ML20058L829

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Fort St Vrain Midcycle 4 Core Performance
ML20058L829
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 09/08/1988
From: Malakhof V
GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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ML20058L807 List:
References
909750, NUDOCS 9008080216
Download: ML20058L829 (61)
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SUMMARY

TITLE FORT ST. VRAIN OR&D MIDCYCLE 4 CORE PERFORuNCE 3 DV&S APPROVAL LEVEL 2 O DESIGN DISCIPLINE SYSTEM 000. TYPE PROJECT DOCUMENT NO. ISSUE NOJLTR. N 18 RGE 1900 909750 N/C QUALITY ASSURANCE LEVEL 'SArETY CLASSIFICATION SEISMIC CATEGORY ELECTRICAL CLASSIFICATION I FSV-I PSV-I NA APPROVAL ~ PREPARED ISSUE SSUE DATE FUNDING APPLICABLE DESCRIPTION / BY ENGINEERING QA PROJECT PROJECT CWBS NO. Y3h-( "1" a(InitialRelease N/C APR e Igge ,3, p. a6 9M m ~ ** D. Albersta in ). Albersta in m -W / R elf i CONTINUE ON GA FORM 14851 N EXT IN DENTU RED OOCUMENTS

  • See List of Effective Pages.

N-9470 9008080216 900803 PDR ADOCK 0500026'7 P FDC G A PROPRIETARY INFORMATION THIS DOCUMENT IS THE PROPERTY OF GENERAL ATOMICS. ANY TRANSMITTAL OF THIS DOCUMENT OUTSIDE GAWILL BE IN CONFl0ENCE. EXCEPT WITH THE WRITTEN CONSENT OF GA. (1) THIS DOCUMENT MAY NOT BE COPtE0 IN WHOLE OR IN PART AND WILL BE RETURNED UPON REQUEST OR WHEN NO LONGER NEEDED BY RECIPIENT AND (2) INFORMATION CONTAINEO HEREIN MAY NOT BE COMMUNICATED TO OTHERS AND MAY BE USED BY RECIPIENT ONLY FOR THE PURPOSE FOR WHICH IT WAS TRANSMITTED. \\ J NO GA PROPRIETARY INFORMATION PAGE 1 0F g

s. 909750 N/C LIST OF EFFECTIVE PAGES Pane Nussber Pane Count Revisions Issue Summary 1 N/C i through iv 4 N/C 1-1 through 1-2 2 N/C 2-1 through 2-4 4 N/C 3-1 through 3-19 19 N/C 4-1 through 4-22 22 N/C 5-1 through 5-6 6 N/C 6-1 1 N/C 7-1 through 7-2 J N/C Total Pages 61 1 i

i ~. 909750 N/C 1 CONTENTS l 1. IN'.10 DUCTION......................... 1-1 2.

  • /0WER HISTORY 2-1

- 3. TESTING 3-1 3.1. Initial Criticality 3-1 3-1 j 3.2. Temperature Coefficient of Reactivity 3.3. Control Rod Group Reactivity Worth........... 3-6 3.3.1. Different Worth 3-6 3.3.2. Integral Worth. 3-12 3.4. Fluctuation Testing 3-13 3.5. Detector Decalibration. 3-16 4 3.6. PIE of FTE-2................... 3, 4. POWER OPERATIONS. 4-1 l .a 4.1.- Reactivity Discrepan'cy.. 4-1 4.2. Region Peaking Factors. 4-4 e 4.3. Axial Power Distribution. 4-16 4.3.1. Relative Integral Group Worth 4-16 4.3.2. Asial Peaking Factors 4-19 i 5. FUEL MANAGEMENT 5-1 5-1 5.1. Fuel Accountability 5-1 5.2. Fuel Particle Burnup. 6. CONCLUSIONS 6-1 7. REFERENCES... 7-1 FIGURES 2-1. FSV operation history 1984-86... 2-2 [ i 2-2. FSV operation history 1987 2-3 2-4 2-3. FSV. operation history 1988 l-3 1. FSV - BOC4 - bank 2B position at criticality 3-2 11 l l

J 909750 N/C FIGURES (Continued) 3-2. FSV - Cycle 4 - temperature coefficient 3-4 3-3. FSV - Cycle 4 - temperature defect 3-5 3-7 3-4. Group 2B rod worth 3-8 3-5. Group 4E rod worth 3-6. Group 4A rod worth 3-9 3-10 3-7. Group 3C rod worth 3-11 3-8. Group 3A rod worth 4-1. FSV reactivity discrepancy in Cycle 4. 42 4-2, 107 MW(t) (13%) 54.7 EFPD in Cycle 4 4-6 4-3. 390 MW(t) (46%) 56.8 ETPD in Cycle 4 4-7 4 4. 250 MW(t) (30%) 58.0 EFPD in Cycle 4 4-8 4-5. 650 MW(t) (77%) 96.8 EFPD in Cycle 4 4-9 4-6. 247 MW(t) (29%) 116.5 EFPD in Cycle 4. 4-10 4-7. 670 MW(t) (801) 120.3 EFPD in Cycle 4. 4-11 4-8. 250 MW(t) (30%) 126.8 EFPD in Cycle 4. 4-12 4 9. 670 MW(t) (80%) 140.2 ETPD in Cycle 4. 4 13 4-10. 673 MW(t) (801) 154.7 ZFPD in Cycle 4. 4 14 1 TABLES 3-1. Comparison of Cycle 4 measured control rod worths and gauge predictions 3-14 3-2. Measured and calculated decalibration factors for control rod withdrawal....................... 3-18 4-1. Fractional absorptions at 112.4 EFPD............ 4-5 4-2. Relative integral group worth 4-18 4-3. Summary of control rod insertiors and power fraction in top fuel sone........................ 4 20 4-4. Summary of control rod insertions and axial power factors in bottom block 4-22 5-1. Loadings at beginning of Cycle 4. 5-2 5-2. Total core heavy metal loadings for period ending July 5, 1988.... 5-3 5-3. Core heavy metal inventory. 5-4 111

i ] 909750 N/C i TABLES (Continued) i 1 5-4 Core heavy metal-loadings by segment for period ending i July 5, 1988.....................'... 5-5 I 5-5. Maximum particle burnup (% FIMA) for standard blocks at 112.4 EFFD. 5-6 1 1 r t r l + 5 1 t t IV

909750 N/C 1. INTRODUCTION The Fort St. Vrain (FSV) core was shutdown from January 20, 1984 (the end of Cycle 3) to May 16, '984 for the third refueling. During this outage the fuel elements of segment 3 were discharged and those of Segment 9 loaded into the core. The initial " cold" criticality at the beginning of Cycle 4 (80C4) was achieved on May 16, 1984. The core operated at about 35% of rated power and by June 22, 1984 achieved a burnup of about 5 effective full power days (EFPD). From June 22, 1984 to February 14, 1986, the core remained shutdown for the refurbishment of control rod drives. After February 14, 1986, the core operated at powers of 35% or less and by May 31, 1986, achieved a burnup of about 25 67FD. From May 31, 1986 to April 17, 1987, the core remained shut-down for the environmental qualification of varioits reactor systems and components. After April 17, 1987, the core operated at powers of 70% or.less and by July 28, 1987 achieved a burnup of about 54 EFPD. From July 28, 1987 to December 10, 1987, the core remained shutdown for cir-culator repair and for repair of damage resulting from a hydraulic oil fire in the turbine building. After December 10, 1987, the core oper-ated at powers of 80% or less and by July 5, 1988, achieved a burnup of about 155 ETPD. On July 5, 1988, the core was shutdown for the refur-bishment of all circulators and remains shutdown at the time of the writing of this report. The initial criticality at the 80C4 was predicted quite accurately. This indicates that the modeling of segment 9 fuel and lumped burnable poison (LBP) has been done correctly. This also indicates that the "as-built" fuel is very close to the design. During the power oper-ation, however, the calculations have systematically overpredicted the core excess reactivity. This apparent loss of core reactivity is not a safety related issue since it results in greater than predicted shutdown i 1-1

I L' ) ] i 909750 N/C' i margins. Furthermore, the monitoring of core reactivity indicates that the reactivity discrepancy (RD) between the cold and hot cores has been decreasing with burnup. This assures that the design burnup of 300 EFFD is achievable in Cycle 4. The control rod reactivity worth and temperature defect measure-monts conducted by Technical Services are in good agreement with the calculations. However, to achieve this good agreement it was necessary to perform core reactivity analyses with the 7-group GAUGE model for the exact core conditions for which the measurements were made. This still verifies the validity of the methods and nuclear cross section database for the analysis of the Fort St. Vrain core operations. Comparisons of measured and calculated region peaking factors (RPF) indicate some relatively large discrepancies. These discrepancies, how-ever, are not out-of-line f rom those observed in the previous cycles. Furthermore, the results of postirradiation examination (FIE) of the ] test element FTE-2 indicate that the calculated power is accurate (i.e., the observed isolated RPF discrepancies are most likely associated with the uncertainties in the measured data). The fluctuation detection tests (RT-500) indicated that the core I stability is going te be maintained at high powers even at the middle of Cycle 4 when the pressure drop across the core is expected to be at its highest. In summary, on the basis of tests and the monitoring of core oper-ation in the first half of Cycle 4, it may be concluded that the core performance is well within the requirements of the FSV Technical Speci-fications and it is in a good agreement with the projections made in the Segment 9 Safety Analysis Report (SAR). 1-2

909750 NIC 2. POWER HISTORY i The initial cold criticality at the beginning of Cycle 4 was achieved on May 16, 1984. The reactor thermal and electrical power gen-eration for years 1984, 1985, and 1986 is shown in Fig. 2-1, for year 1987 in Fig. 2-2, and for year 1988 in Fig. 2-3. Note, that 1988 core operation was terminated on July 5 with the core achieving a burnup of about 155 EFFD in Cycle 4. The power history is provided for general information. Data logger records can be processed to provide moro detailed information. f n 2-1

l i FIC. 2-1 FORT ST. VRAIN OPERATION HISTORY t i 100 100 i CYCL 2 4 r THERMAL POWER . _EEASMJCM_ MER._ 80- -80 e 70 - -70 ~ N N 5 60-5 o -60 o fl. D. ~ E-50 - I -50 E-Zrc Z w t O 40-Z -40 O i M Z W r O-. 30-O 1 /. j -30 7 l 20 - J 1' 0 r '? - a) i r ,1 /,, o t v 1 10 - 1 R,,,', -10 I l ii',) 0 -0 22:s s e4 :s s es es r a se e son ne 3 24 4 s u as r m se e asee m a az ie 2 2a m s e7 se a a t A M J J A S O'N D J F M A M J J A 3 0 N-D J F M A M J i 1964 1985 1986 I t - - - - = - - - - - - - --- - - - - - - - - - - ^ ' - ~ ~ = - - - ~ ~ ~ ~~~ ~ ~ ~

m t FIG. 2-2 l l FORT ST. VRAIN OPERATION HISTORY' i l 100 100 CYC4E 4 i THERWAL POWER N~ _ _E_L_E_C_U_R_I_C_A_L_ M_ _N_ _E_R_ _ -N b 80- -80 i l-70-B- M -70 ? N M t N M 60-O L -60 i e ? A y O i A 1, w o E-. 50-- -50 E-. l ~ Z Z i EQ O r M l O 40 _. ',1 l -40 O l a: a: W ' 'It W } a r;.','.- o L 30- ' u., "', P m 1 -30 6 a ~,. C, l,',- >?.f,, ',,', ' E,'. p a l '>. 20-l -20 '.- '(',*.*,* *, ' - ?. o r ,p,','*,' l g(. l -10 i 10 - H; '/ ,- (,,, < > p,,,s A y . :r:.,, - / 1 ,. - 5,. 1;- ) n 0 i O is zs iz as sz as s 23 r 2: 4 is z as a is a m 24 a 22 s es a ri as JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC 1987 - =-

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i 909750 N/C =_s 3. TESTING 3.1. INITIAL CRITICALITY The initial cold criticality at the beginning of Cycle 4 was achieved on May 16, 1984. The position of control rod group 28 at criticality depends on the average core temperature. The criticality position of 28 as a function of core average temperature was calculated with the 7-group GAUGE model (Ref. 7) prior to the initial criticality as shown in Fig. 3-1. The critical position of 28 measured at critical-ities in the temperature t. age between 100' and 120'F are in excellent agreement with the predictions as also shown in Fig. 3-1. Even the maximum diacrepancy of 2 in. translates into a reactivity discrepancy l (RD) of less than 0.001 Ak. On the basis of this measured data, it was j concluded thats (1) the amount of fuel and lumped burnable poison (L8P)- in Segment 9 fuel elements is in a very good agreement with the design, (2) the refueling operation distributed these elements in the active core as required by the specification 18-R-24 (Ref. 1), and (3) the calculational models do not need any adjustments to correct the biases specific to Cycle 4 core operation. 3.2. TEMPERATURE COEFFICIENT OF REACTIVITY The temperature coefficients (TC) for Cycle 4 were calculated prior to the beginning of Cycle 4 (Ref. 2). They were based on assumed Cycle 4 core operation as projected from the middle of Cycle 3. This projection required that certain simplifying assumptions be made in the calcula-tional model. These assumptions do not necessarily model the actual core conditions at the time of the measurements. One of the assump-tions that has the greatest impact on calculated TC is that the core temperature is uniform irrespective of core power. This is not unreal-istic for core operations at relatively high powers since the orificing 3-1

F'g. 3 t FSV - BOC4 - BANK 28 POSITION AT CRmCAUTY so ~ Bonk 28 L CALCtM ATED o wasueco 70- - ~~ O O O i C r O L 0 ^ t L., m "e .c l g 60 4 -- - ~ ~ ~ - se i e i .co C t l t i g. .~...j........- j......... m c> .c c. s. L 40 80 10 0 12 0 14 0 Mio Average Fuel Temperature, *F I i .j ,n., .en

i 909750 N/C 3 1 l l of refueling regiora is done to flatten the temperature distribution. However, at. powers lower than 8% to 10% of rated, the orifices are set for equal flow. At these powers the heterogeneity of tLe core begins to pley a significant role and, therefore, needs to be taken into account in the comparisons of measured and calculated TCs. The TC measurements were corrected by using a POKE and GAUCE calcu-lation. The calculations were done so that the reactivity difference between a flat fuel temperature distribution and a distributed fuel tem. perature distribution could be accounted for. The phenomena was most notable when high temperature mismatches occurred before the core was orificed for equal exit gas temperatures. By correcting the measure-ments in this manner, better agreement between measured and calculated TC was obtained. Furthermore, there is no direct measurement of TC. The procedure is to change the core power, which causes the changs in the core temper-ature, which in its turn, causes a change in control rod position. The " measured" temperature defect is then defined in the first approximation + as the change in the reactivity worth of control rods. Consequently, any inaccuracy of control rod worth predictions will directly affect the accuracy of TCs. In the next section it will be discussed that special care has to be taken to assure that the measured and calculated control rod worths er consistent (i.e., that the calculations are done for the core conditione at the time >f measurements). The predicted and measured TCs (Refs. 17 and 18) are shown in Fig. 3-2 and the predicted and measured temperature defects for the operating core temperatures (from 80' to 1500'F) are shown in Fig. 3-3. The agreement between the measured and calculated results is very good. The scatter of TC measured data is caused primarily by the differences in the reactivity worth of individual control rod groups. The overall temperature defect is slightly underpredicted. This will result in a small overprediction of consequences of a rod withdrawal accident (RWA) 3-3

\\ l l i i t i i i i Fig. 3-2 FSV - CYCLE 4 - TEMPERATURE COEFFICENT 0 t .. Y. ...... ~.. 1-- .... +. .. f.. - ... ~.. Predected - O Moosured - - -+- - - + - - - ---+'---i----- m .-------i------+- I 5 ..:.._...}... .. f... ..}.. ........b... .. f...-..... .5...w..... .....7.. . }.. X l v e c, _4_ .....,.s........ _... ..O....:... 3 O O U ) i w c_ 5 t e D 9 ....i.... a.. .._..a....... ....1....a.. _.1._. . 1. - O ~ O O. 05 Oi i GD i - + " - - + - - - - - + -+~

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909750 N/C ( ),y hdI discussed in-Segment 9 SAR (Ref. 3). The defect underprediction has an h ' nsignificant effect on the shutdown margins (SDM) discussed in Ref. 3. i 3.3. , CONTROL ROD GROUP REACTIVITY WORTH 3.3.1. Differential Worth he differential worth of various control rod groups was measured in accordance with the surveillance requirement SR 5.1.5 of the FSV Technical Specifications (Refs. 20 and 21). The measurements were done -for the groups that are involved in the initial cold criticality, the-rise-to-power, and steady state high power operation. The groupe 1A, .4F, 4D, and 45 could not be measured because the core is suberitical when they are inserted. The groups 3B and 3D have not'been measured because they are used to control the hot excess reactivity.- In the. second half of Cycle 4, as the hot excess reactivity decreases, one or both of these groups may be measured. co complete the surveillance. In addition to these groups,'4C has not been measured because it is involved in the steam generator boil-out process. In this process, due-to the rapid changes in the group 4C position with power changes, it is not possible to gather meaningful differential worths. The limitation of measuring only five groups out of 13 in the FSV core is not conse-quantial. The rods were designed to be fully interchangeable within their high/ low boronated compact grouping. The three rod pair, consti-tuting a group, are assigned based on core symmetry considerations and do not correlate systematically to any fuel loading or state of burnup. Consequently, the measuremen's cover all the possible heterogeneities in the core and there is no reason to believe that the groups that were not measured could not be calculated within the same accuracy as those that-were measured. The measured and calculated (with the 4-group GAUGE model) differ-ential control rod reactivity worth for group 25 is shown in Fig. 3-4, for group 4E in Fig. 3-5, for group 4A in Fig. 3-6, for group 3C in ' Fig. 3-7, and for group 3A in Fig. 3-8. As the results indicate the 3-6

8 1. o, .n. 1 ~ z.. l-l l Fig. 3-4 FSV - CYCLE 4 - GROUP 28 ROD WORTH 4 bdicted -- ---- - -+ ----- - - - - - (-- f - - - - -. - -- F ..;--...-.-i..-..-.. O seeosured 3_ .........._..;,... p.... .. p... 4.. . p. .. ; _.,4.. ." 0: o.. n i w iO O: o o-o: ~.. n'. x !. O. u ...O' ~'"I""~--- ......................y.. g --C: O - N Q. 4 .. :..... ;...p... .......+...:.....p.....: . p.. .g . g. .q..._, g '.....p.. ..;.... ;.....;........;.... 4.... w, Y. o i. o. e 4 -O i 15 0 200 0 50 10 0 - inches Withdrawn L.

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O Measured b ...; _.,...;.. i _ _.. ..i o...;.... A......;.. a. ,A -- -{ - -i- -i- -i-- - 0 0.8 - i - -i - --I"O I) -i-- - 8 x e-V - oi 5-- c -i--~: 7 f ---'- 50.6-5-~- 2 R <3 + .4....,......_..+.....:....: + 3 .-.3....-. o,.. 3 .+ n t-- -o" 0.2 - ----i---'-'---i---- - - -. - - - - - +. - - - - - ~--i----+- . -- F - (A c, 4 ~ o-0 o so 10 0 iso 200 inches Withdrawn

909750 N/C-L differential worth was predicted in magnitude and distribution with a-good degree of accuracy except for group 25. As it will be discussed in Section 3.3.2, the calculated integral worth of a control rod' group depends on the-GAUGE model. The number of flux energy groups, conse : quently, affects the differential worth. The group effect is the most pronounced for Group 25. As it will be discussed later in Sections 4.1 g and 4.3, this group represents a special case where the difference in the calculated and measured worth causes a rapid change in the reactiv-ity. discrepancy of cold criticalities. The exact cause of such behavior of group 2B is not well understood. It is different by virtue that it is.the only control rod group with partially inserted rods that can interact with other such rods in an immediately adjacent region (i.e., the regulating cod in Region 1). Furthermore, cold criticality is achieved with this group (i.e., a substantial portion of this group's withdrawal cannot be measured).- In addition, what is measured may be subject to large measurement uncertainties due to unfavorable noise to signal ratios. As will be discussed later in Section 4.1, the discrep-ancy of measured and calculated worths of 2B has essentially no impact on-eactivity prediction in the hot operating core. 3.3.2. Integral Worth G The contral rod group worths for Cycle 4 were calculatsd prior to-the beginning of Cycle 4 (Ref. 2), based on assumed Cycle 4 core opera- . tion as projected from the middle of Cycle 3. This projection required that certain simplifying assumptions be made in the calculational model. These assumptions do not necessarily model the actual caso cencicions at R the time of the measurements. The comparison of measured (Ref. 21) and calcuirced results, given in Table 3-1, indicated that the assumed core conditions results in Cycle 4 in either large reactivity worth discrepancies (e.g., the worth of 2B is overpredicted by the 4-group GAUGE model by about 0.005 Ap) or in noncompliance with the acceptance criteria (e.g., the worth of 4E is 3-12

.o. 909750 NIC x underpredicted by the 7-group GAUGE model by about 31%). Diacrepancies of such magnitude have not been observed in the previous cycles. Investigations into the cause of these discrepancies has revealed that control rod worth measurements can vary greatly depending on the actual core' conditions at-the time of the measurements. Consequently, in order to reduce uncertainties, it was found in. Cycle 4 that.it'is better to' represent the actual core conditions at the time of measure-monts as close as possible in the calculational model. If this is done then both the 4-group and 7-group GAUGE models calculate control' rod worth within the acceptance criteria (see Table 3-1). Furthermore, the more accurate 7-group results reduce the RD for the group 2B, which results in a more accurate prediction of cold criticalities. Note,'however, that the total reactivity worth of five control rod groups given in Table 3-1 is systematically overpredicted, irrespective of the GAUGE model or.the core conditions chosen. Such systematic over- - prediction has been observec in the previous cycle (Ref.19), but to a much smaller degree. The "best" calculated results (the.7-group GAUGE model with the actual core conditions) indicate a core excess reactivity ~ about o.006 Ap greater than the measurements. Thi-malculational bias results in the apparent loss of' hot excess reactivacy (see Section 4.1). One cause of systematic overprediction of control rod group worths may be-that the-burnup of boronated compacts has been neglected. The appar-ent loss of excess reactivity does nor. affect the safety of the core since it tends to increase the shutdown margins. 3.4. FI.UCTUATION TESTING Cycle 4 fluctuation testing under RT-500 (Ref. 6) was conducted during July 1987 and March 1988. Testing was performed by taking power rises of approximately 3% of rated thermal power and carefully monitor-ing core conditions during and for one hour after the power rise. No fluctuations or redistributions were observed during Cycle 4 testing. 3-13

1. -

= TABLE 3-1 COMPARISON OF CYCLE 4 MEASURED CONTROL ROD WORTHS AND GAUGE PREDICTIONS ~ 4-Group 7-Group 4-Gr uP (Corrected for .7-Group (corrected for - SR 5.1.5-( **""* ( ***** Cycle 4 Acceptance ~ Rod Measured Group Worths Worth Discrepancy. Worth Discrepancy Worth Discrepancy Worth Discrepancy Criteria 25 0.0320 0.0368 -131 0.0352 -91 0.0360 -111 0.0348 ' -31.

  • 201-4E 0.0093 0.0087

+71 0.0100 -71 0.0071 +311 0.0089 +4%

  • 201 4A 0.0081 0.0080

+11 0.0082 -11 0.0070 +161 0.0072 +131 +1001/-501-3C 0.0176 0.0200 -121 0.0200 -121 0.0202 -131 0.0203 -131

  • 201 3A 0.0120 0.0133

-101 0.0147 -181 0.0129 -71 0.0143 '-161-

  • 201-Total 0.0790 0.0868

-91 0.0881 -101 0.0832 -51 0.0855 -81 Measured - Calculated NOTE: Discrepancy = Calculated F ~

1 C._ 909750 N/C L. i i Initial conditions for RT-500 were established on July 7, 1987. Testing began at approximately $5% core thermal power, with a core pres- ~ c p sure drop of 2.39 paid. Testing was placed on hold following the second [ powerErise that day due to rising primary. coolant oxidants. The condi-E tions at the time were 60% core thermal power, with a core pressure drop o of 2.74 paid. 1. Testing was resumed on July 19, 1987. Beginning at 58% power and 2.30 paid, testing continued through July 21, 1987. The maximum steady y state power reached was 70% with a core pressure drop of 3.37 paid.- ~ l . Testing was again placed on hold, this time due to the trip of."C" circulator. L Following replacement of "C" circulator and the return of FSV to power operation, testing was resumed on March 12, 1988.- Initial con-q i-ditions for the resumption of testing were 61% power and 3.07 paid.. These conditions were selected toJprovide overlap with previous testing.- l A maximum steedy state power of 79% was reached on March 13. At'this L power the core pressure drop was 4.35 paid. l The test plan for RT-500 called for~ repeating the power rise sequence a second time along-a' higher core resistance line.

Howsvar, at the conclusion of the first power rise sequence,-the core tempera-

~! ture rise was approximately 800*F. A em?iew of the plant operating con-ditions indicated that it would not be poss151e to achieve the higher circulator flow rates required to test along ths higher resistance line without changing plant setpoints. A subsequent evaluation indicated that the tested maximum core pressure drop of 4.35 paid is higher than the maximum pressure drop projected for normal operation for the rest of Cycle 4. .This evaluation assumed that the plant would be operated with i the same setpoints (e.g., feed-water flow schedule, the reheat steam camperature, attemperation flow, etc.) that-were used during RT-500 for the remainder of this cycle. Therefore, the testing that was performed is adequate to cover the remaining Cycle 4 operation. 3-15 4

909750 N/C 1 4

l 3.5.

DETECTOR DECALIBRATION L There are six plant protective-system (PPS) nuclear detectors, one-in each of the six detector wells' located azimuchally at 60 des inter-vals around the core at the axial midplane. The.six linear detectors .] are' grouped in pairs to make three channels for PPS use, to provide a-l high level' scram et or below 140% of full power, and a rod withdrawal prohibit (RWP) at or below 120% of fu11' power. Both measured data and analyses have shown that decalibration i of these detectors can occur due to the motion of control ~ rods. This detector decalibration~is primarily the result of the geometric posi-tioning of the detectors around the core in the prestressed concrete reactor: vessel (PCRV) where'esch detector " sees" neutrons from only a few. fuel-columns located on.or very near the core boundary. This means that when control rods are moved to change reactor power, the detector response represents only the change in flux (power) in a few adjacent 3 fuel columns, and therefore, may not indicate the true core power level. Detector decalibration is defined as the ratio of the detector-indicated ( . power. level to the true power. i Since the control rod withdrawal sequence is different for each I fuel cycle, and since the radial and region power distribution is strongly. dependent upon the control rod configuration, the detector decalibration'also varies for each cycle. Decalibration factors are calculated for sequential control rod withdrawal and insertion followed by-a single control rod withdrawal (RWA) prior to each fuel cycle using the analytical methodology specified in the FSAR Section 7.3.1.2.1. -The procedure is to evaluate the RWA resulting in the " worst case" delay in a PPS trip. The PPS high level scram and RWP setpoint schedule as a j function of indicated power is thereby specified. \\ "[i 3-16 4 ,4

909750'N/C 1 For Cycle 4, the calculated decalibration data and the specified PPS trip setpoint'and calibration schedules are given in Ref. 2. Mea-s surements of the decalibration facters for sequential control rod group withdrawal and insertion are performed to confirm the adequacy of the analytical methodology and the conservatism in the PPS setpoint schedules. Measured data was collected for the test on two separate rise-to-7 power ascensions during 1987 and 1988, namely April to July 1987 and December 1987 to January 1988 In Cycle 3 testing, the data collected . was frequently found to be invalid for a variety of reasons, such as nonequilibrium power level or inaccurate or missing calibration data. This deficiency was minimized for Cycle 4 testing by requiring a Tech-nical Services person to be available whenever data collection was in progress. There were instances when data collection was not possible becausetof plant' conditions, such as steam generator boil-out or ini-tiation of attemperation flow, but these limitations did not severely limit the test data. Decalibration data-were collected whenever a control rod. group was fully inserted, partially withdrawn (intermediate positions), and fully withdrawn for contro1~ rod groups 25 and 3A. A full set of data was col-1ected,for all control rod groups after the initial cold criticality, except for Group 3D, which will not be withdrawn until later in the cycle,'and Groups 3A and 3B, for which data were available for only a portion of=the group withdrawal. These data were sufficient to complete the evaluation. A comparison of the measured and calculated data is provided in Table 3-2, where it can be seen that the agreement is generally within 10%. For two of the three groups where the data is outside the 10% value, namely 2B and 3A the measured decalibration factor was extrapo-laced for the fully withdrawn group, which increases the uncertainty of the measured data. The comparison between the measured and calculated 3-17

w :. ~ p '~ -e g.. I i l~ ( TABLE 3-2 MEASURED AND CALCULATED DECALIBRATION FACTORS FOR CONTROL ROD WITHDRAWAL Linear Power Channel l' l Control Rod Group III IV- 'V VI VII VIII t i l j 0.92 0.98 0.93 1.00 0.98 1.00 Measured l 2B (120.5 in. + Fout). 0.87 0.87 0.85 0.85 0.88 0.89 Calculated (*) l -6.00 -13.00 -9.00 -18.00 -11.00- -12.00 1 c 1.08 1.14 1.43 1.74 1.20 l.15 Measured ( 4E (0 + Fout) 1.12 1.16 1.51 1.71 1.16 1.13 Calculated c-m +4.00 +2.00 +5.00 -2.00 -3.00 -2.00 I 4 c 1.51 2.17 1.68 1.07 1.03 1.03 Measured 4A (0+ Fout) 1.43 2.16 1.57 0.96 0.94 0.92 Calculated c-a u -6.00 -0.5 -7.00 -11.00 -10.00 -12.00 I c (n 0.91 0.90 0.98 1.31 1.59 1.63 Measured 4C (0 + Fout) 0.85 0.89 0.94 1.46 1.75 1.68 Calculated c-a t -7.00 -1.00 -4.00 -10.00 +9.00 +3.00 1 1 c 0.87 0.84 0.86 L,04~ 1.12 1.06 Measured 3C (0 + Fout) 0.84-0.78 0.M

  • 10 1.06 1.05 Calculated c-m

-4.00 -8.00 0.0 +5.tw -6.00 -1.00 I c 1.29 1.08 1.42 0.32 1.03 0.95 Measured o-3A (30.2 in. + 103 in. withdrawal) 1.13 - 1.17. 1.20 0.83 0.86 0.82 Calculated (b)- S -c - = .) -14.00 +7.00 -18.00 +1.00 -20.00 -16.00 I o c. 0.86 0.86 0.86 0.86 0.84 0.86 Measured n 3B (0 + 133 in. withdrawal) 0.88 0.90 0.90 0.94 ' O. 96 - 0.93 Calculated (b) c-a -4.00 -1.00 -1.00 +5.00 +10.00 +4.00 I c (a) Calculated value interpolated from fully withdrawn value. .(b) Measured value extrapolated f rom part f ally withdrawn ' value. .-. z. =

J! '909750 NIC. decalibration data for control rod group withdrawal is consistent 1y'as good'as that for previous cycles and supports the analytical methods. In addition, the reevaluation to confirm the PPS 140% trip _setpoint schedule using measured and calculated data demonstrates the adequacy of ~ the schedule given in Ref. 2. A less restrictive calibration schedule than that shown in Ref. 2,= flexibility, up to two control rod groups being withdrawn or inserted before calibration, has also been demon-strated to be accepwable. The reevaluation using measured decalibration data pointed out a deficiency for the 120% RWP setpoint. For reactor powers greater than 95%, the minimum setpoint requirement of 102% (rod group 35) is less than the-setpoint schedule of 107%. However, as long as the reactor power is limited to-the current value of 82%, the RWP setpoint schedule is also adequate. 3.6. PIE OF FTE-2 Eight fuel test elements (FTE) were inserted in the,FSV core dur-ing the first refueling in 1979. FTE-2 was one of those test elements. It was irradiated for 483 EFPD in the core location 22.06.F.06. It was brought to the General' Atomics (GA) hot cell in' January 1986 to undergo destructive postirradiation examination (PIE). The examination was to verify the performance of the element by evaluating the irradiation data-of the graphite block and the FSV reference fuel. The-PIE established (Refs. 4 and 5) emong other things that the calculated power is accurate within 10%. The calculational model that was used for the analysis of FTE-2 PIE is the same as that being used in Cycle 4 for calculating the temperature coefficients (see Section 3.2), the control rod worths (see Section 3.3), the core power distribution, and the core burnup. Consequently, the FTE-2 PIE results provide an added indication of the accuracy and validity of methods and models used for the analysis of the FSV core. 3-19

s f. 909750 N/C: 4. POWER OPERATIONS Power operation in Cycle 4 has been somewhat erratic as shown in Figs. 2-1, 2-2, and 2-3. In the years before 1988, most of the cycle. was carried out at power levels of about 30%. This'1evel of operation is associated with relatively large uncertainties in various core per-formance parameters, such as the coolant flow rates, the exit tempera-tures, the primary side heat balance, etc. These measurement uncer-tainties affect the validity of comparison between the measured and calculated results. In 1988, the' core operation was conducted fairly. consistently at power as high as 80% (see Fig. 2-3). This provides a better set of data for comparison of measured and calculated results. Consequently, in the following sections, the major emphasis will be on-the core operation in 1988, i.e., for the period where core underwent a burnup of nearly 100 EFPD out of the total of 155 EFPD in Cycle 4. 4.1. REACTIVITT DISCREPANCY The RD calculated with the 7-group GAUGE model.as a function of burnup in Cycle 4 is shown in Fig. 4-1. In previous cycles the RD of " cold"-(0% power) criticalities was 0.0065 e.0.0015 Ak, and-the " hot" (>30% power) was 0.009

  • 0.001 Ak. Note, that the average RD reflects the reactivity bias of the calculational model which, because of its insensitivity from cycle to cycle may be discounted.

The accuracy of the calculational model, i.e., its reactivity uncertainty is, therefore, ' defined by the standard deviations associated with the average RD. The change in the calculational bias between the cold and hon criticalities is not well understood. However, since the reactivity biases are invariant with burnup, the RD of cold and hot criticalities has only academic significance (i.e., the prediction of cold and hot criticali-ties.can be done with essentially the same degree of accuracy). The RD 4-1

  • * * +

4-. j t Fig. 4 FSV: REACTIVITY DISCREPANCY IN CYCLE 4: [7-Group GAUGE Model Through July 5,1988 0.014 e en 0.013 Ho,.. cow i 0.0i2_ r y Z 4 0.01:. .y. w m o . in w 0 0.010 e_- i i i Avg. Cycle 3 - Hot i i i i i i t-0.009_ t o w m 0.008 i o. o w 0.0 t v. ce. b b b i .b 3 0*006 ~ y 5 5 -5 5 5 I I I B '5 3 3 5 I I 5 5 5 4 0 10 20 30 40 50 60~ 70 80'190 10 0-110 12 0 13 0 14 0 150-.160 170 18 0 190.200 i, t EFPD Sa5C PIB .,,.,s.m e = m ' ' " - - - - - - - " - - - - - " - ^ - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - m

H b {! 'i 909750 N/C of cold criticalities in Cycle '4 was in excellent agreement with gp i previous cycles until about 115 EFPD.- As results in'Tig. 4-1 indicate, 'the RD'of cold criticalities increased significantly with further burnup. The-RD of hot criticalities initially indicated a rather large reactivity bias over the previous cycles. As discussed before, this large bias is caused by the measurement uncertainties due to relatively low power operation during the first 60 ETPD in Cycle 4.. As the reactor power increased and became more steady the RD of het criticalities decreased significantly to the point where it has been in excellent i agreement with the previous cycles. The behavior-of RD bias between the hot and cold criticalities -indicate that the sudden increase of RD of cold criticalities at burnups higher than 115 EFPD is not caused by undetected or anomalous changes in the core or the control system. If this were the case,-the RD of hot criticalities would have also shown a corresponding sudden increase. Instead, they show a steady decrease of RD with burnup. The change in the behavior of cold criticalities comes about due'to uncertainty in the total worth of control rod group 25. As was. discussed in Section 3.3, this group has the highest calculated worth and shows, in absolute terms,.the' highest reactivity discrepancy with measurements. The cold. criticalities at 115 EFPD and beyond were achieved at relatively high core shutdown temperatures. This, as well as the excess reactivity loss due to burnup, resulted in cold criticalities being achieved with group 25 in the nearly fully withdrawn position. Consequently, the dif-forence between the calculated and measured worths of this group appears as the increased RD of cold criticalities. Once the group 2B is fully = withdrawn this RD change stops (i.e., the RD of hot criticalities is not affected). The overall RD decrease of hot criticalities with burnup is-due to esuses unrelated to the worth of group 25. The FSV core calculations are done not only with the 7-group GAUGE model, but also with the 4-group GAUGE and GATT models. The first of these models is used for scoping or survey calculations involving a 4-3

I

r 909750 N/C-1 c

I i large nasber of cases. The calculational time of the 4-group GAUGE is more than a factor of two cheaper than of the 7-group model. Since a high degree of accuracy is not needed for survey es*.lations, the - 4-group model has been proven to be a very cost-et,ctive tool for cal-culating the FSV core. The second of these models (i.e., GATT) issused for fuel accountability (FA) calculations. This is a few energy group.- three-dimensional model of the FSV core where the burnup of each fuel element is monitored individually. The accuracy of these two models for predicting the fractional absorptions and the core reactivity is good as given for's typical case in-Table 4-1. The GATT results (Ref. 13) appear to be somewhat different from the 4-group GAUGE results despite the similarity in the. group structure. The primary cause for these discrepancies is the neutron leakage. A recalculation of the Cycle 4 ' burnup with the three-dimensional nodal model FAN 3D (Ref. 14) indicated. 1 that the GATT model tends to overpredict core leakage, which then 1 results in a difference in the effective, and in the infinite multi-plication constants. Note that control rod worth is predicted accu-rately by the two-dimensional models even though the representation-of partially inserted rods has to be done by indirect methods. 4.2. REGION PEAKING FACTORS c r i - The measured and calculated (with the 7-group GAUGE model)'RPFs as 7 a function of burnup in Cycle 4 are shown in Figs. 4-2 through 4-10. 1 The results indicate thati 1. Most RPFs can be predicted with accuracy better than *10%. i 2. Irrespective of the power level or core burnup a few regions (e.g., 2, 12, 22, and 31) are consistently showing discrepan-a cies over *10%. The cause for these isolated discrepancies is f not known. They appear not to correlate to the type or age of fuel in the region or to the control rod configuration. The l L e 1 4-4 1 m i

l. ! -

'jt i:l b 909750 N/C m TABLE'4-1 FRACTIONAL ABSORPTIONS AT 112.4 ETPD GAUGE GAUCE GATT. -Nuclide (7 Groups) (4 Groups) (4 Groups) + ' ' 1 Th-232 0.2266 0.2252 0.2253 i t-2 Pa-233 0.0032 0.0033 0.0036 3 U-233 0.1284 0.1321 0.1339 9 Th-232 0.0865 0.0861 0.0861-10 Pa-233 0.0012 0.0013 0.0014 11 lU-233 0.0467 0.0480 0 0485: .12 U-234 0.0028 0.0028 0.0028-13 U-235 0.3402 0.3352 0.3329. 14 U-236 0.0083 0.0083 0.0085 15 U-238-0.0045 0.0044 0.0045 30 Xi-135-0.0192 0.0188 0.'0191 38 -Sm-149-0.0043 0.0050 0.0049 46.LBP1 0.0119 0.0122 0.0085 47 LBP2 0.0076 0.0077 .0.0106 48 BIMP(a) 0.0002 0.0002 0.0002 49 Silicon 0.0094 0.0096 0.0096 50 Carbon 0.0128 0.01t9 0.0132 51 Control rod 0.0244 0.0244 0.0223 52 NBIMP(b) 0.0015 0.0016 0.0016 k gg 1.0100 1.0101 1.0078 LTOT 0.0594 0.0587 0.0623 k.gg o

k. =

1.0732 1.0732 1.0748 (a) Burnable impurities. (b)Nonburnable impurities. 4-5

t 909750 N/C .e ~0 ~I -3 - g8 ~U +2

  1. g g

- -e -e. g O +a -e g 4 O-. _3 +2 .g-0 +10 0 3 g- _t + ~* +2 gO +17_, g -10 +g -6 0 ~I Control Rod Pattern

7. Diser

[(Measured /Calchateh-1]_ x 100 Os5 Os t40 ose og7 W-2 m e m a72 0,73 08 088 0a3 us 0,78 _tos 37 077-OBB 496 E OM Om a7 033 0.M 130. q t 031 W-Oss ose Oss E 0% Os0 a71 oso og W 158 t71 L34 128 Ose 09 096 L30 Oss taa gg Los OB4 OaB O 0a8 022 0.79 oag ~ 032 08 L2B 036 ua Oss -GAUGE RPF Measured RPF k = 1.0113 Fig. 4-2. 107 !N(t) (13%) 54.7 EFPD in Cycle 4 4-6

I- '.i 909750.S/C A +3 &_ g 4 -13 g r H g 'v -e -m g .7 +15 0 +11 - +10 'I 0 +1 +7 +16 O 'e -w .a g 4 +17 _3 -9 +3 i -4 -14 -e Control Rod Pattern % Discrepancy .[(Measured / Calculated)--1] x 100 1 O 0 L38 L42 l '30 L19 L38 103 .g g3 088 G71 a 135 -l g tea os, g 0% as 0 E '036 080 .g g - OSE) 083 0 g-g 140' 082 034 99 020 093 9 gg. gg 93g 130 140 l 091 089 a?7 toe osa uo-g q, ta ta t01 o 0 21 0.09 023 G71 - 084 081 taa las L39 124 GAUGE RPF Measured RPF k = LO114 ,u i Fig. 4-3. 390 MW(t) (46%) 56.8 EFPD in Cycle 4 4 4-7 2

it, 9c9750 :;;; g.R. - 4-0 44 g- . 10 + l +10 -14 L g 4 _g +9 + +2 3 N ~0 ~ +2 +m- +3 g O 0 - +2' -18' -2 g g. +16 -2 -10 +3- -10 -13 Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1] x 100 - 084 G78 137 oss 031 las-12 3 L35 106 q3 087 cas 120 m a~M Los is os7 1.36 0.73 ca0 0 OSO a2 03 Om 05 0 'L41 0.99 12 8 L10 L43 148 ) 034 02 023 020; om om ur og ue -93, tp 081 0 91 Oa2' u3 Om u 083 0.73 137 g 08 086 090 0.73 Oai 0.2 t i OM 0.78 L58 2 g GAUGE RPF Measured RPF

I k = 1.0105 Fig. 4-4.

250 MW(t) (30%) 58.0 EFPD in Cycle 4 4-8 i

= 909750 N/; 7; -4 -4> 4 -6 -13 +1 +2 ~0 +11 +13 +7 _t +14 -10 -,s +10 e +e -1 _3 9 -1 _14 _7 g 4 +15 -4 'O +4 -13 -4l Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1] x 100 038 038 080 023 g g u2 128 1DS IIS g g G70 G71-OSB 153 OSi! L30 a77 087 473 a79 'Qa8 g q L45 u0 s L44 125 G71 a*4 M M M M gy y g3 1D4 Oas u0 0a1 UI 0'70 '03 38 m m W G73 a75 0.75 02 125 u4 u7 1.09 GAUGE RPF Measured RPF k = 1.0108 Fig. 4-5. 650 tN(t) (77%) 96.8 EFPD in Cycle 4 4-9 n.--

g

. c -

.9097$0 N/C -: t: 0 +13 4 ~I4 g@ +11 ,7 -7 -16 g +14 +6 g 4 -4 +6 +3 0 + 0 +13 +15 .l 4 l g 4 ~8 ~I gQ M2 - 11 +10 -13 -6 Control Rod Pattern % Discre [(Measured /Calcufatek-1] x 100 L' QW M QW G77 L40-m a I L34 L13 L41-0,gy gg L 134 te a73 110 L72 102 t44_ t q3' aM 2 gg g W 12 8 m .136 1 L48 152 02 a3 .j E W OSE m OSE 101 q3 4 02 u6 05 0 91 07J 101 OS LOO. 1m L06 104 g 090 023 080 Om L-O G73 1R L32 L46 -2 GAUGE RPF Measured RPF l-k = 1.0111 1: Fig. 4-6. 247 MW(c) (29%) 116.5 EFPD in Cycle 4 4-10

9097$0 !!/C 4 {d - 10 0 _g +3 -12 N 0 - -16 .g ,g

  • O

-10 -2 +u +3 +10 N 4 +3 +n g. +7 0 -18 -12 +14 g 4 -4 -1 N -u -4 Control Rod Pattern

7. Diser

[(Measured /Calchatek-1] x 100 Osa Cae an Om x u, U2 125 -US 1.10 la g G70 a70 ose ' # tse Os4 _ tag 12 8 a, g a77 agg Oae 0' t4e tu

  • Lg Oss 123 G73 a75 022 05 0S8 0.W 12 g.73 g

123 too W a e. m 080 LOG 02 Ose 087 gg, U3 021 OSe 03 - a75 aM 0,% 2 U3 120 LOS GAUGE RPF Measured RPF k = 1.0112 1 Fig. 4 7. 670 MW(t) (80%) 120.3 EFPD in Cycle 4 4-11

e:

909750 ;i; +4 +1 g Q hI +7 +7 +2 -6 -16 i +9 +5 -8 _7 4 + +3 j +g + +2 0 +15 g G +2 _16 _3 gO +21 -2 _n - H .o -10 .-4 J Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1] x 100 j 0.3 Os 0.79 039 g g L33 L15 133 . LOS a72 OW 02 0.73 g g til: LW 1.04 L46 4 g73 - 0.76 E79 0.85 025 033 ~ LO1 037 093 g3 g g a79 031 E 058 01 a61 0 03 0.74 03 0.85 tzi t3s bl4 U8 - 02 090 0.73 LOO 0.61 Os ? 036 02 LO6 LO4 0.90 0.62 -{0B0 0.62 0.71 or tg g g g i h GAUGE RPF Measured RPF l k = 1.0101 Fig. 4-8. 250 IN(t) (30%) 126.8 EFPD in Cycle 4 4-12 i

$N 909730 :ii: d -8 +1 _g +2 -13 + '1 + ,1

  • U

+14- -5 +10 +12 -5 +2 +10 g k -l _19 _g +10 N -4 0 N -10 -6 Control Rod Pattern % Discrepancv [(Measured / Calculated)-1] x 100- ~f f OS4 0 84 0.7/ 0.86 g 109 L ui; 124 u3 LOS g g 3 0.69 0.70 LOO L58 0.96 L31 OR 088 - p;4 - OBO OS2 0 0.99 0.< g g g 0.72 0.75 0.2 0.93 OE. On 0.% 0.73 0.80 03 UO 0B0 LOO OM 0 57 U7 'O.90 0.99 0.78 0.~. ' O.75 0.73 0a2 086 t34 ua 121 LO6 L GAUGE RPF Measured RPF' k = 1.0 Fig. 4-9. 670 MW(t) (80%) 140.2 EFPD in Cycle I. I;- 4-13 l~ L

4' p 900750 N/: ~6 +2 + y -8 o +2 -15 i b +3 g n., g _g. +u h s 413 +7 -5 +10 d +12

  1. 0

+3 +10 g d 0 A -18 -13 +15 +7 1g -3 0 -9 -6 s. Control Rod Patte. n % Discrepancy [(Measured / Calculated)-1] x 100 .4 41 080 0.91 0?M Om 13-g LW 123 1.09 LOS UO 02 087 LO1 1.60 0.95 L31 0.91 LO1 gp 037 0 OM L56 ' LO1 L10 L47 121 0 LOS

0. ',

0.71 0.75 OM 0" 0 1.19 0.2 0.W LO4 L10 02 0.77 089 087 100 0.91 LOO ..o a75 0.71 > 0.73 0.71 W 1.14 d20 LM 0 02 L GAUGE RPF Measured RPF k = 1.0 Tis. 4-10. 673 W(t) (80%) 154.7 EFPD in Cycle 4 4-14

909750 N/C i l i severity of these randomly scattered RPF discrepancies, how-over, is very similar to that observed in the previous cycles. i 3. The RPFs of inner core regions (1 through 19) generally tend to bc tnderpredicted, while the RPF of outer core regions-(20 through 37) tend to be overpredicted. This is due to the calculational bias of the 7-group GAUGE model. This model was adjusted at the beginning of the initial cycle (based on stud-ies with multigroup/ fine mesh methods) to compensate for its bias. The adjustment has worked quite well in the first cou-pie of cycles, but appears to become less and less effective with further core burnup. The further adjustment of the model, based on the measured database, needs to be considered. Such adjustment should be validated by the measurements in Cycle 4 and then incorporated in the reference model at the beginning of Cycle 5. 4 The magnitude and/or distribution of RPF discrepancies at low powers are not significantly different from those at high i powers (e.g., see Fige. 4-2 and 4-7). 5. The magnitude and/or distribution of RPF discrepancies are q reproducible. For example, at about 30% power these discrep-ancies are essentially the same at 58.0 ETPD (Fig. 4-4), at 116.5 EFPD (Fig. 4-6), and at 126.8 EFPD (Fig. 4-8). At about 80% power these discrepancies are also the same at 96.8 ETPD (Fig. 4-5), at 120.3 EFPD (Fig. 4-7), at 140.2 EFPD (Fig. 4-9), and at 154.7 EFFD (Fig. 4-10). The stability of RPF discrepancies with burnup assures the accuracy of pre-dictions and assists the reactor operators in developing i effective orificing strategies. e The monitoring of RPFs through calculatione and measurements during burnup at various power levels indicates that the requirements of the l ] 4-15

I 909750 N/C TSV Technical Specifications are met with more than sufficient margin. The RPF patterns observed in the first half of Cycle 4 are similar to support of Segment 9 SAR. Consequently, it may N those calculated ' e the second half of Cycle 4 will also follow be concluded tb s RPFt 3 the predicted p *:- and stay well within the requirements, 4.3. AXIAL POWER DISTRIBUTION Direct measurements of axial flux distribution (and by inference, of axial power distribution) were done as a startup test A-7 at the beginning of Cycle 1 (Ref. 15). The core lacks necessary instrumen-tation to conduct such measurements after core irradiation. Conse. quently, the axial power distribution can be either indirectly deduced from the differential control rod worth measurements, or it can be gen-ersted with the three-dimensional model as a part of fuel accountability calculations. The following sections address those items. 4.3.1. Relative Differential Group Worth In Section 3.3.1 it was discussed that the differential control rod group worth measurements agree well with the predicted values, except for group 28. This good agreement is indicative of the agreement between the calculated and actual axial power distributions. If the actual dis-tribution were significantly different from the calculated, the differ-ential worth would also be different due to changes in the neutron flux distribution. The good agreement between measured and calculated group worth and its axial distribution also allows the accurate representation of partially inserted rods in the two-dimensional model. Such represen-tation is important for accurate prediction of the core reactivity (see Section 4.1) and the RPFs (see Section 4.*). The control rod groups in the second ring of refueling regions (i.e., 2A and 2B) have always been special cases. Whereas the groups in the third and fourth ring of regions could be represented by a unique 4-16

_~ - 909750 N/C axial distribution of relative integral group worth (what is e,resonly known as the "S curve"), the ring two groups has required their own spe-cial S-curve (Ref. 2). The uniqueness of these groups is causel by their much greater worth than any other group and by their protimity to the regulating rod. A partially inserted group in ring twe will in effect create a concentration of four control rods, adjacent to each other, at the core center. Such a configuration has a very strong effect both on RPFs and the axial power distribution. No such configu-ration is possible with any other group (i.e., the use of other groups results in a more even distribution of control rods in the core). T'ais characteristic of the ring two groups is why they are intentionally withdrawn prior to generation of significant power (typically they are out at less than 51 power). i L The relative integral group worth of ring two groups as a func..on l-withdrawal from the core is given in Table 4-2. Note thats (1) the measured and calculated results are in good agreement in Cycle 4, except for short withdrawal distancesI (2) the Cycle 4 data are in good agree-ment with the Cycle 1 data, especially if one considers the inevitable changes in the axial power distribution due to burnup over three cycles and (3) both measured and calculated Cycle 4 data are significantly dif-forent from Cycle 3 data. Why would group 2A be significantly different from group 25 is not known. The axial distribution of worth of group 2A was not calculated in Cycle 3, so it is not possible to state whether the calculations support the measurements. Furthermore, this group 2A axial distribution was used to predict the differential worth ef group 28 in Cycle 4. This resulted in seemingly large discrepancy as shown in Fig. 3-4. Of course, the use of the calculated S-curve for Cycle 4, or the seasured S-curve from Cycle 1 would have resulted in much better agreement between calculated and measured results for Cycle 4. 4-17

) 909750 N/C I J TABLE 4-2 j RELATIVE INTEGRAL GROUP WORTH J Cycle 4 Effective Cycle 1 Cycle 2 Cycle 3 Core Measured Measured Measured Measured Calcul.ted Layer 28 2A 2A 28 28 1 0.01 0.02 O.01 0.06 0.01 i 2 0.14 0.14 0.11 0.21 0.15 3 0.37 0.37 0.30 0.42 0.42 i 4 0.64 0.63 0.54 0.67 0.67 5 0.84 0 83 0.78 0.88 0.86 6 1.00 1.00 1.00 1.00 1.00 f i ? 1 4-18

909750 N/C On the basis of the above consideration it can be concluded that the axial power distribution in Cycle 4 is essentially the same as cal-culated for the Segment 9 SAR (Ref. 3). This contention is supported by the three-dimensional calculations that provide the fuel accountability l for Cycle 4, as is discussed in the next section. 4.3.2. Axial Peakina Factors t The axial peaking factors (APF) are calculated as a by-product of the semiannual fuel accountability (FA), with the three-dimensional model. The reporting periods are at the end of March and September of each calendar year. Although the core achieved about 155 EFPD by July 5, 1988, the FA calculations for this burnup will not be available until October 1988. The results discussed in this section are, there-fore, for the latest FA (i.e., March 31, 1988 which correspond to the Cycle 4 burnup of about 112 EFPD). The axial power distributions as a function of core location and burnup (reported in Refs. 8 through 13) are too extensive to be pre-sented in this document. Consequently, the discussion here will focus I on: (1) the power fraction generated in the top half of any region and (2) on the APF in the bottom layer of any region. The former parameter is presented for historical reasons. The extensive studies in support of the FSV FSAR indicated that power fractions in the range of 0.60 to 0.55 assure the optimum axial distribution of fuel temperatures in unrodded and fully redded regions. The latter parameter is present because of the requirements provided by the basis of the FSV Technical Specification LCO 4.1.3. Among other requirements, this specification basis stipulates that the APF of unrodded or fully rodded regions shall not exceed 0.90, and the APF of partially rodded regions shall not exceed 1.23. l The power fraction in the top half of each region as a function of burnup in Cycle 4 is given in Table 4-3. The results indicate that 4-19 . - -... ~

~ R + 909750 N/C l TABLE 4-3

SUMMARY

OF CONTROL ROD INSERTIONS AND POWER FRACTION IN TOP FUEL ZONE Cycle 4 ETPD = 4.9 24.9 59.2 80.1 112.4 i 1 2 0.497 2 0.495 2 0.494 2 0.494 2 0.472 2 0 0.566 0 0.558 0 0.557 0 0.549 0 0.521 l 3 0 0.571 0 0.564 0 0.564 0 0.573 0 0.546 4 0 0.591 0 0.587 0 0.588 0 0.581 0 0.555 5 0 0.586 0 0.582 0 0.584 0 0.592 0 0.564 6 0 0.563 0 0.557 0 0.557 0 0.550 0 0.520 7 0 0.548 0 0.540 0 0.538 0 0.549 0 0 519 8 6 0.632 6 0.621 6 0.621 0 0.596 0 0.570 9 6 0.577 6 0.560 6 0.560 4 0.514 3 0.461 10 2 0.540 2 0.526 6 0.563 6 0.548 0 0.585 11 6 0.572 6 0.563 6 0.563 6 0.584 6 0.561 12 6 0.595 6 0.585 6 0.586 0 0.563 0 0.533 13 6 0.597 6 0.583 6 0.585 4 0.543 3 0.484 14 2 0.505 2 0.492 2 0.495 0 0.576 0 0.545 15 6 0.597 6 0.589 6 0.592 6 0.615 6 0.593 16 ', 6 0.606 6 0.596 6- 0.597 0 0.572 0 0.543 17 6 0.618 6 0.602 6 0.602 4 0.555 3 0.498 j '18 2 0.538 2 0.522 2 0.519 0 0.598 0- 0.569 19 6 0.378 6 0.566 6 0.563 6 0.885 6 0.562 20 0 0.596 0 0.582 0 0.582 0 0.563 0 0.543 21 0 0.634 0 0.620 0 0.621 0 0.599 0 0.572 l 22 0 0.615 0 0.599 0 0.600 0 0.608 0 0.581 23 0 0.547 0 0.532 0 0.533 0 0.573 0 0.554 j 24 0 0.556 0 0.540 0 0.540 0 0.578 0 0.559 25 0 0.631 0 0.621 0 0.622 0 0.624 0 0.607 26 0 0 590 0 0 576 0 0.577 0 0.562 0 0.540 27 0 0.584 0 0.570 0 0.571 0 0.548 0 0.515 28 0 0.605 0 0.590 0 0.591 0 0.601 0 0.570 29 0 0.603 0 0.590 0 0.590 0 0.627 0 0.606 .j 30 0 0.555 0 0.541 0 0.541 0 0.579 0 0.559 31-0 0.584 0 0.573 0 0.574 0 0.576 0 0.557 32 0 0.657 0 0.648 0 0.648 0 0.631 0 0.611 33 0 0.648 0 0.638 0 0.638 0 0.616 0 0.585 34 0 0.578 0 0.563 0 0.560 0 0.571 0 0.537 35 0 0.590 0 04580 0 0.574 0 0.611 0 0.590-36 0 0.615 0 0.603 0 0.599 0 0.630 0 0.613 37 0 0 5A9 0 0.577 0 0 575 0 0.5'J 0 0.555 1 Average 0.585 0.573 0.573 0.580 0.552 4-20

909750 N/C the core has been maintained within the desired optimum range. The regions trith power fractions somewhat in excess of 0.60 were all recently refueled (i.e., are typically associated with high RPFs). The relatively open orifices in these regions assure adequate cooling,. espe-cially in the top half where fuel temperatures on average are always lower than in the bottom half. Consecuently, the excess power genera. tion in the top half of these regions will not have any adverse effect on fuel temperature. By the same token, the regions with the power fractions somewhat less than 0.55 are all depleted regions (i.e., are typically associated with lower RPFs). These regions are overcooled (i.e., they do not determine the maximum temperature in the core). l The APF in the bottom fuel layer in each region as a function of burnup in Cycle 4 is given in Table 4 4. The results indicate that the requirements of the FSV Technical Specifications are met with a wide margin. 4-21

909750 N/C j TABLE 4 4 SUte%RY OF CONTROL ROD INSERTIONS AND AXIAL POWER FACTORS IN BOTTOM BLOCK Cycle 4 EFPD = 4.9 24.9 59.2 80.1 112.4 1 2 0.725 2 0.740 2 0.738 2 0.797 '? 0.843 2 0 0.640 0 0.660 0 0.661 0 0.740 0 0.782 3 0 0.622 0 0.638 0 0.634 0 0.683 0 0.723 1 4 0 0.596 0 0.611 0 0.609 0 0.686 0 0.732 5 0 0.632 0 0.648 0 0.645 0 0.696 0 0.744 6 0 0.650 0 0.670 0 0.669 0 0.750 0- 0.796 1 7 -0 0.670 0 0.691 0 0.693 0 0.745 0 0.790 '8 6 0.539 6 0.563 6 0.566 0 0.658 0 0.703 ) 9 6 0.618 6 0.651 6 0.648 4 0.913 3 0.907 1 2 0.687 2 0.719 2 0.714 0 0.640 0 0.682 10 11 6 0.619 6 0.640 6 0.639 6 0.644 6 0.691 12 6 0.592 6 0.616 6 0.615 0 0.711 0 0.762 13 6 0.590 6 0.618 6 0.613 4 0.850 3 0.863 14 2 0.727 2 0.758 2 0.752 0 0.683 0 0.734 15 6 0.603 6 0.622 6 0.619 6 0.617 6 0.665 l 16 6 0.561 6 0.584.6 0.579 0 0.678 0 0.725 17 6 0.572 6 0.603 6 0.602 4 0.860 3 0.865 18 2 0.654 2 0.685 2 0.690 0 0.624 0 0.669 19 6 0.616 6 0.642 6 0.648 6 0.650 6 0.699 20 0 0.578 0 0.609 0 0.614 0 0.589 0 0.736 21 0 0.525 0 0.556 0 0.557 0 0.665 0- 0.?l2 22 0 0.524 0 0.552 0 0.545' 0 0.592 0 0.630- .23 0 0.663 0 0.694 0 0.689 0 0.657 0 0.698 24 0 0.663 0 0.699 0 0.701 0 0.660 0 0.709 25 0 0.514 0 0.539 0 0.542 0 0.567 0 0.613 26 0 0.583 0 0.613 0 0.614 0 0.683 0 0.734 27 0 0.617 0 0.648 0 0.648 0 0.761 0 0.815 28 0 0.583 0 0.615 0 0.616 0 0.662 C 0.716 29 0 0.547 0 0.573 0 0.569 0 0.538 0 0.582 30 0- 0.662 0 0.693 0 0.690 0 0.649 0 0.698 31 0 0.611 0 0.636 0 0.634 0 0.657 0 0.707 32 0 0.457 0 0.474 0 0.474 0 0.540 0 0.587 33 0 0.467 0 0.483 0 0.480 0 0.576 0 0.621 34 0 0.608 0 0.638 0 0.645 0 0.690 0 0.746 35 0 0.590 0 0.616 0 0.633 0 0.602 0 0.655 36 0 0.532 0 0.560 0 0.574 0 0.549 0 '0.596 37 0- 0.595 0 0.623 0 0.631 0 0.667 0 0.714 Average 0.602 0.628 0.628 0.679 0.723 L l 4 22 o

J x-- 909750 N/C 1 l I 5. FUEL MANAGEMENT 5.1. FUEL ACCOUNTABILITY The core total fuel loading at the beginning of Cycle 4 is given in Table 5-1. Both the two-and three-dimensional models of the core were normalized to the given data. The core fuel loading at about 155 EFr0 of burnup, as cliculated with the two-dimensional model, is given in Table 5-2. As tras mentioned in Section 4.3, the results of three-dimensional calculations are not available for this burnup. The latest FA calculatiors with such a model (Ref. 13) are for the period ending March 31, 19P4 for the core burnup in Cycle 4 of about 112 ETPD. -The cor. total heavy metal inventory at this burn is given in Table 5-3. As the results in this table indicate, the agreement between the two-and three-dimensional models is quite good. The fuel loading and burnup of reload segments in the core at 155 EFPD calculated with the GAUGE-model are given in Table 5-4 The maximum burnup (in Segment 6, one of'the initial core sogner.6s) is about 43,000 MWD / tonne (i.e., signifi-cantly lower than the 110,000 MWD / tonne required by the FSV Technical Specification LCO 4.1.1). 5.2. FUEL PARTICLE BURNUP [ The maximum burnup in terms of fissions per initial metal atom (FIMA) for the fissile and fertile fuel particles is given in Table 5-5 as a function of reload segment and the active core layer. The maximum burnup of fissile particles is 151, which is significantly lower than the 231 projected for the v.tJ-of-life of Segment 9 (Ref. 3). The maxi-mum burnup of fertile particles is less than 3%, which is a factor of two lower than the 7% projected for Segment 9 (Ref. 3). 5-1

1 909750 N/C l 1 TABLE 5-1 ] LOADINGS AT BEGINNING OF CYCLE 4 Th--232 14,241.530 kg U-233(*) 184.355 kg ) U-235 559.883 kg Uranium 896.738 kg U-233 enrichment 20.56% U-235 enrichment 62.44% Not enrichment-83.00% (*) Includes full decay of Pa-233. 1 1 1. e B 5-2

p 909750 NIC i TABLE 5-2 TOTAL CORE HEAVY METAL LOADINGS POR PERIOD ENDING JULY 5, 1988 (Cycle 4 Burnup 154.7 EPPD) Th-232 14,146.60 kg Pa-231 52.95 g U-232 45.06 g U-233(a) 226,008.90 g U-234 25,543.56 U-235 452,213.57 g U-236 94,163.14 g U-238 54,583.33 g Pu-238 1,295.53 g Pu-239(b) 1,364.51 g Pu-241 397.40 g Pu-242 220.15 : Total uranium 852.558 k; U-233 enrichment 26.51% U-235 enrichment 53.04% Net enrichment 79.55% PPM U-232 52.85 (*) Includes full decay of I Pa-233. g l (b) Includes full decay of Np.239. l 5-3 11

I o 909750 N/C TABLE 5-3 CORE HEAVY METAL INVENTORY (112.4 ETPD of Cycle 4) GAUGE GATT Nuclide (kg) (kg) Th-232 14,173.3 14.173.8 Pa-233 14.1 14.3 U-233 201.3 198.7 U-234 24.1 24.3 U-235 480.3 482.3 U-236 89.4 89.1 U-238 55.0 55.0 Np-237 5.5 5.5 s Np-239 0.0 0.0 Pu-238 1.1 1.1 Pu-239 1.3 1.3 Pu-240 0.4 0.4 Pu-241 0.4 0.4 Pu-242 0.2 0.2 Total 15,046.4 15.046.5 5-4

e- .a. TABLE 5-4 CORE HEAVY ME'.AL LOADINGS BY SECMENT FOR PERIOD ENDING JULY 5, 1988 (Cycle 4 Burnup 154.7 EFPD) . Segment 7 Segment 8 -Segment 9 Segment 4 Segement 5 Segment 6 Th-232 2,210.40 kg 2,237.89 kg 2,259.79 kg 2.325.06 kg 2,791.18 kg 2,322.28 kg Pa-231 9.38 g 8.37 g 4.27 g 9.67 g 11.63 g 9.63 g U-232 ),39 g 4.59 g 0.74 g 10.11 g 12.07 g 10.16 g U-233(*) 38,667.96 g 32,602.06 g 14,814.25 g 43,692.14 g 52,510.52 g-43,722.05 g U.234 4,002.63 g 2,965.04 g 2,015.49 g 5,180.00 g 6,386.48 g 5,193.93 g U-235 71,201.21 g 95.939.10 g 154,396.48 g 42,079.53 g 46,458.41 g 42,138.99 g U-236 18,215.34 g 15,568.53 g 7,471.72 g 17,004.49 g 18,876.40 g 17,026.70 g y U-238 9,563.81 g 10,224.31 g 11,277.03 g 7,544.11 g 8,419.65 g 7,554.44 g Pu-238 218.02 g 101.49 g 6.67 g 310.81 g 347.06 g 311.49 g Pu-239(b) 258.68-g 270.75 g 214.97 g 199.22 g 222.88 g 198.03 g Pu-240 93.49 g 88.31 g 33.47 g 78.80 g 88.38 g 79.18 g Pu-241 87.38 g 67.94 g 11.00 g 74.02 g 82.89 g 74.18 g l Pu-242 39.62 g 19.11 g 0.72 g 51.77 g 57.03 g 51.91 g MWD / metric ton 39,850.20 30,497.54 11,501.85 42,795.54 41,301.54 42,949.72 (*) Includes full decay of Pa-233. (b) Includes full decay of Np-239. 'd C.

o ,. + 909750 N/C TABLE 5-5 MAXIMUM PARTICLE BORNUP (*. PIMA) FOR STANDARD BLOCKS AT 112.4 EFPD Core Segments 4 Particle Layer to 6 Segment 7 Segment 8 Segment 9 Fissile 4 10.4 8.3 6.8 2.2 5 13.0 11.1 9.2 3.0 6 14.2 12.0 9.9 3.1 7 15.0 13.9 10.6 3.5 8 13.7 12.8 9.4 2.9 9 10.9 10.0 7.2 2.0 Fortile 4 1.1 0.5 0.3 0.0 5 2.1 1.0 0.6 0.1 6 2.6 1.4 0.7 0.1 7 2.8 1.8 0.9 0.1 8 2.1 1.4 0.7 0.1 9 1.1 0.7 0.3 0.0 ETPD 769.7 595.7 406.9 112.4 I 5-6

O 909750 N/C 6. CONCLUSIONS On the basis of results provided in this document it may be con-ciuded that the startup tests confirm the predictions made in the ses-taent 9 SAR. The monitoring of core burnup from 0 to 155 EFFD also con. firms that the core is operating well within the envelopes described in the SAR and required by the FSV Technical Specifications. Consequently, it is expected-that the core physics performance in the second half of Cycle 4 will continue to follow the established pattern. There.is no evidence that the maximum burnup of 300 EFPD could not be achieved in Cycle 4 due to restraints imposed by the excess core reactivity, the .powar distribution, the fuel burnup, or any other core physics parameter. i 1 6-1

909750 N/C 7. REFERENCES 1. "FSV Reactor Core General Assembly." 18-R-24, Issue H. December 5, 1983. 2. "FSV Cycle 4 Data for Reactivity Calculations and Physics Surveil. lance Tests," 907177A, Tebruary 7, 1984 3. " Safety Analysis Report for Fuel Reload 3 (Segment 9 - Cycle 4)," GA-C17128, May 1983 4. " Destructive Examination of TSV Fuel Test Element FTE-2," 908909 N/C, July 11, 1986. 5. " Thermal Diffusivity of H-451 Graphite and Temperature Monitor Anal,yses in FSV Fuel Test Element FTE-2," 908928 N/C, August 22, 1986.* 6. "RT-500 Test Plan (for Beginning of Cycle 4 Rise-to-Power Test-in8)," T-359, June 16, 1987. 7. Archibald, R. J., et al., "A User's and Pro 8 rammer's Guide to the GAUGE Two-Dimensional Neutron Diffusion Protram " GA-A16657, July 1983. 8.

Malakhof, V., " Fuel Accountability at 4.9 EFPD of Cycle 4 (Septem-bor 30, 1984)," 907716 N/C. November 14, 1984 9.

Lefier, W., " Fuel Accountability at 8.5 EFPD at Cycle 4 (March 31, 1986)," 908823 N/C, June 24, 1986. 10.

Rucker, R., " Fuel Accountability at 24.9 EFPD of Cycle 4 (May 30, 1986)," 908925 N/C, October 24, 1986.

11.

Rucker, R., " Fuel Accountability at 24.9 ETPD of Cycle 4 (March 31, 1987)," 909409 N/C, April 9, 1987.

12.

Rucker, R., " Fuel Accountability at 53.5 ETPD of cycle 4 (Septem.

ber 30, 1987)," 909560 N/C, November 18, 1987. 13.

Rucker, R., " Fuel Accountability at 112.4 EFPD af Cycle 4 (March 31, 1988)," 909657 N/C, June 20, 1988.

7-1 .3

r T 909750 N/C i l 14.

Rucker, R., et al., " Validation of FAN 3D Model for FSV Fuel Accountability Calculations " 909436 N/C, January 18, 1988.

15.

Brown, J., et al., " Neutron Flux Distribution Measurement in the x

i ,\\ FSV Initial Core (Results of FSV Startup Test A-7)," GA-A13176/ UC-77, February 1975. 16. Nirschl, R. J., "Close-out.of Cycle 4 Detector Decalibration Test T-216," PSC Memo PPC-88-1833, June 3, 1988. 17. " Letter of Authorisation for Middle of Cycle Fuel Operation Infor-mation," PG-1763, August 11, 1988. 18. "$R$.1.3 RX/4.1.7-Temperature Coefficient of Reactivity Surveil. lance Test for Cycle 4," June 8, 1988. 19. "FSV Cycle 3 Core Performance," GA-C18103, November 1985. 20. "SR5.1.5 RX-Control Rod Reactivity Worth Surveillance Test for Cycle 4," April 13, 1987. 1 21. Stuart, D., "SR5.1.5-Control Rod Reactivity Worth," PPS-88-0605. February 19, 1988. I 7-2 _ -.}}

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