TÜRKİYE ATOM ENERJİSİ KURUMU
ÇEKMECE NÜKLEER ARAŞTIRMA VE SÖİTİM MERKEZİ
Ç.N.A.E.M. A .R 308
CORE CONVERSION CALCULATIONS FOR THE TR - 2 REACTOR
PART A : NEUTRONICS CALCULATIONS
Mehmet Hulusi TU R G U T , T a n ı » TÜRKER
Nüclear Engineering Department
February - 1993
P.K. 1, Hava Alanı, İSTANBUL
B asım tarihi Eylül - 1993
TÜRKİYE ATOM ENERJİSİ KURUMU
ÇEKMECE NÜKLEER ARAŞTIRMA VE EĞİTİM MERKEZİ
'
Ç.N.A.E.M. A.RI308
CORE CONVERSION CALCULATIONS FOR THE TR - 2 REACTOR
PART A : NEUTRONICS CALCULATIONS
Mehmet Hulusi TURGUT, Tanzer TÜRKER
Nuclear Engineering Department
February - 1993
P.K. 1, Hava Alanı, İSTANBUL
Basım tarihi Eylül - 1993
Abbreviations :
ÇNAEN : Çekmece Nuclear Research and Training Center TR-2 : Turkish Reactor 2
HEU Highly Enriched Uranium (%93 U-235) LEU : Low Enriched Uranium (%19.75 U-235) 2D, 3D : 2 and 3 Dimensional
BOC : Beginning of Cycle EOC : End of Cycle
SR-1 : Safety Rod 1 CR-1 : Control Rod 1
ÖZET
TR-2 REAKTÖRÜ İÇİN KALP DÖNÜŞTÜRME HESAPLARI KISIN A : NÖTRONİK HESAPLAR
TR-2 reaktörü için kalp dönüştürme çalışmaları, 1980 yılında, uygun olabilecek yakıt tiplerinin araştırılmasıyla başlamıştır. Bunun için çeşitli yakıt tipleri analiz edilmiş ve yeni yakıt dizaynları i- çin optimizasyon hesapları yapılmıştır.
Silisli yakıtlarda kaydedilen son gelişmeler sayesinde, şim diki yakıtlarda hiçbir geometri değişikliği gerektirmeyen U,Si2 y a kıtların kullanılmasına karar verilmiştir. Yüksek zenginlikli13 yakıt larla aynı performansı verecek düşük zenginlikli U^Si^ yakıtlar için yoğunluk araştırması yapılmıştır.
En son hesaplar şimdiki kalp performansım daha da iyileşti ren 4.8 gr/cnr'lük uranyum yoğunluğu için gerçekleştirilmiştir. Kalp teki yakıtların şu andaki yanma dağılımlarım tesbit edebilmek için üç boyutlu yanma hesapları yapılmıştır. Mevcut yüksek zenginlikli y a kıtlardan azami yararlanmayı sağlıyabilmek için çeşitli denge kalbi dizaynları denenmiştir. Karışık Ye tam düşük zenginlikli yakıtlarla yüklü TR-2 denge kalbi hesapları yapılmıştır.
ABSTRACT
CORE CONVERSION CALCULATIONS FOR THE TR-2 REACTOR PART A: NEUTRONICS CALCULATIONS
Core conversion activities started in 1980's, with the inves tigation of the appropriate fuel type for the TR-2 reactor. Different possible fuel types have been tested, and optimization calculations have been covered for a new fuel design.
Based on the development of the siliside fuels, it has been decided to use U,Si2 fuels, since no modification was necessary in the fuel element geometry for the conversion from HEU to LEU. Density search calculations had been carried out to find a suitable density for the U,Si9 fuels, to achieve the same core performance with the HEU fuels: ù
The latest calculations were carried out for an uranium den sity of 4.8 g/cc, which increased the core performance. Three dimen sional burnup calculations have been made for the HEU fuels to update the present core status. Many equilibrium core designs have been tes ted for the optimum utilization of the available HEU fuels. Mixed and fully converted TR-2 equilibrium core calculations have been per formed .
CONTENTS
Page
1. INTRODUCTION... 1
2. C ARACT E RIZ AT I ON OF THE PRESENT STATUS OF THE TR-2 C O R E ... 1
3. EQUILIBRIUM CORE CALCULATIONS...3
4. CONTROL ROD W O RT HS... 4
5. CALCULATION OF SAFETY PARAMETERS... 5
6. SU M M A R Y ... 6
RE FE RE NC ES... 8
FIGURES Figure- 1 : The initial loading of the TR-2 reactor... 9
Figure- 2 : The calculated and experimental reactivity values of the standart fuel elements... 9
Figure- 3 : Core configuration of TR-2 reactor for cycle-10... 10
Figure- 4 : TR-2 equilibrium core (Design No :l )... 10
Figure- 5 : TR-2 equilibrium core (Design No : 2 ) ... 11
Figure- 6 : TR-2 equilibrium core (Design No :3 )... 11
Figure- 7 : Isothermal feedback coefficients of reactivity for HEU fuel loaded TR-2 small c o re ... 12
Figure- 8 : Isothermal feedback coefficients of reactivity for LEU fuel loaded TR-2 small c o re ... 12
Figure- 9 : Reactivity effect of percent void for HEU and LEU fuel loaded TR-2 small cores... 13
Figure-10 : Reactivity feedback coefficients for LEU fuel loaded TR-2 equilibrium core (Design No.: 3, B O C ) ... 13
Figure-11 : Reactivity feedback coefficients for LEU fuel loaded TR-2 equilibrium core (Design No.: 3, EOC)... 14
Figure-12 : Comparison of reactivity feedback coefficients from BOC to EOC for TR-2 equil. core (Design No.: 3 ) ... 14
TABLES
Page Table- 1 : The results of diffusion and Monte-Carlo calculations
for the TR-2 reactor initial loading (Full core)... 15 Table- 2 : The results of diffusion and Monte-Carlo calculations
for the TR-2 reactor small core loaded with LEU fuels (Full core)... 15 Table- 3 : Criticality experiment results for the TR-2 reactor--- 16 Table- 4 : The calculational and experimental reactivity values
of the Be blocks... 16 Table- 5 : The calculational and experimental burnup values of
the standart and control fuel elements during 9 cycles...17 Table- 6 : The discharge burnup levels of the different fuel e-
lements for the TR-2 equilibrium cores (Design No: 1, 2) ... 18 Table- 7 : The excess reactivities, power peaking values and the
discharge burnup levels of the different fuel ele ments for the TR-2 equilibrium core (Design No: 3 ) .... 18 Table- 8 : The control rod worths for different TR-2 cores loa
ded with HEU or LEU or HEU and LEU mixed fuels... 19 Table- 9 : Control rod worths and power peaking values for the
BOC and EOC of the TR-2 equilibrium core (Design No: 3 ) ... 19 Table-10 : Kinetic parameters for the TR-2 reactor small core
loaded with fresh HEU fuels...20 Table-11 : Kinetic parameters for the TR-2 reactor small and e-
quilibrium (Design No: 3) cores loaded with fresh LEU fuels... 21
1.INTRODUCTION
Core conversion activities at ÇNAEM started in 1980's. Diffe rent possibilities for the TR-2 reactor core had been investigated. The fuel types UA1X , UgOg, UZrHx (TRIGA), U3Sİ2 have been s t u d i e d ^ .
Number of fuel plates per element and the meat thicknesses have been varied to find an optimum design for the TR-2 reactor core.
After the development of the siliside fuels, it has been d e cided to use UgSig fuel for the TR-2 reactor, since there would be no modification necessary in the fuel element designs for the conversion from HEU to LEU fuels. Density search calculations had been carried out to find a suitable density for the U,Si, fuels, to achieve the
121 c
same core performance with the HEU fuels1 J . An uranium density of approximately 4.0 g/cc seemed to satisfy the above criteria for the HEU equilibrium core.
The latest calculations were carried out for an uranium d e n sity of 4.8 g/cc, which has exactly the same design with the French OSIRUS reactor, which increased the core performance for the TR-2 r e actor. This fuel design is also identical with TR-2 HEU fuels, except there is no extra cladding on the two outer plates, which is unneces sary for todays technology.
The main steps of this study are :
- Optimization of the usage of the available HEU fuels until the import of LEU fuels.
- Mixed core calculations to figure out a fuel shuffling strategy for the maximization of the discharge burnup le vels for the HEU fuels.
- Equilibrium core calculations with LEU fuels that will s a tisfy the operational needs.
2. CARACTERI ZAT I ON OF THE PRESENT STATUS OF THE TR-2 CORE
The few group cross-section libraries for HEU and LEU fuels were generated for different regions in the core with the EPRI-CELL c o d e ^ . The RABANL integral transport option of the M C ^ - 2 ^ code
2
-was used to accurately account for the resonance self-shielding of U-238. Two- and three-dimensional diffusion and burnup calculations were carried out by D I F 3 D ^ and R E B U S - 3 ^ codes. lOppm Bo equiva
lent was taken to account for the impurities inside the fuel mate rials. It corressponds to approximately 300 pcm anti-reactivity for the TR-2 reactor core initial loading which is shown in Fig.-l.
The first criticality experiment and some typical critical configurations from the initial laoding of the TR-2 reactor were cho sen to check the accuracy of the calculations. The results of 3D Mon t e - C a r l o ^ and diffusion calculations for the small core with the HEU fuel are given in Table-1. The deviations between the 2 calcula tions are within the range of statistical errors. This good agreement
is also observed for the small core loaded with LEU fuel (Table-2). The criticality experiment results are given in Table-3. The reactivity worths of Be blocks are given in Table-4, and the reacti vity worths of the standard fuel elements are given in Fig.-2. The deviations may come from the experimental estimate of the excess re activity for the full core. Different rod combinations give quite different excess reactivities (Max.: 6462pcm, Min.: 5067pcm).The lar gest deviation is seen in CR-2 calibration experiment. The Be block reactivity worths seem to be overestimated in the calculations.
Burnup calculations started with tracing the TR-2 reactor up to date. The first initial loading was conserved during the first 9 cycles. Only 2 A1 blocks were replaced by 2 Be blocks in cycle 3 in order to increase the excess reactivity to overcome the Xe poisoning p r o b l e m ^ . During the first 2 cycles 3 dry irradiation facilities were inserted at the core periphery. Two of them were dislocated du ring cycle 7 and 8.
The calculated and experimental burnup values for standard and control fuel elements during the 9 cycles are given in Table-5. Calculations were carried out in 3D for control rod CR-2 %50 inserted situation in order to account for the flux depressions due to adjust ments of the CR-2 control blades for criticality during burnup.
3
-The two Be blocks were replaced by two standard fuel elements in the tenth cycle. Also one standard fuel element at core position 45 was replaced by an irradiation element and 4 additional Be blocks were located at the other edge of the reactor core in this cycle. The core configuration for cycle-10 is given in Fig.-3. This core is in tended to be used for a cycle length of 240 MWD's and then a transfer to the equilibrium core will be made.
3. EQUILIBRIUM CORE CALCULATIONS
The REBUS code has an equilibrium core philosophy in such a way that each shuffling pattern repeats itself in every cycle. If one has some patterns that repeats itself in every two or more cycles, then one should represent it in terms of the above criteria. We will have a fixed shuffling pattern for the standard fuel elements that repeats itself every cycle, for the TR-2 reactor. But, the pattern for the control fuel elements will be repeated in every 4 cycles and the pattern for the irradiation fuel elements will be repeated in mo re than 4 cycles. In order to simulate the shuffling patterns more accurately, we started from fresh loading and repeated the desired patterns for many cycles until a reasonable convergence is obtained. Generally, the minimum number of iterations needed for convergence is a few more then the number of shuffling positions of the longest pat tern. This new iteration procedure was followed in all of the equi librium core calculations.
Three major equilibrium core designs were considered with the LEU fuels. Different shuffling schemes and different loading strate gies were investigated for the minimization of power peaking factors and for the maximization of the discharge burnup levels. The first one had 4 control, 2 irradiation and 14 standard fuel elements. It is reflected on both opposite sides by 10 Be elements. The shuffling scheme and the fuel loading strategies for the different fuel ele ments are given in Fig.-4. The discharge burnup levels of the dif ferent fuel elements for a cycle length of 260 MWD's are given in Table-6.
4
-The second equilibrium core had 4 control, 1 irradiation and 16 standard fuel elements. It is also reflected by 10 Be elements on both sides (see Fig.-5). The discharge burnup levels of the different fuel elements for the cycle lengths of 200 and 250 MWD's are given in Table-6.
The third equilibrium core had one additional irradiation fuel element then the second one. One Be block is located inside the core for irradiation purposes (see Fig.-6). The BOC and EOC k-effec- tive values, power peaking factors and the discharge burnup levels for different fuel elements, for cycle lengths of 250, 260, 270, and 280 MWD's are given in Table-7.
Mixed core calculations were made for the second equilibrium core. The same shuffling schemes which is given in Fig.-5 were follo wed for HEU and then LEU fuel elements. The power peaking values are sligthly higher in mixed core then HEU and LEU equilibrium cores, but still below the safety margins. The maximum power peaking value is observed when a fresh irradiation element is inserted into the middle of the core.
4. CONTROL ROD WORTHS
The control rod worths for different cores (small, intermedi ate, equilibrium) loaded with HEU or LEU or HEU and LEU mixed fuels were calculated by diffusion theory using effective cross-sections for the control rod regions. The difference between full and half- symmetric core calculations in 3D differ only by less then 4%, so half-symmetric core model was used in most of the calculations. Re sults are summarised in Table-8. In general, rod worths increase from BOC to EOC. This is demonstrated for the third equilibrium core in Table-9.
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5. CALCULATION OF SAFETY PARAMETERS
Safety related neutronic parameters needed for transient ana lysis of the TR-2 reactor cores loaded both with HEU and LEU fuels have been calculated. Ten group structure was used in the generation of cross-sections, with four thermal groups below 0.625 eV, which is the upper energy limit of group 5 in the standard 5-group set, to increase the accuracy of the coefficient of spectral feedback caused by increasing the water temperature.
Reactivity feedback coefficients were calculated for HEU and LEU fuel loaded small, intermediate and equilibrium cores. Three phy sical effects have been seperately encountered in these calculations:
- The hardening of the neutron spectrum resulting from an in crease in the water temperature only.
- The increase in the neutron leakage resulting from a change in the density of water as it heats (or boils).
- The increase in U-238 epithermal absorption as the tempera ture of the fuel meat increases (i.e., the Doppler effect). The Doppler effect has not been calculated for HEU fuel, be cause of the low U-238 content. Also, reactivity effects associated with the expansion of fuel and structural materials have not been encountered. Cross-sections generated for various water and fuel tem peratures were used to evaluate the feedback coefficients. The re sults of these calculations were plotted in Fig.-7, Fig.-8 for HEU, and LEU fuel loaded TR-2 small core, respectively. The reactivity ef fect of % void in the moderator also shown in Fig.-9 for both HEU and LEU fuels. There is a slight change in the feedback coefficients from BOC to EOC. This is demonsrated in Fig.-10 and Fig.-11 for the third equilibrium core. There is almost no change in reactivity between BOC and EOC cores due to changes in water density, and fuel temperature, but there is a slight inrease from BOC to EOC due to changes in water temperature. This is shown seperately in Fig.-12.
Kinetic parameters such as prompt neutron generation times, delayed neutron fractions for different TR-2 cores were calculated
U 6 U
-sing 3D diffusion theory perturbation capability of the ARC system. Two calculations were made for HEU fuel loaded TR-2 small core. The first one was made for all rods out situation in 3D half-symmetric core. The second one was made for the control rods CR-1 and CR-2 %50 inserted situation in full 3D geometry. The results are given in Tab le-10. Except the neutron generation time, the variation between the two cases is not considerable. The kinetic parameters were also cal culated for LEU fuel loaded small, half-symmetric core for all rods out situation. They were calculated for BOC and EOC of the half-sym metric, LEU fuel loaded third equilibrium core for CR-2 fully inser ted situation. The results of these 3 calculations are given in Table -11. The kinetic parameters for cycle 10, mixed and other TR-2 equi librium cores were also calculated and found in the ranges given in Tables 10 and 11.
6.SUMMARY
The main objectives attained in this study can be summarized as follows :
1) Safe operation of the reactor during all phases(HEU small - HEU intermediate-HEU and LEU mixed-LEU equilibrium cores): To realize this the fuel shuffling and loading strategies were optimized in all stages, in order to keep the power peaking factors under the safety margins defined by thermo-hydraulic calculations. The power peaking factors become important, especially for mixed cores. Besides this, the excess reactivities for the BOC's of all core loadings were ad justed in such a way that they could be easily compansated by the an ti-reactivity of any of the 3 control rod combinations.
2) Economical usage of the HEU and LEU fuel elements: Be ref lectors were used in two opposite sides of the TR-2 reactor core to increase the burnup levels. Different loading patterns and cycle lengths were used to optimize the discharge burnup levels of diffe rent type of fuel elements. In this way the discharge burnup levels were increased up to around %60.
7
-3) Establishment of the best irradiation positions : The ir radiation element and Be block positions were optimized in order to obtain the maximum efficiency for the irradiation samples.
4) Facilitate the operation of the reactor : The loading pat terns were adjusted in such a way that the BOC and EOC excess reacti vities don't differ much from cycle to cycle and also able to compan- sate the Xe+Sm anti-reactivity for continious operation of the reac tor during the whole cycle. Since there is a fixed shuffling scheme for each type of element, the flux values throughout the reactor will remain nearly the same at each cycle. This simplifies the figuring out of the best irradiation positions. The control rod worths in the second LEU equilibrium core are almost the same because of symmetry. This is advantageous for reactor operation, since there will be the same safety margin for any rod-stuck problem.
8
-REFERENCES
[1] T. Aldemir, M.H. Turgut, M.M. Bretscher, J.L. Snelgrove, "A Fea sibility Study Concerning the Conversion of the TR-2 Reactor from Using Highly Enriched Uranium to Light Enriched Uranium", ÇNAEM- R-217 (1982)
[2] M.H. Turgut, "Density Search Calculations for the Siliside Fuels" ÇNAEM Technical Report No:37 (Dec. 1986)
[3] B.A. Zolotar, et. al., "EPRI-CELL Code Description", Advanced Re cycle Methodology Program System Documentation, Part II, Chapter 5, Electric Power Research Institute (Sept. 1977)
[4] H. Henryson III, B.J. Toppel, and C.G. Stenberg, "MC^-2 : A Code to Calculate Fast Neutron Spectra and Multigroup Cross-Sections", Argonne National Laboratory Report, ANL-8144 (June 1976)
[5] K.L. Derstine, "DIF3D : A Code to Solve One-, Two-, and Three-Di mensional Finite-Difference Diffusion Theory Problems", Argonne National Laboratory Report, ANL-82-64 (Apr. 1984)
[6] B.J. Toppel, "A Users Guide for the REBUS-3 Fuel Cycle Analysis Capability", Argonne National Laboratory Report, ANL-83-2 (March 1983)
[7] R.E. Prael and L.J. Milton, "A User's Manuel for the Monte-Carlo Code VIM", Argonne National Laboratory Technical Memorandum FRAM- TM-84 (Feb. 1976)
[8] M.H. Turgut, "Neutronics Calculations of the TR-2 Reactor Present Core", ÇNAEM Technical Report No: 30 (Sept. 1986)
9 + 32 42 52 * 62 Be Be Be Be 33 ** 43 SRI 53 63 SR2 1 CE FE CE 34 CR1 44 54 64 CE FE FE 2 35 45 55 CR2 65 4 3 CE 5 36 46 56 66 Al 7 6 Al +) Core position Be : Berilium block ★ Be : Berilium block with hole **) Loading step No. SRI : Safety rod 1 SR2 : Safety rod 2 CR1 : Control rod 1 CR2 : Control rod 2 CE : Control element FE : Standart element A1 : Aliminium block 1. step 2. step 3. step 4. step 5. step 6. step 7. step
Reactor not critical Reactor not critical Reactor not critical
Reactor critical; SRİ, SR2, CR1 out, CR2 at 683
Reactor critical; SRİ, SR2, CR1 out, CR2 at 441
Reactor critical; SRİ, SR2, CR1 out, CR2 at 271
Reactor critical; SRİ, SR2 out, CR1 at 780, CR2 in
Figure-1 : The initial loading of the TR-2 reactor
•k Be Be Be Be 2533* 3770 2783+ SRI 4649 SR2 4475 4415 2804 CR1 5469 5358 2964 2387 3270 CR2 2260 2462 3558 2307 1841 2062 A1 1809 1994 A1
*) Be element with hole
#) Experimental value [pern] +) Calculational value [pem]
Figure-2 : The calculated and experimental reactivity values of the standart fuel elements
10 -Be Be ★ Be Be ¥ S107 C013 SI 15 C011 SRI SR2 20.13 7.72 10.33 10.41 C012 SI 18 S105 S108 CR1 11.91 1.87 0.0 20.68 S114 1001 C014 S109 19.30 0.0 CR2 2.34 18.21 SI 12 S117 SI 16 SI 13 14.93 3.25 5.38 17.50 Be Be ★ Be Be
*) Be element with hole #) Fuel element ID. No. +) Calculational burnup
value (%)
Figure-3 : Core configuration of TR-2 reactor for cycle 10
* Be Be Be Be Be 6 9 11 d 5 1 b 12 0 0 4 A 3 13 14 c 2 7 0 0 a 10 8 B * Be Be Be Be Be
1, 2, 3, ... : Standart fuel element loading stages
a, b, c, d : Control fuel element loading stages
A, B : Irradiation fuel element loading stages
★
Be, Be : Be blocks without and with hole
11
-Be ★ Be Be Be 9 11 8 Be 6 c 16 b 5 2 13 0 0 14 1 3 a 15 d 4 Be 7 12 10 Be Be * Be BeFigure-5 : TR-2 equilibrium core (Design No : 2)
Be CD (D * Be Be 9 13 12 Be 6 c ★ Be b 5 2 0 0 A 16 15 1 3 a 0 0 B d 4 7 11 14 10 8 Be Be Be Be Be
D E L T A R H O * 1 0 0 0 -D E L T A R H O ’ lO O O
12
-TEMP (C)Figure—7 : Isothermal feedback coefficients o f reactivity for HEU fuel loaded TR—2
sm all core
Figure—8 : Isothermal feedback coefficients of reactivity for LEU fuel loaded TR—2
D E L T A R H O * 1 0 0 0 -D E LT A R H O ’ lO O Q
13
-HEU LEUFigure—9 : Reactivity effect of percent void for HEU and LEU fuel loaded TR—2 sm all cores I 12 10 8 6 4 2 0 DENS EFFECT DOPL EFFECT TEMP EFFECT 0 50 100 150 TEMP (C ) 200
Figure—10 : Reactivity feedback coefficients for LEU fuel loaded TR—2 equilibrium
D E L T A R H O * 1 0 0 0 -D E L T A R H O 'l O O Q
14
-LEU fuel loaded TR—2 equilibrium core (Design No.: 3, EOC)
i BOC EOC 0 50 100 150 TEMP (C) 200
F ig u r e -12 : Comparison of reactivity feedback coefficients from BOC to EOC for TR—2 equil. core (Design No.: 3)
15
-Table-1 : The results of diffusion and Monte-Carlo calculations for the TR-2 reactor initial loading
(Full core) Explanation k-effective A k of Monte-Carlo Diffusion /o kl k2 All rods at 520 0.98626 0.99265 0.65 CR-1 at 780, CR-2 in 0.99964 0.99863 0.10 4 Be out, All rods at 671 0.99109 0.98972 0.14 All rods are out 1.06549 1.06805 0.22 SR-1 in, others out 1.00519 1.00434 0.08 SR-2 in, others out 1.04009 1.03202 0.75 CR-1 in, others out 1.02650 1.02195 0.43 CR-2 in, others out 1.00236 1.00428 0.19
Table-2 : The results of diffusion and Monte-Carlo calculations for the TR-2 reactor small core loaded with LEU fuels (Full core) Explanation k-effective A k 0/ Monte-Carlo Diffusion kl k2 All rods are out 1.07009 1.06913 0.08 SR-1 in, others out 1.01588 1.01244 0.33 SR-2 in, others out 1.04054 1.03670 0.36 CR-1 in, others out 1.02860 1.02864 0.004 CR-2 in, others out 1.01492 1.01412 0.08
Table-3 : Criticality experiment results for the TR-2 reactor Loading step No. U-235 content Total U-235 content [gr] Excess reactivity [pcm] |E-C| 0/ Cgr] Calc.(C) Expr.(E)* 7o E Initial loading 1676.54 - - -1 281.23 1957.77 - - -2 281.83 2239.60 - - -3 280.74 2520.34 - - -4 280.13 2800.47 650 821 20.83 5 280.47 3080.94 2649 2791 5.09 6 281.33 3362.27 4528 4828 6.21 7 281.56 3643.83 6337 6324 0.21
Table-4 : The calculational and experimental reactivity values of the Be blocks Removed Be block core positions Calculated k-effective Reactivity worth [pcm] |E-C| 0 / Calc.(C) Expr.(E) / o E 32 1.05793 861 698 23.35 42 1.05047 1532 1248 22.76 32+42 1.04279 2233 1784 25.17 32+42+62 1.03460 2993 2348 27.47 32+42+52+62 1.02308 4081 3089 32.11
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Table-5 The calculational and experimental burnup values of the standart and control fuel elements during 9 cycles
Fuel Burnup values (in percent) during the cycles
e lemeriL ID. No. 1 2 3 4 5 6 7 8 9 S101 E 7.91 13.80 20.16 _ _ 23.20 25.19 26.95 _ C 8.02 14.02 21.35 - - 24.46 26.82 28.81 -S102 E 7.04 12.37 20.58 - - 22.86 25.50 - -C 6.79 11.90 20.28 - - 23.07 25.70 - ~ S103 E 7.70 13.12 - 14.80 17.31 19.96 22.69 24.76 26.27 C 8.19 14.29 - 16.01 18.44 20.89 23.17 25.25 27.30 S104 E 6.84 11.56 19.43 21.12 - 23.95 - - -C 7.33 12.82 21.08 23.03 - 26.03 - - -SI 05 E - - - - - - - - -C - - - - - * - - - -S106 E 6.38 10.41 17.14 19.08 22.04 - - - -C 6.20 10.93 17.47 19.26 22.29 - - - -SI 07 E 5.60 - - 7.51 10.14 12.59 15.19 17.30 19.50 C 6.25 - - 8.06 10.85 13.66 15.86 17.70 20.13 S108 E 5.52 10.03 17.59 19.66 22.07 - - - -C 5.09 8.85 15.85 17.91 20.68 - - - -S109 E 4.71 9.27 15.67 17.11 19.67 - - - -C 4.18 8.08 13.94 15.53 18.21 - - - -S110 E 5.43 10.54 17.17 19.11 22.20 - - - -C 4.92 9.65 16.61 18.33 21.47 - - -Sill E 6.04 10.42 18.24 - - 21.18 23.89 25.34 27.11 C 6.17 10.84 17.98 - - 20.64 23.35 25.20 27.50 SI 12 E - 4.09 - 5.89 8.13 10.51 12.73 14.74 16.44 C - 3.31 - 4.80 6.81 8.88 10.95 12.82 14.93 S113 E - - 6.27 7.65 9.98 12.48 15.14 16.91 18.75 C - - 5.24 6.55 8.90 11.30 13.70 15.56 17.50 S114 E - - 6.49 8.02 11.05 13.88 16.12 17.76 19.89 C - - 6.04 7.56 10.34 13.13 15.52 17.22 19.30 S115 E - - - - 2.48 4.81 7.18 8.75 10.95 C - - - - 2.35 4.73 6.54 8.18 10.33 S 116 E - - - - - - 2.33 3.94 5.63 C - - - - - - 2.09 3.52 5.38 SI 17 E - - - - - - - 1.78 3.57 C - - - - - - - 1.64 3.25 S118 E - - - - - - - - 1.99 C - - - - - - 1.87 coil E - - - - 3.22 6.20 8.94 10.81 12.78 C - - - - 3.35 5.51 7.36 8.80 10.41 C012 E - - - - - 3.14 6.19 8.36 10.51 C - - - - - 3.40 6.66 9.14 11.91 C013 E - - - - - - 2.87 5.03 7.14 C - - - - - - 2.96 5.22 7.72 C014 E - - - - - - - - 2.41 C - - - - - - - - 2.34 C015 E 8.52 14.55 21.86 23.70 26.37 28.89 - - - ’ C 8.99 15.65 24.52 26.61 29.78 32.93 - -C016 E 6.98 11». 95 19.29 20.97 23.62 26.24 28.65 30.24 -C 7.06 12.34 19.51 21.20 23.77 26.36 28.59 30.29 -C017 E 7.20 12.33 19.42 21.16 - - - - -C 7.59 13.28 21.16 23.05 - - - - -C018 E 5.12 9.23 15.93 17.55 19.98 - - - -C 4.69 8.26 13.55 14.84 16.82 - -
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-Table-6 : The discharge burnup levels of the different fuel ele ments for the TR-2 equilibrium cores (Design No: 1,2) LEU core Design No.: 1 Design No .: 2
* Fuel loading strategy
2 1 1SE+— CEh— IE 7 10 1 1 1SE+— CE+— IE 4 13 Cycle length 260MWD 2Û0MWD 250MWD Standart fuel element (SE) 55.2 46.4 57.2 Control fuel element (CE) 63.3 50.9 62.1 Irradiation fuel elem.(IE) 48.8 56.1 67.5
2 1 *) ISEh— CEh— IE :
7 10
1 standart fuel element per cycle 2 control fuel element per 7 cycles 1 irradiation fuel element per 10 cycles
Table-7 : The excess reactivities, power peaking values and the discharge burnup levels of the different fuel elements for the TR-2 equilibrium core (Design No: 3)
1 1 Fuel loading strategy : 1SE+— CE+— IE
4 4 Cycle length [MWD] Excess reactivity [pern] P max [w/cm^]
Discharge burnup levels(%)
SE CE IE BOC EOC 250 7310 2600 251.5 54.2 63.3 46.0 260 6831 2026 253.5 56.4 65.6 47.8 270 6388 1435 255.8 58.4 67.6 49.6 280 5929 796 257.9 60.3 69.5 51.3
19
-Table-8 : The control rod worths for different TR-2 cores loaded with HEU or LEU or HEU and LEU mixed fuels
Control Rod Small core Fuel:HEU Cycle 10 HEU Equil .* core 2 HEU Equil. core 2 HEU Equil. core 2 HEU+LEU Equil. core 2 LEU SR-1 5939 4191 4133 4290 4164 3368 SR-2 3268 2402 4145 4230 4051 3344 CR-1 4223 3536 4279 4592 4142 3568 CR-2 5945 6906 4200 4340 3835 3324
*) 4 Be blocks adjacent to 2 standart fuel elements at the cor ners are removed
Table-9 : Control BOC and
rod worths and power peaking values for the EOC of the TR-2 equilibrium core (Design No:3)
Explanation k-effective
S
[pan] Rod worth [pcm] P max [w/cm^] All rods are out BOC 1.06582 6176 - 221.7EOC 1.01238 1223 229.3
SR-1 in BOC 1.02838 2759 3417 230.4 others out EOC 0.97805 -2245 3468 214.0 SR-2 in BOC 1.03047 2957 3219 241.2 others out EOC 0.97943 -2100 3323 242.6 CR-1 in BOC 1.02147 2102 4074 214.4 others out EOC 0.97051 -3039 4262 223.6 CR-2 in BOC 1.02758 2684 3492 242.5 others out EOC 0.97459 -2608 3831 251.5
20
-Table-10 : Kinetic parameters for the TR-2 reactor small core loaded with fresh HEU fuels
Layout
Half-symmetric core
All rods out
Full core CR-1 and CR-2 50% inserted eff 7.96035 E-3 7.98599 E-3 Effective delayed neutron lifetime 12.75971 12.76041 Prompt neutron lifetime 5.01064 E-5 4.96271 E-5 Generation time 4.69636 E-5 4.87960 E-5
*
3.0598 E-4 3.0704 E-4 ^ 2 1.6969 E-3 1.7023 E-3 /$3 1.4914 E-3 1.4961 E-3 /34 3.2400 E-3 3.2504 E-3 ^ 5 1.0191 E-3 1.0224 E-3^6
2.0705 E-4 2.0772 E-4\
1.2720 E-2 1.2720 E-2'h
3.1740 E-2 3.1740 E-2 ^3 1.1600 E-l 1.1600 E-l ^4 3.1101 E-l 3.1101 E-l ^ 5 1.4000 1.4000 ^ 6 3.8701 3.8701- 21
Table-11 : Kinetic parameters for the TR-2 reactor small and equilibrium (design No:3) cores loaded with fresh LEU fuels
Layout
Half-symmet. Small core All rods out
Full core Equi. core 3 B0C, CR-2 in Full core Equi. core 3 E0C, CR-2 in eff 7.92846 E-3 7.35751 E-3 7.28252 E-3 Effective delayed
neutron lifetime
12.68650 12.70643 12.70742 Prompt neutron
lifetime
3.93040 E-5 4.86082 E-5 4.91285 E-5 Generation time 3.69189 E-5 4.73101 E-5 5.04096 E-5
/bx
3.0174 E-4 2.7828 E-4 2.7487 E-4/H
1.6815 E-3 1.5701 E-3 1.5561 E-3/33
1.4827 E-3 1.3805 E-3 1.3671 E-3A
3.2249 E-3 2.9805 E-3 2.9471 E-3A
1.0259 E-3 9.5036 E-4 9.4119 E-4 ^ 6 2.1168 E-4 1.9771 E-4 1.9621 E-4 ^1 1.2722 E-2 1.2726 E-2 1.2727 E-2 ^2 3.1743 E-2 3.1718 E-2 3.1711 E-2 *3 1.1623 E-l 1.1663 E-l 1.1674 E-l ^ 4 3.1156 E-l 3.1204 E-l 3.1222 E-l^ 5 1.4002 1.3988 1.3987
