RESULTS AND DISCUSSIONS - OPTIONAL MULTILAYER CASK FOR TRANSPORTATION OF SPENT

Chairmen F. FALCI

OPTIONAL MULTILAYER CASK FOR TRANSPORTATION OF SPENT

4. RESULTS AND DISCUSSIONS

The mulMgrouo neutron and gamma radial attenuations through different zones of shield materials and the transport index calculations for the different transport container configurations - designed to transport Cf-252 and Pu-Be spent sources having different radioactive quantities varrying from 0.1A1 to 1A1 type A container and from 1A1 to 10A1 type B container - are computed using the ANISN code and FE-CH package with DLC-75 and ISOTXS data Libraries. A resume of results are shown in Figs.(2) to (7).

Figure(2) shows the multigroup neutron (7 groups, Fig.(2-a)) and gamma (21 groups, Figs.(2-b) and (2-c)) attenuations in shield design configuration(I) for type A container to transport 0.1A1 quantity Cf-252 spent source. Other similar sets of curves are performed for the other radioactivity quantities 0.2A1, 0.3A1, .... and up to 1A1. From all these curves, the total neutron and gamma fluxes <|>tns and <j>tys °n the outer surface of the container shield are then computed together with the corresponding transport index for both neutron (TI)n and gamma (TI)- radiations.

o\ Table ( I I I ) Detailed Design Connguration Models Pb ^B-10^Pb

The results of these calculated neutron and gamma transport^

indices are drawn against the corresponding variation of the radioactivity quantities (0.1A1 to 1A1) and shown in Fig.(3-a) for design configuration(I) type A container. As observed from the figure, The (TI)y curve has higer values than that of (TI)n

curve. Since, according to the regulations, the transport index of any designed system is always taken for the higher value of any of neutron transport index or gamma one, so in Fig.(3-b), the neutron total flux at the outer surface of the container ( type A, design configuration(I)) is drawn against the garama transport index. As clearly seen from the figures, at total surface neutron flux TSNF §tns °^ 3.484E04 n/cm^.sec, the gamma (or system ) transport index is 1.6695E-02 which is far below the regulatory

-,o

Fig (2) Neutron and Gamma Fluxes Attenuation in Shield Design Configuration (I) for Type A Container to Transport 0.1A1 Quantity of Cf-252 Source

safety limits (TK10) of transporting radioactive materials. This means, the design configuration(I) for type A container may carry a Cf-252 spent source of radioactive quantity up to 1A1 (2Cit1l) to yield a system transport index up to 1.6695E-02.

The deep penetration of the multigroup neutron fluxes in the three shield design configurations(II, III&IV) proposed for type B container to transport spent Cf-252 source of radioactivity

B - 1 0 Pb

0 J O 5 O B 1 0 O OU 0 1O 0 2O O JO 0 -»O 0 5O AI Tl

Fig (3-a) Fig (3-b) Fig | 3 ) Type A Container Shield Design Parameters for

C o n f i g u r a t i o n (I| and Safety L i m i t a t i o n s of Cf-252 Source

£o o

Fig ( 4 - c )

Wesh Number

Fig (4) Neutron Fluxes Attenuation in D i f f e r e n t Shield Design Configurations lor Type B Container to Transport 10A1 Q u a n t i t y ol Cf-252 Source

=1 A i i l .§ _? IP -U, 12 =

10 20 JO

Mesh Mumber

10 ro 30 40 Mesh Mumber Fig (5-o) Fig (5-b)

Fig (5) Gamma-rays Attenuation in Shield Design Configuration (II] for Type B Container to Transport 10A1 Quantity of Cf-252 Source

quantity 10A1 are shown in Fig.(4). In order to optimize between these three options to choose the suitable design configuration capable to safely transport such 10A1 quantity, the total surface neutron flux TSHF C'l'tns^ and *-he corresponding transport index

(TI)n are computed for each design. It is found that TSNF (ij>tn3) equals 3.1277E04, 3.4557E04 and 6.99908E4 n/cm².sec with a corresponding (TI)n of 0.12924315, 0.1655432 and 0.33489 for configurations(II), (III) and (IV) respectively. Since configuration(II) has the lowest values of TSNF (4>tn3) and (TI)n, so it is chosen and recoraended for type B container, since it maintained the largest safety reuirsment margin. For this chosen design configuratiun(II), the multigroup (21 groups) gamma attenuation curves are shown in Fig.(5) which shows a sharper cut off for gamma fluxes. This reveals assuring that the design configuration(II) type B container is the best and efficient shield container for multigroup neutron and gamma attenuations as well as it is the safest design through transportation.

The design paramaters important for safe transportation of spent Cf-252 source to be carried in type B container of shield design configuration(II) are shown in Fig.(6) which reflects its capability to transport spent Cf-252 sources of radioactive quantities up to 10A1 values. It must be noted that curves fo Fig.(6) are constructed in a similar way to that of Fig.(3). As clearly observed from Fig.(6) the maximum achievable transport

-ooo

3 0 0

< T I )

(NJ

£u

c 1 C '

-( T I ) r 0 00

5 O A I

2 0 0 Tl

Fig (6-0) Fig (6-b) Fig (6) Type B Container Shield Design Parameters for Configuration (I I) and Safety Limitation of Cf-252 Source

index for 10A1 ( 2 0 C i ) spent Cf-252 source is TI= 3.7651652 for the chosen design c o n f i g u r a t i o n ( I I ) . At this TI value, which is far below the permitted value (TK10) recommended by the IAEA for safe t r a n s p o r t a t i o n ( 1 ~2J , the total neutron f l u x on the outer surface TSKF of type B container is 3.1277E04 n / c m2. s e c .

S i m i l a r results are also obtained for Pu-Be spent source. An example for these results is presented in F i g . ( 7 ) . In this f i g u r e the m u l t i g r o u p neutron and gamma f l u x e s a t t e n u a t i o n in shield design con£iguration(I) - for type A container and proposed to transport a Pu-Be spent neutron source of 0.2CÎ remained a c t i v i t y - are shown. The results shows that this design c o n f i g u r a t i o n ( I ) type A container is suitable to transport 0.2Ci spent Pu-Be source whereas the design c o n f i g u r a t i o n ( I I ) type B container showed also to be the recommended design to transport a spent Pu-Be source of 2Ci remained r a d i o a c t i v i t y .

Mesh Number Fig(7-a)

t/1esh Mumber F , g ( 7 - b )

Fig (7-c)

t.'lesh Mumber

Fig ( 7 | Neutron and Gamma Fluxes A t t e n u a t i o n in Shield Design C o n f i g u r a t i o n ( I ) for Type A Container to Transport 0 2Ci A c t i v i t y of Pu-Be Source

5. CONCLUSIONS

From the a n a l y s i s of the results of the four design c o n f i g u r a t i o n options for m u l t i l a y e r container (or cask) to transport spent sealed neutron sources, the f o l l o w i n g conclusions are obtained:

1. According to the regulatory s a f e t y r e q u i r e m e n t s , the design c r i t e r i a for spent source t r a n s p o r t i n g container design are attained by keeping to a m i n i m u m both the s y s t e m transport index and the total neutron f l u x on container outer surface.

2. These c r i t e r i a are achieved for spent C f - 2 5 2 in the design c o n f i g u r a t i o n ( I ) proposed to type A container ( f o r 0.1A1 - 1A1 r a d o a c t i v i t y q u a n t i t y range) and the design c o n f i g u r a t i o n ( I I ) proposed to type B container ( f o r 1A1 - 10A1 q u a n t i t y r a n g e ) . 3. For 0 . 2 C Î spent Cf-252 source (1A1 q u a n t i t y ) , the total neutron f l u x on container surface ( t y p e A, design c o n f i g u r a t i o n ( I ) ) is found to be 3.4854E04 n / c m ^ . s e c and the system transport index is 4 . 5 7 4 5 6 E - 1 . On the other hand, for 20 Ci spent C f - 2 5 2 source (10A1 q u a n t i t y ) , the total neutron f l u x on c o n t a i n e r

surface (type B, design configuration(II)) is 3.1277E04 n/cm^.sec and the system transport index is 3.7651652.

4. The two coaxial laminated shells (design configuration(II)) results in the best attenuation characteristics for both neutrons and gamma radiations for the higher range of radioactivity quantities ( 1A1 - 10A1 ).

5. The type B container shield design configuration(II) shows capability to satisfy the design criteria and the regulatory safety limits are adequately achieved for both Cf-252 (2Ci-20Ci) and Pu-Be (2Ci) spent sealed sources. Also, the type A container design configuration(I) is suitable to transport (0.2Ci) spent sealed Pu-Be source.

[9] AHMED, Ensherah E. and RAHMAN, F.A.; "FE-DLM: An Artificial Intellegence Design Logic Module for One and Two Dimensional Multigroup Multiregion Discrete Ordinate Calculation Control", to be Published (1994).

[10] AHMED, Ensherah E. and RAHMAN, F.A.; "Shield Safety Criteria for Spent Fuel Elements Package Design" to be Published.(1994) .

REFERENCES

[1] IAEA SAFETY STANDARDS; "Regulations for the Safe Transport of Radioactive Materials", S. S. No. 6, IAEA, Vienna (1990).

Also, IAEA S.S. No. 7 (1987) and IAEA S.S. No. 37 (1987).

[2] IAEA; "Safe Transport of Radioactive Course No.l, IAEA, Vienna (1990).

Material", Training

[3] RSIC Computer Code Collection CC-514 MICRO; " ANISN/Pc, One Dimensional Discrete Ordinate Transport Code", contributed by EG&E, Idaho (1987).

[4] SAHMAN, F.A., EL-KALLA, E.H. and AHMED, Ensherah E.; " A Theoretical Model for Gamma-ray Attenuation and Distribution in a Laminated Shield", Int.J. Radiât. Appl. Instrum., Fart A, Appl.

Radiât. Isot., 40, Ho. 9, Great Britain (1989) 783-787.

[5] RSIC Data Library Collection DLC.-75/BUGLE-80; contributed by ORNL, Oak Ridge, USA (1985).

[6] ISOTXS Data Library, In Ref. [3].

[7] COURTNEY, J.C. (ED.); "Handbook of Radiation Shielding Data", ANS/SD-76/14 (1976).

[8] MEGMUD, R.M., ISMAIL, M.S. and KAMEL,S.A. ;" Pulse Shape Discrimination Set-up and Investigation of N-Gamma Discrimination", AREAEE./Rep.-281, AEE,Cairo, Egypt (1983).

STATUS OF THE BENEFICIAL USES SHIPPING SYSTEM CASK (BUSS)*

H.R. YOSHIMURA, R.G. BAKES, D.R. BRONOWSKI Sandia National Laboratories,

Albuquerque, New Mexico, United States of America Abstract

The Beneficial Uses Snipping System cask is a Type B packaging developed by Sandia National Laboratories for the U. S. Department of Energy. The cask is designed to transport special form radioactive source capsules (cesium chloride and strontium fluoride) produced by the Department of Energy's Hanford Waste Encapsulation and Storage Facility. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations The cask prototype has been fabricated and Certificates of Compliance have been obtained.

INTRODUCTION

This paper presents a status report on the development of the Beneficial Uses Shipping System (BUSS) cask. The purpose of U. S. Department of Energy's (DOE) Beneficial Uses of Nuclear Byproducts Program is to develop and encourage beneficial uses of nuclear byproduct isotopes such as cesium-137 and strontium-90. Applications include thr use of gamma irradiation to improve the quality of certain food products, to disinfect municipal sewage sludge, and to sterilize medical products.

The transportation of cesium chloride or strontium fluoride capsules produced by the DOE Waste Encapsulation and Storage Facility (WESF) at Hanford, WA, to and from commercial licensed facilities is performed in a Type B packaging (cask) certified by the U. S. Nuclear Regulatory Commission (NRC). Sandia National Laboratories was funded by the DOE to develop the BUSS cask to support transportation of these sources

The BUSS cask was designed to maximize payload within prescribed weight and size limits established by WESF and to serve as a safe, reliable, and efficient alternative to existing

transportation systems. The cask design must be consistent with DOE policies for containment and as low as reasonably achievable radiation exposure and must comply with applicable regulations.

A major goal of the BUSS cask development program was to obtain regulatory approval of the design through verification by means of state-of-the-art analysis techniques.

CONTENTS DESCRIPTION

The approved contents to be transported in the BUSS cask are special form capsules of either melt-cast cesium chloride or press-filied strontium fluoride [ 1,2]. Each source is doubly encapsulated with a 316-L stainless-steel outer layer and an inner steel capsule made from Hastelloy C-276 for the strontium fluoride or from 316-L for the cesium chloride. The capsule assemblies are about 7 cm in diameter, 53 cm long, and weigh about 8 kg. A cutaway sketch of a typical capsule is

shown in Figure 1 Containment for the BUSS cask is provided by the special form nature of the source capsules.

CAPSULE CAPS

CESIUM CHLORIDE OR STRONTIUM FLUORIDE

INNER WALL

OUTER WALL

Figure 1. Schematic of a typical Waste Encapsulation and Storage Facility capsule.

CASK DESCRIPTION

The major components of the BUSS cask system include the cask body and lid, basket, impact limiters, personnel barrier, and shipping skid. Figure 2 shows an exploded view of the BUSS cask.

BASKET WITH CAPSULE

MSBE-TURNBUCXLE 4REO.U

LID

-IMPACT LIMITER

IMPACT LIMITER TAPE GROOVE TRUNNION

CASK TAPE GROOVE

This work was performed at Sandia National Laboratories, a US Department of Energy facility, under contract DE-AC04-94AL85000.

Figure 2. Exploded view of the BUSS cask.

The cask body is constructed from a one-piece 304 stainless-steel cylindrical forging. The wall and end of the cask body are a minimum of 33 cm thick. Eleven integral circumferential fins are

machined on the outer surface of the cask body for heat dissipation The cask closure is a one-piece, 33-cm-thick 304 stainless-steel forged lid, weighing about 680 kg. The lid is bolted through the 10-cm-thick lid flange to the cask body with 12 ASME SA-286 steel 1-1/2 in. (3.81-cm) dia bolts All openings into the cask interior are fitted with bolted-on lids, each having a combination metallic-elastomenc double seal. The inner containment seals used in the cask closure and port lid covers are high-temperature copper Helicoflex seals rated for a 450° C operating temperature [3J The outer elastomenc seal serves to provide the test volume for leak testing.

The cask's contents (capsules) are carried in one of four removable solid stainless-steel basket configurations An example of a 16-hole basket is shown in Figure 3.

(3.17 en)

MATBUl. STEH, STAINLESS TYPE 3O4

Figure 3. Typical basket configuration (16-capsule CsCI).

Depending on the thermal power level of cesium or strontium capsules to be transported, one of tour different basket configurations may be used (see Table 1 for cask content limits).

Table I Beneficial Uses Shipping System Cask Radioactive Material Limits Bisket Capacity/Maximum Thermal

Capsule Thermal Power Power AclivHy Capsule Type tno of caosu'es fWll (kWI (millions of CO*

Cesium chlonde (Cs-137) Strontium fluoride

(Sr-90)

16 12 6 4

(250) (333) (650) (850)

4 0 4 0 39 34

085 085 065 056

* 1 O = 37 GBq

Steel-encased polyurethane-foam impact hrruters are attached to each end of the cask body to provide impact protection to the cask system These impact limiters are retained by four turnbuckles and two tape joints [4]. The tumbuckles are used to secure the impact limiters to the cask body during normal handling opérations. The tape joints become effective during accident environments to hold the limners onto the cask body during impact. 1 hese tape joints are loose fitting and do not take effect until large forces are imposed on the impac: hmi'.ers The tape joints

are unique in cask design and are normally found in applications required to withstand large shear loads. Because of the near one-to-one aspect ratio of the cask body, the ends of the cask do not extend into the impact lirriters sufficiently to produce large resisting forces to counteract the moments developed on the impact limners during side drops. It was determined analytically that these moments may be practically resisted through devices which generate large shear forces such as (ape joints.

The assembled cask with impact limiters is transported on its shipping skid as shown in Figure 4.

A personnel barrier is used to prevent unauthorized access to the hot surfaces of the loaded cask.

The weight of the loaded cask including its shipping skid and personnel barrier is approximately 15,310 kg. The cask can be dry- or wet-loaded and unloaded, depending on the facility.

*ED CA» BOOT

Figure 4. Cask assembly showing personnel barrier and shipping skid.

CASK DESIGN AND DEVELOPMENT

The design process included shielding, structural, and thermal analyses of the BUSS response to regulatory normal and hypothetical accident conditions. Results of these analyses for the final design were incorporated in the Safety Analysis Report for Packaging (SARP) [5].

The principal design criteria for structural integrity and shielding used during development of tlie BUSS cask were specified in 10 CFR Part 71 [6]. Of those conditions, the hypothetical accident conditions are the most stringent. They include 9-m drop, 1-m puncture, 30-mmute thermal, and water immersion tests.

Shielding Analysis

We performed detailed shielding analyses of the BUSS cask loaded with 16 cesium chloride capsules to determine the radiation environment external to the package. The 16-capsule cask was found to be a more extreme shielding problem than (he system loaded with six strontium fluoride capsules. The shielding assessments included multienergy group discrete ordinales and Monte Carlo computer analyses [7,8,9], The radiation transport analyses of the BUSS cask were

S

performed ( l ) to evaluate the shielding capabilities of the package for both normal and accident conditions, and (2) to determine the energy deposition profiles in the container for use in the thermal evaluation of the system. Separate one-, two-, and three-dimensional ( 1-D, 2-D, 3-D) finite-difference models were developed for both cases. Table 2 gives the calculated results for normal operation and post-accident radiation levels for the BUSS cask loaded with 16 cesium chloride capsules as well as maximum levels specified by the regulations

Table 2. Summary of Radiation Levels (mrem/hr) for the Beneficial Uses Shipping System formal Conditions

Source Package Surface 2 m from Surface F.nd Side End Side Gamma" 64 29 29 12 10CFR71 200 10 Regulation

Accident Condition 1 m from Surface End Side

11 46 1000

As shown above, the cask meets the applicable shielding performance requirements specified in 10 CFRPartVl.

Structural Analysis

The principal structural members of the BUSS cask include the body, lid, and the impact Jimiters.

The integrity of the system is assured for both normal operation and accidents by the performance of the structural members and the presence of high-quality metallic seals at every opening into the cask interior. In combination with the impact limiters, we show that this boundary is virtually unaffected by the normal and hypothetical accident conditions specified in 10 CFR Part 71.

9 Meter Free-Drop

We evaluated the performance and structural integrity of a BUSS cask subjected to the hypothetical accident free-drop test with 2-D and 3-D finite-element analysis techniques. Three orientations at impact were evaluated. (1) end, (2) side, and (3) center of gravity over comer. The finite-element models were generated by using QMESH [10] and PATRAN [11], they were analyzed with Hondo II [12] and DYNA3D [13]. The deformed shapes and stress distributions were plotted with MOVIE BYU [14]. Accelerations were also obtained to evaluate lid integrity.

Table 3 shows the predicted values for impact limiter crush, cask body acceleration, and von Mises equivalent stress in the cask for three impact orientations at the most severe operating temperature condition.

Table 3 Predicted Foam Crush and Peak von Mises Stress for the Beneficial Uses Shipping System Cask in the Hypothetical Accident 9 Meter Free-Drop at -40° C Orientation Crush fcml Acceleration (el Stress (MPal End

Side Corner

15.5 191 28.7

105 97 75

593 15.2 20.0 As shown above, the cask wall is stressed to values significantly less than yield minng the 9-m drop event. Evaluation of the bolting arrangement using deceleration values on the cask lid

indicated that the seal will maintain its integrity in all impact orientations. Thus, the cask body is essentially undamaged when subjected to the second event of the hypothetical accident sequence, the 1-m drop onto a mild-steel pin.

1 Meter Puncture

We determined the structural response of the BUSS cask to the hypothetical accident puncture test with analyses similar to that used for the drop. To ensure analysis of the most severe accident, we evaluated puncture in three orientations: (1) cask impacting the punch on its side, (2) corner of the cask directly below the center of gravity impacting the punch, and (3) closure end impacting the punch. Each analysis was performed without the impact limiter in place tc produce maximum damage to the cask. For the side punch, the cooling fins were not modf.led.

In every case, we found that either the cask body or lid during closure end impact would be plastically deformed near the impact point. The damage would be limited to a shallow circular indentation corresponding to the cross-sectional dimensions of the end of the puncture bar.

Elsewhere in the cask, the bulk of the material remains elastic (The average stress is 15 MPa or less).

Since the cask body is only moderately stressed and retains its containment and structural integrity, the cask body configuration (geometry) when subjected to the hypothetical accident thermal test is virtually unchanged.

Thermal Analysis

We evaluated the thermal responses of the BUSS cask for normal conditions of transport and hypothetical accidents with finite-difference modeling techniques and pre- and post-processing software. From the geometric description of the cask, we used PATRAN to generate a 2-D finite-element mesh into finite-difference data and Q/TRAN [ 15] to analyze the finite-difference model.

SYMMETRY LINE'

CASK WALL

GAP WITH H»

AIR GAI

BASKET GAP WTTH H«

\'SYMMETRY LINE

Corresponds to the total dose rate since contents do not include neutron-emitting materials Figure 5. Cask thermal analysis model (16 capsule basket).

Once the model was analyzed, an inverse translator was used to convert the data back into a finite-element representation lor post-processing. We used PATRAN to post-process the analysis data The finite-difference models used for the evaluation of normal and accident conditions were similar The model used in the hypothetical thermal accident evaluation was essentially the same as the normal condition model except that the exterior boundary condition was changed to simulate thermal input from the fire as defined by the regulations. We modeled the body, basket, capsules, cooling fins, and the effects of the circumferential gaps between the body components (Figure 5) We conservatively assumed that heat was transported across the gaps only by conduction and radiation Omitting convective transport across the gaps ensures greater predicted temperatures for the cask interior than expected under actual conditions. Heat loss from the exterior surface of the

Dans le document Developments in the Transport of Radioactive Waste | IAEA (Page 186-195)