ML19257D354
| ML19257D354 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/1979 |
| From: | Ju J, Kapuschinsky D, Lozito E VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML19257D353 | List: |
| References | |
| VEP-FRD-34, NUDOCS 8002040190 | |
| Download: ML19257D354 (45) | |
Text
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'!E P- TRD t NORTH .-uiNA UNIT l, CYCLE 1 CCRE PERFORMANCE REPORT 3Y J. R. JU D. M. KAPUSCHINSKY Approved:
[ / ~~ '
Nuclear Fuel Operatica Group E. J. Lozita, Director Fuel Resour:es Department Nuclear Fuel Operati:n Group virginia Electric 5 Power Ccepany Richmond, 'li r g in ia December, 1979 1868 172 8002040 19 gy o
TABLE OF CONTENTS I Section List of Tables . . .
Page No.
ii List of Figures . . . . . iii 1 Introduction and Summary . 1 2 Surnup Follow . . 7 3 Reactivity Depletion Follow . . ... 12 4 Power Distribution Follow 14 5 Primary Coolant Activity Follow . . . 35 6 Conclusions . . . . 39 7 References . . . . ;g Acknowledgements I
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1868 173 ,
LIST OF TA3LES
!able I 4.1 Title Su:=ary of Incore Flux Maps for Routine Operation p,zg I
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g 1868 174
I I LIST OF FIGl*RES I Figure Title Page '!c.
1.1 Core Loading . 4 1.2 Movable Detector and Thermocouple Locations . . . 5 1.3 Control Rod Locations . 6 2.1 Core Burnup History 3 2.2 Monthly Average Load Factors 9 2.3 Assemblywise Accumulated Burnup: Comparison of I 2.4 Measured with Fredirted Satch Burnup Sharing . . .
. 10 11 3.1 Critical Soron Concentration versus Burnup - HF?-ARO 13 4.1 Assemblywise Power Distribution - N1-1-25 . . 20 4.2 Assemblyvise Power Distribution - N1-1-49 . 21 4.3 Assemblywise Power Distribution - N1-1-73 22 4.4 Hot Channti Factor Normalized operating Envelope . . 23 4.5 Heat Flux Hot Channel Factor, F (2) - N1-1-25 . 24 4.6 Heat Flux Hot Channel Factor, F'(Z) - ';i-1-49 25 Q
I 4.7
?
Heat Flux Hot Channel Factor, F'(2) - N1-1-73 Q
. 26 4.3 Maximum Heat Flux Hot Channel Factor versus Burnup 27 I 6.9 N
Rod Sow Penalty on F;h,
. . . . 28 10 Enthalpy Rise Hot Channel Factor versus Burnup 29 a.11 Target Delta Flux versus Burnup 30
.12 Core Average Axial Power Distribution - N1-1-25 31 4.13 Core Average Axial Power Distribution ';1-1-49 . 32 4.14 Core Average Axial ?cwer Distribution - N1-1-73 33 4.15 Core Average Axial Peaking Factor versus 3crnup 34 5.1 Dose Equivalent I-131 Concentration versus ice- 37 I 1.2 :-131/:-133 Ratio versus Time ...
1868 175 35 n
v Section !
INTRODUCTION AND SUSDd.ARY On September 25, 1979 after more than eighteen months of Operation, North Anna Uni: 1 completed Cycle 1. Since the initial criticality of Cycle 1 6
on April 5, 1973, the reactor core produced approximately 94 x 10 MBTU (15,892 Megawatt days per metric ton of contained uranium) which has resulted in the 0 9 generation of approximately 8.7 x 10' kwhr gross (S.: x 10 kwhr ne:) of electri-cal energy. North Anna 1, Cycle 1 reached the end of full power reactivity at a core burnup of approximately 15,150 MWD /MTU at which point power operation was continued through a power coastdown. The unit was operated in the power coas:down mode and at reduced power levels achieving ar additional 742 MWD /MTU
=
burnup prior to shutting down for refueling. The purpose of this report is to present an analysis of the core performance for routine operation during Cycle 1. The physics tests that were performed during the startup of this cycle were covered in the North Anna 1, Cycle 1 Startup Physics Tes Report and, therefore, will not be included here.
I The first cycle core consisted of three fresh batches of fuel. The North Anna 1, Cycle 1 core loading map specifying the fuel batch identifica-tion, fuel as sembly locations, burnable poison locations and sour:e assembly locations is shown in Figure 1.1. Movable detector locations and :hermoccuple
__ lo:stions are identified in Figure 1.2. Control red locations are shown in Figure . 3.
Rou:ine core f o '. l o w involves the analysis of four prin:ipal perfor-man:e indica:crs. These are burnup dis:ribution, reactirity deple:i:n, power a
distribution, and primary coolant acti<ity. The core burnup dis:ribu:icn ;s followed to verif. both burnup s7 e:rv and proper ba::h burnup sharing, thereby, L868 176 b
I E
ensuring that the fuel held over for the next cycle will be cccpatible with the new fuel that is inserted. Reactivity depletion is monitored to detect the existence of any abnormal reactivity behavior, to determine if the . ore is depleting as designed, and to indicate at what burnup level refueling will be required. Core power distribution follow includes the monitoring of nuclear hot channel f actors to verify that they are within the Technical Specifica-I tions' limits thereby ensuring ths adequate margins :o linear power density and critical heat flux thermal limits are maintained. Lastly, as part of I normal core follow, the primary coolant activity is monitored to verify that the dose equivalent iodine-131 concentration is within the limits specified by the North Anna Unit 1 Technical _pecifications, and to assess the integrity of the fuel.
Each of the four perfor ice indicators is discussed in detail for the North Anna 1, Cycle 1 core in the body of this report. The results are summarized below:
- 1. Burnup Follow - The burnup tilt (deviation from quadrant syn-metry) on the core was no greater than r0.8% with the burnup accumulation in each batch deviating frca design prediction by less than 2%
- 2. Reactivitv Depletion Follow - The critical boron concentration, used to monitor reactivity depletion, was consisten:13 within 20.6% ;K/s of the design prediction which is :ee;l within :he :1% ;K/K rar;in allcwed by See:ica 3.1 1.1 of the Technical 3pecifications.
- 3. Power Distribution Folicw - inc ore flu:< =a p s :2 ken each conth indicated that the assemblyw se radial power distributi:ns deviated from the design predictions by an average difference of approximatel 1.27% All ho:
I channel fac: ors met their respec:i/e Technical Specifica: ions .ici:s.
I 1868 177 I
I I 4 Primary Coolant Activity Follow - The dose equivalent iodine-131 activity level in the primary coolant at the end of Cycle 1 was approximately 2.53 x 10~ >
Ci/gn. This corresponds to less than 37; of the operating limit for the concentration of radiciodine in the primary coolant .
In addition, the effects of fuel densification were monitored through-out the cycle. No densification effects were cbserved.
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I 1868 178 d,
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.O R~'d AT.:A 1 - CYCLE 1 ---t"--
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FIGl.'RE 1. 2 NORTu 12:A L7:IT l-CYC' E 1 MOVABLE DETECTOR A';D THERMOC01'DLE LOCATIONS I --
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1868 180
SORT:i A.CA 1 - CYCLE 1
'a CO:iTROL tOD LOCATIONS R P N n L g J M C F E D C 3 A
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Ag-In-Cd Functier ';u.ber cf Clusters Contr:1 Sank 3 3 Centrol Bank C 3 Centre; 3ank 3 3 Centrol Sank A 5 Shutdem cank 53 3 sam:g m :enk s a 1868 181 5 nr: . ens:n 2_ .. 3 SP (Spare Rod Locations) 5
Section 2 y BURNUP FOLLOW The burnup history for the North Anna Unit 1, Cycle 1 core is graph-ically depicted in Figure 2.1.
~
The North Anna 1, Cycle 1 core achieved a burnup of 15,892 MWD /MTU. As shown in Figure 2.2, the average load factor for Cycle I was 76% when referenced to rated thermal power (2775 MW(t)).
_ Radial (X-Y) burnup distribution maps show how the core burnup is shcred among the various fuel assemblies, and thereby allow a detailed burnup 3
distribution analysis. The NEWTOTE computer code is used to calculate these y assemblywise burnups. Figure 2.3 is a radial burnup distribution cap in which the asse=blywise burnup accumulation of the core at the end of Cycle 1 opera-tion is given. For comparison purposes, the design values are also given.
As can be seen from this figure, the acc==ulated assembly burnups were gener-ally within :2.52 of the predicted values. In addition, deviation from quad-i rant symmetry in the core, as indicated by the burnup tilt factors, was ~1ess than :0.5% The deviation frca quadrant symmetry was slightly higher than expected. It was a direct result of a slight quadrant power tilt that devel-oped approxi=ately two thirds through the cycle. The power tilt will be addressed further in Section 4 The burnup sharing on a batch basis is conitored to verify that the
~
core is operating as designed and to enable a:: urate end-of-cycle batch burnup predictions to be =ade for use in relcad fuel design studies. As seen in Figure 2.4, the batch burnup sharing for North Anna Unit 1, Cycle 2 followed
- design predictions very closely with each batch deviating less than 2% fr m
, design; this is considered excellent agreement. The good agreecent between actual and predicted assemblywise burnups and batch burnup sharing indicate that the Cycle 1 core did deplete essentially as designed.
1868 182
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I 1868 184 3
._ NORT'd AN';A U:'IT 1 - CYCLE 1 FIGURE 2.3 ASSE'13L'GISE ACC'.'MULATED SUR';UP :
~'
COMPARISON OF MEASURED %IT'i PREDICTED 3
(10 MWD /MTU)
E F N .at L r. J n C F I D C 3 A 9.26 11.63 9.18 l l l l l f
, d.90 11.33 8.90 1
-4 . 0 I I
-1.1 1.'
9.49 13.67 15.60 15.21 15.57 13.53 9.51 9.26 13.43 15.46 15.17 15.46 13.43 9.06 2
-2.4 +1.3 -.7 . : -0.3 a.? -1.1 -2.' , ',
10.31 15.39 16.57 16.99 18.05 17.C1 16.72 15.63 10.45 10.18 15.23 16.39 3 17.03 18.25 17.CS 16.39 15.23 10.13
+1.3 -i * +1.1 -0.5 + ,._
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Section 3 REACTIV! t DEPLETION FOLLOW The primary coolant critical boron concentration is monitored for the purposes of following core reactivity and to identify any anomalous re-
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activity behavior. The FOLLOW computer code was used to normalize " actual" critical boron concentration measurements to design conditions taking into consideration control rod position, xenon and samarium concentrations, modera-tot temperature, and power level. The normalized critical boron concentration
._ versus burnup curve for the North Anna 1, Cycle 1 core is shown in Figure 3.1.
T. t can be seen that the measured data compare to within 60 ppa of the design prediction. This corresponds to less than :0.6% 2K/K, which is well within the 21% ;K/K criteria for reactivity ancmalies set forth in Section 3.1.1.1 of the Technical 3pecifications. In conclusion, the trend indicated by the critical boron concentration verifies that the Cycle 1 core depleted as expected without any reactivity anocalies. 1868 187
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Section 4 POWER DISTRIBUTION FOLLOW Analysis of core power distribution data on a routine basis is neces-sary to verify that the ho: channel factors are within the Technical Specifica-tions limits and to ensure that :he reactor is operating without any abnormal conditions which could cause an " uneven" burnup distribution. Three-dimensional core power distributions are determined from movable detector flux map measure-5 ments using the INCORE computer program. A summary of all monthly flux maps taken since the completion of startup physics testing for North Anna 1, Cycle 1 is given in Table 4.1. Power distribution maps were generally taken at monthly intervals with additional maps taken as needed. Radial (X-Y) core power distributions for a representative series of incore flux maps are given in Figures 4.1 thruegh 4.3. Figure 4.1 shows a power distribution map that was taken early in cycle life. Figure 4.2 shcws a power distribution map that was taken near mid-cycle burnup. Figure 4.3 shows a map that was taken late in Cycle 1 life. Most of the radial power distributions were taken under equilibrium operating conditions with the unit operating at approximately full power. In each case, the measured relative assembly powers were generally within 2.2% of the predicted values with an average percent difference of less than 1.3'; which is considered good agreement. With the core burnup at cpproximatel:. 10,300 MWD /MTU, the quadrant power tilt ratio was determined to be 1.3% at full power. This represented a significant increase with respect to earlier values. The Technical Specifica-tions limit for the quadrant power til: ratio i s 2 *; Po:ential mechanisms 1868 189 14
I for causing the tilt indication including measurement inadequacies, fuel varia-tion, system perturbations, and reactivity control component anonalies were reviewed. The results of this review suggested that several rod cluster control assembly (RCCA) rodlets may have dropped into the core. However, based on an inspection that was performed during the refueling shutdown it was determined that the RCCA rodlets were intact. At this time, the mechanism causing the tilt indication has not been identified. An important aspect of core power distribution follow is the moni-toring of nuclear hot channel factors. Verification that these factors are within Technical Specifications limits ensures that linear power density and critical heat flux limits will not be violated, thereby providing adequate I thermal margins and maintaining fuel cladding integrity. The initial Cycle 1 Technical Specifications limit on the axially dependent heat flux hot channel I factor, FT (Z), was 2.05 x K(Z), where K(Z) is the hot channel factor norma-lized operating envelope. On May 19, 1978 with the core at a burnup of approxi-mately 601 MWD /MTU, the Technical Specifications limit for F T(Z) was changed to 2.21 x K(Z) by taking advantage of available margin to the 2200 F LOCA limit on peak clad temperature.6 Figure 4.4 is a plot of the K(Z) curve asse-ciated with the 2.21 Fq (Z) limit. This curve is representative of the K(Z) curves used throughout Cycle 1 since K(Z) changes only slightly with changes in the F (Z) limit. The axially dependent heat flux hot channel factors, F (Z), for a representative set of flux caps are given in Figures 4.5 through 4.7. T Throughout Cycle 1, the measured values of F (Z) were within the Technical q Specifications limit. A summary of the maximum values of all heat flux hot channel factors measured during Cycle 1 is given in Figure 4.S. As can be seen from this figure, there was approximately 3% margin to the limit at the beginning of the cycle, with the margin increasing substantially throughout cycle operation. 1868 190 15
The value of the enthalpy rise hot channel factor, g, 'tich is F}N . the ratio of the integral of the power along the rod with the highest inte-grated power to that of the average rod, is also routinely followed. The Technical Specifications limit for this parameter is set such that the critical heat flux (DN3) limit will not be violated. Additionally, theF{I g limit ensures that the value of :his parameter used in the LOCA-ECCS analysis is not exceeded during normal operation. The Cycle 1 Technical Specifica: ions limit on the enthalpy rise hot channel factor was set at 1.55 x (1+0.2(1-?)) x (1-R3?(3U)) where R3?(SU) is the thi=ble cell rod bcw penalty and ? is percent thermal power. The R3?(3U) values specified in the Technical Specifications are given in Figure 4.9. Figure 4.10 snows that all measured values forF}Ig, ere within the Techni:al Specifications limits during Cycle 1 The Technical Specifications require that target delta flux
- values be determined periodically. The target delta flux is the delta flux which would occur at conditions of full power, all rods out, and equilibrium xenon.
Therefore, the delta flux is measured with the core at or near these conditions and the target delta flux is established at this =easured point. Since the target delta flux varies as a function of burnup, :he target value is updated con:hly. Operational delta flux limits are : hen established about this target value. 3y maintaining the value of delta flux relatively constant, adverse axial power shapes due to xenon redistribution are avoided. The plot of delta flux versus burnup, given in Figure 4.11, shows :he value of this parameter
- Delta Flux = ?:-?b x 100 where ?: = power in top of core 'Mw( t))
2775 ?b = power in 5c :02 of :cr
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I to have bect approximately -S% at the beginning of Cycle 1. By the middle of the cycle, the value of del:a flux had shifted to approximately -3%, and then shifted to approxi=ately -1.5% by the end of Cycle ! This power shift can also be observed in the corresponding core average axial power distribution for a representative series of =aps given in Figures 4.12 thrcugh 4.14. In Map N1-1-25 (Figure 4.12) taken at approximately 300 'GD/>CU, :he axial power distribution had a cosine shape with a peak toward the bottom of the core and a peaking factor of 1.39. In Map N1-1-49 (Figure 4.13) taken at approximately 7,260 WD/>EU, the axial power distribution had flattened considerably and was peaked slightly toward the bottom of the core with an axial peaking factor of 1.17. Finally, in Map N1-1-73 (Figure 4.14) taken at approximately 14,577 WD /MTU , the axial power distribution was peaked very slightly toward the bottom of the core with a peaking factor of 1.16. The history of Y during I the cycle can be seen = ore clearly in the plot of i versus burnup given in Figure 4.15. In conclusion, the North Anna 1, Cycle 1 : ore perfor=ed satisfac-torily with power distribution analyses 7erifying that design predictions were accurate and that the values of the hot channel factors were within the limits of the Techni:al Specifications. 1 I I l l 1, I h 1868 192 1
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J l FIGURE 4.2 l UOP.Til ANNA UNIT l-CYCLE 1 ASSEMBLYWISE POWER DISTRIEUTION N1-1-49 A P N M L ( J H 0 F E O C S A
. PR E D IC T E D . . 0.55 . 0.69 . 0.55 . . PREDIC7ED .
ME ASUR E O -
--- - 0 . 5 7 .- 0. 72-.-0. 51 . .---M E A SUA E 0 .- t-
. PCT O!FFERENCE.- . 3.9 . 3.9 . 2.1 . . PCT DIFFERENCE.
. 0. 55 - . -0. 8 3 - .- 0 . 9 6 - .-0. 9 3 - . - 0. 4 .-0 .41, - 0. 5 4 .
. 0.59 . 0.85 . 0.97 . 0.94 . 0.96 . 0.34 . 0.55 . 2
. 0.7 . 1.7 . 1.0 . 0. 9 . 0.4 . 0.2 . 0.2 .
I . 0.64
. 0 64 0.96 . 1.04 , 1.06 . 1.14 . 1.06 . 1.04 . 0.96 . 0.64 .
0.96 . 1.04 . 1.06 . 1.13 . 1.06 . 1.05 . 0.97 . 0.64 . 3
-. -0.1 -.W . 2 - . - 0 3 -. - 0.1 - r-0. 3,--0.1 .- l . 0 r- 1. 0 -0. 3 . -
. 0.64 . 3.37 . 1.07 . 1.10 . 1.18 . 1.14 . 1.18 . 1.10 . 1.07 . 0.57 . 0 64 .
.-0. 6 3 --.- 0. 4 7-.- 1. 0 7 -,-1.11 - 1.16 .- 1.14-.-4.18,-1.12,--l,0 3 ,-4. 8 6, -0. 64 . .
. -1.3 . -0.4 . 0.6 . 1.1 . 0.4 . 0. 3 . -0 .1 . 1.9 . 1.2 . 0.8 . -0.3 .
- 0. 5 &- 0. 9 6 -.- l .0 7 . - l.11 . -l . 20 -M .16-A .2 0-.-l.16-M .2 4,-1 11 -l .0 7 e-0.9 Fv-0. 5 F.
. 0.56 . 0.93 . 1.06 . 1.11 . 1.21 . 1.18 . 1.22 . 1.13 . 1.22 . 1.12 . 1.07 . 0.96 . 0.5s . $
l . - 3.1 . -3 .1 . -0.8 . 0.5 . 10 . 1.5 . 1.5 . 1.4 . 1.9 . 1.2 . 0.7 . -0.3 . -0 . 3 .
. 0. 5 3 . 1.04 . 1.10 . 1.20 . 1.17 . 1.23 . 1.18 . 1.23 . 1.17 . 1 20 . 1.10 . 1 04 . 0.83 .
. 0 83 . 1 03 . 1.11 . 1.22 . 1.19 . 1.26 . 1.20 . 1.26 . 1 20 . 1.20 . 1.10 . 1.03 . 0.23 . 6
. - 0. 5 . -0 . 5 - . - 0. 4 - . - 2 . 0 - . -- 2 0 . 2.0 . 2. 0 . 2.0 . 2.1 . 0.2 .-0.1 .--0.8 .--0.3 .
I ..........................................................................................................
. 0.55 . 0.96 . 1.06 . 1 1S . 1.16 . 1.23 . 1.16 . 1.22 . 1.18 . 1.23 . 1 16 . 1.13 . 1.06 . 0.96 . 0.55 .
0.56 . 0. 9 6 - .- l . 0 6-.- l .17 . 1.15 . 1.24
. 1 21 r t . 24- .- l . 21 . - l .2 6 .-l .16 .-l .16 . 1 04 v-0. 94 .-0,5 6 . F-
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l . 0.63 . 0.9 3 . 1.14 . 1.13 .
. -1.9 . 01.
1.1$ . 1.18 . 1.24 . 1.21 . 1.23 . 1.19 . 1.23 . 1.12 . 1 12 . 0.94 . 0.71 . 0.0 . -0.4 . -1. 4 . 0.3 . 2.0 . 2.1 . 1.1 . 1.1 . -0.3 . -1.4 . -1.4 . 0.4 . 2.5 . 8
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. 0.56 . 0.94 . 1.04 . 1.'6 . 1.15 . 1.21 . 1 16 . 1.22 . 1.19 . 1.24 . 1.16 . 1.16 . 1.04 . 0.96 - 0.55 . 9
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. 0.53 1.04 . 1.10 . 1.20 . 1.17 . 1.23 . 1.18 . 1.23 . 1.;/ . 1.20 . 1.10 . 1.04 . 0.93 .
- . 0.82 .-1.02-. 1 08 -. - l.1 S - r 1 15 . 1. 21 - c i .17 - . - 1. 2 2 .~ 1. 7 .- 1 19 .- l .12,- 1 0 2 .-C. 4 - . 14-
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. O.' . 3.3 . 1.6 . -1.1 . 0.5 .
1 ............................................................................................
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. 0.59 . 0.97 . 1.07 . 1.09 . 1.19 . 1.14 . 1.18 . 1.16 . 1.20 . 1 13 . 1.09 . 3.98 . 0.58 . 11
. 0.7 . 0.7 . 0.7 . -1.7 . -1 4 . -1.S . -1.9 . -1.9 . 0.1 . 1.6 . 2.4 . 1.5 . -0.0 .
. 0.64 . 0.87 . 1.07 . 1.10 . 1 15 . 1.14 . 1.18 . 1.10 . 1 07 . 0.57 . 0.64 .
. 0.66 . 0.93 . 1.10 . 1.09 . 1.15 . 1.11 . 1.15 . 1.10 . 1.18 . 0.90 . 0.66 . 12
? .1 -.--3.1-. - 3.1 -. -1. 0 -- -2 . 2 . -2 3 - -2 . 0 .-0. 2 - 1 0 . 2.5 . 2.7 . _
. 0.64 . 0.96 . 1.04 . 1.06 . 1 14 . 1.06 . 1 .v. . 0.96 . 0.64 .
0.66-.- 0. 9 9 - .- l . 0 3--r 1. 04-r 1 11 - r l .0C A .0 2 r O . 44 .-0.6 6- . 11-
. 3.1 . 3 1 . -1.0 . -2.1 . -2.3 . -2 . 2 . - 1. 6 . 1.7 . 2.9 .
- 4. 58-. - O. & 3 + 0. 9 b.-C. 9 3-. -O. 9 b.-C .8 3-.-O. 54 .
. 0.59 . 0.54 . 0.95 . 0.92 . 0.94 . 0.32 . 0.53 . 14
. 1.3 . 13 . -0.9 . -1 5 . -2 . 2 . - 1 6 . - 0.9 .
..-........n , - ..,,,. m ......,......, , .,,...... ......., w ,,..
. ST ANDARD . . 0.55 . 0.69 . 0.55 . . AVERA~E .
. CEVIATION . . 0 56 . 0.69 . 0.54 . . PCT O!FF ERENCE. 15
. =0.015 . . 1. 3 . -0 . 5 . -2 . 4 . . = 13 .
MAP 50. N1-1 '.9 DATE 12/18/78 POWER '2689 CONTROL ROD POSITIONS N INCORE TILT F BANK C AT 22S STEPS ah. =1.337* AT G6-IB NW - 1.004 BANK D AT 224 STEPS F =l*60* AT G6-IB Q NE - 1.005 BANK P/L AI 228 STEPS F., i,
=1.163 SW - 0.993 A.0.= .,.3 /s SE - 0.998
- Includes uneartainties
]}{} }}h 21
NORTH A'?!A UNIT l-CYCLE 1 "# ICUP~r **3 m ASSEMBLWISE POWER DISTRIBUTION N1-1-73 A P N M L K J H G F E O O 6 A
. PaED::TED . . 0.59 . 0.72 . 0.5* . . PaE01:Tia .
_ . MEASAE0 . . 0.61 0 75 . 0.31 . .
- E A 5 v R E ;., . 1
. PCT 01Afiti.;E. . 6. 2 . .2 . 3.6 . . POT CIFFE4ENCE.
. 0 .62 . 0.66 . 1.01 . 0.95 . 1.01 . 0.66 . 0.62 .
. 0.s* . 0.58 . 1.02 . 0.96 . 1.02 . 0.6v . 0.6 . 2
. 2.7 . 2 .1 . 11 . 1.1 . 1.= . 3.3 . 3.2 .
__ . 0.66 . 1 .03 . 1. 09 . 1.06 . 1.13 . 1.06 . 1.09 . 1.03 . 0. 6 S . T . 0.69 . 1.33 . 1.09 . 1.00 . 1.13 . 1.07 . 1.12 . 1.06 . 0.70 . 3
. 0.8 . 0.6 . 3.3 . -0.i . ~) . . . 3. 7 . 2.? . 3.1 . 3.* .
. 0.63 . 0.91 . 1 .11 . 1. 09 . 1.16 . 1.10 . 1.16 . 1.09 . 1.11 . 0.01 . 0.6B .
. 0.63 . 0.91 . . 12 . 1.10 . 1.17 . 1.10 . 1.16 . 1.12 . 1 1* . 3.13 . G.70 . 6
. 1 . 0.0 . 0.6 . 0.9 . 0.3 . 0.2 . 0.0 . 2.5 . 2.3 . 2.2 . 3.5 .
. 3.62 . . 03 . 1.11 . 1.07 . 1.17 . 1.10 . 1.16 . 1.10 . 1.17 . 1.09 . 1.11 . 1.03 . 0.62 .
. 0.61 . 1.00 . 1.10 . 1 .03 . 1.16 . 1.11 . 1.17 . 1.12 . 1.20 . 1.12 . 1 13 . 1.06 . 0.t6 . 5
.-26 . -2.4 . -1.5 . - 1. 0 . -0.4 . 0. 7 . 0. 7 . 1.5 . 2.S . 2.2 . 1. 5 . 3.6 . 62 .
. O.de . .0) . 1.S9 . 1 17 . 1.13 . 1.16 . 1.10 . 1.lo . 1,.0 . 1.17 ..09 . . 09 . 3.36 .
. 0.46 . 1.CS . 1.0a . 1.15 . 1.10 . 1.17 . '.11 . 1.18 . 1 12 . 1.16 . 1.10 . 1.10 . 0.37 . 6
. 05 . -0.5 . -3.7 . - 1.1 . 3.2 3.9 . 0.9 . 1.3 . 2.0 . 1.2 . 0.5 . 1.3 . 2.8 .
. 0.5* . 1.01 . 1.06 . 1.16 . 1.10 . 1. 16 . 1.10 . 1.16 . 1.10 . 1.16 . 1.10 . 1.16 . 1.06 . 1.01 . 0.59 .
. 0.63 . 1.00 . 1.05 . 1.1* . 1.O s . 1.14 . 1.10 . 1.17 . 1.11 . 1.15 . 1.11 . 1.14 . 1.05 . 1.00 . 0.5S . 7
. 1. 5 . -0.5 . -0.4 . -2.2 . - 3. 9 . -2.3 . 0.5 . 0.7 . 1.1 . 1.5 . 0.6 . -1.5 . -1.2 . -1.1 . -0.7 .
. 0.72 . 0.95 . 1.13 . . 10 . . 13 . 1. 10 . 1.16 . 1.13 . 1.10 . 1.'O . 1.16 . 1.13 . 1.13 . 0.95 . 0.72 .
. 0.70 . 0.v5 . 1.12 . 1.03 . 1.1 . 1.09 . 1.17 . 1.11 . 1.16 . 1.10 . 1.15 . 1.05 . 1.12 . 0.96 . 0.73 . 6
_ . -2. 7 . -0.7 . -0.S . -1.5 . - 2. = . -1.2 . 0.9 . 0.9 . 0.0 . -0.0 . -1.1 . -1.5 . - 1. 3 . 0.5 . 1.7 .
. 0.59 . 1.01 . 1.06 . 1.la . 1.13 . i.16 . 1.10 . 1.16 . 1 10 . 1.13 . 1.10 . 1.16 . 1.06 . 1.31 . 0.59 .
. 0.57 0.vv . 1.33 . 1.1 . 1 .3s . 1. 13 . 1.00 . 1.14 . 1.09 . 1.10 . 1.10 . 1.17 . 1.04 . 1.02 . 0.60 . 9
_ . -2.0 . -23 . -2.3 . -2.2 . -2.0 . -2.6 . -3.7 . -1. 6 . -0.= . -0.S . 0.1 . 0.3 . -0.0 . 0.7 . 2. 0 . 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......................................................................
. 0.56 . 1.09 . 1.09 . 1.17 . 1.13 . 1.16 . 1.10 . 1.16 . 1.10 . 1.17 . 1.05 . 1.09 . 0.66 .
. 0.85 . 1.06 . i.04 . 1.17 . 1. )* . 1.12 . 1.37 . 1.14 . 1 03 . 1.17 . 1.11 . 1.09 . 0.33 . 13 3 . -2 0 . -2.0 . -0.4 . 0.5 . -1.0 . -3.0 . -2.2 . -1.7 . -3.5 . 0.0 . 2.0 . 0.3 . 1.- .
. 0.s2 . 1.03 . 1.11 . 1.; v . 1.17 . 1.10 . 1.1o . 1.10 . 1.17 . 1.09 . 1.11 . 1.03 . 3.s2 .
. 0.61 . 1.33 . 1.11 . 1.13 . 1.1- . 1.36 . 1.12 . 1.07 . 1 16 . 1.11 . 1.14 . 1.05 . 0.63 . 11
. 01 . 3.1 . 0.3 . 0.6 . -2.. . -3., . -3.5 . -2.7 . -0.1 . 1.5 . 2. 7 . 2.0 . 1.1 .
. 0.65 . 0.91 . 1.11 . 1. s9 . 1.le . 1.10 . 1.16 . 1.0* . 1.11 . 0.01 . 0.68 .
. 0.09 . 0.93 . 1 12 . 1. 03 . 1.12 . 1.G6 . 1.16 . '.S9 . 1.12 . 0.9* . L .7 0 . 12 3
. 2.1 . 1. . 0.6 . -1.- . -3.* . -3.4 . -2.3 . M.1 . 1.0 . 2.9 . 3.3 .
. 0.36 . 1 .33 . 1 09 . 1.36 . 1.13 . 1.36 . 1.01 . 1.03 . G . c,8 .
. 0.69 . 1.0 . 1. 07 . 1.02 . 1.10 . 1.0 . . 02 . .0.. L. 73 . 13
. 1.3 . 1.1 . -1.0 . -3.* . -2.7 . -1.6 . -3.6 . 1.7 . 3.3 .
0 .32 . 0.66 . 1.01 . 0.91 . 1.01 . 0.36 . 0.62 .
. 0 .33 . 0.43 . 1.03 . J.i . 0.99 . 0.46 . 0.62 . 14
. 1.1 . 1.5 . -0.6 . -1.0 . -1.6 . -3.6 . -G .1 .
. S TANO& A 3 . . 0.59 . 3.72 . 0.59 . . ave 9&LE .
. Osv1.T10N . . 0. c 3 . 0.72 . 0.53 . .P;T CIF5EAENCE. 15
. 0.013 . . 19 . 3.6 . -1.2 . . = 1.5 .
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FIGURE 4.5 NORTH A'UTA UNIT 1-CYCLE 1 HEAT FLUX HOT CHANNEL FACTOR, F1(Z) s N1-1-25
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FIGURE 4.6 NORTH ANNA UNIT l-CYCLE I T HEAT FLUX HOT CHANNEL FACTOR. F (Z) Q N1-1-49 w l i . I ! fl I ! l ! i i ! l' l l ! { l 4 j i ; i i 3 2.5: i
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FIGURE 4.12 _ NORTH ANNA UNIT l-CYCLE 1
. CORE AVERAGE AXIAL P0b'ER DISTRIBUTION N1-1-25 1
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; EOTTOM TOP AXIAL FOSITION (NCDES) w 31 1868 206
FIGURE 4.13 NORTH ANNA UNIT 1-CYCLE 1 g CORE AVERAGE AXI AL POWER DISTRIBUTION . e N1-1-49
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m 9 ~ Section 5 PRIMARY COOLANT ACTIVITY FOLLOW Activity levels of iodine-131 and 133 in the primary coolant are important in core performance follow analysis because they are used as indica-tors of defective fuel. Additionally, they are also important with respect to the offsite dose calculation values associated with accident analyses. Both I-131 and I-133 can leak into the primary coolant system through a breach in the cladding. As indicated in Section 3.4.8 of the North Anna Technical Specifications, the dose equivalent I-131 concentration in the primary coolant was limited to 1.0 pCi/gm for normal steady state operation. Figure 5.1 shows the dose equivalent I-131 activity level history for the North Anna 1, Cycle 1 core. Reactor ecolant system activity data indicated that two discrete fuel defect events occurred within one week at approximately 2500 MWD /MTU burnup (end of July, 1978). The failure mechanism is unknown at this time; however, it is believed that the defects were not a result of a design deficiency or core operation tactics. A visual (binocular) inspection of the fuel, which was performed during the refueling shutdown, did not reveal the identity of the defected fuel. The data on Figure 5.1 shows that the core operated substantially below the 1.0 pCi/gm limit during steady state operation (the spike data is _ associated with power transients and/or shutdowns), and that the equilibrium activity levels tended to decrease following the initial defect events. The average equilibrium dose e.quivalent I-131 concentration during Cycle 1 was 3.9 x 10" pCi/gm which is less than 4% of the Technical Specifications limit. 35 1868 210
- The ratio of I-131 to I-133 is used to characterize the type of fuel failure which may have occurred in the reactor core. Use of the ratio for this determination is feasible because I-133 has a short half-life (approxi-
_ mately 24 hours) compared to that of I-131 (approximately eight days) so that _ for pinhole defects where the diffusion time through the defect is on the order of days, the I-133 decays out leaving I-131 dominant in activity, thereby caus-ing the ratio to be 0.5 or more. In the case of large leaks, uranium particles in the coolant, and/or " tramp" uranium *, where the diffusion mechanism is
] negligible, the I-131/I-133 ratio will generally be less than 0.1. As shown in Figure 5.3, the I-131/I-133 ratio data points associated with equilibrium operation following the defect events are greater than 0.5 indicating pinhole defects.
*" Tramp" uranium consists of small particles of uranium which adhere to the outside of the fuel during the manufacturing process.
36 1868 211
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() I-AFR MAY TJ2. ~M AUG SEP CCT NCV 0E0 JAN ?I3 M.AR A?R Y _ .~JL SE? 1973 1979 I x
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*?
Section 6 CONCLUSIONS The North Anna 1 core has completed Cycle 1 operation. Throughout the cycle, all core performance indicators compared favorably with the design predictions and all core related Technical Specifications limits were met with significant margin during full power operation. No abnormalities in reactivity or batch burnup accumulation were detected. However, the indicated increase in quadrant power tilt is somewhat anomalous. The analysis of radiciodine data for Cycle 1 indicates that there are pinhole leaks in the fuel cladding. _ However, based on the coolant activity level and the inability to observe any fuel defects during the refueling shuffle, it is concluded that the fuel defect level was low. E w m E
.g 1868 214
_ 39
l I l
! REFERENCES I 1) T. K. Ross and J. H. Leberstien, " North Anna Unit 1, Cycle 1 Startup Physics Test Report," VEP-FRD-31, September, 1978.
- 2) North Anna Power Station Unit 1 Technical Specifications.
- 3) T. K. Ross, "NEWTOTE Code," NFO-CCR-6, August, 1978.
- 4) R. D. Klatt, W. D. Leggett, III, and L. D. Eisenhart, " FOLLOW Code," WCAP-7482, February, 1970.
l 5) W. D. Leggett, III and L. D. Eisenhart, "INCORE Code," WCAP-7149, December, 1967. i 6) Letter frem O. D. Parr (NRC) to W. L. Proffitt (Vepco), dated May 19, 1978 (Docket No. 50-338). I 1 i
'I 1
I I I 5I 1868 215
.I w
I I I ACKNOWLEDGEMENTS I The authors would like to acknowledge the cooperation of the staff at North Anna Power Station in supplying the basic data for this report. Special thanks are due Messrs. J. P. Smith, R. R. Etling, and A. K. White. Special thanks is also duc to Ms. C. E. Bullock for her patience and accurate typing of this report. E 1868,216}}