ML20151H445
| ML20151H445 | |
| Person / Time | |
|---|---|
| Site: | 07109181 |
| Issue date: | 03/31/1983 |
| From: | NUCLEAR PACKAGING, INC. |
| To: | |
| Shared Package | |
| ML20151H434 | List: |
| References | |
| 22167, NUDOCS 8305040110 | |
| Download: ML20151H445 (91) | |
Text
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i SAFETY ANALYSIS REPORT FOR THE NUPAC PAS-2 PACKAGING TO 10 CFR 71 TYPE "B"
PACKAGING REQUIREMENTS l.
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L PREPARED BY:
Nuclear Packaging, Inc.
815 South 28th Street Tacoma, Washington 98409 (206) 572-7775 8305040110 830425 PDR ADOCK 071*****
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NOTICE Tais Safety Analysis Report for the NuPac Model PAS-2 packaging and all associatad drawings incliding amendments thereto are the property cf sue. ear Packaging, Inc., Tacoma, Washington.
This r.aterial is being made available for the purpose of obtaining required certifications from the U.S.
Nuclear Regulato:ry Commission and to enable others to register with the U.S.N.R.C. as a user of this package.
No other use f
of this material is authorized unless by written consent i
of Nuclear Packaging, Inc.
Parties who may come into possession of this material are cautioned that the information is PROPRIETARY to the interests of Nuclear Packaging, Inc.
and is not to be reproduced in any form without the prior written consent of Nuclear Packaging, Inc.
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PROPRlHTARY DAl*A em a j
1 1.0 General Information 1-1 1.1 Introduction 1-1 1.2 Package Description 1-1 1.2.1 Packaging 1-1 1.2.2 Operational Features 1-6 1.2.3 Contents of Packaging 1-8 2.0 Structural Evaluation 2-1 2.1 Structural Design 2-1 2.1.1 Discussion 2-1 2.1.2 Design Criteria 2-1
.2. 2 Weights and Center of Gravity 2-2 2.3 Nechanical Properties of Materials 2-3 2.4 General Standards for All Packages 2-4 2.4.1 Chemical and Galvanic Reactions 2-4 2.4.2 Positive Closure 2-4 2.4.3 Lifting Devices 2-4 2.4.4 Tiedowns 2-7 2.5 Standards for Type B and Large Quantity Packaging 2-7 2.5.1 Load Resistance 2-7 2.5.2 External Pressure 2-9 2.6 Normal Conditions of Transport 2-11 2.6.1 Heat 2-11 2.6.2 Cold 2-11 2.6.3 Pressure 2-15 2.6.4 Vib ra t ion 2-16 2.6.5 Water Spray 2-16 2.6.6 Free Drop 2-16 2.6.7 Corner Drop 2-16 2.6.8 Penetration 2-16 2.6.9 Compression 2-16 2.6.10 Conclusion 2-17 i
pRoPRBETARY DATA INDEX (Con't)
PAGE 2.7 H)Pothetical Accident Conditions 2-17 2.7.1 Free Drop Events 2-18 2.7.2 Puncture 2-23 2.7.3 71ermal Analysis 2-23 2.7.4 Fater Immersion 2-23 2.7.5 Summary of Damage 2-24 2.8 Special Form 2-24 2.9 Fuel Rods 2-24 2.10 Appendix 2-24 2.10.1 General Arrangement Drawing of the NuPac PAS-2 Packaging 2-24 2.10.2 NnPac N-55 General Arrangement Drawing Certificate of Compliance No. 9070 2-26 2.10.3 Drop Test Results 2-28 3.0 Thermal Evaluations 3-1 3.1 Discussion 3-1 3.2 Summary of Thermal Properties of Materials 3-2 3.3 Technical Specifications of Components 3-4 3.4 Thermal Evaluation for Normal Conditions of Transport 3-5 3.4.1 Maximum Temperatures 3-8 3.4.2 Minimum Temperatures 3-8 3.4.3 Kazimum Internal Pressure 3-8 3.4.4 Thermal Streeves 3-9 3.5 Bypothetical Accident Thermal Evaluations 3-9 3.5.1 Maximum Temperatures 3-10 3.5.2 Maximum Pressures 3-11 3.5.3 Maximum Thermal Stresses 3-11 3.5.4 Evaluation of Package Performance for the Hypothetical Accident Thermal Conditions 3-11 11
PP20PR]HTARY DE'A INDEI (Con't)
PAGE 4.0 Containment 4-1 4.1 Containment Boundaries 4-1 4.1.1 Containment Vessels 4-1 4.1.2 Containment Penetrations 4-1 4.1.3 Seals and Welds 4-1 4.1.4 Closure 4-2 4.2 Requirements for Normal Conditions of Transport 4-2 4.2.1 Release of Radioactive Material 4-2 4.2.2 Pressurization of Containment Vessel 4-2 4.2.3 Coolant Contamination 4-3 4.2.4 Coolant Loss 4-3 4.3 Containment Requirements for the Hypothetical Accident Condition 4-3 4.3.1 Fission Gas Products 4-3 4.3.2 Release of Contents 4-3 5.0 Shielding 5-1 6.0 Criticality Evaluation 6-1 7.0 Operating Procedure 7-1 7.1 Operational Requirements (unless other-wise noted) 7-1 t
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7.2 Operational Narrative 7-3 l
8.0 Acceptance Tests and Maintenance Program 8-1 8.1 Acceptance Tests 8-1 8.2 Maintenance Program 8-1 8.3 Appendix 8-4 8.3.1 Discussion of Gamma Scan Procedure 8-4' 8.3.2 LT-09, Helium Leak Test Procedure 8-6 9.0 Quality Assurance 9-1 l
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!J$ b [h_h,Y hk APPLICATION FOR NRC CERTIFICATE OF COMPLIANCE AttrHORIZING SHIPMENT OF RADIOACTI 1 MATERIAL IN THE NUPAC PAS-2 PACKAGING 1_q GENERAL INFORMATION M Introduction The NuPac PAS-2 packaging has been developed by Nuclear Packaging, Inc. as a safe means of transporting Type B quantities of radioactive liquids from commercial reactor coolant systems to chemical analysis f acilities off-site.
These liquid coolant samples would be taken subsequent to a reactor accident, and would be used to help determine tne condition of the reactor core. Thus, the chemical and isotopic breakdown of the samples would be somewhat varied.
Authorization is sought for shipment by cargo vessel, motor v-hicle and rail.
M Packare Descrintion 1.2.1 Packmains 1.2.1,1 General pescrintion The NuPac PAS-2 packaging consists of a shielded sample cask designed to interface with a post accident sampling system (PASS), a secondary contain-ment vessel to provide redundant containment, and inner and outer overpack systems surrounding the secondary containment to provide insulating and shock absorbing capabilitics.
The inner overpack consists of a foam-lined standard 17H 55 gallon drum.
The outer overpack is a standard N-55, designed by Nuclear Packaging, Inc., Certificate of Compliance No. 9070.
The overpack system is designed to protect the sample cask from the effects of normal transport and hypothetical accident conditions.
1-1
MOPRBTARY DATA 1.2.1.2 Naterials 91 Construction. Dimensions. M Fabricatina Methods General Arrangement Drawings of the NuPac PAS-2 packaging system are included in Appendiz 2.10.1.
They show overall dimensions of the N-55 outer overpack, the foam-lined 55 gallon drum inner overpack, the sample shield and the sample vial.
4 The N-55 outer overpack is f abricated of ductile low carbon galvanized steel in the outer shell, and high impact fiberglass in the inner shell.
The volume between the inner and outer shells of the N-55 is filled with thermal and shock insulating material consisting of rigid polyurethane foam having a density of approximately three pounds per cubic foot. The N-55 has a 32 inch diameter and is 48 inches in height.
The insulating material is poured into the cavity between the two shells l
and allowed to expand, completely filling the void.
Here it bonds to the shells creating a unitized construction for the packaging. Mechanical properties of these materials are further described in Section 2.0, below.
The inner overpack consists of a standard 55 gallon drum (DOT 17H) lined with dense (ten pounds per cubic foot) rigid polyurethane foam cut to closely fit the dimensions of the secondary contaissent vessel and the I
inside of the 55 gallon drum.
Ile secondary containment vessel provides a redundant level of containment
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for the NuPac PAS-2 package.
It is fabricated from ASTN 516 grade 70 steel, which has excellent low temperature properties. The vessel consists of a rolled 3/8 in. plate body with a 3/8 in, plate welded across the botton; and closed with a close fitting 0-ring seal between the body and the lid.
The lid is provided with a test port for use in testing the seal integrity of the secondary containment vessel.
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peoPRFETARY DATA The sample shield consists of steel inner and outer shells with the cav;ity between the shells filled with lead. The cask closure l a e f f e c t e d ley a
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lead-filled steel plug.' There are four penetrations.through the cask walls; an inlet and outle't; port for interf acing with the PASS and valve actuator penetrationssto al[igw tl.e user to operate the inlet and outlet
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valves from outside the radiatien shield.
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All radioact ive material is c ut[ained within a short section of tubing between the inlet and outlet valvas.
This tubing, together with the valve plugs on each end, is the sample vial, constituting the-primary containment bound a ry of t he NuPa c PhS-2,p ack ag e.
The void space around the valve and tubing assembly is packed with a mixture of lead shot and vermiculite as a -
means of providing additiona'l shielding and absorbing any leakage cecuring from f ailure of primary containment. The vial holds approxima tely-50 milliliters of coolant.
s 1.2.1.3 Containment Vessel The overpacks are not. intended to be cont ainment' vessels. Their prime function is to reduce the severity of the hypothetical accident conditions.
However, the NuPac PAS-2 packaging is provided with several levels of containment and confinement to assure that no contaminated material may ercape to the environment.
The cample v.ial,itself constitutes the primary containment of the radioactive materials. This report will demonstrate that under all defined normal and accident events, this sample vial will remain intact with primary containmqnt unviolated. However, should the vial fail in some fashion, two furthe'r boundaries would Love to be breached s
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before any release of radionuclides to th~e environment could occur. The first redundant boundary is the cavity lining of the samp1'e shield. All penetrations for valve operators and plugs into this region have high 4
quality 0-ring seals. Although these seals are not testable. the seals are i
i designed to confine liquid within the cavity.
The vermiculite included in 8
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f the cavity will fully absorb the spillags.
Outside of the sample shield, the secondary containment vessel provides a fully leak-tight, testable containment boundary to fully insure that under no credible circumstances will any contaminated material escape to the environment.
1.2.1.4 Neutron Absorbers There are no sources of neutron emmissions and therefore, there are no materials used as neutron absorbers or moderators in the NuPac PAS-2 packaging.
1.2.1.5 Packame Weiaht Gross weight of the package is approximately 2400 pounds.-
1.2.1.6 Pecentacles There are no receptacles on the outside of the NuPac PAS-2 packaging. A test port receptacle is located on the lid of the secondary containment vessel. The port is sealed with a bolt head 0-ring (Parker Stat-0-Seal) seal and covered with a stainless steel cover. The sample vial has two receptacles which are plugged wj th face-sealed stainless plugs extending from the vial through the side of The simple shield.
1.2.1.8 Tiedowns Tiedowns are not a structural part of the package. The N-55 lugs are not required for this application and therefore are covered at all times.
The package is secured to a dedicated use pallet by means of a frame placed over but not attached to the NuPac PAS-2 package assembly.
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-=ue 1.2.1.9 Liftina Devices i
Lifting devices are present in various parts of the NuPac PAS-2 packaging.
The N-55 outer overpack is equipped with four lif ting lugs which have been qualified for lif ting the 750 pound payload o.f the N-55, Certificate of Compliance No. 9070.
Since these lugs are not te be used to lift the weight of the sample shield, they will be covered during use by two fender washers held together by a small bolt.
The secondary containment vessel is equipped with a lifting bail on its lid for handling the lid only. Finally, the sample shield closure plus is fitted with a second lif ting bail, used to lif t both the plug and the
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Ioaded shield through the plug closure bolts.
Refer to Section 2.4.3 for a detailed st=uctural analysis of this device.
.1 2.1.10 Pressure Relief System There are no pressure relief valves in the NuPac PAS-2 packaging.
1.2.1.11 Hg,31 Dissination The NuPac PAS-2 package will safely transport 3 watts of internal heat.
'1.2.1.12 Coolants The NuPac PAS-2 packaging does not involve the use of coolants.
1.2.1.13 Protrusions There are no outer or inner protrusions on the NaPac PAS-2 packaging.
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1.2.1.14 Shieldina Shielding on the NuPac PAS-2 cask is provided by the 6 in. thick lead-filled sample shield, as well as lead shot and specially fitted lead pieces inside the sample shield.
1.2.2 Onerational Features l
The NuPac PAS-2 packaging involves numercus features important to the safe operation and transport of radioactive material.
Figure 1.2.2 shows a schematic drawing of the system showing all containment systess, seals and containment penetrations in the packaging system.
Note that the shield is penetrated in four places:
twice for the valve operators and twice for the sample vini port plugs. Close-toleranced bore seals are used in each of these penetrations to retain any spillage from the sample vial safely within the shield.
The secondary containment vessel envelopes the shield and completely prevents any release of payload to the environment. The vessel lid is sealed with a bore-seal 0-ring.
This lid is equipped with a test port such that both the lid bore seal as well as the test port seal may be tested before the package is allowed to be shipped, l
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TEST PORT H-55 55-GALLON DRUM
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7 STAT-O-SEAL
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FACE SEAL
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LID BORE SEAL N
SECONDARY x
CONTAINMENT VESSEE VALVE OPERATOR N
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EORE SEAL N
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BORE SEAL Q,y FACE SEAL s\\
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VALVE PLUG
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VALVE BODY
- SAMPLE SHIELD PRIMARY CONTAINMENil (SAMPLE VIAL)
FIGURE 1.2.2 SCHEMATIC, NUPAC PAS-2 PACKAGING l-7
Surrounding the secondary centainment are two foam-filled shells which are designed te mitigate the effects of the hypothetical accident events.
The inner shell is simply a standard DOT 17H 55 gallon drum with specially cut foam between the drum and the secondary containment vessel.
The outer-shell is identical to standard N-55 packaging, Certificate of Compliance Number 9070.
1.2.3 Cputents 21 Packmaina The PaPac PAS-2 system is designed to transport a 50 cc sample of reactor coolant taken from the post accident sampling system. Maximum activity of this sample is estimated to be 5 curies /cc or 250 curies total.
Relative concentrations of isotopes present are given in Table 1.2.3-1.
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'.a-j TABLE 1.2.3-1 NUPAC PAS-2 MAXIMUM PAYLOAD CONCENTRATIONS (CURIES /CC)
(AT START OF ACCIDENT)
Nuclide Nuclide Nuclide BR-84 4.59E-02 RU-106 2.64E-04 IE-135 1.48E-01 BR-85 6.43E-02 TE-129M 3.44E-04 CS-135 3.69E-10 KR-85M 1.27E-01 TE-129 1.05E-03 CS-136 5.72E-06 l
KR-85 2.08E-03 TC-99M 7.29E-04 IE-137 5.78E-01 KR-87 2.39E-01 I-129 3.34E-09 CS-137 1.33E-04 KR-88 3.50E-01 I-131 1.44E-01 IE-138 5.75E-01 RB-88 3.53E-03 IE-131M 1.98E-03 CS-138 6.55E-03 KR-89 4.54E-03 TE-132 4.27E-03 CS-140 5.81E-03 RB-89 4.69E-03 I-132 2.14E-01 LA-140 6.26E-03 Sk-89 4.66E-03 TE-133M 3.443-03 BA-143 5.14E-03 SR-90 2.96E-04 TE-133 3.63E-03 LA-143 5.81E-03 Y-90 2.96E-04 I-133 3.31E-01 CE-143 5.81E-03 SR-91 5.72E-03 IE-133 6.33E-01 PR-143 5.81E-03 Y-91M 3.37E-03 CS-134 3.31E-05 CE-144 4.05E-03 Y-91 5.75E-03 TE-134 6.81E-03 PR-144 4.05E-03 NB-95 6.10E-03 I-134 3.85E-01 ZR-95 6.01E-03 MO-99 6.07E-03 I-135 3.03E-01 RU-103 2.95E-03 IE-135M 1.77E-01 i
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,~~s M STRUL N EVALUATION This Section identifies and describes the principle structural design o f i
the packaging, components, and systems important to safety and to compliance with the ' performance requirements of 10 CFR 71.
M Structural Desian 2,1,1 Discussion The principle structural members and systems in the NnPac PAS-2 packaging are:
(1) the primary containment vessel, the sample vial described in Section 1.2.1; (2) the sample shield; (3) the secondary containment vessel; (4) the inner overpack; and (5) the oster overpack consisting of a standard N-55.
The above components are identified on the drawing as noted in Appendix 2.10.1.
They work together to satisfy the package standards set forth in Subpart C of 10 CFR 71. A detailed discussion of the structural design and performance of these components is provided below.
W Desian Criteria The primary function of the inner and outer overpacks is to protect the secondary containment, sample shield and vial from the hypothetical acci-dent conditions.
Since the sample shield is very robust, the NnPac PAS-2 packaging has been desig'ned to reduce the severity of the hypothetical accident to limits which will not affect the structural integrity of the secondary cantainment vessel and sample vial.
In order to demonstrate the packaging's ability to survive the regulatory transport and accident conditions, a combination of full scale testr, and engineering analyses has been performed.
Since the packaging is very
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' 7 r-3 ; :,, ~ _., $f f? {} i ~ bI !) l similar to the N-55 in heat transfer effectiveness, a comparison to that package's thermal response to 10 CFR 71 requirements has been made. Full scale drop tests have been performed to verify the impact attenuating capability cf the package. The irspac t absorbing capabilities of the N-55 packaging have been tested several times for a package gross weight of 750 pounds. These tests are documented in the Safety Analysis report for the N-55 syeten (Certificate of Compliance No. 9070). In order to absorb the kinetic energy of the 2100 pound NuPac PAS-2 sample shield, vial and containment, additional foam is used between the N-55 and the shield. The foam is contained by a standard DUT 17H 55 gallon drum. Impact energies are absorbed in several ways. The polyurethane foam inside the N-55 and the 55 gallon drum deforms at a fairly constant stress on t impact and has been used successfully in many currently licensed Type B packages as the primary energy absorption medium. Additionally, the steel shell of the N-55 as well as the steel drum shell of the inner overpack also deform plastica 11y on impact. Full scale drop tests were performed to demonstrate the effectiveness of this system. M Weinhts and Center pl Gravity The weight of the NuPac PAS-2 packaging is appretimately 2400 pounds. The outer overpack weight (the N-55) is approximately 180 pounds; the inner overpack weight is approximately 120 pounds; the secondary containment vessel weighs 300 pounds, and the sample shield and vial assembly weighs approximately 1800 pounds. The center of gravity for the assembled package is located at the approximate geometric center of gravity. A reference point for locating the center of gravity is shown on the General Arrange-ment Drawing. (Appendix 2.10.1) 2-2
1 PROPR]HTARY DA7A 2.3 Mechanical Properties pl Materials Mechanical properties of the various materials used in the NuPac PAS-2 packaging are shown in Table 2.3-1 below: TABLE 2.3-1 MECHANICAL PROPERTIES OE MATERIALS ULTIMATE YIEID STRENGTH STRENGTH E 6 MATERIAL (EST) (ESIl (110 PSIl LOW CARBON HOT ROLLED STEEL 63 46 29 ASTM A516 GR. 70 ASTM A513 70 38 29 II) STAINLESS STEELS (304,316) 80 35 26 3 PCF FOAM 50 PSI 1,700 T'SI 10 PCF FOAM 375 PSI 10,000 PSI II) Values shown are typical of stainless steels used in the NuPac PAS-2 packaging. Note that all structural stainless used in the NuPac PAS-2 is l tested to 150% of its design load without visible deformation. l l 2-3 l
PROPRDETARY. DATA 2.4 General Standards for All Packanes Ilis section demonstrates that the general standards for all packages are met. 2.4.1 Chemical and Galvsnic Reactions The materials from which the packaging is f abricated (steel, lead, fiber-glass, and polyurethane foam) along with the contents of the package will not cause significant chemical, galvanic, or other reaction in air, nitro-gen, or water environments. 2.4.2 Positive Closure Positive closure is effected by a valve operator cover plate which holds the valves closed during s hip m e n t. Additionally, the valve ports are plugged with f ace sealed 0-ring seals installed during transit. Secondary containment vessel lid is fastened by eight, 5/16 in, bolts. An 0 ring seal prevents leakage from the vessel. The outer overp6ck is equipped with suitable locks and tamper indicating seals to prevent inadvertent and undetected opening. 2.4.3 Lifting Devices l l 1 The outer overpack is equipped with four lif ting lugs capable of lif ting the entire system except when the sample shield is installed in the inner 1 I overpack. These lugs will be rendered inoperable during transit by use of two flat washers held to the lug with a 1/4 in, bolt to act as a cover. These lugs are the standard N-55 lif ting lugs and can be used to lif t a total of 750 lbs. of packaging, according to the N-55 Certificate of Compliance Number 9070, 2-4
PROPRDETARY DATA The NuPac PAS-2 packaging, without the 1800 pound sample shield, weighs 600 pounds. Therefore, the Margin of Safety on the lugs compared to the N-55 lug rating is as follows: M.S. = 750 - 1 =.25 600 The Carr Lane hoist ring welded to the secondary containment vessel lid is rat ed for 2,000 lb s. (see Figure 2 4.3-1), but is only used to lift the secondary containment vessel lid. This lid weighs only 130 lbs., so there is a.'arge Margin of Safe ty. The lid is placarded to prevent lif ting it while attached to the vessel body. The top of the sample shield is fitted with a 2500 lb. Carr Lane hoist ring (see Figure 2.4.3-1). The sample shield weighs approximately 1800 lbs. and the rated load of the ring is one fourth of its ultimate strength. Since the yield point of mild steel is typically 62 percent of the ultimate strength, the yield load on the ring can be given as: (.62)(4)(2500) = 6,200 lbs. 10 CFR 71 requires that lif ting fixtures be capable of lif ting three times the package weight without exceeding the fixture's yield point. Therefore, the margin of safety of the hoist ring is: 6200 M.S. = 3(1800) - 1 = + 1.5 e 2-5
SWlVEL HOIST RINGS FEATURING 3 LOAD CAPACITIES 1000,2500 AND 5000 POUNDS WITH A DESIGN SAFETY FACTOR OF 4 TIMES THE RATED LOAD CAPACITY ~~~ SWlVEL HOIST RINGS... Forged clloy base ~ ~ block actuates 360 while the forged alloy lift l ring pivots 180 in the base slot which is off center, allowing the ring to be so!id construc-c ) [I r T tion and assuring the base block rotating in ( =% T [g/ the direction of the applied load. ^ _ bg . - _.L_ 9 C ALLOY STEEL, HEATED AND -rd d CERTIFIED TO MIL-1-6868 -G Part No. A lB C D E F G H J R CL.104HR 1.1/ 8 5/8 1.1/2 6 'S 5/8 11/4 13'16 5/16 3/516 TM'D. 5. 8 LG. 13/16 CL.25-Sed 1.58 F, 8 IJ/S F/8 F/0 134 1.1/4 F/16 SeS 11 TM'O 7 8 LG. 1.1/ 4 / CL.50 5HR 2 3/16 11/8 23/4 1" 1" 2 1.11/16 3, 8 3/410 TN'D 1" LG. 13/8 HOIST RINGS BLACK OIIDE FINISH PER MIL. SPEC.-C-13924A 4130 - FCEGED ASEENBLY h_, x N @RU2TARY DATA /'( R G F ~ g h L'.~_ ---y-r-'1 l l [d-( [.-kh=-1-.. ) . f- ~- 7 l
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QJ Yf CA C SI -B l--- E SAFI LOAD CAP.R LANE HOITT RINGS ARE AN P W NO. A B C D E F G H R IN LBS. EFTICIENT MEANS OF PROVIDING ' ~ CL - 20. HR 2000 1 3/4 1 5/16 3/4 2 2 1/2 5/16 1/2 MAD CARRY W CAPACI'rY RE ,.THE ES 'TS a 25 HR i2 M b 2 1/4 1/8 3/8 7/8 2 1/2 3 3/16 3/8 5/ 8 l a. 50 - HR 51TFT 5/ 8 1/2 L/2 1/8 3 3 7/8 1/2 3/4 % LESS OF M DIRECTION OF PULL 2 1 l/ d 6 I 5 6 3/16 1 j I ,M RMS W N ~ TAKING UP A MINIMUM AMOUNT OF SPACE. app g/pe MANUFACTURING CO.
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~ FIGURE 2.4.3-1 2-6
F60PRCHTARY DATA The four shield plug attachment bolts, made to SAE J429 grade 2 specifica-tions or better, must react the weight of the shield body whenever the shield assembly is lif ted by the hoist ring. These 1/2 inch bolts (13 UNC threads) have a stress area for tension given as.1416 in.2 and a yield point given as 36,000 psi. The total allowable load through the four bolts is then: (.1416)(4)(36,000) = 20,390 pounds The margin of safety is then (using the total shield weight): M.S. = 20.390 - 1 = + 2.77 3(1800) 2.4.4 Tiedowns The NaPac PAS-2 packaging does not employ tiedowns as a structural part of the package. 3 j, Standards & T,ygt B_ gas) Larne Quantity P.3,g1;s alua This section demonstrates that the standards for Type B and Large Quantity Packaging are met per the requirements of 10 CFR 71.32. 2.5.1 LEst Resistance The NnPac PAS-2 packaging has been analyzed below to the requirements of 10 CFR 71.3 2, to demonstrate that it could satisf:ctorily withs tand the specified loading. It shon1d be noted that the package is not intended to be shipped in any position but vertical. However, assuming that it was placed on its side and loaded as follows, the critical area would be that of latch capability: W = (2400)(5)/48 = 250.0 lb/in 2-7 l.
t w, a <a-;a i P r r p 32 ~ h 30 / g* FIGURE 2.5.1-1 Moment at latch is given by: M = (218.8)(30) (48-30) 2 M = 67,500 in-Ibs. Conservatively assume only two latches are effective: n IL l b o v
- P FIGURE 2.5.1-2 2-8 2
^ s D = 32 inches 32P = M P = 67,500/32 P = 2,109 lbs. Rated cepacity of each latch is 4500 pounds. Therefore, the latches and the package can safely withstand these loads. Margin of Safety = 4500 - 1 = + 1.13 2109 2.5.2 External Pressure An external pressure of 25 psig will be reacted by the overpack. The double wall construction acts as a stress skin design forcing the two shells to work together. Sides react pressures in hoop compression while the thicker skin ends act in plate bending. Stresses produced in the circular ends due to external pressure loads will be carried in the form of plate bending with the outside sheet is in compression and the inner sheet is in tension. Foam compression strengths in excess of 50 psi are present at all points. The maximum bending moment in a circular uniformly loaded plate is given in Timoshenko, Theorv 91 P.La h 331 Shells, page 61, as: M, = Mt " 3 +" QA2 16 Where: p = Poisson's Ratio =.3 Q = external pressure, (25 psig) A = radius of plate (16 in.) Then: M, = 3 +u (25)(16)2 = 1,320 in-lbs. 16 in. Consider an element in the circular container end, subjected to 25 psi external pressure: 1 2-9
m "".mCPRGETARY Ogyp C7" 2 FIGURE 2.5.2-1 The moment of inertia I for a sandwich structure, using the transfer formula is equal to: 2 I, = I,, + 2Ad o Where: I,, is negligible A is Area per unit width of laminae d is distance from the centerline For a 5-1/2 in. foam thickness, I, = 2 (.048)(2-1/4)2 =.485 in4 o Stress in the steel shells will be equal to: b. h. (1320)(2 1/4) = 6,110 psi t I .485 l l l j 2-10
iGROPRDETARY DATA If the allowable stress for the steel is taken at 46,000 psi (reference Table 2.3-1) then: M.S. = 46.000 - 1 = + Large 6,100 Therefore, the outer overpack is capable of reacting the 25 psi external pressure. 2.6 Normal Conditions 2f.Transnort The NuPac PAS-2 Packaging has been designed and constructed, and the con-tents are so limited (as described in Section 1.2.3 above) that the perfor-mance requirements specified in 10 CFR 71.35 will be met when the package is subjected to the normal conditions of transport specified in Appendix A of 10 CFR 71. The ability of the NuPac PAS-2 Packaging to satisf actorily withstand the normal conditions of transport has been assessed and des-cribed below: 2.6.1 Heat A detailed thermal analysis can be found in Section 3.4 where the package was exposed to direct sunlight and 130 F still air. The steady state analysis conservatively assumed a 24 hour day maximum solar heat load. The maximum inner overpack temperature was found to be less than 240 F. Such temperatures will have no detrimental effects on the package. 2,6,2 Cold The two major concerns relating to the structural performance are brittle fracture and the possible freezing of the sample liquid. Brittle fracture is examined according to the recommendations of NUREG/CR-1815 UCRL-53013. 2-11
f ) /h; b L Of the three major containment components, only the secondary containment vessel and the sample shield are fabricated from brittle-frac?.ure-sensitive ferritic steel. The sample vial is made from austenitic stainless steel, and thus is of no concern for brittle fracture at the temperature in question. Both the secondary containment vessel and the shield shell are fabricated from ASTM A516 grade 70 which has a nil-ductility transition at -10 F, according to Table NC-2311(a)-1 of Section III, Division 1, Sub-section NC of the ASME Boiler and Pressure Vessel Code. Brittle fracture, being a dynamically induced problem, is of concern at temperatures not less than -20 F, since according to Reg. Guide 7.8 the initial temperature of the cask during any of the normal loads is given to be this value. According to NUREG/CR-1815, the lowest allowable service temperature (LST) for a given steel is given by the equation: LST = TNDT + A Where T@ T is the nil ductility transition temperature and A is found from Figure 7 of the NUREG/CR-1815, shown here as Figure 2.6.2-1. Since the thichest ple.te in the design is the 2.38 inch thick Secondary Containment Vessel lid, A for the lid can be found from the Figure to be -10 F, allowing for the 70 F shif t allowed for steels with yield stresses less than 60 ksi in impact limited situations. Therefore: LST = -10 + (-10) = -20 F Therefore, the NuPac PAS-2 packaging meets brittle fracture requirements over the range of temperatures required by the applicable regulatory l guides. i l 1 2-12
= -. PROPRlHTAT4Y DATA I I I I 70 F fo'r oY < 60 ksi l I I I TN O T = LST-A [Th,n sect,on ru!cs apply i i 30 F 60 ksi < oy < 100 ksi K,93/oyd 1.5 y p=0.6 je. - - - + -+- g E I 1.0 2 f .w t 0.5 E l l I h l s I i I i I I b I I I I I I 8 I 0 -40 70 0 20 40 60 80 100 0 l 1 2 2.38 3 4 A ( F) Thickness (in.) DESIGN CIIART FOR CATEGORY II FRACTURE CRITICAL COMPONENTS FIGURE 2.6.2-1
~ N:j=,UL*,WUfT/j(Qj7 gg.; A m,n ~ The second concern relating to structural performance at low temperaturcs is the effect of freezing of the sample. The intended use of the NuPac PAS-2 packaging (to transport reactor coolant samples to a qualifted labor-atory for analysis) demands that the sample be tightly sealed within the sample vial. Since the internal heat of the coolant is not large enough to cause a 72 F temperature gradient between the sample and the outside of the cask, the sample will free e if its packaging is exposed to a steady state ambient temperature of -40 F. Although the sample vial is fabricated from ductile austinctic stainless steels, the sample vial will not satisfactor-ily withstand such conditions on a normal basis. However, the following characteristics of the packaging and its intended use mitigate the conse-quences of this situation: I 1. The package would be used only in response to an emergency situa-tion at a commercial reactor site. Such situations would not occur frequently in the life of a plant. 2. The package would be shipped on a priority basis whenever used, due to the time value of the contents. Therefore, even if the package were unprotected from the environment during shipment, the foam-filled overpacks would insulate and retain the stored heat held in the shield mass for a considerable period af ter exposure to cold. This would maximize the time between initial exposure and solidification of the sample. 3. Because of the relatively small size of the package, providing the package with an enclosed, heated compartment would not be difficult when freezing weather conditions would not allow stan-dard truck shipping procedures. ] l L l l 2-14
4. Should f ailure of the sample vial occur, all contaminated mat-erials would renain fully contained within the secondary contain-ment vessel, with most remaining inside the sample shield cavity. There would therefore be no release of contamination to the environment. 5. All materials needed in the construction of all the components of the NuPac PAS-2 packaging have been shown to be brittle fracture resistent according to the requirements proposed for Category II applications in NUREG/CR-1815, UCRL-53013. Therefore, neither the secondary containnent, the sample shield nor the overpacks 0 would exhibit reduced capabilities at -20 F, the lower tempera-tere limit of operational loads. 2.6.3 Pressure Only the sample vial portion of the NuPac PAS-2 packaging and the secondary containment vessel are designed to be pressure tight. Seals on the N-55 outer overpack as well as the lid closure on the inner overpack are designed to minimize the entrance of external environmental elencats such as rain, dust, etc. The valves are rated at 250 psig at a temperature of 450 F. The piping components of the vi21 assembly are rated per ANSI B.31.1 to 1000 p s i or g re a te r. The sample vial is tested to an internal i pressure of 375 psi to verify its integrity after f abrication; therefore,, the.5 atmosphere pressure requirement does not present a hazard. The secondary containment vessel is tested to 1 atmosphere pressure and demon-i strated to be leak tight. Therefore, the.5 atmosphere pressure in the secondary containment also is not more severe than the fabrication verifi-cation tests on it. t 2-15
[+hkr.- R8tETARY DAyg 2.6.4 Vib ra t ion Shock and vibration normally incident to transport are considered to have negligible effects on the NuPac PAS-2 Packaging. 2.6.5 Water Sorav s Since the package exterior is co.sstructed of steel, this test is not re-quired. 2.6.6 Free Dg_qp, The four foot drop requirement is not applicable in light of the more stringent 30 foot drop requirement of Appendix B of 10 CFR 71. Refer to Section 2.7, below. 2.6.7 Corner Dg_qE This requirement is not applicable since the NuPac PAS-2 Packaging is fabricated of steel. 2.6.8 Penetration From previous container tests as well as engineering judgement, it can be i concluded that the 13 pound rod would have a negligible effect on a heavy 20 gauge foam backed steel shell. 2.6.9 Comoressio3 10 CFR 71 requires that packages under 10,000 pounds gross weight be cap-able of supporting a compressive load equal to the greater of 5 times the package's loaded weight or 2 psi uniformly applied the top and bottom surf aces of the package in the position which the package is normally transported. Since the NaPac PAS-2 is designed to be t ransported vertic-ally, the surfaces in question are the 32 inch diameter circular ends: i 2-16
9 PROPMETARY DATA n(32)2 = 804.25 in.2 4 Since the package weighs 2400 pounds, the compressive stress for a load of 5 g's is: 5(2400) = 14.9 psi 804.25 Since the entire external shell is backed with rigid polyurethane foam with a compressive strength of approximately 50 psi, the Margin of Safety is: M.S. = 50 - 1 = 2.35 14.9 2.6.10 Conclusion As tre result of the above assessment, it is concluded that under normal conditions of transport: 1. There will be no release of radioactive material from the containment vessel. 2. The effectiveness of the packaging will not be reduced. 3. There will be no mixture of gases or vapors in the package which could, through any credible increase in pressure or su explosion, reduce the effectiveness of the package. 2.7 Hvnothetical Accident Conditions The NuPac PAS-2 package has been designed and its contents are so limited that the performance requirements specified in 10 CFR 71.36 will be met if the package is subjected to the hypothetical accident conditions specified in Appendix B of 10 CFR 71. To demonstrate the structural integrity of the package and its ability to withstand the hypothetical accident conditions, detailed analyses and in11 2-17
u
- \\
scale tests were conducted. It is important to note that the techniques, analysis methods, assumptions, sed routines employed follow closely those used for other petitions such as: 1. DOT 6400 Super Tiger 2. DOT 6553 Paducah Tiger 3. DOT 6272 Poly Panther 4. DUT 6679 Half Super Tiger 5. DUT 6744 Poly Tiger 6. AECB - Resin Flask 7. NuPac Model N-55, certificate of Compliance #9070 8. T-3 Spent Fuel Shipping Cask, Certificate of Compliance No. 9132. 9. 1-13C(II) Shipping Cask, Certificate of Compliance No. 9152. These are proven techniques that agree closely with full scale tests as well as other publicized standards such as ORNL-NSIC-68. In all cases, the k analysis has been proven to be conservative when compared with full scale testing. i 2.7.1 Free _ Drop Events e Full scale prototype drop tests were conducted on the NuPac PAS-2 system to demonstrate the cask's survivability of the hypothetical accident condi-tions postulated in Appendix B of 10 CFR 71. Nuclear Packaging, Inc. has considerabic experience designing and drop testing overpack systems, having conducted several successful drop tests since 1977 on the N-55 Package as well as the 1-13C and the Paducah Tiger. The test pad used was specially designed to simulate as much as practical an unyielding surf ace, and has been the target for drops of as much as 46,000 pounds (the Paducah Tiger). It is approximately 6 feet wide, 12 feet long, and 8 feet deep; constructed of concrete, with a 2 inch thick steel plate inlaid into the top surf ace. The test pad has not exhibited any discernable deformation from any of the protected and unprotected drop tests conducted on it. 2-18
PROPROMTARY DATA One NuPac PAS-2 prototype was subjected to 3 drop tests. Firs t, the pac-kage was dropped 30 feet with its centerline parallel to the surface of the te s t pad. Second, the same package was dropped 40 inches onto a 16 inch long, 6 inch diameter steel post, per Appendix B, paragraph 2 of 10 CFR 71. Finally, the same package was subjected to a second JO foot drop with its centerline oriented perpendicular to the surface of the pad, with the top end of the package impacting the pad. Photographs of the drop tests can be found in Appendix 2.10.3. By examining the results of previous drop tests, the ef fects of various drop orientations can be evaluated. The particular drop orientations tested maximize the potential damage to the secondary containment vessel. In a side impact, there is the least foam thickness to dissipate the drop energy, causing the highest accelerations to the package. The 40 inch drop onto the 6 inch diameter post was positioned such that the post would directly impact the surface of the package corresponding to the secondary containment lid seal. End impact was effected on the top end again maximizing the potential damage to that critical seal. Acceptance criteria were based on the ability of both the secondary containment vessel and the sample vial to pass a highly sensitive leak test before and after subjection to the drop tests. 2.7.1.1 Side Drop. Results The first drop test performed on the NuPac PAS-2 packaging was the side impact drop from 30 feet. Before and after photographs are shown in Appe nd ix 2.10.3. Permanent deformation of the packaging was measured as l l follows: l t 2-19
b f j [f $g Reduction to outside diameter of outer overpack at impact 2.5 in. Reduction of inside diameter of outer overpack at impact .8 in. Reduction of inner overpack foam thickness 1.3 in. TOTAL FOAM DEFORMATION: (2.5 .8 + 1.3) 3.0 in. The circumferential stif fening hoops on the drum were flattened on the impact side, and the fiberglas inner skin on the outer overpack also was cracked in the area of impact. One of the outer overpack lid latches was directly on the point of impact; after impact the latch held, but was rendered flush with the side of the package. An 8 inch long crack in the side of the 55 gallon drum also appeared, but since the drum is not a containment boundary, this crack is of no consequence to the safety of the packaging. 2.7.1.2 Puncture Droo Results After the side drop from 30 feet, the package was dropped 40 inches onto a 6 inch diameter post. The package impacted the post on the opposite side of the package than was impacted during the side drop. Before and af ter photographs are shown in Appendix 2.10.3. Post impact caused the following permanent deformations: 1 2-20
PROPRf12TARY DATA Reduction of outside diameter of outer overpack local to impact: 1.0 in. Reduction of inside diameter of outer overpack: .3 in. Total local deformation of the outer overpack: .7 in. There was no permanent deformation to the foam inside the drum. The druz experienced minor deformation local to impact. This drop also impacted a latch, which was severely damaged. In spite of this damage, the latch still held. The fiberglass inner skin of the outer overpack cracked slightly local to the impact. 2.7.1.3 Er.d Dron Results The final drop test had the package impacting directly onto its top sar-fece. This orientation caused the secondary containment vessel lid to react the full impact load from the 1800 pound sample shield. Be fore and after photographs are shown in Appendix 2.10.3. Permanent deformations to the packaging are summarized below: l l l 2-21
i,,. ~ FROPRETARY DATA x \\ Reductio of overall length of outer overpack: .8 19 t s \\. a kL g Increase of cavity length at top of outer 1 overpeck: 5.5 in. \\ s n Total decrease in end thickness of h \\ i outer overpack: 6.3 in. s s< Deflection of drum cover center: '2.'0 i:re ' l' s s Reduction in thickness of top inner p. overpack f'oam: .8 in. N Total Overa1(Foam Deformap on (6.3 +.8) 7.1 in. ~' ~ The inner firerglass liner on the outer overpack experienced a shear / ten-x. sion failure circumferentially where the drum impacted the outer ovespack. Examination of the damage indicates that the drum cover bulged out. against g the top of the outer overpack causing a dome shaped def' rmation in,'the o foam. Then the fiberglass failed as described above. Thes, failure of k he t fib e rg la s s liner contributed significantly to the energy $issipation sprop-erties of the outer overpack. i s. -A s Interestingly, one of the previously undamaged latches, opened on impact, ( while both puviously impacted latches continued to hold firm. Apparently,s the impact drove the\\ outer overpack base isto the Ild on one side f a r 'o -c s. \\,s enough to unhook the-pul'I-down togg$ s'. Exi*yination of t,he lid lip neEr the '\\ s e i s, failed latch revealed that tho.,foain in,the' lid had shattered. dither in the end impac t or one of the ' previous' drops, allowing f airly free movement 't between the two halves of'the overpack near the failed latch... s i, ( s 2 7.1.4 Conformance to Acceptance Criteria N j; ) ( Survivability of the containment boundaries and shield' is's sacial to the L packages conformance to paragraph 71.36 of 10 CFR 71. Care ful attention y-N 4- . ~ s ', N 2-22 tu
c 1; PROPRCHTARY DATA was given before and af ter the ' drop test to insure that any leaks present in the systems af ter the drop test would be found, and would be as a result of the cumulative effects of the drop tests rather than po s s ib le mishandling or assembly. A leak rate from any leak test in excess of 1 x 10-7 standard cubic centimeters per second would be considered unac-ceptable. All leak tests, both before and af ter the drop tests, on both the sample vial and secondary containment, tested leak tight to better than 1 X 10-9 standard cubic centimeters per second, the limit of sensitivity of the mass spectrometer leak detector used for the test. Neither the shield lnor the secondary containment vessel sustained any pe rceptible de forma tion. The only indication on either item that any unusual loading had becured was on the inside of the secondary containment lid, where the shield plus closure bolts scratched the highly polished surface of the lid. Therefore, it can be concluded that shielding will be unaf fected by the ef fects of a 30 foot drop or 40 inch drop onto a 6 inch diameter post. 2,7,2 Puncture The effects of a 40 inch drop onto a 6 inch diameter steel post are documented in Sec tion 2.7.1 above. Prototype testing showed no damage to any containmentsboundaries from this event. 2.7.3 Thermal Analysis w Thermal evaluations'for the NuPac PAS-2 Packaging have been conducted for both normal transport and hypothetical thermal accident conditions. These evaluations are presented in Section 3.0 of this report. 2.7.4 Water Immersic_n .x This section is not applicable, since the NuPac PAS-2 package will not b x contain fissile materials. p s ~ L! u %/A ,) 1 (2-23
. = - h E3\\, L.L1 Summary 21 Damen ' From the above analyses and tests, it can be seen that there will be no significant damage to any of the NuPac PAS-2 containment systems or radia-tion shields from the hypothetical accident event sequence set out in 10 CFR 71. Damage would, in f act, be limited to the inner and outer over-packs, and there would be no reason to require any rework or remanufacture of any other part of the packaging systems af ter such an event, should it ac tually occur. L.J. Special Eorm Not applicable, since no special form is claimed. L1 Ensi Esd.1 i Not applicable, since no fuel rods would be packaged in the NuPac PAS-2. L L. Apnendix 2.10.1 General Arranaesent Drawina sf. tht NuPac PAS-2 Packmains. l l l l l 2-24 i
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2.10.3 Drop Test Results The following pages present a pictorial record of the drop tests performed on the NuPac PAS-2 package. The photographs are numbered, and explanations appear below: Photorrach No. 11 The NuPac PAS-2 package, just prior to drop tests. Photograph h 21 The NuPac PAS-2 package, subsequent to the final 30 foot drop, impacting on the package top end. Photographs 1 and 2 provide a good before and af ter comparison for the cumulative effects of the three drops on the packs;ing. Photograph No. 31 Just prior to first 30 foot drop. Pi st ou rsch No.11 Package shown approximately one foot above ground on rebound after initial impact. Photorrach h 51 Nessurement of damage from the side drop. Undeformed diameter is 31-3/8 inches, photo shows 2-1/2 inch deformation at bottom end. "hotograoh h fi Impact zone of side drop. Note newly recessed latch. Photorraoh h "g1 Puncture impact height verification. Photograph h 81 External damage from puncture impact. Note popped rivots. Photomrath No.11 NuPac PAS-2 rigged for final 30 foot drop. See photo-graph No. 2 for resultant damage. 2-2S t
PROPRCETARY DATA Photorrach h 10: Exterict of 55 gallon drum af ter tests. White areas are due to abrasion with' N-55 fiberglass liner. Note horizontal crack in center of picture. Photograph h 11: Inner overpack dcmage af ter drop. Note cra:ks and crushed foam on f ar side. Blue streaks results from glue used to glue foam pie c e s together. Photorraoh h 12: Damage to inside of outer overpack from end drop. Note fiberglass separation on far side. Photorraph No. 13: Inside of Secondary Containment Vessel Lid. Note scratches from shield plug closure bolts. i 2-29
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} b? i y M THERMAL EVALUATIONS M Discussion The NuPac PAS-2 packaging exhibits ther; mal properties and characteristics nearly identical to those used in the thermal analysis of the N-55, Certi-ficate of Compliance No. 9070. Some significant similarities and differences in the two packages are noted below: 1. The N-55 and the outer overpack of the NuPac PAS-2 are identical. 2. Damage to the NuPac PAS-2 from any of the regulatory 30 foot drops does not exceed the cumulative damage used in the analysis of the N-55. As a result, the features of the hypothetical accident thermal model of the N-55 conservatively approximate the response of the NuPac PAS-2. 3. The N-55 Classification II payload model includes the thermal effects of the vermiculite packing used in that package. This packing is thermally enalogous to the inner overpack of the NuPac PAS-2. 4. The N-55 Classification II has been analyzed to handle 5 watts of internal decay heat. The NuPac PAS-2 packaging will carry only 3 I watts maximum. 5. The NuPac PAS-2 packaging's heavy sample shield is a vary effec-tive heat sink not present in the N-55 taalysis. This difference l means that the NuPac PAS-2 will not be affected by the hypothetical fire transient as much as the N-55. As a result, temperatures in the NuPac PAS-2 packaging will be somewhat less than those predicted for the N-55. l l 3-1
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Comparison with the N-55 thermal analysis yielded the following conserva-tive assessment of temperatures for normal and accident conditions: pneMar. COBBITIONS Outer Overpack 0 External Surface 162.02 F Internal Surface 192.0 F Primary Containment 236.6 F EYP011ETICAL.iCCIDENT C0fSITIONS 0 Secondary Containment 236.3 F Primary Containment 263.3 F(1) III Analysis conservatively ignored the thermal mass of the NuPac PAS-2 sample shield, rendering this an extremely conservative estimate of the primary containment temperature. L.2 Summary pl Thermal Pronerties 91 Materials The following thermal properties were taken from the Safety Analysis Report for the N-55 packaging, Certificate of Compliance No. 9070: i l l l 3-2 l l
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/;, Thermal Conductivity, k (BTU /Hr-ft *F) 26.0 Specific Heat Capacity, Cp (BTU /lb *F) .113 3
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3.0 (outer overpack) YERNICULITE Thermal Conductivity .0375 (BTU /Hr-ft *F)
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~ ~ {.ri3,.3 Q q=. L Additionally, appropriate corrections wars made to the thermal model of the N-55 to account for damage that would occur from the 40 inch drop onto a 6 inch dismeter post, as well as 30 foot drop. Very conservative assumptions were made concerning the thermal decomposition point and thermal conductivity of the foam. As a result, the analysis of the N-55 resulted in a very conservative assessment of its thermal response. J.,,j, Technical Srecificaticas rd Connonents Temperature sensitive components of the NuPac PAS-2 packaging include several 0-rings as well as the two valves in the sample vial. Sealing 0-rings on this package are optionally viton or silicone rubber with the following allowable temperature ranges as published by Parker Hannifin Corporation, a major manufacturer of 0-rings: NATERIAL TEMPERATURE B6KSE 1*E). Viton -65 to +400 (+600 for short periods) Silicone -175 to +450 (+700 for short periods) The bore seal around each valve operator is a Bana-N 0-ring, which exhibits a working temperature limit of -65 to 250 F. The Whitey ball valves are rated for working temperatures of 450 F at 250 0 0 psi, and 400 F at 375 psi, as shown in their product data sheet. All other elements within and including containment are of the following materials: I 3-4
MiOPRCETARY DATA MATERIAL MELTING MI L E Lead 621 Steel 2400 Vermiculite 2200 Jd Thermal Evaluation in Normal Conditions d Transnort The analysis of the N-55 packaging, Certificate of Compliance 9070, is published in the Safety Analysis Report of that package. Two thermal analyses were performed on the N-55, corresponding to general payloads (Classification I) and a mixed oxide Inne-Containment Assembly (Classifi-cation II). The analysis of general payloads was performed using conserva-tive assumptions, and found that payloads of at least 3 watts could be transported without causing the inside of the N-55 to exceed 212*F, even during and af ter the hypothetical fire transient. The mixed oxide payload, with 5 watts of decay heat, was found to remain below 220 F using a slightly less conservative analysis. A comparison of the pertinent analy-tic features of the two N-55 calculations with the features of the NuPac PAS-2 packaging appears below: e a 3-5 ,, _ _. -. -.. - - ~ - - _ _ -. _ _
l i dh( I/ .b),,$1 ~ [\\_ N-55 N-55 Cla a. I Gn 11 NuPac PAS-l Outer overpack N-55 type N-55 type N-55 type Enclosed drum with gaps between it and overpack Present Present Present Internal Insulation None Vermiculite Foam Inner overpack Internal Decay Heat 3 watts 5 watts 3 watts Secondary Containment None None Prestat 0 Loading Temperature Ambient Ambient 400 F The following worst case temperatures were reported in the N-55 Safety Analysis Report for normal conditions (130*F Ambient air): I1) N-55 N-55 l l Class. I Class. II External Surface 162.02 F 170.6 F 0 Internal Surface of N-55 172.69 F 192.0 F 55 Gal. Drum Not Modcled 193.0 F l Inner Containment 236.6 F (1) Temperatures were conservatively derived by adding 30 F to the tempera-tures resulting from a steady-state analysis performed using 100 F as ambient temperature. 4 i 3-6 l l
_ _ = _ - f 3 For purposes of steady state analysis, the NuPac PAS-2 configuration is more similar to the N-55 Classification II packaging than the Classifica-tion I configuration. The most significant difference between the NuPac PAS-2 and the N-55 Classification II lies in the differing thermal proper-ties of the packaging within the 55 gallon drum. In the N-55, vermiculite is placed into the void between the mixed oxide containment and the drump while the NuPac PAS-2 employs polyurethane foam in the region between the secondary containment and the drum. From Section 3.2 above, it can be seen that the thermal conductivity of the foam is significantly higher than the vermiculite, meaning that heat will more easily flow out of the NuPac PAS-2 than out of the N-55 Classification II. Therefore, the thermal steady state analysis of the N-55 Classification II yields a conservative estimate of temperatures found in the NuPac PAS-2. Note also that the NuPac PAS-2 is loaded with 3 watts versus 5 watts in the N-55 Classification II. The high relative conductivity of lead and steel when compared to the polyurethane results in a relatively small thermal gradient betwc:n primary containment and the inner overpack foam. From Equation 2-7 of Kreith, Princinles 21 h Transfer, First Edition, page 26, the gradient through the sample shield can be calculated by the following: AT = a ln(d d of i) 2nkl Where: q = heat load = 3 watts = 10.2 BTU /hr d, = outside diameter = 16.5 inches dg = inside diameter = 3.75 inches L = length - 21 inches = 1.75 feet t l k = thermal conductivity of lead = 19 BTU Hr-ft *F l 3-7
So, AT = 10.2 ln(16.5/3.75) =.073*F 2n(19)(1.75) In the steady state, then it can be assumed that the temperature of the sample shleid is essentially constant. l 3.4.1 Maximum Temperaturgi It is important to note that the NuPac PAS-2 sample vial is designed to be loaded, under normal conditions, at temportures as high as 400*F and pressures as high as 250 psi. Thus, the most severe normal thermal loading is not the steady state hot environment, but the initial loading condition. Since all the valves and important 0-ring seals in the system are capable of handling this temperature, the NaPac PAS-2 will suffer no undesirable effects from any of the normal thermal loads. 3.4.2 Minimum Tenneratures t The minimum temperature that the NcPac PAS-2 will be allowed to experience 0 is 32 F, since it is possible that freezing of the sample might damage the primary containment boundary (the sample vial itself). Should this occur, however, the package is designed with attention to confining any leakage to the shield cavity and absolutely containing the contamination to within the secondary containment vessel. As a result, should the NuPac PAS-2 be subjected to -40"F, the packaging would not allow any escape of contamin-ated materials to the environment, since none of the materials used in the i package are significantly affected by this temperature. 3.4.3 Maximum Internal Pressure The maximum internal pressure within primary containment (the sample vial)is 250 psi, corresponding to the partial pressure of water at 400 F. The sample vial will not be subjected to pressures or temperatures greater than these values. 1 3-8 I_
-__7 .-c. 7 v v h,g g U7=-v g~ "b'.!)Q#l5 U b. ' {i~iihr f.4 LC-- d (I? S v;! Nd L' ~ ~ ' A 3.4.4 Thermal Strasse1' i Because the inthrnal eat load in the NuPac PAS-2 sample vial is,so' low -l ~ (only 3 watts), and the thermal conductivity of steel and lead is so high relative to.th? fome in the inner and. outer overpacks, significant thermal sq gradients developed Vn the via!, the shield, or the secondary conta'inment vessel are insignificant (less than.1'F). Te'apora ture estimates for the N-55 Classificatjon II package under normal conditions reveal a 2,1.4 F ~ i thermal gradient between the outside and inside,of the N-55 outer overpack, and a 43.60 gradient between the mirsd ox[de payload and the 55 gallon drum. If a. similar temperature distribution were assumed on the'NuPac PAC ' 2 packaging on analogous' features, these gradients, although mod,erately s s severe for isotropic,seidl structures, do not pose a severs problem'for combination foam, fiberglass and metal inner and outer' overpecks. The high flexibility of the foam will prevent significant stresses from developing. The primary containment will be kept at a uniform. temperature, and thus will not develop thermal stresses. s I . g i 4 Lj, Hvoothetical Accident Thermal Evaluation 3 s' l go A comparison of the thermal response of the N-55. configurations to the NuPac PAS-2 configuration is also possible for evain2 ting the efft. cts of y the hypothet ical accident conditions. It is conservative for' the acc ident-( 3 -. x conditions, however, to compare the NuPac PAS-2 response to the N-55 Classi / 12 \\ ification I response, since in the accident conditions, the worst case ~ the one which has the least thermal insulation, yielding tha worst (maxi-mum) payload temperature. ~ An examination of the thermal analysja of the hypothetical accident condi,- tions as presented in the N-55 Safety Analysis Report rave 3 s the following ~ 1 significant facts: \\~ l 1. The iazimum allowable payload of the N-55 Classificdion II is L watts, exactly the same as the NuPau PAS-2 cask. 4 s l A W ,g 3-9 t ~ ? hh
_ _ _. y. t N s ^ f !h k A 4 a l 2.~ Damage to the overpack modeled it the analysis included the r i [ cumulative ef fec ts of a top, bot to's, and side drop. Although damage to the outer orsrpack of the NuPac PAS-2 system was more severe than that incurred by the N-55, the ' damage from any one 30 foot drop of the NuPac PAS-2 in tandem with! the damage from the 40 in, drop onto the 6 in. post is not as severe as the cumula-tive damage to the N-55. Further mitigating;this increased danage is the continuing insulating effectiveness of the inner overpack fone. Therefore, the thermal model of the N-55 Classi-fication I package is a conservative representation of the NuPac PAS-2 packaging. 3. Analysis of the N-55 Classification I assumed an ambient initial l ir temperature of 130 F while Regulatory Guide 7.8 suggests using 0 [,caly 1(s0 F for an initial condition. 0 Thus the thermal analysis given for the N-55 Classification I package is more conservative t thsn required. ~., 'q 3, 't
- 4..' The sample shield represents a significant heat sink not present in the N-55 analyses.
As a result, temperatures within the sample shield will be significantly lower than predicted for the center of the N-55 during the fire transient. 3.5.1 Maximum Tenneratures According to the N-55 Safety Analysis report, the inner wall of the Classi-fication I configuration rises from its steady-state temperature of 172.7 F 0 to 199.4 F during the accident scenario. This represents an increase of 2 6.7 *F. If one conservatively assumes that this increase is equivalent to the increase of the steady state temperature of the NuPac PAS-2 payload, and that steady state payload temperature is as estimated above from the N-55 Classification II package (236.6*F) then the maximum temperature one could expect the payload to reach is: t '{ 1 236.6 F 26.7'F u 263.3 F 0 s 3-10
h A! Y i .\\ This temperature is significantly less than any critical temperature of any of the components in the NuPac PAS-2 containment or shield. Further, due to the heat sink represented by the therms 1 shield; this temperature is a very conservative estimate of the internal temperature of the NuPac PAS-2 sample shield. 3Ad Mazina Pressures The hypothetical acciJent conditions will not increase internal pressures in the NuPac PAS-2 above the design pressure of 250 psi. 3.5.3 Maximum Thermal Stresses Thermal stresses induced during the hypothetical accident condition are insignificant for the same reasons the normal condition thermal stresses are not significant. Sec Section 3.4.4 for a complete explanation. 3.5.4 Evaluation 21 Packase Performance Igx 1hg, Hvoothetical Accident Thermal Conditie The hypothetical accident conditions have been analyzed using an extremely conservative analytic comparison to the N-55 thermal analysis as presented in the Safety Arealysis Report on the N-55, certificate of Compliance Number 9070. The maximum temperature was conservatively estimated tre be 263.3*F. Only the Buna-N bore seal 0-ring around the valve operators are rated less i than this tersprature (250 F). These 0-rings are not, however, a part of a l l containment boundary. No shielding-critical elements of the NuPan PAS-2 l system are affected by temperatures less than 621 F (lead melt). From the above analysis and comparisons, it can be concicded that the hypothetical accident thermal conditions will not cause any structural damage to containment boundaries or any breach of containment, nor will any loss of shield result from those conditions. Damage will be limited to minor charring of the foam overpacks and possible damage to two inconsequential 0-rings around the valve operators. 3-11
NOPRUETARY OATA M CONTAIPOGDir 1 M Containana_t Boundaries 4M Containment Vessels There are two separate containment boundaries in the NuPac PAS-2 packaging. Primary containment consists of the sample vial, constructed of various stainless steel tubes and fittings, and two high pressure valves tested to leak tightness of greater than 1 x 10-10 standard cubic centimeters per t second. Two chrome plated stainless steel plugs f ace seal against the valve outflow ports to insure the effectivenecs of the closure. Secondary containment is provided by the secondary containment vessel, which fits around the outside of the sample shield. The secondary contain-ment vessel is constructed from ASTN A516 grade 70 steel;.375 inch sheet on its sides and bottom, and a 2-3/8 inch thick machined plate for closure. The lid seals with an 0-ring bore seal between the lid plate and the vessel side walls. A test port is provided into containment to allow testing of the secondary containment vessel seal. The test port is so designed that the test port closure seal may also be tested to verify proper closure. 4.1.2 .C_9.af ainment Penetrations There are no penetrations into prir.ary containment except the inflow and outflow ports on the valves. These are sealed using f ace-sealed plugs screwed into the ports. Penetrations into the secondary containment include the test port described above, as well as the containment lid, which is bore sealed with the body of the containment vessel, and secured by eight 5/16 inch bolts. 4.1.3 Seals And Welds Seals affecting containment are as described above. Welds on the primary containment are sealed and the assembly verified to withstand 150% of the 4-1
PROPMSTARY MTA design pressure (250 psi x 1.5 = 375 psi.' without any perceptible defc:ma-tion or leaks. Both the primary and secondary containment boundaries are verified by 'eak test to a leak rate le s s than 1 x 10-7 standard cubic contimet. ors per second. 4.1.4 Closure Closure on the secondary containment vessel is effected by eight 5/14 inch hex head bolts, which are tightened to 18 f t-Ibs. torque. The sample vial plugs use NPT threads and seal with a f ace-sealing 0-ring. Therefore, these plugs shall be hand tight for transit. ! J Reanirements IgI Normal Conditipas.91 Transnort 4.2.1 Release 91 Radioactive Material The results of analysis performed in Chapters 1 and 2 indicate that there will be no release of radioactive materials under any of the normal condi-tions of transport. Q:}. Pressurization 91 Containment Vessel Primary Containment is tested to 150% of the highest pressure anticipated i under any normal or accident conditions. There are no vapors or gases i l formed which could mix explosively within the primary containment. i Should a leak develop in the primary containment, the payload could be forced out under pressure and could flash to steam. However, the excessively large thermal mass provided by the shield and the large surface area for heat exchange provided by the lead shot within the shield cavity ensure that the temperature of the steam will not exceed the equilibium temperature of the package, or 237 F. The absolute pressaro of baturated steam at this temperature is less than 25 psia, or only.70 atmospheres 4-2
10= m?!(, & O "% - wb 'n/IIi== y ~ L;7 Lbb4Tp I m - gage. The secondary containment vessel is tested to be leak tight at one atmosphere, so even a breach in primary containment at maximsa design temperature and pressure would not exceed the pressure required to test for leak tightness. 4.2.3 Coola=t Contaminatioq There are no coolants in the NuPac PAS-2 system, so this is not applicable. 4.2.4 Coolant Len There are no coolants in the NuPac PAS-2 system, so this is not. applicable. A.J. ContainELE1 Reanirements M& Hyno thet fM Accident Conditions Drop tests and thermal analyses presented in Chapters 2 and 3 indicate that the containment boundaries will suffer no damage from the hypothetical accident conditions. Therefore, all contents shall be completely contained throughout the entire hypothetical accident scenario. 4.3.1 Fission f1A1 Products The quantity of fission gases in the primary containment vessel available for release is insignificant. 4.3.2 Release gf, Contents Becauss no damage is incurred to the containment structures from the hypo-thetical accident s c e n a. r lo, there will be no release of radioactive atterials to the environment during or after this scenario. 4-3
{ [jk;jhs5 h [../p., '. ' c.h P' ! ~ 1,.,0 SHIELDING 0 The payload for which the NuPac PAS-2 is designed is given in Table 5.0-1. These isotopea can be grouped into groups according to gamma emission energy levels. The resulting grouping with their concentrations are given below: MAXINUM ENERGY CONCENTRATION (Nev.) (Ci/cc) 0.4 1.6 4 0.8 1.3 1.3 1.2 1.7 .45 2.2 .25 2.5 .2 Assuming an effective shield thickness of approximately 6.3 inches of lead and assuming the source to be a line source 6 inches in length, a dose contribution from each of the energy groups above can be calculated. The resulting sum is 9.5 mR/hr at 6 feet from the surface of the package. The dose rate calculated includes several areas of conservatism.
- First, the lead shot was not included in the shielding calculations, nor was the small amount of self shielding provided by the water in the vial.
Second. l many of the isotopes given in Table 5.0-1 have relatively short half-lives, 1 on the order of one hour or less resulting in a shipping payload of less I activity than the initially collected payload. Third, in classifying the isotopes into energy groups, the gamma energy level was rounded upward to the next highest energy group, thus increasing the calculated payload activity. 5-1
~. v,, =,, - fy _), ' W...,,l :G; ' p,f ! l These calculations were meant only to verify the adequacy of the design. Prior to any shipment, dose readings will be taken to verify that the package meets transport limits given in 49 CFR 173. Nuclide Nuclide Nuclide BR-84 4.59E-02 RU-106 2.64E-04 XE-135 1.48E-01 BR-85 6.43E-02 TE-129M 3.44E-04 CS-135 3.69E-2 0 KR-85M 1.27E-01 TE-129 1.05E-03 CS-136 5.72E-06 KR-85 2.08E-03 TC-99M 7.29E-04 XE-137 5.78E-01 KR-87 2.39E-01 I-129 3.34E-09 CS-137 1.33E-04 KR-88 3.50E-01 I-131 1.44E-01 XE-138 5.75E-01 RB-88 3.53E-03 XE-131Li 1.98E-03 CS-138 6.55E- 03 KR-89 4.54E-03 TE-132 4.27E-03 CS-140 5.81E-03 RB-89 4.69E-03 I-132 2.14E-01 LA-140 6.26E-03 SR-89 4.66E-03 TE-133M 3.443-03 BA-143 5.14E-03 SR-90 2.96E-04 TE-133 3.63E-03 LA-143 5.81E-03 Y-90 2.96E-04 I-133 3.31E-01 CE-143 5.81E-03 SR-91 5.72E-03 XE-133 6.33E-01 PR-143 5.81E-03 Y-91M 3.37E-03 CS-134 3.31E-05 CE-144 4.05E-03 Y-91 5.75E-03 TE-134 6.81E-03 PR-144 4.05E-03 NB-95 6.10E-03 I-134 3.85E-01 ZR-95 6.01E-03 MO-99 6.07E-03 I-135 3.03E-01 RU-103 2.95E-03 XE-135M 1.77E-01 TABLE 5.0-1 NUPAC PAS-2 MAXIMUM PAYLOAD CONCENTRATIONS (CURIES /CC) 1 l 1 l [ 5-2
=..-,2.. ~ IIMR]',,l))ikif'jjiG-W l.) d(=t }l.li.)) 'l,, \\ ;l ;.d}1 ~
- d L.%
-i s.,.0 CRITICALI1T EVALUATION The NuPac PAS-2 packaging will not contain significant quantities of fis-sile material; therefore, this section is not applicable. 6-1
- 5
- *,;%p:biith I
r'N b g[ l7 {,. rU-a . s,, i L 9, OPERATING PROCEDURE L,1 Onorational Reanirements (unless (therwise noted)_ 7.1.1 The secondary containment vessel base shall never be removed from the inner overpack assembly. The hoist ring on the seconary containment lid shall never be used to lift the containment base. 7.1.2 During all assembly operations, the face seal 0-riegs on the fill tubes (Part No GF-20-02D-A2) and on the shield closure plugs (Part No. GF-20-02D-A3) shall be wiped clean. Vacuum grease shall be sparingly applied to the 0-ring groove prior to mounting the 0-ring. The 0-ring shall then be mounted in the groove taking care not to allow grease to contact the free surface of the 0-ring. Threads shall be coated with a suitable lubricant to prevent galling. 7,1.3 To prevent accidental or inadvertant spillages, the actuator rod closurer plate shall be installed at all times while the system is not in use, and shall also be installed at any time the system is transported. 7.1.4 Flow path through the valve is indicated by arrows on the valve operators. To alter the flow path through either valve, rotate the operator 180* and continue rotating until operator cannot rotate further. Then turn operator back to vertical, with arrows pointing the desired direction. See Figure 7.1.4-1. I 7-1 1
=-,x__. TiY!?h0[lyg,9 f 3. n;-. 3, CAUTION l ACTUATOR ROD COVER PLATE, GF-20-02D-A6, MUST BE INSTALLED I DURING TRANSIT. SAMPLE LOOP I! h ' 4 FLUSH LOOP O 4 l [Vn i FLUSH LOOP b i
- 1
) 4 SAMPLE LOOP l i VALVE OPERATORS SHOWN IN POSITION TO ACCEPT ACTUATOR ROD COVER PLATE. i (IN TRANSIT, OR CLOSED POSITION) FIGURE 7.1.4-1 7-2
!) L' \\ f}){h\\ 7.1.5 The shield assembly can only be lif ted using the hoist ring on the top of the shield. 7.2 Ooerational Narrative 7.2.1 Storage Procedure When not in use, the system shall be stored as follows: 7.2.1.1 The outer overpack shall be on a steel pallet, with its upper portion set aside. 7.2.3.2 The inner overpack shall be set in the lower portion of the outer overpack with the bottom and side foam pieces in place. The top foam piece and the drums closure ring and plate shall be set aside. 7.2.1.3 The empty secondary containment vessel shall be positioned in the inner overpack. All unpainted surfaces shall be kept coated with a high quality vacuum grease to prevent oxida-tion and to assure a tight seal when in use. The 8 second-ary containment vessel lid bolts shall not be installed at any time the shield is not contained within the secondary containment vessel. 7.2.1.4 The shield assembly shall be positioned on a cart in such a way that the fill ports are positioned conveniently with respect to the PASS sampling ports. 7.2.1.5 The fill tubes (Part No. GF-20-02D-A3) shall be installed as pe r Se c t ion 7.1.2 above. 7.2.1.6 Modified eye-bolts shall be installed in the top of the shield body. 7.2.1.7 The actnator rod closure plate shall be installed. 7-3
'hl f? p l 7.2.2 Filling Procedure 7.2.2.1 Remose the actuator rod closure plate. 7.2.2.2 Set the valves using the procedure in Parai;raph 7.1.4 abeve, such that the arrows indicate flow through the sample loop. 7.2.2.3 Position cart and shield assembly so that the pig-tail assemblics may be easily attached to the PASS ports. Attach pigtails to the appropriate fitting. 7.2.2.4 Run the coolant through the sample loop for a period to be determined so that the internal temperatures of the vial is very close to the temperature of the coolant and a represen-tative sample may be taken. 7.2.2.5 Set the valves, using the procedure in Paragraph 7.1.4 above, out flow valve first, so that the arrows indicate flow through the flush (bypass) loop. 7.2.2.6 Shut down reactor coolant floa to the NuPac PAS-2 Cask. 7.2.2.7 Direct non-radioactive flush water approximately the same temperature as the coolant water through the sample. Flush for several minutes. I 7.2.2.8 Install actuator rod closure plate. 7.2.2.9 Direct a compressed gas source through the vial. The gas should not be pressurized through the vial to a pressure greater than 250 psi. 7.2.2.10 Allow gas to flow until passages are dry. 7-4
f I ll-l- I 7.2.2.11 Remove pigtail assemblies and fill tubes. Be cure that the fill tube 0-rings are on the tubes after removal. 7.2.2.12 Install shield closure plugs per Paragraph 7.1.2. 7.2.2.13 Transport shield assembly on cart to the partially assembled overpacks. 1 ( 7.2.2.14 Remove the secondary containment vessel lid from its base I and place the shield assemblies inside. 7.2.2.15 Replace the secondary containment vessel lid and secure with its 8 bolts. Verify that the test port bolt is open. 7.2.2.16 Carry out the assembly seal ter.s required on the secondary containment vessel. (See Section 7.2.4). Secure the test r port closure bolt and test port cover. 7.2.2.17 Place the top foam piece into the inner overpack. Secure the inner overpack closure plate. 7.2.2.18 Install the upper portion of the outer overpack and secure. 7.2.3 Emptying Procedure 7.2.3.1 Remove the covers of the outer and inner overpacks, and the secondary containment vessel cover. 7.2.3.2 Remove shield assembly and place where vial is to be emptied. 7.2.3.3 Remove shield closure plugs and install Fill Tubes. See Paragraph 7.1.2 above. 7-5
PROPR:2TARY DATA 7.2.3.4 Connect upper fill tube to a source of pressurized gas and lower fill tube to e testing storage tank. Pressure shall not exceed 250 psi. 7.2.3.5 Af ter removing the valve operator cover, turn upper valve operator to the sample loop, then set lower valve operator to the sample loop, thus releasing the sample into the testing storage tank. See Paragraph 7.1.4 above. 7.2.3.6 Af ter entire sample has been transferred into the stcrage tank, redirect the flow from the lower fill tube into the facility waste water system. 7.2.3.7 Attach the upper fill tube to a source of warm demineralized water. 7.2.3.8 Flush the sample loop with water for suf ficient time to i remove any contamination from the inside of the system. l 7.2.3.9 Redirect the pressurized gas through the sample loop to dry the vial. l 7.2.3.10 Reset the valve operators to the flush loop and reinstall the valve operator cover. 7.2.3.11 Redirect the source of warm water through the fill tubes and flush for 30 seconds. 7.2.3.12 Again direct the pressurized gas through the fill tubes until passages are dry. 7.2.3.13 Disconnect the fill tubes and remove from the shield assembly. 7.2.3.14 Reinstall Shield Closure Plugs. See Pa ragraph 7.1.2. 7.2.3.15 Perfara s t e p s 7.2.2.15, 7.2.2.17, a nd 7.2.2.16. 7-6 1 L
Il ~' ') b' 0 0 \\ 7.2.4 Assembly Verification Leak Test 7.2.4.1 Verify test port as being open. 7.2.4.2 Install test port tool. 7.2.4.3 Attach pressurized freon source. 7.2.4.4 Pressurize cavity with freon to 20-25 psig and allow setting time of approximately 5 minutes. 7.2.4.5 Verify that pressure is holding (no gross leaks). 7.2.4.6 Close test port plag, captaring pressurized freon inside the cavity. 7.2.4.7 Remove the test port tool and allow the released freon to dissipate. 7.2.4.8 Using a freon detector, check the area around the lid cir-cumference and the test port plug. 7.2.4.9 If the leak rate is below the sensitivity of this leak test, assembly has been verified. If the leak rate is shown to be greater than 10-3 sta-em /sec., the secondary containment 3 vessel should be checked for improper assembly and retested. 7-7
Pn0PRg2TARY DATA M ACCEPTANCE TESTS ate MAINTENANCE PROGRAM 1 l M Accentance Tests Prior to the first use of the packaging, the tests and evaluations called out on the General Arrangement Drawings, (Appendix 2.10.1 and 2.10.2) shall be performed. Shield integrity shall be verified using the gsama scan procedures des-cribed in Appendix 8.3.1. Wold integrity on the sample vial shall be demonstrated by subjecting it to an internal pressure of 375 psi for 5 minutes, after which the welds shall be visually inspected for any signs of deformation. Then, both the sample vial and the secondary ccntainment vessel shall be tested to the requirements of LT-09, Helium Leak Test (Appendix 8.3.2). Acceptance tests for the N-55 outer overpack are identical to those for the N-55, Certificate of Compliance No. 9070. M Maintenance Picaras General maintenance procedures are as follows: Painted Surfaces A. Painted surfaces may be wiped clean using standard chemical solutions and procedures. B. Chipped or scratched surfaces shall be repainted as follows: l I e 8-1 L
> 1 ,a 1. Remove rust or loose coatings and sand edges so they f air into sound conting. 2. Apply two costs Mobil Chen 89W9 or suitable equivalent to bare surfaces, following manufacturer'. recommendations. Unnafnted Surfaces l A. Unpainted carbon steel surfaces shall be coated with a generous coat of high quality vacuum grease. These areas include: 1. 0-ring glands on the Secondary Containment Vessel lid and body. 2. Secondary Containment test port. 3. Valve operator recess on side of sample shield. B. Vacuum grease shall be removed and replaced yearly. Grease may be removed using solvents recommended by the manufacturer of the grease. All 0-rings shall be removed prior to the use of any solvents. Fasteners All threaded parts shall be inspected yearly and after each use for deformed or stripped' threads. Any damaged parts shall be replaced prior to l further use. l l O-Rinns All 0-rings shall be inspected yearly for unusual wear, damage and dimen-sional consistency. 0-rings shall be replaced at least every five years, or as required from inspection. 8-2
f I l l 0-ring glands shall be inspected yearly for rust, chips, burrs, scratches, etc. Any such condition shall be corrected and the gland retested for Icak-tightness per the applicable portions of LT-09 (Appendix 8,3.2). Lead-Vermiculite Filler The lead shot and vermiculite mixture shall be replaced yearly to insure the vermiculite's absorption capabilities. The following procedure shall be used: 1. Remove old shot and vermiculite by inverting the shield with the shield closure plug removed. A nylon choker strap is re c o m:s end e d. Trace quantities of lead shot and/or vermiculite may remain with no adverse effects. 2. Measure out 350 to 375 cubic centimeters of commercial vermica-lite and grind into a fine powder. 3. Nix ground vermiculite thoroughly with cnough lead shot to fill the shield cavity. 4. Carefully pour the mixture into all spaces within the shield cavity and fill to the upper edge of the cavity. 5. Replace shield plug. Valve Operators Valve operators shall be checked for free operation immediately following replacement of the lead-vermiculite filler. Operators shall be turned through their complete range of movement three times to verify free move-If any 'bnormal resistance to movement exists, or if the operator ment. a cannot be turned through its entire range, contact the manufacturer of the sample shield (Nuclear Packaging, Inc.) for direction to correct the mal-function. 8-3
l - ~ A_{3 W r m ef('P,., s 1 ,.i2. w' # $'ICQ j,,2;,1 f, -u t M APPENDII APPENDIX 8.3.1 DISCUSSION DE GA)DIA SCAN PROCEDURE Lead shielding integrity shall be confirmed via gamma scanning. There are two gamma scan techniques utilized. The main difference is in the method utilized to deterairo acceptance criteria. Both Gamma Scan Techniques are exactly the same in all other respects and are conducted as follows. An Eberline E120 probe or equivalent is used to scan the outer surf ace of the cask while an Iridium 192 or Cobalt 60 source of sufficient strength is present in the center of the cask. The source is first placed on the bottom of the cask while the surface is scanned around its circumference parallel to the source. The source is then moved up a pre-determined distance and tbs circumference scanned again. This sequence is repeated until the entire cask surface is scanned. For these tests, the cask surface is gridded and a chart is made to reflect the gridded cask surf ace. The readings obtained in the cask grid as des-cribed above are recorded in the corresponding grid on the chart. This data then serves as the raw gamma scan results. All readings are in Milliroentgens (MR) The readings are evaluated by comparing them to predetermined MR values for nomical, or as designed, lead thickness and nominal -10% lead thickness. The two dif ferent methods utilized to determine acceptance criteria are discussed belee. The Laboratory Calibration Method (NuPac Procedure GS-001) utilizes test blocks of the cask wall made up of Icad and steel sheets. The test blocks simulate nominal or as designed and -105, lead thicknesses. The source is placed behind the test block at a distance equal to the inside radius of 8-4 i n -n.- ,, - - - +..., _. _. - -. _, - -,..,,., - -. ,,.,....--_,-..n,-,.,. ,n
~_, the cask. The probe is then placed on the outside of the test block and readings are taken. This sequence is repeated on the nominal and -10% test blocks and the data is recorded. The resultant values are then averaged. A ratio of the values is also developed. Then the average value is multiplied by the ratio. The value so derived is the maximum acceptable value for the shielding to be inspected. s The Field Calibration Method (NuPac Procedure GS-002) utilizes a specially fabricated test lid which incorporates a holder for various lead and steel sheet thicknesses. This fixture is installed onto the cask to be scanned. The test lid is then set up to simulate the nominal lead thickness, the source is placed below the test lid in the cask at a distance equal to the inside radius of the cask. Readings are then taken. The test lid is then set up to recreate the -10% lead thickness configuration, and readings are again taken. Other readings are then taken in 1/8 inch lead thickness increments between and beyond the two base readings until four to eight readings are obtained. The data is thea plotted on a chart of readings versus lead thickness. The value for nominal lead -10% is then utilized as l the maximum acceptable reading during the actual gamma scar., l l l l 8-5
APPENDIX 8.3.2. I NUCLEAR FN PACKAGING, INC. ) PROPR:ETARY DATA NUPAC PAS-2 POST ACCIDENT SAMPLING CASK HELIUM LEAK TEST PROCEDURE LT-09 Rev. 1 March 30, 1983 si e4 L
- 3. MO e
pnp ed Bi ' Date h /lN4All' 3/ /W /Elrfgineerincj Date 7308 30 Qual W ance Date ab Mahu acturing/ Production Date Other Date s 35/83 Document Conf.Y61/ Release Date 8-6
PROPRDETARY DATA 1.0 SCOPE This procedure describes the requirements for performing a helium leak test on the NuPac PAS-2 Post Accident Sampling Cask system. 2.0 REFERENCE DOCUMENTS 2.1 ANSI N14.5 Leakage Tests on Packages for Shipment of Radioactive Materials 2.2 QP-5 NuPac Quality Procedure, Quality Planning 2.3 QP-6 NuPac Quality Procedure, Inspection and Verification l 2.4 QP-7 NuPac Quality Procedure Descrepency Reporting and Control 3.0 REQUIREMENTS 3.1 All internal surfaces shall be flushed per ANSI l N45.2.1, Class A & B, after final as.9embly acceptance as follows: 1 l a. Flush at =1.0 GPM until a 20 mesh screen shows no more than slight particle specking j or rust staining. No particles larger than 1/32 inch in any dimension shall be allowed, except that fine hairline slivers less than 1/32 inch up to 1/16 long may be allowed. 3.2 Leak detector shall be capable of detecting a -8 leak of 5 x 10 standard cubic centimeters per second or smaller. 1-8-7. [
NOPRETAsy DATA 4.0 PROCEDURE 4.1 Sample Vial Leak Test 4.1.1 Calibrate leak detector according to manu-facturer's recommendations, such that the leak -8 detector sensitivity is 5 x 10 standard cubic centimeters /second (scc /sec) or better. 4.1.2 Clean the sample vial and secondary contain-ment vessel so that they are free from all foreign materials such as dirt, grease or oils. 4.1.3 Set the valve such that the flow direction through both valves is directed through the large sampling cavity (arrows pointing opposite each other). 4.1.4 Install a plug (GF-20-02D-A3) in one flow port 6n the vial. 1 4.1.5 Using appropriate fittings attach the leak detector to the other flow port on the sample vial. 4.1.6 Evacuate sample vial cavity for one hour, or until vacuum is sufficient to operate the leak detector according to manufacturer's recommendations. Note any adjustments made to the connecting - fittings required to achieve this. 4.1.7 Envelop the sample vial as assembled in plastic and fill the enclosed space with helium, taking care to purge all other gases from any pockets or cavities adjacent to the sample vial. 8-8
PROPRDETARY DATA 4.1.8 Determine the leak rate of the system using the leak detector manufacturer's recommendations and so note. 4.1.9 If leak rate of the system is determined to be greater than 1 x 10' sec/sec, inspect system for cleanliness and tightness of connecting fittings. Return to step 4.1.6 above. Continue testing until the vial passes the test or it is apparent that the system cannot be made to be leak-tight to the indicated level. Note the best system leak tightness observed. 4.1.10 Release vacuum and disassemble from leak detector. 4.2 Secondary containment Vessel Leak Test 4.2.1. Install 0-ring on the lid of the secondary containment vessel. Install the lid onto the body of the secondary containment vessel and l ' secure with the proper bolts. l 4.2.2 Install the test port sampling tool in the test port on the lid. Attach to the leak detector using appropriate fittings. Using sampling tool, adjust test port closure bolt such that the stat-o-seal does not seal. 4.2.3 Evacuate the system for one hour or until vacuum is sofficient to operate the leak detector as per manufacturer's recommendations. Note any adjustments made to the connecting fittings required to achieve this. 8-9 l
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^ PROPMETARY DATA 4.2.4 Envelop the secondary containment vessel in plastic and till the enclosed space with helium, taking care to purge all other gases from any pockets, or cavities adjacent to the vessel. 4.2.5 Determine the leak rate of the system using the leak detector manufacturer's recommendations and so note. 4.2.6 If the leak rate of the system is determined ~ to be greater than 1 x 10 sec/sec, inspect system for cleanliness, tightness, and proper assembly. Return to step 4.2.3, above, and continue testing until the vessel passes the test or it is apparent that the system cannot be made leak-tight to the indicated level. Note the best leak tightness observed. l i .4.2.7 Using the sampling tool handle, adjust the Test Port Closure Bolt such that the stat-o-seal is l properly seated. l i 4.2.8 Release the vacuum to the sampling tool. l l 4.2.9 Connect sampling tool to a helium source, taking care to adequately seal connecting fittings to prevent significant leakage of air into the system. 4.2.10 Adjust test port closure bolt to allow free passage of helium into the Secondary Containment Vessel. i l 4.2.11 When the internal pressure of the system reaches one atmosphe: e, again adjust the Test Port Closure Bolt so that the stat-o-seal is properly seated. 4.2.12 Disconnect from helium source and reconnect to the leak detector. L -(\\- fil-JL@
. m._ _-.4.- _,c PROPRETARY DATA 4.2.13 Repeat steps 4.2.3 through 4.2.6. 4.2.14 Release vacuum and disassemble from leak detector. i 5.0 ACCEPTANCE CRITERIA 5.1 An inspection report shall be prepared in accordance with QP-5 and QP-6, describing the system and giving the part names and numbers for each component tested. The brand name, serial number and calibration date of the leak test shall be recorded. Actual leak rates obtained shall also be recorded for each unit tested 5.2 To be acceptable, the units shall exhibit a leak rate less than 1 x 10-sec/sec. Any leaks greater than that rate shall be recorded, corrected and retested a maximum of three times. After the third failure, a' Quality Descrepancy Report /Supolier Disonnition Request (QDR/SDR) shall be prepared for disposition in accordance with QP-7. If testing is,, performed sub-sequent to any destructive testing, such as a drco test, a QDR/SDR shall be prepared after the first failed leak test. i w e ~--~ w m--, n y-,+-n u
E (DD $5?k?qy I;.- a
- ,, in 9.0 QUALITY ASSURANCE NuPac's - qua li t y assurance program used for the design, fabrication, assembly, testing, use and maintenance of the NuPac PAS-2 cask is designed and administered to meet the 18 criteria of 10 CFR 71, Appendix E.
A description of this program has been submitted to the NRC under NuPac letter QA-78-1, Rev. 1, dated July 31, 1980, and has received Quality Assurance Program Approval No. 0192. e 4 9-1 . - - - -}}