ML20010G709

Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in BWR Storage Pool

ML20010G709
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Site:	Clinton
Issue date:	07/21/1978
From:	BROOKS & PERKINS, INC.
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References
577, NUDOCS 8109220437
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Report No. 577 -

m+f Brooks & Perkins. Incorporated

\\U Prepared by:

Brooks & Perkins, Inc.

' Advanced Structure s Div.

12633 Inkster Road l

Livonia, Michigan 48150 I

e Date:

July 21,1978 i

THE SUITABILITY OF BROOKS & PERKINS' SPENT FUEL STORAGE MODULE FOR USE IN BWR STORAGE POOL I

4 8

4 8109220437 810914 PDR ADOCK 05000461 A

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Repori 577 Brooks & Perkins. Incorporated THE SUITABILITY OF 3 ROOKS & PERKINS SPENT FUEL STORAGE MODUL C FOR USE IN BWR STORAGE POOL PURPOSE The purpose of this report ic to exhibit test results and literature research that illustrates the suitability of the Brooks & Perkins Spent Fuel Storage Module (SFSM) for use in a boiling water reactor (BWR) storage pool.

BACKGhCUND Soent Fuel Storace Module:

The SFSM is a slender square-shaped tube with open ends that is used for the storing and the shielding of one spent fuel assembly in a light water nuclear reactor storage pcal. The tube is constructed with the inside and outside coverings being made of type 304 stainless t : eel. These two stainless steel surfaces are welded together at the top and bottom of the tube over an inner t

layer of a thermal neutron shielding material called BORAL m.

Boral is a sand-which type panel that has outer surfaces of type 1100 aluminum and a core of boron carbide uniformly dispersed in a matrix of type 1100 aluminum.

A group of SFSM's are assembled into a tightly packed array called a high-density storage rack. A network of horizontal and diagonal members separate the modules within the rack and provide the necessary lateral support. The racks stand in a vertical position on the bottom of a 40-foot deep storage pool.

The water in the storage pools is constantly circulated through a ' eries of filters which causes a constant water flow within the pool. The water is monitored and controlled for pH and temperature within specific limits depending on the type of nuclear reactor.

Environment of SFSM: In a BWR, the high density ste rage rack is exposed to the following conditions.

Radiation Exposure 10" rads gamma total.

2 104 neutrons /cm /sec average flux.

Water Type demineralized.

Water Temperature 700 to 1500F (21 to 660C).

pH at 77 F (25 C)

5. 8 to 7. 5 Chloride ion, ppm, max.

O. 5 m

Report 577 e

-\\Uf Brooks & Perkins, Incorporated Tot'al Heavy Element, ppm, max.

O.1 Total Suspended Solids, ppm, max.

1. O Solids Filtration, Microns, max.

25.0 The storage racks are expectedto withstand these conditions over a 40-year

. p e rio d.

Shielding Cacability of Boral:

The shielding capability of a Boral panel is due to its ability to capture thermal neutrons. The capture of thermal neutrons is accomplished by the B IO (boron-ten) isotopes that are contained within the boron carbide particles. These boron carbide particles are chemically inert (unreac-tive), heat-resistant, highly crystalline and nearly equivalent to diamond in hardnecs.

In order for corrosion tc cause a reduction in the shielding capability of a Beral panel, the boron carbide particles have to be physically displaced from the panel. A displacement of the boron carbide particle: to occur would require the following sequence of events.

( ) ',

The coinplete removal of the outer protective aluminum aurfaces on the Boral panel.

(2)

The complete removal of the aluminum matrix surrounding each boron carbide particle.

I(3)

The physical displacement of the boron carbide particles.

TESTING AND RESEARCH Testing and research were cond"cted to substantiate the ability of the Brooks &

Perkins SFSM to satisfactorily resist the environn2ent of a boiling water reactor spnt fuel storage pool. The following is an outline of the investigation.

1.

Corrosion Resistance Testing and Research 1.1 SFSM Without Leak in Stainless Steel Covering: The corrosion resist-ance of the stainless steel covering of the SFSM has been investigated through research of published data. The following information indicates that the Brooks & Perkins SFSM (namely 304 stainless steel) provides adequate corrosion resistance to achieve a life expictancy of 40 years when used in a BWR storage pool.

.;2

Report 577 hh Brooks &Perkins. Incorporated 1.1.1 Stainless Steel - Type 304:

A.

General CorrosionI Water Type BWR pH

7. O to 11 0

Temperature 572 F (300 C)

Oxygen, ppm

<. 01 to 2 Chlorides, ppm

<.I Corrosion Ratc, mpy

<2 Er.timated Corrosion Rate 0

@ 150 F, mpy

<.6 Expected Life (at 36 mils thicknes s)

> 60 years 2

B.

General Corrosion After 3000 Hours Water Type high purity, demineralized Hydrazine, ppm

. 01 to. 0 7 Oxygen, ppm

<. 005 Chlorine, ppm

<.05 pH

6. 95 to 9. 58 Flow R :.te, gal /hr 3.5 0

Temperature 320 F (160 C)

Corrosion Rate, mpy

.01 Expected Life (at 36 mils thicknes s)

> 60 years I National Assoc. of Corrosion Engineers, Corrosion Data Survey,1974, pp. 34 and 252, 2 A.P. Larrick, Corrosion Studies in Simulated N-Reactor Secondary System Water Environment. Atomic Energy Commission Research and Development, Report HW-76358, Hanford Atomic Products Operation,..;ay 1963, pp. 7,10 J

and 22. ~

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Report 577 I

hk Brooks &Perkins,lacorporated C.

Stress-Corrosion-Cracking after 3000 houg water quality same as "B" above Stress % of.2% yield 120 re sults "Metallographic examination of selected samples also failed to 1

reveal any cracking. "

expected life (at 36 mils thickness)

> 60 years

1. 2 SFSM With A Leak In The Stainless Steel Covering 1.2.1 Resistance to Corrosion From Published Data f

The corrosion resistance of BORAL and of Stainless Steel (Type 304) coupled with Aluminum (Type 1100F) has been investigated to determine the corrosion resistance of the SFSM under conditions of the stainless steel covering con-taining a leak during use in a BWR Storage Pool. The published data reviewed indicates the materials used in the Brooks &

Perkins' SFSM (namely BORAL, 304 Stainless Steel and 1100F Aluminum) previde z.dequate corrosion resistance to achieve a life expectancy of forty years without a reduction.of neutron absorbing capability when used in a BWR Storage Pool with a leak in the stainless steel covering.

CORROSION DATA BCRA L i

A.

2000 Hour Test Results water type BWR pH

7. 0 temp.

190 F (88 C) 3L. Marti-Balaguer and W. R. Smalley, " Evaluation of Control Rod Materials: CVTR Project", CVNA-86. Carolinas-Virginia Nuclear Power Associates, Inc. (1960) 1 Report 577 Wh Brooks &Perkins. Incorporated

\\

l corrosion rate, mpy

1. 2 ta 2. I estimated corrosion rate

@ 70 to 15. 'F., mpy

.18 to. 32 r

expected life (at 15 mils thickne s s)*

45 years STAINLESS STELL (tvoe 304) coupled with ALUMINUM (type 1100F)

A.

Crevice and Galvanic Corrosion water type high purity, demineralized oxygen, ppm 4 to 3 pH

5. O to 6. O flow rate, fpm O. 5 te mp.

194 to 356 F (90 to 180 C) time, hrs.

1100 1775 2000 Al max. pit depth, mils 2

<3

<5 Al corrosion rate, mpy O.1 0.1 0'. 1 S. S. Corrosion rate,...py 0

0 0

expected life (at 15 mils

> 60 yrs 760 yrs 760 yrs thickness of A1)

CONCLUSION: A thorough review of the published te ot data indicates the materials used in the Brooks and Perkins, Inc. spent fuel storage module (namely 304 Stainless Steel and 1100F Aluminum) provide adequate corrosion resistance to achieve a life expectancy of forty years without a reduction of

~

neutron absorbing capability when used in a BWR storage pool with a rupture in the stainless stee! covering.

• 10 mils of Clad plus 5 mils of Matrix Holding Boundary Layers of B C.

4 4.L. English and J.C. Griess, Dynamic Corrosion for the High - Flux J

- - - Isotope Re. actor, ORNL -TM - 1030, September, 19 66, pg. 1,2, 3, 3,2 3, 26,27,31.

o -

Report 577 Wh Brooks &Perkins. incorporated I

1. 2. 2 Corrosion Resistance of BORAL A test was conducted to determine the physical changes to unprotected samples of BORAL after one year of exposure to an aqueous solution representing a BWR Storage Pool water.

Test Method: Three bare and unclad samples of BORAL were placed in a covered beaker containing demineralized water.

The samples were the standard 35% B C type BORAL panc1::

4 that measured.177 x 2 x 2 inches. The samples and the solution were periodically examined and the progression of the change s were recorded. The samples were carefully left undisturbed during the test period without any change or alteration being made to the water solution. -

Re sults: The BORAL samples experienced weight losses through general corrosion with no evidence of pitting, galvanic or intergranular type s of corrosion. The average rate of weigh:

loss after one year of exposux e was 1.91 milligrams per square centimeter per year or. 28 mils per year.

The pH reaction and other test data are included in Appendix 1.

The test concluded that the period of time necessary for tue total loss of the oute r cladding of the BORAL by general corrosion at the corrosion rate after one year wculd be considerably more than 40 years.

2.

Irradiation Resistance Testing and Research

2. I Gas Generation Tests of BORON CARBIDE / ALUMINUM MATRIX BLEND An experiment was conducted to determine if gas would be evolved by the matrix material utilized in BORAL with hygroscopic moisture during neutron and gamma irradiation.

The experiment consisted of irradiating a 26 gram granular boron; carbide / aluminum sample containing 2,673 ppm moisture.

The sr.mples were exposed to radiation over 5 minute and 75 hour8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> periods. The radiation exposure of these tests is listed-in Table 1. '

Report 577 hh ' Brooks & Perkins. Incorporated TABLE 1 40 YEAR AND TEST CUMULATIVE RADIATION EXPOSURE 40 Year 5 Min. Te st 75 Hr. Test Radiation Tvue Exposure Exposure Exposure Neutron (n/cm )

j 13 13 16 Thermal (

.1 Mev)

4. 29 x 10

4. 50 x 10

4. 05 x 10 i

1. 38 x 10

1. 24 x 10 Epithermal 12 1$

6. 00 x 10

5. 40 x 10 Fast ( 1 Mev)

Gamma (rad)

8. 00 x 10

1. 67 x 10

1. 50 x 10 l

Test resulted in no gas evolution detected during or following the 5 minute and 75 he,ut irradiation.

f A complete description of the experiment is included in Appendix 2.

1

2. 2 Irradiation of SFSM with and without leak in Stainless Steel Covering Experimental observations were made of BORAL plates encased in stainless steel jackets. Sample s were tested dry and with 25 ml distilled water injected within the stainless steel jacket.

Under irradiation fluxes and water conditions expected in a power reactor spent fuel pool, the BORAL samples exhibited no detectable gas evolution, pressure buildup or damage due to temperature or other effects.

A listing of the experimental results is given in Table 2.

A description of the experiment is included in Appendix 3.

' ?: '~

hh Brooks &Perkins, incorporated SAMPLE 1 SAMPLE 2 9" x 9" Boral Plate 9" x 9" Boral Plate Stainless Steel Jacket Stainless Steel Jacket Dry

. 25 ml Distilled Water I

( CONDITION 1 f 42 Hours 25 Hours Spent Fuel No Detectable Effect No Detectable Effect 7 - 2 x 105 Rad / ri h

N - Negligible

CONDITION 2 24 Hours 6 Hours s

Reactor at 2 MW No Detectable E!!ect. No Detectable Effect i

I - 4 x 107 Rad /hr l

Rad /h; l 7

t N - 1 x 10 i

l CONDITION 3 4 Hours 4 Hours Reactor Shutdown No Detectable Effect No Detectable Effer' i

6 T - 1. 2 x 10 Rad /hr' i N - Negligible TABLE 2 Observed Effects of Irradiation Conditions on BORAL Samows --

Report 577

$h Brooks &Perkins.Inco.porated 4

2. 3 Helium Generation The previously discussed irradiation tests have shown no detectable gas generation by BORAL, but concern has been expressed due to the well known rcaction I

4 N

B 4

7 He o

+S 3Li

+2 and the possibility of pressure build-up in an enclosed environment. This concern may be satisfied by considering that all neutrons which strike the BORAL are thermal neutrons and are absorbed by boron-10 and considering the following calculations:

4 2

f = 10 n/cm /sec Average Flux 4

2 BORAL Area / tube = 3.4 x 10 cm Void between BORAL and tube = 130cc Void in BORAL are/ tube - 300cc 9

Seconde in 40 years = 1. 26 x 10 4

9 13 2

{ = 10 x 1. 26 x 10 = 1. 26 x 10 molecule s /cm over 40 years i

2

~

1. 26 x l'1

6. 023 x 10

= 2.1 x 10 mole s/ cm of He over 40 yrs 2.1 x 10-11 x 3. 4 x 104 = 7 x 10-7 moles / tube of He over 40 yrs 7 x 10-7 x 22,4 x 103 = 1. 6 x 10-2 cc/ tube @ STP of He in 40 yrs Pressure @ 150 F = 1 atm x 1. 6 x 10-2 x (273 + 66) * (273 x 430)

= 4. 6 x 10-3 atm The pressure rise lor the 40 year period of 4.6 x 10-5 atmosphere s or.0007 pounds per square inch is insignificant when considering the internal gauge pressure to cause a buckling of the tube walls is in excess of 5 pounds per square inch and that the pool water exerts an external pressure of 17 psi under 40 feet of water..

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i Report 577

-(hh Brooks & Perkins,lacorporated A PPENDIX 1 Test Data from Brooks & Perkins Report 511 " Corrosion Resistance Tm of BORAL to One Year of Exposure to BWR Storage Pool Water."

1 Total Unit TVt.

Surface Area Iritial Final Weight Loss lhtration !

Size Two Sides

\\%igt Weight Loss Per Yeag Ihr Year 2

Samole (cm)

(cm )

(gms)

( g me,)

(gms)

(mmsicm /vr)

(mis /vr)

A

.45x4.92x4.92 48.4 29.8564 29.7612

.0952 1.97

.29 B

.45x5.08x5.08 51.6 31.2759 31.2042

.0717 1.39

.20 C

.45x5.08x5.24 53.2 32.3222 32.1962

.1261 2.37

.34 Ave rage 1.91

.276 Observation of e[ of Test Solution i

The pH of the solution increased in number from the original S. 6 to 7. 7 in the first two and one-half months and remained at that approximate level for the remainder of the year. A graph of this observation follows.

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Report 577 hh Brooks &Perkins.Incomorated APPENDIX 2 Gas Generation Tests of Boron Carbide / Aluminum Matrix Blend Test Descriotion Sample material was tightly packed, but not compressed, into a cylindrical-aluminum container as illustrated in the followitig fig tre of the BORAL ir-adiation container. Minimum lengths of tubing and connectors were attached to a threaded fitting at the top of the cylinder to enable: 1) Attachment of a pressure relief valve set at 30 psig and 2) Attachment of a valved gage tapoff for pressure measurement. The total volume of tubing and connectors was approximately O. Sec. The in line relief valve relieved to a long aluminum tube that extended from the irradiation position to the pool surface. A gage was attached to the tube above the pool surface.

If gas pressure built up during irradiation above the point of lifting the relief valve, the pressurc would have been detecte? on the surface gage. Following irradiation, a gage was attached to the container te measure pressure built up by gas evolution during irradiation.

Samples were subjected to gamma and neutron fluxes in the Ford Nuclear Reactor at the University of Michigan, Ann 4 tbor, Michigan.

4. --

/'

3 Surface Gage i

j x%,

1 I

.... l ;-.

Pool Surface- - -.. - - -

x N

l 30 psig.

Relief '

l

- ~

f x

i Post

~'

l V s,-

1 Irradiation ss

/ Gage s

ii o

_.1 It Irradiation 'l i.

i Container '

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BORAL IRRA DIATION CONTAINER -

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Report 577 hk Brooks &Perkins. Incorporated APPENDIX 3 Irradiation of Boral

• te s encased in St.'inle ss Steel Test

Description:

Each BORAL sample was a 9 inch x 9 inch plate of 0.36 inch thickne s s.

Each plate was encased in a thin, watertight jacket of stainles i steel welded around the 3dge s.

A threaded connect'.on was welded in the upper right corner of the face on one side of the stainless steel jacket. Irradiations were conducted in the Ford Nuclear Reactor pool at depths of 12 and 20 feet.

An alurninum tube was run from the connection to the surface of the reacter pool for pressure measurements and gas collection.

Prior to testing, each sample plate was baked at 200 C for seven hours in a vacuum oven to remove meisture.

Each sample was tested to 10 P SIG internal pressure. Experimenti prc ssures were limited to 5 P SIG as a reactor safety precaution.

Experimental measurements were made of pressure within each sample. Gas evolved during the tests was collected and analyzed. It was decided that temperature would not be measured. Each sample was observed after irradiation for damage due to pressure, temperature, or other effects.

Each sample was pressurized momentarily to 10 P SIG a3 it was inserted into the reactor pool to verify watertightne ss. Once each sample was placed in its experimental position, a 30 inch Hg vacuum was drawn to evacuate as much air as possible. The starting pressure for each test was the 30 inch Hg vacuum.

Exuerimental Conditioris: The two samples were subjected to two different irradiation conditions. Sample 1 was a sealed, dry sample vented only through the gas collection line to the surface of the reactor pool. Sample ?. was identical to Sample 1 except that 25 ml of distilled water was injected within the stainless steel jacket.

Initially, in Condition 1, each sample was irradiated adjacent to spent reactor 5

fuel in a gemma flux of 2 x 10 rad /hr. In conditica ?., each sample was placed The in a holder adjacent to the reactor pperating at a power level of 2 MW.

as 4 x 10 rad /hr and thermal neutron flux was Condition 2 gamma g w/cm /sec, or 1 x 10 2

7 approxf mately 1 x 10 N

rad /hr. Finz',1y in Condition 3, each sample was left adjacent to the reactor core immediately after shutdown.

Neutron flux was quite low, approximately five orders of magrgitude below operating levels, while gamma flux was measured as 1. 2 x 10 rad /hr.

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