ML19248D544

LACBWR Fuel Cycle 6 Startup Test Program Results

ML19248D544
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	08/08/1979
From:	Raffety S DAIRYLAND POWER COOPERATIVE
To:
Shared Package
ML19248D541	List: ML19248D540 ML19248D544
References
LAC-TR-070, LAC-TR-70, NUDOCS 7908160453
Download: ML19248D544 (14)

Text

{{#Wiki_filter:.- LAC-TR-070 LACBWR FUEL CYCLE 6 STARTUP TEST PROGRAM RESULTS PREPARED BY: SEYMOUR J. RAFFETY AUGUST 1979 DAIRYLAND POWER COOPERATIVE 2615 EAST AVENUE SOUTH LA CROSSE, WISCONSIN 54601 ",- ;) L 02;d) 7908160403

LAC-TR-070 TABLE OF CONTENTS Page 1.0 I N T R O D U C T I O N - - -- -- - - -- - - - - -- - - -- - - - - -- -- -- -- -- -- -- -- -- -- 1 2.0 TE ST P ROG RAM AND RE SU LTS---------- ----------- - -- -- -- ---- 1 2.1 VERIFICATION THAT THE LACBWR CORE CONFIGURATION FOR CYCLE 6 AGREES WITH THE DESIGN CONFIGURATION--- 1 2.2 CONTROL ROD DRIVE FUNCTIONAL AND SCRAM TESTS AND CHECK FOR CRITICALITY WITH ONE ROD OUT------------- 2

2. 3 VERIFICATION OF PREDICTION OF CRITICAL ROD P AT T E R N- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - -

2 2.4 SHUTDOWN MARGIN VERIFICATION WITH MOST REACTIVE ROD WITHDRAWN-------------------------------------- 3 2.5 VERIFICATION THAT THE TEMPERATURE COEFFICIENT OF REACTIVITY IS NEGATIVE AT ALL TEMPERATURES FROM AMBIENT TO OPEPATING TEMPERATURE------------------- 4 2.6 INITI AL POWER ESCALATION FOR FUEL CYCLE 6---------- 4

2. 7 COMPARISON OF CALCULATED POWER SHAPE WITH INCORE MONITOR DATA--------------------------

4

3. 0 CONCLUSIONS---------------------------------------------

5 ylDYUd' A -~nc .4 ii

LAC-TR-070 FIGURES Figures Page 1 LACBWR RELOAD CONFIGURATION FOR CYCLE 6------------- 7 2 29 ROD BANK CRITICAL POSITION VS. REACTOR T E MP E RA TU RE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- 8 3 LACBWR POWER HISTORY DURING INITIAL STARTUP AND POWER ESCALATION FOR FUEL CYCLE 6------------------- 9 4 EXPERIMENTAL AND CALCULATED RELATIVE AXIAL INCORE '10NITOR SIGNAL-------------------------------------- 10 4a EXPERIMENTAL AND CALCULATED RELATIVE AXIAL INCORE MONITOR SIGNAL-------------------------------------- 11 ~. ) v: 2 ;c,.; v,_4'9., ..x iii

LAC-TR-070

1.0 INTRODUCTION

On May 25, 1979, the NRC issued Amendment No. 16 to Provisional Operating License No. DPR-45 (Reference 1) authorizing operation of the LACBWR for Fuel Cycle 6. The LACBWR core configuration fcr Cycle 6,shown in Figure 1, consists of 26 fresh Type III (Exxon) fuel assemblies, 2 fresh Type I (Allis-Chalmers) assemblies, 32 previously irradiated Type III assemblies, and 12 previously irradiated Type II (Allis-Chalmers) assemblies. On May 25, 1979, at 2140 hours, the LACBWR was taken critical for the beginning of Fuel Cycle 6. A summary report of the plant startup and power escalation testing outlined in References 1 and 2 is presented below.

2. O TEST PROGRAM AND RESULTS 2.1 VERIFICATION THAT THE LACBWR CORE CONFIGURATION FOR CYCLE 6 AGREES WITH THE DESIGN CONFIGURATION As each fuel assembly was loaded into the LACBWR core, it was identified by the serial number on the lower end fitting by a Fuel Accountability Delegate and its position in the core documented.

After all of the previorsly irradiated fuel assemblies haa been positioned in the core according to the design configuration, the core was examined visually by the Fuel Accountability Representative to verify that an irradiated fuel assembly of the proper type was in each location designated for a previously exposed fuel assembly, and that locations designated for fresh fuel were still empty. The visibility during the examination was excellent and there was no problem in identifying the Type II and Type III assemblies by the dif ference in upper end fittings. After this examination was satisfactorily completed, the fresh fuel assemblies were loaded into the core according to the design configuration with a Fuel Accountability Delegate identifying each assembly by serial number and documenting each move. When the refueling was completed, a new high resolution TV camera was used to read the serial number on the handle of each fuel assembly in place in the core. The numbers were easily distin-guished with the new TV equipment and the as loaded core configuration for fuel Cycle 6 was verified to be in agreement with the design configuration. 7;3SO.'O x

LAC-TR-070 2.2 CONTROL ROD DRIVE FUNCTIONAL AND SCRAM TESTS AND CHECK FOR CRITICALITY WITH ONE ROD OUT Before the reactor was taken critical, all of the control rod drives were functionally checked. Each control rod was exercised over its full range of travel while carefully observing the count rate on the source range nuclear instrumentation channels. Only ve.ry small changes in the count rate due to subcritical multi-plication of source neutrons was observed during individual control rod withdrawals indicating that the reactor was signifi-cantly suberitical with any single control rod fully withdrawn. Each of the control rods was scram tested with an average scram time of 1.98 seconds and minimum and maximum scram times of 1.82 and 2.13 seconds respectively observed. The Technical Specification limit for control rod scram time is 3.0 seconds. Both scram solenoids were observed to operate as each rod was scramed. All control rod position indicators were functionally checked. All interlocks and scram functions associated with the control rods, such as high reactor power trips, short reactor period trip, high reactor pressure trip, reactor coolant flow rate abnormal trip, reactor coolant flow rate low trip, reactor water level high trip, reactor water level steam isolation valves not fully open trips, turbine stop valve not fully open trip, low oil level or low gas pressure in any control rod drive accumulator trips, low voltage trips, the interlock to prevent increasing recirculation pump speed while withdrawing a control rod and interlocks controlling reactor start permits and control rod withdrawal permits were functionally tested and/or calibrated as required by Technical Specifications before reactor startup.

2. 3

{ERIFICATION OF PREDICTION OF CRITICAL ROD PATTERN On May 25, 1979, the LACBWR was taken critical to begin Fuel Cycle 6. Criticality was achieved at 2140 hours with all 29 control rods w ithdrawn as a bank to 12. 819 dial inches. The reactor temperature was 950F and the recirculation flow rate was 7.1 x 106 lb/hr. The predicted 29 control rod bank critical position for these conditions was 12.590 dial inches. The deviation between actual and predicted critical position was, therefore, 0.229 dial inches or 0. 334% a k/k. This value is well within the LACBWR Technical Specification requirement that the reactivity worth of the dif ference between the observed and predicted critical rod pattern not exceed 2% Ak/k following core reloading and/or core rearrangement. M,0! Q LcC-TR-070 .4 SHUTDOWN MARGIN VERIFICATION WITH MOST REACTIVE ROD WITHDRAWN As reported in Reference 3, the stuck rod shutdown margin at the beginning of Fuel Cycle 6 has been calculated for each rod (excluding symmetric duplicates) using the TRILUX model. The least shutdown margin determined from this model was 1.4% ak/k with control rod No. 14 out of the core and the other 28 control rods fully inserted. The LACBWR Technical Specification requirement for shutdown margin with the most reactive control rod stuck out of the core is 0.5% Ak/k. In order to experimentally confirm that the required shutdown margin is available, the following experiment was performed: (1) Control Rod No. 14 was withdrawn from the critical reactor while criticality was maintained by inserting the remaining twenty-eight control rods as a bank. Stable critical condi-tions were obtained with Rod No. 14 fully withdrawn and the remaining twenty-eight control roc's banked at 12.679 dial inches. (2) The twenty-eight rod bank was inserted 0.221 inches so that the reactor was subcritical by 0. 358% A k/k as determined from the known worth of the 28-rod bank, and the steady state suberitical count rate of each of the startup channels was obtained. (3) The twenty-eight rod bank was inserted in several increments, and the steady state count rate of each startup channel was obtained after each incremental insertion. Theoretically, the keff of the system and therefore the shutdown margin after each control rod insertion (x) in Step (3) above may be determined from the following equation: 1-(1- (kef f)l] (CR)1 (keff)x = (CR) x WHE RE : (keff)1 and the count rate (CR)1 are determined in Step (2) above and (CR)x is the steady state subcritical count rate observed after control rod insertion (x). Af ter each incremental insertion of the twenty-eight rod bank, the count rate in the startup channels decreased indicating a continuous insertion of negative reactivity as the rods were inserted. With the twenty-eight rods fully inserted and rod 14 fully withdrawn, the above procedure yielded values of 1.27% and 1.96 % Ak/k shutdown margin from startup channels 1 and 2 respectively. The average of these two values, 1.62% ak/k, agrees quite favorably with the cal-culated value, 1.4% ak/k, determined from the TRILUX model. 3- ?;%i}%

LAC-TR-070

2. 5 VERIFICATION THAT THE TEMPERATURE COEFFICIENT OF REACTIVITY IS NEGATIVE AT ALL TEMPERATURES FROM AMEIENT TO OPERATING TEMPERATURE The moderator temperature coefficient of reactivity in the LACBWR was shown to be negative at all temperatures from ambient to operating temperature as required by the LACBWR Technical Specifi-cations by determining that continuous withdrawal of the 29 control rod bank is required to maintain criticality as the reactor is heated from ambient to operating temperature.

During the heatup of the system, the position of the 29 rod bank for criticality at essentially zero power was determined approximately every 500F and the results are plotted in Figure 2. 2.6 INITIAL POWER ESCALATION FOR FUEL CYCLE 6 Upon completion of the system heatup on the 27th of May, reactor power was escalated and an attempt was made to put the turbine-generator on line. However, problems with the turbine oil and governor system delayed power escalation for approximately two days. Instrument problems causing reactor shutdowns on May 30, June 1, and June 18 also delayed the power escalation. Figure 3 shows the power history for the LACBWR during the initial startup and power escalation for Fuel Cycle 6. In order to limit stresses in the fuel rod cladding, the initial power escalation was limite d to 10% or rated power per day between 30% and 50% of rated power, 5% of rated power per day between 50% and 70% of rated power and 3% of rated po"er per day above 70% of rated power.

2. 7 COMPARISON OF CALCULATED POWER SHAPE WITH INCORE MONITOR DATA On June 25, 1979, approximately 4 days after recching a steady state operating power of 92 % of rated power, data was obtained from the traversing incore flux monitoring system for canparison with valces calculated by the TRILUX model.

The locations of the nine incore flux monitors are shown in Figure 1. Data was obtained from only 8 of these monitors since the channel 1 detector is not operational at the present time. These incore monitors measure the neutron reaction rate in a 10B or 2 35 U lined detector as it is moved vertically in the center of a Zirconium post along the common junction line of four fuel shroud cans and the accuracy of the data is dependent on amplifier and recorder calibration and on the ability to knav the detector position in t ae core. 7dO."*b

LAC-TR-070 On the other hand, the calculation with the TRILUX model determines the nodal powar dis tributica in each of the fuel assemblies surrounding the nonitor. In order to compare the two, it is necessary to convert one to the other by means of a calculation. The procedure selected was to convert the calculated power dis tribu-tion in the fuel assemblies to predicted reaction rates in the detectors in the Zirconium posts and compare results with the incore monitor data. This reaction rate is a function of neutron spectrum and neutron density and, there f ore, is a function of finsion rato in the surrounding fuel, isotopic composition of the fuel, voids in the moderator, type of shroud can, presence of control rods, etc. A computer code called BF3, written specifically for LACBWR, is used to calculate this reaction rate at the center of the Zirconium posts from i.'RILUX output. Since the LACBWR incore monitors cannot be intercalibrated, they can only be used to deLermine axial relative flux chapes and the comparison of the amplitude of one monitor output to another has no valid basis. Therefore, it was decided that both the incore monitor data and the calculated data would be normalized to an average value of one for each channel to f acilitate their comparison. The comparison of the normalized incore monitor data and the normalized results calculated by the TRILUX and BF3 model are shown in Figures 4 and 4a. As can be seen in these figures, the agreement between the calculated and the measured axial flux shape is quite good.

3.0 CONCLUSION

S The startup testing of the LACBWR was completed on June 25, 1979, and the reactor was operating at a nominal steady state power level of 92% of rated power. All operating parameters were conservatively within Technical Specification limits and consistent with design predictions. All reactor systems were operating satisfactorily. Y.N]QX7 LAC-TR-070 REFERENCES 1. NRC Letter, Ziemann to Linder, dated May 25, 1979. 2. DPC Letter, LAC-6274, Linder to Director of NRR, dated May 9, 1979. 3. DPC Letter, LAC-6130, Linder to Director of NRR, dated February 26, 1979. 7[5dO!sO LAC-TR-070 A B C D E F G H K L 2-l.7 NT (1.45 13.ob P6 O 2-61 3 29 J-17 3-V7 3-V3 3 -13 3-/ 2-J'1 2 fl. y k 2.4Y V.J9 0 o

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g O .. gi:' t- ~ @Q L"' D 1. 2 es. = ' w..

.p gi e 0

. ap ~ M 5= g figg gG:,asz 4

_.i +:

~ = e c to:, a n h .b_L r :. _ i. i.._.4._h' M h di . l!:1.i:as==iiI} i. = I / j E =i::: 7 iamam 7+ m i-y T w .8q + + .8 o g]

._ = A1 :

8 --i g E._. h_. _ .ap:-- 5:-s ss : .gs l O. h y _.. q. a ..-- h,. +-__ __g. _d - g - -h .:H N!! ;iqij==j.[._;; p= - -- -l==H

== c-l s rm n-m .g;;.d!b l Ch-4 r - - -- J ---- - ~[ . gi;. Ch-5 + = o o .... - p -

u e

p z

._.-_LT!_. Z dikes a.:. yI:1.=:pi.1 ;..; ;4: 'p;; ..J

ns u_

b^ i I O @ E 0 i '- J 0 20 40 60 80 0 20 40 60 80 INCHES FROM TOP OF CORE INCHES FROM TOP OF CORE 1 _4 ~Jq

1. 6,-- -- ~

1.6 {WFF:FEi.n/i_..:.g+:@=f5Fpm6%Es:! . I

---

+- ]_-.;--

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I- #

'~~1 ~ o O i. - l_ _ : Q o q). {c-n =:

'yy 4 9 7-. = -.

=: =9 gg =- -- +=n g % (._ ' $ a f p -h I k G O* Q ~ % g+= =ikh M bo g y.

2.,, p. y 9__h g

.8 = -_i g 3 J.. a a 4:3 7. l.p.:; ~ g' gu;; a h) @ ---g---- 3 ^ ]# $i J i J " ~. = g-? g .g~==g p.. g g q s =.I .-).4y ; ~ .4Q. g T _ L M ---i Ch-6 -f.+--- _ _.O_- o

-..:l='"

..qp;i Ch-7 ed 4==- -?+ - -i ""-G 9 C 4=

.=

2 -i -_ j _[---J Z _jj:tjg @ O i ~ ~ ~ ',' F ~~ != ~ ~ O G;j6i=i T T'i'~ 7if ~ ~ ~J~~~ ir; 0 20 40 60 C0 0 20 40 60 80 INCHES FROM TOP OF CORE INCHES FROM TOP CF CORE REACTOR OPERATING CONDITIONS: Power = 91.83%, Equi ibrium Xenon, Recire. Flow = 10.6 x 106 lb/hr., Subcooling = 17.98 Btu /lb., ~ Control Rods: 1@ 16" 6, 10 0 50.25" 7, 11 @ 16.0" M M@. 8, 12 0 75.25" - Normalized TRILUX Dai a o Normalized Incore Monitor Data EXPERIMENTAL AND CALCULATED RELATIVE AXIAL INCORE MONITOR DATA TAKEN: '25 June 1979 SIGNAL FIGURE 4 LAC-TR-070 i L i :!,;, ;.,, .j 1.6 [ _-il .6 u!d .Ni:. [' ;hu. 2 i" ..it. A

.i;

.. S : id11Q:4; i .. a..j g 1 "OO #' b .2 i iW ' g 12 O hh.h"- Sf+n;7 'Ufkr; dih+ riif:iN 'i= N "b-9N , [;" 5E':f-WM h N-g-i:+ " + +" i+. -i - v .8 e.8 . 4E !nDE-_ d k]E{c:Eu31:. ll @di.,$ . 21.1.Ti !]iji;j!si:

. !- !r.:4 1

a .n ~ i=u n 1 ,1 = r 1.ia n ra i gi o n g.i!!!kg=. sins!!!L s in..AWy 9% n . 4.JM!!!sW nSF g.,;;gy,+1.n;. ;; :h n.:g m a+:

tsn:5r---

3 f =n. ~g P +- ch-8 =ful p :: ~ ~ ~ ~ lg au A 4...iii:s _ " . i.5 m-- d-ii]{ !!*"""D:i .. si g (. 3 Ch-9 g g O ':ini. ji!!5]iiFi:L.....i$E'il til its O i-i..i:l jh.. Mil..ll ti :..j@! iii J i +{bO"I:!IlUI INI 'i F 0 ' "" " E !" h " ' NI' iii N "C 0 O 20 40 60 80 0 20 40 60 80 INCHES FROM TOP OF CORE INCHES FROM TOP OF CCRE @) k,, s?\\ \\p r 8 ,G ,t t y l v' : 11J r d C - y n\\ \\ g A, \\(,(U f\\\\ 1 JW y h Lg 1 73sina MACTOR OPERATING CCNDITIONS: Power = 91.83%, Equilibrium Xenon, Recire. Flow = 10.6 x 106 lb/hr., Subcooling = 17. 98 Btu /lb., Control Rods: 1@ 16" 6, 10 0 50.25" 7, 11 0 16.0" 8, 12 9 75.25" 9, 13 0 25.12" - Normalized TRILUX Data o Normalized Incore Monitor Data EXPERIMENTAL AND CALCULATED RELATIVE AXIAL INCORE MONITOR DATA TAKEN: 25 June 1979 SIGNAL FIGURE 4a }}