i: Table

- 55 - Table 7. Estimated mixing ratios of fine aggregates and ceramic fillers for maximum packing density of powder mixtures.

4m*H ^^^^«l^(wt%) 4]M (vol%) 81 19 27 Calcined IH^IM- 76 24 28 Tabular IH^IM- 68 32 21 76 24 23 73 27 19 82 18 29 Calcined ^HHM- 77 23 31 Tabular OT'14 69 31 23 77 23 25 73 27 21

Table 8. Estimated compositions for polymer concrete samples with maximum packing density.

(^*13 0.5wt%) 71% 16% 13% 0.65 Calcined IH^IM- 66% 21% 13% u Tabular ^-^14 62% 29% 9% u 68% 21% 11% u 66% 25% 9% « 71% 16% 13% u Calcined IH-13)^ 67% 20% 13% u

tt Tabular ^^-TII^ 62% 28% 9% 68% 21% 11% u 67% 24% 9% u

- 56 - Table 8<% M-E{-\+ St^SLS, PC Table 9^4 Fig. 5^ 6^1 AA

PC ^1^^ ^-3q^JE7V 13-15% -x^ 3.711 .17 g/ast)7\ JE.efi(2.60

fe o^ 188

A}- PC PC

Table , calcined alumina, -te]^*- AA ^^ PC 60"C ££7HA-] 24A)?> -B-^)^; ^i 30^ ^o> AJ-^ ^^-A]^1 365±2 kg/crf, 331±10 kg/crf, 272±11 calcined aluminaS. Af-g-^ ^-f, Table 8

ig. 7, 8).

- 57 - Table 9. Properties of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates. «-«^-£ Cg/crf) (kg/crf) (GPa) 2.05±0.01 365±2 24.2+0.6 Calcined W*>14 2.17±0.02 331±10 26.0+0.4 Tabular W*14 2.19±0.03 210±2 39.5±2.3 A):£E 1.88±0.04 103±l 17.4±0.4 2.05+0.02 272+11 28.3±L2 2.32±0.04 321±9 31.6+1^ Calcined IHHM- 2.46±0.01 303+22 38.2±33 3rSF?f4; Tabular IH^M 2.57+0.02 372±4 57.3+1.6 A]:SE 2.34±0.04 342±12 47.8 ±0.4 2.38+0.03 324+7 48.9±0.5

S1O2 CaCO3 C.A T.A filler Fig. 5. Bending strength of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates.

- 58 - SiO2 CaCO3 T.A filler Fig. 6. Elastic modulus of polymer concrete samples prepared with two different ceramic fillers and various fine aggregates.

tabular alumina 372±4 kg/crf-S. -^^3-Efe 342±12 kg/cnf, 324±7 kg/crf 9%, 11%,

(Fig. 8) ^ PC

3.71 PC

44 44 PCs] 44011 4UU

O •

f 300 • • 2? O m o n-i 200 oSgW

i nn 1 1*, 1 1 1.6 1.8 2 2.2 2.4 2.6 2.8

Fig. 7. Bending strength versus bulk density of PC samples.

4UU

• o

* o

f 300 • o m SO o r}i 200 oHSAf

1 r\n a a 8 9 10 11 12 13 14 ^Xltfi^ (wt%)

Fig. 8. Bending strength versus polymer resin contents of PC samples.

- 60 - 60

50 . m

H 40

o 30 o uSJ o o 20 o

• 4 a 10 1.6 1.8 2 2.2 2.4 2.6 2.8

Fig. 9. Elastic modulus versus bulk density of PC samples.

40 ± • FT 30 O O o o 20 o

1 n 8 9 10 11 12 13 14 .A. xi fQ at ( «/t9f,>

Fig. 10. Elastic modulus versus polymer resin contents of PC samples.

PC PC

3.7] Fig. 9^ PC^

- 61 - PC

., pc , o] PCS] 11)

Table 104 Fig. 12^ 13*^1 -L ^ 66%

PCS]

PCS] PCS] fe PC

- 62 - 400 o m • • ? 300 • o IH SO o 200 oagw

100 10 20 30 40 50 60

Fig. 11. Bending strength versus elastic modulus of PC samples.

Table 10. Effect of polymer resin and coupling agent on the bending strengthand elastic modulus of PC samples.

7c}-£(kg/cm2) ^Tfl-rCGPa)

X 243±1 22.6±0.4 A O 403 ±22 21.9±0.8 X 226+26 22.6+1.3 B O 333±11 20.1+0.8 X 367±7 22.2±0.1 C O 398±36 21.3±0.7

A : B : C :

- 63 - 450 DCoupllng agent X 400 • Coupling agent O 350 sg-300 5250 • 1 1 §150 100 50 - I 1 1 0 LJL * Li , 1LI 1

Fig. 12. Effect of polymer resin and coupling agent on the bending strength of PC samples.

24 DCoupling agent X • Coupling agent O 23

Fig. 13. Effect of polymer resin and coupling agent on the elastic modulus of PC samples.

- 64 - 4.

Table 112}- Fig. 14, 4

Table 11. Bending strength and elastic modulus of PC samples with and without the cyclic freezing treatment

^^(kg/cm2) ^^^l^rCGPa) A]f^ 272±11 28.3±1.2 *)W 296±10 27.5+1.6 324±7 48.9±0.5 356+15 49.0 ±1.4

Table 12. Bending strength and elastic modulus of PC samples prepared with different polymer resin content.

(^l€:7l§) (g/cnf) (kg/cm2) (GPa) 100 2.07 303±13 28.9±2.5 120 2.12 544±10 35.2±0.7

^# 100 2.10 480+15 26.8±0.7 ^:# 120 2.12 591 + 12 28.1 ±0.5

- 65 - 400

Fig. 14. Bending strength of PC samples with and without cycEc freezing treatment

SSAKS1O2 SiC+SiO2 AII g w + m a JH Fig. 15. Elastic modulus of PC samples with and without cyclic freezing treatment.

- 66 - . PC

Table 124 Fig. 16-19*11 PC

1.2tiH PC A fe 303 544 kg/cnf, 3.71] fe 480 591 kg/cafS.

Fig. 16. Effect of polymer resin content on bending strength of PC samples prepared with silica filler.

- 67 - 700

600

500

a 400 ft

fa 3oo so ® 200 •

100 •

0 x 1 1.2

Fig. 17. Effect of polymer resin content on bending strength of PC samples prepared with CaCC>3 filler.

£25 •

K ffiJ 10 -

5 •

0 x1.2

Fig. 18. Elastic modulus versus polymer resin content of PC samples prepared with silica filler.

- 68 - za

28 •

• -s?27 a. O -fl-26 *

SSJ 25 •

x 1 x 1.2

Fig. 19. Elastic modulus versus polymer resin content of PC samples prepared with CaCC-3 filler.

13wt%

PC 480 kg/cnfS. ^ 303 ^ 60% ^^ Fig. 164 PC 20% 303 544 kg/aiS, 3.X] A]-g- 480 10%

9wt% x 1.2 = 10.8wt%5. 13wt% .711

Fig. 184 19^:

29 35 GfeS. 20%

- 69 - PC

100 kg/cnf2 51 y v 50 MRad o 4^2:A> ^°\] %SJ\ 30 kg/crf ^7>^t|-7]- SA>^o] 100 ^-dbs) ^^(Table 13). ;$.$-, W^-A}^ 2:A>7V 100 MRad 4^1 fe ^1^1 ^H crosslinking fLtfiLS. ^l«flA^ ^^j^sj- 7l7i]^7j-£ s^ ^7}^, 150M Rad .S 44^4(Fig. 20, 21). crosslinking 7>5L bond^l degradation^ 3*IW ^ol-^l3L

14, Fig. 22, 23)

- 70 - 510 kg/cnf°M 460 31 M*] 27 'QhS. ^ 13% #4

PC PC 60X: PC , 60^?}:

Table 13. Effect of r -radiation on the bending strength and elastic modulus of PC samples prepared with two different polymer resin.

(M Rad) (kg/cm2) (GPa) 0 303+13 29.0 ±2.5 ^*\ 50 398±33 32.3±0.8 100 406±15 34.1 ±1.4 1 150 403±26 32.2+0.2 0 486±17 31.0±0.3 50 516 ±16 31.3+0.9 100 498±29 30.2±2.8 150 471 ±11 29.9±1.8

- 71 - sou

500 o- g450 •

400 / 60 ta 350

300 5£IOfl ^(K AJ) 250 0 50 100 150 Ah (M Rad)

Fig. 20. Effect of intensity of r -radiation on bending strength of PC samples prepared with two different polymer resin.

34 / \ / • 33 / \

32 > (GPa ) / 31 / 30 3 ~o

29 • eeioi

28 0 50 100 150 SAl-d S AV (M Rad)

Fig. 21. Effect of intensity of r -radiation on elastic modulus of PC samples prepared with two different polymer resin.

- 72 - Table 14 Effect of acid treatment on the bending strength and elastic modulus of PC samples prepared with two different ceramic fillers.

(day) (kg/cm2) (GPa) 0 303±13 29.0±2.5 7 415+16 31.8±0.6 €371- 14 408+16 31.2±2.3 21 406±17 30.3+0.2 0 480±15 26.8+0.7 7 510 ±54 31.4±1.4 14 496+40 29.7 ±0.7 21 464±38 26.5±2.5

550

500

J 450

2 400 m 350 mi 300

250

Fig. 22. Bending strength versus acid treatment time for PC samples prepared with two different ceramic fillers.

- 73 - 33 32 31 30 29 28 27 26 25

Fig. 23. Elastic modulus versus acid treatment time for PC samples prepared with two different ceramic fillers.

AA PC 60"C -, 60

30^ ^ 60

80°C A}

- 74 - Table 15. Effect of high temperature ageing atmosphere on the bending strength and elastic modulus of PC samples prepared with two different ceramic fillers.

(day) (kg/cm2) (GPa) - 303±13 29.0±2.5 30 410±29 31.7±1.2 -gem 80iC^7l (AAV 1-3ESJ- 60 441 ±15 31.6+15 €-eH^3) 30 260±ll 13.8±1.3 80TC1: 60 145±27 8.9±0.5 - 479.9 26.8+0.7 30 444±3 27.9±0.8 80^7} 60 494±48 29.7±0.4 30 384±33 20.1+8.8 ^#«:# 801C1- 60 374±10 18.9 ±1.8 7 182±11 13.3+0.5 80 TC 14 183 ± 2 12.8±0.4 autoclave 21 183 ± 1 12.5+0.2 28 177 ± 7 11.6±0.7

Fig. 24. Effect of ageing atmosphere on the bending strength of PC samples prepared with silica filler.

- 75 - 80"C 80TC sot: (30 i*) (60S) (30§J) (60^) Fig. 25. Effect of ageing atmosphere on the elastic modulus of PC samples prepared with silica filler.

600

80X; 801C 80 "C (30^) (60^) (30iJ) (60 i!)

Fig. 26. Effect of ageing atmosphere on the bending strength of PC samples prepared with CaCC>3 filler.

- 76 - 25 -

^20 - 0- -(1-1o 5 -

5 -

n 80*0 S7IS sot: 80t; ^ sot: (60g!)

Fig. 27. Effect of ageing atmosphere on the elastic modulus of PC samples prepared with CaCO3 filler.

550

450

cs w 350 rjt) m 250

150 14 21 28

Fig. 28. Effect of ageing time on the bending strength of PC sample immersed in water(autoclave, at 80V,).

-17 - 14 21

Fig. 29. Effect of ageing time on the elastic modulus of PC sample immersed in waterCautoclave, at 80V).

0 0 II n R-C-O-R + H,0 R-C-OH + HO-R

S.^7]-

(1)

PC Fig. 304 3H PC Fig. 30^ (A), (B), (C), (DH PC

- 78 - Fig.

30-50 fe PC

Fig.

(A) Fig. 30. Micrographs of a PC sample prepared with CaCCfe filler : (A) loose . packing of CaCC>3 particles : (B) dense packing of CsCOs particles: (C) Magnified image of Fig. 30 (A): (D) Magnified image of Fig. 30 (B).

- 79 - (B)

60*. (C)

(D) Fig. 30. (continued)

- 80 - (A)

(B) Fig. 31. Micrographs of a PC sample prepared with SiO2 filler. The unsaturated polyester resin 120% of theoretical resin loading (A),(B) and the unsaturated polyester resin contents was 100% of theoretical resin loading (C), (D).

- 81 - (C)

(D) Fig. 31. (continued) Micrographs of a PC sample prepared with SiC>2 filler, the unsaturated polyester resin contents was 100% of theoretical resin loading (C), (D).

Fig. PC

PC (2) PC 12.4£ . 33, 34,

PC ^^^ PC ^-f, SL-S-g; . 34 (A), (B)).

(A) (B) Fig. 32. Micrographs showing the polished surface of PC samples • (A) with coupling agent : (B) without coupling agent

- 83 - (A) . (B) Fig. 33. Micrographs showing the fracture surface of PC samples prepared with CaCCfe filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading.

(A) (B) Fig. 34. HNO3-treated fracture surface of PC samples prepared with CaCC>3 filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading.

(A) (B) Fig. 35. NaOH-treated fracture surface of PC samples prepared with CaCO3 filler. The unsaturated polyester resin contents were (A) 100% and (B) 120% of theoretical resin loading.

- 84 - PC . 35).

PC -S.s. ^-S-^- $14- PCS)

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= 160exp( ) + 82exp(-

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- 89 - 2000-

80000 1600-

60000-

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350 355 360 365 370 375 350 355 360 365 370 375 380 Temperature (°K) Temperature (°K)

300 year S ifiE ofl^j- *| =46.4kg/cn?

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- 92 - 4.

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ASTM ^71- potato dextrose agar

7\. (1) £3.

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~~53.38%, 24.91%S. €32)1 15m 9J~~

~~. Specimens for Testing Resistance to Fungi.~~

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~~(4) ^^1 (potato dextrose agar) 42g# 1000 ml~~

Petri dishes «4# ^ s\A] ^4. Petri dish aspergillus niger, gliocladium virens, aureobasidium pullulanslt 43 fe^-. (4)Water shaker^) #-§• ^8-f 2. *£t 30 'CS. ^^ 43^: Petri dish* water shaker °1]S. dl§-^J-°^ Bfl<

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~~Spore suspension^: pore suspension tfl^H agar 30 1C, ^vtfl^£ 100 . Spore suspension^: , Nutrient-salts-agar~~

~~Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar(Aspergillus Niger).~~

~~- 98 - Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar(Gliocladium Virens).~~

~~. Photograph of Specimen after 20 days' Experiment on the Surface of Nutrient-salts-agar (Aureobasidium Pullulans).~~

~~(2) Potato dextrose agaHH Spore suspension -S] potato dextrose agar^Mi~~

~~potato dextrose agar~~

~~- 99 - 10, 11,~~

~~Photograph of Specimen Surface before treated with Aspergillus Niger.1000 X.~~

~~Photograph of Specimen Surface before treated with Gliocladium Virens,1000 X.~~

~~Photograph of Specimen Surface before treated with Aureobasidium Pullulans.lOOO X.~~

- 100 - ft 13, 14, 15-8: potato dextrose agar aspergillus niger, gliocladium virens, aureobasidium pvdlulans^ , aureobasidium pullulans 3. ^^^^1 16, 17, 18£r ^1^14 potato dextrose agar ^^1 ^r^l-i- ^.<^§4. Aspergillus niger, gliocladium virensSrte 4S. aureobasidium pullulansS

~~19, 20, 21~~

~~Photograph of Specimen after 8 days' Experiment on the Surface of Potato-dextrose-agar(Aspergillus Niger).~~

~~- 101 - Photograph of Specimen after 8 days' Experiment on the Surface of Potato-dextrose-agar(Gliocladium Virens).~~

~~Photograph of Specimen after 8 days' Experiment on the Surface of Potato-dextrose-agar(Aureobasidium Pullulans).~~

~~- 102 - Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agar(AspergiIlus Niger).~~

~~Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agar(Gliocladium Virens).~~

~~- 103 - Photograph of Specimen after 20 days' Experiment on the Surface of Potato-dextrose-agar(Aureobasidiutn pullulans).~~

~~Photograph of Specimen Surface after 20 days' Experiment with Aspergillus Niger,1000 X.~~

~~- 104 - Photograph of Specimen Surface after 20 days' Experiment with Gliocladium Virens,1000 X.~~

~~Photograph of Specimen Surface after 20 days' Experiment with Aureobasidium Pullulans.lOOO X.~~

~~- 105 - :3 *HH l, -Ml~~

~~4. #SL 1, 2, A, B, C~~

~~Specimens after Testing Resistance to Fungi.~~

~~- 106 - ASTM ^ G 21~~

~~gel nutrient-salts-agar 7} ^-^]7\ $4^7)<%~~

~~SLA>«|-71 ^«B potato dextrose agar , aureobasidium pullulans ^ potato dextrose agar *H1~~

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~~3-2-1. "Immobilization of Low and Intermediate Level Radioactive Wastes with Polymers" International Atomic Energy Agency, Vienna, 1988, Technical reports series No. 289, 45.~~

~~3-2-2. http://isl.cps.msu.edu/ "TRP-Liquid Molding-Cure-CrossUnMng of Unsaturated Polyester Resins~~

~~3-2-3. "Plastics Materials", J.A. Brydson, Butterworth Scientific, Fifth Edition, 1989.~~

~~3-2-4. "Polymers in Concrete" Satich Chandra and Yoshihiko Ohama, Inc. CRC Press, 1994.~~

~~3-4-1. Kloker W. and Dolfen E., "Rigid Foam Light Weight Concrete Based on Unsaturated Polyester Resins." Proceedings of the First ICPIC, London, England, 378-396 (1975)~~

~~3-4-2. Imamura K., Toyokawa K., and Murdi N., "Precast Polymer Concrete Manhole Development" Proceedings of the Second ICPIC, Austin, Texas, 173-186 (1978)~~

~~3-4-3. Ohama. Y. Kawakami. M., and Fukuzawa K., Polymers in Cncrete, Proceedings of the Second East Asia Symposium on Polymers in Concrete, E & FNSPON, London (1997)~~

~~3-4-4. «W, "ig*Ma si]7j# *)e]-g- 3L#$.s$ -8-7] (1995)~~

~~3-4-5.~~

~~- 117 - . 10C1], 93-101 (1993)~~

~~3-4-6. Stevens. M. P, Polymer Chemistry. 2nd Ed., Oxford University Press, Oxford, 305-306 (1990)~~

~~3-4-7. Mikhailov. K. V. et aL, Polymer Concretes and Their Structural Uses, 288-290 (1992)~~

~~3-4-8. McCrum. N. G., Buckley. C. P., and BucknalL C. B., Principles of Polymer Engineering, Oxford Science, NewYork, 1988, P. 210~~

~~3-4-9. £TM, S.^, "#el^$3.SlEi. ol-g.^- ^ jfl^j 7^ 1991 (1993)~~

~~3-4-10. (1994)~~

3-4-11. Soh. Y. S., Jo. Y. K, and Park. H. S., "Effect of Filler on the Mechanical Properties of Unsaturated Polyester Resin Mortar.", Proceedings of the Second East Asia Symposium on Polymers in Concrete, Koriyama, Japan, 67-74 (1997)

~~3-4-12. Milewski. J. V., "Efficient Use of Whiskers in the Reinforcement of Ceramics," Ad. Ceram. Mater., 1[1] 36-41 (1986)~~

~~3-4-13. Starr. T. C, "Packing Density of Fiber / Powder Blends." Am. Ceram. Soc. Bull., 65[9] 1293-1296 (1986)~~

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(1) ^^:^?1 ^-^A}^: 20.2001 (General requirements) (a) A licensee shall dispose of licensed material only — (1) By transfer to an authorized recipient as provided in 20.2006 or in the regulations in parts 30, 40, 60, 61, 70, or 72 of this chapter; or (2) By decay in storage; or (3) By release in effluents within the limits in 20.1301; or (4) As authorized under 20.2002, 20.2003, 20.2004, or 20.2005. (b) A person must be specifically licensed to receive waste containing licensed material from other persons for: (1) Treatment prior to disposal; or (2) Treatment or disposal by incineration; or (3) Decay in storage; or (4) Disposal at a land disposal facility licensed under part 61 of this chapter; or (5) Disposal at a geologic repository under part 60 of this chapter.

(2) ^7l# *13# 3$: M ^: 20.2002 (Method for obtaining approval of proposed disposal procedures) A licensee or applicant for a license may apply to the Commission for approval of proposed procedures, not otherwise authorized in the regulations in this chapter, to dispose of licensed material generated in the licensee's activities. Each application shall include: (a) A description of the waste containing licensed material to be disposed of, including the physical and chemical properties important to risk evaluation, and the proposed manner and conditions of waste disposal; and (b) An analysis and evaluation of pertinent information on the nature of the

- 148 - environment; and (c) The nature and location of other potentially affected licensed and unlicensed facilities; and (d) Analyses and procedures to ensure that doses are maintained ALARA and within the dose limits in this part

(3) A?\Mr sr^T1 TOM: 205003 (Disposal by release into sanitary sewerage.) (a) A licensee may discharge licensed material into sanitary sewerage if each of the following conditions is satisfied- (1) The material is readily soluble (or is readily dispersible biological material) in water; and (2) The quantity of licensed or other radioactive material that the licensee releases into the sewer in 1 month divided by the average monthly volume of water released into the sewer by the licensee does not exceed the concentration listed in table 3 of appendix B to part 20; and (3) If more than one radionuclide is released, the following conditions must also be satisfied: (0 The licensee shall determine the fraction of the limit in table 3 of appendix B to part 20 represented by discharges into sanitary sewerage by dividing the actual monthly average concentration of each radionuclide released by the licensee into the sewer by the concentration of that radionuclide listed in table 3 of appendix B to part 20; and (ii) The sum of the fractions for each radionuclide required by paragraph (a)(3)(i) of this section does not exceed unity; and (4) The total quantity of licensed and other radioactive material that the licensee releases into the sanitary sewerage system in a year does not exceed 5 curies (185 GBq) of hydrogen-3, 1 curie (37 GBq) of carbon-14, and 1 curie (37 GBq) of all other radioactive materials combined. (b) Excreta from individuals undergoing medical diagnosis or therapy with radioactive material are not subject to the limitations contained in paragraph (a) of this section.

~~(4) sfj7j# 4^*13 &#: 20.2004~~

- 149 - (Treatment or disposal by incineration) (a) A licensee may treat or dispose of licensed material by incineration only: (1) As authorized by paragraph (b) of this section; or (2) If the material is in a form and concentration specified in 20.2005; or (3) As specifically approved by the Commission pursuant to 20.2002. (b) (1) Waste oils (petroleum derived or synthetic oils used principally as lubricants, coolants, hydraulic or insulating fluids, or metalworking oils) that have been radioactively contaminated in the course of the operation or maintenance of a nuclear power reactor licensed under part 50 of this chapter may be incinerated on the site where generated provided that the total radioactive effluents from the facility, including the effluents from such incineration, conform to the requirements of appendix I to part 50 of this chapter and the effluent release limits contained in applicable license conditions other than effluent limits specifically related to incineration of waste oiL The licensee shall report any changes or additions to the information supplied under 50.34 and 5034a of this chapter associated with this incineration pursuant to 50.71 of this chapter, as appropriate. The licensee shall also follow the procedures of 50.59 of this chapter with respect to such changes to the facility or procedures. (2) Solid residues produced in the process of incinerating waste oils must be disposed of as provided by 20.2001. (3) The provisions of this section authorize onsite waste oil incineration under the terms of this section and supersede any provision in an individual plant license or technical specification that may be inconsistent.

(5) #3 s|7l# *!e) *» 20.2005 (Disposal of specific wastes) (a) A licensee may dispose of the following licensed material as if it were not radioactive-' (1) 0.05 microcurie (1.85 KBq), or less, of hydrogen-3 or carbon-14 per gram of medium used for liquid scintillation counting; and (2) 0.05 microcurie (1.85 kBq), or less, of hydrogen-3 or carbon-14 per gram of animal tissue, averaged over the weight of the entire animal. (b) A licensee may not dispose of tissue under paragraph (a)(2) of this section in a manner that would permit its use either as food for humans

~~- 150 - or as animal feed, (c) The licensee shall maintain records in accordance with 702108.~~

(6) *|7l€- ^e|# 3\& *r#: 202006 (Transfer for disposal and manifests) (a) The requirements of this section and appendix G to 10 CFR Part 20 are designed to— (1) Control transfers of low-level radioactive waste by any waste generator, waste collector, or waste processor licensee, as defined in this part, who ships low-level waste either directly, or indirectly through a waste collector or waste processor, to a licensed low-level waste land disposal facility (as defined in Part 61 of this chapter); (2) Establish a manifest tracking system; and (3) Supplement existing requirements concerning transfers and record-keeping for those wastes. (b) Any licensee shipping radioactive waste intended for ultimate disposal at a licensed land disposal facility must document the information required on NRC's Uniform Low-Level Radioactive Waste Manifest and transfer this recorded manifest information to the intended consignee in accordance with appendix G to 10 CFR Part 20. (c) Each shipment manifest must include a certification by the waste generator as specified in section H of appendix G to 10 CFR Part 20. (d) Each person involved in the transfer for disposal and disposal of waste, including the waste generator, waste collector, waste processor, and disposal facility operator, shall comply with the requirements specified in section m of appendix G to 10 CFR Part 20.

~~(7) Appendix G to Part 20 (Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests)~~

(7r). &W (Manifest) A waste generator, collector, or processor who transports, or offers for transportation, low-level radioactive waste intended for ultimate disposal at a licensed low-level radioactive waste land disposal facility must prepare a Manifest (OMB Control Numbers 3150 - 0164, - 0165, and - 0166) reflecting

- 151 - information requested on applicable NRC Forms 540 (Uniform Low-Level Radioactive Waste Manifest (Shipping Paper)) and 541 (Uniform Low-Level Radioactive Waste Manifest (Container and Waste Description)) and, if necessary, on an applicable NRC Form 542 (Uniform Low-Level Radioactive Waste Manifest (Manifest Index and Regional Compact Tabulation)). NRC Forms 540 and 540A must be completed and must physically accompany the pertinent low-level waste shipment Upon agreement between shipper and consignee, NRC Forms 541 and 541A and 542 and 542A may be completed, transmitted, and stored in electronic media with the capability for producing legible, accurate, and complete records on the respective forms. Licensees are not required by NRC to comply with the manifesting requirements of this part when they ship: (a) LLW for processing and expect its return (i.e., for storage under their license) prior to disposal at a licensed land disposal facility; (b) LLW that is being returned to the licensee who is the "waste generator" or "generator," as defined in this part; or (c) Radioactively contaminated material to a "waste processor" that becomes the processor's "residual waste." For guidance in completing these forms, refer to the instructions that accompany the forms. Copies of manifests required by this appendix may be legible carbon copies, photocopies, or computer printouts that reproduce the data in the format of the uniform manifest

A. ^#3 Aj-IKGeneral Information) The shipper of the radioactive waste, shall provide the following information on the uniform manifest 1. The name, facility address, and telephone number of the licensee shipping the waste; 2. An explicit declaration indicating whether the shipper is acting as a waste generator, collector, processor, or a combination of these identifiers for purposes of the manifested shipment; and 3. The name, address, and telephone number, or the name and EPA identification number for the carrier transporting the waste.

~~B. -83 313L (Shipment Information) The shipper of the radioactive waste shall provide the foEowing~~

- 152 - information regarding the waste shipment on the uniform manifest: 1. The date of the waste shipment; 2. The total number of packages/disposal containers; 3. The total disposal volume and disposal weight in the shipment; 4. The total radionuclide activity in the shipment; 5. The activity of each of the radionuclides H - 3, C - 14, Tc-99, and I - 129 contained in the shipment; and 6. The total masses of U - 233, U - 235, and plutonium in special nuclear material, and the total mass of uranium and thorium in source material.

C. sS7]#-g-7l ^ s57]#3};£ (Disposal Container and Waste Information) The shipper of the radioactive waste shall provide the following information on the uniform manifest regarding the waste and each disposal container of waste in the shipment: 1. An alphabetic or numeric identification that uniquely identifies each disposal container in the shipment; 2. A physical description of the disposal container, including the manufacturer and model of any high integrity container; 3. The volume displaced by the disposal container; 4. The gross weight of the disposal container, including the waste; 5. For waste consigned to a disposal facility, the maximum radiation level at the surface of each disposal container; 6. A physical and chemical description of the waste; 7. The total weight percentage of chelating agent for any waste containing more than 0.1% chelating agent by weight, plus the identity of the principal chelating agent; 8. The approximate volume of waste within a container; 9. The sorbing or solidification media, if any, and the identity of the solidification media vendor and brand name; 10. The identities and activities of individual radionuclides contained in each container, the masses of U - 233, U - 235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material. For discrete waste types (Le., activated materials, contaminated equipment, mechanical filters, sealed source/devices, and wastes in solidification/stabilization media), the identities and activities of individual radionuclides associated with or contained on these waste types within a disposal container shall be reported;

- 153 - 11. The total radioactivity within each container; and 12. For wastes consigned to a disposal facility, the classification of (he waste pursuant to ?1.55 of this chapter. Waste not meeting the structural stability requirements of ?1.56(b) of this chapter must be identified.

D. «]£.# A y\ § ^J£(Uncontainerized Waste Information) The shipper of the radioactive waste shall provide the following information on the uniform manifest regarding a waste shipment delivered without a disposal container: 1. The approximate volume and weight of the waste; 2. A physical and chemical description of the waste; 3. The total weight percentage of chelating agent if the chelating agent exceeds 0.1% by weight, plus the identity of the principal chelating agent; 4. For waste consigned to a disposal facility, the classification of the waste pursuant to ?1.55 of this chapter. Waste not meeting the structural stability requirements of ?1.56(b) of this chapter must be identified; 5. The identities and activities of individual radionuclides contained in the waste, the masses of U - 233, U - 235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material; and 6. For wastes consigned to a disposal facility, the maximum radiation levels at the surface of the waste.

E. (Multi-Genrator Disposal Container Information) This section applies to disposal containers enclosing mixtures of waste originating from different generators. (Note: The origin of the LLW resulting from a processor's activities may be attributable to one or more "generators" (including "waste generators") as defined in this part). It also applies to mixtures of wastes shipped in an uncontainerized form, for which portions of the mixture within the shipment originate from different generators. 1. For homogeneous mixtures of waste, such as incinerator ash, provide the waste description applicable to the mixture and the volume of the waste attributed to each generator. 2. For heterogeneous mixtures of waste, such as the combined products from a large compactor, identify each generator contributing waste to the disposal container, and, for discrete waste types (ie., activated materials, contaminated equipment, mechanical filters, sealed source/devices, and

- 154 - wastes in solidification/stabilization media), the identities and activities of individual radionuclides contained on these waste types within the disposal container. For each generator, provide the following: (a) The volume of waste within the disposal container; (b) A physical and chemical description of the waste, including the solidification agent, if any; (c) The total weight percentage of chelating agents for any disposal container containing more than 0.1% chelating agent by weight, plus the identity of the principal chelating agent; (d) The sorbing or solidification media, if any, and the identity of the solidification media vendor and brand name if the media is claimed to meet stability requirements in 10 CFR 61.56(b); and (e) Radionuclide identities and activities contained in the waste, the masses of U - 233, U - 235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material if contained in the waste.

(Certification) An authorized representative of the waste generator, processor, or collector shall certify by signing and dating the shipment manifest that the transported materials are properly classified, described, packaged, marked, and labeled and are in proper condition for transportation according to the applicable regulations of the Department of Transportation and the Commission. A collector in signing the certification is certifying that nothing has been done to the collected waste which would invalidate the waste generator's certification.

~~(4) s$7l# #5=- ^ £:4KControl and Tracking) A. qq7]Mr *m~~

- 155 - 2. Label each disposal container (or transport package if potential radiation hazards preclude labeling of the individual disposal container) of waste to identify whether it is Class A waste, Class B waste, Class C waste, or greater then Class C waste, in accordance with ?1.55 of this chapter; 3. Conduct a quality assurance program to assure compliance with 6155 and 61.56 of this chapter (the program must include management evaluation of audits); 4. Prepare the NRC Uniform Low-Level Radioactive Waste Manifest as required by this appendix; 5. Forward a copy or electronically transfer the Uniform Low-Level Radioactive Waste Manifest to the intended consignee so that either (i) receipt of the manifest precedes the LLW shipment or (ii) the manifest is delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable; 6. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph A.5 of this section; 7. Receive acknowledgement of the receipt of the shipment in the form of a signed copy of NRC Form 540; 8. Retain a copy of or electronically store the Uniform Low-Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR Parts 30, 40, and 70 of this chapter; and 9. For any shipments or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix. B. AA& ^7\A7\ m ^: Any waste collector licensee who handles only prepackaged waste shall: 1. Acknowledge receipt of the waste from the shipper within one week of receipt by returning a signed copy of NRC Form 540; 2. Prepare a new manifest to reflect consolidated shipments that meet the requirements of this appendix. The waste collector shall ensure that, for each container of waste in the shipment, the manifest identifies the generator of that container of waste; 3. Forward a copy or electronically transfer the Uniform Low-Level Radioactive Waste Manifest to the intended consignee so that either: (i) Receipt of the manifest precedes the LLW shipment or (ii) the manifest is

- 156 - delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable; 4. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph B.3 of this section; 5. Receive acknowledgement of the receipt of the shipment in the form of a signed copy of NRC Form 540; 6. Retain a copy of or electronically store the Uniform Low-Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR parts 30, 40, and 70 of this chapter; 7. For any shipments or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix; and 8. Notify the shipper and the Administrator of the nearest Commission Regional Office listed in appendix D of this part when any shipment, or part of a shipment, has not arrived within 60 days after receipt of an advance manifest, unless notified by the shipper that the shipment has been cancelled C. J57lf Z\B\<3^7\ ^

- 157 - evaluation of audits); 6. Forward a copy or electronically transfer the Uniform Low-Level Radioactive Waste Manifest to the intended consignee so that either: (i) Receipt of the manifest precedes the LLW shipment or (ii) the manifest is delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable; 7. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph C.6 of this section; 8. Receive acknowledgement of the receipt of the shipment in the form of a signed copy of NRC Form 540; 9. Retain a copy of or electronically store the Uniform Low-Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR parts 30, 40, and 70 of this chapter; 10. For any shipment or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix; and 11. Notify the shipper and the Administrator of the nearest Commission Regional Office listed in appendix D of this part when any shipment, or part of a shipment, has not arrived within 60 days after receipt of an advance manifest, unless notified by the shipper that the shipment has been cancelled. D. s|7m *)£-*! £<8*}-7r tb <3: The land disposal facility operator shall: 1. Acknowledge receipt of the waste within one week of receipt by returning, as a minimum, a signed copy of NRC Form 540 to the shipper. The shipper to be notified is the licensee who last possessed the waste and transferred the waste to the operator. If any discrepancy exists between materials listed on the Uniform Low-Level Radioactive Waste Manifest and materials received, copies or electronic transfer of the affected forms must be returned indicating the discrepancy; 2. Maintain copies of all completed manifests and electronically store the information required by 10 CFR 61.80(1) until the Commission terminates theHcense; and 3. Notify the shipper and the Administrator of the nearest Commission Regional Office listed in appendix D of this part when any shipment, or

- 158 - part of a shipment, has not arrived within 60 days after receipt of an advance manifest, unless notified by the shipper that the shipment has been cancelled. E. && *\&7Y SHr 3-f: Any shipment or part of a shipment for which acknowledgement is not received within the times set forth in this section must: 1. Be investigated by the shipper if the shipper has not received notification or receipt within 20 days after transfer; and 2. Be traced and reported. The investigation shall include tracing the shipment and filing a report with the nearest Commission Regional Office listed in Appendix D to this part Each licensee who conducts a trace investigation shall file a written report with the appropriate NRC Regional Office within 2 weeks of completion of the investigation.

~~(1) ^-^ ^ s|7l# &^-: (6155) (Waste classification)~~

~~(a) (Classification of waste for near surface disposal)~~

(1) jDL^AV*J-(Considerations) Determination of the classification of radioactive waste involves two considerations. First, consideration must be given to the concentration of long-lived radionuclides (and their shorter-lived precursors) whose potential hazard will persist long after such precautions as institutional controls, improved waste form, and deeper disposal have ceased to be effective. These precautions delay the time when long-lived radionuclides could cause exposures. In addition, the magnitude of the potential dose is limited by the concentration and availability of the radionuclide at the time of exposure. Second, consideration must be given to the concentration of shorter-lived radionuclides for which requirements on institutional controls, waste form, and disposal methods are effective.

~~(2) s|| 71* ^(Classes of wast). (i) Class A waste is waste that is usually segregated from other waste~~

- 159 - classes at the disposal site. The physical form and characteristics of Class A waste must meet the minimum requirements set forth in ?1.56(a). If Class A waste also meets the stability requirements set forth in ?1.56(b), it is not necessary to segregate the waste for disposal. (ii) Class B waste is waste that must meet more rigorous requirements on waste form to ensure stability after disposal The physical form and characteristics of Class B waste must meet both the minimum and stability requirements set forth in 61.56. (iii) Class C waste is waste that not only must meet more rigorous requirements on waste form to ensure stability but also requires additional measures at the disposal facility to protect against inadvertent intrusion. The physical form and characteristics of Class C waste must meet both the minimum and stability requirements set forth in 61.56. (iv) Waste that is not generally acceptable for near-surface disposal is waste for which form and disposal methods must be different, and in general more stringent, than those specified for Class C waste. In the absense of specific requirements in this part, such waste must be disposed of in a geologic repository as defined in part 60 of this chapter unless proposals for disposal of such waste in a disposal site licensed pursuant to this part are approved by the Commission.

(3) Classification determined by long-lived radionuclides. If radioactive waste contains only radionuclides listed in Table 1, classification shall be determined as follows- (i) If the concentration does not exceed 0.1 times the value in Table 1, the waste is Class A. (ii) If the concentration exceeds 0.1 times the value in Table 1 but does not exceed the value in Table 1, the waste is Class C. (iii) If the concentration exceeds the value in Table 1, the waste is not generally acceptable for near-surface disposal, (iv) For wastes containing mixtures of radionuclides listed in Table 1, the total concentration shall be determined by the sum of fractions

- 160 - Table 1 Concentration curies Radionuclide per cubic meter C-14 8 C-14 in activated metal 80 Ni-59 in activated metal 220 Nb-94 in activated metal 0.2 Tc-99 3 1-129 0.08 Alpha emitting transuranic nuclides with 'lOO half-life greater than 5 years Pu-241 '3,500 Cm-242 '20,000 'Units are nanocuries per gram.

(4) #71%^ SJ^ 4# sj|7l«- £%• Classification determined by short-lived radionuclides. If radioactive waste does not contain any of the radionuclides listed in Table 1, classification shall be determined based on the concentrations shown in Table 2. However, as specified in paragraph (a)(6) of this section, if radioactive waste does not contain any nuclides listed in either Table 1 or 2, it is Class A. (i) If the concentration does not exceed the value in Column 1, the waste is Class A. (ii) If the concentration exceeds the value in Column 1, but does not exceed the value in Column 2, the waste is Class B. (iii) If the concentration exceeds the value in Column 2, but does not exceed the value in Column 3, the waste is Class C. (iv) K the concentration exceeds the value in Column 3, the waste is not generally acceptable for near-surface disposal.

~~(v) For wastes containing mixtures of the nuclides listed in Table 2, the total concentration shall be determined by the sum of fractions rule~~

- 161 - Table 2 Concentration, curies per cubic meter Radionuclide CoL 1 CoL 2 CoL 3 Total of all nuclides with less 700 (1) (1) than 5 year half-life H-3 40 (1) (1) Co-60 700 (1) (1) Ni-63 3.5 70 700 Ni-63 in activated metal 35 700 7000 Sr-90 0.04 150 7000 Cs-137 1 44 4600 There are no limits established for these radionuclides in Class B or C wastes.Practical considerations such as the effects of external radiation and internal heat generation on transportation, handling, and disposal will limit the concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in Table 2 determine the waste to the Class C independent of these nuclides.

(5) Classification determined by both long- and short-lived radionuclides. If radioactive waste contains a mixture of radionuclides, some of which are listed in Table 1, and some of which are listed in Table 2, classification shall be determined as follows: (i) If the concentration of a nuclide listed in Table 1 does not exceed 0.1 times the value listed in Table 1, the class shall be that determined by the concentration of nuclides listed in Table 2. (ii) If the concentration of a nuclide listed in Table 1 exceeds 0.1 times the value listed in Table 1 but does not exceed the value in Table 1, the waste shall be Class C, provided the concentration of nuclides listed in Table 2 does not exceed the value shown in Column 3 of Table 2.

(6) Table 1, Table 2 *]$$ qfH Hflfc sfl7}# Classification of wastes with radionuclides other than those listed in Tables 1 and 2. If radioactive waste does not contain any nuclides listed in either Table 1 or 2, it is Class A.

- 162 - (7) The sum of the fractions tule for mixtures of radionuclides. For determining classification for waste that contains a mixture of radionuclides, it is necessary to determine the sum of fractions by dividing each nuclide's concentration by the appropriate limit and adding the resulting values. The appropriate limits must all be taken from the same column of the same table. The sum of the fractions for the column must be less than 1.0 if the waste class is to be determined by that column. Example: A waste contains Sr-90 in a concentration of 50 Ci/m3. and Cs-137 in a concentration of 22 Ci/m3. Since the concentrations both exceed the values in Column 1, Table 2, they must be compared to Column 2 values. For Sr-90 fraction 50/150=0.33; for Cs-137 fraction, 22/44=05; the sum of the fractions=0.83. Since the sum is less than 1.0, the waste is Class B.

(8) Determination of concentrations in wastes. The concentration of a radionuclide may be determined by indirect methods such as use of scaling factors which relate the inferred concentration of one radionuclide to another that is measured, or radionuclide material accountability, if there is reasonable assurance that the indirect methods can be correlated with actual measurements. The concentration of a radionuclide may be averaged over the volume of the waste, or weight of the waste if the units are expressed as nanocuries per gram.

~~- 163 - (2), 3-*] *J^ 34 7] # #^: (61.56) (Waste characteristics)~~

(a) Vq *^ ^7]^^ %& SL? The following requirements are minimum requirements for all classes of waste and are intended to facilitate handling at the disposal site and provide protection of health and safety of personnel at the disposal site. (1) Waste must not be packaged for disposal in cardboard or fiberboard boxes. (2) Liquid waste must be solidified or packaged in sufficient absorbent material to absorb twice the volume of the liquid. (3) Solid waste containing liquid shall contain as little free standing and noncorrosive liquid as is reasonably achievable, but in no case shall the liquid exceed 1% of the volume. (4) Waste must not be readily capable of detonation or of explosive decomposition or reaction at normal pressures and temperatures, or of explosive reaction with water. (5) Waste must not contain, or be capable of generating, quantities of toxic gases, vapors, or fumes harmful to persons transporting, handling, or disposing of the waste. This does not apply to radioactive gaseous waste packaged in accordance with paragraph (a)(7) of this section. (6) Waste must not be pyrophoric. Pyrophoric materials contained in waste shall be treated, prepared, and packaged to be nonflammable. (7) Waste in a gaseous form must be packaged at a pressure that does not exceed 1.5 atmospheres at 20HJ. Total activity must not exceed 100 curies per container. (8) Waste containing hazardous, biological, pathogenic, or infectious material must be treated to reduce to the maximum extent practicable the potential hazard from the non-radiological materials.

(b) The requirements in this section are intended to provide stability of the waste. Stability is intended to ensure that the waste does not structurally degrade and affect overall stability of the site through slumping, collapse, or other failure of the disposal unit and thereby lead to water infiltration. Stability is also a factor in limiting exposure to an inadvertent intruder, since

- 164 - it provides a recognizable and nondispersible waste. (1) Waste must have structural stability. A structurally stable waste form will generally maintain its physical dimensions and its form, under the expected disposal conditions such as weight of overburden and compaction equipment, the presence of moisture, and microbial activity, and internal factors such as radiation effects and chemical changes. Structural stability can be provided by the waste form itself, processing the waste to a stable form, or placing the waste in a disposal container or structure that provides stability after disposal. (2) Notwithstanding the provisions in ?1.56(a) (2) and (3), liquid wastes, or wastes containing liquid, must be converted into a form that contains as little free standing and noncorrosive liquid as is reasonably achievable, but in no case shall the liquid exceed 1% of the volume of the waste when the waste is in a disposal container designed to ensure stability, or 0.5% of the volume of the waste for waste processed to a stable form. (3) Void spaces within the waste and between the waste and its package must be reduced to the extent practicable.

~~(Purpose and scope)~~

(a) This part establishes — (1) Requirements for packaging, preparation for shipment, and transportation of licensed material; and (2) Procedures and standards for NRC approval of packaging and shipping procedures for fissile material and for a quantity of other licensed material in excess of a Type A quantity. (b) The packaging and transport of licensed material are also subject to other parts of tiiis chapter (e.g., 10 CFR parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the U.S. Department of Transportation (DOT) and the U.S. Postal Service(D) having jurisdiction over means of transport. The requirements of this part are in addition to, and not in substitution for, other requirements. (c) The regulations in this part apply to any licensee authorized by specific or

- 165 - general license issued by the Commission to receive, possess, use, or transfer licensed material, if the licensee delivers that material to a carrier for transport, transports the material outside the site of usage as specified in the NRC license, or transports that material on public highways. No provision of this part authorizes possession of licensed materiaL (d) Exemptions from the requirement for license in ?L3 are specified in ?1.10. General licenses for which no NRC package approval is required are issued in 71.14 through 7154. The general license in ?1.12 requires that an NRC certificate of compliance or other package approval be issued for the package to be used under the general license. Application for package approval must be completed in accordance with subpart D of this part, demonstrating mat the design of the package to be used satisfies the package approval standards contained in subpart E of this part, as related to the tests of subpart F of this part The transport of licensed material or delivery of licensed material to a carrier for transport is subject to the operating controls and procedures requirements of subpart G of this part, to the quality assurance requirements of subpart H of this part, and to the general provisions of subpart A of this part, including DOT regulations referenced in ?1.5. (e) The regulations in this part apply to any person required to obtain a certificate of compliance or an approved compliance plan pursuant to part 76 of this chapter if the person delivers radioactive material to a common or contract carrier for transport or transports the material outside the confines of the person's plant or other authorized place of use. (f) This part also gives notice to all persons who knowingly provide to any licensee, certificate holder, quality assurance program approval holder, applicant for a license, certificate, or quality assurance program approval or to a contractor, or subcontractor of any of them, components, equipment, materials, or other goods or services, that relate to a licensee's, certificate holder's, quality assurance program approval holder's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of 71.11.

~~(2) (Definitions) Al means the maximum activity of special form radioactive material permitted in a Type A package. A2 means the maximum activity of radioactive material, other than special form,~~

- 166 - LSA and SCO material, permitted in a Type A package. These values are either listed in Appendix A of this part, Table A - 1, or may be derived in accordance with the procedure prescribed in Appendix A of this part

Low Specific Activity (LSA) material means radioactive material with limited specific activity that satisfies the descriptions and limits set forth below. Shielding materials surrounding the LSA material may not be considered in determining the estimated average specific activity of the package contents. LSA material must be in one of three groups'. (1) LSA - L (i) Ores containing only naturally occurring radionuclides (e.g., uranium, thorium) and uranium or thorium concentrates of such ores; or (ii) Solid unirradiated natural uranium or depleted uranium or natural thorium or their solid or liquid compounds or mixtures; or (iii) Radioactive material, other than fissile material, for which the A2 value is unlimited; or (iv) Mill tailings, contaminated earth, concrete, rubble, other debris, and activated material in which the radioactive material is essentially uniformly distributed, and the average specific activity does not exceed 106 A2/s. (2) LSA - IL (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/liter); or (ii) Material in which the radioactive material is distributed throughout, and the average specific activity does not exceed 10-4 A2/g for solids and gases, and 10-5 A2/g for liquids. (3) LSA - HL Solids (e.g., consolidated wastes, activated materials) in which: (i) The radioactive material is distributed throughout a solid or a collection of solid objects, or is essentially uniformly distributed in a solid compact binding agent (such as concrete, bitumen, ceramic, etc.); and (ii) The radioactive material is relatively insoluble, or it is intrinsically contained in a relatively insoluble material, so that, even under loss of packaging, the loss of radioactive material per package by leaching, when placed in water for 7 days, would not exceed 0.1 A2; and (iii) The average specific activity of the solid does not exceed 2 xlO-3 A2/g.

~~Surface Contaminated Object (SCO) means a solid object that is not itself classed as radioactive material, but which has radioactive material distributed on~~

- 167 - any of its surfaces. SCO must be in one of two groups with surface activity not exceeding the following limits: (1) SCO - I: A solid object on which-" (i) The non-fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 Bq/cm2 (10-4 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 0.4 Ba/cm2 (10-5 raicrocurie/cm2) for all other alpha emitters; (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4x104 Bq/cm2 (1.0 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4x103 Bo/cm2 (0.1 microcurie/cm2) for all other alpha emitters; and (iii) The non-fixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 40x104 Bq/cm2 (1 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 40x103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters. (2) SCO - H: A solid object on which the limits for SCO - I are exceeded and on which: (i) The non-fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 400 Bq/cm2 (10-2 microcurie/cm2) for beta and gamma and low toxicity alpha emitters or 40 Bq/cm2 (10-3 microcurie/cm2) for all other alpha emitters; (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8x105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8x104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters; and (iii) The non-fixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8x105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8x104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters.

~~Transport index means the dimensionless number (rounded up to the next tenth) placed on the label of a package, to designate the degree of control to be~~

- 168 - exercised by the carrier during transportation. The transport index is determined as follows^ (1) For non-fissile material packages, the number determined by multiplying the maximum radiation level in millisievert (mSv) per hour at one meter (3.3 ft) from the external surface of the package by 100 (equivalent to the maximum radiation level in millirem per hour at one meter (3.3 ft)); or (2) For fissile material packages, the number determined by multiplying the maximum radiation level in millisievert per hour at one meter (3.3 ft) from the external surface of the package by 100 (equivalent to the maximum radiation level in millirem per hour at one meter (3.3 ft)), or, for criticality control purposes, the number obtained as described in ?1.59, whichever is larger.

Type A quantity means a quantity of radioactive material, the aggregate radioactivity of which does not exceed Al for special form radioactive material, or A2, for normal form radioactive material, where Al and A2 are given in Table A - 1 of this part, or may be determined by procedures described in Appendix A of this part :

~~Type B quantity means a quantity of radioactive material greater than a Type A quantity.~~

Package means the packaging together with its radioactive contents as presented for transport (1) Fissile material package means a fissile material packaging together with its fissile material contents. (2) Type B package means a Type B packaging together with its radioactive contents. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating pressure of more than 700 kPa (100 Ib/in2) gauge or a pressure relief device that would allow the release of radioactive material to the environment under the tests specified in 71.73 (hypothetical accident conditions), in which case it will receive a designation B(M). B(U) refers to the need for unilateral approval of international shipments! B(M) refers to the need for multilateral approval of international shipments. There is no distinction made in how packages with these designations may be used in domestic transportation. To determine their distinction for international transportation, see DOT

~~- 169 - regulations in 49 CFR Part 173. A Type B package approved before September 6, 1983, was designated only as Type B. Limitations on its use are specified in 71.13.~~

Packaging means the assembly of components necessary to ensure compliance with the packaging requirements of this part It may consist of one or more receptacles, absorbent materials, spacing structures, thermal insulation, radiation shielding, and devices for cooling or absorbing mechanical shocks. The vehicle, tie-down system, and auxiliary equipment may be designated as part of the packaging.

(3) (Transportation of licensed material) (a) Each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 170 through 189 appropriate to the mode of transport (1) The licensee shall particularly note DOT regulations in the following areas: (i) Packaging — 49 CFR part 173: Subparts A and B and L (ii) Marking and labeling — 49 CFR part 172: Subpart D, 172.400 through 172.407, 172.436 through 172.440, and subpart E. (iii) Placarding — 49 CFR part 172: Subpart F, especially 172.500 through 172519, 172556, and appendices B and C. (iv) Accident reporting — 49 CFR part 171: 171.15 and 171.16. (v) Shipping papers and emergency information — 49 CFR part 172: Subparts C and G. (vi) Hazardous material employee training — 49 CFR part 172: Subpart H. (vii) Hazardous material shipper/carrier registration — 49 CFR part 107: Subpart G. (2) The licensee shall also note DOT regulations pertaining to the following modes of transportation: (i) Rail ~ 49 CFR part 174: Subparts A through D and K. (ii) Air — 49 CFR part 175. (iii) Vessel -- 49 CFR part 176: Subparts A through F and M. (iv) Public Highway — 49 CFR part 177 and parts 390 through 397. (b) If DOT regulations are not applicable to a shipment of licensed material, the

- 170 - licensee shall conform to the standards and requirements of the DOT specified in paragraph (a) of this section to the same extent as if the shipment or transportation were subject to DOT regulations. A request for modification, waiver, or exemption from those requirements, and any notification referred to in those requirements, must be filed with, or made to, the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555 - 0001.

~~Subpart C -- Qitq -8-71*1 «| ^(General Licenses)~~

(4) NRC 3^ -8-71 (71.12) (General license: NRC-approved package) (a) A general license is hereby issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package for which a license, certificate of compliance, or other approval has been issued by the NRC. (b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part (c) This general license applies only to a licensee who — (1) Has a copy of the certificate of compliance, or other approval of the package, and has the drawings and other documents referenced in the approval relating to the use and maintenance of the packaging and to the actions to be taken before shipment; (2) Complies with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part; and (3) Submits in writing to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555 - 0001, before the licensee's first use of the package, the licensee's name and license number and the package identification number specified in the package approval. (d) This general license applies only when the package approval authorizes use of the package under this general license. (e) For a Type B or fissile material package, the design of which was approved by NRC before April 1, 1996, the general license is subject to the additional restrictions of 71.13.

- 171 - (5) 71 €^ *W€ -§-71(71.13) (Previously approved package) (a) A Type B package previously approved by NRC but not designated as B(U) or B(M) in the identification number of the NRC Certificate of Compliance, may be used under the general license of 71.12 with the following additional conditions: (1) Fabrication of the packaging was satisfactorily completed by August 31, 1986, as demonstrated by application of its model number in accordance with 71.85(c); (2) A package used for a shipment to a location outside the United States is subject to multilateral approval, as defined in DOT regulations at 49 CFR 173.403; and (3) A serial number that uniquely identifies each packaging which conforms to the approved design is assigned to, and legibly and durably marked on, the outside of each packaging. (b) A Type B(U) package, a Type B(M) package, a low specific activity (LSA) material package or a fissile material package, previously approved by the NRC but without the designation "-85" in the identification number of the NRC Certificate of Compliance, may be used under the general license of 71.12 with the following additional conditions: (1) Fabrication of the package is satisfactorily completed by April 1, 1999 as demonstrated by application of its model number in accordance with 71.85(c); (2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403; and (3) A serial number which uniquely identifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging. (c) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided — (1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified

- 172 - in 71.71 and 71.73; (2) The modifications of a fissile material package are not significant, with respect to the prevention of criticality, when the package is subjected to the tests specified in 71.71 and 71.73; and (3) The modifications to the package satisfy the requirements of this part (d) NRC will revise the package identification number to designate previously approved package designs as B(U), B(M), AF, BF, or A as appropriate, and with the identification number suffix "-85" after receipt of an application demonstrating that the design meets the requirements of this part

(6) ^r^r^ (Department Of Transport) *|3i -§-71(71.14) (General license: DOT specification container) (a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a specification container for fissile material or for a Type B quantity of radioactive material as specified in DOT regulations at 49 CFR parts 173 and 178. (b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part (c) This general license applies only to a licensee who — (1) Has a copy of the specification; and (2) Complies with the terms and conditions of the specification and the applicable requirements of subparts A, G, and H of this part (d) This general license is subject to the limitation that the specification container may not be used for a shipment to a location outside the United States, except by multilateral approval, as defined in DOT regulations at 49 CFR 173.403.

(7) (General license: Use of foreign approved package) (a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package the design of which has been approved in a foreign national competent authority certificate that has been revalidated by DOT as meeting the applicable requirements of 49 CFR 171.12. (b) Except as otherwise provided in this section, the general license applies only to a licensee who has a quality assurance program approved by the

- 173 - Commission as satisfying the applicable provisions of subpart H of this part (c) This general license applies only to shipments made to or from locations outside the United States. (d) This general license applies only to a licensee who — (1) Has a copy of the applicable certificate, the revalidation, and the drawings and other documents referenced in the certificate, relating to the use and maintenance of the packaging and to the actions to be taken before shipment; and (2) Complies with the terms and conditions of the certificate and revalidation, and with the applicable requirements of subparts A, G, and H of this part. With respect to the quality •assurance provisions of subpart H of this part, the licensee is exempt from design, construction, and fabrication considerations.

~~(8) iZ\7\ ^^r<% i£.SL® A>%#(71.31)~~

(a) An application for an approval under this part must include, for each proposed packaging design, the following information: (1) A package description as required by 71.33; (2) A package evaluation as required by 71.35; and (3) A quality assurance program description, as required by ?1.37, or a reference to a previously approved quality assurance program. (b) Except as provided in 71.13, an application for modification of a package design, whether for modification of the packaging or authorized contents, must include sufficient information to demonstrate that the proposed design satisfies the package standards in effect at the time the application is filed. (c) The applicant shall identify any established codes and standards proposed for use in package design, fabrication, assembly, testing, maintenance, and use. In the absence of any codes and standards, the applicant shall describe and justify the basis and rationale used to formulate the package quality assurance program.

(9) -8-71^ -g^ (71.33) (Package description) The application must include a description of the proposed package in sufficient detail to identify the package accurately and provide a sufficient basis for

- 174 - evaluation of the package. The description must include — (a) With respect to the packaging — (1) Classification as Type B(U), Type B(M), or fissile material packaging; (2) Gross weight; (3) Model number; (4) Identification of the containment system; (5) Specific materials of construction, weights, dimensions, and fabrication methods of — (i) Receptacles; (ii) Materials specifically used as nonfissile neutron absorbers or moderators', (iii) Internal and external structures supporting or protecting receptacles; (iv) Valves, sampling ports, lifting devices, and tie-down devices; and (v) Structural and mechanical means for the transfer and dissipation of heat; and (6) Identification and volumes of any receptacles containing coolant (b) With respect to the contents of the package — (1) Identification and maximum radioactivity of radioactive constituents; (2) Identification and maximum quantities of fissile constituents; (3) Chemical and physical form; (4) Extent of reflection, the amount and identity of nonfissile materials used as neutron absorbers or moderators, and the atomic ratio of moderator to fissile constituents; (5) Maximum normal operating pressure; (6) Maximum weight; (7) Maximum amount of decay heat; and (8) Identification and volumes of any coolants.

(10) -8-71^1 sg7r (71.35) (Package evaluation) The application must include the following: (a) A demonstration that the package satisfies the standards specified in subparts E and F of this part; (b) For a fissile material package, the allowable number of packages that may be transported in the same vehicle in accordance with ?1.59; and (c) For a fissile material shipment, any proposed special controls and precautions

~~- 175 - for transport, loading, unloading, and handling and any proposed special controls in case of an accident or delay.~~

(11) ^€ ^$1(71.37) (Quality assurance) (a) The applicant shall describe the quality assurance program (see Subpart H of this part) for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the proposed package. (b) The applicant shall identify any specific provisions of the quality assurance program that are applicable to the particular package design under consideration, including a description of the leak testing procedures.

(12) *\7\ &*\ 3£fe ^£ SKISJ ^(71.38) (Renewal of a certificate of compliance or quality assurance program approval) (a) Except as provided in paragraph (b) of this section, each Certificate of Compliance or Quality Assurance Program Approval expires at the end of the day, in the month and year stated in the approval. (b) In any case in which a person, not less than 30 days before the expiration of an existing Certificate of Compliance or Quality Assurance Program Approval issued pursuant to the part, has filed an application in proper form for renewal of either of those approvals, the existing Certificate of Compliance or Quality Assurance Program Approval for which the renewal application was filed shall not be deemed to have expired until final action on the application for renewal has been taken by the Commission. (c) In applying for renewal of an existing Certificate of Compliance or Quality Assurance Program Approval, an applicant may be required to submit a consolidated application that incorporates all changes to its program that, are incorporated by reference in the existing approval or certificate, into as few referenceable documents as reasonably achievable.

(13) -¥-7> ?§£.$] .8-^(71.39) (Requirement for additional information) The Commission may at any time require additional information in order to enable it to determine whether a license, certificate of compliance, or other approval should be granted, renewed, denied, modified, suspended, or revoked.

- 176 - (14) 3^ ^ £(71.41) (Demonstration of compliance) (a) The effects on a package of the tests specified in 71.71 ("Normal conditions of transport"), and the tests specified in 71.73 ("Hypothetical accident conditions1'), and 71.61 (Special requirement for irradiated nuclear fuel shipments"), must be evaluated by subjecting a specimen or scale model to a specific test, or by an her method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered. (b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper, the Commission may permit the shipment to be evaluated together with the transporting vehicle. (c) Environmental and test conditions different from those specified in 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment.

(15) SL^ 3L##<*3 tfl^: g# fl-^g (71.43) (General standards for all packages) (a) The smallest overall dimension of a package may not be less than 10 cm (4 in). (b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons. (c) Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package. (d) A package must be made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from inleakage of water, to the maximum credible extent Account must be taken of the behavior of materials under irradiation. (e) A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an

- 177 - enclosure to retain any leakage. (f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in 71.71 ("Normal conditions of transport") there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging. (g) A package must be designed, constructed, and prepared for transport so that in still air at 38C (100F) and in the shade, no accessible surface of a package would have a temperature exceeding 50C (122F) in a nonexclusive use shipment, or 85C (185F) in an exclusive use shipment (h) A package may not incorporate a feature intended to allow continuous venting during transport

~~(16) S-^ -g-71^ ^-^-^ #?1 ^MH- fl-^ (71.45) (Lifting and tie-down standards for all packages) (a)~~

- 178 - package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices. (3) Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of this part

(17) 5L^ -§-71 <* « 4}-»f ifi-XHr ff;g(71.47) (External radiation standards for all packages) (a) Except as provided in paragraph (b) of this section, each package of radioactive materials offered for transportation must be designed and prepared for shipment so that under conditions normally incident to transportation the radiation level does not exceed 2 mSv/h (200 mrem/h) at any point on the external surface of the package, and the transport index does not exceed 10. (b) A package that exceeds the radiation level limits specified in paragraph (a) of this section must be transported by exclusive use shipment only, and the radiation levels for such shipment must not exceed the following during transportation: (1) 2 mSv/h (200 mrem/h) on the external surface of the package, unless the following conditions are met, in which case the limit is 10 mSv/h (1000 mrem/h): (i) The shipment is made in a closed transport vehicle; (ii) The package is secured within the vehicle so that its position remains fixed during transportation; and (iii) There are no loading or unloading operations between the beginning and end of the transportation; (2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the vehicle, including the top and underside of the vehicle; or in the case of a flat-bed style vehicle, at any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load or enclosure, if used, and on the lower external surface of the vehicle; and (3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the outer lateral surfaces of the vehicle (excluding the top and underside of the vehicle); or in the case of a flat-bed style vehicle, at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the vehicle (excluding the top and underside of the vehicle); and

- 179 - (4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except that this provision does not apply to private carriers, if exposed personnel under their control wear radiation dosimetry devices in confonnance with 10 CFR 20.1502. (c) For shipments made under the provisions of paragraph (b) of this section, the shipper shall provide specific written instructions to the carrier for maintenance of the exclusive use shipment controls. The instructions must be included with the shipping paper information. (d) The written instructions required for exclusive use shipments must be sufficient so that, when followed, they will cause the carrier to avoid actions that will unnecessarily delay delivery or unnecessarily result in increased radiation levels or radiation exposures to transport workers or members of the general public.

(18) Type B-§-7H « ^H?] -8-TT *W (71.51) (Additional requirements for Type B packages) (a) Except as provided in 71.52, a Type B package, in addition to satisfying the requirements of 71.41 through 71.47, must be designed, constructed, and prepared for shipment so that under the tests specified in: (1) Section 71.71 ("Normal conditions of transport"), there would be no loss or dispersal of radioactive contents — as demonstrated to a sensitivity of 106 A2 per hour, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and (2) Section 71.73 ("Hypothetical accident conditions"), there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 rem/h) at 1 m (40 in) from the external surface of the package. (b) Where mixtures of different radionuclides are present, the provisions of appendix A, paragraph IV of this part shall apply, except that for Krypton-85, an effective A2 value equal to 10 A2 may be used. (c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system.

~~(19) ^HW9 -§-7H « &A 3^ (71.52) (Exemption for low-specific-activity (LSA) packages) A package need not satisfy the requirements of 71.51 if it contains only LSA or~~

~~- 180 - SCO material, and is transported as exclusive use, but is subject to 71.41 through 71.47, including 71.43(f). This section expires April 1, 1999.~~

(20) «*€r^ #€^} £*| 3=^ (71.53) (Fissile material exemptions) Fissile materials meeting the requirements of one of the paragraphs in (a) through (d) of this section are exempt from fissile material classification and from the fissile material package standards of 7155 and 71.59, but are subject to all other requirements of this part These exemptions apply only when beryllium, graphite, or hydrogenous material enriched in deuterium is not present in quantities exceeding 0.1 percent of the fissile material mass, (a) Fissile material such that less than or equal to 1 for an individual consignment, where X and Y are the mass limits defined in table following paragraph (a)(3) of this section, provided that (1) Each package contains no more than 15 g of fissile material. For unpackaged material the mass limit of 15g applies to the conveyance; or (2) The fissile material consists of a homogeneous hydrogenous solution or mixture where the minimum ratio of hydrogen atoms to fissile radionuclide atoms (H/X) is 5200 and the maximum concentration of fissile radionuclides within a package is 5 g/liter; or (3) There is no more than 5g of fissile material in any 10 liter volume of material and the material is packaged so as to maintain this limit of fissile radionuclide concentration during normal transport

Fissile materials mass (g) Fissile material mass (g) mixed with substances having mixed with substances Fissile Material an average hydrogen density having an average hydrogen less than or equal to water density greater than water Uranium-235(X) 400 290 lOther fissile 250 180 material(Y) The Requirements for Packages Containing Fissile Material

~~(b) Uranium enriched in uranium-235 to a maximum of 1 percent by weight, and with total plutonium and uranium-233 content of up to 1 percent of the~~

- 181 - mass of uranium-235, provided that the fissile material is distributed homogeneously throughout the package contents and does not form a lattice arrangement within the package. (c) Liquid solutions of uranyl nitrate enriched in uranium-235 to a maximum of 2 percent by weight, with a total plutonium and uranium-233 content not exceeding 0.1 percent of the mass of uranium-235, and with a minimum nitrogen to uranium atomic ratio (N/U) of 2. (d) Plutonium, less than 1 kg, of which not more than 20 percent by mass may consist of plutonium-239, plutonium-241, or any combination of these radionuclides.

(21) n&Q #^ 3£=3--8-7lo| rfl^ <£#Z\ _£L^- A\*£ (71.55) (General requirements for fissile material packages) (a) A package used for the shipment of fissile material must be designed and constructed in accordance with 71.41 through 71.47. When required by the total amount of radioactive material, a package used for the shipment of fissile material must also be designed and constructed in accordance with 71.51. (b) Except as provided in paragraph (c) of this section, a package used for the shipment of fissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained: (1) The most reactive credible configuration consistent with the chemical and physical form of the material; (2) Moderation by water to the most reactive credible extent; and (3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging. (c) The Commission may approve exceptions to the requirements of paragraph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to ensure that the containment system does not leak. (d) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in

- 182 - 71.71 ("Normal conditions of transport") — (1) The contents would be subcritical; (2) The geometric form of the package contents would not be substantially altered; (3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under ?1.59(a)(l), it has been assumed that moderation is present to such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and (4) There will be no substantial reduction in the effectiveness of the packaging, including: (i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed; (ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer surface of the packaging; and (iii) No occurrence of an aperture in the outer surface of the packaging large enough to permit the entry of a 10 cm (4 in) cube. (e) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in 71.73 ("Hypothetical accident conditions"), the package would be subcritical. For this determination, it must be assumed that: (1) The fissile material is in the most reactive credible configuration consistent with the damaged condition of the package and the chemical and physical form of the contents; (2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and (3) There is full reflection by water on all sides, as close as is consistent with the damaged condition of the package.

- 183 - (22) ^*<* ^-fi-^: i##s) t#3 S3(71.71) (Normal conditions of transport) (a) %7h Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section. Separate specimens may be used for the free drop test, the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests. (b) 3,7} &&. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature preceding and following the tests remaining INSOLATION DATA Total insolation for a Form and location of surface 12-hour period (gcal/cm2) Flat surfaces transported horizontally; Base None Other surfaces 800 Flat surfaces not transported horizontally 200 Curved surfaces 400 constant at that value between -29C and +38C which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable.

~~(c)~~

(1) Heat An ambient temperature of 38C (100F) in still (2) Cold. An ambient temperature of -40C (-40F) in still air and shade. (3) Reduced external pressure. An external pressure of 25 kPa (3.5 Ibf/in2) absolute. (4) Increased external pressure. An external pressure of 140 kPa (20 Ibf/In2) absolute.

- 184 - (5) Vibration. Vibration normally incident to transport (6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour. (7) Free drop. Between 15 and 25 hours after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. CRITERIA FOR FREE DROP TEST (WEIGHT/DISTANCE) Package weight Free drop distance Kilograms (Pounds) Meters (Feet) Less than 5,000 (Less than 11,000) 1.2 (4) 5,000 to 10,000 (11,000 to 22,000) 0.9 (3) 10,000 to 15,000 (22,000 to 33,100) 0.6 (2) More than 15,000 (More than 33,100) 0.3 (1)

(8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lbs). (9) Compression. For packages weighing up to 5000 kg (11,000 lbs), the package must be subjected, for a period of 24 hours, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load must be the greater of the following: (i) The equivalent of 5 times the weight of the package; or (ii) The equivalent of 13 kPa (2 Ibf/ln2) multiplied by the vertically projected area of the package. (10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, dropped from a height of 1 m (40 in) onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.

- 185 - (23) (Hypothetical accident conditions) (a) €^ &H. Evaluation for hypothetical accident conditions is to be based on sequential application of the tests specified in this section, in the order indicated, to determine their cumulative effect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraph (c)(6) of this section.

(b) With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of this part during testing, the ambient air temperature before and after the tests must remain constant at that value between -29C (-20F) and +38C (+100F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the maximum normal operating pressure, unless a lower internal pressure, consistent with the ambient temperature assumed to precede and follow the tests, is more unfavorable.

(c) Tests for hypothetical accident conditions must be conducted as follows: (1) *r-n-^"3KFree drop). A free drop of the specimen through a distance of 9 m (30 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. (2) #iKCrush). Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding, horizontal surface so as to suffer maximum damage by the drop of a 500 kg (1100 pound) mass from 9 m (30 ft) onto the specimen. The mass must consist of a solid mild steel plate 1 m (40 in) by 1 m and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 500 kg (1100 lbs), an overall density not greater than 1000 kg/m3 (62.4 Ibs/ft3) based on external dimensions, and radioactive contents greater than 1000 A2 not as special form radioactive material. (3) ^-f-(Puncture). A free drop of the specimen through a distance of 1 m (40 in) in a position for which maximum damage is expected, onto the

- 186 - upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (6 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not less than 20 cm (8 in) long. The long axis of the bar must be vertical. (4) *fl*i (Thermal). Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 800C (1475F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800C. The fuel source must extend horizontally at least 1 m (40 in), but may not extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must be positioned 1 m (40 in) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the cohvective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally. (5) 41 T1 — ^•S-'I #€ (Immersion — fissile material). For fissile material subject to 71.55, in those cases where water inleakage has not been assumed for criticality analysis, immersion under a head of water of at least 0.9 m (3 ft) in the attitude for which maximum leakage is expected (6) ^^r — 2-TE- -%r7} (Immersion — all packages). A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of water of 150 kPa (21.7lbf/in2) gauge is considered to meet these conditions.

~~(24) LSA - m ^4f HMMI #€^1 #^«3 (71.77) (Qualification of LSA - HI Material) (a) LSA - IE material must meet the test requirements of paragraph (b) of this~~

- 187 - section. Any differences between the specimen to be tested and the material to be transported must be taken into account in determining whether the test requirements have been met (b) Leaching Test (1) The specimen, representing no less than the entire contents of the package, must be immersed for 7 days in water at ambient temperature; (2) The volume of water to be used in the test must be sufficient to ensure that at the end of the test period the free volume of the unabsorbed and unreacted water remaining will be at least 10% of the volume of the specimen itself; (3) The water must have an initial pH of 6 - 8 and a maximum conductivity 10 micromho/cm at 20C (68F); and (4) The total activity of the free volume of water must be measured following the 7 day immersion test and must not exceed 0.1 A2.

(25) <*"! ^3 (71.85) (Preliminary determinations) Before the first use of any packaging for the shipment of licensed material — (a) The licensee shall ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce the effectiveness of the packaging; (b) Where the maximum normal operating pressure will exceed 35 kPa (5 Ibf/in2) gauge, the licensee shall test the containment system at an internal pressure at least 50 percent higher than the maximum normal operating pressure, to verify the capability of that system to maintain its structural integrity at that pressure; and (c) The licensee shall conspicuously and durably mark the packaging with its model number, serial number, gross weight, and a package identification number assigned by NRC. Before applying the model number, the licensee shall determine that the packaging has been fabricated in accordance with the design approved by the Commission.

~~(26) W8 -g-7H « ^£*1~~

- 188 - of the license. The licensee shall determine that — (a) The package is proper for the contents to be shipped; (b) The package is in unimpaired physical condition except for superficial defects such as marks or dents; (c) Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects," (d) Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid; (e) Any pressure relief device is operable and set in accordance with written procedures; (f) The package has been loaded and closed in accordance with written procedures; (g) For fissile material, any moderator or neutron absorber, if required, is present and in proper condition; (h) Any structural part of the package that could be used to lift or tie down the package during transport is rendered inoperable for that purpose, unless it satisfies the design requirements of 71.45; (i) The level of non-fixed (removable) radioactive contamination on the external surfaces of each package offered for shipment is as low as reasonably achievable, and within the limits specified in DOT regulations in 49 CFR 173.443; (j) External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in ?1.47 at any time during transportation; and (k) Accessible package surface temperatures will not exceed the limits specified in 71.43(g) at any time during transportation.

(27) 24 ^ *1^ (71.93) (Inspection and tests) (a) The licensee or certificate holder shall permit the Commission, at all reasonable times, to inspect the licensed material, packaging, premises, and facilities in which the licensed material or packaging is used, provided, constructed, fabricated, tested, stored, or shipped. (b) The licensee shall perform, and permit the Commission to perform, any tests the Commission deems necessary or appropriate for the administration of the regulations in this chapter. (c) The licensee shall notify the Administrator of the appropriate NRC Regional

- 189 - Office listed in appendix A of part 73 of this chapter, at least 45 days before fabrication of a package to be used for the shipment of licensed material having a decay heat load in excess of 5 kW or with a maximum normal operating pressure in excess of 103 kPa (15 Ibf/in2) gauge.

(28) 2£ 3L (71.95) (Reports) The licensee shall report to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, within 30 days — (a) Any instance in which there is significant reduction in the effectiveness of any approved Type B, or fissile, packaging during use; (b) Details of any defects with safety significance in Type B, or fissile, packaging after first use, with the means employed to repair the defects and prevent their recurrence; or (c) Instances in which the conditions of approval in the certificate of compliance were not observed in making a shipment

(29) #€ SL^ A?- W (71.101) (Quality assurance requirements) (a) Purpose. This subpart describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging that are important to safety. As used in this subpart, "quality assurance" comprises all those planned and systematic actions necessary to provide adequate confidence that a system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to control of the physical characteristics and quality of the material or component to predetermined requirements. (b) Establishment of program. Each licensee shall establish, maintain, and execute a quality assurance program satisfying each of the applicable criteria of 71.101 through 71.137 and satisfying any specific provisions that are applicable to the licensee's activities including procurement of packaging. The licensee shall apply each of the applicable criteria in a graded approach, i.e., to an extent that is consistent with its importance to safety.

- 190 - (c) Approval of program. Before the use of any package for the shipment of licensed material subject to this subpart, each licensee shall obtain Commission approval of its quality assurance program. Each licensee shall file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied, with the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555 - 0001. (d) Existing package designs. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, and which have been designed in accordance with the provisions of this part in effect at the time of application for package approval. Those packages will be accepted as having been designed in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section. (e) Existing packages. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979; have been at least partially fabricated prior to that date; and for which the fabrication is in accordance with the provisions of this part in effect at the time of application for approval of package design. These packages will be accepted as having been fabricated and assembled in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section. (f) Previously approved programs. A Commission-approved quality assurance program that satisfies the applicable criteria of Appendix B of Part 50 of this chapter, and that is established, maintained, and executed with regard to transport packages, will be accepted as satisfying the requirements of paragraph (b) of this section. Before first use, the licensee shall notify the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555 - 0001, of its intent to apply its previously approved Appendix B program to transportation activities. The licensee shall identify the program by date of submittal to the Commission, Docket Number, and date of Commission approval. (g) Radiography containers. A program for transport container inspection and maintenance limited to radiographic exposure devices, source changers, or packages transporting these devices and meeting the requirements of ?4.31(b) or equivalent Agreement State requirement, is deemed to satisfy

~~- 191 - the requirements of 7U2(b) and 7Ll01(b) of this chapter.~~

(30) ^ S.^ 71^(71.103) (Quality assurance organization) (a) The licensee shall be responsible for the establishment and execution of the quality assurance program. The licensee may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part of the quality assurance program, but shall retain responsibility for the program. The licensee shall clearly establish and delineate, in writing, the authority and duties of persons and organizations performing activities affecting the safety-related functions of structures, systems, and components. These activities include performing the functions associated with attaining quality objectives and the quality assurance functions. (b) The quality assurance functions are — (1) Assuring that an appropriate quality assurance program is established and effectively executed; and (2) Verifying, by procedures such as checking, auditing, and inspection, that activities affecting the safety-related functions have been performed correctly. (c) The persons and organizations performing quality assurance functions must have sufficient authority and organizational freedom to — (1) Identify quality problems; (2) Initiate, recommend, or provide solutions; and (3) Verify implementation of solutions. (d) The persons and organizations performing quality assurance functions shall report to a management level that assures that the required authority and organizational freedom, including sufficient independence from cost and schedule, when opposed to safety considerations, are provided. (e) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms, provided that the persons and organizations assigned the quality assurance functions have the required authority and organizational freedom. (f) Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portion of the quality

~~- 192 - assurance program, at any location where activities subject to this section are being performed, must have direct access to the levels of management necessary to perform this functioa~~

(31) #€ SL^ >m (71.105) (Quality assurance program) (a) The licensee shall establish, at the earliest practicable time consistent with the schedule for accomplishing the activities, a quality assurance program that complies with the requirements of 71.101 through 71.137. The licensee shall document the quality assurance program by written procedures or instructions and shall carry out the program in accordance with those procedures throughout the period during which the packaging is used. The licensee shall identify the material and components to be covered by the quality assurance program, the major organizations participating in the program, and the designated functions of these organizations. (b) The licensee, through its quality assurance program, shall provide control over activities affecting the. quality of the identified materials and components to an extent consistent with their importance to safety, and as necessary to assure conformance to the approved design of each individual package used for the shipment of radioactive material. The licensee shall assure that activities affecting quality are accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. The licensee shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test (c) The licensee shall base the requirements and procedures of its quality assurance program on the following considerations concerning the complexity and proposed use of the package and its components: (1) The impact of malfunction or failure of the item to safety; (2) The design and fabrication complexity or uniqueness of the item; (3) The need for special controls and surveillance over processes and equipment; (4) The degree to which functional compliance can be demonstrated by inspection or test and

- 193 - (5) The quality history and degree of standardization of the item, (d) The licensee shall provide for indoctrination and training of personnel performing activities affecting quality, as necessary to assure that suitable proficiency is achieved and maintained. The licensee shall review the status and adequacy of the quality assurance program at established intervals. Management of other organizations participating in the quality assurance program shall review regularly the status and adequacy of that part of the quality assurance program which they are executing.

(32) -8-71 *|*f *|$ (71.107) (Package design control) (a) The licensee shall establish measures to assure that applicable regulatory requirements and the package design, as specified in the license for those materials and components to which this section applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from standards are controlled. Measures must be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the materials, parts, and components of the packaging. (b) The licensee shall establish measures for the identification and control of design interfaces and for coordination among participating design organizations. These measures must include the establishment of written procedures, among participating design organizations, for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures must provide for verifying or checking the adequacy of design, by methods such as design reviews, alternate or simplified calculational methods, or by a suitable testing program. For the verifying or checking process, the licensee shall designate individuals or groups other than those who were responsible for the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, the licensee shall include suitable qualification testing of a prototype or sample unit under the most adverse design conditions. The licensee shall apply design control measures to items such as the following:

- 194 - (1) Criticality physics, radiation shielding, stress, thermal, hydraulic, and accident analyses; (2) Compatibility of materials; (3) Accessibility for inservice inspection, maintenance, and repair; (4) Features to facilitate decontamination; and (5) Delineation of acceptance criteria for inspections and tests. (c) The licensee shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design. Changes in the conditions specified in the package approval require NRC approval.

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~~-§-71^} PES. € M3H . -8-717>~~

~~#4.~~

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~~$14. ^ $14. *|7ll- -§-71 fc~~

~~. PEfe $14. PE71-~~

~~- 205 - 4.~~

~~# ^^^ #$£ "dunk tesf'e)- ASTM D-543 <>]4. °~~

~~«!-€• Jl^l- 1- *r $14. Dunk ^-^^1 ^^4^ ^4. Dunk~~

~~H t- ^:6>ilfe ^7>^I ^w>^-^ 'g-'JW SlfecO, ASTM D-1693 D-2552<>!4-. D-~~

~~D-25527> Creep X|^4- ^^fc 3H4. A]^^S] -§-7] »1 3|7>^l£ D-25527V~~

~~(It is actual failure of the IIIC, not cracking or crazing which may not penetrate the container under field conditions, that is of concern.)~~

~~-§-7] j ^fecf. PE~~

~~^- o]£-52.-^*13- ^S. PE -g~~

~~a^: ^AfAiofl sis)] PE7> 108 rad ol^o] i] ^4^. ^4. 44^ 108 rad °1^ ^l7ll:^ PES. ^ aL^^^J -g-7lfe~~

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~~- 206 - D-543). jl^-tf 1£S #€• ?l^A]^-ir n?]^:*} €• 50 oC ASTM D-638<>!| 43- *l^sHr 3H4. (4Nl ASTM~~

~~-§-7} ASTM D-25527}~~

~~fe 0.01 inS. T^^^-S Sa4. €^1 at^i^^ -§-7]fe b| & ^A7> 5acf. iL^ *)#& 0.05~0.1 in 711 $4.~~

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~~^714 #^1^1 ^A]^ A] 3^31 €r)-4iL4~~

~~- 207 - 3.~~

~~7\. 10s rad ASTM C394 ASTM D10441 44~~

~~- Cement~~

~~9 JfllS. A>^-^I:K 10~~

~~- VES lCf racial . 7m -§-# A]^ S-^A^S 108 rad °M$] 3b*\*&<%~~

~~- Bitumen 714 SI4. "M-Ji SAH ^l^r Bitumen^ . Bitumen4 1O9 rad ^* €4 hard iEfe soft O}Ê<>II tfiÂ^ IOÎO8^ «^>^^: £4« 3-f, soft hard ^f^^Sfe ^-^^ ^^(36% vol.)~~

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~~|| 108 rad 107 rad 108 rad«H>| 44\+4ai~~

~~- 208 - 4 ^M, Bitumen^ bitumen^)~~

~~(1)~~

~~1H<>)4 ^l^-^-S-^: ^7l#S] ^J^-d] ^A^- jlii) A> 5xlO7 rad ^£ £Afl- %fe4at %q% ^, 55 gal drum -8-^=^1 bitumen^ f^ 10 cm3 ^£^ 71^11~~

~~4 ^^T0! ^«H47fl €4. 7f~~

~~9 A^(IO7 rad " 10 rad) WW ^71-^-fc^, o|~~

~~7\ ^A^4fe ^^: £14^:4. -8-t- ^HJ^l ^n^\^ 108 rad~~

~~(2)~~

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~~- 209 - 3.7),~~

~~. PE 8 3^^ 10 md °MHH 3&AHa M^ -g-7]7>~~

~~f. 108 radfe A-] Al^^~~

~~EPicor-H -§-712} T^^-Â-] ^t. tg-Aî)70%fc «J|^- êfâL ^-tj-. neiq- -§-7l ^.^. hydn 1/2 in T- bulkfe #*}&<% £W Â>^43. ^ ^ S14. ^^-41^: 2mm~~

~~"T- Xicj-. Cs-137014~~

~~(1) PE^^fij~~

~~modulus7> jt^- ^7>^4. ^-«1 °13l £.41-^ 7>5L?1 HDPES~~

~~ASTM~~

~~(2)~~

~~- 210 - 1^A) iA}^ IIDPE oflAi ?1^- A]^ - ductile/brittle~~

~~10 °C 4. 60 °C #4. *g-«*8 «•« ^^5}-^ ^^o] ^r^^o] ^€- ^^ Xi4 u]«HA|~~

~~014. ne}4 ^3^|^ -8-7]~~

~~-S-^, creep modulus) f-^ ductile^ ^€^r ^-^^1^ -B-e]€^-7> #3I€4. °l^f!: ductiUty^~~

~~ASTM 814.~~

~~fraction CL-100 A]^6|] tfl^ gel fraction gel fraction^ «o>AV^ 3:A>ofl~~

~~gel fraction~~

~~(3)~~

~~-g-71 o>^A^ ^-aj^SL o]^.^X|fe ^o|^ -8-7H~~

~~-211 - ASTM 4- 61~~

~~ASTM D- -10 tad (5\1^ 7l?V^£)^ SAV^^I: ^-^ ^-f, ^^ 3.7]*) ^^ 7le}- ^^- 3*HH ^^^l*^ol ^^=^-^1 ^^4fe ^-1: SLIi €A7l- 5U 4. ^^> #£7]- ii^s- brittle/ductile~~

~~.§-71 ^l^->q-fe ^-^^r^-^ tn^ #AKi 5-41- sg7V^-^-^ #^«1 -g-7]~~

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~~- 212 - 4. -g-#(Leach) ^ M ^(immersion)~~

~~ANS 16. J^J ASTM C394- ASTM D10741: 4§4. -¥-7>^^.S ANS 16. (demineralized waterH*]^ Alf ^ 714~~

~~. ANS 16.1 > ^4. 50 psi3 ^^ ^H* - ^ ea«i4.~~

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~~- 213 - . 4~~

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~~- n ^ R, W factors-~~

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~~- 214 - IFR us (tn172 - tn-i^2)^ <83HA| t^^Hr *J4i $•« £H<>># #42 3:4. 4 nan as.~~

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~~M-. Cements -g- Chem-Nuclear Systems ^^Jite Cesiutn(Cs) ^ Strontium(Sr)«~~

~~ANS 16.li |]| fl«fl^ IFR vs~~

~~220L2] ^^11 37H tfl# A]^^4011 tfl*HA^51 Csfe 10%, BftE 9= 2%^£S sa4. cements] ^^-ofl u^sq.^ fy*}% 2.o}3. $a4. ^.^^ cement^ ##i~~

~~4. Bitumen^ -§-# A] ^~~

~~IAEA A|^ ^2)-6il 4§ jpR vs Tc ^jg.^. ^-s^^.^ bitumen^l^si Cs-137, Co-60 5J C-14^ ^^o] V R, 534. 4eM bitumen^~~

~~4. VSE^ -§-~~

~~Cs-137, Co-60~~

~~- 215 - . ANS 16.1~~

~~BNL ANS 16.~~

~~( s||71 Cement^ 1: -t- 5114.~~

~~ANS 16.1 ANS 16.~~

~~- 216 - 5.~~

~~71 Til3~~

~~VESS.~~

~~o]4.~~

~~wet/dry -i: 60C~~

~~wet/dry~~

~~- 217 - 6.~~

~~ASTM #4. o] *9-^ eoc~~

~~Al^Al^t;- 71-g- A] O] *^4\~~

~~-8-71 fe Class B, Class C *8*M§ Î7l#^ ^-7] TP^lîfe -fe ASTM ol -ôi| -§-71^1~~

~~7K Jl^^^J -§-7)*l ^^ ^^ sî^^ -§-7]fe ^s)7l Sfl7]-§r£ ^m AJ-E]]^. f-^ôllA-] ^ ^^ iô] s|7l~~

~~- 218 - -40C -§-71^ ^la^ AB-SM flsfl Phillips Chemical Company^*! brittleness ££.(-118C, ASTM D7846)4 vicat softening ^£(+116C, ASTM -40C " +60C~~

~~creep stress-strain~~

~~4. ^^^- (embrittiement) -118CS. ASTM~~

~~0%7>~~

~~F07> -30C~+40C A>O]O)1 ^4^7 HDPE7]- ^tj-. PE~~

~~notch HDPE a^i^^ -g-71 «H ^-g-1" ^ $l#^lfe ^^^14. °1^1 ^^ PE~~

~~, si, 9X) HDPE^l tfl«|| SA}t «>ol| 4S.^ HDPE3 creeps*€-~~

~~ASTM~~

~~s)J7l!-~~

~~- 219 - TP^Hfe ^3 ^SS «l«fl*l -8-71~~

~~creep -M-^1^O14 stress-strain curve!: M ^r^sVS.^- «|-ji 5^4. 1M ^:^: ^S creep^ stress-strain PE~~

~~91 4 o] -g-^-i: -f-tilAi -g-7]7> ^rfe -%-%o] 300^d -^^> ^^^1-1- ^^1 &o}° creep 7]^^ ^^] 5L#%$ -%-7]sS\~~

~~. -8-71 fc ASTM 0553^1 44 t^ ^S.# 71^14. «l-g-7]4 ^fl^l -§-7l~~

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~~71 Tfl^ 4 -8-713 ^-°J:^ -§-713 41^-, ^1^, ^i«- ^^1^3~~

~~- 220 - PE~~

~~-40~60 X:~~

~~PE g~~

~~PE-8-7H PE pipe^~~

~~-§-7121 2.^^^^- Jg7>S)-feiJl) ^A«|-4. #**,~~

~~PES. € 31^4^ -§-717}~~

~~3.7]$~~

~~-8-7H~~

~~PE-§-7l4 ^*1 #4.~~

~~- 221 - 7.~~

~~ASTM G21 70*1 &:a ASTM G22-76°l G21,~~

~~4. Bitumen^ 4~~

~~, G21,~~

~~50psi~~

~~. U-P~~

~~300M7M~~

~~. B-P~~

~~B_P G21j G22~~

~~bitumen, PE 3.^^-^ -8-7H1~~

~~7\. boric acid, sodium sulfate, beadH ^1, Power^ G21, G22 3.7\7\~~

~~- 222 - G21, 4. cement 3L%^^ G21«14 SL<>1 bitumen^~~

~~G212?- i+Bj-\}tl-. Boric acid^ 3-f^A^ ^ boric acid7|- -i-S^lS ^--g-^-fe ^~~

~~G22 PE -g-7]~~

~~Cement~~

~~A 300^-?}:~~

~~. ^(primary)~~

~~Bife ^- . <£*} ^^ >B Sifc~~

~~(1) ("Potential") (2) *}<& ("Simulated") (3) -£*!] ("Field")~~

~~("inherent") ilfe Xi4. 4^-4~~

~~(1)~~

~~(2)~~

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~~(1) (2) (3) Co-metaboUsm^ (4)~~

~~-g-7]7V ASTM G21«]4 G22# ^-^A] ^^-«1|7> ^-#^4^ B-P~~

~~4-t-^ G214 G22 ft (1)~~

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~~- 224 - B-P Ai^o] ^7H #^4. a-fr ^^-^s^^ ^£4 2^1 3H& 3W~~

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~~- J17] 4.~~

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~~- 225 - 4.~~

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~~(3)~~

~~(4) TP^A-] ia^sj-fc- ASTM G214 G22^r 4-§-4 ^-^^M 544. (a) X|^ a^i^ ready~~

~~(b)~~

~~(5) 2.-^- ^€r«fl£ll' Jg7l-€*7] ^|t(lAi G214 G22« '1^-i-^ co-metabolism4 €^^-§-^-1: fe<^ ^ ^ $a4.~~

~~(1) TP^^i ^^J-Sl-fe G214 G22~~

~~(2) G21 4 G22-~~

~~(3) 4. (4) 714 '9-^-5-^. ^^i -@-fe #$£ ASTM G214 G22~~

~~(5) alsHS ^7l^^.S. £°^1A] ^^-sfl ^3E» SAf^fe B-P~~

~~- 226 - 21)~~

~~(1) <33 mold~~

~~ASTM~~

~~•fr^l 1. SI~~

~~(2) ASTM standard(D618) Conditioning Plastics and Electrical Insulators for Testing~~

~~(3)~~

~~(2) 3*13: -B-7]^S A]^ ^^, (3) %, (4) ^^-cfl « AI^ ^t ^ 7i^, (5)~~

~~2. A]~~

~~(4) (71-)~~

~~^rl-B-(lubricant), ^>^ ^1 (stabilizer) ?J ^-^^1] (colorants) 2 - 38 1C, #tfl<^E 60 - 100~~

~~-1-n *& 3.7]~~

~~(4)~~

~~- 227 - ol4.~~

~~weathering, t~~

~~(5)~~

~~(4) ^S«l! ^7] (incubator)- SL€- 'Stl1^^ ^^ ^-^-^«1fe 28 - 30 TC fe. 85% ^1^*1 <^«> ^4 3, ^~~

~~(6) (7})~~

~~(4)~~

~~(4) 0} Agar ^-^r Nutrient-Salts-Agar~~

~~Potassium dihydrogen orthophospate(KH2PO4) 0.7 g~~

~~Magnesium sulfate (MgSO4 • 7H2O) 0.7 g~~

~~- 228 - Ammonium nitrafce(NIl4NO3) 1.0 g Sodium chloride(NaCl) 0.005 g~~

~~Ferrous suIfote(FeSO4 • W)) 0.002 g~~

~~Zinc sulfate(ZnSO4 • 71W) 0.002 g~~

~~Manganous sulfate(MnSO4 • 7HaO) 0.001 g Agar 15.0 g~~

~~autoclave v pH 7> 6.0 4 6.5 *>°H Sa^^- 0.01 N NaOH~~

~~•£•*!- ^^-^(Mixed Fungus Spore Suspension):~~

~~3. ^^ •§•€ ^°ls.~~

~~Fungi ATCC MYCO No. No.~~

~~Aspergillus niger 9642 386 Penicillium funiculosum 9644 391 Chaetonxium globosum 6205 459 Gliocladium virens 9645 365 Aureobasidium pullulans 9348 279~~

~~ATCC : American Type Culture Collection MYCO : Mycological Services~~

^r^l(potato dextrose agar) 4 ^-^ ^^ nfl^ ${<%/$ %7\<>\ ^*8-°m& -S-B^H «H°^M4. ^^ «fl° ^i^Alfe #£4. 2.^^(spore suspensions) # tefe ^| Xl«H^1, 28-30 r-»1i^ 7-20 QQ ^-^-€ ^- ^^^l(subcultures incubated)-i-

~~- 229 - 4\ A %- *VM(subcultures) <*8SL 10 mL sodium dioctyl sulfosuccinate *)%! H]-^$ $.%*)^K wetting agent) 0.05 g/L~~

~~- 45 mL 3 #5^4 ^^o] 5 mm 91 %•$'£ 10 - 15 7V\ &AA SU M 125 mL 3.7]*] glass-stoppered Erlenmeyer flask •I: ^fe^l-. Fruiting 711~~

~~7] fl«ll, ^SrTfl ^1-S ^^4^- -H-^ funnel~~

~~50 mL~~

~~-g-«q (nutrient-salts solution, -fr £] 4) «.S. s]^«>o] counting chamber sL ^^ A) 1,000,000 ± 200,000 spores/mL 7~~

~~4. ^^^r-^ -g-64^ agar7l- 4^ ^€- ^l^^-SLfe- 6.3^ ^^^1 nutrient salts agar ty~~

~~3 - 10~~

~~(7) $^n (7}) It ^Wfe IliS, a.7]7> 25 mm x 25 mm eJ ^ <^^1 4 ayq-3 444- ^2l^l~~

~~nfl&o]t^ DeVibiss No.~~

~~- 230 - 154 W7l ^£ *^fc ^¥S] £"¥-7)7}- 40HSJ «H ^4. #tfl ££ 85 -S", 28 - 30 TC ^~~

~~(8) (7» 7}-^- 13:^^: A]^^ 50 x 50 mm, 33*1 50 mm ^4i 76 mm~~

~~^r 50 x 25 mm castings]-^ , i-Cglass fabric)€- impregnating(^:^«1~~

~~(Cf)~~

~~^<>I4- ^, 7]- ^ e ^^-(degradation)* 44^fe ^bl-^1 7V^- ^A«- ^«5]4(4.4). 'Data and Control Chart Analysis STP 15D $] &*}' <^1 tfl^ ASTM Manual «1~~

~~(9) ^^r^ «^S] nutrient-salts-agart- 3 - 6 mm tl ^^^: agar ^-i- ^r€-4 Agar 7> ^ f. 110 kPa i^ ^^-ajAS. ^J£^- ^Al3g4 agar AS~~

~~- 231 - l 28 - 30 85%~~

6. Nutrient agarl- 41-MS*

~~Si 0 10% ^^) 1 - 30%) 2 ^^"(30 - 60%) 3 4 7. n (1) (2)~~

~~8. fe*.~~

~~71 5000) «H . o] Al^-i- Method D 618 «H~~

~~(10) Ji~~

~~- 232 - (11)~~

(1) Bagdon, V. J., Military Specification Mil-P-43018(CE), "Plastic Sheets: Polyethylene Terephthalate, Drafting, Coated," June 13, 1961. (2) Baskin, A. D., and Kaplan, A- M., "Mildew Resistance of Vinyl-Coated Fabrics,: Applied Microbiology, APMBA, Vbl 4, No. 6, November 1956. (3) Berk, S., "Effect of Fungus Growth on Plasticized Polyvinyl Chloride Films," ASTM Bulletin, ASTBA, No. 168, September 1950, p. 53(TP 181). (4) Berk, S., Ebert, H., and Teitell, L., "Utilization of Plasticizers and Related Organic Compounds by Fungi," Industrial and Engineering Chemistry, EECHA, VoL 49, No. 7, July 1957, pp. 1115-1124. (5) Brown, A. E., "Problem of Fungal Growth on Synthetic Resins, Plastics, and Plasticizcrs," Modern Plastics, MOPLA, Vol 23, 1946, p.189. (6) Ross, S. H., "Biocides for a Strippable Vinyl Plastic Barrier Material," Report PB-151-119, U. S. Department of Commerce, Office of Technical Services.

-Fungi: ^l^-#, xt, -g^a], a^= include yeastCunicellular organisms), molds (multicellular filamentous organisms such as mildews, rusts, and smut), and flesh fungi(include multicellular mushrooms, puffballs and coral fungi) - Aspergillus niger(*-H*|): mokK^ol, W3\ ATCC 9642) <=fl #«H ^llrvfi °fl -^--J citric acid-i- -*3-£h Amylase, cellulase, glucise oxidase, lipase, pectinase^ £•& £-£• $]2z*\] *]•%•$. Rennin 4 ;££

- Penicillium funiculosum: - Chaetomium globosum: ATCC 6205, Growth temperature 24 V, S-^Ej ^"^^1-^, cellulose ^-c: glucanohydrasc ^^ - Gliocladium virens: ATCC 9645, Growth temperature 24*0, £#£. - Aureobasidium puUulans: ATCC 9348, Growth temperature 24 t;, pullulan ^£ fructosyl transferase

~~- 233 - tYlb4o^ fe-tfrla lv fife ' •fc?~~

~~jo 3OUB5S1S9H Surarau^aQ aoj sorjOBOtj JZ O pure sot)SBU Bunioppuco 8T9 G sjxrepuE>s JMXSV (4 (Z)~~

~~^ is -~~

~~9 Mn ibH^Pr V ^^-tn #4^ M^r to a~~

~~(T) (4) ^^(transmission)~~

~~(4) 7}~~

~~, weathering,~~

(5) 7 4. 75 mm *W3 Al^ofl tfl«fl 150 mm^ ^%$X^ Petri dish 75 mm o]7l4 rL o^ 3.7l?l ^-f, ^-^sj- 3£71- Site Petri dish, borosUicate glass $&, A>Z}-^ %•$% ^^ 4 S<^^1 1-^-fe 3A} (4) *fl-3"»H ^71 (incubator)- 5L€- ^^%^#<=fl ifl^: ^^-^Blfe 28 - 30 1C s] fe 85% o]4>ol<^o> ^-4. £t $ ^<% t\)

~~(6) A]^ ^~~

~~- 235 - nutrient-salts-agarfe L ^~~

~~Potassium dihydrogen orthophospate(KIl2P04) 0.7 g Potassium monohydrogen orthophospate(K2HPO4) 0.7 g~~

~~Magnesium sulfate (MgSO4 • 7H2O) 0.7 g Ammonium nitrate(NH4NO3) 1-0 g Sodium chloride(NaCl) 0.005 g~~

~~Ferrous sulfate(FeSO4 • 7H2O) 0.002 g~~

~~Zinc sulfate(ZnSO4 • 7H2O) 0.002 g~~

~~Manganous sulfate(MnSO4 • 7H2O) 0.001 g Agar 15.0 g Distilled water 1000.0 mL~~

~~autoclave 121~~

~~3. Nutrient-salts-agarfc (1) 0.15 % ^ cycloheximide* nutrient-salts-agar °1] (2) ethylene oxide i ^#^1^)^.~~

~~(e}-) ^^-^(bacterial cell suspension): bacteria Pseudomonas aeruginosa, ATCC 13388, MYCO B 1468. *%<$$- nutrient agar^~~

~~- 236 - Nutrient agar slants* ^}2:^-7) < 1 0.5% peptone 3J 1.5% agar* 7r I 3:4 ^rlraL, ^^ ^ 121 TC % 103 fcPa 15g- •I:~~

- nutrient broth(agar, nutrient-salts-agar, nutrient brothfe ^#^ <&%$&. *^*|S^.^ ^'S 7]-^BV^ brothfe *€- soup^l ^^fl> tfl^l^ <£^r3£.S 2^*1^ olÂlôH ^)«g <^^ ^ ^cj-. Nutrient brothfe 03 %*H &3L7] ^%%A 05 % ^ peptone-i: ^^-^r^ -S-^ l^^lM- l-e^fl4| ^^«1 ^r«l|^:^. ^-^r ^ 103 121 13 5] £.£# -f}-^^4. ir^^-S- #^^1^1 w>^€- nutrient agar slants-?-^ nutrient brith^S oj-S-^îtq-. 24^1 # . o] broth «fl *<*«•&- #^€ nutrient broth . Brothiî «fl*^ -i:^^: %-§&?\ ^1^14. Broth ^-lr( 0.8 % NaCl)

~~a)^^(photoelectric colorimeter)~~

~~(counting)^-~~

~~(4) Petri ] nutrient agar %7\}(loop-full) . 48 -72 A]^)-, 85% , 35 -37~~

~~$14.~~

~~(7)~~

~~(7r) 50 x 50 mm, ^^*1 50 mm ^41 75 mm ^^13~~

~~- 237 - fe =17}7}- ^4:^ 50 x 50 mm~~

~~(Cf)~~

~~-^ 4.~~

~~^- €• 1 ^-(degradation)* M-Ef^fe ^i-o] 7]-^-~~

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- 252 - L "Containers for Packing of Solid and Intermediate Level Radioactive Wastes", Technical Reports Series No. 355, Vienna, IAEA (1993) 2. "Evaluation of Hydrogen Diffusion from Low-Level Radioactive Waste Container", EPRI NP-6244, Palo Alto, EPRI (1989) 3. B. Siskini, "Gas Generation from Low-Level Radioactive Waste: Concerns for Disposal", BNL-NUREG-47144 Q992) 4 R. Morrow et al., "Disposal/Storage Container Development Experience", CONF-881054-i6 (1988) 5. R. Chapman and R. Haelsig, "Filter Assembly Developed for a High Integrity Container Off-Gas Vent System", EGG-PE-6759 (1984) 6. J. McConnell, et al., "Disposal Demonstration of a High Integrity Container Containing an EPICOR-H Prefilter from Three Mile Island", GEND-45 (1985)

~~- 253 - *l *1 * it * 4~~

~~INIS ^MlSJE.~~

~~KAERl/RR-1945/98~~

~~# ?t *1 ^r*37l^: KAERI 1999~~

~~253p. a 7l A4 •98 «^* ^oiw.(«*ra*™*Am~~

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~~t T^^ll ^ 7^1 TQ ~" Js»s.7l«^*7U ******* *^^*aaiM, 7i BIBLIOGRAPHIC INFORMATION SHEET Performing Org. Sponsoring Org. Stamdard Report No. INIS Subject Code Report No. Report No. KAERI/RR-1945/98~~

~~Title/Subdtle Development of Polymer Concrete Radioactive Waste Management Containers~~

Project Manager H. Chung (KAERI) and Department Researcher M.S. Lee, DJH. Aha, RJ. Won, H.S. Kang, H.S. Lee, SP. Lim, Y.E. Kim, B.O. LEE (KAERD and KJP. LEE, B.Y. Min, J.K. Lee, W.S. Jang, W.B. Sim (Kye Lim, Ltd.) Department J.C. Lee, MJ. Park, YJ. Choi, HJE. Shin, H.Y. Park, C.Y. Kim (Myong Ji University) Publication Publication Seoul Publisher MOST 1999 Place Date Page 253 p. 111. & Tab. Yes(V), No ( ) Size A4

Note Open(V), Restricted( Classified Report Type Research Report Class Document Performing ' Contract No. Organization Abstract (15-20 Lines) A high-integrity radioactive waste container has been developed to immobilize the spent resin wastes from nuclear power plants, protect possible future, inadvertent intruders from damaging radiation. The polymer concrete container is designed to ensure safe and reliable disposal of the radioactive waste for a minimum period of 300 years. A built-in vent system for each container will permit the release of gas. An experimental evaluation of the mechanical, chemical, and biological tests of the container was carried out The tests showed that the polymer concrete container is adequate for safe disposal of the radioactive wastes.

~~Subject Keywords radioactive waste, high-integrity container, ceramic filler, polymer -concrete, (About 10 words) mechanical properties, gamma radiation, particle size distribution, hydrothermal~~