ML20002D157

Forwards Request for Proposed Change 10 to Tech Specs of License DPR-6 to Insert Reload C Fuel in Reactor

ML20002D157
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/29/1966
From:	Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Boyd R, Doan R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101190567
Download: ML20002D157 (19)

Text

{{#Wiki_filter:,/' COPY CONSUMERS POWER COMPANY General Offices - Jackson, Michigan July 29, 1966 i ' V1F t@_C.105 Dr. R. L. Doan, Director Re: Docket 50-155 Division of Reactor Licensing United States Atomic Energy Commission Washington, D. C. 20545

Dear Dr. Doan:

Attention: Mr. Roger S. Boyd Transmitted herewith are three (3) executed and nineteen (19) conformed copies of a request for a change to the Technical Speci-fications of License DPR-6, Docket No. 50-155, issued to Consumers Power Company on May 1,1964, for the Big Rock Point Nuclear Plant. This proposed change (No. 10) will enable Consumers Power Company to insert Reload "C" fuel into the Big Rock Point reactor. This fuel will incorporate vibratory compacted UO2 powder but is otherwise physically identical to the previous reload fuel. Shutdown of the plant for refueling is scheduled pre-sently for around September 5, 1966. Yours very truly, Robert L. Haueter (Sigrad) RLH/wf/mp Robert L. Haueter Attach. Assistant Electric Production Superintendent - Nuclear to S g If eo vp 2c -3: DO$ETED 7 L %g.gtp,g q g SAEC j f g. y v/ l 2 AU 1956

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ff en;,,, CONSUMERS POWER COMP r, uum Docket No. 50-1$$ m m^ CM tm,6 s -< b; b. Request for Change to the Technical Specifications License No. DPR-6 For the reasons hereinafter set forth, it is requested' that the Technical Specifications of License DPR-6 issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant, be changed as follows: I. Section 5 A. In Section 515, change "(a)" to read as follows: "(a) Enrichment of Fuel, approximate weight percent U-235 from 2.6 to 5 2, inclusive." Under "(c) Fuel Bundles," change "UO Density, Percent 2 Theoretical" for the reload fuel to read as follows: " Pellet Fuel - 94 1 1 Pcvder Fuel - Approx 85" L. Add a new section - 5 1.8: "5 1.8 Thin Clad Powder Fuel Bundle Two of the Reload "C" fuel bundles may contain standard rods with Zr-2 cladding of 0.025" thickness; otherwise, they will 1)e the same as the remaining Reload "C" fuel bundles." II. Discussion - Reload "C" Fuel The proposed changes in Section 51 vill enable Consumers Power Company to refuel the Big Rock Point reactor with vibratory com-pacted UO powder fuel. Experience has shown powder fuel performance to g be at least equal to pellet fuel performe.nce, but with significant poten-tial economic gain. At the upcoming refueling, it is planned to inscr' ; to b renining Reload "B" fuel and a portion of the Reload "C" fu ~ o lTItry r 1 e t Al.c 21w s c k.. - ,3 s .~ \\ 'i Q $D q \\ c. 3

1 2 A. Fuel Description ~ The Reload "C" fuel is similar physically to the first reload fuel'(Reload "B") except that vibratory compacted UO powder will g be used in place of-the UO pellets. The basic fuel bundle design re-2 mains the same - same cage, same fuel cladding material and dimensions, same spring clip; spacers,.etc. The Reload "C" fuel bundle is shown on Figure 1 and described in Tuble 1. The powder fuel was designed to the same cladding-stress criteria as the pellet fuel. All of this' fuel.will have.a cobalt rod at each of the four corners of the fuel bundle. Included with the "C" fuel bundles are two pilot bundles with thinner cladding. They are the same physically as the "C" bundles - except for this thinner cladding and associated increased fuel diameter. This will result in a slight increase in fuel weight and a slight de-crease in water-to-fuel volume ratio. Central fuel temperature at the maximum licensed heat flux vill remain the same. The cladding thickness . has been reduced from 0.034" to 0.025". (See Table 1.) only the large rods in the bundles have been changed; the small corner rods will remain unchanged. The clad was designed as a self-supporting tube using standard design criteria but with the stress intensity limits raised to a higher fraction of the ultimate strength. Continuing study of the stress system on fuel cladding indicates that all stresses are being taken into account so that lower design margins are permissible. Use of higher stress intensity limits results in a reduction in clad thickness for prescribed operating conditions and associated savings in the cost of Zircaloy material for a fuel rod. The current stress intensity limits were established on the assumption that failure of the clad would occur if the ultimate strength of the. clad were exceeded. Recent examinations show appreciable ductility for Zr-2 cladding irradiated up to fast flux exposures of about 21 1 5 x 10 nvt greater than 1 Mev. Although greater fast flux exposures will be experienced in typical fuel clad applications, irradiation effects 21 on ductility appear to saturate at lower exposures of about 0 5 x 10 nyt greater than 1 Mev. The ductility of the clad means that clad collapse onto-the fuel, rather than rupture, will occur should the strength be 4 -n- -,,, ~ ,ne.

+ L ~ 3 I exceeded by the external compressive forces which are limiting. (Clad col- ' lapse onto the. fuel is not expected to result in clad perforation. In-Lternal pressure is maintained below limits by providing plenum space for -fission gases. i It.is expected that further reduction in design margin and' 71ad thickness will be possible. -These two pilot bundles are an inter-mediate step in conservatively evolving Italistic design bases. B. Fuel Thermal Data The thermal conductivity data and the heat transfer coef-ficient between fuel.and clad for the "C" powder fuel are the same as s previously described for. ten pilot bundles included in the Type III-f reload fuel for Cycle 4 in the Dresden Nuclear Power Station. The previous thermal data were obtained from irradiations performed under the High Per-formance UO Program sponsored by the Joint U.S.-Euratom Research and 2 . Development Program..The UO p vder conductivity data, as submitted for 2 the Presden reload pilot bundles, yield an integral KdT from the fuel surface temperature (500 C by definition) to-the melting temperature (2800 C) of h9 v/cm. (This was reported as preliminary data in the Dres-den submittal.) In the su= mary report of the High Performance Program (GEAP-5100-1,"UO Powder and Pellet Thermsl Conductivity During.Irra-2 diation,".by M. F. Lyons, et al, March 1966), the value of the integral. as used for the Dresden powder bundles was confirmed. Based on post-irradiation analysis, Lyons, et al, recommended an integral value from surface to melt (as defined above) I h9 w/cm. (See Figure 2.), The establishment of an appropriate thermal conductivity curve for powder UO is a deduction from this. The approach used to 2 4 establish the vorhing curve for the Dresden and Big Rock Point powder fuel is based on the following line of reasoning. During the 11rst few minutes of operation, the powder conductivity will be uncertain and sintering probably change rapidly due to sintering of powder above the UO2 temperature (N1600 C) and. densification. However, as these processes continue, the conductivity of the fuel above the temperature for the onset of grain growth will become very close to that determined earlier for pellets. The remaining differences in conductivity and in the conductivity integral values from surface to melting between powder and pellets are . attributable to the poorer conductivity of the unsintered powder rim l -~

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operating below the grain growth temperatures., Working backwards from the previously established pellet UO e nductivity curve, 'the powder 2

conductivity was assumed to'be identical, at the melting point-temperature, to that for pellets.. Below this temperature, the powder curve was assumed - to gradually and smoothly fall below that for pellets. The deviation be-tween the two curves was ad, justed so that, upon reaching reasonable value forLthe surface temperature (%500 ' C), the difference in area under the. two curves equals the difference in the integral values from surface to melting. Below this temperature, the curve was simply extended smoothly to provide a reasonable mean through the out-of-pile data. (See Figure 3.) The powder fuel has a higher UO to clad gap conductance 2 than pellet fuel. This tends to counteract the lowered UO nductivity. 2 In this particular instance, the two effects cancel. (See Table 2.) C. Fuel Physics Data The characteristics of the "C" fuel with annular cobalt target rods in each of the four corners of the bundle have been calculated a and compared to the "B" or pellet fuel. 1. Reactivity (k=) "B" Fuel "B" Fuel "C" Fuel Temperature-Without Cobalt With Cobalt With Cobalt 68 F, Zr Channel 1.275 1.215 1.2hh 572 F, Zr Channel 1.303 1.2h1 1.272 572 F, Zr Channel 20% Steam Void 1.296 1.231 1.256 2. Moderator Temperature Coefficient (Ak /k per F in Zr Chan-gf eff nel at TT F) "B" Fuel "B" Fuel "C" Fuel Without Cobalt With Cobalt With Cobalt -6 -6 -6 Start of Cycle +3.2 x 10 +8.3 x 10 +2.6 x 10 -5 -6 End of Cycle +5 5 x 10 ' +5 1 x 10 +h.9 x 10 3 Void Coefficient (Akg /k,77 per Unit Void Within the Channel) "B" Fuel "B" Fuel "C" Fuel Without Cobalt With Cobalt With Cobalt Cold -0.0h -0.06 -0.06 Hot -0.09 -0.09 -0.08 4 ,,_-_m

L. r. 5 h. Doppler Coefficient (Ak /k fgper-F) ff "B" Fuel Fuel: "A" Fuel With or "C" Fuel Temperature - Moderator (SS Clad) Without Cobalt With Cobalt {' 68 F 68 F, 0 Voids -1.47 x'10 -1.4 x 10 -1.2 x 10 1323 F 550 F,- 0' Voids -1.03 x 10 -1.1 x 10 -1.0 x-10 ~ -5 -5 1323 F 550 F, 20% Voids -1.15 x 10 ' -1.4 x 10 -1.1 x 10 It can be seen from the above comparisonIthat the modera-tor temperature and void coefficients for the "C" fuel are not sig-nificantly different from those for the "B" fuel. Although the "C" fuel has a slightly lower Doppler Coefficient than the "B fuel, it is very similar to the initial ("A") fuel load of stainless steel . clad fuel at all conditions except the cold, zero-power case. The enrichment of the "C" fuel has been increased to par-tially-offset the loss of reactivity due to the cobalt targets. As seen above, & reactivity lies between "B" fuel without cobalt and "B" fuel with cobalt. 5 Control Rod Worth. Control rod worth for the "C" fuel is slightly less than that for the "B" fuel containing cobalt. This is due primarily to the increased enrichment of the "C" fuel which is designed to attain a discharge exposure of 15,000 Mvd/T of uranium with cobalt targets in place for three quarters of the design life. (The cobalt will be replaced by standard UO fuel r ds of suitable enrichment to control 2 power peaking.) D. Operational Safety for Powder Fuel Factors relating to the operational safety of UO p vder fuel 2 in power reactors include those which can cause cladding failures, propagate existing failures or limit the perfomance of. the plant itself. i Mechanical and chemical interaction between UO fuel and clad, 2 if not properly controlled, can result in failure of the cladding. Propagation of a failure by such phenomena as waterlogging, gross oxidation of the UO, r internal hydriding of Zircaloy cladding, are 2 no more severe for UO p vder fuel than for UO P"11** I"*1* 2 2 d s

p. ~ ( m UO -Clad Mechanical Interaction- - 1. 2 The temperature' rise and the : thermal expansion of the lUO are much greater than those of the cladding when a UO fuel 2 2 rod is heated from ambient temperature to steady-state full power operating conditions. The' degree of mechanical contact between - the fuel and;the cladding depends upon the type of fuel (pellet or compacted powder),.the; initial fuel-cladding gap, the fuel rod ,v thermal rating, the fuel exposure and the type of cladding material. The mechanical' contact between the fuel and the cladding can vary _ from zero to a point where both diametral and. axial plastic strain - .of.the cladding ~ occur. ^ The premise that fuel-cladding mechanical interactions can be' eliminated. completely by the use of freestanding cladding with a nominal clearance between the pellets and the cladding is not valid. A significant aspect of recent measurements is that circumferential ridging occurs in all pellet-fuel rods of current designs, even though the ' cladding is freestanding wit.h diametral - gaps'of up to 0.013" between the' pellets and the cladding. ha-surements made on various types of freestanding cladding, including Zircaloy-2, stainless steel, Incoloy and Inconel, show ridging at the pellet-interfaces when operated at nominal power levels. Ridging of freestanding, Zircaloy-2-clad fuel rods operated for a short time in the Consumers' Big Rock Point reactor has been mea-sured. The ridges are visually accentuated by selective crud depo-sition on the. cladding surface. The magnitude of the ridging appears to be a function of initial cold gap and fuel rod specific power. Ridges of up to 0.001" in height have been measured in 0.h25" OD, 2 -Zircaloy-2-clad fuel rods having a vall thickness of 0.030", an as-fabricated pellet-cladding gap of 0.007", and operated at a peak l surface-heat flux of up to 350,000 Btu /hr-ft. No comparable ridging has been observed on fuel rods-containing compacted UO Powder fuel. 2 A comparison of the thermal expansion of compacted . powder fuel with the thermal expansion of small diametral gap 4 ,--w. e..w- - - - - - -..

y 3 pellet-fuel indicates that th'e'diametr'al strain from' compacted g powder fuel is about.one third.less than the strain. at midpellet and'about'h0% less th'n-the maximum' strain at the pellet inter-a faces. 2. UO2-Clad-Chemical-Interaction There. have been no confinned ' cladding failures re-lated to impurities in either pellet or powder fuel in which the . fuel material has been within the allowable impurity levels. However, clad failures have occurred in both pellet and compacted . powder fuel rods in which excessive amounts of impurities have been present. During.early' tests of Zircaloy-2-clad pellet fuel -rods inithe VBWR, presence of up to 1,000 ppm of fluoride in the UO2 pellets resulted in severe cracking of the-cladding when water, entered the fuel rod through:a cladding defect. In the PRTR' pro-gran-at Battelle Northwest Laboratories, failures in the Zircaloy-clad compacted powder fuel rods have been attributed to the pres-ence of fluoride and other. contaminants in the powder. 3. Structural changes During Irradiation Although there are differences in the physical char-acteristics of as-fabricated sintered pellets and compacted ~ powder, they tend to be eliminated or diminished in magnitude as irradia- -tion proceeds. The nigher the thermal conditions or fuel operating temperatures, the more alike the two fuel types beccme as exposure proceeds. Powder. fuel becomes essentially identical with pellet fuel when operated above about 1600* to 1800 C (the temperature range above which sintering, grain growth, void formation'and void migration begin to occur). There is experimental evidence that powder fuel irradiated at bulk temperatures below sintering tem-peratures (1600 C) undergoes a phenomenon in which the mechani-cally pressed particles are more finnly bound together. This is attributed to " fission-sintering" in which very high local tem-peratures are achieved as a result of fission spikes. As pellet fuel is ' subjected to high thermal stresses during irradiation, the pellets tend to crack-into smaller pieces in the direction of compacted powder fuel. The powder fuel which i l t. l l

~ '8L starts out' as a mechanically bonded compact of sea 11 particles tends to become a more cohesive body in the direction'of pellet fuel. ' Thus, the morphologies of the two fuel types tend to be-come more alike as irradiation proceeds. Initially, there had been-some concern that, when relatively low-density (80% to 90% TD) compacted powder fuel rods vere subjected to high-temperature irradiation,- the powder would densify into'a snaller diameter mass. However, compacted powder fuel rods behave the same as pellet fuel with respect to densi-fication during irradiation. As grain growth and void migration occur in the central fuel portion operating above about 1600* to 1800 C, the voids migrate toward the hotter central region of the fuel resulting in a central void surrounded by a densified annular region. At typical BWR fuel and coolant conditions, the outer rim of pellet or compacted powder fuel never operates at temperatures high enough for structural changes to occur other than the fission-sintering. h. Propagation of Failures In the operation of intentionally defected pellet and compacted powder BWR type rods, there do not seen to be any sig-nificant differences in the performance characteristics of the two types of fuel with respect to fuel washout, UO e lant re-2 action, or susceptibility to waterlogging. E. Background of Experience With Powder Fuel Process development and irradiation testing have been con-ducted as part " many programs to reduce fuel fabrication costs and to investigate alternate fabrication methods showing promise of lov fabrica-tion costs and potential improvement in fuel performance. Of these, the powder compaction process shows the greatest potential. The relative fabrication cost economics of compacted powder fuel has been the subject of frequent reviews. It appears to be favorable. Fuel rods made by powder l-compaction techniques were first tested by General Electric Company, APED, in the Vallecitos toiling water reactor (VBWR). They are now under irra-l diation in the Big Rock Point, Dresden and JAERI reactors. Compacted l l-powder fuel rods comprise a large portion of the plutonium recycle test I

( 9 reactor-(PRTR) core at the Battelle Northwest Laboratories. Other tests ' of this fuel concept are being conducted by Bettis, Westinghouse ' (Sexton), ORNL, Combustion Engineering and APED (EVESR superheat). Consumers Power Company has participated in the High Power Density Fuel Development Program which has as one of its primary goals the -develop.nent and irradiation testing of new and/or improved processes for fabrication of UO fuel for power reactors. Compacted powder UO fuel has 2 2 been irradiated in the VBWR and the Big Rock Point reactor as part of the HPD program. In addition, many developmental _ fuel bundles containing com-pacted UO p vder fuel are currently being irradiated in various GE BWRs. 2 Table 3 summarizes some of this experience with compacted powder fuel rods. No verified failures have occurred in any of the developmental bundles being irradiated in power reactors except for failures of FFD fuel rods in the VBWR and some GETR loop tests. The HPD program 304-ss clad powder UO f"*1 """ f"D#i "t'd 2 by various powder compaction techniques such as one, two-and three-pass svaging, hot svaging, tandem rolling and vibratory compaction. Approxi-mately 150 powder fuel rods were irradiated in the VBWR until its shutdown in December 1963. ~ Exposure achieved by the powder fuel was approximately: -*9100 Mvd/T peak for cold svaged powder fuel, 7350 Mvd/T peak for hot svaged powder fuel, 9500 Mvd/T peak for tandem rolled powder fuel and 8000 Mvd/T peak for vibratory compacted powder fuel. Early in the HPD program,-high-temperature erosion stability tests were performed in conditions simulating boiling water reactor envi-ronment and in the VBWR utilizing fuel rode with intentional defects. The out-of-pile tests resulted in no significant loss of UO fr m ld svaged 2 (s92% TD) or hot svage,. (s95% TD) specimens and variable loss (<50 mg to 1-1/2 grams in 60- to 70-gram specimens) of UO fr m tanden rolled (%88% TD) 2 and vibratory compacted samples (65% TD and 92% TD). No unusual swelling of the UO was apparent in any of the compacted powder specimens. In-reactor 2 tests in VBWR utilizing compacted powder with densities ranging from 90% TD (cold svaged) to 95% TD (hot 's'vaged) resulted in no loss of UO by erosion 2 and no UO swelling. 2 " Average exposure is lower than the peak value by a factor of about 1.55

10 s Three in-service' failures related to the stress-assisted intergranular corrosion of 304-SS occurred in compacted po' der fuel rods. v Post-irradiation examination of these fuel rods has verified the satis- . factory perfomance of defected UO p vder fuels in reactor service. A 2 thin clad (10 mil SS) cold svaged powder rod was operated in the VBWR for approximately 300 hours at 1000 psi, Sh5 F, with an 8" long longi-tudinal defect. Although the UO was co=pletely exposed to the reactor 2 coolant, no significant amount of the 93.1% TD compacted powder was washed out. Operation of a 16 mil clad rod in -the same assembly with a less severe defect resulted in no loss of the UO ' 2 A failed 86.8% TD vibratory compacted powder fuel rod had -been exposed to flowing steam and water in VBWR for at least 72 hours. Again, no significant loss of UO was observed. 2 -In all cases, there was no evidence of " waterlogging" or swelling of the UO. Dial gauge and profilameter dimensional measurements 2 confirmed these results. The good dimensional and chemical stability of the powder UO fuel, when exposed to flowing steam and water, is attributed to in-2 reactor bonding between the powder UO particles. This la due to the 2 combined effects of thermal sintering and fission sintering. Experience at Other Sites The most extensive testing and application of compacted powder fuel has been associated with the PRTR program at Battelle North-west Laboratories. Since start-up of the PRTR in July 1961, a total of 66 UO2 and 90 UO -Pu0 19-r d fuel bundles have been irradiated in the PRTR. 2 2 These, plus a small number of other experimental elements, comprise ap-s proximately h,h00 individual full-length rods. The current PRTR fuel exposure status is: Average Exposure of Type of Fuel Leading Bundle (Mvd/T) I. UO -Pu02 (Vibratory Compacted) 8300 2 2 (Vibratory Compacted) 6100 UO There have been over 35 fuel rod failures in the PRTR. All of the failures have been attributed to deficiencies in the fuel processing

Ec 11 such as: fluoride impurities in plutonium, gas phase'hydriding in sus-ceptible cladding regions caused by irradiation deco = position of residual-water in the fuel and hydrocarbon impurities, or both. No failures can be attributed to inherent-problems with the design of fuel rods utilizing UO2 p vder fuel. ' Oak Ridge has conducted irradiation _ tests of 26 vibratory compacted Th0 -UO eaPsules containing powder made by are fusing and by 2 2 the sol-gel process. The peak burn-up achieved'vas approximately 81,000 Mvd/T in a 10" long stainless-steel clad fuel specimen containing 85% TD Th0 -UO ' 2 2 Most of the ORNL irradiations were conducted in the NEX and MTR low-temperature process water with cladding temperatures of about 100 C. Therefore, these are tests of the.-fuel meats and fuel-cladding inter-actions rather than cladding-environment tests. No failures and no sig-

nificant changes in dL ension of the irradiated spectmens were found.

III. Hazards Considerations The Reload "C" fuel bundles described abcve utilize the same hardware as the Reload "B," the Phase I and the Phase II R&D fuel bundles. This hardware continues to give excellent performance in the Big Rock Point reactor. A great deal of experience with powder fuel is accumulating as discussed above. Based upon this experience, we believe that powder fuel is now a commercially acceptable fuel design and that its performance-should be as good as, if not better than, pellet fuel. The nuclear characteristics of the Reload "C" fuel are not significantly,different from previous fuels, and its performance under normal and transient conditions should be comparable. The thermal-hydraulic performance vill be identical to the Reload "B" fuel. Based upon the above considerations, we have concluded that use of Reload "C" fuel in the Big Rock Point reacter does not

12 present a significant change-in'the hazards considerations' described or implicit in.the Final Hazards Summary Report. CONSUMERS POWER COMPANY By Vice President Date: July 29,.1966 5 Sworn and subscribed to-before me this 29th day of July 1966. +- w . hwes) ~ Notary Public, Jackson County, Michigan 'My commission expires February 16, 1968 1 e

i [K es....-., , }l %EY. %il _ } TABLE 1 Reload "C" Fuel Data Sheet Fuel Rod, Cold Rod' Type (See Figure 1) 1 2 3 C Thin Clad Fuel Diameter, Inches 0 381 0 381 0.282 0 399 CladdinC Thickness, Inches 0.034 0.03h 0.031 0.031 0.025-Cladding'Outside Diameter, Inches 0.449 0.449 0 344 0 344 0.449 70 Active Fuel Length, Inches TO 70 70 Fuel Material UO2 UO2 UO2 UO2 Cladding Material Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Number of Rods per Bundle 37 T2 8 4 109* Enrichment,W/0U-235 52 - 29 29 Fuel Bundle Number of Fuel Bundles 40 Fuel Rod Array 11 x 11 Weight UO2 per Bundle, Lb 305 i Water-to-Fuel Volume Ratio 2.6 5 i 37 at 5 2 W/0 U-235 72 at 2 9 W/0 U-235 109 Rods Total

t i ....h. t, {. i TABLE 2 Thermal Performance Comparison of Pellet and Powder Fuel r')ellet

• Powder Fuel-Diameter, Inches-0 373 0 381 Cladding Thickness, Inches o.034 0.034 Cladding Diameter, Inches.

.o.h49 0.M9 FuelDensity,'% Theoretically 95 85

Nucleate Boiling Heat Transfer Coefficient, Btu /Hr-Ft - F 10,000 10,000 2

-Heat Transfer Coefficient Between Fuel and 2 Cladding, Btu /Hr-Ft - F 1,000 3,000 .UO2 Condretivity Integral-T = 2800 C KdT, W/Cm (See Figure 2) 59 49 T = 500 C Incipient Melting Temperature of UO2, F 5,080 5,060 2 HeatFluxforIncipientMelting,Bhu/Hr-Ft 550,000 550,000 Fuel Linear Heat Generation, Kw/Ft (For Incipient Melting) 19 19 4 l. I [

• Data are given for large fuel rods only as they are most limiting from fuel temperature considerations, i

L TABLE 3 GE-APED Compacted Powder Developmental Fuel Irradiation Experience Clad Clad Avg Eurn-Up of. Approximate - No. of Clad 'oD Thickness Uo Density Peak Q/A' -Leading Bundle Peak Burn-Up Reactor Rods Material (Inch) (Inch)~ 2(kTD) (Rated Power) ' Mwd /T-Mwd /T Pdmarks Tests: ^ 0.400 VBWR 150 304-ss to 0.005 to 83 to 95 UP to 527,000 6,000 9,100 3 Failed I we o.565 0.016 VBWR N 210 304-ss o.250-o.028 53 to 67 20,000 30,000 - GEI'R-PWL 18 Zr-2 o.560 0.030 . Up t 6 20,000'

30,000:

3 Failed Rods. 1.4 x lo Power Reactor Demonstration: Dresden 350 -Zr-2 0 5625 0.035 84.T 330,000 5,700 . 8,550.. JPDR 72* Zr-2 0 557 0.030 S90.0 300,000 2,500 '3,750 Consumers' '363 .Zr-2 o.425 0.030 85 340,000 6,200 9,150 Big Rock Point " 484 304-ss o.425 0.010 91 384,000 T,680' 11,200 a 726 Incoloy-o.425 -0.011 91 430,000 6,400 9,45o 800 s J W j 'oTwo segments per rod. 'g \\ k i

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.l*,.- >M : y; - n= L r .'. 5. Ossernment Print. tog Omes: 1965 -769445 9 I I FROM' NM *M r8 MCr M DATE OF DOCUMENT: DATE RECEIVED: NO.a l Jr.cl.t nr., richip r. 7 29-M P-PJA ??hl I E.Oltrt L. [hutttr LTR.: MEMO: R EPO RT: OTH ER : I 0; ORIG.: CC: OTH ER : Dr. I ichcrd L. Doan 1 (3) executed arri (19) comrom ' e { CONCURRENCE D D ATE ANSWERED: ACTION NEC.ESS ARY { NO ACTION NECESSARY O COM +4 ENT O evs t CL ASSI FIC ATION, POST OFFICE FILE CODE: Docket ?% 50-155 DESCRIPTION: (Must Bt UNCLASSW?ED) p p % *:d Ed I Ltr. trnrs. f v/all extr.a cyc...F1 A" Tr' f ENCLOSURES: f i hequr:t for Chanre (?c.10) to the Techn5 cal Speciff eationt Lic. fio. Doon $.? p%C y/ int % q. Notariycd July 25, 1966. (2P cyc. received) I g g ' } e fj!;--[j-REM ARKS: }j Distritution: 1-forrAl guypl. J1-U0 E 2-Co=plisinec l-WC 1-Stoner U. S. ATOMIC ENERGY COMMISSION Mall CONTROL FORM W""gr7,:s .jhyw w (Yi?* C ]. k[."-. i 'N ?'f. _ 'f As Ty[, ' . N,'b ik l l l -}}