ML20058L833

Fort St Vrain Mid to End of Cycle 4 Core Performance

ML20058L833
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/02/1990
From:	Malakhof V GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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ML20058L807	List: ML20058L803 ML20058L809 ML20058L829 ML20058L833
References
910078, NUDOCS 9008080218
Download: ML20058L833 (48)

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18 RGE 190 910078 N/C TSV MID TO END OF CYCLE 4 CORE PERf0RMANCE i

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N-9470 9008080218 900803 FDR ADOCK 05000267 F-FDC C GA PROPRIETARY INFORMATION THl3 OOCUMENT IS THE PROPERTY OF GENERAL ATOMICS ANY TRANSMITTAL OF THIS DOCUMENT CUTSIDE GA WILL M IN CONP10ENCE. EXCEPT WITN THE WRffTEN CONSENT OF GA. (1) THIS DOCUMENT MAY NOT BE COPIED IN WHOLA OR IN PART i

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NER$1N MAY NOT BE COMMUNICATED TO OTHERS AND MAY BE USED BY RECIPfENT ONLY POR THE PURPOSE FOR WHICM IT WAS TRANSMITTED 0 NO GA PROPRIETARY INPORMATION PAGE 1 OF *

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4-910078 NIC LIST OF EFFECTIVE PAGES i

Pete Number Pete Count Revision Issue Susumery 1

N/C 2 through 48 M

N/C Total Pages 48 4

2

e 910078 N/C CONTENTS 1.

INTRODUCTION 5

2.

POWER HISTORY.........................

8 3.

POWER OPERATIONS 10 3.1.

Reactivity Discrepancy 11 3.2.

Region Peaking Factors 16 3.3.

Aaial Feaking Factors.

20 4

FUEL MANAGEMENT........................

26 4.1.

Fuel Accountability..

.26 4.2.

Fuel Particle Burnup and Exposure.

29 4 3.

Power Coastdown...............

29 5.

CONCLUSIONS.....................

32 6.

REFERENCES 32 AFFENDIX As RPF DISCRIPANCIES IN CYCL? 4 (155 To 232 EFFD) 34 FIGURES 2-1.

FSV operation history.....................

9 3-1.

FSV reactivity discrepancy in Cycle 4.

12 3-2.

666 MW(t) (791) 194.8 EFFD in Cycle 4 18 3-3.

666 MW(t) (791) 232.0 EFFD in Cycle 4 19 TABLES 3-1.

Fever fraction in top fuel zone....

23 3-2.

Aniel pasking factors in bottom elements 25 4-1.

Loadings as beginning of Cycle 4 27 4-2.

Total core heavy metal loadings for period ending AuEast 18, 1989 Cycle 4 burnup 232.0 EFFD.

27 3

s 910078 N/C TABLES (Continued)

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4 3.

Core heavy metal loadings by segment for period ending

~ 28 August 18, 1989 Cycle 4 burnup 2332 0 ETPD 44 Maximum particle burnup (2 TIMA) for standard blocks at 232 ETPD 30 4-5.

Time at power during coastdown 31 e

i 4

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.s 010078 NIC 4

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1.

INTRODUCTION AND SUMARY The performance of the Fort St. Vrain (TSV) core during.the first half of Cycle 4 from May 16, 1984 to July 5.1988, has been previously documented (Ref. 1).

During this period the core achieved a burnup of about 155 effective full power days (EFFD).

The nominal design burnup I

of Cycle 4 is 292 EFFD (Refs. 2 and 3).

Segment 9 SAR (Ref. 3) allowed the exter.sion of Cycle 4 to 300 ETPD, provided the subsequent cycle was correspondingly shortened.

The reactor was shutdown on July 5, 1988 for the repair and refur-L bishment of all helium circulators.

The core power operation was resumed on March 26,1989.

Thereafter, the core operated at about 80%

of rated power until August 18, 1989.

Note, that core operation at 801 was a restriction imposed by the NRC and not by any physical limitation of the core such as inadequate reactivity, excessive temperatures, etc.

The core was shutdown on August 18, 1989, when a surveillance test indi.

cated that a control rod was malfunctioning and subsequently declared inoperable.

During the investigation of the probles steam generator ring header ' cracks were discovered, which prompted Public service Company of Colorado (PSC) to terminate core operation permanently.

At the time of the final shutdown, the core burnup was 232 EFFD which

, i falls somewhat short of the nominal design burnup of 292 EFFD.

I The core operation at the time of final shutdown for all practical l

purposes corresponded with the segment 9 SAR analysis of core condition I

at the end of Cycle 4.

In the design of Segment 9 (Ref. 2), it was con.

1 sidered that toward the end of Cycle 4 rated power operation might not be possible and some dorating of power might become necessary.

Such dorating was necessary only if the reactivity discrepancy (RD) in Cycle 4 exceeded the RD of 0.009 delta k, which was assumed in the design of Segment 9.

As discussed in Section 3.1, the observed RD toward the end of Cycle 4 did indeed exceed that assumed for design level.

Because of this unexpected RD increase at the time of final 5

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shutdown, the core was operated with the last control rod group in the sequence.

This means that due to diminished hot excesa reactivity, the core operation at 80% of rated power could have been continued for only l-a relatively short period.

Of course this does not mean that power operation had to be terminated.

By operating at powers (at least ini.

tially) somewhat lower than 801, the burnup of Cycle 4 could have been extended considerably as discussed in Section 4.3.

At the time of final shutdown, the core, therefore, effectively achieved the analysed design at end of Cycle 4.

Any subsequent core operation in reality would have been a coasting down mode, i.e., exten-sion of core operation at derated powers.

In this report the burnup of 232 EFFD will be treated synonymous with the end of Cycle 4 (EOC4).

i Consequently, the purpose of this report is to document the core perfor-mance between the middle and the end of Cycle 4, which compliments the I

information provided by Ref. 1.

g There were no F8V Technical Specification requirements to con-duct further temperature defect measurements during the second half c' l

Cycle 4.

These measurement were judged unnecessary because the maximum l

fuel temperatures emperienced during the second half of Cycle 4 were not k

appreciably different from those esperienced during the first half, when temperature defect measurements were performed.

In addition during the second half of Cycle 4, the emit helium temperatures at certain times reached 1450'F, i.e., which are at the design level discussed in the FSV FSAR.

Consequently, the neutronic behavior of the core at 801 power in Cycle 4 should not be appreciably different from the core operation at rated power in the equilibrium cycle which is discussed in the FSAR.

l.

This makes the core performance information in Cycle 4 an important benchmark case for the verification and validation of HTGR methods and

• models.

There were no FSV Technical Specification requirements to conduct further differential and integral control red measurements during second I:

6

910076 Nic half of Cycle 4.

These measurements were judged to be unnecessary because the measurements conducted during the first half of Cycle 4 indicated that the agreement between measured and calculated integral control rod group worths were within the acceptance criteria.

There-fore, control rod groups beyond Group 3A in the Cycle 4 sequence were not measured.

The total reactivity worth of control rod groups up to and including Group 3A (see Table 3-1 of Ref. 1) indicated that Cycle 4 calculations systematically overpeedicted measurement.

Therefore, it i

is not inconceivable that the worth of groups that follow JA in the sequence are also overpredicted although they meet the acceptance cri-j teria. With the total control rod bank worth of over 0.20 delta k and the acceptance criterion of 101, the total effect on the excess reac-tivity can be as high as 0.02 delta k if the deviations are systematic.

This will manifest itself during the monitoring of the core reactivity status as an anomalous change in RD (see Section 3 1).

l The comparison of measured and calculated region peaking factors (RPF) during the second half of Cycle 4 shows that several regions in the core have relatively large (>101) discrepancies. These discrepen-cies, however, are similar in magnitude and ince:e distribution to those observed during the first half of Cycle 4 Due to high reproducibility of both measured and calculated results, one has to consider the possi-bility that the observed large RPF discrepancies are not caused by the uncertainties in the measured or calculated results, but by their incon-sistency.

It appears that, for reasons not known, in some regions what is being measured is not what is being calculated and visa versa. Due to a complicated network of coolant flow paths in the FSV core and the location of temperature measuring instrumentation, the interpretation of

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data and then forming correctly inferred RPFs is very difficult.- The use of " comparison" regions on the North West edge of the core, how-ever, clearly resulted in smaller RPF discrepancies in Cycle 4 (see Section 3.2).

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l The calculation of the axial peaking factors (AFF) indicated that the stability of the axial power distribution is the same or even better than was predicted by the design.

This means that the fuel and lumped burnable poison (LSP) soning techniques developed at General Atomics (GA) are based on a solid foundation. The APFs appear to be fairly insensitive to burnup, and the effect of partially inserted control rods was less severe than anticipated. The stability of axial power distribu-tion in no small way is due to the graded fuel cycle.

The grading of fuel into segments allows the core designer to take corrective mea-sures by means of soning fuel and LSP in each reload segment (see Section 3.3).

To sumanarise, on the basis of information compiled from monitoring core operation during the second half of Cycle 4, it may be concluded that the FSV core performance was well within the requirements of FSV Technical Specifications.

However, unlike the first half of Cycle 4, the second half was characterized by a relatively large and rapid decrease of the hot excess reactivity.

Similar reactivity loss at sero or low power operations was not detected.

Due to the permanent shutdown of the core, the cause of het and cold RDs could not be fully established.

Note, however, that although the observed RD behavior may have resulted in less efficient electricity production, it had abso-lutely no adverse impact on the safety of core operation.

l 2.

POWER HISTORY Pow r operation of the FSV core Jaring the second half of Cycle 4, i.e., between 155 and 232 EFFD, was carried out during the Spring and Sasser of 1939.

Figure 2-1,shows the power history between restart on March 26, 1989 and final shutdown on August 18, 1989.

This power his.

tory, based on one data point per day, is provided for quick and easy reference.

Data Logger records should be examined to provide more detailed information.

In Fig. 2-1, it is shown that from the beginning of June and until the final shutdevn in August the core was operated at I

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eaaeauoava 4 e a e a e n a se e a oea y a JAN TE8 MAR APR MAY -JUN JUL AUG SEP OCT NOV DEC 1989 Fig. 2-1.

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910078 N/C about 801 prm e essentially without interruption. Consequently, this

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period "4 presents the highest core burnup (in terms of power level and the duration of its application) in the history of FSV power operation.

Because of this steady state power operation, various core parameters such as RPF's, RD's, etc., are especially useful for comparing the au-sured and calculated results. Such comparisons are valuable for the validation of methode and verification of'the design.

3.

POWER OPERATION Power operation during the second half of Cycle 4 improved sig-nificantly when compared with the first half (compare Fig. 2 1 of tris report with Fige. 2-1, 2-2, and 2-3 of Ref. 1). Continuous operation c.

1 steady power over relatively long periods should result in the reduction.

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of uncertainties, both measured and calculated, in various core parame.

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ters such as the coolant flow rate, the emit coolant temper.t.ures, the primary-side heat balance, the controi rod group position, etc.

The measureewnt uncertainties adversely affect the validity of comparing the measured ar.d calculated results.

In July and August of 1989 the core l

operation was conducted essentially at a steady state of about 80%

power.

This provides a better than average set of data for comparison.

f Consequently, the following sections concentrate on this period in the core power history, i.e., mostly addressing the last 50 EFFD of Cycle 4 i

As discussed in Section 3.1, just prior to the final shutdown, the monitoring of RDe indicated that the excess reactivity at high power i.

started to decrease faster than predicted.

This means that the core operation could not have been continued at 801 power auch longer, even if the core had not been forced into the final shutdown acde.

However, other RD monitoring at zero or low power indicated that cold excess reactivity c.an be predicted with good accuracy. This means that the l~

converse is also true, i.e., the core gains excess reactivity, f aster l

than predicted, as power level decreases. Therefore, the projected i

extended power operation (coastdown) could have been carried to the i

10

910078 NIC predicted burnup level (see Section 4.3).

Of course, it is impossible to be absolutely certain of this since only the continuation of power operation would have provided such certainty.

Nevertheless, the results discussed in the next section should provide assurance that, at least so far as reactivity is concerned the objectives of the final coastdown would have been achieved.

3.1.

REAC'"IVI7"' DISCREPANCY The RDs calculated with the reference 7-group GAUGE model as a function of burnup in Cycle 4 e.re shown in Fig. 3-1.

In previous cycles the RD of " cold" (0% power) criticalities was 0.0065 e 0.0015 delta k, and the " hoc" (>30%) was'O.090 a 0.001 delta k. The RD at low and inter-mediate power (0% to 30%) has not been accurately determined, primarily because at these power levels the core is in a transient mode.

In such modes the uncertainties of measured data are relatively high, and the accuracy of any steady-state code, like GAUGE, may also be low.

Never-theless, the reactivity status monitoring by Dats-Logger indicates that even at these pcwore the RD lies somewhere between e.he cold and the hot RDs.

The average (i.e., the statistical mean) RD determines the reactiv-icy bias of the calculational model. This bias, because of its insensi-tivity to burnup frem cycle to cycle, can be discounted when reactivity related predictions are made.

Therefore, the reactivity uncertainty of predictions is essentially the same as the standard deviation of RD.

Therefore, only RDe that lay outside the range of established statistics significantly indicate the anomalous behavior of reactivity.

The change in the calculational bias of the GAUGE model between cold and hot criticalities, shown in Fig. 3-1, is not well understood.

Since the reactivity biases have been fairly invariant with burnup, the determination of cause for this difference he.s only acadetric signifi-cance.

In practice, by discounting, appropriate biases, the prediction 11

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'v of cold and hot criticalities can be done with essentially the-same degree of accuracy.

The RD of cold criticalities during Cycle 4 is in excellent agree-

, ment with previous cycles, except for a relatively short burnup period at 120-EyPD.

The unexpected rapid increase et.ald RDs at this-burnup level was discussed in the previous core performance report (Ref.

1)..

The-cause of this anomaly could not be explained, by either changes in--

% w e operation..or. calculational errors. The rapid change in cold-RDs a yuite disturbing, yortunately, subsequent core: operation indi-cated that the RD of cold criticalities were again in the expected range.

This provided assurance that neither the calculations; model nor

. physical changes in the core geometry and/or materials were the probable.

causes of-the. observed anomaly.

The acet likely explanation is that the-measured data that' feeds into the calculational model somehow became temporarily faulty. The reason this occurred has not been determined.

As part of the lessons learned program for future generations of.the HTOR, it appears to be prudent to develop.a special trending procedure which will alert the core operators to any sudden systematic changes 4

in the core reactivity. With such a procedure in place, the causes of "w

,m reactivity changes could be found-with greater ease,.and corrective sea-sures could be taken before the changes become-poteatially serious.- It Y

is'very difficult to establish the cause(s).for anocalous discrepancy s

between measured and calculated data if the investigation depends solely

'a on data retrieval sad /or operator's recollection of past events.

J At the beginning of Cycle 4,-the RD of hot'criticalities started

'Y outside the espected reactivity range. With burnup:it slowly decreased.-

- <a and at the middle of the cycle it was within range.

This behavior is ya.~

explained (Ref. 1) by prolonged core operation at relatively low powers, which was interspaced with frequent shutdowns it. the first quarter of g

Cycle 4 Toward the middle of the cycle, core power operation stabil-ised at relatively high power with fewer interruptions. As was stated b

before, under these core conditions, the accuracy of both measurements i

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and calculations >fs expected to improve, resulting in the. normalization-of RD behavior toward the middle of cycle.

Note that the improvement of p

~ hot RDs with burnup in Cycle 4 was accompanied with anomalous behavior l.

of cold RDs at the middle of Cycle 4.

The relatively large difference b'etween the cold and hot RDs at the beginning of the cycle essentially

-disappeared by the middle of the cycle.

g) 4 Based on the behavior of hot criticalities during.the first half.of the cycle, it was even more pusaling to observe a steady increase of hot J

RDs during the second half of the cycle.

The RD trend at the end of

' cycle was unprecedented, and yet this burnup period was characterized by prolonged and uninterrupted operation at about 801 power.

Because of this stability there were no shutdowns and, therefore, the behavior of m

cold criticalities were not available. Without the benefit of such-measured data, it is impossible to be sure that there were no physical L'

changes to the core which would manifest themselves in the observed anomalous behavior. However, based on the experience gained from pre-vious. cycles, as well as from the first' half of Cycle 4, it is difficult to accept that the explanation lies in the faulty measured or calculated u.

results.

It is more appropriate to question in what way the burnup

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between 190 and 230 EFFD in Cycle 4 is different fr a other burnup

. g.,

periods'in that cycle or the previous cycles.This period, unlike any other period, was characterised by the prolonged use of partially inserted out-of-sequence control rods.- The rods in Regions 13, 22, and l

33 were kept at 165 to 170 in. of withdrawal'to achieve a more optimal balance of steen generator modules.

The partial insertion of up to six

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rods is allowed.by the FSV Technical Specifications.

This was intended W

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to provide-a fast and easy way to control excessive RPFs while the core f [.

is being reorificed. The long-term use of such rods was not considered m P l'

in Segment 9 SAR (Ref. 3), and the reactivity worth at such insertions 7 '

has not been measured, either in Cycle 4 or in the previous cycles.

The 4,

GAUGE model is not suitable for calculating rod insertions because the m

model'uses the measured integral rod worth curves for representation of o

partial rod insertions. As was stated above there are no measured data 4

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910078 NlC for control rod withdrawals of 150 to 180 in.

The three-dimensional calculational models at GA and PSC were not used and may.not be suitable to calculate such~ rod insertions.

Consequently, the reactivity uncer-tainty associated with partial rod insertions may be significant (sev.

eral 0.001 delta k). Therefore, it is not unexpected that the RD will increase as the number of out-of-sequence partially inserted rods inc rease'.. Still, the trend of RD toward the end of Cycle 4 was unex.

p pected.

Since the same out-of-sequence rods were inserted into the same regions and to the same-position, the RD change was expected to occur.

g stepwise rather than in the manner that was observed.

The reversal of the RD trend right before the final shutdown'is only 0.001 delta k, which is not statistically,significant. : Continued-core operation would have established for certain whether the observed RD trend was temporary and/or reversible.. As a speculation..however, it.

should be noted that sometime the RD trends are associated with the dis-crepancy between measured and calculated control rod group worths...Such trend was observed at low powers when Group 2B was in.use.

.(The. worth of this group was overpredicted more than that of any other group.)

Group 3B'was in use toward the end of the cycle during core operation at 80% power..As.was mentioned before, the worth of this group was not

' measured.

Therefore, it is not. inconceivable that its worth could have been overpredicted by as much as OiOO2. delta k (such overprediction-would still have met'the acceptance-criteria).

The reversal of the RD ut trend-right before the final' shutdown then can be explained by.the-fact that Group 3B was' fully withdrawn and the core operation started to use the last group (which in its turn could have been somewhat underpre-i dicted).- As was the. case for anomalous. behavior of cold RDs, a trending procedure should serve well,the future generation of HTGRs by providing a means to detect and investigate unexpected trends-at the earliest possible time.

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3.2.

REGION PEAKING FACTORS The measured-and calculated (with 7-group GAUGE model) RPFs as a function of=burnup during the second half of Cycle 4 are given in Appen-o dix A.

The results are very similar to those reported for the first

-t half of Cycle 4 (Ref. 1).

Results indicate that most RPFs can be pre-dicted with accuracy better than *10%. As an exception, the discrepan-cies in a few regions (e.g., 2, 12, and 31) are consistently over 10%

irrespective of the core burnup or core power level.

The cause of these isolated discrepancies.is not known, especially since they do not cor-relate 1to fuel 1 age or control rod configuration, which are the two core

.i parameters =that have the greatest effect on RPFs.

The magnitude of these randomly scattered RPF discrepancies, however, is similar to that

-observed during previous cycles.

l

'l core, operation during the second half of Cycle 4 was different from-the first halflLn one other important aspect.. As was mentioned before, a

in order to achieve a more favorable exic gas temperature distribution, and to be able toLbalance steam generators better, tho' core-operators.

used partially inserted control rods in-Regions 12, 22', and;33. The:use i

of these out-of-sequence rods was started at about 190 ETPD and con-

.n tinued to the end of Cycle 4.

As was the case with the ' worth of. these 1

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rods'(see Section 3.1), their impact on the radial and axial power dis-t.iousions after prolonged burnup periods.was.not assessed in the Seg-ment 9 SJR. Although FSAR calculations indicated that the impact of out-of-sequence partially inserted rode would be minimal, a slight i

increase in che power peaking and reactivity uncertainties is unavoid-able.- The effect of such rods on reactivity was discussed in the pre -

vious section.

In this and the next sections their possible ef fect on

.che power distribution will be discussed.

As'results in Appendix A indicate, the RPFs of the inner core regions (1 through 19) are generally underpredicted, while the RPFs of the outer core regions (20 through 37) are generally overpredicted.

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910076 N/C

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These inaccuracies are caused by the well-known calculational bias of the 7-group GAUGE model. At the beginning of core operation, the model was adjusted to compensate for this bias.

The adjustment was based on a comparison with a multigroup, fine-meshed method.. The adjustment worked quite well-for the first couple of cycles, but'became less and less effective with further core burnup.

PSC authorized, in 1988 (PG-1767),

a study to determine a means for enhancing the accuracy of RPF predie-tions.

A simple modification of the GAUGE model was found-(GP-3209).

which resulted in a significant improvement of RPF accuracy.

However, by the time this study was completed,.the burnup of Cycle 4 was substan-tial.

It was, threfore, decided that modificatica of the model should-l be. delayed until the next' cycle or next phase of core operation, i.e.,-

the power coastdown.

It is interesting to: note that the RPF discrepancies, calculate'd by PSC with the three dimensional nodal model, FAN 3D, as a part of their fuel accountability. calculations, show better accuracy than the refer-Lence CAUGE model.. Note, that the FAN 3D was approved by the NRC for fuel accountability calculations only. 'As shown in Figs. 3-2 and 3-3-the:RPF dircrepancies in the inner.part of_the core, especially in Regions 2.and 12, are much lower,cand the RPF' discrepancies in the outer part of the-core are also lower, except for Regions 22 and 31.

Since the FAN 3D model has not been accepted by the NRC for power distribution calcula,

tions, all further discussion of RPFs will be confined to the-_ GAUGE results.

It is, howiever, regrettable that such an excellent analytical-model' as FAN 3D hass been barred-f rom generating - such an important-parame-ter for core operation as RPFs in comparison regions.

The RPF discrepancies in Regions 20, 21, 22, and 23, which are associated with the same thermocouple " string," were significant--

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throughout the cycle. Note that the low discrepancy of Region 20 is caused-by the use.of calculated, rather than measured, exit gas temper-ature for derivation of its " measured" RPF.

The fuel temperature mes-re::::r.;. obtained during the Post Irradiation Examination of FTE-2 17

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U9 gg L32 L30 u0 UO 02 02 G71 0.99 [ 0 0.93 l

141.

p 0.98 LO1 0.72 0.09 0.%

0.67 38 120 us u9 LOS

,4 FAN 3D RPF Measured RPF

.k k = 1.00 Fig. 3-2.

666 MW(t) (.79 % ) 194.8 EFPD in Cycle 4 18

- - -. ~

'910079 N/C 3,

'10

+8

-10

+3

+3

+2

-16 3

+9 f

g O

-15

(

'N

-1 5

4

+I

~I

+5

+6 j

-3 9

o

- -9

-3 1

-1.

~'

0 M

-4

\\

-5

-5 c

Control Rod' Pattern

7. Discrepancy

[(Measured /Calcu, lated)-1] x 100-~

j M

M OM ON gy g

-0.91

.u2 038 g

U2.;101 0.85

.O M OS9 U4 Oa0-ON

'US 03 082 1.52 ; g 0'

W.

0.

u2 L58 120 tg 0.75 t,

O0 0.75 0.94 O

7 0.72 W

02 Os.025' '

• t3g 035 Om 056 t9. ago

' 1,71 156-tot 101 gg 06'?

GM 0.64 0m Oa5 124 120 us u4 FAN 3D RPF Measured RPF i

k = 1.00 Fig. 3-3.

666 MW(t) (791) 232.0 EFPD in Cycle 4

.s,

19

,1

c 910078 N/C.

p g,

(Ref.-4)" indicated that indeed the measured exit has temperature is too low in Region 22.

Since Region 22 is-neutronically no different from p

Regions 20, 21, and 23, there=is a compelling reason to believe that'the

[

measurements for these regions are faulty.

The reason for measurement bias 'for this particular thermocouple string, however, is not known. -

The thermocouple recalibrated on several occasions without finding any-.

thing wrong with it.

The faulty region exit gas temperature reading is

~

probably due to the mixing of intraregion coolant with an unknown quant'-

ity of bypass coolant flow.

Such mixing tends to lower the temperature measurements and, therefore, is erroneously interpreted as.systemati-cally lower RPFs.

The comparison of. calculated and measured.RPFs, during Cycle 4 burnup at various power levels, indicates that the FSV Technical Speci-fication requirements were met with more than adequate margin.

Tho'RPF

j patterns ' observed during the second half of Cycle 4 are similar to those E'

observed during the first half, or those which'were calculated in sup--

port of the Segment 9 SAR. There is no evidence that the use of par-tially; inserted,out-of-sequence control rods adversely affected the radial power distribution. On the basis of this observation'it may be coneludeds chat RPFs_during, subsequent core operation (e.g., power coast-U down)'also would have meet the requirements.

q f

3.3.

AXIAL PEAKING FACTORS lu Direct measurements of axial flux = distribution (and by interfer-ence of asial~ power distribution) were performed in the FSV core during t

Startup Test A-7 atrthe-beginning of Cycle 1 (Ref. 5).

The core lacks-necessary-instrumentation to conduct such measurements af ter core irra,

,T

'diation has started. Consequently, the axial power distribution can be, either indirectly deduced from the differential control rod worth mes-l surements, or it can be generated with the three-dimensional model as a part of fuel accountab.ility calculations.

a 5f.

20 r

4l v.

-910078 N/0 LThe 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 e ch calendar year.

The FA calculations for the first half of cycle-4 were done with the GATT model (Ref. 7) at GA.

Upon validation of the FAN 3D model (Ref. 8), which was developed on the basis of-DIF 30 nodal code, the FA calculation for the second half of. Cycle 4 was performed at PSC with the FAN 3D model. Note, however, that unlike the GA's GATT.

-model, the FAN 3D model validation was restricted by the NRC to-the fuel accountability, i.e.,'it did not extend to the radial and axisi power distributions.

To verify their FA calculations, as well as the= accuracy of the power distribution calculated with the FAN 3D model, the PSC authorised GA to conduct an independent. review.This independent review

,.4 was performed with the GATT model (Ref. 9) and results indicated that.

the AFF's calculated with the FAN 3D model are:in very good agreement with the GATT results.

This review was confined to one time point in u

Cycle 14 b'urnup, however, it is reasonable to assume that the agreement between FAN 3D and GATT'results would have also been good for all other:

1 time points. Consequently, the discussions below are based on the assumption-that the AFF's calculated with the FAN 3D model are valid.

y

.j

-The axial power distributions as a function of core location and d

burnup (provided by PSC in EPG-0165) are too extensive to be presented

- k

- q'c in this document.

Consequently, the discussion here'will focus on:

• 7 (1) the' power frection generated.in the top half of any region and

.i Y

i (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 FSAA indicated that power. fractions in the range of 0.60:co I

0.55 assure the optimum. axial distribution of fuel temperatures in unrodded and fully rodded regions.

The latter parameter is present because of the requirements provided by the-basis of FSV Technical k

Specification LCO 4.1.3.

Among other requirements, this specification basis stipulates that the APF of unrodded or fully redded regions shall i

21 i

sI'

[!i

910078 N/C m

I not exceed 0 90, and.the APF of partially rodded regions shall not

- exceed 1.23.

The. power fraction in the top fuel zone (top half) of each reload region as a function of burnup-in Cycle 4 is given in Table-3-1.

At a burnup of 195-ErPD the fractions are within'the desired range in all regions except Regions 11, 9, 13, and 17.

These regions concein par-tially inserted rods, and are not expected or required to maintain the optimum axial pow.

distribution determined for unrodded regions. The

.burnup of 195 EFPD was. selected for presentation in this report because the shim group was' half;way out of the core.

FSV experience has-shown that the maximum perturbation of the axial power distribution occurs at-

- such's shim group position.

Consequently, it may be concluded that the power-fraction in the top half of the core, as a function of Group 35 withdrawal with burnup, was no less than that given in Table 3 1.

At.burnup of 232 EFpD (EDC4), Group 35 was fully withdrawn and the next-in-sequence (and last), Group 3D, was just starting to be with-drawn.

So at this burnup, the core did not contain, at least in the calculational model, any other partially withdrswn rods, axcept for.the regulating rod in Region 1.

Actually at this. time, as was mentioned before, there were-several

'out-of-sequence rodo partially inserted into the core.

Due to the:dif-ficulty of introducing changes to the model, and other practical consid-erations, the out-of-sequence rods were not modeled.

The effect of

. neglecting these< rods on the axial power distribution is discussed

.below.

Due to the absence of partially inserted rods in the calculational model as expected the axial power distribution tilted toward the core top, causing the-power fraction in the top half to exceed somewhat the upper limit of the optimum range of 0.60 K.

Unlike power tilting toward the core bottoa, the high power fraction in the core top should not 1

22

. j-...,

010078 NIC-TABLE 3 1 POWER FRACTION IN TOP FUEL ZONE--

Region 195'EFPD 232 EFPD.

Region-195 EFPD 232 EFPD 1

0.49 0.56 20<

0.54

-0.61 2

0.53

-0.62 21 0.58 0.66 3-0.56 0.64 22-0.58 0.66--

'4 0.57

-0.65 23 0.55 0.62 5

0.58-0.67 24 0.56 0.'66 6

0.53 0.62 -

25 0.60 0.60 7

0.53-0.62 26 0.54 0.62 8

0.58 0.66 27 0.52 0.66 9

0.49 0.63 28 0.57 0.66~

10 0.59 0.67 29 0.60 0.66' 11-0.57 0.63 30 0.56 0.62i 12-0.54 0.63 31 0.55 0.61

'13 0.51 0.65 32 0.60-0.66-14

.0.55 0.63 33 0.58 0.66

+

15 0.60 0.66'-

34 0.54 0.62 16 0.55.

,0.63 35' O.58 0.64

'17 0.52 0.67

'36

0. 61..

0.66-18 0.57 0.65 37 0.56 0.~ 61

'19

0. 56 --

0.63 Group 3D-Fully.in Fully in Core average 0.56

0.64 Group 38 Half in Fully out d

23 i,

' 47 ;

J L

910078 H/C i

l l-result in higher fuel temperatures and will nt,'have a potentially I

adverse impact on the fuel performance.

The coolant enters the active t

t core from the top, which causes the average fuel temperature in the top f

half to be substantially lower than in the bottom half of any region.

u s'

Furthermore, the axial power tilting to the core top las a relatively I

short period, since the withdrawal of a' shim group with further burnup

'will inevitably' result in the flattening of axial power distribution.-

In addition to these mechanisms for moderating excessive tilts to the y

core top, the E0C 4 was characterised by the partial insertion of rods-in Regions 12, 22, and 33.

Such insertions have a small-but definite effect on controlling power tilts to'the top.

This means the data given in Table'3-1 for EOC 4 core condition are most likely systematically overestimated.

In the case of partially inserted out-of-sequence rods, theaxial' power. distribution should improve somewhat, putting it closer.

to the optimum range.

I 9i The APF in the bottom fuel layer in each reload region as a frac-tion of burnup during Cycle 4 is given in Table 3-2..The results indi-cate that the-requirements-of the-Technical l$pecifications were met with h

h a' wide. margin.

The nonrepresentation of.out-of-sequence partially u

o

. inserted rods should have very little affect on the axial power dis -

j tribution. At the EOC4 these rods could produce some power flatten-

)

ing, but the margin between the allowable 0.90 and the calculated core j

average of 0.53.is so large that a few percent increase imposed ~upon i

0.53 will not change the above conclusion.. At other'burnups.(e.g., at 195'EFFD) the partial insertion of several rode up to two feet produces-

.very small' perturbational-effects to the axial power distribution as compared with three-rode-inserted halfway into the core.

To summarise this section. the use of out-vi-sequence partially J

P inserted rods toward the end of Cycle 4 appears to haves (1)'some s

effect on increasing the RD, (2) no detectable effect on the RPF dis-crepancies, and (3) a small, but beneficial effect, on the APFs.

'l 24 2 ---

910078 MIC TABLE 3-2

'AIIAL PEAKING FACTORS IN BOTTOM ELEMENTS Region 195 EFFD 232 EFPD Region

'195 ETPD 232 ETPD i

1 0.79' O.63 20 0.73 0.59 2

0.72 0.55 21 0.68.

0 52 3

0.66 0.51 22 0.60 0.46 4

0.68 0.51 23 0.68 0.55 5

0.68 0.50 24 0.69 0.58 6

.0.74 0.56 25 0.62 0 51 7

0.73 0.56 26 0.72

0 58 8

0.67' O.51 27 0.76 0.56 9

0.78 0.53 28 0.69 0.51 10 0.64-0.49 29 0.58 0.46

'11-0.66 0.53 30 0.70 0.56 12 0.72 0.54 31 0.71 0.58 -

13 0.74 0.48 32 0.60-0.48 14 0.69 0.52 33 0.60 0.46 15

~0.64 0.50 34 0.72 0.55 16 0.69 0.53 35 0.66

-0.54

'17 0.75 0.49 36 0.61 0.51 18 0.64 0.49 37 0.71 0 59 19 0.68 0.54 Group 3D.

Fully'in Fully in; Core average 0.69'

.0.53 Group 3B Half in Fully out

+

I i

25 h%

,j i ;

910078 N/C.

F 4.

FUEL MANAGEMENT 4.1.

FUEL ACCOUNTABILIT?

The total fuel loading o? the FSV_ core at the beginning of Cycle 4 is_given in Table 4-1.

It is bcsed on fuel accountability (FA) informa-

' tion produced at the end of Cycle 3 for the irradiated segments as well as for the fresh Segment 9.

Both the two-dimensional (2-D) and three-dimensional (3-D) models of the FSV sore were normalized to these data (Ref. 1).

Such_ periodic normalizatior. is needed to keep the_2-D_model from deviating too much from the reference 3-D FA Model (GATT in the first half of Cycle 4, and FAN 3D in the second).

The 3-D models were developed to monitor the FA of each of the 1482 fuel elements in.

the core.

The volume of information they produce is very large and detailed, so it is typically stored on a magnetic tape, or in a computer mass-storage file.' Since the differences in the FA of relatively;1arge portions of the core (e.g., the reload sessents) between the GAUGE and 3-D models are relatively small for the purposes of this report it is sufficient to present only the GAUGE results..

The core fuel icading at the end of Cycle 4 (232 ETPD) is_given in Table 4-2. 'Although the inventory of U-235 decreased due to burnup by about 150 kg, the inventory of U-233 increased by about 59 kg,;resulting in a 50% increase in its enrichment. 'Since, U-233 has botter neutronic characteristics than U-235, its buildup testifies to the capabilities of the FSV core to act as an effective converter.

The building of U-233 also substantially increased the core potential for prolonged power.

coast down (see Section 4.3).

The total fuel loading 'of each segment in the core at the EOC4 is given in Table 4-3.

The results indicate that.the maximum burnup was about 46,000 MWD / tonne, which is substantially lower than 100,000 MWD /

tonne limit established for fuel burnup by the FSV Technical Specifica-tions.

A large margin in burnup also supports the contention that the core potential for prolonged power coast down was indeed substantial.

26 t

.., m ? ;

• 910078 N/C L

l; TABLE 4-1 LOADINGS AT BEGINNING OF CYCL'. 4-Th-232 14,241.530 kg U-233(a) 184.355 kg t-U-235 559.883 kg.

Uranium 896.738 kg-U-233 enrichment 20.56%

Net enrichment 83.00%.

(*) Includes full decay of Pa-233.

TABLE 4-2

-TOTAL CORE HEAVT METAL LOADINGS FOR PERIOD ENDING AUGUST-18, 1989 CYCLE 4 BURNUP 232.0 EFPD Th-232 14,097.98 kg Pa-231 56.05 g U-232 52.27 g

'U-233(a) 243,435.53 g U-234 2,8401.13 g U-235 403,810.73 g,

p U-236 102,'165.27 g

--j U-238 53,879.08.g Pu-238-1,624.89 g

-w:

Pu-239(b) 1,400.41 g Pu-240 507.'17 g Pu *, 1 448.02'g

'iir a;

-u-242 263.88 g Total uranium 831.744 kg S

U-233 enrichment 29.27 %

c-l U-235 enrichment 48.55 %

Net enrichment 77.82 %

ppa U-232 62.84 (a) Includes full decay of Pa-233.

(b) Includes full decay of Np-239.

m 27 i

=-;.-

-y=:

u; 2 -

TABLE 4-3

. CORE HEAVY METAL 1. OAT'INGS SY SEQ 9ENT-FOR PERIOD ENDING AUGUST 18,'1989 CYCLE 4 80RNUP-232.0-EFPD Segment 7 Segment'8 Segment 9 Segment 4 Segment 5.

Segment 6 Th-232 2,201.92 kg 2,230.33 kg 2,250.69 kg 2,318.09 kg.

2,781.53 kg 2.315.43 kg Pa-231 9.71 g 8.97 g 5.69 g 9.83 g 11.82 g.

9.76 g U-232 8.84 g 5.81 g 1.66 g 11.16 g 13.58 g 11.21 g U-233(a) 41,242.56 g 35.838.59 g 21,481.47 g 45,154.15 g 54,577.43 g 45.141.40 g U-234 4,560.63 g 3,393.81 g 2,294.73 g.

5,649.99 g 6,845.31 g 5,656.67 g U-235 62,850.48 g 86,267.89 g 137,336.27 g 37,999.97 g 41,259.52 g 38,096.72 g U-236 19,519.63 g 17,194.85 g 10,666.57 g 17,578.22 g.

19,608.95 g 17,597.08 g U-238 9,429.66 g 10.095.82 g 11.100.47.g 7.463.28 g 8,314.64 g 7,475.23 g Pu-238 292.48 g 147.37 g 21.65 g 369.63 g 424.22 g 369.53 g-Pu-239(b)-

254.36 g 267.82 g 159.21 g 199.27 g 222.42 g 197.33 g Pu-240

'98.03 g.

96.30 g 58.40 g 81.29 g 92.53 g.

81.53 g Pu-241 9,618.00 g 80.52 g 28.76 g-77.30 g 87.84 g 77.42 g Pu-242

'50.33 g 27.10 g 3.18 g 58.60 g 65.97 g 58.72 g MWDfeetric ton 44,616.38 35,068.34 17.641.48 45,956.60 44,824.76

.46,075.99

(*) Includes full decay of Pa-233.

(b) Includes full decay of Np-239.

8 Eo

91007B N/c'

-4.2.._ FUEL PARTICLE BURNUP AND EXPOSURE The maximum burnup, in terms of fissions per initial metal atom.

(FIMA),-for fissile and fertile particles at the EOC4 is given in Table 4 4.

Because the core location and'the length of irradiation affect burnup, the FIMAs are givac as a function of active core layer and segment.. Since Segment 4, 5, and 6 were inserted into the initial core at the same time, their burnup is very similar. Therefore, there-is no need to supply individual results for these segments.

As the results indicate, the maximon FIMA of fissile particles is less than 16%, which in significantly lower than the 23% projected for the end-of-life of Segment 9 (Ref. 2).

The maximum fast flux fluence (exposure) of-any element in the core at the EOC4'is 4.3 E + 21 nyt, and the average exposure of any element is 2.4E + 21 nyt.

The' maximum exposure is a factor of two lower than the 8.4E + 21 nyt projected for the end-of-life of Segment'9-(Ref._2).

4.3.

POWER C0ASTO N Prior to the occurrence of problems that-led to the final shut-down on August 18, 1989, core operation'was expected to: continue to the-designed EOC4, i.e., to the burnup of 300 EFPD.

Subsequent to the EOC4 operat!on. st was intended to continue to operate at powers less than 80% until either lack of reactivity or poor economics 1 forced tho' final

.n shutdown. A study was undertaken in support of the coastdown SAR (Ref. 4) to show that the core coastdown operation would be-in com-pliance:with the FSV Technical Specification requirements.

The time-intervata and the corresponding power -levels for-the projected coastdown are:given in Table 4-5.

The data in this table are for illustration purpose only, since the actual coastdown would have been conducted by decreasing power smoothly.rather than stepwise.

w 29 a1

(

, ~.

f:

910078 N/c TABLE 4-4

. MAXIMUM PARTICLE SURNUP.(1 TIMA) FOR STANDARD BLOCKS AT 232-ETPD 1

Particle core Layer Segments-4 to-6 Segment 7 Segment-8 Segment 9 Fiscile 4

11.5-9.7 8.3 4.3-5 14.0.

12.7 10.8 5.8 6-15.2-13.5 11.4 6.0

=7-15.9 15.4 12.6 6.7 8

14.6 14.2 11.5 5.7-9 11.8 11.4 9.0 4.2-

' Fertile:

4 1.4 0.7 0.5 0.1 '

.i 5

2.5 1.5 0.9 0.2 6

3.1 1.9 1.2

.0.3 7

3.3 2.3 1.4-0.3-8-

2.5 1.8 1.1 0.2 9

1.3 1.0 0.6 0.1

'J' ETPD-890 716 527 232 U

30

!l'-

t

'to 910078 N/C

?

1 TABLE 4-5 TIME AT POWER.

DURING COASTDOWN Power

-y Day

(%)

EFPD 75 80 60

'1 43 70 30 i

58 60 35 60 50

'30 88 40 35 j

100-30 30-424 220 1

,1

'l l

4 1

+

N(~

-1

^l

.l

\\

I i

.t

', f o

j I

.y 31 3

.m 1

i$,

l.;. g.

%l5IO

. r.

Lw 910078 N/C i

'In the previous sections, it was pointed out that there are no technical reasons why the coastdown could not be carried out.

More than adequate margins exist for the-importsnt core parameters, such as excess reactivity, RPFs, APFs,'FIMAs, etc.

Since coastdown was not conducted there' is no certainty tt at the coastdown operation would have performed as predicted. However calculations, as.well as core operating experi-ence, indicate that it is quite probable that over 200 EFFD could have-i been.used for' electricity production in addition to the 300 EFPD burnup designed for Cycle 4.

5.

CONCLUSIONS

-On the basis of results provided.in this report, it may be con-s cluded that the core opsrated during the second half of Cycle 4 well within'theienvelope described in the Segment 9 SAR within the and requirements of the FSV Technical Specifications.

The hot excess reac-l a

tivity decreased faster than predicted, but there is no evidence that the design b'urnup of 300 EFPD could not have been achieved in CycleL4 1

-Furthermore, if premature final shutdown on-August 18, 1989 had not-occurred, the core-would have been capable of achieving the objectives of the power coastdown.

6.-

REFERENCES-1 1.

Malakhof, V., "FSV Midcycle 4 Core Performance," GA Document-909750.

Issue N/C, April 5. 1989.

'2.

Malakhof, V., et al., " Segment 9 Design Document (Core' Physics),"'GA-Document 906500,- Issue A, March 10, 1983.

- i, 3.

" Safety Analysis Report for Fuel Reload 3 (Segment 9 - Cycle 4)," GA s

Report GA-C17128, May 1, 1983.

4.

" Cycle 4 Coastdown Design Support Document (Core Physics)," GA

-f Document 909768, Issue N/C, February 22, 1989.

5.

Brown,'J., et al., " Neutron Flux Distribution Measurements in the FSV. Initial Core (Results cf TSV Startup Test A-7)," GA Report GA-A13176/UC-77 February 1975.

32 t

91007B N/C c.

6.

McCord, F., " Destructive-Examination of TSV Fuel-Test Element FTE-2," GA Document-908909, Issue N/C, July 1985.

' +

7.

Wagner, W.

R., et al.. "GATT'- A Three-Dimensional.Few-Group Neutron Diffusion Theory Program for Hexagonal-Z Mesh," GA' Report GA-8547, January-1969.

8.

Rucker, R.,

et al.,." Validation of FAN 3D Model for FSV Fuel Account-

' ability Calculations,"'CA Report 909436 Issue N/C January 1988.

9.

" Independent Review of FAN 30 Results," GP-3250, December 21, 1988.

i, e

e

('_a s i

\\ '; '

li; * '

f.

5 33

.i i

)

3

i-,-

t'

.:p:

L 910078 N/C e.

7, y-

-p

.6 l:

4:

AFPENDIX A-RPF DISCREPANCIES IN CYCLE 4 (155'To 232 EFPD) u.

3:

e t

'( l' n

i 4

1 m

f_

l 34

• i:

3) r

_~t j

~

9 '.00 ? B '4 t c 4

1

-2

-9

-2

-10

-5 g

G.

+4 4

f g

)

4

-9

+11

/

4 o

0

+4

+10

u 0

\\

'13

+3 e.

-w

~

~9

~4

+17

-i

+9

~

l

-3

+1 j

-4

-1

+4 1

Control Rod Pattern

7. Discrepancy 1

[(Measured / Calculated)-1] x 1.00 J

j 083 0 77

- 081 0.70 gg gg q

M 1.03

.1.10 0.93 1.04 OM 085 0'69 0.73 OM 0.88 1

OB1 1.10 0lM OM ON. 082 0.84' 084 l

\\.69 0B5 02 085 0.

1 O.78

. g,75 _

036 y

1.03 OW s

(

0.93 0'95

-}

y, 2

OM M-OB6 j

u 1,4g 084 pg i

L86 1.71 -( 0.95 t34 g,33 swy 0.9d. 0.W ON 0.72

/ 0.84

,1.39 j b

124 9

N~h 036 0.78 0.83 e-f 0.94 [ 033 [

030 q

L31 0.95 x.

GAUGE RPF Measured RPF k = 1.0110

,_.f Fig. 1.

75 MW(t) (9%) 155.7 ETPD in Cycle 4

, j J

1{-

j' 3 } r

s,.

f 910078 N/C

-4

~4

,) e $

+8

+3

_g

_14 e

a

'V

• I4 s

+10

~0/

-7

+12 Q

+2 h A

'(

+3 r~

-3 g.

+9

+t7

+2

_tg

_4 C. $

+

0

+7

_13

+5

-10

- 11

-10 Control Rod Pattern

% Discrepancy j--

[(Measured / Calculated)-1] x 100 T.

~ 0.M ~ 0.89 0.73 0.72 g

g S

129 U6 139 1.00 4

- 0.68 0.72 U2 LW LT.

L52

/

E e

s 0.79 OM O

OM 129['

I L10 US L38 s

g.

62 0

0.76 0.83 OM {

08 0.61 0.75 u8

. 086 E

[I~39 0.88 I'

db ~~

-- P U5 1.17 (

0.M -

-0 000

~N~% 0

(-

100 0.00

1. ' IM /. 0,y )g.96 058 L(T/

\\

0.88 0.83 0.77 0.78-O.70 2 Lee 129 1.44 e$

GAUGE RPF Measured RPF J

k = 1.0118

)

Fig. 2.

240 MW(t) (28%) 157.9 EFPD in Cycle 4 36 J

.}

' l' )

9

]

910078 N/C "h

+7 '

y

+3

_a

_tg 4

J

+1s O

+4

-14

_a

+9 0

+3

+u 0

+10

+17 4

+8 d

-6I

-20 g$

+21 -

1 4

+5_)

-12

-B Control Rod Pattern 7.-Discrepancy

[(Measured / Calculated)-1]'x 100 0.75 0.67 0.75 0.71 gg g

080 134 1.14 L43 0.72 0.77' 088 a73-tse t04 3

1.12 L80 t03 L48 OM 0.M 0.85 y

O L40 129.

1.77

.40 g

g 0.79 0.81 0

OM O

0.75 L10 0.88-

- 00 g

ON 0.70 OM 0

056 0

LO6 1.06 L-OSa 0.60 031

'-4.63, )

0 t03 127 L43 L17 GAUGE RPF Measured RPF k ' = - 1.0120 Fig. 3.

359 W(t) (43%) 160.4 ETPD in Cycle 4 I

37 l

910078 NIC l

~4

~e

-2 3*g

-12

-2 4

.e ?

+e h

+16 Y

-1

-9 4

+5-

"' x __, %

p

,y

+5

-3

-2

,g O

+1

-18

-5 g

g

+21

-1 g

-5

-13

-12 Control Rod Pattern

.[(Measured /Calcule.t.ed)-1] x.100

7. Discrepa'.tcy 0.3 ON 0.73 0.66 g

g 129 0.M 0E0 0.68

.aM gg gg Lil L78-

' LW g3 084 O.?B OM E

1.48.'

L16' L46 129 082 086 OM 000 E

W 0.75 06

_OBO.

gg OR 0.M.

0 0.70 ) 0 058 g

1.06 0.88 - #

0.63 0.78 E*

L62 129 1.41 L14 GAUGE RPF Measured-RPF-k = 1.0119 Fig. 4.

235 MW(t) (28%) 161.6 ETPD in Cycle 4 38 a

k e10078 N/C i

j

~4

-7

-1

~0 i

+10 g

-4

- 17 V

G

+16 s

4

+t

+11

+5

+3 g

+18

+7

~

4 q.

g

+1 g@

(~

+1

~5 ' 4

-4 1

-13

-6 Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1] x 100 n

0.79 0.73 0.76 0.68 g

g 127 LIO LI7 088 097 0.70 0.73 LOO L73 1D5 L4 0

s 0.77 084 0.84 0.9E>

0.89

-s tg g

g s-g 0

'040 N

~

0.75 0.70 057 0.75

( 64 f I

124 LOS l4 OM W

'- /

O.2 0

0.67 081 l ' I LOS 106

.%f ON 085 0.64 081. #

'N 030

.c\\.73 L gT 0.7e 0

V y

GAUGE RPF Measured RPF k = 1.0123 Fig. $.

425 MW(t) (503) 165.2 EfPD in Cycle 4 39 i

1

~

I 910076 NIC i

i

-u4-10s;

,s

-10 4

O

+9

~

+1e f

-5

+14 e

0

+4

+12 s

'II 5

+7 4

+!8

+4

+u

-3

-7_)

+4

_a

_e Control Rod Pattern

,. Discrepancy

[(Measured / Calculated)-1] x 100 02 0.84 Q76 0.76 g

ul 11 3 -

L22 LW LOS gg g

0.71 0.73 1.03 L80 0.98 OM 0.M 0.77 0.80 d'

0%

081 088 0

0.95 0.

gg g

g d

E 0.73 E

0.76 fg 0.95 0

081 /

0.73 L19 d

g LO9 f 1.31

,D LO1 s

088 O00

~

.92 (

0.81 1.04 OS6

/0 0.85

' 8 k100h-to' as0 V

t00 0.?9 0.72 0.77 0.67 >

03 0

L34 toe 123 LO3-GAUGE RPF Measured RPF k = 1.133 i

Fig. 6.

575 MW(t) (681) 172.7 EFFD in Cycle 4 i

L 40

f

,}l 910078 N/C

-5

-5.

-18 g

.g n.-

+16

-3

+2 4

f

(

+12 O

~4 4

i

+7 G

f

+7

-20

-14

+15 g

4

-1

-6

-9

-5 Control Rod Pattern

% Discrepancy

[(Measured / Calculated)-1) x 100 083 0.83 0

0 gg 13 0 uo 123 LO5 1.05 US 0.X) 070 1,3g t

LT L62 OM 02 0.93 07/

05 0*

0.81 093 0.

LOO 0.

g g

ig g

0

$ g3 0.77 0.88 0.94 08 LOO 0.M 0.73 ua 0 81 122

/ I'3I LO2 1.09 OM 0.79 LO3 063

,s%-0.99 )' 089 f 02 2

0.90 L0n' 0.78 0.71 077 Om L34 111 122 1.06 i

GAUGE RPF Measured RPF k = 1.0124 Fig. 7.

666 HW(t) (791) 175.0 CFFD in cycle 4 41

'q 910078 N/C b

-e

-e n

( -5

-14/

+3 t'~

4 s

-3 g

7

-18,

)

+16 a

a

+g

+

-2

+11

+3 4g

-5

-5

+11

)

7 O

0

-11]

-18

.g4 Y

4

~'

+10 s

-1

-7

'3

-9

-5 Control Rod Pattern P. Discrepancy

[(Measured / Calculated)-1] x 100 1

083 0.83 0*

g u4 ul 122 LOS 1.05 u7 0%

0.72 LO2 L62 0.99 1.33 081 0.2 Om 0*

081 05 OE 0.95 0.8 iM U2 gg g

0.72 0.M 0.94 0.80 LO1 08280 gg g,39 080 0* l 02 e'

' '*I LGB ON LO6 l

ON y

02 OM 0.79 0 72 0.78 Op 080 082 L36 til 124 LO6 L

GAUGE RPF Measured RPF k = 1.0132 l

Fig. 8.

495 W(t) (59%) 179.0 ETPD in Cycle 4 42

910078 N/C

_J7\\

-5 0

-4

+1

-16 3

+2 2

o

-20 i

.e

.g

+13 i

4

+4 4

q

+tg

+9 1

g

+2

.g1 (

e

+2 s

G

-1

- 21

-7 y,

h

+10

-1

-3 i

+

- 11

-3 Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1] x 100 i

0.80 080 OM US L10 LW 122 LOS LO3 U7

_/-

0.es 0.09 I'

LO2 OSp 07/

EU ON O.7'5 0.81 15/

W W

g4g 0.73 076 L46 0.W 0.94 E"

g 0.72 22 08^

g LO3 OM 0.85 y g_nn *0%/,

0.77 W

0 61 9,,

U9

/~

g LOO 0.7e 0.71 a75 One tm 1.13 u9 U0 GAUGE RPF Measured RPF k = 1.0115 Fig. 9.

666 MW(t) (79%) 184.9 ETPD in Cycle 4 43

9too:s 9/c i

1 1e c

0

/

/ +1

-19 s

+5

+m e

i 5

1

+10 d ~II 4

s c

+1 C

9

.ee-g 4

A

+5 0

/

-19

_g

+l4

+4

~

+9 0

-1 )

i

+2

\\

-9

-4 i

Control Rod Pattern

% Discrepancy

[(Measured / Calculated)-1] x 100 O

O O

O 1JO LOS a

1.04 121 LOG 1.03 l

U6 U7 O.67 0 67 L*

1.01 a9s L83 LT 1.32 1.00 0.78 gj7 g3 ggt 12 ) 1.05 g

1.56 122 EN O.

UO I'

V 0.M 0

gg 085 OM 0

0 E)

M7 0.72 24 0.2 LO9(3 LO4 4

085 0%

099 0.60 } OM

• ~, 0 87 /

0.87 0.99 <

I'04 0.91 099 A ~~

0.M 0.70 0.M s

069g OM

~

1.31 U4 1.19 L10

,i_'

GAUGE RPF Measured RPF k = 1.0119 j

Fig. 10.

665 MW(t) (791) 194.4 EFFD in Cycle 4 i

44 L

2.

m.

m m

-. +

- +., -

=

910078 N/C A'

-4 0 rm

_3

+2

-13

+3

+g g/

-3

-17 3

[2

+

1. 3

~lI

+11

+1

+10

'3

+3

+10

~

,3

+5

+1

-18

-7

+I4

+4

+4

-3_)

-7

-4 Control Rod Pattern

7. Discrepancy

[(Measured / Calculated)-1) x 100 0 71 0M 0.08 0.M gg LOO 1.00 U9 LW LO3 U4 3

og 057 LO3' s

LS2 LOO L35 UO 0.75 OM OM 0.

1.61 LOG 158 121 0.72 0.73 OM f OM 0.M 02 M3 0.00 U6 0.7B LOS U0 02 0.71 OM OM V

0.99 087 OM,

M W

sV og too 0.71 OM 0.M OM 02 OM 12 1.15 L19 L10 i

GAUGE RPF Measured RPF k = 1.0139 Fig. 11.

665 MW(t) (793) 215.4 EFFD in Cycle 4 45

.~

l 7

910078 N/C

'A 3

'*0 p

sg 0

~l4

-s

+12+10

-(3 V*

1 g

9 j+1 10 o

{

-17

+I4

+,3 0 _)

-3 0

Control Rod Pattern

% Discre

[(Measured /Calcufatedf-1] x 100 t'

0.72 0.75 02 0.2 LO6 0.98 LO3 121 LO1 W

l U7 g4 0.&

0.68 LO6 L80

'O.98 L30 OE 0.&

V3 0.79 OE 0.2 L61 UO L53

'121 g

gg 0.R 0.D

. f 0.79 0.96 g77 7 046 {

084 US Os6 ua 0.68 ul 1.30 LT U0 sy i OM (

L54[5 0.?O ON 02 084 099 L43 e-

- 'E' wv' 02 g

02 0.65 0.7/

066 >

OR O

122-UO UB UO GAUGE RPF Measured RPF k = 1.0145 ris. 12.

665 W(t) (792) 218.5 ErPD in Cycle 4 46

e 910078 N/C

~

-6

-3 4

/

-1

-13

~I

+2

1. g

-3

- 17 4

e'

+10

+5 I

)

[

0

+11

+,,

(s

+1 9

+3 4

-2

+13

+3

+5 A~

+4

-19

- 11 4

+4

-4h l

'~

+2 4

4 Control Rod Pattern

% Discrepancy

[(Measured / Calculated)-1] x 100 0.70 0.72 a06 070 gg 02 0.99 M7 02 LO2 U4 M3 0.65 OM l41 1.02 1.56 0.99 130

.UO M9 0.79 OM g.

O.78 082 g

0.

L33 0'

15/

LOS IN 120 E

~

0.79 0.86 OM f 0.92 088 G2 00 0.69 g7 0.78 l01

~

I 1.08 U1

, " 0.80 <

082 02 0.60 0.90}

056 058 156 g

LO1

~

0op OM 0.73 OM h.64J 0

085 12/

L16 M9 til

~

GAUGE RPF Measured RPF k = 1.0152 1

Fig. 13.

665 MW(t) (79%) 223.3 EFFD in Cycle 4 47

0 4

i 910078 N/C g E llN._,

'A

+5 d

-4 n H'

-18 i

.g 7

-3

+102 1

[

M

+1 f

+9

1. 3

-3

+16[

3 h

+7

+

_t7

_g

+15

+3

+10 4

+2

_e 4

i Control Rod Pattern

% Discrepancy

[(Measured / Calculated)-1) x.100 t

0.*10 0.72-0.68 0.09 LO2 LOO 099 U6 0.93 0.99 U3 0.86 I

b4I 0.89 LO2 L56 0.98 128 UO U8 0.3 OR 0.78 0.82 g

0' 1.31 0~

L57 LO6 L53 U9 g

gg 0.73 0.M b

0.92 0.69 0.95 OM 0.3 OM 06

. ) 0 81 LO6 e

1.11 OM 0.90 0.89 0.91 OSI

1. j LO1 / OM s,

OM D~

0.93 LT

',.89 OM 0.73

. A' 0.64J _

OM ON 171 U6 gg gg GAUGE RPF Measured RPF k = 1.0142 Fig. 14.

565 MW(t) (19%) 232.0 ETPD in Cycle 4 48

-