ML19281B144
| ML19281B144 | |
| Person / Time | |
|---|---|
| Site: | 07106345 |
| Issue date: | 04/02/1979 |
| From: | Patterson D ENERGY, DEPT. OF |
| To: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML19281B145 | List: |
| References | |
| NUDOCS 7904260062 | |
| Download: ML19281B144 (5) | |
Text
epn 1
as
~
%9
.6D\\\\ y /,
I Department of Energy GF'
.,,i Washington, D.C. 20545
,g.g April 2, 1979 ussRC 3
~
D' APR S U M2 " l3
,i tivfis j '-
g MAR s'cc0CM
/[-
Mr. Charles E. MacDonald v coo 21 C L Q?
U.S. !!uclear Regulatory Commission y
!!ashington, D.C.
20555 m
Dear ftr. MacDonald:
'In accordance with our past agreement, we are forwarding for your review the following for packaging Certificate #6345:
1.
8 copies of the SARP for packaging.
THIS DOCUMENT CONTAINS 2.
8 copies of the field office evaluation.
POOR QUAUTY PAGES 3.
8 copies of the Certificate of Compliance i
Remarks:
These TREAT 100P racke t.,ill ne4 he used for Prototype Fast Reactor (PFR) fuel pins.
There may ba international shipments to United Kingdom.
We expect that you will issue an fiRC certific M o for licensee use.
This packaging review should be completed as early as possible.
l!e will appreciate being advised of your planned schedule for this review.
Sincerely, N,
[
$N Ik$h5 DavidE.fPattersoi f n b"
3 M.,
6.f d,u d fi
//
j<
Acting Branch ChieV Occupational Safety Branch L
Operational and Environmental Safety Division
Enclosures:
bcc:
R. R. Rawl, DOT (1 each)
As stated E. Loop, OESD (1 each)
R. I. Elder, CH Safety Divi _sion - ltr. only 7904260O G
{
us w-f.fg DOCKETED h USNRC Qg DOE Evaluation Ap20 91979 9C Safety Analysis Report for Packaging on g
/ _ jl the Shipment of Irradiated Fissile Materials in the
%x, oMAg gN//
ANL-402-SPM and ANL-403-SPM Shipping Casks
\\
Revision A
,s l g )y w
I.
Introduction A revision to the original Safety Analysis Report for Packaging (SARP) has been prepared for llanford Engineering Development Laboratories (llEDL) to cover shipment of a special pressure vessel containing Prototype Fast Reactor (PFR) fuel pins in the ANL-402 and ANL-403 shipping containers.
These contents are shipped as part of the US/UK fast reactor fuel research program.
The revision was prepared by Battelle Memoria! Institute - Columbus Laboratories (BCL), who also prepared the original SARP.
The work was donc under contract to HEDL.
The SARP revision, coordinated by M.
E.
Balmert, demonstrates shipment of the PFR fuel special containments in the RAS casks complies with the requirements of DOE Manuals 0529 and 5201 and provides information required by 10 CFR 71.
The format for the revision conforms to the U. S.
NRC Rehulatory Guide for SARP preparation.
II.
Reviews An independent review of the revision to the SARP uns made by HEDL.
This revicu determined that appropriate channes were made to the SARP and shipment of the added contents complies with DOE requir(,ents for shipping containers.
The formal DOE review of the SARP revision was performed by the Chicago Operations Office (Cil).
The CH evaluation was made by R.
I.
Elder.
Mr. Elder has a BA degree in Physics and an MS degree in Nuclear Engineering.
III.
Evaluation Results The conclusions reached from the Cll review of the SARP revision are summarized below.
This review covered only the portions of the original SARP affected by the shipment of the special contents described above.
Chapter 1.0 - Structural Evaluation The basic structure of the cask has not been altered.
This revision covers only information on an additional set of contents and its containment.
As a result, the conclusions ebout the structural integrity of the RAS casks are not changed.
The casks can continue to meet the regulatory requirements.
Chapter 2.0 - Thermal Evaluation In the revision to the SARP,the thermal t,
cacteristics of the cask loaded with 3 PFR fuel pins in their containment structure were evaluated.
The TRUMP conputer code was again used to predict the steady state temperature conditions for the loaded cask, as well as the consequences from the hypothetical accident fire.
. The temperature of the pressure vessel holding the fuel pins was calculated to be 235 F.
The inner capsule walls would reach 341 F.
These are for the normal conditions of transportation.
The corresponding pressure in the inner contain-ment is 11.1 psig.
In the hypothetical accident case, the pressure vessel reaches 405 F, 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the fire starts and the inner capsule reaches 479 F, 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the fire begins.
The viton o-rings used in the pre wure vessel holding the fuel pins would tolerate this temperature.
The welds of the inner capsule would be unaffected by the peak temperatures mentioned above.
No release of contents would result from this containment structure.
The input data used for these code calculations was determined to be uutisfactory.
Chapter 3.0 - Containment The revision to the SARP provides an analysis of the containment structure that will be used to hold the PFR fuel pins.
The structure used gives two levels of containment, a welded stainlecs steel capsule and an outer carbon steel pressure vessel with scaled closures.
The first part of the analysis evaluated the ability of the containment to withstand the normal conditienc cf the transportation.
Stresses due to internal pressure, thermal expansion, thereal gradients, vibration and a free drop were calculated.
The structure is shown to with-stand these stre:aes with no yielding of the material and consequently no release of the contents.
Straight-forward engineering equations were used in making the analysis.
Conservative assumptions were made in developing the analysis.
The containment structure also was evaluated to the conditions imposed by the hypothetical accident.
The ability of the containment to withstand the 30 foot drop, while within the cask, was demonstrated.
The impact energy absorbed in the drop would cause some yielding of the containment structures but there would be no rupture of them, or dispersal of the radioactive contents.
A thorough set of calculations was made to support this determination.
The equations used were verified as vaJid.
Chapter 4.0 - Shielding Evaluation j
The SARP revision contained an added shiciding evaluation to assess irradiated PFR fuel pins as cask contents.
The radiation source which the fuel pins comprise was determined by using the ORIGEN computer code.
The output of the code calculation provided the information needed to characterize the source strength.
The shielding capability of the cask, with the PFR fuel as contents, was evaluated with the QAD code.
The source strength for this calculation came from the OR1 GEN results.
The QAD code results are in terms of the radiation Icvels at the cask exterior.
The radiation levels from the PFR fuel are shoun to be well within the DOT chipping regulations.
. The input data used for these calculations was reviewed.
It was verified the data used was appropriate and a suitable modeling of the cash was developed.
Chapter 5.0 - Criticality Evaluation The fissile contents of the PFR fuel in the cask is quite small.
The safety of these contents was determined by reference to nuclear safety guides which shows the quantities are for below masses needed to reach criticality.
It is concluded criticality safety is assured and an adequate analysis was provided.
Several corrections, to be included in the final printing of the SARp, have been agreed to by BCL.
These changes are'11sted below.
1.
P. 0-10 A current version of the Certificate of Compliance will be included.
2.
P.
1-6 and P.
1-7 Data on static and dynamic properties of 6061-T6 aluminum will be included in Tables 1.3 and 1.4.
3.
P.
2-5, Section 2.2 - Thermal proper. ties of materials BCL will add tables which will provide data on the thermophysical properties of materials used in the containment structures used for tb 1)FR fnal pinc.
4.
P.
2-16 The following will be added to the end of the first paragraph.
"The consequences of the free drop and puncture accidents are not significant enough to require altering the analytical model used for the TRUMP calcula-tions, either for the TREAT Loop contents or the PFR fuel pins as contents."
5.
P.
2-22, Section 2.5.6 Line 8 is revised to read:
. was conservatively assumed to be 750 F.
The seal area.
6.
P.
3-62 The title for Table 3.11 will be corrected to read:
Comparison of Two Foot Free Drop Stresses with Vibration Stresses The heading line will be:
2 Ft. Drop G
Stress 7.
P.
3-66 The last line will be revised to read:
., to an impact in any orientation, has been calculated previously to be 19,200 psi.
There will.
8.
P.
3-82 The sentence in mid-page will be revised to read:
"This is below even the static yield strength of 6016-T6, which is psi (see Table 1. 3). "
9.
P.
3-10 The equation at mid-page is corrected to be:
~3 2
5 4
3 2
2 2X + 4X LX +LX W
X
~ 4L 9
6 36 Z
4EI
_S L
.