ML20059E426

Certificate of Compliance 9256,Rev 0,for Model Ynps SG Package

ML20059E426
Person / Time
Site:	Yankee Rowe, 07109256
Issue date:	10/28/1993
From:	Chappell C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20059E411	List: ML20059E408 ML20059E426
References
NUDOCS 9311030197
Download: ML20059E426 (12)

Text

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' j U.S. NUCLEAR RESULATORY COMMISSIEN l

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CERTIFICATE OF COMPLIANCE l -

l So crR rt FOR RADIOACTIVE MATERIALS PACKAGES g

l D. REVISION NUMBER

c. PACKAGE IDENilFICATION NUMBER o PAGE NUMBER

e. TOTAL NUMBER PAGE'S

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USA /9256/A 1

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cert:8 cate is issued to cerbfy that the packagmg and contents described in item $ below, meets the apphcable safety standards set forth in Title 10, Code I

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of Federal Regulations, Part 71," Packaging and Transportation of Radioactive Matenal' I

h' This certihcate does not rebeve the consignor from comphance with any requirement of the regulations of the U.S. Department of Transportation or other I

apphcable regulatory agencies, meluding the government of any country through or into which the package will be transported.

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l t TmsgTi,gys;ssgUy THE BASIS OF A SAFETY ANALYSIS RE RT OF THE G

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! Yankee Atomic Electric Company Yankee Atomic Electric Company application l

l 580 Main Street dated April 12, 1993, as supplemented, y

i Bolton, MA 01740-1398 I

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~71-9256 l

c.oOu eTNUusER l

1. E ek%ie ia condiconai upon fuiviihng the requirements of io crR eart 71. as apphcabia and the conditions specified be'ow-I I

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(a)

Packaging i

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l (1) Model No.:

YNPS Steam Generator l

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I (2)

Description g

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I A steam generator filled with low density concrete, with seal-welded nozzle l

l covers, impact limiters, and. shear. keys.; The steam generator consists of the I

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steam generator shell, internal U-tubes and tube supports. -The steam l

generator is essentially cylindrical with an 00 of approximately 8 feet 6-1/4 g

I inches in the steam drum region-and an OD of approximately 7 feet 1 inch in g

I the tube bundle region,1The shell. wall thickness is -approximately 2-3/4 I

I inches in the tube bundle: region'and 3-1/8 inches in-the steam drum region.

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The length of the steam generator package is approximately 40 feet 7-3/8 I

I approximately 3/4 inch, and an average wall thickness of 0.072 inch. The

' l inches. The tube bundle is composed of 1620 U-tubes, which have an OD of i

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nozzles and other penetrations are covered with welded closures. The steam I

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generator is filled with low density. concrete on both primary and secondary l

l sides. Three shear keys are welded to'the steam generator shell bottom for I

axial restraint. Toroidal impact limiters, which are polyurethane foam with a I

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carbon steel shell, are fixed at each end of the steam generator.

Steel I

plates are welded onto the steam generator shell for radiation shielding, as i

i required. The maximum weight of the package is approximately 273,000 pounds.

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(3) Drawings j

The package is constructed and assembled in accordance with the following l,

I Chem-Nuclear Systems, Inc. Drawing Nos.:

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,e,

I C-110-B-46063-1, Rev. 1 W

l C-110-B-46063-2, Sheets 1 through 4, Rev. 0 W

i C-110-D-46063-3, Sheets 1 and 2, Rev. 0 I

C-110-A-46063-4, Rev. 0 I

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l 9311030197 931028 l

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[DR ADOCK 07109256 [pg _______________-------------- ---

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i jdj Page 2 - Certificate No. 9256 - Revision No. 0 - Docket No. 71-9256 d

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l (b) Contents h

ll (1) Type and form of material y

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Steam generator containing radioactive contamination, filled with low density l

g concrete, meeting the requirements of low specific activity radioactive material.

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H l li (2) Maximum quantity of material per package n

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Greater than a Type A quantity of radioactive material.

Fissile material may l

l be present provided the fissile material meets the exemption standards of 10 l

CFR 671.53.

l 6.

In addition to the requirements of Subpart G of 10 CFR Part 71:

H l

(a)

The package must be prepared for shipment and transported in accordance with Chapters 7 and 8 of the application.

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l (b)

The package must be transported in accordance with the operational controls of i

Table 1.1 of the application.

(c)

Prior to transport, shielding must be welded onto the package in accordance with Chem-Nuclear Systems, Inc. Drawing No. C-110-B-46063-1, Rev.1, as j

necessary, such that the package meets the external radiation standards of 10 CFR 971.47.

7.

This certificate authorizes a one-time shipment for each of four packages from the Yankee Nuclear Power Station to a point near Barnwell, South Carolina.

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8.

The package authorized by this certificate must be transported on a motor vehicle j

and on a railroad car assigned for the sole use of the licensee.

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i 9.

The package authorized by this certificate is hereby approved for use under the j

general license provisions of 10 CFR 671.12.

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10.

Expiration date:

September 30, 1998.

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i REFERENCES Yankee Atomic Electric Company application dated April 12, 1993.

5 Supplements dated:

April 20, July 30, August 2, and September 10, 1993.

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FOR THE U.S. NUCLEAR REGULATORY COMMISSION

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Cass R. Chappell, ection Leader i

Cask Certification Section N

i Storage and Transport Systems Branch l

Division of Industrial and i

OCT 2 81953 Medical Nuclear Safety, NMSS y

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION o'

WASHINGTON. D.C. 20565

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SAFETY EVALUATION REPORT Model No. YNPS Steam Generator Package Certificate of Compliance No. 9256 Revision No. O j

SUMMARY

By application dated April 12, 1993, as supplemented, Yankee Atomic Energy Company requested approval of the YNPS (Yankee Nuclear Power Station) Steam Generator as a transportation package for greater than a Type A quantity of low specific activity radioactive material. Based upon the statements and representations contained in the application and the conditions listed below, i

we have concluded that the YNPS Steam Generator package meets the requirements of 10 CFR Part 71.

REFERENCES Yankee Atomic Electric Company application dated April 12, 1993.

Supplements dated: April 20, July 30, August 2, and September 10, 1993.

DESCRIPTION A steam generato-filled with low density concrete, with seal-welded nozzle covers, impact

.aiters, and shear keys. The steam generator consists of the steam generator shell, internal U-tubes and tube supports. The steam generator is essentially cylindrical with an OD of approximately 8 feet 6-1/4 inches in the steam drum region and an OD of approximately 7 feet 1 inch in the tube bundle region.

The shell wall thickness is approximately 2-3/4 inches in the tube bundle region and 3-1/8 inches in the steam drum region.

The length of the steam generator package is approximately 40 feet 7-3/8 inches. The tube bundle is composed of 1620 U-tubes, which have an OD of approximately 3/4 inch, and an average wall thickness of 0.072 inch. The nozzles and other penetrations are covered with welded closures.

The steam generator is filled with low density concrete on both primary and secondary sides. Three shear keys are welded to the steam generator shell bottom for axial restraint. Toroidal impact limiters, which are polyurethane foam with a carbon steel shell, are fixed at each end of the steam generator.

Steel plates are welded onto the steam generator shell for radiation shielding, as required. The maximum weight of the package is approximately 273,000 pounds.

DRAWINGS The package is constructed and assembled in accordance with the following Chem-Nuclear Systems, Inc. Drawing Nos.:

C-Il0-8-4606*i-1, Rev. 1 C-Il0-B-46063-2, Sheets 1 through 4, Rev. O C-110-D-46063-3, Sheets 1 and 2, Rev. 0 C-Il0-A-46063-4, Rev. 0 4

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YNPS STEAM GENERATOR PACKAGE

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3 CONTENTS A.

Type and form of material Steam generator containing radioactive contamination, filled with low density concrete, meeting the requirements of low specific activity radioactive material.

B.

Maximum quantity of material per package Greater than a Type A quantity of radioactive material.

Fissile material may be present provided the fissile material meets the exemption standards of 10 CFR 971.53.

STRUCTURAL EVALUATION The YNPS Steam Generator package contains greater than a Type A quantity of low specific activity radioactive material. Consequently, the structural evaluation of the package must demonstrate that the design meets the performance requirements of 10 CFR Part 71 for normal conditions of transport.

A.

Gineral Standards for all Packaaes Minimum oackaae size No dimension of the package is less than 10 cm (four inches).

The package meets the requirements of 10 CFR $71.43(a) for minimum size.

Tamper-oroof feature The steam generator package closures are all seal-welded. The intact welds are evidence that the package has not been opened by unauthorized-persons.

i Positive Closure Since the steam generator package closures are seal-welded, unintentional opening of the package is precluded.

Chemical and Galvanic Reactions There are no materials which will create significant chemical, galvanic, or other reactions among the packaging components, or between the packaging components and package contents, including reactions resulting from inleakage of water to a maximum credible extent.

Valves or other Devices There are no valves or other devices on the package. All containment penetrations are welded closed.

. B.

Liftino and Tie-down Standards for all Packaae liftina Devices The package is placed in a horizontal position while being prepared for shipment and is not lifted or moved into any other orientation during transport.

There are no permanent welded attachments to the package that can be used in lifting the package during transport.

Tie-down Devices During transport, the package rests on two support saddles made of structural steel. The package is secured to each saddle by five,1-3/8 inch diameter pre-tensioned steel cables. The saddles, along with the pre-tensioned steel cables, resist inertia forces associated with transverse and vertical accelerations.

The inertia force associated with the longitudinal acceleration of the package is resisted by the shear key assembly. Three longitudinal shear keys, which are 6 inches wide and 26 inches long, are welded to the bottom of the package. The shear keys are the only parts of the tie-down system that are structural parts of the steam generator shipping package. Therefore, the shear keys are designed to withstand a longitudinal acceleration of 10 g, as required in 10 CFR 671.45. Tie-down system components that are not a structural part of the package were designed in accordance with the acceleration criteria of ANSI N14.2, " Proposed American National Standard For Tie-Down for Truck Transport of Radioactive Material." The design allowable stresses were based on the AISC standards, which are more conservative than ANSI N14.2.

C.

Normal Conditions of Transoort diill There will be no significant stress increase in the package under the normal transport heat condition.

fald Thermal stress in the package was evaluated by assuming that the ambient temperature was reduced from 100" F to -40 F in one hour (the maximum range of hot and cold environments specified in 10 CFR Part 71).

Thermal stress in the steam generator shell was well below the allowable design stress. The primary material of construction of the steam generator vessel is ASTM A-212, Gr. B, which meets " normalized to fine grain practice," as specified in Regulatory Guide 7.11 for prevention of brittle fracture. The penetration covers are constructed of A516, Gr.

70, steel which also meets the " normalized to fine grain practice" standard.

The tube sheet and channel head forging, which are constructed of ASTM A-266, Gr. II, are greater than 4 inches thick. The criteria of Regulatory Guide 7.11 are only applicable to sections less i

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. than four inches thick. However, the stress intensity in the tube sheet and channel head forging is small for a one-foot drop (below 6,000 psi),

and ASTM A-266 is authorized for use under ASME Section III. Coupled with the additional safety controls used during transport of the package (Table 1.1 of the application), it can be concluded that brittle fracture will not occur during transport.

Reduced External Pressure ard Increased External Pressure The applicant performed finite element analyses to show that stresses in the steam generator vessel were small due to reduced and increased external pressure, and were well below allowable values.

Shock and Vibration There are no valves or other devices on the steam generator vessel that are susceptible to shock and vibration loads.

Also, the radioactive contents are encapsulated in concrete There will be no significant adverse effects on the package from the shock and vibration loads normally incident to transport.

Water Soray The steam generator package closures are seal-welded. This test condition has no effect on the package.

Free Droo Because of the large size and weight of the steam generator package, the only realistic drop orientation is the orientation in which the package is positioned during transport.

Since the steam generator package will always be oriented in the horizontal position during transport, the package was evaluated for a one foot drop in the horizontal orientation.

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The YNPS Steam Generator package has two, 12 lb/ft rigid polyurethane foam impact limiters to absorb the drop energy and to reduce the impact force on the steam generator. The stress-strain data for the

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polyurethane foam was provided by the vendor (General Plastic).

To account for changes in material properties due to density and

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temperature variations, a 115 percent margin was assumed to bound the response of the impact limiters.

An energy balance approach was used to estimate the amount of crush (deformation) and reaction force developed in the impact limiters during the one foot free drop onto an unyielding horizontal surface.

The maximum impact limiter deformation was used to confirm that the maximum strain of the polyurethane foam was below the design limit of 70%, and to confirm that none of the steam generator penetrations would directly contact the rigid surface during the free drop. The stresses in various components of the steam generator package during the one-foot free drop event were calculated with a combination of hand calculations and finite element analyses.

Stresses in the steam generator vessel shell away

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. from the impact region were calculated by assuming the vessel acted as a beam. A finite element analysis using ANSYS was performed to evaluate the stresses in the package at the point of impact, i.e., underneath the impact limiter. The maximum stress intensity of 16.0 ksi occurred in the steam generator shell. This stress intensity was well within the allowable stress of 1.5 S, (34.95 ksi) as prescribed in Regulatory Guide 7.6.

Stresses in all other components of the steam generator package were also well within the allowable design stresses. Thus, the package meets the one foot free drop condition.

Corner Droo This test is not applicable because the weight of the package exceeds 200 pounds and neither wood nor fiber board is used as a material of construction.

Comoression This test is not applicable for packages weighing more than 5000 kg.

Penetration The applicant provided an analysis to show that the penetration test will have negligible effect on the package.

D.

Conclusion The package was evaluated against the applicable requirements of 10 CFR Part 71. Based on the results of the evaluation, it was concluded that the structural design of the package meets the performance criteria of 10 CFR Part 71.

THERMAL The decay heat in the package is negligible.

CONTAINMENT The containment vessel boundary is defined as the steam generator shell and the nozzle closures. The containment vessel is welded closed. Void spaces within the vessel are filled with low density concrete. The containment vessel remains intact under the normal conditions of transport as described above, and there is no loss or dispersal of radioactive material.

EVALUATION OF RADI0 ACTIVE CONTENTS AND SHIELDING The radioactivity present in the package is in the form of a layer of activated corrosion products on the primary side surfaces of the steam generator (inside of the tubes and the channel head). Dose rate and contamination measurements were made to determine the magnitude and isotopic distribution of the radioactivity present in the steam generators.

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i A.

Contents Determination The applicant performed radiation dose rate measurements of the steam generators in January and February, 1993.

The maximum contact dose rate was 340 mR/hr. The source term was based on the maximum measured dose rate. The radionuclide distribution was determined from a contamination sample taken from the primary side of the steam generator in 1992.

The computer code MICR0 SHIELD was used to calculate the dose rate on the outside of the steam generator per curie, using the isotopic distribution of the contamination survey. The measured dose rate was divided by the dose rate per curie to obtain the source term in the steam generator tubes.

The source term in the channel head was calculated using the ratio of the surface area of the tubes to the surface area in the channel head. The shielding analysis used the sum of these two source terms.

The total radioactivity was estimated as 441 curies. The predominant nuclides were iron-55 (229 curies), nickel-63 (102 curies), and cobalt-60 (93 curies).

Fissile materials were present in trace quantities, but were well below the fissile exempt quantities of 10 CFR 071.53.

The NRC staff performed a confirmatory analysis to determine the source term, using the applicant's values for the dose rate and the isotopic fractions in the application.

The source terms were verified by the staff using the MICR0 SHIELD computer code.

Using the calculated source terms, the applicant determined the concentration of radioactivity in the package.

Considering only the weight of the low-density concrete in the tube region, the concentration of radioactivity was determined to be within the concentration limits for low specific activity defined in 10 CFR 571.4. The weight of the steam generator tubes, tube supports, and shell was conservatively ignored in determining the concentration.

B.

Shieldino Calculation The source term was used to determine the external dose rates for the package.

The free space in the steam generator is fil ed with low density concrete, with a minimum density of 0.336 g/cm].The tubes, low density concrete, air, and downcomer were homogenized in the computer model. The outer steel wall was modeled explicitly. The applicant used MICR0 SHIELD to calculate the dose rates at the package surface and 2 meters.

Results of the applicant's shielding analysis showed that the external dose rates at the top and bottom ends of the steam generator met the 10 CFR 571.47 external radiation standards. However, the dose rates at the surface and two meters from the conveyance exceeded the limits, without additional shielding.

Shielding will be welded onto the package surface, as necessary, to meet the dose rate limits. The applicant estimated that a maximum of 2.5 inches of additional steel l

s.-

.. shielding may be necessary to comply with the radiation standards in 10 CFR 571.47. The dose rates were verified by NRC staff using the MICR0 SHIELD computer code.

tea table below summarizes the results of the applicant's and staff's shielding analyses.

EXTERNAL DOSE RATES (mrem /hr) location Aoolicant HRG Package surface Side (with 2.5" 21 22 steel shielding)

Top 0.033 Bottom 0.2 2 meters from surface Side (2 m from conveyance) 6.6 7.7 Top (2 m from package) 0.017 Bottom (2 m from package) 0.071 The Certificate of Compliance has been conditioned to require shielding to be added to the steam generators to meet the external radiation standards of 10 CFR f71.47. The shielding is in the form of steel plates up to 2.5 inches thick welded to the steam generator shell.

CRITICALITY Trace amounts of fissile materials may be present. The Certificate of Compliance limits fissile materials to the exempt quantities defined in 10 CFR 971.53. Therefore, criticality is not a concern.

FABRICATION EVALUATION The YMPS steam generators were fabricated in accordance with the ASME Code.

These fabrication criteria are acceptable for a Category III transportation package. New welds will be made and inspected according to the ASME Code as specified on the package drawings referenced in the Certificate of Compliance.

OPERATING PROCEDURES, ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Four steam generators will be shi;~ K o om the Yankee Nuclear Power Station site near Rowe, Massachusetts, to u - disposal site near Barnwell, South Carolina. Each package will be lond J rnto a motor vehicle designed for heavy loads. The steam generator will be,i.*nsferred to the rail car from the trailer by cribbing the steam gener...e over the rail track, moving the trailer from under the steam generator, and positioning the rail car under the package. These procedures will also be used in reverse to transfer the packages from the rail car to the trailer for road transport to the disposal facility.

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Chapter 7 and Table 1.1 of the application describe the operational controls which will be used Juring transport of the package. Some of operational controls are as follows:

(1) Maximum speed of 5 mph for the motor vehicle leg of the shipments, (2) package not lifted by crane during modal transfers, (3) use of a dedicated train for the rail leg of the shipments, (4) shipments accompanied by qualified health physics personnel.

Chapter 8 of the application describes the acceptance tests which will be performed on the package prior to transport. These tests include:

(1) radiation measurements which confirm that the contents meet the concentration limits for low specific activity material, (2) visual inspection of the closure plates and welds, (3) external radiation measurements which show the package meets the standards in 10 CFR 571.47, and (4) contamination surveys which show the package meets the standards in 10 CFR 571.87.

SHIPPING ROUTE The steam generators will be transported by special heavy haul transporter from the Yankee Nuclear Power Station, near Rowe, Massachusetts, to the rail line at the Hoosac Tunnel, near the town of Florida, Massachusetts. After leaving the YNPS site the transporter will cross an earthen dam and a bridge,-

which are privately owned by New England Power Company. A tewporary bridge will be placed over the existing bridge, to prevent overloading the existing bridge. The transporter will then travel approximately six miles over county-owned highway. At the Hoosac Tunnel, the steam generators will be transferred to a rail car for shipment to Barnwell, South Carolina. Shipment will be by dedicated train.

The applicant has evaluated the shipping route from the site to the rail line to assure the adequacy and safety of the roadway, the culverts, the earthen dam, and the temporary bridge (supplements dated August 2 and September 10, 1993). The Federal Highway Administration, in cooperation with the State of Massachusetts Highway Department, has reviewed and confirmed the adequacy of the roadway within Massachusetts and the temporary bridge. The evaluation of the dam was reviewed by NRC staff ' u found to be adequate.

In response to a recommendation by the Massachusetti

'y Department, the applicant will perform a dry run of the roadway rot:

i e r to transporting a steam generator. The dry run will consist ot Juading the transporter with a non-radioactive load exceeding the weight of the steam generator package. During the dry run, the transporter will travel across the earthen dam, across the temporary bridge, and along the same highway route as the actual shipments.

After shipment by rail, the packages will again be moved by the transporter for approximately two miles. A heavy haul permit will be ob' ained from the t

State of South Carolina for this movement.

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. CONDITIONS 1.

In addition to the requirements of Subpart G of 10 CFR Part 71:

)

(a) The package must be prepared for shipment and transported in j

accordance with Chapters 7 and 8 of the application.

(b) The package must be transported in accordance with the operational j

controls of Table 1.1 of the application.

f (c) Prior to transport, shielding must be welded onto the package in accordance with Chem-Nuclear Systems, Inc. Drawing No.

C-Il0-B-46063-1, Rev. 1, as necessary, such that the package meets the external radiation standards of 10 CFR 571.47.

2.

This certificate authorizes a one-time shipment for each of four packages from the Yankee Nuclear Power Station to a point near Barnwell, South Carolina.

3.

The package authorized by this certificate must be transported on a motor vehicle and on a railroad car assigned for the sole use of the licensee.

CONCLUSIONS l

Based on the review of the statements and representations contained in the application, as supplemented, and the conditions listed above, we have concluded that the YNPS Steam Generator package meets the requirements of 10 CFR Part 71.

W Cass R. Chappell, S c n Leader Cask Certification Section Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety, NMSS Date j

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