ML20030A444

Mechanical Design & Testing of Development Fuel for CPC Big Rock Point Reactor

ML20030A444
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/30/1962
From:	Fowler W, Naymark S, Pashos T GENERAL ELECTRIC CO.
To:
References
CON-AT(04-3)-361, CON-AT(4-3)-361 GEAP-3851, NUDOCS 8101090519
Download: ML20030A444 (80)

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-s=y g, GEAP-3851 AEC Research and Development Report April,1962 MECHANICAL DESIGN AND TESTING OF DEVELOPMENT FUEL FOR CONSUMERS POWER BIG ROCK POINT REACTOR by W. D. Fowler Commission U. S. Atomic Energy (04-3)-361 Contract No. AT ATOMIC POWER EQUIPMENT DEPARTMENT GEN ER AL $ ELECTRIC SAh JOSE,CAllFORNIA Printed in U.S. A. Price $2.00. Available from the Office of Technical Services, Department of Commerce, Washington 25, D. C.

GEAP-3851 Prepared by:

$40 A1&L W. D. Fowler

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Fuels and Materials Development Vallecitos Atomic Laboratory Approved:

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T.'J. 'Pashos #

V Manager - Fuel Development 4GEk (,

S. Naymdk Manager - Fuels and Materials Development

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L. K. Holland 446-T102 E Project Engineer v

LEGAL NOTICE This report uas prepared as an account of Gos ernment sponsored work, Neither the United States, nor the Commission, nor any person acting on behalf of the Commission:

A. Makh any uarranty or representation, expressed or implied, uith respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe prisately oscned rights; or B. Assumes any liabilities uith respect to the use of, or for dam.

ages resulting from the use of any information, apparatus, method, or process disclosed in this report.

As usedin the abose," person acting on behalf of the Commission" includes any employee or contractor of the Commission, or em.

playee of such contractor, to the extent that such employee or contractor of the Commission, or employee of such contractor prepares, disseminates, or prosides access to, any information pursuant to his employment or contract uith the Commission, or his employment scsth such contractor.

I.N 1 I

GEAP-3851 EXTERNAL DISTRIBUTION i.

No. of Copies U. S. Atomic Energy Commission 5

Washington 25, D. C.

Attention: Chief, Water Reactors Branch Civilian Reactors, DRD U. S. Atomic Energy Commission 1

Washington 25, D. C.

Attention: Chief, Water Systems Project Branch Army Reactors, DRD U. S. Atomic Energy Commission 1

Washington 25, D. C.

Attention: Chief, Fuels and Materials Development Branch Nuclear Technology, DRD I

General Nuclear Engineering Corporation 1

P. O. Box 245 Dunedin, Florida Attention: J. M. West Combustion Engineering, Inc.

1 Nuclear Division Prospect Hill Road Windsor, Connecticut Attention: W. H. Zinn Westinghouse Electric Corporation 1

Atomic Power Department P. O. Box 355 Pittsburgh 30, Pennsylvania Attention: Dr. W. E. Shoupp Allis-Chalmers Manufacturing Company 1

Atomic Energy Division

. Milwaukee 1, Wisconsin Attention: C. B. Graham Allis-Chalmers Manufacturing Company 1

Nuclear Power Department P. O. Box 8697 Washington 11, D. C.

s Attention: H. Etherington Northern States Power Company 1

Minneapolis 2, Minnesota Attention: D. F. McElroy 11

__-.___________._.___m-

GEAP-3851 No. of Copies U. S. Atomic Energy Commission 2

Chicago Operations Office 9800 South Cass Avenue Argonne, Illinois Attention: C. A. Pursel, Director Reactor Engineering Division U. S. Atomic Energy Commission San Francisco Operations Office 2111 Bancroft Way Berkeley, California Attention: W. H. Brummett, Jr., Director 2

Contracts Division Attention: G. F. Helfrich, Director 1

Reactor Division Meyer Novick 1

Argonne National Laboratory P. O. Box 2528 Idaho Falls, Idaho Hoylande D. Young, Director 1

Technical Information Division U. S. Atomic Energy Commission 9700 South Cass Avenue Argonne, Illinois W. C. Cooper 5

Consumers Power Company 212 W. Michigan Avenue Jackson, Mic bigan U. S. Atomic Energy Commission 3

TechnicalInformation Service Extension (plus masters)

P. O. Box 62 Oak Ridge, Tennessee 111

GEAP-3851 TABLE OF CONTENTS Section I Introduction Section II Summary Section III Big Rock Development Fuel Assembly Design Details 3.1 Development Fuel Design and Performance Criteria 3.2 Development Fuel Design Description 3.3 Fuel Testing 3.4 Development Fuel Design Data 3.5 Advantages of the Design Section IV ConclusionsSection V Bibliography Appendix A Testing A.1 Fuel Rod Vibration Testing A.2 Fuel Assembly Tensile Testing i

A.3 Fuel Rod Fretting Test A.4 Spacer Tests A.5 Handling Tests Appendix B Analysis B.1 Water / Fuel Ratio B.2 Stainless Steel / Fuel Ratio B.3 Fuel Temperature as a Function of Heat Flux B.4 Heat Transfer Area Per Bundle B.5 Average Heat Flux at 45 kw/l Power Density B.6 Hydraulic Diameter Within The Fuel Channel B.7 Fission Gas Release, Containment and Clad Stress Appendix C Drawings l

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l GEAP-3851 LIST OF ILLUSTRA'170NS

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Title Page_

3-1 131g Rock Development Fuel 3-3 Assembly 3-2 Mechanical Arrangement of 3-4 Component Parts 3-3 VBWR Prototype of Big Rock 3-7 Development Fuel A-1 Tensile Testing Structure After A-3 Being Destructively Tested A-2 Specimen and Holder A-4 A-3 Macroscopic Results of Six A-6 Specimens A-4 a and b Metallographic Section A-7 A-4 e and d Metallographic Section A-8 A-4 e and i Metallographic Section A-9 A-5 Spacers Being Currently Tested A-11 in the VBWR A-6 Fuel Bundle Deformation Caused A-13 by Impact (Side Views)

B-1 UO Temperature B-4 2

B-2 Volumetric Mean Temp of UO B-5 2

Within A Fuel Rod B-3 Clad Stress Nomograph - 0.425 OD B-9 Powder Fuel Rod B-4 Clad Stress Nomograph - 0.320 OD B-10

' Powder Fuel Rod B-5 Clad Stress Nomograph - 0.425 OD B-11 Pellet Fuel Rod B-6 Clad Stress Nomograph - 0.320 OD B-12 Pellet Fuel Rod B-7 End of Life Clad Stress for Pellet B-13 RMs B-8 End of Life Clad Stress for Powder B-14 Rods V

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GEAP-3851 LIST OF TABLES 3-1 Design Data for the Big Rock Fuel Development Assembly 3-2 Clad Stress at the End of Fuel Life (10,000 MWD /T) for Pellet and Compacted Powder Type Fuel B-1 Design Data Relating to Fission Gas for Pellet and Compacted Powder Type Fuel Rod B-2 Fuel Rod Clad Stress at the End of Life Assuming A Sudden Loss of Reactor Pressure I

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GEAP-3851 SECTION I INTRODUCTION The development and irradiation testing of low cost fuel fabrication processes and mechanical assembly designs are presently being pursued under the AEC-sponsored High Power Density Program. Under the pro-gram, the irradiation of prototype fuel assemblies holding promise of having reduced fuel fabrication costs, long fuel life, and high power den-sity will first be conducted in the VBWR, Subsequently, irradiation of full-size assemblies will be conducted in the Consumers Big Rock reac-tor.

The development of fuel fabrication processes and fuel assembly designs having good cost reduction potential have been reported.(1-4)

Irradiation in the VBWR of nine fuel assemblies containing these low cost features is in progress. (5)

With the fabrication development program essentially complete, and the irradiation in VBWR well under way, there remains the design of fuel with low cost features for irradiation testing in the Consumers Big Rock reactor. The purpose of this report is to present a fuel assembly design having potential for reduced fuel fabri-cation costs to be irradiated in the Big Rock reactor at a power density of 45 kw/ liter (average) and an exposure of 10,000 MWD /T (average).

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1-1

GEAP-3851 SECTION II 9

SUMMARY

A fuel assembly design for the development fuel to be irradiation-tested in the Consumers Big Rock reactor under concitions of 45 kw/ liter and 10,000 MWD /T has been completed. The design includes novel fea-I tures that represent a 25 per cent reduction in fabrication costs. The fuel assembly design consists of an 11 x 11 array of fuel rods, free floating in a fuel assembly structure (see Figure 3-1). Design data for the fuel assembly are:

Cladding material 304 stainless steel Cladding thickness (inches) 0.010 Fuel rods per assembly 0.425 in. OD 109 0.320 in. OD 12 Fuel UO2 Enrichment (% U-235) 2.7% e 0.050 UO Weight per assembly (pounds) 383-395 2

Average heat flux at 45 kw/1 126,000 2

(Btu /hr-ft )

Water / fuel ratio 2.4 Hydraulic diameter (inches) 0.58 l

The development fuel offers the following advantages over the first core fuel design.

2-1

GEAP-3851 1.

The individual fuel rods are free floating (not fixed at either end of the assembly), and therefore easily insertable or re-movable.

2.

Spacers are of a single layer design not a part of or attached to the assembled fuel rods.

3.

The design permits the use of fuel rods having variations in rod length of about 3/4 inch. This makes the design amenable to the use of rods fabricated by powder compaction techniques where the final lengths vary slightly.

4.

The design permits the use of any type of end plug on the fuel rod without regard to attachment or location provisions to tie plates.

5.

A simplified structure design that represents a 25 per cent cost reduction.

A dummy development fuel assembly was fabricated and success-fully subjected to mechanical and flow tests. Three VBWR prototype assemblies containing features of the development fuel assembly design are presently undergoing irradiation testing. The tests are proceeding satisfactorily after 3000 MWD /T, as of May 1,1962.

Eight development fuel assemblies of the design described are scheduled for delivery to the Big Rock reactor by December,1962.

Fuel rods for four assemblies will be fabricated by the swaged pellet

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GEAP-3851 1

SECTION III l

BIG ROCK DEVELOPMENT FUEL ASSEMBLY DESIGN DETAILS The design of the developmental fuel for the Consumers Big Rock reactor must necessarily be within the parameters set by the physics, hydraulics, and heat transfer considerations established for the reactor first core. Within these boundary conditions, a set of criteria were de-fined for the developmental fuel.

3.1 Development Fuel Design and Performance Criteria 1.

Low cost mechanical assembly structure to hold and space the fuel rods.

4 2.

Individual fuel rods to be easily insertable and removable.

3.

Relaxation of stringent tolerances on fuel rod length.

4.

Elimination of threaded type end plug to reduce cost.

5.

Utilization of cladding purchased to commercial tolerances and specifications.

6.

Use of rod type, stainless steel clad, UO I"'I' 2

7.

Attainment of 45 kw/ liter (average) power density and 10,000 MWD /T (average) burnup.

3-1

GEAP-3851 3.2 Development Fuel Design Description The final fuel design that evolved from the criteria listed above and from development and testing work is illustrated in Figure 3-1.

Design drawings for the development fuel are shown in Appendix C.

Functionally, the component parts of the fuel assembly are shown in Figure 3-2, and break down into five areas:

Part Function Spacer 1.

To laterally position the fuel rods with respect to each other, the corner angles, and the channel in which the bundle will be placed.

2.

To add stiffness to the corner angles and tie them together.

Corner Angles 1.

To transmit tensile loads from the support to the weldment-retainer which forms the handle for the bundle.

2.

To axially position the spacers.

3.

To provide protection for the fuel rods during insertion and removal from channels by being the outermost part of the fuel bundle.

4.

To assist in controlling corner rod flux peaking.

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MECHANICAL ARRANGEMENT OF COMPONENT PARTS llll2:3 446 W

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GEAP-3851 Part Function Support 1.

To support the weight of all fuel rods and transmit this force to the corner angles.

2.

To prevent individual fuel rods from dropping out of the bundle.

J Weldment 1.

To provide a handle for the fuel bundle without obstructing fuel rod removal.

2.

To transmit the weight of the fuel rods from the corner angles to the lifting device.

3.

To position and capture the retainer.

Retainer 1.

To prevent withdrawal of fuel rods.

2.

To add stiffness to the weldment.

3.3 Fuel Testing The fuel assembly design shown in Figure 3-1 has been subjected to mechanical and flow testing. Specific tests, which include handling, tensile, wear, and drop tests, are discussed in detail in Appendix A.

The results of these tests were satisfactory, and confirmed the final design of the Big Rock developmental fuel.

In-reactor testing of VBWR prototypes of the Big Rock development fuel design is presently in progress. The design of the VBWR prototype l

l 3-5

GEAP-3851 Big Rock development fuel is shown in Figure 3-3.

Three fuel assem-blies are currently in the VBWR, with three more similar assemblies scheduled for insertion by mid-1962. The leading assembly has attained a burnup of about 3000 MWD /T, and operated at a power density of 97 kw/ liter, as of May 1,1962.

1 3.4 Development Fuel Design Data Design data for the Big Rock development fuel are presented in i

Table 3-1. In selecting 0.010-inch thick 304 stainless steel as the fuel element cladding, it was necessary to make a slight departure from the

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" free standing"* criterion normally used in fuel design. The use of thin cladding necessarily dictates that the UO fuel column assists in pre-2 venting the cladding from wrinkling or collapsing due to reactor pres-sure. Attendant with the use of thin cladding is the requirement that the clad have sufficient strength at the end of fuel irradiation life to resist deformation from internal pressure due to fission gas. Table 3-2 pre-hents data that show the stress due to internal pressure in 10-mil clad fuel rois of the development rods to be within acceptable levels after 10,000 MWD /T burnup.

Detailed design calculations supporting Table 3-2 are presented in l

Appendix B, along with clad stress nomographs that show the effect ex-posure, fission gas release, gas collection volume, and heat flux exert on the clad stress at zero reactor pressure. Two types of fuel are pre-sented in Table 3-2, in that the first loading of development fuel in the The " free-standing" criterion is based on the tubular cladding mat-erial not collapsing as a result of 125 per cent or rated reactor pressure when the tube is assrmed to be empty.

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GEAP-3851 TABLE 3-1 DESGN DATA FOR THE' BIG ROCK DEVELOPMENT FUEL ASSEMBLY Water / fuel ratio 2.4

- Cladding material 304 St. Steel Cladding thickness (inches) 0.010 Fuel rods per bundle 121 0.425" OD 109 0.320" OD 12 Fuel.

UO2 Enrichment (% U-235) 2.710.050 2

Heat transfer area per bundle (ft )

76.6 Active fuel length (inches) 70.0 Core volume per bundle (liters) 63 UO weight per bundle (pounds) 383 - 395 2

Kilograms of U per bundle 153 - 158 Kilograms of U-235 per bundle 4.14 - 4.26 2

Average heat flux at 45 kw/l (Blu/hr-ft )

126,000 UO cross-sectional area / bundle (in.2) 34,9 2

Moderator area per bundle (in.2) 35.8 Hydraulle diameter (inches) 0.58 Clad area /UO area 0.105 2

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GEAP-3851 Big Rock reactor will contain both swaged pellet and swaged powder fuel.

Maximum UO temperature has not been used as a primary criter-2 lon in this design. It was included in the design as a secondary factor since the UO temperature does affect the mean fission gas temperature 2

in the powder fuel rods, and the gas temperature, in turn, affects the clad stress criteria stated above.

TABLE 3-2 CLAD STRESS AT THE END OF FUEL LIFE (10,000 MWD /T)

FOR PELLET AND COMPACTED POWDER TYPE FUEL i

Pellets Compacted Type of UO2 Powder Rod diameter (inches) 0.320 0.425 0.320 0.425 Mean fission gas temperature (*R) 700 700 800 800 Internal pressure (psi) 1550 1550 850 865 Pressure stress in clad (psi) 24,000 32,300 13,000 18,000 Minimum Y.S. of clad @ 300 *F 34,400 46,000 18,900 25,800 Y S. -1 43%

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Safety margin 3.6 Advantages of the Design 1.

The design permits the use of fuel rods having variations in

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rod length of approximately 3/4 inch. This feature makes it7 design acceptable for fuel rods fabricated from centerless-ground fuel pellets, swaged-over pellets, vibratory-compacted 3-9

GEAP-3851 UO powder, swaged powder, or Calrod rolled powder. None 2

of these processes is expected to yield fuel rods having more than a 3/4-inch length variation.

2.

The removal and replacement of individual fuel rods is accom-plished with relative ease. This is an important advantage for the Big Rock development fuel where frequent gamma-scans will be perfornwd on individual fuel rods to determine the ac-tual power profile within the fuel bundle. Also, it has the ad-vantage of permitting the reloading of the asnembly structure with fresh fuel, if this is desired. In addition, it would be possible to replace some fuel rods in each bundle with poison to hold down the reactivity in a new core until equilibrium is reached. Having reached the equilibrium condition, the poison rods could be removed and replaced with normal fuel bearing rods.

3.

Any type of end plug may be used on the fuel rod without regard to attachment or location provisions to tie plates as long as the end closure diameter is not larger than the clad OD. For the design presented here, inexpensive cold-drawn hemispherical end plugs are used.

4.

Low fuel assembly structure costs are possible as a result of eliminating castings and the need for the fuel rods to be a structural part of the assembly. Comparison of fabrication i

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GEAP-3851 costs shows that the development fuel assembly design repre-sents a cost saving potential of 25 per cent over the first core fuel design.

l 3-11 l

GEAP-3851 SECTION IV CONCLUSIONS Based on the data presented in Section III, the Big Rock develop-ment fuel design shown in Figure 3-1 fulfills the program objectives for a fuel assembly holding promise of having reduced fuel fabrication costs, long fuellife, and capability of operating at high power density.

4-1

GEAP-3851 SECTION V BIBLIOGRAPHY 1.

Lees, E. A. " Fuel Element Fabrication by Swaging", GEAP-3918 (April 1962) 2.

dyer, C. M. " Powder Fuel Processing by Two-Pass Swaging -

The Effect of Particle Size and Distribution", GEAP-3891 (March 1962) 3.

DeHollander, W. R. " Vibrational Compaction of Uranium Dioxide",

GEAP-4032 (April 1962) 4.

Lingafelter, J. W. " Fabrication of Fuel Rods by Tandem Rolling",

GEAP-3775 (July 1961) 5.

High Power Density Development Project Seventh Quarterly Progress Report, October - December,1961. GEAP-4001 5-1

GEAP-3851 APPENDIX A TESTING Various mechanical tests were performed, both on the fuel bundle components and on a full-size mockup, inorder tosimulate the mechan-ical operations to which the bundle will be subjected in actual service.

The scope of these tests were such that the final design has been con-firmed by way of testing in addition to analysis where it was reasonable, and by irradiation under conditions that simulated the Big Rock design i

when analysis was not feasible or where the cost of a detailed analysis exceeded the experiment cost.

A.1 Fuel Rod Vibration Testing The single-and two-phase flow induced vibration that occurs in the fuel rods has been investigated by E. P. Quinn* atthe Atomic Power Equipment Department, San Jose. Quinn's experiments indicate a mid-span amplitude for either swaged pellet or swaged powder fuel rods that is less than 0.001 inch.

It had been assumed at the outset of the design that the vibration amplitude would be held to less than 0.005 inch by adding more spacers if it was necessary. Since the 0.001 inch is much less than the 0.005 inch expected, the four-spacer design is adequate to hold the vibration amplitude well within reasonable limits.

• Report now in preparation.

A-1

GEAP-3851 APPENDIX A TESTING Various mechanical tests were performed, both on the fuel bundle components and on a full-size mockup, inorder tosimulate the mechan-ical operations to which the bundle will be subjected in actual service.

The scope of these tests were such that the final design has been con-firmed by way of testing in addition to analysis where it was reasonable, and by irradiation under conditions that simulated the Big Rock design when analysis was not feasible or where the cost of a detailed analysis exceeded the experiment cost.

A.1 Fuel Rod Vibration Testing The single-and two-phase flow induced vibration that occurs in the fuel rods has been investigated by E. P. Quinn* atthe AtomicPower Equipment Department, San Jose. Quinn's experiments indicate a mid-span amplitude for either swaged pellet or swaged powder fuel rods that is less than 0.001 inch.

It had been assumed at the outset of the design that the vibration amplitude would be held to less than 0.005 inch by adding more spacers if it was necessary. Since the 0.001 inch is much less than the 0.005 inch expected, the four-spacer design is adequate to hold the vibration amplitude well within reasonable limits.

• Report now in preparation.

A-1

GEAP-3851 A.2 Fuel Assembly Tensile Testing A small prototype bundle was destructively tested to determine the force required to cause permanent deformation of the structure. The structure after the test is shown in Figure A-1.

It was interesting to note that permanent deformation occurred in three different places, at loads from 3000 to 3150 lb. Since the normal tensile load to be transmitted by the structure is 393 lb, the safety fac-tor of 7.5 is considered more than adequate.

It should be noted that the final design does not have holes in the corner angles as shown in Figure A-1. The permanent deformation that occurred at a tensile load of 3000 lb was in the thin section presented between the edge of the angle and the holes in the angle. Since the only purpose of the holes in the anglec was to present an opening where the corner fuel rods could be visually inspected without disassembly, it was decided that the inspection criterion would be omitted and the holes could be eliminated to further strengthen the structure and reduce the cost of the angles.

A.3 Fuel Rod Fr:tting Test There was some concern that the lower end plug would wear or fret through the edge weld that connects the end plug to the cladding, as a result of relative motion between the fuel rod lower end and the support rod. To determine whether or not this was a problem, a test was per-formed in the CL-1 loop to simulate the conditions that might be encoun-tered in pile. A schematic diagram of a specimen and holder is shown in Figure A-2.

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GEAP-3851 The inconel sprink was guided by the clad, and constrained on the ends by the top of the specimen holder and the hemispherical end plug.

Using this scheme, the spring-free length was varied to provide each specimen with a different force (1-7 lb) between the edge weld 'and the X-bars at the lower end of the specimen holder.

Test conditions in the CL-1 loop during Run No. 46 were as follows:

Length of time at test conditions 993 hours0.0115 days <br />0.276 hours <br />0.00164 weeks <br />3.778365e-4 months <br /> Water chemistry Chloride

<D.1 ppm pH 7

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1. 6 x 10 ohm-cm Oxygen (time-weighed average) 0.14 ppm Oxygen Range 0.09 - 0.16 ppm Temperature 546 F Pressure 1000 psig Water velocity in specimen section 5.6 ft/sec Six specimens were tested, and the macroscopic results are shown in Figure A-3, where the specimen numbers are identified by the spring force they experienced.

Metallographic sections were taken through each sample, and the results are shown in Figures A-4a, A-4b, A-4c, A-4d, A-4e, A-4f. It is concluded from these results that no serious problem should exist in regard to fretting or wear of the weld when the fuel rod is supported in this manner.

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p i _k FIGURE A.4F mETALLOGR APMIC SECTION 1-LB FORCE 105 DK.10X MAG P00R ORIGINAL

GEAP-3851 Six High Power Density special fuel bundles will use this support I

method in the VBWR to further evaluate this condition.

A.4 Spacer Tests Two types of spacers are presently being tested in the VBWR, as shown in Figure A-5. Irradiation of fuel with each type of spacer is approximately 5000 MWD /T. Both types are functioning satisfactorily, with neither spacer apparently superior to the other. As a result of this test, the single layer spacer configuration, shown in the lower photo in Figure A-5, has been selected as the one to be used for the Big Rock development fuel because it is the least expensive of the two types.

The spacers tested in the VBWR have from 0- to 0.005-inch clear-ance between the fuel rods and spacer openings, and within this range, no adverse effects have been visually noticeable on the cladding.

A.5 Handling Tests Handling tests were performed on a full-size Big Rock development fuel bundle containing fuel rods loaded with lead shot to simulate the ac-tual fuel bundle weight. The bundle was repeatedly loaded off-center into a regulation size Big Rock fuel channel. The misalignment between the fuel bundle center line and the channel center line was as much as 3 inches. Even in the extreme case, no difficulty was encountered with the fuel bundle entering the channel.

Following these tests, the bundle was dropped in air from a height of 18 inches, to simulate the deformation to be expected from a free fall through water where its terminal velocity would be approximately A-10

a

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sg 446 -2 FIGURE A-5 SPACERS BEING PRESENTLY TESTED IN THE VBWR A -11

=

l l

GEAP-3851 l

10 ft/sec. The bundle was dropped squarely onto a reinforced concrete floor. The extent of damage to the bundle was limited to deformation, but not fracture, of the lower support (see Figure A-6).

As a result of this test, the lower support was strengthened to in-crease the bundle's resistance to impact deformation, particularly in the center of the fuel bundle. The details of this change are shown in Appendix C, Drawing No. 985C212 - Base.

It should also be noted that even though there was permanent defor-mation in the fuel bundle structure as a result of this test, there was no deformation at any place other than the base, and the fuel rods could still be removed.

l A-12

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FIGURE A-6 FUEL BUNDLE DEFORMATION CAUSED BY IMPACT (SIDE VIEWS)

P00R ORIGINAL

.-u

1 GEAP-3851 l

APPENDIX B ANALYSIS B.1 Water / Fuel Ratio ZR-2 CHANNEL d

6.7 4 3"S O.

7.40" t

U 7.40" TYPICAL LATTICE 446-TP 2 Data:

109 fuel rods 0.425 in. OD x 0.010 in, wall 12 fuel reijs 0.320 in. OD x 0.010 in, wall Channel thickness is 0.100 in. thick Water area contained in a typical lattice is = Lattice area-channel area-fuel rod area j

P00R ORIGINAL

GEAP-3851 Water Area = (7.40)2

, 4(6.643) (0.10) - j 109(0.425)2

+ 12(0.320)2 2

= 54.76 - 2.66 - 16.50 = 35.60 in Fuel Area Assuming 100 per cent dense fuel and zero diametral gap between fuel and clad ID, fuel area in a typical lattice is

=j 109(0.405)2 + 12(0.300)2 = 14.90 in.

Water / Fuel Ratio 2

Water area _ 35.60 in

- 2 40 Fuel area

~

14.90 in B.2 Stainless Steel / Fuel Ratio Stainless steel area = cladding area + corner angle area 2

2 2

2 Cladding area = j 109(0.425 0.405 ) + 12(0.320

-0.300)

= 1.56 2

Corner angle area = (4) (2.02 in.) (0.031 in.) = 0.25 in Stainless steel / fuel ratio = gk'

0.121 B'. 3 Fuel Temperature as a Function of Heat Flux The temperatures of the fuel were determined by use of the diffusion equation and difference equation techniques. The diffusion equation is NT

B-2

. ~ _

GEAP-3851 where T

= Temperature within the fuel Q

= Volumetric heating rate, used as constant across fuel sec-tion k(7 ) = Full thermal conductivity, data from J. L. Bates, Nucleonics, Vol.19, No. 6, p.83,1961.

Difference equation techniques were used to determine the temperatures of each of a series of thin cylindrical shells. Twenty shells represented the fuel section. The maximum temperature was directly available, and the volumetric average temperature was obtained by numerical integra-tion of the temperatures of the individual shells. The assumptions used were:

2

1. Boiling heat transfer coefficient is h = 10,000 Btu /hr-ft,

and is constant.

and the cladding

2. Thermal contact resistance between the UO2 2

ID is R = 1000 Blu/hr-ft, and is constant.

The UO temperature as a function of heat flux is plotted in Figure B-1.

2 Having generated these data, it was then necessary to determine the volumetric mean temperature of UO within the fuel rods to know at 2

what average temperature the gases contained within a fuel rod exist.

This information is shown in F_gure B-2. It was obtained by dividing the maximum heat flux by 1.5 (the peak / average axial heat flux) to deter-mine the average heat flux in the rod. At this average heat flux, the B-3

I I

' MAX UO2C NTER TEMP.

4000 O.

(EITHER ROD SIZE)

J 3500

/

1. /

s 4-

/

2500 ts

?

O 2000

/

/

1500

=

O.0. Op 002,,,,,,,. -

/Y p **

iOOO

-= "

/

/

s00 2

MAX HEAT FLUX IN ROD (BTU /HR FT )

0 5

5 S

5 5

5 IX10 2X10 3XIO 4X10 SX 10 6XIO 446-8

\\

FIGURE B-1 UO TEMPERATURE 2

"^

P00R ORIGINAL

\\

W.

2500

/

2000 J r

/

1500

• D 0-E

,4 W

/

1000 ASSUMED: PEAK / AVERAGE HEAT FLUX = 1.5 500 2

MAX HEAT FLUX IN ROD (BTU /HR FT )

I I

I O

5 5

5 5

S 5

IX10 2X10 3Xl0 4x10 SXIO 6X10 446 9 FIGURE 8-2 VOLUMETRIC MEAN TEMP OF UO2 WITHIN A FUEL ROD B-5 P00R ORIGINAL

GEAP-3851 volumetric mean UO temperature was calculated again, using the 2

method aid assumptions stated above.

B.4 Heat Transfer Area Per Bundle Data:

109 rods, 0.425 inch OD 12 rods, 0.320 inch OD 70 In. active fuel length

$425 x 109 +.320 x 15 70 2

w

= 76.6 ft 144 B.5 Average Heat Flux at 45 kw/l Power Density Lattice area x core height x C = liters 0.0164 liters (7.4 in.)2 x 70 in. x

= 63 liters in.

63 liters x 45 kw kw

= 2840 bundle liter bundle 2840 kw 0.948 Btu 3600 Sec bundle l

o 2

(

bundlej sec-kwj(

hr/(76.6ft 2

126,000 Blu/hr-ft average B.6 Hydraulic Diameter Within The Fuel Channel A = flow area DH" P

P = wetted perimeter A = channel inside area - fuel rod area (inches)

= 6.543 109(.425)2 + 12(.320)2 2

2

= 42.9 - 16.4 = 26.5 in

~

B-6

GEAP-3851 P = channel wall perimeter + fuel perimeter

= (6.543) (4) + x 109(.425) + 12(.320)

= 26.1 + 157 = 183.1 in.

H = (4) (26.5) = 0.58 in.

D 183.1 l

B.7 Fission Gas Release, Containment, and Clad Stresa l

In the case of either powder-or pellet-type fuel rods, it will be assumed that 100 per cent of the fission gas generated will be released from the UO, and must be contained within the fuel rod. In the case 2

of powder-filled rods, the fission gas is assumed to exist at the temp-erature shown in Figure B-2. It is further assumed that the gas is con-tained witnin the 8 per cent void volume within the fuel rod (i.e., the powder occupies 92 per cent of the volume within the rod; therefore the remaining 8 per cent is void and can be occupied by gas).

1 For the fuel rods containing UO sintered pellets (at 95 per cent of 2

theoretical density) a plenum space has been provided in the fuel rod to collect the fission gases that are released. Since these fuel rods are non-segmented, it is a simple matter to position this plenum either above orbelowthe active core. In this case, the plenum was placed abovethe core (i.e., the upper end of the fuel rods) because of space limitations within the fuel channel at the lov;er end.

A sample calculation for a 0.425-inch OD powder-filled rod and a 0.425-inch OD pellet-filled rod.is presented in Sections B.7.1 and B.7.2 to show how the data for the following items were generated:

B-7

GEAP-3851 1.

Fission gas collection volume, 2.

Total fission gas generated in each fuel rod, 3.

The fission gas temperature in each fuel rod, 4.

The internal pressure and pressure stress in each rod, 5.

The thermal stress in each rod, and 6.

The combined thermal and pressure stress within each fuel rod.

Based on the detailed calculated data, clad stress nomographs were prepared to illustrate the effect of exposure, fission gas release, gas collection volume, and heat flux exerted on the clad stress at zero reac-tor pressure. These nomographs are shown in Figures B-3 through B-6.

The effect of reactor pressure on the clad stresses of swaged pellet and swaged powder fuel at the end of life is shown in Figures B-7 and B-8.

B.7.1 Powder-Filled Rods (0.425 OD) 1.

Gas Collection Volume (V in.3) 2 V = gl.00 - p) j D H

P = fraction of theoretical density of the UO2 D = fuel OD in inches H = 70 inch active fuel V = (1.00

.92) j (.405)2 70 = 0.720 in.3 B-8

I P(ann)

E00,00 k

pg, 4500-

\\

90.000 Y_,

4000 W.

N 3500

\\

3000-N N

T

~

\\

2500 k

2000 xx xa 1500-x \\x

=

~

1000 NNN - - 500 UO20EN51TY

~

3500 3000 2500 2000 1500 1000 500 1

2 3

4 5

6 PV s/IN 500

1. /l

~

/// / N%

,,00

/// l N

//

/ /

~

~

/ /

/

I N

f /

/

/

/ j' il l

,,00

/

/

"/

/

/

/

/

0.42,5o.o.p=

446-4 FIGURE B-3 CLAD STRESS HOMOGRAPH. O.425 0.D.. POWDER FUEL ROD

~

P00R ORIGINAL

P8/IN2

~"*' Wg

-3500-3 Np A

NN\\

.o.

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~~

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m

.,,00_

^

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• • s as 1500-K 3

-1000-N- -

UO2D E N S'.T Y

%s 1

2 3

4 5

6 70 1000 000 600 400 200 0

MAX Q/ A + 10 5(3T /H R-F T 2 )

P V( t-I N I

200

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/A

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- e00 x

1. 7 / /

/

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l

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~ " " ~

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/

/

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-1600-j

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/

/

s f

'gf f

I sE

-fg

.T 446 5 FIGURE 8.4 CLAD STRES$ NOMOGRAPH 0.320 0.D.. POWDER FUEL RDD B-10 P00R ORIGINAL

%4

..,, 2 4,,+

,#+n"

~

~

i

~

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=

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~

=

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n.,30 1200 1000 800 600 400 200 0

1 2

3 4,

5 g 6 l

M AX Q/ A + 105 3TU/HR-FT2 PVie-IN) 1

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-- 200

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. 00 7

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,,.000 _T y'

1600 1800-

/

[

f f

ASSUMPTIONS: GA5 TEMPERATURE j,$

CONST ANT AT 1260'R s8 AEh 446-6 FIGURE B-5 CLAD STRES$ NOMOGRAPH 0.475 0.D. PELLET FUEL ROD 5 P00R ORIGINAL

P 8/lH2

[

N w4 k

4000-Q

%3 000 h

(

3500' 5

s

=

~=

q h

2500- A t

g l

k

-2000-'

\\\\\\

'N k

%eo 1500 n

500 3

i 2'

3 d

140o 1200 1000 000 eco' 400 200 o

pvt /riN) l l

I 200

1. /l pgj 1-

=

15 000 uwo/T 1000

/ / /

I F

/

I

/

P

</

I

/

/

_"/

s

/

/

/

ij I

446 7 FIGURE B 6 CLAD STRESS HOMOGRAPH O L100 PELLET FUEL ROD i

~

P00R ORIGINAL

70,000 60,000

- 0.425 O.D.

. 0.320 0.D.

\\

6 0,0 0 0 x

\\ 04,

1. o 40,000 of

%o

9.,'?

f

% 9/

5 0,0 0 0 l

p K

g N

~ a,Y.

s

1. RO n

24000 -g

\\ \\

i N

\\

u O

I O,0 0 0 A

N N

5 8

l 0 : ;;I O

=

lO Of 20 00 REACTOR PRESSURE s/IN2

-l0,000 446-3 FIGUREB 7 END OF LIFE CLAD STRESS FOR PELLET RODS P00R OR!GINAL B-13

~

s 70,000 60,000

- 0.425 0.D.

- 0.320 0.D.

60,000 s\\

s%#

40,000

-n W

h

'4.-

\\

=

30,000 a

O R+ \\

20,000 -3 b

h 10,000

\\

10M 2

N$5s'uE o-T+

s/IN2

+o00

-10,000 7++

go

-20,000 446 2 FIGUREB 8 END OF LIFE CLAD STRESS FOR POWDER RODS P00R ORIGINAL B-14

GEAP-3851 2.

Total Fission Gas Generated in Each Rod Personal communications with C. N. Spalaris have indicated that 1.31 x 10-3 gm-mols of Kr + Xe are formed per megawatt day (MWD) of exposure. Using this data and assuming all of the gas generated is released from the UO, the total mols of 2

Kr + Xe can ta calculated as shown below.

Tons of U per rod:

tr 0.396 lb.

Ton g (0.405)2 70 in.3 (0.92) 238 lb.U 2000 lb.

- in

• 3 270 lb.UO2

= 1. 45 x 10-3 tons of U/ rod Since the design exposure is 10,000 MWD /T, the total Kr + Xe is 4

1. 31 x 10-3 gm-mols/ MWD x 1.45 x 10-3 T x 10 MWD /T

= 1.90 x 10-2 gm-mols of Kr + Xe/ rod 3.

Fission Gas Temperature in Each Rod For the powder-filled fuel rods, the mean fission gas temper-ature may be obtained directly from Figure B-2, knowing the maximum heat flux on the rod surface.

4.

The Internal Pressure in Each Fuel Rod The internal pressure in the fuel rod is based on a maximum 2

heat fic of 550,000 Blu/hr-ft. Entering (as shown in Figure B-2) at 550,000 Btu, the mean temperature of the UO2 (fission gas also) is 2050 F (2510 R). Then, from B-15

GEAP-3851 PV = n RT V = 0.720 in.3 n = 1.90 x 10-2 gm-mols I"y, mol' R = 40.6 R

T = 2510 R the internal pressure due to Kr + Xe is calculated to be 2690 lb/in.2 Assuming zero reactor pressure at this condition, the maximum tensile stress in the clad would be defined by a = E' t

a = tensile stress in the clad (lb/in.2)

P = internal pressure (Ib/in.2) r = mean clad radius (in.)

t = clad thickness (in.)

or

= (2690) (0.208) = 56,000 lb/in.2 0.010 5.

Thermal stress in the cladding.

l The thermal stress is calculated using the Timoshe ' o* equa-tion for thin-walled cylinders.

Enot "t

= 2(1- )

6 lb/in.2) where E = Young's Modulus (30 x 10 l

l l

• Timoshenko, S., Strength of Materials, Part II, 3rd Edition, D. Van Nostrand Co.,1956.

B-16

GEAP-3851

)

a = coefficient of expansion (10-5 in. /in. )

I At = temperature drop across the clad wall ( F)

= Poisson's Ratio (0.3) and since

= (Q/A)(X)

At

-k 2

where Q/A = Blu/hr-ft 550,000 0.

O X = wall thickness ft 2

k = thermal conductivity 10 Blu/hr-ft we can substitute for at at = E a(Q/A)(X) 2k (1- )

= 30 x 10 x 10-5 x 5.5 x 105 x o,010 6

2 x 10 x (1 - 0.3)(12)

= 9,820 lb/in.2 This will be tension on the clad OD and compression on the ID.

6.

The combined thermal and pressure stress in the clad at an exposure of 10,000 MWD /T, zero reactor pressure, and a 2

heat flux of 550,000 Blu/hr-ft gg a total = a thermal + o pressure

= 9,820 4 56,000 65,820 lblin.2

=

i B-17 l

l l

GEAP-3851 B. 7.2 - Pellet-Filled Rods (0.425 OD) 1.

Gas Collection Volume The gas collection volume (plenum) for these rods was derived from matching, as closely as possible, the clad stress (at 2

10,000 MWD /T, a heat flux of 550,000 Blu/hr-ft, and zero reactor pressure) of the powder-filled fuel rods. This trial-and error matching process yielded a plenum volume for the 0.425 OD fuel rods of 0.362 in.3,

2.

The total fission gas generated is calculated in the same manner density is 95 per cent.

used for the powder rods, except the UO2 On this basis, each rod contains 1.50 x 10-3 tons of U, and 1.96 x 10-2 gm-mols of Xe + Kr are generated and released.

3.

The fission gas is assumed to be contained within the plenum, and is assumed to exist at a temperature of 800 F (a conserva-tive estimate that is greater than the mean surface temperature of the plenum walls).

4.

The Internal Pressure in Each Fuel Rod The pressure is calculated using the same method presented for the powder rods, namely:

p, n RT V

where n = 1.96 x 10-2 Em-mols R = 40.6 in.-lb/gm-mol *R T = 1260 *R V = 0.362 in.3 B-18

GEAP-3851

then, p = 1.96 x 10-2 x 40.6 x 1260 0.362

= 2850 lb/in.2 Assuming zero reactor pressure, the tensile stress in the clad due to Kr + Xe pressure at the end of life is Pr (2850)(0.208) y =T 0.010

= 59,400 lb/in.2 5.

The thermal stress in the cladding will be the same for either powder-or pellet-type fuel rods having the same clad thick-ness and operating at the same heat flux. Therefore, the ther-mal stress will be 9,820 lb/in.2 as calculated for the powder rods.

6.

The combined thermal and pressure stress in the clad at an exposure of 10,000 MWD /T, zero reactor pressure, and a heat flux of 550,000 Blu/hr-ft2, g

" total " " thermal + " pressure

= 9,820 + 59,400

= 69,220 lb/in.2 The analysis method presented above was also used to calculate the conditions in the 0.320-diameter fuel rods.

Table B-1 tabulates data relating to fission gas for the two sizes of pellet of power fuel rods calculated as discussed above. Table B-2 B-19

GEAP-3851 TABLE B-1 DESIGN DATA RELATING TO FISSION GAS FOR PELLET AND COMPACTED POWDER TYPE FUEL ROD C

Pellets Type of UO2 r

Rod diameter (inches) 0.320 0.425 0.320 0.425 UO density (h of T.D.)

95 95 92 92 2

Tons of U per rod ( x 10-3) 0.82 1.50

.79 1.45 Pounds (or kg) of UO 1.86 3.40 1.80 3.30 2

(O.845)

(1.55)

(O.82)

(1. 50)

Location of fission gas collection Plenum Within UO space (out of core) 2 Fission gas volume (in.3) 0.198 0.362 0.396 0.720 Total fission gas generated 0.0108 0.0196 0.0104 0.0190 (moles of Kr i Xe at 10,000 MWD /T)

Assumed % release of fission gas 100 100 100 100 B-20

GEAP-3851 TABLE B-2 FUEL ROD CLAD STRESS AT THE END OF LIFE, ASSUMING A SUDDEN LOSS OF REACTOR PRESSURE C

cted Pellets Type of UO2 Rod diameter (inches) 0.320 0.425 0.320 0.425 Maximum UO temperature (*F) 3375 3950 3375 3950 2

Mean UO temperature ( F) 1800 2050 1800 2050 2

Mean fission gas temperature ( R) 1260 1260 2260 2510 Internal pressure (psi) 2800 2770 2410 2690 Pressure stress in clad (psi) 42,500 56,900 36,700 56,000 Thermal stress in clad (psi) 9820 9820 9820 9820 Tensile stress in clad OD (psi) 52,320 66,720 46,520 65,820 Minimum U.T.S. of clad @ 650 F 65,500 83,500 58,300 82,000 Safety raargin U T.S

-1 25%

25%

25%

25%

1.

G NOTES:

2 (1) Q/A Max. = 550,000 Blu/hr-ft (2) Peak / axial heat flux = 1.5 l

I B-21

GEAP-3851 tabulates individual fuel rod clad stress for the condition of sudden loss of reactor pressure at the end of fuel life. The significant fact from Table B-2 is that, with a sudden loss in reactor pressure, the clad has sufficient strength to prevent rupture of the fuel rod from internal pres-sure due to fission gas.

B-22

. - = -.

~.. _

GEAP-3851 APPENDIX C DRAWINGS The following is a list of the drawings that comprise the Big Rock Development Fuel.

124F725 Fuel Bundle 141F898 Spacer 985C211 Handle 985C212 Base 985C213 Weldment 985C214 Bracket 985C215 Angle 114B5456 Retainer Bolt 114B5457 Hairpin 114B5833 Rod 114B5834 Foot 114B5835 Plate 114B5850 Fuel Rod 117B1233 Plate 114B1450 Fuel Rod 124F725 Assembly Fuel Bundle 149A4071 End Plug 149A4958 Pellet C-1

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~~ GiKi %. NAME DESCRIPTION N A8 49 i POWCER SEE NO TE I. 2 2 ENP FLU 6-14 0A 407/. / 3 TUDE ASTH A$G9 THE 304 STN STL', t1 ClL., i A '"# PLuC'

1. 4 #A # /*

HELIUM = 70 1 5 TUBE .3750.P X.010 WALL 4TOCK &l E o JHIELDED ARC STN srL ASTM A 2(o9 TYPE 304 WELP TYf/Chl. BOTH ENPS 24 1 35 FCf" 0N 124/~725 ~ ~ n v $';_ A.' 5,.Y lO.*f NOTE I. POWDER-ARC rusED V0a. OOI b O

2. i *.a.7

_) \\) ENRICHMENT LATER ~ J .Q. , i; PER PROCES$ 3PEC. L ATER --,.,/r* k *.'. 4 } } DIAMETER ~ ATTER SWAGING. G NO. A. ( .425 2. .320 i s y:w- $qy....mo... .. - -Lga5850 SAN JOSE ~ ~ ~= g 3 l l ~

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I a:Ell JE C.5 Ca= Cll3 " '" "' 8 """i c 1l~7 B 1450 OG1GTjl K Cac ,.<..o uo e,- C".* '. l 1 B 1 4 5 0 FUEL ROD g ,.o-., . rapgs4pi_ j ,,,,,,_,, con s o ME R3 Powea cov. cll3 gg = ._.o.-.- .406 ,0.D A2 i FUEL PELLET 149A4958 PI g 4'. OI5 ** ' WALL 3 k O.D X.OIO NALL i 7 '2 EMD PLUG (49A40Fl CUT LG. TO Suff I 3 TUBE ASTM A769 TP ~504 1 4 PLENUM ASTM A 269 TP 304-AR 5 FULL PELLET [49A4258 P2 + .305~004 2 6 END PLUG I40 A 40ll O D-1 003 g 7 TUBE ASTM A 269 TP 304 6 015 WALL ] 2_I ~ O.D X.010 WALL I O PLENUM ASTM A 169 TP 3O+ CUT L6 To SutT 64 FCF 114 F715 or 1 NN 5 Is } 43 Bo ~ __3 _ 0K .. _ _ _ _/\\ __ A - 4 (HEllUM SHIELDED ARC WELD v

g. d a 3 L

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GENERAL @ ELECTRIC / {pg " "o-W TITLE ASSEMBL Y 722F725 FUEL BUNDLE ~ coat on swert - - en no. FIRST MADE FOR CONSUMERS POWER DEM " " '" "[h / P hAME DRAWING NO., DESCRIPTION, MATERIAL, WEIGHT I I HAbfDLE 985C2IIGI l 2. WELDMEN T 885C 2I3 &I 4 3 ANGLE 985C2ISP) IO 4 FUEL R00 II4 85850Gl IE 5 ruEL R0p //4 85850G2 l l G B A.S E 385C2 /2 GI 4 A 7 SPACER I4/F888&l l I 8 RE7GMEPBou //485/SYo P/ l I 3 HR/ARN /MB5457P/ 4 10 AM61.E 9 85C7.!5 PI 109 11 FUEL ROD ) l l B 14 50Gl 11 l'1 FUEL ROP l l 1 B 1450G2. l IS HRNDLE 785'c 2 // 6 2 l 14 WBDMENT 985C 2/3 G2 4 DESCRIPTION OF GROUPS REVISIONS PRINTS TO G! - A9WDER G2-Pf4LfrJ P00R ORIGINAL I I I I gf p da w u-R$$ _AEER_ _M ]}}f}Qf_ ew saysoss s FFJt01-P 2%f U-37) . ~..

GENERAL @ ELECTRIC /49A 4 0 7 / 8" "0 I "'d 4 M/> 4 TITLE lC0" '" 8"'E7 14 9 A 4 0 7 / ~ END PLUG \\

o. /

FIRST MADE FOR HPD TA $ K I 5 w., .. m -

• 045 '

FCr ON 385CGG/ JAER/, A. __ 4,5 - [ PROFILE To BE NE8\\s PRERIO At paa WIT 4\\M ' c 2. 0 '.' QJA =a' v H

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