ML19263C651
| ML19263C651 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 02/08/1979 |
| From: | Goddard R COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 7902280433 | |
| Download: ML19263C651 (15) | |
Text
_.
<i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
- n.,
BEFORE THE ATOMIC SAFETY AND LICENSIflG BOARD
[
p-a
~ D In the Matter of
($
COMMONUEALTH EDIS0!! COMPANY
)
Docket flos. 50-295 7 y
)
50-304
-i 'j (Zion Station, Units 1 and 2)
)
CERTIFICATE OF SERVICE AND AMENDMEflT TO CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S MOTION FOR
SUMMARY
DISPOSITI0fi," in the above captioned proceeding, were personally served upon John W. Rowe, Esq. for Licensee and on Susan fl. Sekuler, Esq.
for Intervenor, State of Illinois on February 1, 1979.
I further certify that (1) signed copies of the affidavits of Frank M.
Almeter, and (2).Round 2 & 3 questions and responses, referenced in the above Motion, have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 7th day of February, 1979:
Ed'..'ard Luton, Chai rman Atomic Safety and Licensing Board Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Comnission Washington, D. C.
20555 Washington, D. C.
20555 Dr. Linda W. Little Atomic Safety and Licensing Appeal Research Triangle Institute Board Panel P.O. Box 12194 U.S. Nuclear Regulatory Commission Research Triangle Park, N. Carolina 277n9 Washington, D. C.
20555 Dr.' Forrest J.1Mmick Docketing and Service Section 305 E. Hamilton Avenue U.S. Nuclear Regulatory Commission State College, Pennsylvania 16801 Washington, D. C.
20555 John W. Rowe, Esq.
Isham, Lincoln and Beale One First National Plaza Chicago, Illinois 60690 Susan N. Sekuler, Esq.
Russell R. Eggert, Esq.
Assistant Attorney General Environmental Control Division
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188 West Randolph Street, Suite 2315 Chicago, Illinois 60601
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Mr. Rick Konter
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V 617 Piper Lane Lake Villa, Illinois 60046 7902280 9'3.3
s, UrlITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSI0fl BEFORE THE ATOMIC SAFETY AND LICEftSING BOARD In the Matter of
)
Docket Nos. 50-295 COMMONWEALTH EDISOf1 COMPAtlY 50-304 (Zion Station, Units 1 and 2)
AFFIDAVIT OF FRANK M. ALMETER I, Frank M. Almeter, being duly sworn, do state as follows:
I am employed by the United States Nuclear Regulatory Commission as an Applied Mechanics / Material Engineer in the Engineering Branch, Engineering and Projects, Division of Operating Reactors, Office of fluclear Reactor Regulation. A statement of my professional qualifications is attached to this affidavit.
This affidavit addresses the following contentions:
8.
The amendment request and supporting documentation have not analyzed the long term
- electrolytic corrosion effects of using dissimilar alloys for the pool liners, pipes, storage racks and storage rack bases, such as the galvanic corrosion between unadnodized aluminium, as is used in Brooks and Perkins storage racks, and the stainless steel pool liner.
- "long term" storage would include storage during the life of the reactor.
Contention 8 assumes that electrolytic corrosion will affect the long term integrity of dissimilar alloys, such as the galvanic corrosion
, between unadnodized aluminum and stainless steel in the pool water environment.
The spent fuel pool components of the Zion facility that are exposed to the pool water are:
1.
Pool liner - stainless steel, 2.
Spent fuel assemblies - Zircaloy clad fuel rods, stainless steel tie plates, and Inconel spacers, 3.
Storage Rack base - stainless steel, and 4.
Storage racks - Square tubes of inner and outer shrouds of stainless steel completely encapsulating Boral neutron absorber plates.
Boral is a composite panel of B C/ aluminum matrix 4
clad with aluminum.
Corrosion is the deterioration that occurs in metals because of either a chemical or electrochemical reaction with its environment. The Zion spent fuel storage pool environment is oxygen-saturated high purity demineralized water containing boron as Boric acid, normally at a temperature range of 70 to 150 F.
Corrosion by chemical reaction results in a uniform surface attack that forms a protective film which decreases the corrosion rate. The Zircaloy, stainless steel, and Inconel in the spent fuel assemblies removed from
the reactor vessel would have an initial protective oxide layer which would decrease the corrosion rate once the assemblies were placed in the pool water environment.
Zircaloy and Inconel have greater corrosion resistance than stainless steel. A Zr0 layer f less than 3.15 x 10-2 inches was measured on fuel rodt that had been in the Halden (England) reactor for approximately 8 years and also on other fuel rods that had been in the Shippingport reactor after 12 years. According to B. Cox, Atomic Energy of Canada, the initial corrosion kinetics decrease on a quasi-cubic rate and, after formation of the initial protective Zr0 2
-5
-4 layer, become linear after an oxide thickness of 7.9 x 10 to 2.76 x 10 inches is attained.
In the absence of neutron irradiation Zircaloy is quite resistant to oxygen in aqueous environments and the passivity will remain in either weak acid or weak alkaline solutions.
By extrapolation of data to the 70 to 150 F temperature and oxygen-saturated high purity demineralized water environment of the spent fuel pool, it may be cal-culated that the additional linear growth of the Zr0 1 yer should be 2
-5 not more than 2 x 10 inches after 100 years, which is minute relative to the initial thickness.
Stainless steel has performed satisfactorily in the reactor as fuel cladding and in spent fuel pools without.significant deterioration being detected over a 15-year period.
Based on the observations of stainless steel fuel cladding in spent fuel storage pools, the corrosion rate of
, the stainless steel pool liner and the stainless steel storage racks
-5 in pool environments should not exceed 5.96 x 10 inches in 100 years.
J. E. Draley and W. E. Ruther, Argonne National Laboratory, have reported
-5 an average corrosion rate of 3.5 x 10 inches / year for unanodized aluminum in oxygen-saturated high purity water at 120*F which corresponds to an
-3 oxide layer of 3.5 x 10 inches in 100 years. This small amount of corrosion should not impair the structural integrity of the unanodized aluminum components in the spent fuel pool.
The electrochemical (electrolyic) nature of corrosion is a reaction which involves oxidation and reduction.
Acceleration of the reaction in high purity water requires metals that have a large electrical potential differential and which are in close contact with each other or an electrical current flow in the aqueous environment.
The electroytic potential of stainless steel and Inconel are about the same, and they can be coupled without experiencing significant electrolytic corrosion or galvanic effects.
Zircaloy is very resistant to electrolytic corrosion and galvanic effects because of its nonconducting Zr02 protective layer.
Recent surveys by G. Versterlund and T. Olsson in Sweden, A. B. Johnson of Battelle florthwest Laboratories, and J. R. Weeks at Brookhaven National
, Laboratories reveal that Zircaloy or stainless steel cladding, stainless steel tie plates, and Inconel spacers in BWR and PWR fuel bundle ast,blies have been stored in water pools for the past 20 years without evidence of accelerated corrosion. Defective fuel placed in the water pools at Windscale (England) and examined after 9 years storage showed no indicatinn of accelerated corrosion, metallurgical changes, crack propagation and hydrogenation of the Zircaloy cladding, or oxidation of the U0 fuel.
2 Release of fission products from the high burn-up fuel decreased rapidly to a relatively low and steady rate after 100 days. The detection of only 1 microcurie of Cs-137 and less than 10 ppb iodine in the pool water further indicates no degradation during water pool storage of high burn-up fuel.
Galvanic corrosion is an accelerated electro chemical reaction which occurs when dissimilar metals are in contact or near each other and connected by an ionic electrical conductor. Significant deterioration can occur only when one metal is more noble than the other, i.e., where there is a major difference in electrical potential. The aluminum in the Boral neutron absorber plates is more reactive than stainless steel and it will experience galvanic corrosion if the stainless steel tubes encapsulating the Boral are vented to the pool water environment.
Carolinas-Virginia Nuclear Power Associates, Inc, and Exxon Nuclear
corrosion tests of Boral with a leak in the stainless steel covering
-4
-4 have shown a corrosion rate of 1.8 x 10 to 3.4 x 10 inches / year for the aluminum in the Boral composite plates.
The deterioration was in the form of pitting and edge attack confined to the area of the leak path.
Pitting had no effect on the dislodgement of the B C particles 4
in the Boral core.
In fact, the B C particles are inert to pool water 4
environment and galvanic corrosion and became embedded in the aluminum oxide corrosion product which forms on the edges of the Boral plate.
The more noble stainless steel showed no attack by the galvanic coupling.
Although galvanic corrosion does occur in the unanodized aluminum of the Boral plates, it should not have any significant effect on the neutron absorption capability of the Boral, and certainly no effect on storage rack structural integrity for a period far in excess of 40 years.
The stainless steel pool liner uuld not be affected by interaction with the aluminum in the Boral plates for the following reasons:
1.
Stainless steel is more noble than aluminum and will not suffer galvanic or electrolytic corrosion, 2.
The Boral plates are completely encapsulated in the stainless steel tubes of the storage rack module, thus isolating them from the pool liner. The stainless steel storage rack base.
fonns a further protective layer between the Boral plates and the floor of the pool.
, 3.
The spacing between the storage racks (containing the Boral) and the pool liner is sufficient enough to cause electrical discontinuity.
4.
The high purity pool water is not a sufficiently strong electrolytic solution to provide a conducting path which would allow galvanic or. electrolytic corrosion to occur between any of the components in the modified pool which are not in actual physical contact with each other.
CONCLUSION Although acknowledgement has been made that corrosion will occur in the Zion spent fuel storage pool environment, it will be of no significance for at least 40 years. All the components in the Zion spent fuel storage pool, excluding the aluminum in the Boral neutron absorber plates, are constructed of alloys with the same electrical potential (or a minute differential) that have a high resistance to general chemical corrosion, electrolytic corrosion, and galvanic corrosion. The only spent fuel pool components of concern are the storage rack modules which have a galvanic coupling between the stainless steel tubes and the unanodized aluminum in the Boral.
The deterioration of the aluminum in the Boral by galvanic corrosion, however, would not be of such significance as to affect neutron shielding properties of the Boral.
The B C neutron absorber particles 4
are inert to the pool water environment.
~
, Based on the preceding facts, it can be concluded that such corrosion as occurs will have no significant effects upon the spent fuel pool components.
PROFESSIONAL QUALIFICATIONS OF FRANK M. ALMETER I joined the Commission in October, 1974 as a Materials Engineer and I am presently an Applied Mechanics / Material Engineer in the Engineering Branch, Engineering and Projects, Division of Operating Reactors, Office of Nuclear
' Reactor Regulation. Since October, 1974 my duties and responsibilities have involved the review and evaluation of materials application in nuclear power plants with specific emphasis on corrosion and water chemistry in PWR and BWR systems.
I have been appointed to the Electrical Power Research Institute (EPRI) Corrosion Advisory Committee and the NRC Corrosion Review Group for Reactor Systems.
I have the primary responsibility for the safety evaluation regarding the corrosion problems of PWR steam generator tubing, spent fuel storage pools, BWR and PWR piping systems, and snubbers.
I also have the responsibility for the evaluation of reactor coolant chemistry in both Pressurized Water Reactors and Boiling Water Reactors.
I have provided the Division of Regulatory Standards with the technical bases required for the revision of Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors."
9-I presented testimony on " Steam Generator Tube Integrity" at the Beaver Valley Unit 1, Pilgrim Station Unit 2, Jamesport Station Units 1 and 2, Byron /Braidwood Stations Units 1 and 2, and Prairie Island public hearings.
I also assisted in the preparation of testimony on this same subject for the South Texas Project Units 1/2 and the Washington Nuclear Project One public hearings.
I have a Ph.D. in metallurgy from the University of London (1959) and a D.I.C. degree in metallurgy from the Imperial College (London 1956).
I received a B.Sc. degree in Metallurgical Engineering from the University of Missouri at Rolla in 1953.
From June,1973 to October,1974, I was associated with the U.S. Consumer Projects Safety Commission as a metallurgist responsible for the evaluation of engineering, manufacturing and quality control procedures within the consumer product industry to insure production cf non-hazardous products.
I developed safety tests and basic engineering factors relative to the modification of product safety standards.
In 1971 I joined the Office of Saline Water, Department of the Interior, as Assistant to the Chief, Materials Division. My duties and responsi-bilities were the planning and directing of contracts for the development, 9
, testing, and evaluation of materials utilized in the various desalination processes.
I prepared contracts for the development of economic materials to reduce the capital and maintenance costs of desalination plants and increase their reliability.
I also conducted inspections to evaluate the corrosion performance of materials in operating plants.
I performed highly technical studies of the corrosion, mechanical, physical, and fabrication properties of a wide range of materials.
From 1968 to 1971 I was Chief Metallurgist of corporate materials technology for the Burndy Corporation with duties and responsibilities for the technical / administrative management of materials pertinent to process and product development. As manager of the metallurgical R & D laboratory, I was responsible for program planning, cost estimates, budget control and recruiting.
I established, staffed and managed a new Metallurgical Service Center to support Engineering, Manufacturing, Purchasing, and Sales / Marketing Departments.
Before I became Chief Metallurgist with the Burndy Corporation, I was a research scientist for 10 years in the aerospace industry where I conducted basic and applied research in the areas of surface science, precious metal coatings, corrosion of metals, mech anical/ physical metallurgy, fibrous composite materials, simulated high altitude environ-mental effects-on materials, fracture and surface damage in metals, alloy
,-e development, heat treating, ferrous and nonferrous alloys, ceramic /
dielectric materials, and HERF forming of metals.
From 1955 to 1958 I was a Consulting Metallurgist in the United Kingdom.
I specialized in the areas of precipitation-hardening, fatigue and tensile properties of Beryllium Bronzes.
I am listed in the American Men of Science,12th edition and Who's Who in America, 14th edition.
I was Guest Lecturer, Fairleigh Dickinson University course on " Desalination Operations," Dec.1972.
I was invited by the Electrical Power Research Institute (EPRI) to be secretary to the "First U.S. - Japan Joint Symposium on Light Water Reactors" (May 29 -
June 2,1978).
I have authored 17 publications in my professional field, Current Publication:
"An Overview of Water Chemistry for Nuclear Power Plant Safety by F. M. Almeter, Vol 28, pp 582-583,1978 Transactions of the American Nuclear Society.
I am a member of the American Society for Metals, AIME Metallurgical-Society, and National Association of Corrosion Engineers.
. I have prepared the foregoing affidavit and swear that it is true and correct to the best of my knowledge.
Em< k 9h&ElwY, Frank M. Almeter Subscribed and sworn to before me this M day of Ja cary, 1979.
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ffotary Public My Conunission expires:
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSI0ti BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
COMMONWEALTH EDIS0ff COMPANY Docket Nos. 50-295
)
50-304 (Zion Station, Units 1 and 2)
)
AFFIDAVIT OF FRALK M. ALMETER I, Frank M. Almeter, being duly sworn, do state as follows:
I am employed by the United States Nuclear Regulatory Commission as an Applied fiecharics/ Material Engineer in the Engineering Branch, Engineering and Projects, Division of Operating Reactors, Office of Nuclear Reactor Regulation.
A statement of my professional qualifications is attached to my affidavit regarding Contention 8.
This affidavit addresses the following contention:
9.
The Applicant has not discussed whether the proposed modification and long tem
- storage may cause the following effects on the stored fuel:
accelerated corrosion, micro-structural changes, alterations in mechanical properties, stress corrosion cracking, intergranular corrosion, and hydrogen absorption and precipitation by the zirconium alloys.
I have read the Licensee's Motion for Sunmary Disposition of this contention, which deals in part with matters which I discussed in my affidavit in regard to Contention 8.
I have also reviewed the affidavit of Dr. Wyvil R. Kendall, in support of Licensee's Motion.
Based upon my analysis of the contention, and my review of the foregoing materials, I am of the opinion that the proposed modification will not result in any significant effects upon the spent fuel as set forth in the contention.
I have prepared the foregoing affidavit and swear that it is true and correct to the best of my knowlerige.
5Anuk $h eG&na. En Frank M. Almeter Subscribed and sworn to before me this ( Aday of January 3 1979.
6%
3,a L f h. h r-~t Notary Public
//
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My Commission expires: hM
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~-e--.e e-m..,. _ _.
Commonwealth Edison
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'. i onefyst rm onal Plata. Ch'cago Hhnois fj Address Reply to Post Office Box 767 J
.~
Chicago, Illinois 60690 January' 24, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Station Units 1 and 2 Additional Information on Proposed Expansion of Spent Fuel Storage Capacity NRC Docket Nos. 50-295 and 50-304
Dear Mr. Denton:
The NRC Staff requested Commonwealth Edison Company to provido additional information in support of its request to expand the storage capacity of the Zion Units 1 and 2 spent fuel pool.
The request consisted of two sets of questions telecopied from the Staff on November 14 and 28, 1978.
Attachments 1 and 2 to this letter contain Commonwealth Edison's responses to these questions.
Please address any additional questions that you might have to this office.
One (1) signed original and thirty-nine (39) copies of this letter are provided for your use.
Very truly yours, "j
William F. Naughton Nuclear Licensing Administrator Pressurized Water Reactors attachments JAK 796 AMo <a s y'
SPENT FUEL POOL CAPACITY EXPANSION ZION NUCLEAR POWER PIANTS, UNITS 1 AND 2 DOCKET NOS. 50-295 AND 50-304 ROUND 2 QUESTIONS m
1.
QUESTION In regard to your response number 20, a limit on the fuel assex;bly loading is more inclusive than a limit on the enrichment.
Also, this maximum fuel loading can be obtained from just an arithmetical calculation of quality assurance data.
For these reasons, we find that a technical specification on these racks which limits the fuel loading to 39.4 grams of Uranium-235, or equivalent, per axial centimeter of fuel assembly is an acceptable method of limiting the uncertainty in keff whereas a limit on the enrichment is not.
ANSWER The appropriate technical specification change is being drafted and will be submitted after approval by our On-Site and Off-Site Review functions.
e e
1.1 9
NRC Docket Nos. 50-295/304 ATTACHMEtIT 1 2.
QUESTION In regard to your response number 21. state the bases for the dimensions of the cylindrical supercell (Figure 3) for the first benchmark calculatf]n.
ANSWER Radii Ri and R2 of the cylindrical supercell (Figure 3) is obtained by conserving the corresponding areas of the 9 x 9 basic fuel pin assembly.
Radius R3 is obtained by adding the aluninum wall thickness to the fuel pin assembly and then conserving the area. Ra'dius Ru is obtained by adding to R 3 half the thickness of the boral core.
Constants used for cylindrical supercell dimensions:
Basic fuel pin assembly array 9x9
=
Fuel rod pitch 0.75 inches
=
Thickness of AL wall
.041 inch
=
Thickness of boral plate 0.168 inch
=
h(0.75-2x0.041-0.168)
Thickness of water layer
=
0.25 inch
=
Area of fuel region:
(fuel rod pitch x 9)2
=
(0.75 x 9)2 sq. in.
=
(6.75)2 sq. in.
=
(6.75 x 2.54)2 sq. cm.
=
293.95102 sq. cm.
=
nR 2
=
3 (293.95102S gg
=
cm.
9.67303 cm.
=
G 9
2.1
NRC Docket Nos. 50-295/304 ATTACIE4ENT 1 2.
ANSWER (continued)
Area of (H O + fuel) re91on 2
nR 2
=
2 (6.75 + thickness of water layer)2 sq. in.
=
(6.75 + 0.25)2 sq. in.
=
(7 x 2.54)2 sq. cm.
=
316.1284 sq. cm.
=
(316.1284 h
R 2
10.03129 cm.
=
Area of (H O + At + fuel) region 2
(7 + thickness of AL layer)2 sq. in.
=
(7.041 x 2.54)2 sq. cm.
=
319.84246 sq. cm.
=
HR 2
=
3 (319.8 )cm.
R
=
3 n
10.09005 cm.
=
R3+
thickness of boral plate Rg
=
10.09005 +
x 0.168 x 2.54
=
10.30341 cm.
=
2.2
NRC Docket Nos. 50-295/304 ATTACHMEllT 1 4
Al H0 BORAL R4 b
/
"2 N
rua.
(
s y=9.67303CM R - 10.03129 CM 2
R = 10.09005 CM 3
R = 10.30341 CM 4
Figure 3.
1-0 Supereell Ccnfiguratien for First Sanch:::srk Calculations 2.3
oc e os.
-. /
ATTACIIME!TT 1 3.
ppESTION In your calculations of the two critical assemblies with Boral, describe how you accounted for the self-shielding of the boron carbide particles in the aluminum matrix.
ANSWER In the benchmark calculations, the Boral core was homogenired and then the cross sections were obtained from XSDRN, which is a one dimensional discrete ordinates spectral averaging code.
There was no account for pa'rticle self-shielding since the range of particle sire is60-200 mesh with a mean sire of 175 mesh and, therefore, self-shiciding ef fects are negligible.
3.1 9
NRC Docket Nos. 50-295/304 ATTACHME17f 1 4.
QUESTION In regard to your response Number 23, the NRC requires an on-situ neutron attenuation test to verify the presence of the boron.
This is in addition to the Quality Assurance Program you described.
Provide a description of the neutron attenuation test that you will perform at the Zion plant to statistically show with 95 percent confidence that the boron i:e not missing from more than one out of every sixteen plates.
ANSWER A neutron posion verification test will be conducted at the Zion plant after the; racks are installed in the pool.
This will be a qualipative test to statistically show with 95 percent confidence that the boron is not missing from more than one out of every sixteen plates.
This procedure is similar to the poison verification tests conducted at Montecello and TVA by National Nuclear Corporation utilizing their proprietary equipment.
O 4.1
NRC Docket Nos. 50-295/304 ATTACHMENT 1 4
5.
QUESTION For the proposed type of racks, a surveillance program is required to show the continued presence of boron throughout the life of the racks.
Provide a description of the boron surveillance program that you will perform.
ANSWER See attachment "A",
Neutron Absorber Sampling Plan - In Pool.
9 5.1
NRC Docket Nos. 50-295/304 ATTACHMENT 1 ATTACHMENT "A" NEUTRON ABSORBER SAMPLING PLAN - IN P00L A sampling plan to verify the ability of a neutron absorber material employed in the high density fuel racks to withstand the long-tem environment is described.
The test conditions represent a restricted flow of water over the neutron absorber material. The samples will be supported adjacent to and suspended from the fuel racks. Eighteen (18) test samples are to be fabricated in accordance with Figure 1 and installed in the pool when the racks are in-stalled.
The procedure for fabrication and testing of samples shall be as follows:
1.
Samples shall be cut to size and dried in an oven for five hours at 175'F, followed by a cycle at 600'F for three hours.
2.
Samples shall be weighed imediately following removal from the oven and weight in milligrams recorded for each sample.
- 3. ' Samples shall be fabricated in accordance with Figure 1 and installed in pool.
4.
Two samples shall be removed per schedule shown in Table 1.
5.
Carefully cut samples apart at the weld without damaging the neutron absorber. Wash with a soft brush in a mild abrasive and detergent solution, imerse in nitric acid to remove surface products, followed by a rinse of clean water and alcohol. Dry in a 175'F oven for five hours, followed by a cycle at 600*F for three hours.
6.
Weigh the samples and evaluate the weight change in the neutron absorber material in milligrams per square centimeter per year.
5.A-1
NRC Docket Nos. 50-295/304 ATTACHMENT 1 7.
. If pitting is present, the depth of the four major pits are to be reccrded and the average pit penetration in mils of an inch per year determined.
8.
Retain two (2) samples.
9.
Prepare report of sample test results and observations.
9 m
O 5.A-2
a.
ATTACHMENT 1 TABLE 1 Date Installed INITIAL FINAL WEIGHT PIT SAMPLE WEIGHT WEIGHT CHANGE PENETRATION
' NO.
SCHEDULE (mg/cm2-Yr)
(mg/Cm -Yr)
(mg/Cm -Yr) mil /Yr 2
2 1
2 90 day
'r 3
4 180 day v
5 6
1 year V 7
8 5 year - V 9
10 10 year v 11 12 15 year V 13 14 20 year V 15 16 30 year
'r 17 18 40 year y O
5.A-3
s.
ATTACIB1ENT 1 d
q Em o
a R
a:
2 N
I b
(TYP-SEAL WELD
$~
0.075X.125-J {g.V' '304Su c
'" 4 S' 'S 0.062" DIA.
k HOLE-TYPICAL TOP & BOTTOM ORILL BEFORE ASSEMBLY k
NEUTRON 7
ABSORBER 5%
%x h
6 x
.030" 304 SST 18 SAMPLES Figure 1 5.A-4
NRC Docket Nos. 50-295/304 ATTACIU4IRTT 2 i
SPENT FUEL P00L CAPACITY EXPANSION ZION NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-295 AND 50-304 ROUND 3 OUESTIONS QUESTION NUMBER 1:
Provide a more detailed description of the inter-tube welded connection; include drawings if possible. Specifically discuss if the tubes are welded continuously to each other the full length of the tube or only at discrete intervals. Also discuss the structural members or plates used in this connection.
RESPONSE
The tubes typically have bars (flat plates) attached to the specified corners as shown on Drawing No. 1000483. These bars are welded the full length of the tube.
The tubes with the bars attached are welded into cluster subassemblies per Drawing No. 1000484. Again, they are welded together the full length.
These clusters are then welded to other clusters and the base assembly as shown on Drawing No. 1000490, which is typical of the other rack size assemblies. These cluster attachment welds are again the full length of the tube.
1.1
NRC Docket Nos. 50-295/304 ATTACHMENT 2 OUESTION NUMBER 2:
Provide a detailed description of the analysis or con-siderations used to establish that t he tube inner com-partment containing t he Boral remains scaled against leakage. What are the potential consequences of pool water leaking into the area containing the Boral?
_ RESPONSE:
Consideration was given to maintaining the inner compartment containing the Boral sealed against leakage, and on the basis of the information available, it was decided to vent the Boral containing compartment and allow pool water to enter and exit without restriction.
The consequences of pool water in the area containing the Boral are discussed in the Brooks and Perkins' report, "The Suitability of Brooks & Perkins' Spent Fuel Storage Module for Use in PWR Storage Pool," Report No. 578 dated July 7, 1978, and did confirm the Boral panels are capable of meeting a forty year service life.
2.1
NRC Docket Nos. 50-295/304 ATTACIIMENT 2 QUESTI0fl !! UMBER 3:
What considerations have been taken to prevent off-gas from the Boral and swelling of the tube?
RESPONSE
The off-gas from the Boral will not be a problem in a vented tube, thereby eliminating any swelling of the tube.
G e
O 3.1 8
NRC Docket Nos. 50-295/304 ATTACHMENT 2 QUESTION HUMBER 4:
Provide the basis for concluding that an empty rack will slide further than a full rack under seismic loadings.
Provide a drawing of the equivalent stick model used in the sliding analysis and indicate the points where these loads were applied.
RESPONSE
Since the spent fuel racks are stored under water, their seismic movements are caused by the horizontal inertia of " virtual mass" which is the sum of the body mass and the " hydrodynamic mass".
The magnitude of the hydrodynamic mass depends on the shape of the rack body and the density of water, and so is independent of whether the rack is loaded or empty. The friction force resisting the seismic movement is proportional to the buoyant weight of the rack and its contents, but because of larger horizontal " virtual mass" per unit weicnt, the ratio of inertia force to friction force is more for empty racks. For this reason, it was concluded that empty racks will slide further than a loaded rack under seismic loadings.
Figure _4.1 shows the equivalent stick rodel used in the sliding analysis. Time history of SSE seismic ovement was applied at Node 8 which represents the pool floor.
4.1
NRC Docket Nos. 50-295/304 A_TTACIIMElff 2 EL. 169.8 O1 O'
113.0
(),2 66.0 3 3 36.0
( )4 Sliding Element 0.2
=
y 16.0 5
6.0 6
0.0 7
8 FIGURE 4.1 LUMPED MASS STICK MODEL FOR SLIDING ANALYSIS e
4.2
NRC Docket Nos. 50-295/304 ATTACIU4EITT 2 QUESTION NUMBER 5:
Provide the value of the " rattling factor" used in the seismic analysis.
RESPONSE
Rattling factors account for the nonlinear effects of the fuel bundles moving within the spent fuel rack cells.
The magnitude of these factors depend on the structural and damping properties of the rack and fuel bundles as well as on the level of excitation.
The rattling factors used for the Zion rack evaluation ranged from 1.10 (for SSE loading of 10 x 11 size rack in the direction parallel to the longer side) to 2.57 (for OBE loading of 5 x 10 size rack in the direction parallel to the shorter side). These are upper bound factors computed using a conservative assumption that all the fuel bundles inside a rack " rattle" in phase.
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NRC Docket Nos. 50-295/304
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ATTACHMENT 2_
QUESTION NUMBER 6:
Provide a description of the thermal gradient analysis considera-tions, include the thermal gradient considered and a discussion on why this was considered a conservative estimate of the worst case, i.e., the gradient between a full and empty cell.
RESPONSE
The thermal gradient due to the placement of a hot fuel bundle in an empty cack is as shown in Figure 6.1 (0 F at the rack bottom and 32.380F at the top). Stresses caused by this thermal gradient were computed using a finite element model which is also shown in Figure 6.1.
To minimize the computation cost, only the central part of the rack body was modeled with the sides restrained from lateral translation, thus representing the worst case and predicting conservative stresses.
It is important to note here that, near the top of the rack the thercal gradient, and hence, the resulting thermal stresses are maximum, but the dead load and seismic stresses are minimum. Thernal gradient near the bottom is very small where the seismic and dead load stresses are maximum.
6.1
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ATTACHMENT 2 QUESTION NUMBER 7:
The fuel bundle drop analysis considered a drop at the most " critical" location on the rack, provide a descrip-tion of this location and drawings to illustrate the postulated configuration of the fuel bundle at impact.
Discuss the procedure for limiting the height of the fuel bundles above the racks to 24 inches.
Discuss the conse-cuences of a fuel bundle dropping straight through the tube and bmpacting the bottom of the rack.
RESPONSE
The top corners of the racks were found to be the most critical locations for evaluating the consecuence of dropping a fuel bundle.
When tha fuel bundle drops on the rack, the cross-sectional araa of tho' cell walls absorbing the impact energy increases as the load is transmitted downward.
Since this gradually-increasing cross-sectional area is minimum when the fuel bundle drops on a corner, the latter ccnstituted the most crit ical location.
For evaluating the consecuences of fuel bundle drop, the bundle configuration was assumed to be vertical at impact 7.1
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ATTACIIMENT 2 (Figure 7.1).
An inclined drop was judged to be less critical f rom the following considerations:
(a) The in pact area will be larger, (b) The impact will be " softer" because of the flexibility of the fuel bundle itself.
-The length of the fuel handling tools and interlocks on the fuel pool bridge hoist limits the distance bitween the top of the rack and fuel assembly to less than 24."
Fuel assemblics thus cannot be raised above the 24" limit.
Consequences of the fuel bundle dropping straight through the tube and impacting the bottom of the rack have been invesitgated.
The method of analysis and the results obtained are briefly described below:
The fuel bundle will drop approximately 164 inches from the top of the rack to the rack base plate.
If the fluid drag on the bundle is neglected (a conservative assumption), the impact energy will be approximately 254,000 in-lbs.
This energ: will be absorbed by the following mechanisms:
(a) Since the fuel bundle is " soft" as compared to the rack, a large part of energy will be absorbed by the collapsing of the fuel bundle, thus limiting the maximum load transmitted to the rack.
7.2
NRC Docket Nos. 50-295/304 ATTACHMENT 2 (b) A part of the enrgy will be absorbed in bending the base plate inside the fuel cell.
If.it is conservatively assumed that, in the extreme case, the bending of the base plate causes a localized plastic hinge to form at the intersection of the tube wall and the base plate, the upper bound stress due to the accidental fuel bundle drop can be evaluated by applying at the cell wall the load required to form such a localized plastic hinge.
This was done using a finite element model of a portion of the rack in the vicinity of the bundle drop.
Loads were computed and stresses were determined at the bottom of the tube wall.
The poison material is capsulated at a height of 4.26 inches from the base plate.
Maximum stress in the outer tube wall at that level was computed to be 18.9 ksi, well below the yield stress limit of the material.
Also, it has been observed that the loads dissipated rapidly in the structural panels, indicating that the overall structural integrity of the rack is not impaired.
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NRC Docket Nos. 50-
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7.4
NRC Docket Nos. 50-295/304 ATTACHMEtTr 2 QUESTION NUMBER 8:
The results from the sliding anaiysis indicated that one rack could potentially slide 1.31 inches and that the minimum gap between any two adjacent racks is 2.4 inches.
Discuss the basis for concluding that two adjacent racks could never slide out of phase, actually slide towards each other, and impact.
They potentially could close a gap of 2.62 inches (180 degrees out of phase).
RESPONSE
Each rack can potentially slide a distance of 1.31 inches towards each other. Powever, providing a space between the two adjacent racks less than twice this distance was justified from the following considera tions:
(a) 1.31 inches is the peak movement of the rack obtained from a time history analysis. Under identical conditions, the adjacent rack would be in phase and would also move 1.31 inches in the same direction, in whic't case the original gap between the two racks would remain unaltered. However, since the adjacent rack is not likely to have identical conditions, the gap status is likely to change. Only if the two racks have identical conditions and their movements are exactly 1800 out of phase, the minimum required gap to preclude impact would be the absolute sum of the movements of two racks, i.e., 2.62 inches. However, the probability of satisfying both these conditions simultaneously is extremely small, which justifies the use of a lesser gap.
If SRSS method is applied to account for the low probability of the phenomenon, the required gap would be 1.414 times 1.31, i.e.,1.85 inches which is less than the 2.4 inches gap provided.
8.1
NRC Docket Nos. 50-295/304 ATTACIIMENT 2 (b) 1.31 inch is the predicted movement of the empty rack computed using the minimum coefficient of friction.
It is judged that the loaded racks would slide significantly less than the empty racks. The reasons have been outlined in the response to Question No. 4.
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NRC Docket Nos. 50-295/304 I
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l All structural materials, with the exception of Boral, are stainless steel grade 304.
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