ML20037B027
| ML20037B027 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/05/1963 |
| From: | Wade I COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8009030713 | |
| Download: ML20037B027 (26) | |
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August 5, 1963 f
Str. Robert Lowenstein, Director Division of Licensing and legulation U. S. Atomic Energy Commission Washington 25, D. C.
D e a r.' f r. Lowenstein:
Pursuant to Paragraph 3.a.(4) of License DPR-2, as amended ("DPR-2"), Commonwealth Edison Company herewith requests that Appendix "A" of DPR-2 be amended to allow loading of an improved fuel design (Type III) in the Dresden Nuclear Power Station reactor.
The amendments' proposed are as follows:
Amendment No. 1 Amend item "2.
.';uclear Core" cf section "R.
DESIO:
FEATilRES" of Appendix "A" to DPR-2 to read in its entirety:
"2.
Nuclear Core "staximum Core Diameter (circumscribed circle) 129 in.
\\taximum active fuel length - cold 112 in.
tiaximum number of fuel assemblies by types Type I 352 Type II 107 Type III 200 Type PF-8 through PF-12 (one each) 5 Staximum total number of fuel assemblies 488 "The various fuel assemblies may be located in any position of the reactor, provided overall core symmetry is preserved and provided that fuel assemblies Type PF-8 through 12 are each separated from any Ne other such assembly by at least four Type I, Type II, or Type III fuel assenblies.
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n Mr. Robert Lowenstein August 5, 1963 "The reactor may be operated at any power up to and including rated. power with any number of the various types of fuel assemblies installed, pro-vided the maximum number and location are within b
the limits specified above."
Amendment No. 2 Amend the second paragraph of item "3.
Fuel" of section "B.
DESIGN FE ATURE9" of Appendix "A" to M -2 to read in its entirety:
"3.
Fuel "The minimum fuel pellet density averaged over a fuel segment is 941 of t!icoretical for all fuel assemblies except PF-8 and PF-9 which have fuel densities 90% of theoretical."
Amendment No. 3 Amend item "3.
Determination of Maximum Reactor Power" of section "D.
P'niiR OPIRATION" or Anpendix "A" to DPR-2 to read in its entirety:
"3.
Determination of 'laximum Reactor Power "The rated power of the reactor shall be limited to a maximum steady state value of 700 'lW(t).
"The naximum allowable steady stage heat flux limits expressed in units of Bru/(hr)(f t-) shall never ex-ceed the following values:
(~FuelType I 320,000 Fuel Type II 410,000 g.w ]L Fuel Type III
-330,000 3
Fuel Type PF-8 and PF-9 470,000 3
Fuel Type PF-10 through PF-12 510,00p_]
"The reactor shall be operated within the above limits such that a minimum burnout ratio of at least 2.0, evaluated at 125 per cent of rated power, will be maintained in each type of fuel closest to burnout in the hottest channel in the core based on a uniform stean qualitv over the cross section of the channel.
This burnout ratio shall be based l
upon the correlation in Edison's ' Recommended Curves of Burnout Limit for Design and Operation of Boiling Water Reactors', dated.ianuary 5, 1962.
The reactor shall be operated always w 4,withintheboundsof
m 3
Sir. Robert Lowenstein August 5, 1963
" stability, as evidenced by the operation itself and any experimental data produced."
Amendment No. 4 Amend Appendix "A" to DPR-2 by deleting Table II (revised 12/31/61) attached thereto and substituting Table II (revised 6/15/63) attached herewith.
In accordance with Paragraph 3.a.(4) of DPR-2, a Description and Hazards Evaluation Report in support of the proposed amendments to Appendix "A" is attached hereto as Exhibit I.
In our opinion, the data in this document in-dicates that operation of Type III fuel in.the Dresden reactor will.not involve hazards greater than or materially different from those considered by the Commission in authorizing License DPR-2, as amended, nor will use of Type III fuel constitute a material alteration to the facility.
Very truly yours,
- COM'iONWEALTli EDISON CO.*iPANY l
// /l V y'}./ I"l##h By I.
L. riace Administrative Engineer Submitted and sworn to before me this Jf 4' day of August, 1963 by said I. L. Wade.
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Notary Public Attachments:
Table II Exhibit I l
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I EXHIBIT I DRESIEN NUCLEAR POWER STATION DESCRIFfION AND HAZARDS. EVALUATION REPORT OF TYPE III FUEL s
CONIENTS SECTION I Description of Proposed Aneni::ents to Appendix "A" to DPR-2 SECTION II Rysical Characteristics and Mechanical Desisn of Fuel SECTION III Nuclear Characteristics of Fuel and Core 1
SECTION IV Thermal and Hydraulic Characteristics SECTION V Safety Evaluation This report provides technical infor=ation in support of the attached application for amendment of Dmsden Operating License DPR-2, as amended.
It is not intended that the material contained herein constitutes
" Technical Specifications" in the sense of the Licensing regulations (10 CFB, Part 50, Section 50 36).
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SECTION I DESw..IPTION OF PROIOSED AMENINENTS ib AEPENDII "A" TO DPR-2 These four amendments are proposed primarily for the purpose of obtaining authority to utilize Type III fuel assemblies in the. Dresden reactor.
Current plans are to load up to 200 Type III fue2. assemblies in the Dresden reactor during tne refueling at the end of the present operating cycle. The physical, thermohydraulic, and nuclear properties of the fuel and safety implications involved in the use of Type III fuel in a mixed core conciating of Type I, Type II, Type III, and five PF. assemblies are covered in Sections II throu6h V of this report.
The secondary purpose of the proposed a:andments is to provide a correct I
description of fuel loaded in the Dresden core by eliminating reference to development fuel assemblies FF-1 through PF-7, which were removed during the last refueling. Since this change does not affect the safety of Dresden operation, no further evaluation of this change is submitted herewith.
?
1
-SECTION II PHYbICAL CHARACTERISTICS AND MECHANICAL DESIGN.0F DRESDEN FUEL i
Three principal types of fuel assemblies have been designed for the. Dresden reactor. In addition, twelve slightly different developnental fuel assem-blies have been designed and operated in tha Dresden reactor. Seven of the developmental fuel assemblies, assemblies FF-1 through PF-7, were removed from the reactor during the refueling shutdown of November-December, 1962. After the next refueling, the reconstituted Dresden reactor vill contain a maximum of five developmental fuel asserblies, plus Type I, Type II, and Type III fuel. - The basic design features of the various fuel types are tabulated in Table II (revised 6/15/63) as part of the proposed amend-ments to License DPR-2.. Description of the details of Type I, Type II, 7
and PF asse=blies which have been previously submitted vill be referenced.
Only Type III fuel vill be described in this report.
A.
Type I The first Dresden core fuel consists of Zirealcy-clad fuel rods in a "6x6" = atrix. Details of the fuel asse=bly are given in Table II and Reference 1*.
B.
Type II The TyIe II fuel consists of 304 stainless steel clad fuel rods in a "7x7" matrix. Details of the design are given in Table II and Reference 2++.
C.
Tyre III The Type III fuel asse=blies consist of six rows of six Zircaloy-clad non-segmented fuel rods except, as described below, one rod is seg=ented "a7 requ M W or positioning spacers. The basic constituent of Type III fuel is a sintered solid cylindrical pellet of about 9h% theoretical density.
The pellet size is 6iven in Table II. The pellets are enclosed by a Zircaloy jacket forming a fuel rod about nine feet long. The fuel rods are neld in position at top and bottom by stainless steel tie plates.
Between the plates, rod position is maintained by five wire-type spacers, located axially along the length of the fuel assembly. These spacers serve to minimize deflection and vibration of the fuel rods.
The five spacers are hel1 in longitudinal position by the segmented spacer support fuel rod. J.500 p yas;.,er._;::1111on,Er2 3 are added to the uniformly 0
enriched fuel as a burnable poison to increase the exposure potential within the capabilities of the control system.
"m mas in each fuel assembly hgX2 thickened clad and reduced diameter fuel pellets to reduce local pover_gg.
GEAP-10hh, " Preliminary Hazards Summary Report for the Dresden Nuclear Power Station", by G. Sege, May 1,1957 Letter to R. *.ovenstein;. Director, Division of Licensing and Regulation, IJ.S. AEC, from I.L. Wade, Administrative Engineer, Co=monwealth Edison Co=pany, dated January 5,1962.
The water-to-ruel ratio is 2 3:1. The fuel enrichment is set for a design average fuel exposure of approximately 13,500 MWD /T.
Details of the Type III design are given in Table II and Figure 1.
D.
Development Fuel Assemblies Type FF-8 through PF-12 Design features of Type FF-8 through FF-12 fuel are given in Table II atd Reference 2**.
Fuel assemblies FF-8 through 12 are currently operating in the Dresden reactor and their performance does not significantly affect the nuclear or ther:nal performance of the entire core.
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c SECTION III NUCLEAR CHARACTERISTICS OF THE FUEL AND Cole The detailed nuclear characteristics of the various types of fuel assemblies have been deter:nined. This data was used to study the secttered fuel loading which is of interest for this document. Two typical scattered core loadin6s am shown in Figures 2 and 3 The results of these studies are given in the following paragraphs.
A.
Fuel Assembly Characteristics t
The basic physics lattice data for the Type I, Type II and Type III fuel assemblies is su=cari::ed in Table 1.
The data was developed by computing the nuclear characteristics of individual fuel rods in each type of fuel separately. Ther=al group nuclear constants are averaged over the thermal neutron spectra calculated by the *,{ilkins ecuation. Details of the ther nl neutron behavior in and around individual rods are computed using the Pisphericalhar=onics approx 1:ation to the neutron transport equation. Individual fuel rods and surrounding =oderator are then ho=ogenized for the cell calculations. Epither=al group transport properties are obtained from spectrum-veighted, multigroup cross section filee using the M*JFT-I7 calculational model. Resonance absorption and
~
X fission para =eters are derived ~fMii~~ a'sufed resonance integrals with 1/v or other appropriate co=ponents added to acco==cdate nuclear events in the energy range i==ediately above ther:al. Fission in the higher energy group is treated by a fast fission factor multiplier of the diffusion theory source. The resonance absorption for U-238 is calcu-lated from Hellstmnd's =easured resonance integral data. To account for fuel rod resonance interactions EDIEdfWthe~sffeit's of water
.Iaps, a Dancoff interaction correction is applied to the individual fuel rods through the U-23d resonance integral.
The analysis described in the pmceding paragraph provides three neutron group nuclear constants for the individual fuel rods which comprise a fuel assembly. Nuclear constants for the re=aining =aterials, i.e.,
vater in the water gaps, clad and channels, are obtained using similar techniques. These data, together with a geometric mpresentation of the fuel assembly, are used to obtain assembly average lattice data. Two-di=ensional, three group diffusion theory is used to compute the assembly average lattice data. All =aterials in the calculation are treated as diffusing regions with the exception of control blades. For this case, a logarithmic derivative boundary condition, simulating a near black neutron boundary, is applied to the ther=al neutron group and an absorp-tion cross section is used to acco==odate epither=al absorption in the control =aterial.
B.
Core Characteristics 1.
Pcver Distribution Two scattend fuel loadings representative cf what =ay be used and typical core power dietributions, computed for these loading arrange-ments, are indicated in Figums 2 and 3 A three neutron energy
.h.
4 group, two-di=ensional diffusion theory method was used to obtain the data. An actual Dresden control blade configuration was used in the computation.
To simulate the effects of boiling and three dicensions, the distri-butions shown in Figures 2 and 3 vere synthesized from distributions computed independently for the top, center and bottom sections of the reactor. A typical operating Dresden control blade configumtion, the axial power distribution of which had been measured by vire irradiation, was selected. Void concentrations were computed for the core sections using this experimental axial data. Fuel bundle nuclear characteristies repmsentative of the depletion and moderator conditions vero entered in the computer calculations directly for the mdial power solutions. The axial coupling or synthesis was accomplished using the experimental axial data. The power distri-bution in a Type III fuel assembly is shown in Figure 4 2.
FF Assemblies Data for developmental assemblies FF-8 through FF-12 was included in the request for amendments to Appendix A of License DPR-2, dated January 5,1962. The FF assemblies have been treated as depleted Type II fuel in the calculation of void and temperature moderator mactivity coefficients for the next fuel cycle. The similarity of lattice characteristics and the scattered distribution of these five assemblies provide assurance that the =oderator coefficients vill not be significantly affected.
3 Temperature and void coefficients Temperature and void coefficients teceme lass negative with exposure in the Dresden reactor. Thus, the limiting moderator coefficients occur at the end of a fuel cycle. Moderator coefficients mpresent-ing mixtures of Type I, II and III fuel at the end of the third cycle are listed in Tables 2 and 3 The mixed lattice calculational model used to determine-these coe'fficients successfully predicted the mixed lattice temperature coefficient of the Dresden reactor at the beginning of the second fuel cycle.
Af ter refueling is completed and before the reactor is brought to operating power levels, the temperature coefficient vill be measund.
It vill be verified by experimental measurements that the temperature coefficient is in accordance with the license require =ents.
4.
Reactivity Control The fuel assembly lattice data presented in Table 1 indicates that the material buckling decreases and the control strength increases as the tempemture and void fraction increases. The effective reactivity levels of these fuel assemblies decrease in a uniform manner with exposure.
It han bee M M n*ed +. hat the burnable _
poiscn, erbium ox1de (erbia)uin _the. Type III fuel vill not deplete in reactivi,ty_.vorth faster than tha effective _ fuel. depletion.
Thus, the cold ahutdown condition at the beginning-of-cycle *m es j
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the control systes ruore severely than any other condition. If the control system satisfies cold shutdown argin requirements, adequate reactivity control is available for any other operating condition.
Before loadin6 is consnenced, sufficient calculations vill have been perfomed to provide the basic inforation required to demonstrate compliance with shutdown margin requirements.
Detailed calculations performed on the maximum Type III leading con-laval ta_
figuraticn shown in Figure 3 indicate that the zwactiy4+v less than that of the initiaLI)resden core. The difference in reac519Ey~is greater tha5 the reductiYo4i control strength indicated f
by Table 1.
This assures an adequate shutdown margin for the proposed configuration. Any other scattered loading configuration containing a unifom mixture of Types I, II and III fuel, would be less reactive than a loading of the :::aximum number of Type III fuel assemblies, i
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T SECTION IV THERMAL AND HYDRAULIC CHARACTERISTICS The thermal and hydraulic characteristics of a mixed core consisting of Type I, Type II, and Type III fuel have been investigated. These investi-gations included analysis of the Dresden core at 125% of rated power con-ditions. The power distributions for typical single and double batch loadings of Type III fuel shown in Figures 2 and 3 are representative of the expected distribution at the end of the fuel cycle.
Type III fuel has slightly i= proved ther:nal capability relative to Type I fuel. The use of thickened clad and small diameter pellets in corner rods f
has reduced comer rod peaking and the use of through rods has eliminated the local power pesking at fuel rod segment connectors which must be con-sidered in Type I fuel.
Tables k and 5 present results of two of the thermal and hydraulic calcu-lations. The cases presented are for single and double batch loading with m imum primary steam flow. Maximum primary steam flow represents the
=ost severe condition from the standpoint of =inimum burnout ratio since increased prinary steam flow results in greater average void fraction in the reactor as well as higher exit void fraction and quality.
The PF assemblies produce a negligible change in the peak heat flux in the Type I, II, and III fuels. Othezeise the PF asse:blies have no effect on operation of the reactor.
Peak analytical results have been compared with license limits for peak heat flux and minimum burnout ratio for each type of fuel.
It is concluded that all fuel can be operated within the limits specified in the. license.
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SECTION V-SAFETY EVALUATION The potential hazards of operating the Dresden Nuclear Power Station have been re-evaluated to deter =ine the effect of Type III fuel assemblies on reactor safety.. Due to the similarity between Type I and Type III fuel, the re-evaluation has led to the conclusion that use of Type III fuel has a negligible effect on the safety or performance of the Dresden reactor.
. Evaluation of the effect of Type III fuel on potential hazards of operating the Dresden reactor included an erehtion of the mechanisms by which power could be rapidly increased and by which removal of heat could be adversely affected in the reactor. Specific cases presented in this safety evaluation include :
(1) additions of reactivity by withdrawal of contrcl rods or addition of fuel, and (2) loss of coolant flow.
The severe conditions assu=ed for analyses of the above cases are expected to result in closest approach to the thermal limits of the fuel. These conditions are believed to have an extre=ely s=all probability of occurrence. Even if the assumed conditions were to occur, the analyses show that there would be no fuel cladding failures.
In addition to considering the effect of additions of reactivity and loss of pumps, this safety evaluation also includes discussion of the effects of:
(1)
Type III fuel on system stability, (2) Type III fuel design on fuel cladding failures, and (3) Type III fuel on the consequences of the -av4 mum credible accident which was analyzed for the initial core loading of Type I fuel.
A.
Additions of Reactivity On a co=parative basis, it can be igvn that,_the transients arizing from additions of reactivity are Isss severe with the use of Type III fuel, _
interspersed with Type Tarid' TI fuel, thaE were analyzed for-th~e original core loading.
This conclusion is based upon the lattice data discussed in Section III, above and set forth in Table 1, which shows that the control rod worths are considerably smaller for Type III fuel than for Type I fuel.
It follows, therefom, that the to,talmac.tiv_ity_chega ed_ rate of.
rnac_tivity addition resulting_from_vithdrawing a_.controLrod_from a. core consiiEng,_of. mixtures-of_ Type I,..II and III fueLaust.msalt.in_a maxi-num contml_ rod. vorth less than the mari-mm control. rod _vorth_for a core composed of only Type I fuel. Therefore, the approach to thermal limits of the fuel during contml md runout or the step insertion of reactivity, vill be less severe than was considered in the safety analysis of the original com loading (see GEAP-LCM, p.131).
Regarding the step insertion of reactivity, the original analysis involved computer studies of step inputs of reactivity of 0.4% to 0.6% 4 k.
These results showed that even if a scram should not occur, the transient would settle out very quickly (about two seconds) at a slightly higher power with no serious system effects and that even much lar6er step changes in reactivity would not produce serious results.
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The mechanisms which limit the severity of the transient caused by additions of reactivity are (1) the prompt increase in neutrori absorption in the U-238 by Doppler broadening and (ii) the effects of void formation.
These mechanisms are applicable equally to Types I, II and III fuels.
The Doppler broadening effect of the Type III fuel is about that of the Type I because of their similarity of design. TheJ1se of erb 1_a,_as a_,,
burnable poison in the Type III fuel makes the void mactivity response En~e~gative Edh either-TypWTor Tise 11.
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~N An accident in mfueling involving the maximum reactivity change has been examined. Again, solely for purposes of considering the worst f
possible condition, it was assumed that during the loading procedure the highest worth fuel position was left vacant in the double batched core. In addition it is postulated that control rods adjacent' to the vacant fuel position have been withdrawn to bring the reactor critical. The withdrawal is =ade in such a =anner as to assure the s=allest possible uncontrolled region, since the reactivity worth of a single assembly increases as the size of the just critical core decreases. Further assumptions =ade in the analysis of this hypothetical accident are:
1.
The source level instrument flux counter fails, or the operator fails to observe increase in count rate 'gon withdrawal of the rods.
2.
A new Type III fuel asse bly is inserted in the vacant fuel position at the =aximum design rate of the hoist (~ 12 inch /sec. ).
3 Ihe period scram circuitry fails.
4
"'he =oderator tecperature is 68'F.
The accident was simulated by assuming the.t a.C27dk bundle reactivity addition is represented by the maximum incertion rate of 0.CC8dk/sec.
The resulting me.:timum center fuel te=perature is co=puted* to be about 2600*F. Thus, center melting in the hottest fuel rod will not occur, and no cladding failures are expected. The reac*o r is automatically shut down by the high neutron flux scras circuitry with some localized boiling occurring.
Mechanical and procedural zeasures (presented in the original accident analysis of Dmsden, see GEAP-3076, p. 66) are utilized to prevent this type of accident.
B.
Ioss of Coolant Flow Of the r.everal hypothetical accidents which adversely affect the removal of heat from the core, the =ost severe in terms of the themchydraulic transient is the si=ultaneous loss of power to all recirculating pumps.
- The Doppler coefficient used in the calculations was based on the data of Hellstrand. This coefficient is consistent with the SPERT experi=ents of 1961. i i
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l The analysis for this accident is based on the investigation of the steady state thermal-hydraulic performance of the core at the vorst possible end condition. This condition assumes that the reactor is at full power and that the total recirculation flow is determined by the water density distribution in the total primary coolant loop prior to the pump trip.
This analysis is considend to be conservative for the following reasons:
1.
The power of the reactor is not assumed to decrease with the associated increase in moderator void fraction caused by the deemased flow.
2.
The transient reduction in flow due to the kinetic energy of the coolant is neglected. This has the effect of instantaneously dropping the I
flow rate to the natural circulation flow component of the total flow existing at the tice of the pump trip.
3 The additional flow which results from the accumulation of voids in the risers is neglected. The steady state natural circulation flow rate is higher than the value used in this analysis.
Without considering these Wree factors, the total core flow rate after 0
the pump trip is 10.8 x 10 lb/hr ec= pared with 2k.8 x 10 lb/hrbefore, giving the msults shown below; Type of Fuel Mini =um Burnout Ratio Before Pump After Pump Trip Trip I
2.81 1.44 II 2.62 1 37 III 3 19 2.29 It is noted that burnout does not occur even under the conservative assumptions on which the analysis was =ade for the complete loss of flow accident.
C.
System Stability The operation of the Dresden Nuclear Power Station has adequately demon-strated the ability of the power plant to respond to a large variety of operating conditions in a stable =anner.
It has been experimentally proved that the Dresden reactor can be operated stably with void content in the core in excess of 30%*. Startup, rod oscillating tests, pmssure regulator tests and system msponse tests all indicated that a large margin from any* stability exists at all operating conditions. Subsequent operation indicates that Type II fuel has no noticeable effect on stability of the Dresden reactor.
- Report on the High Void Test, dated October 31, 1961, submitted with letter from I. L.. Wade to R. Lovenstein, on November 16, 1961.
Since Type III fuel is very similar to Type I fuel in its thermal and hydraulic characteristics, the use of Type III fuel vill have no effect on the stability of the Dresden reactor.
.D.
. Fuel Cladding Failure t about 4000 fuel Evaluation of the initial Dresden core indicated tp% of the noble gas element segments could be leaking at a rate of 10' activity per second without exceeding the maximum permissible stack emission mte.
(See GEAP-lC44, p.141 and GEAP-3076, p. 27) i This analysis is equally applicable to Type III fuel except that since f
the fuel rods are not segmented,1000 Type III fuel rods vould result in approximately the same release as 4000 Type I fuel segments. These same facts were pointed out in connection with licensing of Type II fuel, which is also not segmented.
Although a single failure can result in an increase in gas release, the decrease in the number of welds increases the reliability adequately to compensate for the decrease in number of seg=ents per fuel rod.
E.
Maximus Credible Accident As pmviously dese.-ibed, Type III fuel is very similar to Type I fuel in its basic design para =eters.
The pri=ary factors in the maximum credible accident analysis, that is rated power (and hence core fission product inventory) and total stomd energy within the reactor vessel (pressure and temperature) are not being changed. It is concluded, therefore, that use of Type III fuel does not appreciably affect the consequences of the maximum credible accident.
11
2, TABM 1 ASSIMBLY AVERAGE IATfICE DATA GREEN TYPE I, TYPE II AND TYPE III FUEL (Including the effects of end connectors and vertical leakage)
)
Fuel Type I
I I
II II II III III III Teraperature (*F) 68 546 546 68 546 546 68 546 546 Void Fraction O
O O.20 0
0 0.20 0
0 0.20 k., Uncontrolled 1.132 1.151 1.145 1.13;
1.139 1.131 1.164 1.160 1.150 g
k.. Contmiled 0 946 0.884 0.852 0 986 0.925 0.899 0.900 0 904 0.873
' l Controlworth(Ak/k) p.164 0.232 0.256
_O._131 0.189 0.205 ; o.159 ' O.221 0.241 7
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39.8 63 5 75.4 35 0 55 5 66.7 38.8 61.9 74.1 g
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n TAB 3 2 MOERATOR F2 ACTIVITY CEFFICIEYrS FOR I
FULL COPE MIXZD LATTICE - SIN 0IE BATCED - END OF CYCG 4
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Temperature Coefficient l
Temperature Coefficient Temp. (*F)
(ak/k/*F )
68
+1.4 x 10-5 315 0
1 546
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Void Coefficient 4
Void Fraction Temp.
Void Coefficient (Interior to Flow Channel)
(*F)
( Ak/k/f Void) 0 68
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]i 20 546
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.n TABE 3 MODERATOR REACTIVITY COEFFICIENTS FULL COBE MIIED LA'ITICE - DOUBE BATCED - END OF CYCLE Temperature Coefficient p
Temperr.ture Coefficient Temp. (*F)
(dk/k/*F) 68
+3 1 x 10-5 378 o
-5 5L6
-6 7 x 10 Void Coefficient Void Fraction Temp.
Void Coefficient (Interior to Flow Channel)
(*F)
(Ak/k/% Void) o 70
- 2 x 10-5 10 70
- 2.4 x 10~
o 546
-2 91 x 10-4 10 546
- 4.43 x lo-k 20 546
- 6.o x 10-4 14 4
TABG 4 SINGE BATCH END OF FUEb CYCIE A. General Plant Data lated Power Overpcuer
- 1. Reactor Thermal Power (Wth) 700 875 6
6
- 2. Primary Steam Flow Rate (lbs/hr) 1.575xlo 2.345 xlo 6
6
- 3. Secondary Steam Flow Rate (1bs/hr) 1.350x10 1,350xyg
- 4. Feedwater Enthalpy (Bt lb) 394 394
)
- 5. Core Inlet Enthalpy (Bt lb) 489 486 B. Core Description
- 1. Number of Fuel Assemblies 464 464
- 2. Fraction of Power Generated in Fuel O.97 0 97 2
- 3. Total IIeat Transfer Area (ft) 22,53o 22,530 2
4
- 4. Average Heat Flux (Btu /hr-ft) 106, 132,0006 i
- 5. Total Core Flow Rate (1bs/hr) 25 0x1 25.Ox1g
- 6. Leakage Flow Rate (lbs/hr) 2.Ox10 2.0xlO
- 7. Average Void Fraction in Channels 0.20 0.24 C. Fuel Description
- 1. Type of Fuel I
II III PF-8&9 PF-lokl1 FF-12
- 2. Number of Assemblies 256 107 96 2
2 1
- 3. Rod Diameter (in.)
0 567 0.441 0.555 0.489 0.412 0.412
- 4. Active Fuel length (in.)
106.5 110 5 109 0 111 56 105 34 98 78
- 5. Number of Rods per Assembly 36 49 36 36 64 64
- 6. Heat Transfer Area per 2
Assembly (ft )
47.43 52.09 47.46 39.87 60.60 56.76 D. Results for Peak Channel et Overpower 2
- 1. leak Heat Flux (Btu /hr-ft) 400,000 395,000 :92,000 4360,000 4360,000 4360,000
-2. Minimum Burnout Ratio 2.30 2.20 2.g
>2.0
>2.0
>2.0
~
TABIE 5 DOUBE BATCH END OF FLEL CYCE A. Gene ml Plant Data Rated Power overpower
- 1. Reactor Thermal Ibver (wth) 700 875 6
d
- 2. Primary steam Flow (1bs hr 1.575xlo 2.345do6
- 3. secondary steam Flow lbs 1.350xlo6 1.350x10
- 4. Feedwater Enthalpy Bt 394 394
- 5. Co m Inlet Enthalpy Bt 489 486 B. Com Description
- 1. Number of Fuel Assemblies 464 464
- 2. Fraction of Power Genemted in Fuel o.97
- o.97 2
- 3. Total Heat Transfer Area (ft) 22,532 22,532 2
- 4. Ave nge Heat Flux (Btu hr-ft ) 106,000
- 132,
- 5. Total Core Flow Rate (lbs
)
.25.ox10 25 011 i
f
- 6. Leakage Flow Rate (lbs r) 2.ox106 2.ox1
- 7. Average Void Fraction in Channels
.27 C.' Fuel Description
- 1. Type of Fuel I
II III FF-8&9 FF-lokll FF-12
- 2. Number of Assemblies 160 107 192 2
2 1
- 3. Rod Diameter (in.)
0.567 0.441 0 555 0.489 0.412 0.412
- 4. Active Fuel length (in.)
.106.5 110.5 109 0 111 56 105.34 98.78
- 5. Number of Bods per Assembly 36 49 36 36 64 64
- 6. Heat Transfer Ama per 2
Asceably (ft )
47 43 52.o9 47.46 39.87 60.60 56.76 D. Results for Peak thnnel at overpower 1.1%ak Heat Flux (Btu /hr-ft) 396,000 390,000 400,000 e360,000 4360,000
.c 360,0co 2
2.-Minimum Burnout Ratio 2.30 2.20 2.40
> 2.0
> 2. 0
> 2.0
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[ Control rod position (vater)
Number = Fuel rod power average fuel :od power l
FIGURE 4 Relative Power Distribution in a Type III Fuel Assembly I
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F ROM:
D4 7f er Def'tsurnit DATE PTCrit f D NO.:
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CONCURRENCE O
DALE AN$wlif RED:
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