ML20009H319
| ML20009H319 | |
| Person / Time | |
|---|---|
| Site: | 07109132 |
| Issue date: | 06/05/1981 |
| From: | NUCLEAR PACKAGING, INC. |
| To: | |
| Shared Package | |
| ML20009H318 | List: |
| References | |
| NUDOCS 8108070248 | |
| Download: ML20009H319 (48) | |
Text
_
1 i
1 uP c 4
i i
SAFETY ANALYSIS REPORT FOR THE FFTF FUEL PIN SHIPPING CASK i
MODEL - T3 I
I d
E- /ISION 6
June 5, 1981 1,
i NUCLEAR AlnPb'c PACKAGING,INC.
8108070248 810616 PDR ADOCK 07109132 815 SO. 28TH STREET. TACOMA, WASHINGTON 98409. (206)S72 7775 838-1243
t i
e.
e.
INSTRUCTIONS FOR INCORPORATION REVISION 6 AMENDMENTS TO MODEL T-3 SHIPPING CASK SAFETY ANALYSIS REPORT (CERTIFICATE OF COMPLIANCE 9132)
DATED JUNE 5, 1981 Insert new pages i and vi in Removo old pages i and iv Table of Contents from Table of Contents Insert new page 0-6 Remove old page 0-6 Insert additional page 0-6a Insert new page 0-7 Remove old page 0-7 Insert additional page 0-7a,
Insert a'dditional page'0-7b Remove old page 4-7
(
Insert new page 0-8 Remove'old~page 0-8 l
Insert 1ew page 0-9 Remove old page 0-9 Insert new page 1-6b Remove old page 1-6b l
Insert additional page 1-6b (1)
Insert additional page 1-6b(2)
Insert additional page 1-6b(3)
Insert additional page 1-6b(4) l Insert new page 1-23a Remove old page 1-23a Insert additional page 1-23b Insert new page 4-1 Remove old page 4-1 Insert new page 4-2 Remove old page 4-2 Insert new page 4-4 Remove old.page 4-4
. Insert new page 4-5 Remove old page 4-5 4-6 4-6 4-8 4-8 f
4-9 4-9 4-10 4-10 4-11 Remove old page 4-11 Insert additional page 4-lla l
Insert new page 4-12 Remove old page 4-12
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4-13 Remove old page 4-13 4-14 4-14 5-1 5-1 5-2 Remove old page 5-2 Insert additional page 5-2a Insert new page 5-6 Remove old page 5-6 (Continued)
1 INSTRUCTIONS FOR INCORPORATION REVISION 6 (Continued)
Insert additional page 5-6a-Insert new page 5-10 Remove old page 5-10 Insert new page 5-12 Remc
- old page 5-12 Insert new page 5-13 Remove old page 5-13 Insert new Drawing Ha-61289, Remove old Drawing H4-61289 Sheets 1, 2,
and 3 following following page 1-121 page 1-121 Add new Drawing H4-61289, Sheet 4 l
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Revieion 6
.Tunn 5, 1981 PROPOSED SAR TABLE OF CONTENTS FOR WESTINGHOUSE T-3 CASK Page 0.O GENERAL INFORMATION 0.1 Introduction 0-1 0.2, Package Description 0-2 0.2.1 Packaging 0-2 0.'2.1.1 General Description 0-2 0.2.1.2 Materials Qf Construction, Dimensions,
& Fabricating Methods 0-2
- 0. 2.1. 3 Containrnent Vessel 0-4 i
O.2.1.4 Neutron Shielding and Absorbing Materials 0-4 l
0.2.1.5 Package Weight 0-4' O. 2.1. 6 Receptacles 0-5 0.2.1.7 Drain Port 0-5 l
0.2.1.8
.Tiedowns 0-5 t
O.2.1.9 Lifting Devices 0-5 0.2.1.10 Pressure Relief System 0-5 l
0.2.1.11 Heat Dissipation 0-5 l
0.2.1.12 Coolants 0-6 l
0.2.1.13 Protrusions 0-6 0.2.1.14 Shielding 0-6 0.2.1.15 Liner Holddown 0-6 6
0.2.2 Operational Features 0-6a 0.2.3 Contents of Packaging 0-7 0.2.3.1 Radioactivity 0-8 J
e Revicion 6 June 5, 1981 TABLE OF CONTENTS (Cont.)
Pace 3.3 Containment Requirements for the Hypothetical Accident Conditions 3-11 r
3.3.1 Fission Gas Products 3-11 3.3.2 Release of Contents 3-11 4.0 SKIELDING EVALUATION 4-1 4.1 Discussion and Results A"1 4.2 Source Specification 4-4
- 4. 2'.1 Ganma Source 4-4 4.2.2 Neutron Source 4-4 4.3 Model Specification 4-5 4.3.1 Description of Radial and Axial Shielding Configuration 4-5 4.3.2 Shield Region Densities 4-14 4.4 Shielding Evaluation 4-15 4.5 Appendices 4-16 4.5.1 References 4-16 4
4.5.2 Basic Algorithm for QAC 4-17 1
4.5.3 Finite Source Length Effect 4-19 5.0 CRI77CALITY EVALUATION 5-1 1
5.1 Discussion and Results 5-1 L.2 Package Fuel Loading 52 5.3 Model Specification 5 2a 5.3.1 Description of Calculational Model 5-2a 6
5.3.2 Package Regional Densities 5-6a 5.4 Criticality Evaluation 5-11 5.4.1 Calculational Method 5-11 5.4.2 Fuel Loading Optimization 5-11 vi
f RLvicion 6 June 5, 1981 e'
n-600 watts.
However, this value may be exceeded if it can be l
l demonstrated that actual equilibrium temperatures with the higher heat load are still within allowa'ble limits.
Refer to Section 2.0 for the thermal evaluation of the T-3' cask.
l 0.2.1.12. Coolants There are no coolants involved except the normal transportation atmospnere wnich may include air, argon or helium.
0.2.1.13 Protrusions There are no outer or inner protrusions except the external
- nnions as shown on the drawing, and these are located within
+
the envelope protected by the overpack.
0.2.1.14 Shielding The design of the T-3 cask insures sufficient shielding for neutron and-gamna radiation to comply with the limits on dose rates external to the cask specified in 10 CFR 71.
0.2.1.15 Liner Holddown A modified verison of the T-3 is available with a holddown ring attached to the inside of the containment wall.
The holddown restrains payload liners while the payload is being extracted.
6:
The holddown is held in place by small screws tapped into the i
~
containment wall inner surface.
A modified shield plug is used to provide clearance for the rings (See Drawing H-4-61289).
0-6 l
... - - - -.. o
R3vicion 6 Juno 5, 1981 0.2.2 Operational Features The T-3 cask system is fundamentally a shielded cylindrical container with associated contained contents of high integrity which is capable of complying with NRC and DOT regulations governing the shipments of large quantities of radioactive Its operation in transport is simple and fissile ' material.
passive, requiring no auxiliary cooling or special systems.
I 1
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0-6a i
-g----
-+
y,+.-
- p w- - -r
.m e-
,9 r-,
-yg- - -,
,,,,,.,y m--g,..,,,,9 c.-_
g 9 g
y
,y
l.
R3vicion 6
~-
Juns 5, 1981 t
l j
0.2.3-Contents of Packaging l
l The cask contents will consist of reactor fuel material in-6 solid form, comprised of uranium or plutonium oxides, car-bides or nitrides or mixtures thereof, including fissile l
material and mixed fission products.
The fuel material will be contained'in cladding of austenitic stainless steel, ferritic stainless steel or other high strength alloys as fuel pins or fuel pin sections placed end to end in tubes.
The contents will also include non-fuel bearing reactor core components and test assemblies which may cuntain neg-ligible quantities of fissile material in the form of dosi-metry foils.
Payloads may include absorber material as whole control rods or individual pins or complete test assem-6 blies and specimens.
This cask as prepared for shipment may contain fissile fuel and large quantities of radioactive material.
Fissile fuel may consist of uranium-233, uranium-6 235 and/or plutonium-239 or mixtures of these isotopes.
Restrictions l
l Shipments in this cask shall not exceed the following limits:
1.
The decay heat of contents shall not exceed 600 watts 6
total or 17.5 watts / inch.
2.
The fissile material geometry and quantities shall con-form to the following:
a.
One FTR or experimental assembly with a maximum l
k-effective of 0.85.
The mass density of fissile f
material shall be limited to 110 grams per linear 6
centimeter of uranium-233 plus plutonium-239 plus one-third of the uranium-235 density.
The total 0-7 I
Revision 6 June 5, 1981 mass of fissile material shall not exceed 10 kilo-
- grams, b.
Fuel pins or fuel pin sections encapsulated in tubes, not configured as fuel assemblies, shsll conform to the geometric criteria illustrated in Figure 0.2.3-1.
6 The fissile material mass density shall be limited to 60 grams per linear centimeter for uranium-233 plus plutonium-239 plus one-half of the uranium-235 density.
The total mass of fissile material shall not exceed l
10 kilograms, with the exception of payloads containing uranium-233, which shall not exceed 5 kilograms fis-sile material.
3.
Radiation levels from the cask shall not exceed the Depart-ment of Transportation limits.
l l
4.
Payloads shall not exceed 230 kilograms total nass.
The linits described above envelope the actual payloads which will be transported in the T-3 Cask.
The following i
table presents a representative sample of cask payloads.
Mith the exception of the 217-pin FTP driver assembly, all i
the listed payloads contain well below the 10 kilogram limit of fissile material.
)
0-7a
Ravision 6 June 5, 1981 Fuel OD 10.195 in.
Cladding thickness.*0.010 in.
Stainless steel outer spacer tubing.
1
% (
Thickness 20.012 in.
Stainless steel inner tubing reovired only for encapsulation i
of cut pin sections.
Thickness E0.020 in.
0 ID Shall be 50.025 in. less than two tirnes the fuel pin cladding OD.
i l
l FIGURE 0.2.3-1 GEOMETRIC' CONFIGURATION CRITERION FOR PAYLOADS OTHER THAN DRIVER ASSEMBLY I
0-7b
\\
i
/
R3vicion 6
/
June 5, 1981 6
TABLE 0.2.3-1
/
~
),,/
MOST RADIDACTIW PAYLOADS TO BE TRANSPORTED THE T-3 CASK
'O Fiesile Material Fuel burnup Irradiation Decay Period Payload kg,Pu-239+U-235 WD/kg PowerDepsity days Watts /cc 21 Carbide /
1.0
$80 1,860 90 Nitride Pins 109 Standard FTR Driver Fins 4.4 580 2,056 250 46 FTR Carbide /
4.2
$80 1.880 250 Nitride Pins 46 Pins of 4.4 580 2,056 25C Mixed Type (Fuel Scrap) 50 EBR-II Ex-4.5 S.150 3,600 250 perieental dxide Pins 50 50 EBR-II
'5
<250 2,056 Carbide / Nitride Pins 217-Pin FTR 10.0
<80 2,056 350
~
Driver Assembly 0.2.3.1 Radioactivity The identity and radioactivity of the material which the T-3 cask may contain are given in 10 CRF 71, Appendix C for transport grouping of radionuclides.
The package will not contain more than 200,000 Ci of mixed fission products and activated materials.
0.2.3.2 Chenical and Physical Forn The T-3 cask'is designed to carry a variety of quantities and g
types of spent fuel pins consisting of austenitic or ferritic stainless steels, or other high strength alloy clad elements containing solid fuel forms.
0-8 j
R3 vision 6 Juno 5, 1981 0.2.3.3 Nuclear Safety Configuration Individual fuel pin assemblies are inserted in an array in the T-3 Cask that is geometrically fixed by the basket support structure.
Analyses described in Section 3.1.5.4.1 and Section 3.1.5.4.2 assure the integrity of the support baskets in the worst possible cask drop accident.
Thus it 6
is impossible for the fuel pins to assume a more reactive configuration due to structural failure of the support l
basket.
i l
Analyses described in Chapter 5 show that criticality safety is assured for all conditions of hydrogenous moderation and reflection described in 10 CFR 71.
The presence of the sup-port basket was found to have no~significant effect on either
[
payload criticality or shielding requirements.
In normal transport conditions the cask is of course dry.
The cask i
l has been desi'ned so as not to deform significantly under i
a worst case accident condition, so that the normal config-1 i
uration of fuel pins can always be assumed with respect to criticality and shielding requirements.
t l
l 0.2.3.4 Maximun Weicht The Maximum weight of the T-3 cask contents is 500 pounds.
0.2.3.5 Maximum Decay Heat The maximum decay heat for this proposed cask usage is 600 watts.
0-9
Ravicion 6
. June 5, 1981 l
l l
maximum normal operating temperature ( < 200 F) is 20,000 psi l
(A240, Grade 304).
The expected number of operating cycles (defined as the process of going from an empty cask, to one with a maximun heat load, at the maximum normal operating temperature and back again) for the T-3 Cask is below 3000.
From the ASME Code,Section III, Appendix I, Figure I-9.2, the fatigue allowable stress amplitude, 5, of the alternating stress component (1/2 3
of the alternating stress range) for 3000 cycles is 65,000 psi.
The nonfatigue allowable stress limit ASME Section III Divisic S
1, Subject NB-3222.4, however, is 60,000 psi (3 S ).
Since this m
is less than the fatigue allowable, then the nonfatigue allowable 6
stress criteria will govern, giving an unlimited fatigue life.
Areas of stress concentration must be considered separately, since the effect of stress concentrations are not considered when eval-uating the Normal Loads discussed above.
The worst case stress concentration occurs at the retaining ring screw hole locations in the upper portion of the containment wall (see Drawing H4-61289).
Following Reg. Guide 7.6, an alternating stress intensity, SAlt' is calculated and compared with the appropriate fatigue allowable, 6
S a
The maximum stress range at the screw hole locations results from considering the stress states due to:
1.
Maximum pressure and internal heating load and, 2.
An empty cask at -20 F.
Stress State 1 pressure induced stress,es, calculated using a finite element model of the containment vessel, are in Section 1.6.1.3.7 for a 1000 1
psi pressure.
!!aximum pre ssure for Normal Conditions of Trans-port is only 39 psi (reference 2. 4. 4).
Taking stress values at i
1-6b
Ravicion 6 j
Junn 5, 1981 the screw hole location (Element 20) and correcting for pressure:
-28 psi S
= -704
=
yy 10 0 149 psi S
= 3822 (.039)
=
22 l
S
= 6690 (.039) 261 psi
=
33
-27 psi 5
= -686 (.039)
=
12 Thermal expansion stresses given in Section 1.6.1.2 corresponding to Stress State 1 must be combined with the_ pressure induced stresses.
However, the containment wall thermal stress is calculated for the thin portion of the wall.
This value must be corrected for the I
thicker portion of the wall at the screw hole (.322" vs..782").
This can be done by comparing pressure induced axial stresses at each location taken.from Section 1.6.1.3.7:
l Thinwall Section (Element 104),
S
= 5959 psi 22 l-Hole Location (Element 20),
S
= 3822 psi 6
22 3822 The ratio of these stresses,
.641, can then be used to
=
5959 correct the thermal stresses.
The thermal stress for this condition taken from Section 1.6.1.2:
S
= 8701 psi 22 Correcting for hole location, S22 = (.641) 8701 = 5577 psi 1-6b (1)
=
l i
.R3vicion 6 l.,
Junn 5, 1981 I
i Combining thermal and pressure stresses and applying the Reg.
, Guide 7.C recommended stress concentration factor of 4.0:
i Syy = -28 (4)
= -112 psi l
(149 + 5577)4 = 22,905 psi S
=
22 1,044 psi S
= 261 ( 4 )
=
33
-108 psi 12 = (-2 7 ) 4 S
=
Calculating principle stresses (S is.taken as a principle 33 stress because of axial symmetry):
-113 psi S
=
py 22,906 psi S
=
p2 1,044 psi S
=
p3 6
Stress State 2:
0 The only stress to be considered is the ther al stress at -20 F.
From Section 1.6.1.2 the stress for this condition:
0 Syy =
S22 = -4450 psi 0
8
=
33 l
L Correcting for hole location and stress concentration as before:
4 (.641)(-4450) = -11410 psi S
=
22 1-6b (2 )
l Rsvicion 6 June 5, 1981 Sincc all other stresses are zero, the principle stresses are:
l l
pl = 0 S
L S
= -11410 psi p2 l
5
=0 l
p3 I
l l
.The principle-stresses of Stress States 1 and 2 must now be i
combined following Reg. Guide 7.6 Paragrar1 8.5 to get Salt
- i (22905 - (-11410)) = -34427 psi
( (-112 ) -0 )
S
=
2 6
(1044 - 0) = 33271 psi (22905 - (-11410))
S
=
3
((-112) - 0) = 1156 psi (1044 - 0)
S
=
31 S
is the maximum giving:
i 2
1
-34427
= 17214 psi Salt "
2 This value is less than either S or minimum S showing that a
fatigue requirements are satisfied at stress concentrations as well as other portions of the containment vessel.
t 1-6 h(3)
~R3 vision 6 June 5, 1981 i
1.1.2.3.2.2 Bolts End Closure Bolts - For'the end closure bolts the maxinum cyclic stress is due to the preload.
From " Machine D3 sign", September 11, 1969, P. ige 20, Table 4, the preload is given by:
T I
" 16 Y
Where:
=
P = bolt load, lb.
T = torque, in-lb K = torque coefficient l
l D = nominal bolt diameter, in.
l l
t l
I l
1-6b ( e. )
3
---t 4
,,ee<
,---.3.e--
-e
Re3Pision 6 June 5, 1981 ma N
6 Stress Location A
L LT E
6
-6 outside 78.5 166.7 95.7 28x10 9.2x10
.147 9876 6
-6 I
Lead 394 166.7 99.8 2x10 16.4x10
.276
-770 6
-6 Inside 8.40 166.7 100.1 28x10 9.2x10
.154 8701
(.147) (78. 5) + (. 27) (39 4 ) (2/29) + (. 54) ( 8. 4 0) 23.67
- 2058
=
gg,
8.40 + 394(2/28) + 78.5 115.04 j
6
( 2 Bx10 ) (.205 E.147 ) /16 6. 7 = 9 876 psi (Outer shell) o
=
g 6
( 2x10 ) (.10 58. 27 0) /166. 7 = -107 8 psi (Lead) o
=
t 6
of = (2 8x10 ) (.2 058.15 4 ) /166. 7 = 8701 psi (Inside shell) i s similar calculation for a constant cask temperature of -20 F t
(assuming in this case the ' a
bonded to the cask walls) gives:
a i
Location A
L AT E
a 6
Stress
-6 Outside 78.5 166.7
-9C 28.3x10+6 9.2x10
.138" ' -4450
-6 Lead 394.0 166.7
-90 2.0x10 16.llx10
.246" 981
-6 Inside 8.4 166.7
-90 28.3x10 9.2x10
.138" -4450 1-23a
i,
. Revision 6 June 5, 1981 1.6.1.3 Stress Calculations The T-3 package containment vessel consists of a 147 inch long cylindrien1 tube fabricated of an eight (8) inch schedule 40 stainless steel pipe.
At each end the cylindrical tube is capped by machined end assemblies'with removable caps and plugs sealed by redundant "o"-ring seals.
Evaluation of stresses in the eight (8) inch cylindrical tube, at sc.ae distance from the end assemblies, is straight forward and completely determined by first l
principals ( d =_p R/t or 6 = p R/2t).
In the region of the end assemblies, stress analyses have been performed using an axisymmetric finite element method (FEM) analysis.
i l
l These FEM analyses were performed on a unit load basis using i
a r;ominal pressure of 1000 psi.
The results indic 'e that l
L predicted stresses are well within the allowables prescribed by Section III of the ASME Boiler and Pressure Vessel Code.
1-23b
Revision 6 i
June 5, 1981
- ~
I 4.0 SHIELDING EVALUATION l
4.1 Discussion and Results I
The design of the T-3 Cask insures sufficient shielding for neutron and gamma 6 radiation to comply with the -limits - on dose
~
rates external to the cask specified in 10 CFR 71..The cask l.
l shielding is provided by lead in a steel casing.
The design.
of the stai~nless steel removable plug port at the top of the l
l cask and the push rod insertion port at the bottom'were analyzed i
for gamma streaming through the gap between the port;and the vask wall.
The results of the shielding calculations for the worst-case neutron and gamm'a'scources are. summarized in. Table 4.1.
The cask is intended to carry a variety of types and quantities of spent fuel pins from both the Fast Test Reactor (FTR) and the Experimental Breeder Reactor (EBR-II) cores.
Table 0.2.3-1 6
i on page 0-8 describes the most radioactive payloads which are currently envisioned for transport in the T-3 Cask, including the fuel burnup, power dansity during irradiation and the decay period following irradiation prior to transport.
i 4-1
... -. - _.. _. _.. _..,., _. _.. _, _... _.... _ _ _. -.,... ~ ~.. _,
[*
Rsvision 6 June.S, 1981 l
TABLE 4.1 l
l
'RESULTS OF SHIELDING CALCULATIONS FOR NEUTRON 6 GA.WA SOURCES DOSE MREM / HOUR SOURCE 3 FEET FROM CASK 6 FEET FROM CASK-CASK-SURFACE 21-Pin C/N G:.rra. Source
.16.8 3.7 1.5 l
Dose Through Cask Wall 21-Pin C/N Gamma Source-22.6 5.4 3.0 6-
- Dose Through Cask Top 21-Pin C/N [iamma Source 3.6 1.o-1.0 Dose ~Through Cask Bettom 21-Pin C/N Garsa Source Extended to 8 Length 21.8 4.1 1.5 6
Dose Through Cask Top 21-Pin C/N_ Gamma Source,
Extended to S' Length 9.7 2.6 1.6 Dose Through Cask Bottom
'FTR Driver Assembly Gamma Source 12.5-3.0 1.2 Dose Through Cask Wall FTR Driver Assembly Gamma Source 4.7 1.2 0.44 Dose Through Cask Top FTR Driver Assembly Gamma Source 0.80 0.31 0.16 Dose Through Cask Bottom FTR Driver Assembly Gamma Source 4.69 1.5 0.5 Extended to 8' Length Dose Through Cask ' Top i
l FTh Driver Assembly Gamma Source 2.84 1.5 0.5 l
Extended to 8' Length Dose Through Cask Bottom
' t 4
4-2
_. _ -... _. _ _. _.. ~.. _,.
R* vision 6 June 5,.1981 4.2' Source Specification Separate' spent fuel payloads were selected as the design-basis sources for neutron and crmma radiation shielding calculations.
The design-basis gamma source is a payload of 21 carbide /
nitride pins with a 90 day. decay period following irradiation (to 80 MWD /kg).
Although the 21-pin payload does not contain the maxijun quantity of fuel material which is carried in.the cask, it is the worst-case gamma source because of its short decay period and its subsequent high energy gamma emmission-spectrum.
The design-basis neutron source is the 217-pin FTR assembly.
Because of the're'latively long half-life of the primary neutron precursor in the spent fuel, the payload containing.the most fuel material will be the worst-case neutron source rather than the payload with the shortest decay time following irradiation.
t 4.2.1 Gamma Source The design-basis gamma source consists of 9.5 kg. of fuel material, of 20 percent plutonium-carbide or nitride (PuC or pun) and 80 percent uranium-carbide or nitride (UC or UN).
The pins are 0.315 in. in diameter, with a 36 in. active fuel l
length.
The gamma emission spectrum from the eleven pins is listed in Table 4.2 in a 16 energy group structure.
The 6
spectrum corresponds to a 90-day cooling period following irradiation to a maximum fuel hnenup of 80 MWD /kg.
The gamma source was calculated by Westinghouse Hanford Company, using l
the computer code RIBD-II which converts inventories of radioisotopes in curies to a combined gamma emmission energv spectrum.
4.2.2 Neutron Source The design basis neutron source consists of 38 kg of fuel material composed of 25 percent Puo and 75 percent UO contained in 2
2 217 pins of 0.23 in. diameter by 36 in. active fuel length.
4-4 1
R3 vision 6 June 5, 1981
.The neutron emmission rate for the FTR driver assembly was calculated by Westinghouse Hanford Company to be 8.0 x 10 neutrons /second, following irradiation at 80 MWD /kg.
This source is due primarily to spontaneous fission of the transuranic isotope Cm-242.
This isotope decays by beta emission with a 163 day half-life.
Combined with other sources of decay neutrons the neutron emission rate at the end of 350 days was specified to be 2'. 3x10 neutrons /second.
The emissien of neutrons is apportioned as 90% from spont,aneous fission and 10% from alpha neutron reactions between fission alphas and oxygen in the fuel.
The neutron emission spectra combining both reactions was determined from Reference 5.
Table 4.3 lists the contribution
.f6 from the two neutron source reactions in the 15. group spectrum.
4.3 Model Specification 4.3.1 Description of Radial and Axial Shielding Configuration l
Figure 4.1 illustrates the T-3 Cask shielding design.
Radiation 6
l dose points for shielding calculations were located on the cask surface, at three feet and at six feet, with respect to the cask side, top and bottom.
Details of the ports in the cask top and bottom are shown in Figures 4-2-A&B and Figure 4. 3 The removable plug port in the cask top was represented by two ana-l lytical models.
Figure 4.2-A shows the plug as a solid steel unit, the design that will-be used for most of the cask peyloads.
Figure 4.2-B shows the plug with an aluminum insert, the " hold-6 down ring" required to support certain pay]oads in the cask.
Shielding calculations were run for both models of the remov-able plug port.
No significant " hot spots" were calculated due to the aluminum insert in the plug; the calculated dose rates were essentially the same using both models (Table 4.1 lists maximum calculated dose rates for all shielding calculations).
i l
The cask is structurally designed so as not to rupture or sig-nificantly deform in any credible accident.
Thus no di atinction 6
wds made between shielding for normal and shieldin tor accident conditions.
4-5
s Revision 6 June 5, 1981 The spent fuel canisters which carry both the neutron and gamma sources were modeled as homogenized cylinders for calculational purposes.
To demonstrate the versatility of'the T-3 Cask to a
~
variety of payload configurations,-both sources were modeled first as 36 inch long cylinders, then as 96 inch long cylinders in each shielding calculation for the cask wall and end ports.
The 21 carbide / nitride pins which are the gamma source are mounted-in a circular' array of 7-inches diameter and 36 inches length.
The homogenized source region defined for shielding
- calculations neglected the mass of the canister, considering the fuel and cladding mass only for a homogenized source density of 0.46 grams /cc.
In homogenizing the FTR driver assembly, only the fuel and cladding were considered, neglecting the mass of the-assembly support structure.
The neutron source was modeled as a homogenized cylinder 4 inches in diameter by 36 inches in length with a homogenized density of 6.5 grams /cm.
9 4-6 i.
r y
g-y- - - - --
4-r yw m-g g
g-p%
y
-w-
,---JW
./:
r
!/ "..
R^ vision 6
'/.'
June 5,-1981-6
>+
. TABLE 4.2 6
G/JNA. EMISSION SPECTRUM l
j
, Group Number Mean Energy Photons Hev Sec.
-1
-3.0 1.8E10 2
-2.7 1.3E03 3'
12. 5 4.8E11 4
2.3 5.3E11.
2,3-3.6E12 5
t i
6 1.9 7.0E10 7
1.7 1.3E11 8
2.475 1.9E13 9
- 1.225 4.6E12 '
l 10 1.0 1.3E15 I
11 0.825 1.0E15 l
12 0.65 4.4E14 13 0.475
- 5. 3EI4 14 0.35 4.8F12 15 0.25 7.8E12 -
16 0.15 1.9E14 l
l s
9
'V r
4-8 1-4
..#,..ww
,-.-,.,-gm
..y
,.,,,,,,,,.,9.,-
,.7-e-,m,
--,,,,,--try
,-g--
( -
TABLE 4.3 NEUTRON ENERGY SPECTRUM
~
Energy Mean Spontaneous (a,n)
Total Flux-to-Dose Group Energy Fission of Reaction neut/sec Conversion Factor Mev Cm-242 neut/sec mrem /h.r per neut/cm'see neut/sec 1
13.5 4.96E3 0
4.96E3 0.209 2
11.1 2.15E4 0
2.15E4 0.166 3
9.09 7.29E4 0
7.29E4 0.148 4
7.27 2.59E5 0
2.59E5 0.148
~
5 5.66 5.91E5 0
5.91E5 0.140 6
4.51 3.40E5 1.24E5 4.64E5 0.133 7
3.54 1.72E6 6.38E5 2.49E6 0.139
~
e.
.I.
8 2.74 1.44E6 5.32E5 1.97E6 0.126 9
2.4 3.52E5 9.60E4 3.16E5 0.126 10 2.09
.l.97E6 2.98E5 2.27E6 0.130 11 1.47 3.56E6 1.38E5 3.70E6 0.133 12 0.83 3.42E6 0
3.42E6 0.119 13 0.33 2.49E6 0
2.45E6 0.054 14 0.057 3.19E5 0
3.19E5 0.0065 15 0.002 1.54E3 0
1.54E3 0.0043 yg u<
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~
'l 4
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Revision 6 June.5, 1981
~ ne radial shielding in the cask wall is illustrated in Figure 4.1 l6 A standard schedule 40 steel pipe is the inner liner with a steel wall thickness 0.3 in.
The steel liner is surrounded by a 7.7 in, annulus of lead, followed by a 1 in. layer of steel.
For the gam a strea.ing analysis the removable. plug port and the push-rod insertion port were modeled essentially according to design as shown in Fi gure s 4. 2 and 4. 3 A three-dimensional point kernal code was required 6l to accurately represent the series of concentric cylindrical regions which make up-the ports, plugs and the steel cask ends.
Calculations were made for dose points along the cask centerline and four to eight inches off 3hc centerline to check for paths of significant gan'ma streaming.
At the top of the cask, a 1/4 in. step is provided where the port plug seats against the cask wall to alleviate streaming along the steel liner as shown in Figure 4.2 In addition to the step, at least an 8 inch thickness 6
is required for the steel plug, followed by at least a 4 inch steel cover plate.
De urethane foam overpacks on each of the cask ends were not considered in the neutron shielding analysis, although subsequent calculations show that their hydro-carbon composition is a very effective neutron shield.
~
At the bottom of the cask, a 1-1/4 i.. 4tep is provided in the push rod insertion port to alleviate streaming.
In addition to the step, a tot at least 13 inches of steel is included in the cask botten and cover plate 1
as shown in Figure 4.3 6
Since steel is a slightly better attenuator of neutrons than lead and the steel cask ends are thicker than the lead cask wall, neutron streaming along the port liners was not considered a significant design factor.
Instead conservative calculations were performed to insure the shielding adequacy of the cask ends for neutrons.
Calculations for the neutron radiation penetrating the cask ends were modeled with the cask end regions surrounding the neutron source volume in a cylindrical geo-metry.
Since a one-dimensional discrete ordinates code was used for 4-13 s
Ravicion 6
L.
-June S, 1981
'A neutron calculations, the source was effectively treated as infinite in length.
In this n.anner the self shiciding of the fuel pins and the assembly; structure was neglected as it effects neutron radiation streaming toward either end of the cask. The' purpose of these calculations was to demonstrate the sufficient attenuation of neutrons through the end regions of the cask by a conservative geometric'model.
4.3.2 Shield Region Densities Table 4.1 lists the atomic. densities for the gamma and neutron l6 '
source as well as for steel and-lead.
l6 7ABLE 4. 4 ATOM DENSITIES OF SO'JRCE AND SHIELD COMPOSITIONS Ato'm Density Material Component
~I 26 atom /cc x 10 Neutron Source:
.0022 Homogenized Chromium
.0032 FTR Driver Assembly Iron
.011 Nickel
.0013 p = 6.46 g/cc Plutonium
.0044 U-235
.000059 U-238
.0081 Gamma Source:
.00096 21 C/N pins Chromium
.00013 Homogenized in a Iron
.00049 7 inch. diameter Nickel
.000073
- Cylinder, Pu-239
.00019 p = 0.46 g/cc 8
.00077 Steel Iron
.085 Lead Lead
.033 4-14
Revision 6
. June 5,-
1981
- 5. 0' CRITICA[.1TYEVALUATION 5.1.
Discussion and Results t
The T-3 Cask does not contain in its design any neutron absorbing material for the purpose of maintaining subcriticality of its cont'ents.
Analysis has shown that al] the payloads for which the cask is designed wi11 ' remain subcritical under the conditions of hydrogenous moderation l
and reflection described in 10-CFR-71.
The cask is designed with suf-ficient strength to prevent significant deformation under credibl.e sc-cident co,n di tion s.
In addition the cask allows sufficient heat trans-fer throu;;h the walls without the requirement of a' mechanical cooling system to prevent melting of the contents and a rearrangement of the fuel material into a more reactive configuration than that designed for shipment.
The cask is designed to carry a variety of quantities and types of spent fuel pins.
The restrictions on cask con-6 tents are described in detail under " Contents of Packaging" in Section 0.2.3.
In order to conservatively represent the payload. restrictions outlined in Section'0.2.3, seven worst-j' case payloads were postulated for criticality calculations.
I The seven payloads include the 217 pin FTR driver ascenbly described in Item 2-a.
of the payload restrictions, as well as six idealized payloads representing the range of fuel types which could be carried as cropped pins described in 6
Item 2-b. of the payload restrictions.
All payloads include the maximum allowable fissile mass of 10 kilograms, with the exception cf the uranium-233 fuel payload which is limited to 5 kilograms fissile nass.
The seven postulated payloads are defined to envelope the reactivity of any payload that will actually be carried in the T-3 Cask.
5-1
R3 vision 6 June 5, 1981' i
To address the worst credible accident. condition, the T-3 Cask with each of the seven postulate'd payloads was modeled as completely flooded and submerged in water for all criti -
cality calculations.
To allow maxinum hydrogenous moderation the cask interior as well as the-inner canister including the volume between pins was modeled as completely flooded. Three-I dimensional models of the cask and-its contents, including a i
detailed model of the fuel pin arrangement were used in'each L
L criticality calculation.
The 123 energy group XSDRN cross-l section set was used, corrected by the code NITAWL for accurate I
resonanc' absorption calculations according to the fuel type.
The criticality analysis is summarized in Table 5.1.
I i
5.2 Package Fuel Loading
~
Table 5.2 describes the cask fuel payloads for which criticality i
l calculations were run.
Since the cask is designed not to deform j
significantly in the event of a worst-case accident, and since 6
l the fuel support baskets have been shown to maintain structural integrity, no distinction was made in performing criticality calculations between normal and accident conditions.
The fuel composition of the 217-pin driver assemblg was modeled in its actual isotopic proportions of 0.7% U-235 and 99.3%
U-238 for uranium-dioxide, and 88% Pu-239, 9% Pu-240, and 3%
6 Pu-241 for plutonium-dioxide.
The fuel comoositions of the six cropped pin payloads were idealized to present a more general as well as conservative representation of the various fue]
5-2 I
Rsvicion'6
'+a June 5, 1981 types to be carried in the cask.
The fuel was modeled as con-sisting only of the fissile isotope, U-233, U-235, or Pu-239 in its dioxide or carbide forn.
The fissile linear density l
and the total fissile mass contained in each payload'were then 6
determined only by the restrictions on cask contents outlined in Section 0.2.3.
Table 5.3 describes the conoliance of each of the seven payloads to either Item 2-a or Item 2-b of the cask l
contents restrictions, which limit the total mass and density l
of fissile material in the cask to insure criticality safety.
~
l 5.3 Model Specification 5.3.1 Description of Calculational Model 6
An elevation view of the T-3 Cask as modeled is shown in Figure 5.1.
For all criticality calculations the cask was modeled as submerged in a reflective water region, extending at least l
100 cm. beyond the cask surface.
A cross-section of the ca'sk payloads is illustrated in Figures 5.2 and 5.3.
Figure 5.2 l
111ustrates the arrangement of pins as modeled for the 217-pin l
l assembly.
Figure 5.3 illustrates the 109-pin Ident 69 inner i
container which would hold unassembled pins or cropped pin sections.
The six idealized pin section payloads are all 6
l configured as illustrated in Figure 5.3.
By examining the l
l detail of an encapsulated pin section in Figure 5.3, the com-I i
pliance will be seen with the criterion on fuel geometry specified in Item 2-b of the cask contents restrictions of Section 0.2.3.
The detail of the encapsulated pin section is seen to comply with the geometric configuration illustrated 5-2a
.~. _
TABLE 5.3 REPRESENTATION OF CASK CONTENTS RESTRICTIONS-IN SEVEN POSTULATED PAYLOADS,'
Payload Description Total Fissile Payload Fissile Applicable Restriction Loading Length Linear
-from Section 0.2.3 kg.
cm.
Density gram /cm.
1.
217-Pin Driver 10 91.44 109.4 Item 2-a.
Assembiy 10 kg. max. fissile mass.
110 g/cm. max. fissile linear density 2.
139 U-235 Oxide 10 83.82 119.3 Item 2-b.
Pin Sections 10 kg. max. fissile mass 60 g/cm. max. fissile linear density considering one-half of.U-235 inventory.
3.
109 U-233 Oxide 5
83.82 59.65 Item 2-b.
m Pin Sections 5 kg. max.' fissile mass E
60 g/cm. max. fissile linear density 4
4.
109 Pu-239 Oxide 10 167.6 59.66 Item 2-b.
Pin Sections 10 kg. max. fissile mass 60 g/cm. max. fissile linear density 5.
109 U-235 Carbide 10 33.82 119.3 Item 2-b.
Pin Sections 10 kg. max. fissile mass 60 g/cm. max. fissile linear density considering one-half of U-235 inventory';
6.
109 U-233 Carbide 5
83.82 59.65 Item 2-b.
4x Pin Sections 5 kg. max. fissile mass 4
60 g/cm. max. fissile o e-linear density m&
-oo 7.
109 Pu-239 Carbide 10 167.6 59.66 Item 2-b H
Pin-Sections 10 kg. max. fissile mass 60 g/cm. max. fissile linear density s
Revision f June 5, 1981'
'in_ Figure 0.2.3-1.
l The fuel _ pin arrays were not homogenized in_the criticality l
calculations, but were modeled in their. actual configurations cladding, and for the cropped of unit cells consisting of-fuel, pin payloads, theistainless steel encapsulation tubing.
In l
l all calculations components _of the payload canister lother than the tubing surrounding each' pin'were lanored.
This is a con-servative approximation, since in tha flooded cask model the samil. volume actually occupied by the steel canister is. replaced l
[
by water, increasing the amount of moderator in the cask.
5.3.2 Package Regional. Densities The cenpositions of all materials used in the criticality _
i I
calculations, including fuel, moderator and shielding materials,_
are listed in Table 5.4.
No regions of the cask were homogenized,.
i however, the non-fissile isotopes of uranium and plutonium were neglected in the idealized models of the six pin section pay-loads.
The maximum allowable linear fissile densities and the 9
total allowable fissile mass in the cask determined the length l
of the pin sections in the six payloads.
t 5-6a l
Rsvision 6 June 5,'1981' TABLE 5. 4 6:I MATERIAL COMPOSITIONS. USED IN' CRITICALITY ANALYSIS-i Ma te ri al '
. Constituents-grams /cc Atom Density 24 atoms /cc x 10 U
FTR Dri ve r Fue 1 0xygen 1.2 O'.0451 t;-2 35 0.046 0.00012 U-238.
6.46-0.0160-2.12-0.0053 Pu-239
~ 0.22 0.00054 Pu-240 l
Pu-241
U-235 6.347 0.0164 l
0.216 0.0164 U-233 3.174 0.00S2 Pu-239 0xide Oxygen 0.218 0.0164 Pu-239 3.174 0.0082 U-235 carbide carbon
- 0. 32 7 0.0164-U-235 6.347 0.0164 U-233 carbide carbon 0.163 0.0082 ti-2 33 3.174 0.0062 Pu-239 Carbide Ca rt.on 0.163 0.0082 Pu-2 39 3.174 0.0082 l
i 5-10
'R2 vision 6 June 5, 1981
- Fs.,^
l in its oxide or carbide form.
Thus the competing effect of neutron absorp-i
'tions byLnonfissile isotopes such as U-238 was neglected, giving a more l
general representation of the maximum possible ireactivity of the fuel l
type.
Except for whole assemblies the most reactive payloads to be car-ried-in the T-3 Cask -would consist -of unassembled pins or cropped pin l
l-s e c t' ions.
The high-reactivity of unassembled pin payloads is due pri-l marily to the dispersion of fuel over a large cross-sectional area in the cask by the Ident 69 fuel pin container.
This large dispersion minimizes geometric byckling or neutron leakage from the fissile package..For the six postulated unassembled pin payloads, the fuel was assumed to be at max-imum dispersion, with all 109 tubes of the Ident 69 containing pins. Al-though the encapsulation tubing for cropped pin sections and the individual pin tubes were included in the model, all other steel in the cask interior was conservatively neglected, and its volume replaced by water moderator.
The cask was modeled as both flooded 'and submerged to achieve the maximum possible hydrogenous moderation. Such a flooded condition would require both the cask and the inner container to rupture in the-event that the cask were dropped in water, an extremely unlikely sequence of events.
5.4.3 Criticality Results t
The results of the criticality analysis for the seven payloads con-sidered are summarized in Table 5. 5.
All calculated values of K-effective 6
are well below 1.0.
The most reactive fuel payload is seen to be the U-233 oxide and carbide.
This is probably due to the high fission probability of neutrons entering the U-233 fuel lump.
The ratio of the fission to the total absorption cross-section is higher in U-233 than in U-235 or Pu-239.
Also of interest is the fact that the 217-pin FTR asserbly has the lowest cal-culated K-effective.
The fuel assembly occupies the smallest cross-sectional area in the cask of all the payloads.
It is apparent that the dispersion of I
l-fuel in the cask, ie. the geomtric buckling, has a primary effect on the reac-1 l
tivity of the payload, j
(
5-12'
.._...-4
R3vicion 6
- =>
e, June 5, 1981 TABLL 5. 5 6
CALCULATE D VALUES OF K-EFFECTM FULL PAYLOAD K-L FFE CTI VE 1.
217-Pin FTR Driver Assembly 0.67 2.
109 U-235 0xide Pin Sectirns,
62.5 cm'. Length, 120 g/cm 0,77 Lincar Fissile Density 3.
109 U-233 0xide Pin Sections 125 cm. Length, 60 g/cm
- 0. 55 Linear Fissile Density 4.
109 Pu-239 0xide Pin Sections 125 cm. Length, 60 g/cm 0.76 Linear Fissile Density 5.
109 U-235 Carbide Pin Sections 62.5 cm. Length,120 g/cm
- 0. 79 Linear Fissile Density 6.
109 U-233 Carbide Pin Sections 125 cm. Length, 60 g/cm O. S5 Linear Fissile Density 7.
109 Pu-23.9 Carbide Pin Sections 125 cm. Length, 60 g/cm O. 75 Linear Fissile Density 5-13
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