ML19211D099
| ML19211D099 | |
| Person / Time | |
|---|---|
| Site: | 07105753 |
| Issue date: | 11/14/1979 |
| From: | Travis W ENERGY, DEPT. OF |
| To: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 14762, NUDOCS 8001160355 | |
| Download: ML19211D099 (7) | |
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Department of Energy Oak Ridge Operations o
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P.O. Box E g
Oak Ridge, Tennessee 37830 3
EECE/ygg ha November 14, 1979 E-Nova l
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U. S. Nuclear Regulatory Comission y
ATTii: Charles E. MacDonald, Chief 4
U Transportation Certification Branch Division of Fuel Cycle & Material Safety Washington, D. C.
20553 Gentlemen:
SAFETY ANALYSIS REPORT FOR PACKAGING FOR THE ORNL LOOP TRANSPORT CASK - REPORT ORNL/ENG/TM-11.
Reference is given to the subject SARP and to your Docket No. 71-5753.
In reply to the specific additional information which you requested, ORNL prepared the enclosed supplemental data. The data have been reviewed by the 00E-0R0 Safety and Environmental Control Division and we fully concur with the conclusion.
We request your early review.
Sincerely, Q:--.
YO William H. Travis, Director MS-334:WAP Safety & Environmental Control Division
Enclosure:
Supplemental Data (9 copies) cc w/ encl.:
T. H. Hardin, AD-46
- 0. M. Ross, G-135, HQ-G 1755 252 cc w/o encl.:
C. A. Keller, MS-30 d.s fLd g%
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Structural - We were not aware of any regulatory requirement for evaluation of packages at low service temperatures with respect to the free fall accidents. However, the cask will comply with the requirements of the regulations if subjected to the free fall accidents at -20*F.
It was con-cluded that rupture of the outer shell was credible (see pages 99 and 132) as a result of the hypothetical accident at normal temperatures.
It was postulated that the lead which melted would be lost and the resulting loss of shielding evaluated (see section 4.2).
A catastrophic, cast iron type failure of the shell would not occur with materials in thicknats range
(.68" max thickness) employed in the loop cask at -20*F.
This is because the cooling rate, after hot rolling, from the austenitizing temperature would be rapid enough to assure a microstructure which will exhibit adequate toughness at temperatures as low as -20*F.
This fact is recognized by the ASME Code,Section III which does not require toughness testing for 5/8" and thinner materials (see paragraph NB 2311 al). Also the probability of an impact accident followed by a fire at -20*F is very low.
Thennal - The damage from the impact (30 foot free fall and puncture) would be local in nature hence would not significantly affect the response of the cask to the thermal component of accident. The worst case "ioss of shielding" was considered since the calculated loss from a side impact was greater than the combination corner impact and loss due to melting in the corner.
It is noted that calculations show that met [ing is confined to the corner.
Containment - It should be recognized that the loop transpor't cask is a general purpose package which is used to transport a variety of material (see Secticn 0.2.3).
Hence containment vessels are designed or existing vesseis evaluated when be need arises. As stated in the SARP the primary containment vessels 14762 1755 253
are designed in accordance with Section V!!! of the ASME Code and 00T specification 2R (see Section 1.8). The inspections and tests required by the code are therefore a normal part of the fabrication. Routine inspections of the containment vessel are outlined on the ORNL Radioactive Materials Packaging Information fann (see page 211 of the SARP) under the heading
" Internal Container". Two typical containers were tested to demonstrate com-pliance with the impact tests (see section 3.3 and Appendix D). Calculations
~
in the SARP demonstrate compliance with the other requirements of 10CFR, part 71.
Criticality - We have considered the effect of lead as a reflector on the con-tents of the subject cask. We found that for the proposed loadings of the cask there is no significant increase in the k-eff of the cask due to the presence of lead as a shielding material.
l The analysis used the KEN 0 IV Monte Carlo Criticality Code with the Hansen-Roach cross-section sets.2 This ccmbination of code and cross sections meets the requirements of the American National Standard ANSI N16.11 on Validation of Calculational Methods for Nuclear Criticality Safety.
3 The cask was modeled as described in the SARP.
It is known that lead is a more effective reflector for systems having a high median energy of fission, thus rather than consider fissile materials " optimally moderated " the effect to be examined is emphasized if the fissile material is modeled as metal spaced in water to produce a k-eff larger than one may credibly expect. The metal was described as a rod having an 0.4-cm-diameter and a length of 1.7 m, resulting 233 239 in a mass of 396 g U(93.2), or 388 g U or 410 g Pu. The cask contained a 1cading of nine rods on 2.9 cm centers in a square matrix with and without water as a moderator. The results of the calculations appear in the following tabl e.
The cask was considered submerged.
1755 254
Calculated k-eff for the ORNL Loop Transfer Cask for Various Fissile Materials Cask Loading Water in Material Mass, kg Cavity k-eff ta U(93.2) 3.6 Yes e.463 0.037 No 0 083 0.002 2330 3.5 Yes
- 0. 148 0.006 No 0.128 0.003 239Pu 3.7 Yes 0.484 0.006 No 0.131 0.003 Three additional calculations of the cask were performed with the U(93.2) loading.
In one, the external reflector was removed and an = x
=x = array of packages was calculated. The resultant k,was 0.595 1 0.010. The second was the same case but with the lead replaced by water giving a k, of 0.474 !
0.006. The third was the same as the first rntry of the table but having water in place of the lead in a single package.
This gave a k-eff of 0.457 t 0.005.
In summary, it is concluded that the lead is not a significant source of addi-tional reactivity to the loaded cask and that the cask does satisfy the require-ments of Fissile Class I package for its intended use.
REFERENCES 1.
L. M. Petrie and N. F. Cross, KEN 0 IV - An Imoroved Monte Carlo Criticality Program, ORNL-4938 (November 1975).
2.
G. E. Hansen and W. R. Roach, Six and Sixteen Group Cross Sections for Fast and Intermediate Critical Assemblies, LAMS-2542 (1960).
3.
H. C. Paxton, J. T. Thomas, D. Callihan, and E. B. Johnson, Critical 235 239 233 Dimensions of Systems Containino U
, ?u
, and U
, TID 7028 (June 1964).
14762 1755 255
3 Structural - We were not aware of any regulatory requirement for evaluatic' of packages at low service temperatures with respect to the free fall accidents. However, the cask will comply with the requirements of the regulations if subjected to the free fall accidents at -20*F.
It was con-cluded that rupture of the outer shell was credible (see pages 99 and 132) as a result of the hypothetical accident at normal temperatures.
It was postulated that the lead which melted would be lost and the resulting loss of shielding evaluated (see section 4.2).
A catastrophic, cast iron type failure of the shell would not occur with materials in thickness range
(.68" max thickness) employed in the loop cask at -20*F.
This is because the cooling rate, after hot rolling, from the austenitizing temperature would be rapid enough to assure a microstructure which will exhibit adequate toughness at temperatures as low as -20*F.
This fact is recognized by the ASME Code,Section III which does not require toughness testing for 5/8" and thinner materials (see paragraph NB 2311 al). Also the probability of an impact accident followed by a fire at -20*F is very low.
Thermal - The damage from the impact (30 foot free fall and puncture) would be local in nature hence would not significantly affect the response of the cask to the thermal component of accident. The worst case " loss of shielding" was considered since the calculated loss from a side impact was greater than the combination corner impact and loss due to melting in the corner.
It is noted that calculations show that meAf ng is confined to the corner.
Containment - It should be recognized that the loop transport cask is a general purpose package which is used to transport a variety of material (see Section 0.2.3).
Mcnce containment vessels are designed or existing vessels evaluated when the need arises. As stated in the SARP the primary containment vessels 1755 256 14762
are designed in accordance with Section VIII of the ASME Code and DOT specification 2R (see Section 1.8).
The inspections and tests required by the code are therefore a normal part of the fabrication. Routine inspections of the containment vessel are outlined on the ORNL Radioactive Materials Packaging Information form (see page 211 of the SARP) under the heading
" Internal Container". Two typical containers were tested to demonstrate com-pliance with the impact tests (see section 3.3 and Appendix 0). Calculations in the SARP demonstrate compliance with the other requirements of 10CFR, part 71.
Criticality - We have considered the effect of lead as a reflector on the con-tents of the subject cask. We found that for the proposed loadings of the cask there is no significant increase in the k-eff of the cask due to the presence of lead as a shielding material.
l The analysis used the KEN 0 IV Monte Carlo Criticality Code with the Hansen-Roach cross-section sets.2 This combination of code and cross sections meets the requirements of the American National Standard ANSI N16.ll on Validation of Calculational Methods for Nuclear Criticality Safety.
The cask was modeled as described in the SARP.
It is known that lead is a more 'ffective reflector for systems having a high median energy of fission, thus rather than consider fissile materials " optimally moderated," the effect to be examined is emphasized if the fissile material is modeled as metal spaced in water to produce a k-eff larger than one may credibly expect. The metal was described as a rod having an 0.4-cm-diameter and a length of 1.7 m, resulting 233 239 in a mass of 3961 U(93.2), or 388 g U or 410 g Pu. The cask contained a loading of nine rods on 2.9 cm centers in a square matrix with and without water as a moderator. The results of the calculations appear in the following tabl e.
The cask was considered submerged.
1755 257
Calculated k-eff for the ORNL Loco Transfer Cask for Various Fissile Materials Cask Loading Water in Material Mass, kg Cavity k-eff ta U(93.2) 3.6 Yes 0.463 0.007 No 0.083 0.002 2330 3.5 Yes 0.548 0.006 No 0.128 0.003 239Pu 3.7 Yes 0.484 0.006 No 0.131 0.003 Three additional calculations of the cask were performed with the U(93.2) loading.
In one, the external reflector was removed and an = x = x = array of packages was calculated. The resultant k,was 0.595 1 0.010. The second was the same case but with the lead replaced by water giving a k, of 0.4741 0.006. The third was the same as the first entry of the table but having water in place of the lead in a single package. This gave a k-eff of 0.457 t 0.005.
In summary, it is concluded that the lead is not a significant source of addi-tional reactivity to the loaded cask and that the cask does satisfy the require-ments of Fissile Class I package for its intended use.
REFERENCES 1.
L. M. Petrie and N. F. Cross, KEMO IV - An Imoroved Monte Carlo Criticality Program, ORNL-4938 (November 1975).
2.
G. E. Hansen and W. R. Roach, Six and Sixteen Group Cross Sections for Fast and Intermediate Critical Assemblies, LAMS-2542 (1960).
3.
H. C. Paxton, J. T. Thomas, D. Callihan, and E. B. Johnson, Critical 235 239 233 Dimensions of Systems Containino U
, Pu
, and U
, TID 7028 (June 1964).
1755 258 14762