Text

Renewal of Certificate of Compliance No. 9246 Safety Analysis Report (Redacted)
	ML22039A080
Person / Time
Site:	07109246
Issue date:	10/01/2011
From:	; US Dept of Commerce, National Institute of Standards & Technology (NIST)
To:	; Office of Nuclear Material Safety and Safeguards
	Chris Allen NMSS/DFM/STLB 301-415-6877
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ML22033A231	List: ML22033A239; ML22039A080;
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Safety Analysis Report for Packaging

ST Package Certificate of Compliance Number 9246 NlSI National Institute of Standards and Technology U.S. Department of Commerce

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Safety Analysis Report for Packaging

ST Package Certificate of Compliance Number 9246 October 2011 U.S. Department of Commerce Rebecca M Blank, Acting Secretary National Institute of Standards and Technology Patrick D. Gallagher, Under Secretary for Standards and Technology and Director

Certain commercial entities, equipment, or materials may be identified in this document in order to describe an experimental-procedure or concept adequately. Such identification is not intended to imply recommendation or endorsement by the National Institute of Standards and Technology, nor is it intended to .imply that the entities, materials, or equipment are necessarily the best available for the purpose.

Table of Contents 1 General Information ................................................................................................................................................. 1-1 1.1 Introduction ........................................................................................................................................................ 1-1 1.2 Package Description ........................................................................................................................................ 1-2 1.2.1 Packaging ................................................................................................................................................... 1-2 1.2.2 Contents ...................................................................................................................................................... 1-3 i.2.3 Special Requirements for Plutonium ............................................................................................. 1-3 1.2.4 Operational Features ............................................................................................................................ 1-3 2 Structural Evaluation ............................................................................................................................................... 2-1 2.1 Description of Structural Design ................................................................................................................ 2-1 2.1.1 Discussion .................................................................................................................................................. 2-1 2.1.2 Design Criteria ......................................................................................................................................... 2-1 2.1.3 Weights and Centers of Gravity ................._....................................................................................... 2-1 2.1.4 Identification of Codes and Standards for Package Design ................................................... 2-1 2.2 Materials .............................................................................................................................................................. 2-2 2.2.1 Material Properties and Specifications .......................................................................................... 2-2 2.2.2 Chemical, Galvanic, or Other Reactions ......................................................................................... 2-2 2.2.3 Effects of Radiation on Materials ..................................................................................................... 2-2 2.3 Fabrication and Examination ...................................................................................................................... 2-3 2.3.1 Fabrication ................................................................................................................................................ 2-3 2.3.2 Examination .............................................................................................................................................. 2-3 2.4 General Requirements for All Packages .................................................................................................. 2-4 2.4.1 Minimum Package Size ......................................................................................................................... 2-4 2.4.2 Tamper-Indicating Feature ................................................................................................................ 2-4 2.4.3 Positive Closure ....................................................................................................................................... 2-4 2.5 Lifting and Tie-Down Standards for All Packages .............................................................................. 2-5 2.5.1 Lifting Devices .......................................................................................................................................... 2-5 2.5.2 Tie-Down Devices ................................................................................................................................... 2-5 2.6 Normal Conditions of Transport ................................................................................................................ 2-6 2.6.1 Heat .............................................................................................................................................................. 2-6 2.6.2 Cold ............................................................................................................................................................... 2-6 2.6.3 Reduced External Pressure ................................................................................................................ 2-7 2.6.4 Increased External Pressure .............................................................................................................. 2-7 2.6.5 Vibration ..................................................................................................................................................... 2-7 2.6.6 Water Spray ............... ,.............................................................................................................................. 2-7

2.6.7 Free Drop ...........................................................................................................................................:....... 2-7 2.6.8 Corner Drop .............................................................................................................................................. 2-8 2.6.9 Compression ............................................................................................................................................. 2-8 2.6.10 Penetration ................................................................................................................................................ 2-8 2.7 Hypothetical Accident Conditions ............................................................................................................. 2-9 2.7.1 Free Drop ................................................................................................................................................... 2-9 2.7.2 Crush ............................................................................................................................................................ 2-9 2.7.3 Puncture ..................................................................................................................................................... 2-9 2.7.4 Thermal ...................................................................................................................................................... 2-9

2. 7.5 Immersion - Fissile Material ............................................................................................................. 2-9 2.7.6 Immersion -All Packages ........................................................................................................., ......... 2-9 2.7.7 Deep Water Immersion Test .............................................................................................................. 2-9 2.7.8 Summary of Damage ............................................................................................................................. 2-9 2.8 Accident Conditions for Air Transport of Plutonium ...................................................................... 2-10

2. 9 Accident Conditions for Fissile Material Packages for Air Transport ...................................... 2-11 2.10 Special Form ..................................................................................................................................................... 2-12 2.11 Fuel Rods ........................................................................................................................................................... 2-13 3 Thermal Evaluation ................................................................................................................................................... 3-1 3.1 Description of Thermal Design ................................................................................................................... 3-1 3.1.1 Design Features ....................................................................................................................................... 3-1 3.1.2 Content's Decay Heat ............................................................................................................................ 3-1 3.1.3 Summary Tables of Temperatures .................................................................................................. 3-1 3.1.4 Summary Tables of Pressures ........................................................................................................... 3-1 3.2 Material Properties and Component Specifications .......................................................................... 3-2

. 3.2.1 Material Properties ................................................................................................................................ 3-2 3.2.2 Component Specifications ................................................................................................................... 3-2 3.3 Thermal Evaluation under Normal Conditions of Transport. ........................................................ 3-3 3.3.1 Heat and Cold ........................................................................................................................................... 3-3 3.3.2 Maximum Normal Operating Pressure .......................................................................................... 3-3 3.4 Thermal Evaluation under Hypothetical Accident Conditions ...................................................... 3-4 3.4.1 Initial Conditions .................................................................................................................................... 3-4 3.4.2 Fire Test Conditions .............................................................................................................................. 3-4 3.4.3 Maximum Temperatures and Pressure ......................................................................................... 3-4 3.4.4 Maximum Thermal Stresses ............................................................................................................... 3-4 3.4.S Accident Conditions for Fissile Material Packages for Air Transport ............................... 3-4 4 Containment................................................................................................................................................................. 4-1 ii

4.1 Description of the Containment System ................................................................................................. 4-1 4.2 Containment under Normal Conditions of Transport....................................................................... 4-2 4.3 Containment under Hypothetical Accident Conditions .................................................................... 4-3 4.4 Leakage Rate Tests for Type B Packages ................................................................................................ 4-4 5 Shielding Evaluation ......,. .......................................................................................................................................... 5-1 5.1 Description of Shielding Design ........................................ ,........................................................................ 5-1 5.1.1 Design Features ....................................................................................................................................... 5-1 5.1.2 Summary Table of Maximum Radiation Levels .......................................................................... 5-1 5.2 Source Specification ........................................................................:....... ,....................................................... 5-2 5.2.1 Gamma Source ......................................................................................................................................... 5-2 5.2.2 Neutron Source ........................................................................................................................................ 5-2 5.3 Shielding Model ................................................................................................................................................. 5-3 5.3.1 Configuration of Source and Shielding .......................................................................................... 5-3 5.3.2 Material Properties ................................................................................................................................ 5-3 5.4 Shielding Evaluation ....................................................................................................................................... 5-4 5.4.1 Methods ...................................................................................................................................................... 5-4 5.4.2 Input and Output Data .............. :........................................................................................................... 5-4 5.4.3 Flux-to-Dose-Rate Conversion .......................................................................................................... 5-4 5.4.4 External Radiation Levels ................................................................................................................... 5-4 6 Criticality Evaluation ..................... .'.......................................................................................................................... 6-1 6.1 Description of Criticality Design ................................................................................................................ 6-1 6.1.1 Design Features ....................................................................................................................................... 6-1 6.1.2 Summary Tables of Criticality Evaluation .................................................................................... 6-1 6.1.3 Criticality Safety Index ......................................................................................................................... 6-1 6.2 Fissile Material Contents ............................................................................................................................... 6-2 6.3 General Considerations .................................................................................................................................. 6-3 6.3.1 Model Configuration .............................................................................................................................. 6-3 6.3.2 Material Properties ................................................................................................................................ 6-3 6.3.3 Computer Codes and Cross-Section Libraries ...........................................................~ ................ 6-3 6.3.4 Demonstration of Maximum Reactivity .............. .-......................................................................... 6-3 6.4 Single Package Evaluation ............................................................................................................................ 6-4 6.4.1 Configuration ...........................................................................................................'................................. 6-4 6.4.2 Results ......................................................................................................................................................... 6-4 6.5 Evaluation of Package Arrays under Normal Conditions of Transport ..................................... 6-5 6.5.1 Configuration ............................................................................................................................................ 6-5 6.5.2 Results ......................................................................................................................................................... 6-5 iii

6.6 Package Arrays under Hypothetical Accident Conditions ............................................................... 6-6 6.6.1 Configuratiqn .................................................................;.......................................................................... 6-6 6.6.2 Results ......................................................................................................................................................... 6-6 6.7 Fissile Material Packages for Air Transport .......................................................................................... 6-7 6.7.1 Configuration ............................................................................................................................................ 6-7 6.7.2 Results ......*.................................................................................................................................................. 6-7 6.8 Benchmark Evaluations ................................................................................................................................. 6-8 6.8.1 Applicability of Benchmark Experiments ..................................................................................... 6-8 6.8.2 Bias Determination ................................................................................................................................ 6-8 7 Package Operations ................................................................................................................................................... 7-1 7.1 Package Loading ....................................................................:.............................................................. :........... 7-1 7.1.1 Preparation for Loading....................................................................................................................... 7-1 7.1.2 Loading of Contents .........................................................*.................;.................:.................................. 7-2 7.1.3 Preparation for Transport .................................................................................................................. 7-5 7.2 Package Unloading ........................................................................................................................................... 7-7 7 .3 Preparation of Empty Package for Transport ...................................................................................... 7-9 7.4 Other Operations ......................................................................................................................_...................... 7-10 8 Acceptance Tests and Maintenance Program ................................................................................................ 8-1 8.1 Acceptance Tests ..........................................................:.......................................................... :........................ 8-1 8.1.1 Visual Inspection and Measurements ............................................................................................ 8-1 8.1.2 Weld Examinations ................................................................................................................................ 8-1 8.1.3 Structural and Pressure Tests ........................................................................................................... 8-1 8.1.4 Leakage Tests ........................................................................................................................................... 8-1 8.1.S Component and Material Tests ......................................................................................................... _8-1 8.1.6 Shielding Tests ........................................................................................................................:................ 8-1 8.1.7 Thermal Tests .......................................................................................................................................... 8-1 8.1.8 Miscellaneous Tests ..................................................;............................................................................ 8-*1 8.2 Maintenance Program .................................................................................................................................... 8-2 8.2.1 Structural and Pressure Tests ...............................................................................................;........... 8-2 8.2.2 Leakage Tests ........................................................................................................................................... 8-2 8.2.3 Component and Material Tests ......................................................................................................... 8-2 8.2.4 Thermal Tests*.......................................................................................................................................... 8-2 8.2.S Miscellaneous Tests ............................................................................................................................... 8-2 9 References ..................................................................................................................................................................... 9-1 10 Appendix ................................................................................................................................................................. 10-1 iv

1 General Information 1.1 Introduction The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance ofneutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is operated 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day, seven days a week with routine shutdowns every five-and-one-half weeks for partial refueling and as needed for maintenance. The ST package, used for transporting a single unirradiated NBSR fuel element, directly supports the routine refueling requirement and the research mission of the NCNR.

NIST originally submitted an application requesting approval of the ST package on February 7, 1992. The U.S. Nuclear Regulatory Commission (NRC) reviewed the application and issued Certificate of Compliance (COC) Number 9246 for the Model Number ST Package on February 26, 1992. The package is currently approved and authorized for use as a Type A Fissile package with a Criticality Safety Index (CSI) of 50 under COC Number 924~, Revision 6.

This report contains the results of the analyses provided by the original package application into the format prescribed by NRC Regulatory Guide (RG) 7.9, "Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material." It should be noted that not all sections prescribed by RG 7.9 are applicable. Additionally, not all sections prescribed by RG 7.9 were discussed or presented in the original package application.

NIST maintains an NRC-approved quality assurance program for activities conducted regarding transportation packages, as required by Subpart Hof Title 10 of the Code of Federal Regulations.

This program is currently approved and effective under Approval Number 0390, Revision 9.

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1.2 Package Description 1.2.1 Packaging.

Package Overall Dimensions 1 The ST package consists ofa 0.9525-cm (0.375-in) carbon steel plate, a 0.3175-cm (0.125-in) thick, 13.335-cm (5.25-in) inside diameter carbon steel welded tube, a 0.635-cm (0.5-in) carbon steel flange, and a 0.635-cm (0.25-in) carbon steel cover plate. The overall length of the package is 177.8 cm (70 in). The approximate gross weight (package and contents, as defined, per 49 CFR) of the loaded package is 34 kg (75 lbs). Engineering drawings of the package are provided in Figures 1.2.1.1 and 1.2.1.2.

Package Containment Features The package opening consists of a flange and cover plate. The flange is tapped with eight (8) 1/4-20 holes used to attach the cover plate with screws. A 0.3175-cm (0.125-in) thick neoprene gasket with the same hole pattern forms a seal between the flange and cover plate. The cover screws protrude through the flange and have a 0.15875-cm (0.0625-in) diameter hole used for attaching a tamper seal. When a fuel element is inserted with the support pieces attached, the cover and gasket are applied, bolted, properly torqued, and sealed.

Package Shielding and Barrier Features There are no safety-related shielding features or personnel barriers incorporated into the original package design.

Package Criticality Control Features There are no safety-related features intended for criticality control (e.g. neutron poisons, moderators, or spacers) incorporated into the original package design.

Package Structural Features There are no safety-related structural features incorporated into the original package design; however, there are features intended to orient, support, and restrict movement of the package contents during normal conditions of transport.

Support pieces are assembled on a NBSR fuel element when inserting it into the package. The outside diameter of the support pieces is slightly smaller than the inside diameter of the package.

This permits insertion of the fuel element and supports into the package and restricts movement during normal conditions of transport.

1 In accordance with various Federal Acts, the Code of Federal Regulations, and Executive Order 12770 (see Preface), it is NIST policy that the International System of Units shall be used in all NIST publications.

However, it should be noted that the package design is based in United States Customary System units. All values in SI units provided are converted, not measured, were not used in design or fabrication of the package, and should be considered "for reference only.

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Package Heat Transfer Features There are no features intended for heat transfer incorporated into the original package design.

Package Markings The package will be marked as required by 49 CFR and 10 CFR Part 71.

Summary of Key Design Parameters A summary of key design parameters are provided in Table 1.2.1.1.

1.2.l Contents Identification and Quantity of Fissile Material The maximum quantity of material per package is one NBSR MTR-type fuel element containing not more than 3 60 g of 23su *.

Physical Form The permitted type and form of material per package is one NBSR MTR-type fuel element composed of enriched uranium and aluminum.

Location and Configuration of Contents Support pieces are assembled on a NBSR fuel element when inserting it into the package to restrict movement.

Each NBSR fuel element is wrapped and sealed in a plastic bag (or equivalent) that is resistant to moisture and dust. The plastic bag assists in maintaining cleanliness of the fuel element and is not considered to be part of the packaging.

1.2.3 Special Requirements for Plutonium The package is not authorized to ship plutonium. Therefore, this section is not applicable.

/ 1.2.4 Operational Features The package design is not considered to be complex as it does not have multiples valves, connections, piping, openings, seals, or boundaries. Therefore, this section is not applicable.

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Table 1.2.1.1: ST Package Key Design Parameters Designer NIST Package Model "ST" Series Package Type Type A Fissile Package Identification Number USA/9246/AF Quantity 4 packages Dimensions Overall Length 177.8 cm 70in Outside Diameter 13.97 cm 5.5 in Inside Diameter 13.335 cm 5.25 in Wall Thickness 0.3175 cm 0.125 in Weight - Gross, or Loaded 34kg 75lbs Weight - Unloaded 24.95 kg 55 lbs Materials of Construction Package Body Carbon steel i

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2 Structural Evaluation 2.1 Description of Structural Design 2.1.1 Discussion The package is limited to a Type A quantity of radioactive material. In addition, the quantity of fissile material (two packages with one unirradiated fuel element each) is limited to less than the critical mass of2 35 U. The criticality safety of the fissile material does not depend on the packaging to maintain geometry. Therefore, there are no structural members or systems that received credit or were identified as performing any specific safety functions in the original package application.

2.1.2 Design Criteria There were no design criteria identified in the original package application.

2.1.3 Weights and Centers of Gravity 2.1.3.1 Weights The gross weight of the loaded package is approximately 34 kg (75 lbs). The weight of the unloaded package is approximately 24.95 kg (55 lbs). The weight of the NBSR MTR-type fuel element being transported is 8.26 kg (18.2 lbs).

2.1.3.2 Centers of Gravity The centers of gravity of the package were not discussed in the original package application.

2.1.4 Identification of Codes and Standards for Package Desigri The package was designed, fabricated, assembled, and tested by NIST according to commercially accepted engineering standards. No specific codes or standards were identified in the original package application.

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2.2 Materials 2.2.1 Material Properties and Specifications There were no material mechanical properties.specifically identified for structural evaluation in the original package application.

2.2.2 Chemical, Galvanic, or Other Reactions There were no chemical, galvanic, or other reactions between package materials identified in the original package application.

2.2.3 Effects of Radiation on Materials There were no presumed effects of radiation on the package materials identified in the original package application.

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2.3 Fabrication and Examination 2.3.1 Fabrication The package was fabricated by NIST according to commercially accepted engineering standards.

No specific codes or standards were identified in the original package application.

2.3.2 Examination The package was examined by NIST according to commercially accepted engineering standards.

No specific codes or standards were identified in the original package application.

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2.4 General Requirewents for All Packages 2.4.1 Minimum Package Size The smallest overall dimension of this package is 13.97 cm (5.5 in). This is greater than 10 cm (4 in); therefore, the requirement of 10 CFR 71.43(a) is satisfied.

2.4.2 Tamper-Indicating Feature The tamper-indicating feature is also discussed in Section 1.2.1, "Packaging", under "Package Containment Features". In summary, the tamper-indicating device for the package is provided by a tamper seal that passes through a hole in each of the eight (8) 1/4-20 screws that are used to attach the cover plate and gasket.

To tamper with contents of the package, it is necessary to remove the screws and cover plate; thus, a severed seal will indicate purposeful tampering. With the tamper seal intact, it is impossible to gain access to the contents of the package undetected, satisfying the requirement of 10 CFR 71.43 (b).

2.4.3 Positive Closure The positive closure device is also discussed in Section 1.2.1, "Packaging", under "Package Containment Features". In summary, the positive closure device for the package is provided by the eight (8) 1/4-20 screws that are used to attach the cover plate and gasket. The attaching screws protrude through the flange and have a 0.15875-cm (0.0625-in) diameter hole used for attaching a tamper seal.

By procedure, when a fuel element is inserted with the support pieces attached, the cover and gasket are applied, bolted, properly torqued, and sealed. It is necessary to deliberately loosen the screws with an Allen wrench to facilitate opening. With the screws torqued and tamper seal intact, it is impossible to gain access to the contents of the package unintentionally, satisfying the requirement of 10 CFR 71.43(c).

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2.5 Lifting and Tie-Down Standards for All Packages 2.5.1 Lifting Devices There are no lifting devices incorporated into the approved package design.

2.5.2 Tie-Down Devices There are no tie-down devices incorporated into the approved package design.

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2.6 Normal Conditions of Transport In normal cor;iditions of transport, two (2) packages with one unirradiated fuel element each may be transported at the same time in an exclusive use vehicle.

The package was submitted to a series of tests (NIST, 1990c) in accordance with 49 CFR Part 173 (Type A packaging tests) and 10 CFR Part 71. ;

Prior to the test, the package was loaded with a dummy element. A dummy element is a sample of a NBSR fuel element that is identical to a NBSR fuel element in materials and dimensions except the fuel plates are replaced with plates made from aluminum stock.

Upon completion of the tests, the package was examined for any damage that could affect its integrity. The package was then opened. The dummy element was removed and examined for any damage that would indicate a possible failure of the package containment effectiveness. Finally, the interior of the package was examined for any damage that could affect its integrity.

An inspection of the package revealed no damage, internally or externally, that would affect its integrity. The package exterior did show some cosmetic damage (e.g. areas of chipped paint) that was determined to be acceptable. An inspection of the dummy element also showed no damage from tlie testing.

The test parameters, inspections, and results were documented (NIST, 1990d) and summarized in a report (NIST, 1990c).

2.6.1 Heat The original package application did not discuss subjecting the package to a heat test.

As discussed in Section 3, the package design and contents do not require a certain thermal performance1in order to be transported safely. Considering the limitations placed on the contents and method of transport, exposure to an ambient temperature of 38 °C (100 °F) in still air and in I the shade should not have an adverse effect on the packaging or its contents and surface temperature of the package will not exceed the limitations in 10 CFR 71.43(g).

2.6.2 Cold The original package application did not discuss subjecting the package to a cold test.*

As discussed in Section 3, the package design and contents do not require a certain thermal performance in order to be transported safely. Considering the limitations placed on the contents.

and method of transport, exposure to a cold ambient temperature of -40 °C (-40 °F) in still air and in the shade is not expected or likely.

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2.6.3 Reduced External Pressure The original package application did not discuss subjecting the package to a reduced external pressure test.

Considering the limitations placed on the contents and method of transport, exposure to a reduced external pressure of 25 kPa (3.5 lbf/in 2) absolute is not expected or likely. However, exposure to a reduced external pressure should not have an adverse effect on the packaging or its contents.

2.6.4 Increased External Pressure The original package application did not discuss subjecting the package to an increased external pressure test.

Considering the limitations placed on the contents and method of transport, exposure to an increased external pressure of 140 kPa (20 lbf/in 2) absolute is not expected or likely. However, exposure to an increased external pressure should not have an adverse effect on the packaging or

its contents.

2.6.5 Vibration The original package application did not discuss subjecting the package to a vibration test.

Considering the low weight of the package, limitations placed on the contents, and method of transport, exposure to a vibration incident to normal methods of transportation should not have an adverse effect on the packaging or its contents.

2.6.6 Water Spray The package was subjected to a water spray test that simulated exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (NIST, 1990c). Sprinklers were arranged so that all sides of the package were wetted. Measurement of the simulated rainfall exceeded the 5 cm/h (2 in/h) requirement (NIST, 1990d).

Following the test, the package was examined for *any deterioration or evidence of water entering the container. No water entered the container. No deterioration or other affects from the test were observed or otherwise noted.

2.6. 7 Free Drop The package was subjected to a free drop test from a height of 1.2 m (4 ft) after the conclusion of the water spray test (NIST, 1990c). The package was dropped from a height of 1.2 m (4 ft) onto a poured concrete pad. Due to the geometry of this package, this test was repeated such that: (1) the first drop was on the closed, welded end; (2) the second drop was on the end with the removable plate; and, (3) the third drop was with the package horizontal. As discussed in Section 2.6.8, "Corner Drop", the package was also subjected to a corner drop test prior to the free drop test.

The package was examined for any damage that could affect its integrity. No damage or other affects from the test were observed or otherwise noted.

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2.6.8 Corner Drop Due to the geometry of the package, a corner drop test was required to be conducted prior to the free drop test (NIST, 1990c). The package was dropped from a height of 0.3 m (1 ft) on each corner

-of each end (four drops per end). There was a total of eight drops for this test.

The package was examined for any damage that could affect its integrity. No damage or other affects from the test were observed or otherwise noted.

2.6.9 Compression The package was subjected to a compression test of 325 kg (715 lbs) for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (NIST, 1990c).

This test requires a compressive load applied uniformly to the top and bottom of package in the position the package would normally be transported. Therefore, the package was placed with the longitudinal axis of the package horizontal and each end resting on an unyielding surface.

In accordance with 10 CFR 71.71(c)(9), the compressive load must be the greater offive times the weight of the package or the equivalent of 13 kPa (2 lbf/in 2) multiplied by the vertically projected area of the package. The unloaded weight of the package is 24.95 kg (55 lbs). The projected area of the package (13.97 cm (5.5 in) diameter, 177.8 cm (70 in) length) is 2500 cm2

The compressive load was calculated as follows:

(1) 5 x 24.95 kg (55 lbs.) = 124.75 kg (275 lbs.)

(2) 1300 k~ x 0.25 m 2 = 325 kg (715 lbs.)

m Therefore, a compressive load of at least 325 kg (715 lbs) was required to be placed along the length of the package. This was accomplished by placing layers of lead bricks on top of the package.

Thirty (30) bricks, each weighing 11.95 kg (26.35 lbs)., were placed on top using three layers of eight bricks and a final layer of six bricks (NIST, 1990d). The 357 kg (787.1 lb) load remained in place for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This provided the uniform compression loading on two sides as required by 10 CFR 71.71(c)(9). -

The package was examined for any damage that could affect its integrity. No damage or other affects from the test were observed or otherwise noted.

2.6.10 Penetration The package was subjected to a penetration test by dropping a 3.2-cm (1.26-in) diameter bar, weighing 6 kg (13.23 lbs), from a height of 1 m (3.28 ft) (NIST, 1990c). The package was resting on an unyielding surface, positioned horizontally as it would be during a shipment. The bar impacted the center of the curved portion of the package. The axis of the bar remained perpendicular to the surface of the package throughout the test. The bar did not strike the surface obliquely.

The test bar was examined for any deformity at the impact point; none was found. The package was examined for any damage that could affect its integrity. No damage or other affects from the test were observed or otherwise noted.

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2. 7 Hypothetical Accident Conditions

The package is limited to a Type A quantity of radioactive material. In addition, the quantity of fissile material (two packages with one element each) is limited to less than the critical mass of2 35 U.

The package was not evaluated for accident conditions since the criticality safety of the fissile material does not depend on the packaging to maintain geometry.

2.7.1 Free Drop The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

2.7.2 Crush The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

2.7.3 Puncture The original package application did not discµss subjecting the package to tests based on hypothetical accident conditions.

2.7.4 Thermal The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

2.7.5 Immersion - Fissile Material The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

However, it should be noted that the criticality analysis discussed in Section 6 assumes full moderation and reflection of fuel elements on all sides.

2.7.6 Immersion -All Packages The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

2.7.7 Deep Water Immersion Test The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions. \

2.7.8 Summary of Damage The original package application did not discuss subjecting the package to tests based on hypothetical accident conditions.

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2.8 Accident Conditions for Air Transport of Plutonium The package is not authorized for air transport of plutonium. Therefore, this section is not applicable.

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2.9 Accident Conditions for Fissile Material Packages for Air Transport The package is not authorized for air transport of fissile material. Therefore, this section is not applicable.

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2.10 Special Form The package is not designed or authorized to transport special form radioactive material.

Therefore, this section is not applicable.

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2.11 Fuel Rods The original package application did not specifically discuss the cladding in the context of containment during normal or accident conditions.

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3 Thermal Evaluation 3.1 Description of Thermal Design 3.1.1 Design Features There are no design features that received credit or were identified as important to thermal performance in the original package application.

3.1.2 Content's Decay Heat The contents of the package are limited to unirradiated enriched uranium, which has negligible decay heat. Therefore, the package does not require certain thermal design features in order to transport the contents safely.

3.1.3 Summary Tables of Temperatures There was no discussion of temperatures that could affect the package under normal or hypothetical conditions in the original package application.

3.1.4 Summary Tables of Pressures There was no discussion of pressures that could affect the package under normal or hypothetical conditions in the original package application.

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3.2 Material Properties and Component Specifications 3.2.1 Material Properties There are no thermal properties for materials that received credit or were identified as important to heat transfer in the original package application.

3.2.2 Component Specifications There are no material specifications that received credit or were identified as important to thermal performance of the package in the original package application.

3-2

3.3 Thermal Evaluation under Normal Conditions of Transport 3.3.1 Heat and Cold There was no thermal evaluation of the package in the original package application.

3.3.2 Maximum Normal Operating Pressure There was no thermal evaluation of the package in the original package application.

3-3

3.4 Thermal Evaluation under Hypothetical Accident Conditions 3.4.1 Initial Conditions There was no thermal evaluation of the package in the original package application.

3.4.2 Fire Test Conditions There was no thermal evaluation of the package in the original package application.

3.4.3 Maximum Temperatures and Pressure There was no thermal evaluation of the package in the original package application.

3.4.4 Maximum Thermal Stresses*

There was no thermal evaluation _of the package in the original package application.

3.4.5 Accident Conditions for Fissile Material Packages for Air Transport The package is not authorized for air transport of fissile material.* Therefore, this section is not applicable.

3-4

4 Containment 4.1 Description of the Containment System As stated in the original package specification (NIST, 1990b), the containment system consists of a bolted closure. The bolted closure is synonymous with the package opening, as described in Section 1.2.1, "Packaging", under "Package Containment Features".

The package opening consists of a flange and cover plate. The flange is tapped with eight (8) 1/4-20 holes used to attach the cover plate with screws. A 0.3175-cm (0.125-in) thick neoprene gasket with the same hole pattern forms a seal between the flange and cover plate. The attaching screws protrude through the flange and have a 0.15875-cm (0.0625-in) diameter hole used for attaching a tamper seal. When a fuel element is inserted with the support pieces attached, the cover and gasket are applied, bolted, properly torqued, and sealed.

4-1

4.2 Containment under Normal Conditions of Transport As discussed in Section 2.6, the package prevents the loss or dispersal of radioactive material under normal conditions of transport as demonstrated by physical testing.

\

4-2

4.3 Containment under Hypothetical Accident Conditions There was no evaluation of the containment system under hypothetical accident conditions in the original package application for the reasons discussed in Section 2.7.

4-3

4.4 Leakage Rate Tests for Type B Packages This package is a Type A Fissile package. Therefore, this section is not applicable.

4-4

5 Shielding Evaluation 5.1 Description of Shielding Design 5.1.1 Design Features There are no shielding design features for this package. The contents are limited to unirradiated enriched uranium; therefore, external radiation levels are low and shielding is not necessary.

5.1.2 Summary Table of Maximum Radiation Levels There was no maximum radiation level table provided in the original package application.

5-1

5.2 Source Specification 5.2.1 Gamma Source There was no gamma source discussion provided in the original package application.

5.2.2 Neutron Source There was no neutron source discussion provided in the original package application.

5-2

5.3 Shielding Model 5.3.1 Configuration of Source and Shielding There are no shielding design features for this package. Therefore, this section is not applicable.

5.3.2 Material Properties There are no shielding design features for this package. Therefore, this section is not applicable.

5-3

5.4 Shielding Evaluation 5.4.1 Methods There are no shielding design features for this package. Therefore, this section is not applicable.

5.4.2 Input and Output Data There are no shielding design features for this package. Therefore, this section is not applicable.

5.4.3 Flux-to-Dose-Rate Conversion There are no shielding design features for this package. Therefore, this section is not applicable.

5.4.4 External Radiation Levels The contents are limited to unirradiated enriched uranium; therefore, external radiation levels are low. Operating procedures and routine determinations require that radiation surveys be made prior to each shipment.

There were no external radiation levels discussed in the original package application.

5-4

6 Criticality Evaluation 6.1 Description of Criticality Design 6.1.1 Design Features There are no design features that received credit or were identified as important to criticality control in the original package application.

6.1.2 Summary Tables of Criticality Evaluation The original analyses, performed by Babcock & Wilcox (B&W) (Koudelka, 1991a and 1991b),

assumed a number of fuel elements would be fully flooded and arranged to achieve optimal moderation and reflection geometry. No credit was given to highlight the negative reactivity (poison) effect of the steel in the package and the spacing between the packages in transport.

Summary tables from these criticality ana,lyses are provided in Table 6.1.2.1 and Table 6.1.2.2.

6.1.3 Criticality Safety Index*

The original package application was before the changes to 10 CFR 71.59 in 1996 and therefore did not discuss Criticality Safety Index (CSI) for the package. However, when th(;! NRC and DOT adopted the new regulations, the NRC modified COC 9246 (Revision 2) and included the new CSI provision based on the original criticality analyses.

CSI is based on the number of packages evaluated in the array for the criticality model. The CSI for the package was determined using this equation:

50 CS!= -

N where N = the number of packages used in the array for the criticality model Based on the original criticality analyses, one (1) package was used for N (the array) in the

.criticality model; therefore, the package CSI was set at 50.

6-1

6.2 Fissile Material Contents According to Koudelka (1991a), each standard NBSR fuel element was modeled explicitly to contain 17 fuel plates in the top and bottom fueled region, making a total of 34 plates. The fuel portion of each plate, measuring 6.19 cm by 28.88 cm by 0.05 cm (2.436 in by 11.37 in by 0.020 in), is uniformly loaded with a matrix of 93% enriched U02 and Al to give a total plate loading of 10.294 g 235 U. Clad thickness is 0.039 cm (0.01525 in) and water gap spacing is 0.295 cm (0.116 in). A 15.24-cm (6.0-in) long water-filled center section separates the top and bottom fueled regions of the element. The total 235 U loading for the element is 350 g.

6-2

6.3 General Considerations 6.3.1 Model Configuration The model configuration was not discussed in the original package application.

6.3.2 Material Properties No additional material properties important to the criticality design were identified or discussed in the original package application.

6.3.3 Computer Codes and Cross-Section Libraries To conduct the criticality analyses (Koudelka, 199 la), a model of the fuel element was developed for computer simulations using Standardized Computer Analyses for Licensing Evaluation (Scale) computer software system, KENO V.a three-dimensional (3-D) Monte Carlo criticality computer code, and the Scale 3 16-group Master Library. This was accomplished through CSAS25, a Scale 3 control model.

According to the Oak Ridge National Laboratory (ORNL) website (ORNL, 2011), Scale is a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by ORNL under contract with NRC and US Department of Energy (DOE) to perform reactor physics, criticality safety, radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs. According to the KENO primer (Busch and Bowman, 2003), KENO V.a three-dimensional (3-D) Monte Carlo criticality computer code is the primary criticality safety analysis tool in SCALE.

6.3.4 Demonstration of Maximum Reactivity As shown in Table 6.1.2.1, the upper limit of Keff for four elements arranged two-by-two on a uniform square pitch does not exceed 0.734.

As shown in Table 6.1.2.2, additional analyses were conducted to show that the upper limit of Kerr for seven elements in their most reactive configuration does not exceed 0.881 (Koudelka, 1991b).

6-3

6.4 Single Package Evaluation 6.4.1 Configuration Evaluation of a single package was not presented in the original package application.

6.4.2 Results Evaluation of a single package was not presented in the original package application.

6-4

6.5 Evaluation of Package Arrays under Normal Conditions of Transport 6.5.1 Configuration The analyses used to evaluate the package arrays were based on 10 CFR 71.61, previously titled "Fissile Class III". The regulations have been updated several times since the original submission and these specific requirements are no longer in existence:

§ 71.61 Specific standards for a Fissile Class III shipment.

a) Twice this number of undamaged packages would be subcritical if stacked together in any arrangement, assuming close reflection on all sides of the stack by water; and The assumptions used for the calculations assumed that the given number fuel elements would be fully flooded and arranged to achieve optimal moderation and reflection geometry. No credit was given to highlight the negative reactivity (poison) effect of the steel in the package and the spacing between the packages in transport.

6.5.2 Results This analysis assumed the four (4) fuel elements would be fully flooded and arranged to achieve optimal moderation and reflection geometry with no credit given to the package or spacing. The results are summarized in Table 6.1.2.1 6-5

6.6 Package Arrays under Hypothetical Accident Conditions 6.6.1 Configuration The analyses used to evaluate the package arrays were based on 10 CFR 71.61, previously titled "Fissile Class III". The regulations have been updated several times since the original submission and these specific requirements are no longer in existence:

§ 71.61 Specific standards for a Fissile Class III shipment.

b) This number of packages would be sub critical if stacked together in any arrangement, closely reflected on all sides of the stack by water, and with optimum interspersed hydrogenous moderation. Except as permitted under§ 71.41, each package must be considered to have been subjected to the tests specified in§ 71.73 0-Iypothetical Accident Conditions).

The assumptions used for the calculations assumed that the given number fuel elements would be fully flooded and arranged to achieve optimal moderation and reflection geometry. No credit was given to highlight the negative reactivity (poison) effect of the steel in the package and the spacing between the packages in transport.

6.6.2 Results Two (2) packages contain no more than 720 g 235 U. Criticality cannot be achieved under any condition since this is less than the smallest mass of 235 U required to achieve criticality (NIST, 1992e).

6-6

6. 7 Fissile Material Packages for Air Transport

6. 7.1 Configuration The package is not authorized for air transport of fissile material. Therefore, this section is not applicable.

6. 7.2 Results The package is not authorized for air transport of fissile material. Therefore, this section is not applicable.

6-7

6.8 Benchmark Evaluations 6.8.1 Applicability of Benchmark Experiments According to Koudelka (1991a), the CSAS25 control module has been benchmarked against numerous known-critical systems by B&W and other organizations.

6.8.2 Bias Determination According to Koudelka (1991a), B&W benchmark and validation work shows that the CSAS25 control module does not underestimate the actual KeffValue ofa system by more than 2%.

Therefore, a bias value of 0.02 plus two-sigma (2o-) was added to the calculated values of Keff shown in Tables 6.1.2.1 and 6.1.2.2.

6-8

Table 6.1.2.1- Four NBS Elements on Square Pitch in an Infinite Sea of Water Run ID Element Separation Upper Limit of Kerr NBSG 0.0cm 0.708 NBSH 0.5cm 0.734 NBSI 1.0cm 0.726 NBSJ 3.0 cm 0.590 NBSK 30.0 cm 0.412 Table 6.1.2.2 - Seven NBS Elements on Square Pitch in an Infinite Sea of Water Run ID Element Separation Upper Limit of Kerr NBSB 0.0cm 0.849 NBSC 0.5 cm 0.876 NBSD 1.0cm 0.881 NBSEE 1.5cm 0.874 NBSFF 2.0cm 0.859 6-9

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6-12

7 Package Operations The required determinations and steps presented here were previously outlined in the "NIST "ST" Series Shipping Container Loading, Unloading, and Quality Assurance Procedure" and "NIST "ST" Series Shipping Container Shipper's Checklist" in the original package application. Some changes were required to determinations or steps used for package operations for clarity, consistency with other documents (e.g. certificate of compliance and engineering drawing), and to update steps that reference outdated regulatory requirements.

7.1 Package Loading Local procedures, checklists, or other appropriate documents shall be written and utilized during shipping activities. The following routine determinations (from 10 CFR 71.87) are presented with guidance that can be used to develop steps or inspection points. Any documented results from these local procedures, checklists, or other appropriate documents at the time of each shipment sh.ould be retained in accordance with regulatory requirements, as appropriate.

Previously, the determinations in Section 7.1.1 and Section 7.1.2 were outlined in the "Loading, Unloading, and Quality Assurance Procedure" as presented in the original package application and required by Condition 6 in the Certificate of Compliance.

7.1.l,- Preparation for Loading Before each shipment of licensed material, the users of the package shall make the routine determinations listed in 10 CFR 71.87, as applicable.

Routine determination. 10 CFR 71.87(a):

The package is proper for the contents to be shipped.

Amplifying guidance:

The package is designed and approved for transportation of one NBSR unirradiated MTR-typeJuel element.

Routine determination. 10 CFR 71.87(h):

The package is in unimpaired physical condition except for superficial defects such as marks or dents.

Amplifying guidance:

Inspect the package for any damage that could affect its integrity, e.g. cracks, weld failures, major deformations.

7-1

Routine determination. 10 CFR 71.87(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.

Amplifying guidance:

Contaminations surveys should be conducted on the package in.accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport.

Routine determination. 10 CFR 71.870):

External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in 10 CFR 71.47 at any time during transportation.

Amplifying guidance:

Radiation surveys should be conducted on the package in accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport.

7.1.2 Loading of Contents Before each shipment of licensed material, the users of the package shall make the routine determinations listed in 10 CFR 71.87, as applicable. Non-applicable determinations are included and discussed for completeness.

Routine determination. 10 CFR 71.87(c):

Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects.

Amplifying guidance:

Inspect all bolts, threads, and gasket for condition and damage. Appropriate inspection points that may be considered:

Bolts -Any evidence of binding or galling during operation of the bolt? Any evidence of damage or rounding out to the bolt head?

Threads -Any evidence of thread damage during operation or upon inspection?

Gasket - Are there any portions of the gasket that are torn, ripped, fraying, or appear to interfere with the operation of the threaded fasteners?

Close the package using the cover and gasket. Assure that the cover is snug against the top support on the fuel element.

7-2

Tighten and torque all cover bolts to between 7 ft.lbs and 10 ft.lbs.

The cover bolts are a commercial-off-the-shelf part and are not intended to be torqued in excess of 10 ft.lbs.

Using the hole in the threaded end of the bolt, attach a wire-type tamper seal.

As discussed in Section 2.4.2, installation of a tamper-indicating feature is required to satisfy the general package requirement of 10 CFR 71.43(b).

Routine determination. 10 CFR 71.87(d):

Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid.

Amplifying guidance:

Considering the package design and contents, there is no system for containing liquid; therefore, this routine determination is not applicable.

Routine determination. 10 CFR 71.87(e):

Any pressure relief device is operable and set in accordance with written procedures.

Amplifying guidance:

Considering the package design and contents, there is no pressure relief device; therefore, this routine determination is not applicable.

Routine determination, 10 CFR 71.87(0:

The package has been loaded and closed in accordance with written procedures.

Amplifying guidance:

As discussed, the determinations and guidance in Section 7.1 shall be incorporated into written procedures, checklists, or other appropriate documents that will be utilized during shipping activities. Therefore, this routine determination should assure that those

procedures are followed accordingly.

Routine determination. 10 CFR 71.87(g):

For fissile material, any moderator or neutron absorber, if required, is present and in proper condition.

Amplifying guidance:

Considering the package design and contents, there are no required moderators or neutron absorbers; therefore, this routine determination is not applicable.

7-3

Routine determination. 10 CFR 71.87(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.

Amplifying guidance:

Contaminations surveys should be conducted on the package in accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport.

Routine determination, 10 CFR 71.870):

External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in 10 CFR 71.47 at any time during transportation.

Amplifying guidance:

Radiation surveys should be conducted on the package in accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport..

Aside from these routine determinations, loading the contents of the package requires assembly or installation of supports to the fuel element to limit movement during normal conditions of transport. The following guidance is a suggested outline for assembly or installation of the support while inserting the fuel element into the package.

Assemble nozzle support to fuel element.

Insert fuel element into the package, nozzle end first. Assure that the fuel element is in contact with the bottom support and the rear of the package.

Place top support in the end of the fuel element.

Close the package using the cover and gasket. Assure that the cover is snug against the top support on the fuel element.

7-4

7.1.3 Preparation for Transport Before each shipment of licensed material, the users of the package shall make the routine determinations listed in 10 CFR 71.87, as applicable. Non-applicable determinations are included and discussed for completeness.

Routine determination. 10 CFR 71.87(c):

Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects.

Amplifying guidan,ce:

Verify the tamper seal is installed on the package. As discussed in Section 2.4.2, installation of a tamper-indicating feature is required to satisfy the general package requirement of 10 CFR 71.43(b).

Routine determination. 10 CFR 71.87(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 10 CFR 71.45.

Amplifying guidance:

Considering the package design and contents, there are no structural parts of the package that could be used to lift or tie down the package during transport; therefore, this routine determination is not applicable.

Routine determination. 10 CFR 71.87(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.

Amplifying guidance:

Contaminations surveys should be conducted on the package in accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport.

Routine determination. 10 CFR 71,870):

External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in 10 CFR 71.47 at any time during transportation.

7-5

Amplifying guidance:

Radiation surveys should be conducted on the package in accordance with local written instructions, as appropriate, to ensure the shipment is within the limits. This may require surveys to be conducted during preparations for loading, loading of contents, or preparation for transport.

Aside from these routine determinations, preparation for transport requires the following additional checks. These checks were previously outlined in the "NIST "ST" Series Shipping Container Shipper's Checklist as presented in the original package application and required by Condition 6 in the Certificate of Compliance. The following language is suggested for inclusion in local procedures, checklists, or other appropriate documentation used for preparation for transport.

Verify all required and any local quality assurance procedures are completed prior to shipment.

Verify the package is labeled in accordance with regulatory requirements.

As appropriate, verify that the shipment is in accordance with the Certificate of Compliance and meets any additional regulatory requirements or commitments.

Examples of these requirements may include, but are not limited to: Exclusive use vehicle, Escorted (driver has companion), No more than two packages per vehicle.

Verify that the shipping papers are filled out in accordance with regulatory requirements.

As appropriate, verify all required shipping documentation is complete and provided to the driver.

As appropriate, verify written instructions required for exclusive use shipments have been provided to the driver.

7-6

7.2 Package Unloading In addition to local procedures, checklists, or other appropriate documents, the package shall be received and opened in accordance with the following applicable steps of 10 CFR 20.1906, "Procedure for receiving and opening packages".

Procedure for receiving and opening packages, 10 CFR 20.1906(b)(1):

Monitor the external surfaces of a package labeled with a Radioactive White I, Yellow II, or Yellow III label (as specified in U.S. Department of Transportation regulations) for radioactive contamination unless the package contains only radioactive material in the form of a gas or in special form as defined in 10 CFR 71.4.

Amplifying guidance:

The package does not contain radioactive material in the form of a gas or in a special form; therefore, contaminations surveys should be conducted on the package in accor:dance with local written instructions, as appropriate.

Procedure for receiving and opening packages, 10 CFR 20.1906(b)(2):

Monitor the external surfaces of a package labeled with a Radioactive White I, Yellow II, or Yellow III label (as specified in U.S. Department of Transportation regulations) for radiation levels unless the package contains quantities of radioactive material that are less than or equal to the Type A quantity, as defined in Sec. 71.4 and appendix A to part 71 of this chapter.

Amplifying guidance:

The package contains, but may not exceed, a Type A quantity of material; therefore, radiption surveys should be conducted on the package in accordance with local written instructions, as appropriate.

Procedure for receiving and opening packages, 10 CFR 20.1906(b)(3):

Monitor all packages known to contain radioactive material for radioactive contamination and radiation levels if there is evidence of degradation of package integrity, such as packages that are crushed, wet, or damaged.

Amplifying guidance:

Inspect the package for any damage that might indicate there could be damage to the fuel element inside the package. Suggested inspection points are upon receipt, while opening, and after unloading.

7-7

Procedure for receiving and opening packages. 10 CFR 20.1906(c):

Perform the monitoring required 10 CFR 20.1906(b) as soon as practical after receipt of the package, but not later than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the package is received if it is received during normal working hours, or not later than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months from the beginning of the next working day if it is received after working hours.

Amplifying guidance:

No further amplifying guidance is available.

Procedure for receiving and opening packages. 10 CFR 20.1906(d):

Immediately notify the final delivery carrier and the NRC Operations Center (301-816-5100), by telephone, when: (1) Removable radioactive surface contamination exceeds the limits of 10 CFR 71.87(i); or (2) External radiation levels exceed the limits of10 CFR 71.47.

Amplifying guidance:

No further amplifying guidance is available.

Procedure for receiving and opening packages.10 CFR 20.1906(e):

Establish, maintain, and retain written procedures for safely opening packages in which radioactive material is received; and ensure that the procedures are followed and that due consideration is given to special instructions for the type of package being opened.

Amplifying guidance:

Local procedures, checklists, or other appropriate documents shall be written specific to the package and utilized during package unloading.

Aside from the steps required by 10 CFR 20.1906, package unloading requires the following additional steps. These steps were outlined in the "Loading, Unloading, and Quality Assuranc.e Procedure" as presented in the original package application and required by Condition 6 in the Certificate of Compliance. The following language is suggested for inclusion in local procedures,

checklists, or other appropriate documentation used for package unloading.

Verify the tamper seal is in place and intact. Verify the serial number on the seal is the same as indicated on the shipping papers. (Note: installation of a tamper-indicating feature was required to satisfy the general package_ requirement of 10 CFR 71.43(b))

Remove the tamper seal, cover bolts, cover, and gasket.

Remove the fuel element.

Perform an initial inspection of the fuel element for any observable damage.

Remove the top support and nozzle support.

,I . 7-8

7.3 Preparation of Empty Package for Transport A separate procedure for preparing an empty package for transport was not provided in the original package application.

However, local procedures, checklists, or other appropriate documents shall be written and utilized to ensure the empty package is prepared for transport in accordance with the appropriate regulations and requirements.

7-9

7.4 Other Operations There were no procedures for other operations identified in the original package application.

- I I

7-10

8 Acceptance Tests and Maintenance Program 8.1 Acceptance Tests There are no specific first-use acceptance tests that were identified in the original package application. Additionally, there are currently no plans to construct any new ST packages that would require implementing acceptance tests.

8.1.1 Visual Inspection and Measurements There are no specific first-use visual inspection and measurement tests that were identified in the original package application.

8.1.2 Weld Examinations There are no specific first-use weld examinations that were identified in the original package application.

8.1.3 Structural and Pressure Tests There are no specific first-use structural and pressure tests that were identified in the original package application.

8.1.4 Leakage Tests There are no specific first-use leakage tests that were identified in the original package application.

8.1.5 Component and Material Tests There are no specific first-use component and material tests that were identified in the original package application.

8.1.6 Shielding Tests There are no specific first-use acceptance tests that were identified in the original package application.

8.1.7 Thermal Tests There are no specific first-use thermal tests that were identified in the original package application.

8.1.8 Miscellaneous Tests There are no specific first-use miscellaneous tests that were identified in the original package application.

8-1

8.2 Maintenance Program 8.2.1 Structural and Pressure Tests There are no systems or components that necessitate periodic structural or pressure tests to ensure continued performance of the package.

8.2.2 Leakage Tests There are no systems or components that necessitate periodic leakage tests to ensure continued performance of the package.

8.2.3 Component and Material Tests There is no periodic test or replacement schedule for package components or materials.

Routine use of the package may cause fatigue or damage to the off-the-shelf components. This damage would most likely occur to the socket head cap screws or gasket. The socket head cap screws are used with the cover and gasket to close the package. The screws, which are tightened and torqued, will occasionally round-out or have thread galling issues and require replacement.

The gasket material will occasionally tear from contact with one of the screws used to close the package and require replacement. Damage to these components will likely be identified during receipt, loading, and unloading.

8.2.4 Thermal Tests There are no systems or components which would necessitate a periodic thermal test to ensure continued performance of the package.

8.2.5 Miscellaneous Tests There are no periodic miscellaneous tests currently identified for the package or components.

8-2

9 References Busch, R. D., Bowman, S. M. "The KENO V.a Primer," in proc. American Nuclear Society 2003 Annual Meeting "The Nuclear Technology Expansion: Unlimited Opportunities", June 1-5, 2003, San Diego, California.

Koudelka, A.J., to Baldwin, M.N. 1991a. Four NBSR Elements in Infinite Sea of Water.

Memorandum. Babcock & Wilcox.

Koudelka, A.J., to Baldwin, M.N. 1991b. Seven NBSR Elements in Infinite Sea of Water.

Memorandum. Babcock & Wilcox.

NIST. 1990a. Description of the N.I.S.T. Shipping Cask.

NIST. 1990b. NIST "ST" Series Shipping Container Specification.

NIST. 1990c. NIST "ST" Series Shipping Container-Testing for Normal Conditions of Transport.

2 pages.

NIST. 1990d. Package Type 7A Test Protocol.

NIST. 1990e. NIST "ST" Series Shipping Container Subcritical Analysis (10 CFR 71.61).

"SCALE - A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design,"

Oak Ridge National Laboratory, accessed September 23, 2011,http://scale.ornl.gov/index.shtml.

Slab a ck, Les. 1990a. "Certificate of Compliance". Memorandum. NIST.

Slaback, Les. 1990b. "DOT Type 7A Test of NIST "ST" Series Package". Memorandum. NIST. 2 pages.

Sturrock, Jack. 1990a. "Shipping Container Model ST Series" [Engineering Drawing]; Drawing D-04-048, Sheets 1 and 2, Revision 4. NIST, Gaithersburg, MD.

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9-2

10 Appendix Scanned copies of select documents used in the original package application are provided in the following pages.

10-1

UNITED BTATEB DEPARTMENT CF COMMERCE NI.Sr Nat:lonal tn t:lt:ut:a of Bt:andar-d and Technology Gal I IUl~lih 11'1). Mal *ylet ic1 20099 (301) 975 -6210 FTS 879-6210 FAX (301) 921-9847 February 7, 1992 Mr. Charles E. MacDonald Chief, Transportation Branch Division of Safeguards and Transportation U.S . Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. MacDonald:

The National Institute of Standards and Technology requests approval of its shipping container designated *s1* with the condition that two containers, each containing a single unirradiated NBSR fuel element, may be transported at one time in an exclusive vehicle. Enclosed are a description of the container and supporting documents.

s:;;~

v.:~=-

Chief, Reactor Radiation Division Enclosure 7

10-2

DESCRIPTION OF N.1.S.T. SHIPPING CASK (DOT 1YPE 7A PACKAGE *sr SERIES)

The N.I.S.T. Shipping Cask consists of a carbon steel welded tube, 5 1;2* outside diameter x 1/8" wall x 10* long, welded closed on one end with a 3/8" thick carbon steel plate and a 1/2" thick carbon steel flange welded to the other. This flange is tapped with eight 1/4-20 holes to receive a cover plate and 1/a* thick neoprene gasket with the same hole pattern. The attaching screws protrude through the flange and have a 1/16" diameter hole for attaching a seal. The fuel element has wooden attachments at each end that are assembled to it outside the cask. The wooden attachments are almost the same diameter as the inside diameter of the tube to restrict movement A 3/16* sponge tape is also applied on the flange end as a cushion backup. The fuel element assembly is inserted and the cover and gasket applied and bolted then properly torqued and sealed. The loaded cask weighs approximately 55 pounds. At approximately 1/2" from each end is a 6 1/2' x 1/4* thick square plate with four 3/4" diameter holes for attachment of carrying handles. These holes are also to be used for securing the cask during shipment.

The design of this shipping cask complies with the requirements of 49CFR173.411 and 173.412. Specifically 49CFR173.412(b) is met by the holes for a seal, 173.412(f) is met by the bolted closure and the non-volatile nature of the contents, 173.412(i) is met by the designed limited area opening (22 in~ and eight securing bolts, 173.4120) is met by the design detail that the tie down flange is not an integral part of the containment portion of the package.

10-3

NIST "ST' SERIES SHIPPING CONTAINER CONTAINER SPECIFICATION Model:

ST series Container Description Each container consists of a 5.5" outside diameter carbon steel tube, 70" long, with flanges welded to each end, opening at one end only via a bolted cover plate. See attached drawing for detailed engineering specifications.

Classification: fissile material packaging.

Gross weight: approximately 75 lbs, with element Containment system: Bolted closure Construction materials: see attached drawing for size and material specifications, internal element support structure, and tie down fixtures. There are no valves, vents, or coolants in this system.

Content: one unirradiated NBSR MTR-type fuel element, nominally 350 grams of ...U, not to exceed 360 grams of mu (see attached drawing)

Package Evaluation The design specification and construction controls for this package -are-as -specified in 49CFR173.24, 173.411, 173.412, 173.461 , 173.462, and 173.465. Usage and package maintenance procedures are enclosed.

Shipping Configuration Quantity: two packages per shipment, not to exceed a total quantity of 720 grams of =u.

This amount is less than the least amount of mu that can be made critical under optimum conditions of moderation, reflection, shape, and purity.

Marking, labeling, placarding, shipping papers: as required by 49CFR.

Vehicle: Transport will be via an exclusive use vehicle.

Package tie-down: Each container shall be secured to the interior structure of the transport vehicle.

10-4

I< *I UNITEO STATES DEPARTMENT OF COMMERCE\_.

National lnetltute of Standard and Technology 4*

Ga.* * "" :,1 .. ,*u. M '"" *vtt:tt 11: 12De:1u (301) 975-6210 FTS 879-6210 FAX (301) 921-9847 February 14, 1992 Mr. Charles E. MacDonald Chief, Transportation Branch Division of Safeguards and Transportation United States Nuclear Regulatory Commission Washington, D.C. 20555 Reference : Docket Number 71 -9246

Dear Mr. MacDonald:

Enclosed please find the additional requested information pertaining to criticality (10CFR71.61) , and package integrity 10CFR71 .55. Revised drawings of the container (Revision 2) are attached. This reflects the removal of the lifting and tie-down eyelets from the package design. There are no tie-down or lifting devices which are a structural part of the package in this final design .

. Michael Rowe Chief, Reactor Radiation Division Enclosure 10-5

UNITED STATES DEPARTMENT OF COMMERCE Nation ! ln tltut of Bt nd rd nd Technology (formerly Nation ! l!lu.- au of Bt ndrd)

G a . ~ ~ g . Mi,rytand 20899 October 9, 1990 MEMORANDUM FOR Record From: l..es Slaback Jls1"1., L,1S

Subject:

Certificate of Compliance Package: "ST" Series Container for NBSR Fuel Element Specification: USDoT Type 7A Date Tested: October 4-9, 1990 We herewith warrant and certify that the referenced containers have been manufactured and tested in accordance with:

10 CFR 71 49 CFR 173.411 49 CFR 173.412 49 CFR 173.415 49 CFR 173.417 49 CFR 173.461 49 CFR 173.465 Copies of relevant documentation are attached.

10-6

UNITED STATES DEPARTMENT CF COMMERCE National lnatltute of Standard and Technology (formerly Nation ! Bureau of Standard*)

G&c:.nerGbvrg, M&""yl:8~ 20099 October 9, 1990 MEMORANDUM FOR Record From: Les Slaback 9',f ~II S

Subject:

DoT Type 7A Test of NIST *sr Series Package The *sr series DoT Type 7A Radioactive Material Package is described in Enclosure 1.

This package is designed to transport a single NBSR fuel element in compliance with 49CFR173 and 10CFR71.

The tested package is identical to those that will be used for transport, and in fact is itself expected to be used for transport. Prior to the sequence of tests, the package will be loaded with a dummy element as per the loading procedure (Enclosure 2). All tests required by 49CFR 173.465 are shown in Enclosure 3. These will be performed in sequence and the results recorded on Enclosure 3. Photographs will be taken of each test to document the physical conditions of the test.

Water Test: The sprinklers will be arranged so that all sides of the package are wetted. The simulated rainfall via sprinklers shall be measured and will exceed 2* per hour. A post test time delay to allow 'soak in' is not required since this is a metal container. Following the test, the package will be examined for any deterioration or evidence of water entering the container.

Free Drop Test: The package will be dropped from a height of 4 foot onto a poured concrete pad. Because of the geometry of this package this test will be repeated such that (1) the first drop is on the closed, welded end, (2) the second drop is on the end with the removable plate, and (3) the third drop is with the package horizontal.

Prior to the 4' free drop test the package will be dropped from a height of one foot on each corner of each end (four drops per end), as per 49CFR 173.365(c)(3). The package will be examined for any damage that could affect its integrity.

10-7

Penetration Test: A 6kg, 3.2cm diamater bar will be dropped so as to impact the center of the package from a height of 1 meter. The package shall be resting on an unyielding surface, positioned horizontally as it would be during a shipment. Because the bar must impact a curved surface of the package, care must be taken that the impact is not glancing. If the bar does not substantially recoil (bounce) vertically the test will be repeated.

The test bar will be examined for any deformity at the impact point (none permitted).

The package will be examined for any damage that could affect its integrity.

Compression Test: This test will be applied in the geometry in which the package normally is positioned during shipment, that is with the longitudinal axis of the package horizontal and each end resting on an unyielding surface (concrete). The projected area of the package (5-1/2" diameter pipe, 70' long) is 2500cm 2* This requires the greater of (1) 1300 kg/m 2 x 0.25 m 2 = 325 kg (715 lbs.), or (2) 5 x 551bs. = 275 lbs.

Hence 325kg will be placed along the length of the package by placing three layers of lead bricks, each layer consisting of eight bricks (11 .9 kg/brick), plus four additional bricks for a total of 28 bricks (333kg) on top of the package. This will provide the uniform compresion loading on two sides as required by 49CFR 173.365(d). The load shall remain in place for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Upon completion of this test, the package will be examined for any damage that could affect its integrity.

Upon completion of the test sequence, the package will be opened. The dummy element will be examined for any damage that would indicate a possible failure of the package containment effectiveness. The interior of the package will be examined for any damage that could affect its integrity. Any observed damage will be documented on the testing results form.

10-8

Enclosure 3 PACKAGE TYPE 7A TEST PROTOCOL Package Tested: Shipping Container for Single NBSR Fuel Element Date of Test: /(} iI 1/7(2 Package ID:

FAIL

1. Pretest inspection (complies with drawings, no construction defects, no corrosion, no distortions)

2. Water spray test (Start time: 13: 't 0 (The waler spray test shall simulate exposure to rainfall of app<oxlmately 5 cm {2 i~~

per hour for one hour. Tlme elapsed / ;Ot; hrs. 0..pth ol watw coftec\ecl :>_l'._'. Inches.) Observation: /Jc) £..,,.,/**-, <.t cl,' De Y-f'lf.1i;,,e.:,-I;----

3. Free drop test (Time: IS-i OSJ

{Free faU drop of 1.2 meters (4 feet) onto a flat, horizontal, unyielding surface .)

4. Penetration test (Time: 1:r:1s- )

(One meter (3.3 feet) vertical drop of panetration test bar onto the center of tht weakest part ol packag ing specimen.) Observation: c_>, ,"I'/ n l ;,u,..d p,,ric-ull {l.,,$rr4~*-, /~~

~-,1t...1,,. , ~ C'~*.,t~,~

5. Compression test (Start time: ocr:ou) / &1:,-/f'C S'kfr.-'tl fl':'--iCl,1.-r' (Compression test shaU last 101 at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and consist of 325 kg. of lead bricl~~'11e-, . .e,e,c~~~

Print Name '
£o1:~
Tests Observed By:
? Signature 10-9
NIST *sr SERIES SHIPPING CONTAINER TESTING FOR NORMAL CONDITIONS OF TRANSPORT The "ST' series Fissile Material Package is designed to transport a single NBSR fuel element in compliance with 49CFR173 and 10CFR71 . This package was tested to demonstrate its integrity for normal conditions of transport.
The tested package is identical to those that will be used for transport. Prior to the sequence of tests, the package was loaded with a dummy element. The dummy element is identical to a fueled element except for the absence of the fueled portion of the internal plates. All tests required by 10CFR71.71 were performed in sequence as outlined below.
Photographs were taken to document the physical conditions of the test.
Water Test: The sprinklers were arranged so that all sides of the package are wetted.
The simulated rainfall via sprinklers was measured and exceeded 2* per hour. A post test time delay to allow "soak in" was not required since this is a metal container. Following the test, the package was examined for any deterioration or evidence of water entering the container. No water entered the container.
Free Drop Test: The package was dropped from a height of 4 feet onto a poured concrete pad. Because of the geometry of this package this test was repeated such that (1) the first drop was on the closed, welded end, (2) the second drop was on the end with the removable plate, and (3) the third drop was with the package horizontal.
Prior to the 4' free drop test the package was dropped from a height of one foot on each corner of each end (four drops per end). The package was examined for any damage that could affect its integrity. No damage was found.
Penetration Test: A 6kg, 3.2cm diameter bar was dropped so as to impact the center of the package from a height of 1 meter. The package was resting on an unyielding surface, positioned horizontally as it would be during a shipment. Because the bar must impact a curved surface of the package, care was taken that the impact was not glancing.
The test bar was examined for any deformity at the impact point; none was found. The package was examined for any damage that could affect its integrity. No damage was observed.
Compression Test: This test was applied in the geometry in which the package normally is positioned during shipment, that is with the longitudinal axis of the package horizontal and each end resting on an unyielding surface (concrete). The projected area of the package (5-1/2" diameter pipe, 70" long) is 2500cm2* This requires the greater of (1) 1300 kg/m 2 x: 0.25 m* = 325 kg (715 lbs.), or (2) 5 x 551bs. = 275 lbs. Hence 325kg was placed along the length of the package by placing three layers of lead bricks, each layer consisting of eight bricks (11.9 kg/brick), plus four additional bricks for a total of 28 bricks (333kg) on top of the package. This provided the uniform compression loading on two sides as required by 10CFR71.71. The load remained in place for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
10-10
Upon completion of this test, the package was examined for any damage that could affect its integrity. No damage was observed.
Upon completion of the test sequence, the package was opened. The dummy element was examined for any damage that would indicate a possible failure of the package containment effectiveness. Toe interior of the package was examined for any damage that could affect its integrity. Other than chipped paint on the exterior of the package no effects of the testing were observed on the package. The dummy fuel element also showed no effects from the testing. It was not distorted or marred in any way. The package clearly met all the performance requirements of 10CFR71.57(d).
10-11
NIST ' ST' SERIES SHIPPING CONTAINER SUBCRITICAL ANALYSIS (1 OCFR71.61)
Criteria: Twice this number of packages would be subcritical.
The attached report of a criticality analysis by the fabricator of the NBSR fuel, Babcock and Wilcox, demonstrates that up to seven undamaged NBSR fuel elements, arranged in any undamaged configuration, would be subcritical for any moderation and reflection geometry. No credit is taken for the poison effect of the steel in the package, nor for the spacing between packages.
Criteria: lhis number of packages would be subcritical if stacked together in any arrangement with optimal reflection and moderation.
Two (2) packages contain no more than 720 grams. Criticality cannot be achieved under any condition since this is less than the smallest mass of '"'U required to achieve criticality.
10-12
Babcock & Wilcox Naval Nuclear a McD~rmon company Fuel Division To I A. J. Koudelka - NNFD-lSA From Filt No.
M. N. Baldwin - NNFO-lSA or Ref. MNB91-04 Subj.
FOUR NBSR ELEMENTS IN INFINITE SEA OF WATER 1/31/91 J
REFERENCE l: MEMO FROM J. W. HARWELL TO B. O. KIDD TITLED "NUCLEAR CRITICALITY SAFETY EVALUATION OF RIFLE RACXS DI ml! RTRFE I\Rl!A 'IO INCLUDE ADDITIONAL ELEMENT TYPES,* MARCH 28, 1989.
As you requested in respoDse to a request from the National Institute of Standards and Technology, I have detemined the upper limit of K-eff for four fully flooded NBSR fuel eleJDents arranged 2x2 on a uniform square pitch in an infinite sea of water. The evaluation shows that the maximulll K-eff value does not exceed 0 . 734.
For this evaluation, the basi c NBSR element employed in Reference 1, and modeled in KENO Va was used. Each standard NBSR element was IDOdeled explicitly to contain 17 fuel plates in the top and bottom fueled region, making a total of 34 fuel plates. The 2.436- Inch by 11.37-Inch by 0.020- Inch fueled portion of each plate is uniformly loaded with a matrix of 93\ enriched U02 and allllllinum to give a total plate loading of 10.294 Grams U- 235. Clad thickness is 0.01525 Inches and water gap spacing is 0.116 Inches. A 6.00 Inch long water-filled center section separates the top and bottam fueled regions of the element. Total U-235 loading for the element is 350 grams.
The coaputer code KENO Va and the 16-group Master Library from Scale-3 processed thru BONAM! were used for the calculations. This was acC011plished through the use of the Scale-3 control module CSAS25. Scale-3 is a modular code system for performing standardized con,puter analyses for licensing evaluations. It was prepared for the U. S. Nuclear Regulatory Conaission by Oak Ridge National Laboratory. The CSAS25 control module has been bencmarked against numerous known-critical systems by Babcock & Wilcox Coq>any (B&II) in addition to many other organizations. Since B&II benchmark and validation work shows that this control module (with various option restrictions which B&W imposes on criticality safety calculations) never underestimates the actual K- eff value of a system by more than 2\, and since a statistical uncertainty is always associated with a ICRNO Va calculation, a bias value of 0 . 02 plus two-sigma is always added to the calculated value when c r iticality safety is the considerati on.
Five calculations, representing vari ations in the element spacing were made.
The results presented in Table 1 include the two-sigma uncertainty and the 0.02 bias. K-eff is at a maximum when the element separation is about 0.5 CID.
10-13
KNB91-04 1/31/91 Although this evaluation uses the same calculational techniques, codes, bias value and cross-section library as would an internal criticality safety evaluation, and although the writer is confident that the X-eff quoted is conservative for safety considerations, and although it has been independently reviewed by another criticality safety engineer who has included his QA statement; this memo does not address many items that would be required by our evaluation procedures (S11ch as review of required procedures, double contingency evaluation, posting requirements, etc.). This mmm> is not intended to constitute~ criticality safety evaluation as defined by our procedures, and must not be subject to audit as a critica1ity safety evaluation. It is rather, what our customer requested: a determination of the upper limit of X-eff for four fully flooded HBSR fuel elements arranged 2x2 on a uniform square pitch in an infinite sea of water.
M. N. Baldwin QA statement:
I have reviewed these calculations and concur with the model, the codes used, the calculational techniques, the cross-section library, the results and conclusions. I further concur that this evaluation is not and is not intended to be a criticality safety evaluation as defined by NNFD procedures.
~~
J . W. Harwell cc: FM Alcorn, NNFD-lSA JJ Bazley, NNFD-15A AB Croft, NNFD-15A RL Dunham, NNFD-lSA JW Harwell, NNFD-15A BO Xidd, NHFD-lSA TD Lee, NNFD-35 RB Park, NNFD-lSA LL Wetzel, NNFD-lSA 10-14
ftBLK 1 - RESULTS OF CSAS25 RUNS -
fOOR 1185 ELl!Jll!ll'l'S at sgoARB PITCH DI All lJIYilll"IE SU OF DDJl ELl!HENT UPPER LDUT*
RUN ID SEPARATION OF K-KFF NBSG 0,0 CM. 0.708 NBSl:I 0.5 CM. 0.734 NBSI 1.0 CM. 0 . 726 NBSJ 3.0 CM. 0.590 NBSK 30.0 CM. 0.412
~~calc. ~-:~ff __+ _two-sigma _+ 0.02 . . .~~
10-15
N ,J*l Babcock & Wilcox Naval Nuclear a McDermoa Comp¥1y Fuel Division To I A. J, Koudelka, NNFD-46 From Fll1No.
M. N. Baldwin, NNFD-46 or Atf. MNB9l-08 Subj.
SEVEN NBSR ELEMENTS I N INFINITE SEA OF WATER APRIL 26, 1991 I
Reference 1: MEMO FROMM N BALDWIN TO A J KOUDELKA TITLED "FOUR NBSR ELEMENTS IN INFINITE SEA OF WATER",
JANUARY 31, 1991 ,
Reference 2: MEMO FROM J W HARWELL TO B O KIDD TITLED "NUCLEAR CRITICALITY SAFETY EVALUATION OF RIFLE RACKS IN THE RTRFE AREA TO INCLUDE ADDITIONAL ELEMENT TYPES", MARCH 28, 1989 ,
In response to a request to you trom the National Institute of standards and Technology, I deter~ined and reported in Reference 1, the upper limit of ~-eff for four fully flooded NBSR fuel elements arranged in their most reactive con1'.iguration in an infinite sea of water. The evaluation showed that the maximu.m K-eff value for four elements doea not exceed 0.734.
Recently, a second request was received fron the National Institute of Standards and Technology for the upper limit on K-eff for seven fully flooded and reflected elements. This second eva l uation shows that the maximWI K-eft value for seven elements does not exceed 0.881.
The methods used are basically the same as previously reported, but descriptions are repeated herein tor the convience of the reader.
For this evaluation, the basic NBSR element employed in Reference 2, and modeled in KENO Va was used. Each standard NBSR element was modeled explicitly to contain 17 fuel plates in the top and bottom fueled region, maleing a total of 34 fuel plates.
The 2,436-inch by 11,37-inch by 0.020-inch fueled portion of each plate is uniformly loaded with a matrix of 93t enriched 002 and aluminum to give a total plate loading of 10.294 Gm U*235. Clad thickness is 0,01525 inches and water gap spacing is 0,116 inches. A 6,00 inch long water-filled canter section separates the top and bottom fueled regions of the element. Total U-235 loading for the element is 350 grams .
UNCLASSIFIED a;;;J.~
JP/rL Dllt 10-16
The computer code KENO Va and the 16-group Master Library from Scale-3 processed through BONAM.I were used for the calculations. This was accomplished through the use of th*
scale-3 control module CSAS25. Scale-3 is a modular code system tor perfonning standar~ized computer analyses for licensing evaluations. It was prepared for the u. s. Nuclear Regulatory Commission by Oak Ridge National Laboratory. The CSAS25 control module has been benchmarked against nUJ11erous known-critical systems by Babcock & Wilcox Company (B&W) in addition to many other organizations. Since B&W benchmark and validation work shows that this control module (with various option rest~ictions which B&W imposes on criticality safety calculations) never underestimates the actual K-eff value of a system by more than 21, and since a statistical uncertainty is always associated with a KENO Va calculation, a bias value of 0.02 plus two-sigma is always added to the calculated value when criticality safety is the consideration.
Five calculations, representing variations in the element spacing were made. The results presented in Table 1 include the two-sigma uncertainty and the 0.02 bias.
Although this evaluation uses the same calculational techniques, codes, bias value and cross-section library as would an internal criticality safety evaluation, and although the writer is confident that the 1<-eff quoted is co.n servative for safety considerations, and although it has been independently reviewed by another criticality safety engineer who has included his QA statement; this memo does not address many items that would be required by our evaluation procedures (such as review of required procedures, double contingency evaluation, posting requirements, etc.). This memo is not intended to constitute a criticality safety evaluation as defined by our procedures, and must not be subject to audit as a criticality safety evaluation.
It is rather, what our customer requested: a determination of the upper limit of k-eff for seven fully flooded NBSR fuel elements in an infinite sea of water.
)n. )7. /J~--.:..
M. N. BALDWIN QA stateltlent:
I have reviewed these calculations and concur with the model, the codes used, the calculational techniques, the cross-section library, the results and concl~sions. I further concur that this evaluation is not and is not intended to be a criticality safety evaluation as defined by NNFD procedures.
~~
10-17
TABLE l - RESULTS OF CSAS25 RUNS SEVEN NBS Et.EMENTS IN AN INFINITE SEA OF WATER ELEMENT UPPER LIMIT RUN ro SEPARATION OP K-EFF NBSB 0.o CM. 0.849 NBSC 0.5 CH. 0.876 NBSD 1.0 CM. 0.881 NBSEE 1.5 CM 0.874 NBSFF 2.0 CM 0.859
~~calc. x-ett +two-sigma + 0.02 TOTA.. P.05 10-18~~
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10-19

ML22039A080

Text

Dear Mr. MacDonald:

Dear Mr. MacDonald:

Subject:

Subject:

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