ML023080337
| ML023080337 | |
| Person / Time | |
|---|---|
| Site: | 07103036 |
| Issue date: | 01/29/1999 |
| From: | Chappell C NRC/NMSS/SFPO |
| To: | Boyle R US Dept of Transportation, Radioactive Materials Branch |
| References | |
| -RFPFR NUDOCS 9902030145 | |
| Download: ML023080337 (9) | |
Text
UNITED STATES
-0 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 29,' 1999 Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 400 Seventh Street, S.W.
Washington, D.C. 20590
SUBJECT:
MODEL NO. JRF-90Y-950K
Dear Mr. Boyle:
This is in response to your letter dated November 14, 1996, as supplemented July 29, 1997, July 23 and December 22, 1998, requesting our assistance in evaluating the Model No. JRF-90Y-950K package, for contents shown on the enclosed Table 1, authorized by the Japanese Certificate of Approval No. J/1 19/B(U)F-85, Rev. 1.
Based upon our review of the statements and representations in the application, and for the reasons stated in the enclosed Evaluation Record, we recommend revalidation of Japanese Certificate of Approval No. J/1 19/B(U)F-85, Rev. 1 for the Model No.
JRF-90Y-950K package for contents shown on Table 1, with a minimum transport index of 0.0 for criticality control.
Sincerely, Cass R. Chappell, t Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No.:
Enclosures:
As Stated 9902030145 990129 PDR ADOCK 07103036 c
January 29.
1999 Mr. Richard W. Boyle Radioactive Materials Branch US. Department of Transportation 400 Seventh Street, S.W.
Washington, D.C. 20590
SUBJECT:
MODEL NO. JRF-90Y-950K
Dear Mr. Boyle:
This is in response to your letter dated November 14, 1996, as supplemented July 29, 1997, July 23, and December 22, 1998, requesting our assistance in evaluating the Model No. JRF-90Y-950K package, for contents shown on the enclosed Table 1, authorized by the Japanese Certificate of Approval No. J/1 19/B(U)F-85, Rev. 1.
Based upon our review of the statements and representations in the application, and for the reasons stated in the enclosed Evaluation Record, we recommend revalidation of Japanese Certificate of Approval No. J/1 19/B(U)F-85, Rev. 1 for the Model No.
JRF-90Y-950K package for contents shown on Table 1, with a minimum transport index of 0.0 for criticality control.
Sincerely, Original Signed by:
Cass R. Chappell, Chief Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No.:
Enclosures:
As Stated Distribution (w/encl):
SFPO r/f NMSS r/f NRC PDR NRC File Center ANorris, NMSS/PMDA(Closes 07003036010S, TAC L22366) r
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20665-0001 EVALUATION RECORD Model No. JRF-90Y-950K Package Japanese Certificate of Approval J/1 19/B(U)F-85 Revision No. 1 By application dated November 14, 1996, as supplemented July 29, 1997, July 23, and December 22, 1998, the Department of Transportation (DOT) requested assistance in evaluating the Model No. JRF-90Y-950K package which is authorized by Japanese Certificate of Approval No. J/1 19/B(U)F-85, Revision No. 1. The DOT provided the Competent Authority of Japan Certificate and the Safety Analysis Report for the package.
Based on the statements and representations in the application, as supplemented, we recommend revalidation of the Japanese Certificate of approval for the contents shown in Table 1 for the Model No. JRF-90Y-950K package.
PACKAGING The Model No. JRF-90Y-950K package is a freslh fuel package for transporting JRR-2, JRR-3, JRR-4, JMTR and JMTRC research reactor fuel. The fuel to be transported can be low-enriched, medium-enriched, or high-enrichedi The package is cylindrical in shape and consists of an inner stainless steel containment vessel and an outer stainless steel vessel.
The region between the inner and outer shells is filled with polyurethane foam. The package is equipped with a stainless steel fuel basket for either box type fuel or cylindrical type fuel. The inner containment vessel has a design pressure of 0.0981 MPa Gage (14.2 psig). The approximate dimensions and weights of the package are as follows:
Outside dimensions of packaging Height 1559 mm Diameter 840 mm Inner containment vessel Height 1310 mm Diameter 460 mm Number of assemblies per package box type 10 cylindrical 7
Maximum weight of fuel elements 92 kg Total weight of package 950 kg The outer shell is 3 mm thick stainless steel with a 6 mm thick stainless steel bottom plate.
The inner shell is 10 mm thick stainless steel with a 35 mm thick bottom plate. The space between the inner and outer shells is filled with,igid polyurethane foam which serves as a thermal insulator. There are eight fusible plugs in the outer shell which are provided to relieve pressure from combustion gases during a fire.
The outer lid and the bottom of the
" ENCLOSURE
packag6 contain balsa wood, which is used as an impact limiting material.
I The package is closed by a 55 mm thick stainless steel inner lid and 16 bolts. The inner lid is sealed with double silicone rubber O-rings. The outer lid is a cylindrical dome with a height of 398 mm and a diameter of 840 mm and is bolted to the main body of the package.
DRAWINGS The package is constructed in accordance with Kobe Steel, Ltd. Drawing Nos.
NXP-038-S04, Rev. 1, NXP-038-S06, Rev. 1, and NXP-038-S07, Rev. 1.
STRUCTURAL The ability of the package to meet the requirements of IAEA Safety Series 6, 1985 edition, (As Amended 1990) for structural integrity was demonstrated by analyses and prototype testing. The design standards for the packaging are based on the Public Notification No. 5 of Japan Science and Technology Agency, and Section III, subsection NB of ASME code.
The applicant analyzed the package for the normal condition 1.2 meter drop test using textbook formulas and the CASH-Il computer code which calculates deformation and accelerations. The applicant considered drops in the following orientations: 1) horizontal,
- 2) vertical (lid side and bottom side), 3) corner drop (lid side and bottom side), and 4) inclined drop (lid side and bottom side). The calculated stresses in the package were within allowable limits.
The applicant analyzed the package for the vibration, stacking and penetration tests. The calculated stresses in the package were within allowable limits.
The applicant calculated stresses in the lifting eye-Rlates of the package using textbook formulas. The calculated stresses were within allowable limits.
The full scale packaging was subjected to a 9 meter drop test, a 1 meter puncture test, and a thermal test. The drop orientations selected were: (1) side drop, (2) oblique drop at 1350, and (3) bottom end drop. The drop tests results indicated that the outer packaging and its lid had minor deformations, and that no deformation was observed in the containment vessel and its lid.
The containment vessel was leak tested before and after the drop tests. The test results indicated that the integrity of the containment system was not impaired.
Based upon review of the structural design criteria with the analytical results and the prototype testing results shown in the application, the staff has found that the Japan Atomic Energy Research Institute has adequately demonstrated the structural integrity of the packaging. The applicant concluded and the staff agrees that the package meets the requirements of IAEA Safety Series No. 6, 1985 Edition, (As Amended 1990).
THERMAL EVALUATION The thermal behavior of the package was analyzed using textbook analysis for normal conditions and the TRUMP, non-steady state thermal analysis code for hypothetical accident conditions. In addition, a prototype package was subjected to a hypothetical accident conditions thermal test in a furnace.
The results of the analyses for normal conditions of transport thermal conditions show that the structural strength and containment of the package are not affected by normal test conditions.
The surface temperature is calculated to be 650C which is below the permissible value of 850C.
The maximum pressure under normal conditions is calculated to be 0.016 MPa which is below the design pressure of 0.0981 MPa. The calculated temperatures of the O-ring seals are within the normal service temperature range of the silicone rubber.
The applicant performed both an analytical evaluation and a 30-minute furnace test of the package in order to evaluate the package's performance under hypothetical accident conditions.
In the analysis, the applicant assumed that the impact limiting material was deformed prior to the thermal test. The analysis assumes natural convection and radiation heat transfer between the external surface of the package and the 800 0C heat source, an emissivity coefficient of 0.9 and a surface absorptivity coefficient emissivity of 0.8. The analysis took into account the possibility of hot gasses being generated during combustion of the polyurethane foam and the burning of the balsa wood. Consequently, the calculated external surface temperatures exceed the specified boundary condition temperature of 8000C, due to the combustible gasses passing through the fusible plugs.
For the furnace test, a prototype package was subjected to the 9 meter drop and the 1 meter puncture test and was placed in 8000C furnace for 30 minutes and allowed to cool naturally.
The applicant reported that upon completion of the furnace test, flames appeared in the fusible plugs and within the cracks that developed during the 9 meter drop test. The flames were attributed to the combustion of the polyethylene foam and balsa materials. The higher external surface temperatures were due to the additional heat input from the flow of gasses out of the cracks and the fusible plugs.
The maximum temperatures measured during the thermal test were lower than those calculated using the TRUMP computer code. The maximum temperature were as follows:
Test (°C)
Analysis (0C)
Near O-ring 88.6 161.0 Inner surface of inner shell 396.2 464.1 Outer surface of fuel basket 123.3 182.5 Outer surface of packaging 1051.6 1229.7 The maximum temperature of the stainless steel shell is below the melting point of the steel.
The terrperature of the O-ring seal is below its maximum service temperature. The pressure calculated during accident conditions is 0.065 MPa which is below the design pressure of 0.0981 MPa.
The applicant concluded, and the staff agrees, that, the package meets the requirements of IAEA Safety Series 6, 1985 edition, (As Amended 1990).
CONTAINMENT The containment system of the package is the stainless steel inner vessel shell and lid. The lid is closed by 16 closure bolts, and double silicone rubber O-ring seals. The lid is equipped with a leak test port. There are no penetrations into the containment system. The applicant demonstrated by analysis and test that the structural integrity of the package containment system remained unaffected under the normal and hypothetical accident conditions in IAEA Safety Series 6, 1985 edition, (As Amended 1990). The maximum temperature of the seals under fire test conditions was calculated as 209.8 °C, which is within the operating range for silicone rubber.
Since the package can contain a Type B quantity of radioactivity, the applicant provided a containment analysis for the package, using the methodology of ANSI N14.5-77, "American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials." The maximum allowable leak rate for the package is 1.1 x 10"1 atm-cm3/s. The applicant stated that the temperatures and pressures experienced by the unirradiated, intact fuel assemblies under normal and hypothetical accident conditions will not affect the integrity of the fuel. To evaluate the containment performance of the package, the applicant assumed that the surfaces of the fuel assemblies were contaminated with fuel residue. The applicant assumed that 100 percent of this surface contamination could be released under normal and accident conditions. The applicant showed that the radioactivity that could be released from the package, assuming a leak rate of 1.1 x 10.1 atm-cm 3/s, was less than 1 x 10-8 A2/hr under normal conditions of transport. The total radioactivity present as surface, contamination does not exceed an A2 quantity.
I The applicant reported the leak rate measurements performed on the package test specimens that were subjected to normal and hypothetical accident tests. The leak rates measured, before and after the physical testing, were well within the maximum allowable leak rate of 1.1 x 10-1 atm-cm3/s.
The applicant stated that the package is leak tested prior to each shipment. The applicant specifies a pressure drop leakage test. The package provides adequate containment to meet the requirements of IAEA Safety Series 6, 1985 edition (As Amended 1990).
SHIELDING The applicant performed a radiation analysis of the package with the maximum number of unirradiated fuel elements. The source term was based on the fuel assembly type with the maximum quantity of radioactive material. For that fuel assembly type, the uranium was enriched to 19.5 weight percent U-235. The quantity of U-234 and U-236 were estimated based on previous samples from fuel during fabrication. The maximum calculated dose rate was 0.008 mSv/hr (0.8 mrem/hr) at the package surface, which included gamma rays and neutrons from spontaneous fission. The dose rates calculated by the applicant were well within the limits of IAEA Safety Series 6, 1985 edition, (As Amended 1990).
CONTENTS AND TRANSPORT INDEX FOR CRITICALITY CONTROL The contents for the package are found on the enclosed table 1.
Minimum Transport Index to be Shown on the Label for Nuclear Criticality Control:
0.0 CRITICALITY EVALUATION
The applicant performed a criticality analysis which shows that the package remains subcritical under normal conditions of transport and hypothetical accident conditions.
For the analysis the applicant used the KENO V.a. computer code with a 137 group MGCL cross section library. The applicant assumed that k-eff for the array of damaged packages bounds the k-eff for a single package and an array of undamaged packages. The package is designed to transport up to 10 research reactor fuel elements. The fuel elements have three specified enrichments, 20, 45 and 93 weight percent U-235. The applicant modeled an infinite array of triangular-pitched packages. The applicant conservatively ignored all of the materials and the spacing provided by the materials outside the stainless steel inner shell (i.e., the polyurethane foam heat insulator and the stainless steel outer shell), in the model for hypothetical accident conditions. Ignoring the spacing due to the foam and outer shell reduced the diameter of the package from 84 cm to 48 cm.
The applicant calculated the k-eff of a fully flooded array of damaged package. The analysis considered each type of fuel element, excluding the fuel follower elements since they contain less U-235 than the standard fuel elements of similar design. The applicant determined that the JRR-3 Uranium-Silicon-Aluminum Dispersion Alloy fuel was the most reactive. The k-eff for this fuel element was 0.902 + 0.005. The applicant then simultaneously varied the water density inside and outside the package to find optimum moderation. Optimum moderation was found with the water density equal to 0.02 g/cm 3. The k-eff was 0.9305+/-0.0026.
The applicant performed a benchmarking analysis of the KENO V.a computer code and the 137 group MGCL cross section library. The applicant performed a k-eff calculation for three critical benchmark experiments taken from the "International Criticality Safety Benchmark Evaluation Project (ICSBEP)". The benchmark experiments are of the SPERT reactor which is 93 weight percent enriched fuel plates with a geometry similar to typical MTR fuel assemblies. The smallest value of k-eff calculated by the applicant was 0.98865:t 0.00141. The bias was then determined to be the sum of the difference between the k-eff calculation and unity plus twice the Monte Carlo uncertainty in the benchmark calculation. The bias calculated by the applicant was 0.0142. Although the determination of the bias did not use a large number of critical experiments (greater than 10) or follow a method of determining the bias in NUREG/CR-6361, "Criticality Benchmark Guide for Light-Water Reactor Fuel in Transportation and Storage Table 1, Specification of Contents, Model No. JRF-90Y-950K Fuel Element JRR-3 JRR-3 JRR-4 JRR-4 JRR-4 JMTR Standard JMTR Follower type JRR-2 standard Follower type type B-type L-type JMTRC Standard JMTRC Follower type Cylindrical Gross weight of U 16.12 24.81 10.49 15.86 1.83 11.77 10.75 7.28 21.74 4.78 14.33 3.584 (kg -U/package) or less or less or less or less or less or less or less or less or less or less or less or less T-otaLActLvity-of.
Contents 5.97 9.14 3.89 5.89 5.01 4.36 4.04 6.47 8.05 4.25 5.31 3.185 (GBq/package) or less or less or less or less or less or less or less or less or less or less or less or less Fuel Uranium Uranium Uranium Uranium Uranium Uranium Uranium.
Uranium Uranium Uranium Uranium Uraniut Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Dispersal Dispersal Dispersal Dispersal Dispersal Dispersal Alloy Silicon Alloy Silicon Alloy Alloy Silicon Alloy Silicon Alloy Silicon Alloy Dispersal Dispersal Dispersal Dispersal Dispersal Alloy Alloy Alloy Alloy Alloy Clad, Side Plate, etc Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Alloy Alloy Alloy Alloy Alloy Alloy Alloy Alloy Alloy Alloy Alloy Alloy Burnable Absorber Cadmium Cadmium Cadmium Cadmium Wire Wire Wire Wire (wt%)
19.95 19.95 19.95 19.95 93.3 19.95 19.95 46 19.95 46 19.95 46 or less or less or less or less or less or less or less or less or less or less or less or less Burnup (%)
0 0
0 0
0 0
0 0
0 0
0 0
Total heat (W/Package) 0 0
0 0
0 0
0 0
0 0
0 0
Cooling Time (days) 0 0
0 0
0 0
0 0
0 0
0 0
Gross Weight of Contents 9.2 9.2 6
6 6.3 7.9 6.5 7.6 8.4 5.2 5.8 7.5 (kg/element)
Number of Fuel 10 10 10 10 10 10 10 10 10 10 10 7
Elements/Package or less or less or less or less or less or less or less or less or less or less or less or less
Packages", the bias was accepted by the staff. The staff accepted the bias since the applicant used the lowest of the three benchmark k-effs to determine the bias. The model used by the applicant was clearly conservative and the criticality analysis was verified by staff calculations.
The NRC staff performed confirmatory calculations using KENO V.a and the 44-group cross section set. The staffs model is similar to the applicant's, including an infinite array of packages. The staff determined that the maximum k-eff was 0.9253 for the JRR-3 Uranium-Silicon-Aluminum Dispersion Alloy fuel and 0.8541 for the JRR-4 Uranium-Aluminum Alloy. The staff performed a calculation which determined that neglecting the spacing and the stainless steel of the outer shell increases the k-eff by approximately 6 percent. The staff's results are consistent with the applicant's. The staff agrees with the applicant's conclusion that the package meets the criticality requirements of IAEA Safety Series No. 6, 1985 edition, (As Amended 1990).
OPERATING AND MAINTENANCE PROCEDURES The application specifies procedures for loading and unloading a package. The procedures specify inspections for package damage and leak checking and the contents are proper for the package.
The application specifies a periodic maintenance schedule for the package.
CONCLUSIONS Based on our review, and the statements and representations in the application, we have concluded that the Model No. JRF-90Y-950K package meets the requirements of IAEA Safety Series No. 6, 1985 Edition, (As Amended 1990).
Cass R. Chappell, Chief Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Date 1/29/99