ML19241B721
| ML19241B721 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 05/29/1979 |
| From: | Almeter F, Lantz E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19241B710 | List: |
| References | |
| NUDOCS 7907230226 | |
| Download: ML19241B721 (21) | |
Text
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UNITED STATES OF #1 ERICA NUCLEAR REGULATCRY C0f1 MISSION BEFORE THE ATGMIC SAFETY AND LICENSING BOARD In the itatter of
)
)
COMMONWEALTH EDISON CCMPANY
)
Docket Nos. 50-295
)
and 50-3C4 (Zion Station, Units 1 and 2)
)
NRC Staff Testimony on Contentions 2(e)(3), 2(e)(4), 2(h),
2(i), 2(j), and 2(k)
By Frank M. A' meter and Edward Lantz In this testimony, we address the above listed contentions. The following discussion regarding the experimental data, operational experience, and surveillance provides material facts to show that whatever corrosion occurs will have no significant effect on spent feel pool components.
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 assemolies (Z?rcaloy clad fuel rods, stainless steel tie olates, and Inconel spacers);
3.
Storage rack base (stainless steel); and 4
Storage racks (Scuare tuDes of inner and outer snrcuds of stainless steel completely encapsul ating Boral neutron aosorber plates.
Scral is a cccccsite panel of 3 C/ aluminum matrix clad with 1100 aluminum 4
alloy).
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. The Zion spent fuel storage pool environment is typical of PWR plant systems (i.e., oxygen-saturated high puri ty demineralized water containing boron as Boric acid, normally at a temperature range of 79'F to 150*F and a pH range of 4.5 to 6).
Contention 2(e)(3) and 2(e)(4)
The amencment request and supporting documentation do not adequately discuss monitoring procecures.
In the i.ght of the proposed modification and long term storage of nuclear spent fuel the Applicant should clarify the following:
(e)(3) Methods or detecting the loss of neutron absorber material and/or swelling of stainless steel tubes in storage racks.
(e)(4) Details of a corrosion test prearam to monitor per-formance of materials usea in the construction of the racks.
Commenwealth Edison has committed to an in-gvol neutron absorber surveillance program to verify the ability of a neutron absorcer material employed in the
}
high density fuel racks to withstand the long-term env'ronment.
The test conditions represent a restricted ficw of water over tne neutron absorber material. Samples will be supoorted adjacent to and suspended from the fuel racks. Eignteen vented test samples of 3ar31 enclosed in stainless steel are to ::e fabricated and installed in :ne ; col when the rxXs are installed. Tne test samples are representati <e of material s used in the construction of the storage racks. The orogram is cesigned to evaluate the long-term effects of galvanic corrosion cetween dissimilar metals with a large electrical potential cifference and the integrit.y and III Letter from W. ~. Naughtoa to H. R. Centon, dated January 24, 1979.
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. the capability of Soral as a neutron aosorber material in spent fuel pools.
The welded seams of the stainless steel covering on tne Boral will test the resistance of welded stainless steel to stress corrosion cracking in the spent fuel pool environment. The test samples will be examined every 90 days for the first year of exposure, then every 5 years up to 20 years, and af ter 20 years every 10 years up to 40 years.
Since 'che rate of loss of boron cue to corrosion will be small in any event and this rate generally cecreases with time in the pool, we find that this program is satisfactory for monitoring the condition of the Boral plates and the continued presence of the baron.
Also in its January 24, 1979 submittal, CEC stated that it will vent the Boral containing ccaponents and allcw pool water to enter and exist without restriction.
It is our finding that this venting will eliminate the potential for any significant amount of swelling of the stainless steel tubes.
Contentien 2(h)
The amencment recuest and succorting cocumentation nave not analy:ec the long term
- electrolytic corrosion effects of using cissimilar alloys for tne cool liners, cipes, storage racks anc storage racx cases, sucn as the galvanic corrosion ce: ween unanocized aluminum, as is usec in Brec(s and Perkins storage racks, and tne stainless steel ; col liner.
""Long term" storage wculd include storage during the life of the reactcr.
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. Contention 2(h) assumes the electrolytic corrosion will affect the long term integrity of dissimilar alloys, such as the galvanic corrosion between unanodized aluminum and stainless steel in the pool water environment.
Corrosion is the deterioration that occurs -in metals because of eitner a chemical or electrochemical reaction wi th its envirtnment. Metallic corro-sion, in water, is a combination of anodic and cathodic reactions (electro-chemical oxidation / reduction) which involves the development of protective films. Acceleration of the corrosion reaction in high purity water requires either:
(1) an increase in aqueous temperature, (2) a change in the elec-trolytic nature of the anocic.nc cathodic reactions by metals that have a large electrical potential differential or an eiectrical c.rrrent flow in the aqueous environment, (3) a change in the ionic concentration of the aqueous electrolyte, or (4) coupling of dissimilar metals by direct contact or an electrical concuctor (galvanic corrosion). Significant galvanic corrosion can occur only when one metal is more noble than tne other (i.e.,
where there is a major difference in electrical potential).
The Zircaloy, stainless steel, anc Inconel in the spent fu:-l assamolies removed from the reactor vessel would have an initial protective oxice layer anich aaula cecrease tne corrosion rate once tne assemolies aere placed in the pool water environment. Z'rcaloy and 'nconel nave greater corrosion
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/o resistance than stainless steel. According to B. Cox, Atomic Energy of NI Canaca
, C. Breden, Argonne Natf onal Labs, and E. T. Hay c: U ". Sureau I
of Mines
, the initial corrosion kinetics for Zircaloy increase at a quasi-cubic rate and, after formation of an initial protection Zr0 l aye r 2
of 2.42 x 10-inches af ter 1,250 days at 600*F,'become linear at a corrosion rate of 1.4 x 10-5 inches / year.
A. B. Jahnson has noted that the corrcsion ra te of Zi rcal oy i s muc n l es s a t 195 *F te 212 *F, ( i.e., 1.18 x 10- to 3.54 x 10-7 i nches / year)
In the absence of neutron irradiation, Zircaloy is quite resistant to oxygen in aqueous environments and the passivity will remain in either weak acic or weak alkaline solutions.
If one assumes that the corrosion rates at tne temparature range cited by Johason would be the extreme case for extenced periods in the Zion spent fuel pool, the additional linear growth of the Zr0 lay e should be not more than 4 x 10-b inches af ter 100 years, 2
which is minute relative to the initial thickness.
Stainless steel has performed satisfactorily in the reactor water environment as fuel clacding anc in spent fuel pools witncut significant deterioration being cetected over a 15-year period. Based on tne corrosion cata puolished for stainless steel in more aggressive envircrments than in scent fuel storage ::cols, the corrosion rate of tne stainless steel pool liner anc the UI S. Cox, J. Nucl ear Materi al s, 26, 310 (1965).
(3 )D. E. Thomas, Proc. :nternational Conference on the Peaceful Uses of Atomic Energy, Geneva 1955, Vol. 9, p. 407, United Nations, N.Y. (1956).
(4 }3attelle Pacific Northwest Lac. Report (3NWL-2256, Septemoer 1977).
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. stainless steel storage racks in pool environments should not exceed 5.90 x 10-7 inches per year during the expected plant life.
Unanodized 1100 aluminum also forms an initial protective film within the first 5 days of immersion in distilled water before the corrosion rate becomes linear.
J. E. Draley and W. E. Ruther, Argonne National Laboratory have reported that tne corrosion rate in distilled water (pH7) ranges from 1.1 x 10-inches / year at 122*F to about 7.0 x 10-5 inches / year at 212*F.
-5 In dilute acid (pHS), tne corrosion rate at 212 7 is 3.7 x 10 i nches/ year
-5 and decreases to 3.57 x 10 incnes/ year at 122*F.
The corrosion rates in slightly alkaline solution (pH 8.5) at 212*F were oDserved to be approximately the same as in distilled water, but they are generally higher at 122*F
-5 (5.83 x 10 incnes/ year). This indicates that the corrosion rate of aluminum in aqueous solutions in a ran;. of pH 5-8.5 decreases with decreasing tempera-ture and becomes relatively minute at temperatures below 122*F.
The electrolytic potential of stainless steel and Inconel are about the same, and they can De coucled without ex::eriencing significant electrolytic corrosion or galvanic effects. Zircaloy is very resistant to electrolytic corrosion anc galvanic effects cecause of its nonconducting Zr02 orotective 1ayer. However, tnere is a major cifference in electrical octanti31 aetween aluminum and stainless steel. The aluminum claccing of tne Scral neutron aosarcer plates in contac witn the stain) ass steel tubes encao-sulating the Scral is are reactive and it will ex;:erience galvanic corrosion upon exposure to :ne pool water environment.
dI f roc. International Conference on tne Peaceful Uses of Atcmic Energy, Geneva 1955, Vol. 9, p. 391, Uni ted Na tions, N.Y.
(1956).
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. R. A. U. Huadle, AERE-Harwell( }, states that in a pH 4-8 range the dissolution of the protective film on aluminum is negligible and the slight acidity or alkalinity reduces the resistance of the aluminum to anocic attack in the adjacent area of the galvanic couple, resulting in pitting corrosion (a highly localized penetration at only a.
few spots).
Oak Ridge National Laboratory tests on 1100 Aluminum in contact with stainless steel in an oxygenated demineralized water at a pHS ana at 194*F also show that galvanic corrosion between dissimilar metals results in pitting corrosion of the anodic material (aluminum) ar.d no attack of the cathc lic material (stainless steel). The corrosion process on the aluminum not -
contact with the stainless steel was typical of 1100 aluminum in distilled water (i.e., a rapid increase in corrosion rate price-to a low.. stable linear corrosion rate. This indicates that galvanic coup'ing between stainless steel and alumin;m does not accelerate general corrosion of aluminum).
Exxon Nuclear nas conducted tests on samoles of vented scent fuel storage cells in a baratea deionized water (pri5) at la6*C anc in an alkalinated f
ceionized water (oH9.6) at 153.4 C for periods ao to one year '. The ceterioration was in the forn of pitting if tne aluminum claccing cn the Scral cl ates and eage attack confined to the area of tne leek path. The frequency of pitting did not increase witn icnger ex::osure tme. Pitting had (6)Prcc. International Conference on the Peaceful Uses of Atomic Eneroy,
'ieneva 1955, Vol. 9, p. 403, United Nations, N.Y.
(1956).
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I7IReport No. ORNL-TM-1030, September 1966.
(8) Exxon Nuclear Company, Inc. Report No. XN-NS-IP-00nhP, March 1979.
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. no effect on the dislodgement of the B C particles in the Boral core.
4 In fact, the cermet B C particles are inert to pool water environment 4
and galvanic corrosion and became embedded in the aluminum corrosian product which forms on the edges of the Boral plate. The more noble stainless steel showed no attack Dy the galvanic coupling. The tests also indicated that galvanic coupling does not accelerate the general corrosion of the aluminum cladding on the Boral nor the aluminum in the Boral matrix.
Although galvanic corrosion does occu* in the unanodized aluminum of the Boral plates, it should not have any significant effect on the neutron absception capability of the Boral, and certainly no effect on storage rack structural integrity for a period far is excess of 40 years.
The stainless steel pool liner would not be affected by interaction with the aluminum in the Boral plates for the folicwing reasons:
1.
Stainless steel is more noble than aluminum and will not suffer galvanic or elect clycic corrosion, 2.
The Boral plates are ccmpletely encacsulatec in the stainless steel tubes of :ne storage rack mccule, thus isolating trem frcm the :: col liner. The stainless stee' storage rack case forms a further protective layer tetween the Boral plates and the floor of the pool.
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The spacing between the storage rcks (containing the Boral) and tne pool liner is sufficient to cause electricial discontinuity.
Contention 2(i)
The Applicant has not discussed whether the proposed modification and long term storage may cause the follcwing effects on the stored fuel' accelerated corrosion, micro-structural changes, alterations in mechanical properties, stress corrosici cracking, intergranular corrosion, and hydrogen absorption and precipitation by the Zirconium alloys.
As discussed in the response to contention 2(h), accelerated corrosion of Zircaloy at spent fuel pool temperatures is not expected to occur for twu reasons:
(1) the fuel removed from the reactor has an initial Zr0 layer 2
which retards the corrosion rate, (2) the corrosion rata at pool temperatures is several orders of magnituce less than that at reactor temperatures. Raactor 13 2
9
-I irradiation (fast neutron fluxes > 10 n/cm -sec and gamma 10 Rhr doses) can produce minor accelerated corrosion (e.g., fuel that remained in the Shippingport reactor for 12 years resulted in less than 2 x 10-4 inch I
penetration of the claading). According to R. C. Asher gamma doses of 6
-l 0
-I 6 x 10 Rhr and beta cases of 7 x 10 Rhr have no effect on the corrosion rate of fuel claccing. The maximum gamma doses (10 - 10 Rhr ) ex::ected in the spent fuel pool are many order of magnituces icwer wnich shculd not affect the fuel curing storag. Jo to a0 years.
Zircaloy is quite resistant to general corrosion in aqueous solutions containing chloride (e.g., less than 4 x 10 inches / year in 20,CCO ppm chl rice at 200*F).
In order for pitting corrosion, intergranular corrosion.
L ' Corrosion Science, Vol. 10 (1970).
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or stress corrosion cracking to occur in aqueous solutions containing 0.15 ppm chloride, generally specified in PWR storage p ols, Zircaloy must be polarized anodically to break down the protec lo Zr0 I 7'I' 2
To achieve this, tnere must be galvanic coupling to a Cathodic material with a large electro-poten,tial differential or/and applied current. Since the stainless steel and Inconel in the fuel bundle hava about the same electrical potential as Zircaloy and stray current potential in pool water is minute, accelerated general corrosion, pitting intergranular corrosion, ano stress corrosion cracking of fuel cladding should not occur in the Zion spent fuel pool for at least 40 years.
Alternations in the microstructure and mechanical / physical properties of metal. is achieved oy metallurgical processes (viz., heat treating, welding, alloying, f abrication, etc. ) or by irradiation. High levels or fast neutron 10 2
fluxes such as those occurring in the reactor vessel, >10 n/cm -sec (E > 1 MeV), are required to cause such additional changes in the yield strength, ductility, or resicual stresses of the fuel assemoly bundle.
Furthermore, irradiation creep causes relaxation of resicual stresses and tne nign fracture toughness of irradiatec Zircalcy wculd minimize tne potential for brittle fracture.
Galvanic ccucling of Zircaloy to aluminum will cause catncdic polarization at tne Zircaloy, which makes it sensitive to nydroget absorption at temp-eratures below 212*F.
Hcwever, the only aluminum in the spent fuel pool is tnat in the 3cral neutron absorcer plates which are enclosec in sta...ess m fl f< v L
. steel isolating the aluminum from the Zircaloy. Therefore, tnis eliminates the probability of progressive hydriding (pr ecipitation) and emorittlement of the fuel cladding during long-term pool storage.
Recent surveys by G. Versterlund and 7. Olsson in Sweden (10), A. B. Jonnson of Battelle Northwest Lab;:atories(lli', and J. R. Weeks at Brookhaven Na t' ' nal Labo'ratories(I2, reveal that Zircaloy or stainless steel cladding, stainless steel tie plates, and Inconel spacers in SWR and PWR fuel bundle assemblies have been stored in water pools for the past 12 years without evidence of accelerated corrosion. Defective 'uel placed in the water pools at Windscale (England) and examined after 9 years storage showeo no indica-tion of accelerated corrosion,.netallurgical changes, hydrogenation and crack propagation of the Zircaloy cladding, or oxidation of the UO fuel. Release 2
of fission products from the high burn-up fuel decreased rapidly to a relatively low and steady rate af ter 100 days. The detection of only 1 microcurie of Cs-l?? ana less than 10 ppo iodine in the pool water furthe-incicates no cegracation turing water pool storage of hign Durn-up fuel.
IM ASEA-Aten ? reprint No. R3 73-29 (January 13, 1978)
(11)Satte11e Pacific Nortnwest Lao. Report (3N'al-2256, Septemoer 1977)
II2 NRC Report (3NL-NLREG-23021, July 1977) an r O.,'
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- 12 Contention 2(j)
The amendment request and supporting documentation do not give sufficient data to fully assess the durability and performance of the Boral-stainless steel tubes which form the spent fuel storage racks:
(1) there is inaaequate analysis of :ne co-rosion rate of the tubes, (2) there is no calculation of the effect of water chemistry on tne Soral within the stainless steel, (3) there is no mention of the possible swelling of the Soral within the stainless steel tubes, a condition which could affect, among other things, removal of fuel assemolies from the racks.
Exxon Nuclear has conducted tests of..
les of vented spent fuel storage cells in a borated water (pH5) at 146'F and in an alkalinated water (;H9.6) at 153.4*F for periods up to one ye;r The deterioration was in t-
" arm of pitting of the aluminum cladding on the Soral plates and edge attack confined to the area of the leak path. The fraquency of pitting did not i'ncrease with longer exposure time.
Pitting had no effect on the cis;cdge-ent of the 3 C particles in the Boral core.
In fact, the 3 C particles are 4
4 inert to pool water environment and galvanic corro: ion and became amaecded in tne aluminum corrosion product which foras on tne ecges of tne Soral plate.
The more noble stainless steel showed no at ack ty tne galvanic roupling.
Tne cests also indicated that galvanic couo!1ng coes not accelerate tne general corrosion of the aluminum clacaing an tne Scral nor tre aluminum in tne Soral matrix.
Ucoa exposure of Scral t' the spent fuel col environment gas generation will occur nitnin tne first 5 cays due to tne rapid initial corrosion of the aluminum, after which there is a minute excess gas evolved during the sicw
- l Exxon Nuclear Comoany, Inc. Report No. XN-NS-TP-009M4P, March 1979.
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. steady state corrosion rate of the aluminum. However, the Zion storage racks will De vented at the top allowing the gas to escape, thus minimizing the procability of swelling and any effect on the removal or fuel assemblies The Exxon Nuclear tests on vented storage cells indicated that the possibility of small bulging in the Soral plates would be rare, random, and self-limiting in size minimizing severe defamation of the stainless steel covering.
Boral has been exposed to the reactor coolant water of the Brookhaven Medical Research Reactor for 20 years.vithout experiencing significant de terioration The Boral plates are the typical 3 C/ aluminum core 4
~
with an aluminum cladding, but not covered with stainless steel. The reactor water is nigh-purity cemineralized water at a temperature range of 75'F to ll5'F.
Boral samples examined in September 1978 showed the characeristic protective film on the aluminum cladding and a slight corrosion product on the edges where the aluminum in the Boral core was exposed to the water. The samples showed no loss of the inert 3 C.
Neutron attenuation Msts on five sanpies 4
showed no loss in neutron absorption properties, indicating that no baron was lost frcm tne Soral core during tne 70 years exposure.
Contention 2 f k)
The amencment request and succorting cocumentation co not consider possible cegeneration of tne Soral censity cue eitner to generic aefects or to mecnanical f ailure wntch would ciminisc :ne effectiveness of Soral as neutron aosorcer, thus leacing to criticality in :ne s;;ent fuel pool.
I I NRC Report (3NL-NUREG-25522, January 1979).
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. As stated in Section 2.1 of the Safety Evaluation, the criticality calculations for the modification were made using the stated minimum value of 0.02 grams of the boron - 10 isotope per square centimeter of Boral plate. During our review of the application we asked the Commonwealth Edison Company how the minimum value would be verified for all of the Boral plate area.
Its response, which was transmittea to us by letter dated January 24, 1979, is as follows:
The manuf acture of Boral and fabrication into plates ts controlled by the Brooks and Perkins Quality Assurance Program, which includes detailed procedures for the inspection and verification of boron-ten loading in each Boral plate. The inspection plan for these activities will include the following:
1.
Documented laboratory analysis for chemical boron content and isotopic boron-ten content of each lot of boron carbide powder.
2.
Inspector's verification of the weighing and mixing of baron carbide and aluminum powder into a batch, according to the production plan, and the assecoly of this batch into one or several ingots, all identified and traceable to the batch.
3.
Occumented laboratory ana'ysis of a selected sample of batch mixes to verify ocron Caroide content.
4 The rolling of the ingot into 3 sneet, and tne sucsequent olanking of a sneet into two or nree plates, each icentified anc traceacle to the ingot anc Datcn.
5.
Visual ins::ection of the perimeter cf eacn plate to verify tnat tne core extends to tne ecge (:ne aluminum edge filler has teen ccmoletely shearea away), and a check of plate thickness at several points.
6.
Cocumented laboratory analysis, according to a sampling plan, of tne boron carnice content of coupons cut from each end of each plate.
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, 7.
Documented neutron transmission tests over the surface of a selected sample of plates to veri fy the uniformi ty of coron-ten loading across the entire plate area.
The detailed sampling plans will De established prior to manufacturing, based on the specific production lot sizes to be used, in order to establish a 95 percent confidence that the minimum areal density of 0.02 grams of baron-ten per square centimeter is present over the entire area of each plate.
It is estimated tnat the nominal level of baron-ten areal density will be 0.0220 + 0.0012 grams per square centimeter, to provide an assured minimum level of 0.02.
The exact range will De estaolished in the detailed production plans.
From this reponse we concluded that there will indeed initially be at least 0.02 grams of boron-ten per square centimeter of Boral plate.
Frcm the physical facts and experimental results presented in this testimony we conclude that tne areal density of baron-ten in the Boral plates will not be reduced by chemical means below 0.02 grams of baron-ten per square centimeter of Soral plate over the life of the racks.
In regard to a possible mecnanical failure wnicn could diminish the effectiveress of Boral as a neutron absorcer, Ccmmonwealtn Edison in their April 25, 1975 suomi ttal stated tnat only tre outer stainless steel tuces are wel ded together, and tney form tne load-carrying, structural part of the rac<.
These outer stai11ess steel tuces are designed to carry tne mecnanical loads imposed during a Safe Shutcown EartncuaKe ($$E). Inus the mechanical strengtn of the Boral plates is not relied on in tne design.
Also, tne inner stainless steel tuces are designed to nold tne Soral in place. Thus the mechanical strengtn t, g.
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of the Bor31 is not used in any way in this design. Since the strength of the stainless steel is expected to deteriorate by only a minute, insignificant amount over the life of the racks, there should De no mechanical failures which would diminish the Boral density.
The only other effect ahich could possibly diminish the Soral density in the spent fuel pool is raciation. However, the boron' carbide in tne Boral is very resistant to the types of radiation in the spent fuel pool, where the neutron flux is very low. The neutron flux is the product of the subcritical neutron multiplication and whatever neutron source is present in the pool. For a Keff=0.95 the subcritical neutron multiplication is a f actor of 20; so the neutron flux in the pool will be twenty times whatever source is present. Since even the strongest, non-reactor sources emit only 0
about 10 neutrons per second, the neutron flux in the pool will be many orders of magnitude celow the level where baron depletion effects would be significant in forty years of full time use.
The gamma flux in a spent fuel pool is also mucn less than in an cperating reactor. And boron caroide, even in a reactor, is relatively unaffected oy eitner gamma or Deta radiation.
Thus, exoerierce has snown that tne radiation in the spent fuel pool will not diminish tne Boral censi ty.
rca all of the above, we concluce that ine areal density of Doron in the Boral plates will not be reduced celow 0.02 3... _ of Doron - 10 per square centimeter of Soral plate tnroughout the life of tne racks in the spent fuel pool.
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PRCFESSIONAL QUALIFICATIONS OF FRANK M. ALMETER I joined the Commission in October, 1974 as a Materials Engineer and I am presently a Senior Materials 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 pcwer plants with specific emphasis en corrosion and water chemistry in PWR and SWR systems.
I have been acpointed to the Electrical Power Research Institute (EPRI) Corrosion Advisory Ccmmittee and the NRC Carrosion 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 respcnsibility for the evaluation of reactor coolant chemistry in both Pressurized Water Reactors and Boiling Water Reactors.
I have provided the Division of Regulatory Standar_ds with the technical bases recuired-for the revisicn of Regula cry Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors."
' creser:ed te: imony on Steam 3enerator TRe Integrity" at the Eeaver
'! alley Uni: 1, Pilgrim 5:nion Unit 2, Jamesecr: Sta:ico Units 1 and 2, Byron /3raidwced Stations Units 1 and 2, and prairie Island cublic hearings.
I aisc assisted in tne crecara-icn of testimony on this same subject for the South Texas Project Units 1/2 and the Wa:hington 'fuclear Project Cne public hearings.
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I have a Ph.D. in metallurgy frem the University of London (1959) and a D.I.C. degree in metallurgy from the Imcerial College (Lcndon 195e).
I received a B.Sc. degree in Metallurgical Engineering frca the University of Missouri at Rolla in 1953.
From June,1973 to October,1974, I was associated with the U.S. Consumer Projects Safety Comission as a metallurgist rescoasible for tne evaluation of engineering, manufact: ring and quality control procedures within the consumer product 1duetry to insure production of non-hazardous products.
I develeced safety tests and basic engineering factors relative to the modification of product safety standards.
I4 1971 I joined the Office of r ;line 'Ja a.
Ce:artment of the Interior, as Assistant to the Chief,.v terials Division. My duties and responsi-a bill ies were the planning and directing of c0ntracts for tne develo cent, testing, and evaluaticn of materiais ;tilized in the various desaliration processes.
I cre ared centracts for the develo nent of econcmic.atarials to reduce :ne cacital and maintenanca costs of desalisatic ' clants anc increne twir eliability.
I als: conducted ins ec:icns :: evalua.e tre :--Osi:n e #:r ance of mate-ials in coe-3;ing
'3rts.
- cerfor e:
highly technical studies of the corrosion, mechanical, mysical, and fabri:ati:n crece-ties of a wide -ange of materials.
i L.J J J.
From 1968 to 141 ! was C?tief Metallurgist of corporate materials technology for the Burndy Corporation with duties and responsibilities for the technical /acministrative management of materials pertinent to process and product development. As manager of the metallurgical R & D laboratory, I was rescons-ible 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 Corocration, 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 comcosite materials, simuisted high altitude environ-mental effects on materials, fractore and surface damage in metals, alloy d eve i c p..:e n t, heat treating, fe rous and nonferrous alloys, ceramic /
dielectric materials, and HERF for ning of metals.
Fm - 1955 to 1955 : was a C:nsulting ve t2 i l urg i s t i n t."e L'n i ted < i ngdcm
! s:ecialize: in :ne areas o' crecipitation-hardening, 6-dpa and tensile prc:erties of Eeryllfun Srcnzes.
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. 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 "Cesalination Operaticns," Dec.1972.
I was invited by the Electrical Pcwer Research Institute (EPRI) to be secretary to the "First U.S. - Japan Joint Syrposium en Light Water Reactors" (May 29 -
June 2, 1978).
I have authored 17 publications in my professional field.
Current Pubitcation:
"An Overview of i'ater Chemistry for Nuclear Pcwer Plant Safety by F. M. Almeter, Vol 28, po 582-583, 1978 Transactions of the American Nuclear Scciety.
I am a member of the American Scciety for Metals, AIME Metallurgical Society, and National Asscciatien of Corrosion Enginears.
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ECWARD LANTZ DIVISION OF OPERATING REACTORS U.S. NUCLEAR REGULATORY CCMMISSION PROFESSICNAL QUALIFICATIONS As an Engineering efstems Analyst in the Plant Systems Branch I an responsible for technical reviews and evaluations of comoonent and system designs and operating characteristics of licensed nuclear cower reactors.
I have a Bachelor of Science degree in Engineering Physics from the Case Institute of Technology and a Masters of Science degree in Physics fron Union College and a total of 28 years of professional experience, with over 20 years in tne nuclear field. My experience includes work on reactor transients and safeguards analysis, nuclear reactor analysis and cesign, research and development on nuclear reactor and reactor control concepts and inv::t g=Hnnt n' +acir operational and safety aspects.
i I have held my present position w;tn :ne Commission since December 1975.
My previous position, anich I held for about two and one half years, was Project Manager in :ne Gas Cooled Reactors Brancn, Division of Reactor Licensing, U. S. Nuclear Regulatory Commission, where I was responsible for the technical review, analysis, and evaluation of the nuclear safety aspects of acclicaticns for construction and operation of nuclear power clants.
For about ten years prior to that I was Head of the Nuclear Reactor Section in NASA. My section was responsible for tne develooment anc verification of nuclear reactor analysis comouter programs, conceptual design engineering, and develoorent engineering contracting.
Prior to my employment witn NASA, I was a nuclear engineer at the Knolls Atomic F wer Laboratory 'or aoout six years, wnere I worked s n tne safe ards ano nuclear cesign of tne 53G reactors and the initial develoc: int of the nuclear design of the 55G reactors.
Previous excertence incluces system engineering and electrical engineering with tre General Electric Comoany and electrcric tevelopment engineering aith ne Victoreen Instrurer.:
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