ML20090D926
| ML20090D926 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/30/1972 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| References | |
| NUDOCS 8303110189 | |
| Download: ML20090D926 (19) | |
Text
.
O l
l FUEL ABNORMALITIES AT THE O
R. E. GINNA PLANT AND OTHER 0?ERATING PWRs JUNE 30, 1972 O
.i 8303110189 73063005000244 PDR ADOCK PDR s
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1.0 INTRODUCTION
V Cycle 1B of the Ginna Nuclear Plant was initiated in the spring of 1971. Twelve Region 4A assemblies replaced twelve leaking Region 3 assemblies during the cycle 1 A-1B outage.
The plant operated normally during cycle 18 at the nominal 1300 MWt full power rating until the March 1972 scheduled shutdown in preparation for the 1520 MWt uprating.
The plant then operated at a maximum power of 1455 MWt except for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of operation at 1520 MWt.
Primary coolant activity increased during this time period which indicated additional fuel leakers. Ginna was removed from service on April 14, 1972, for inspection, maintenance, and cycle 18-2 refueling. Examination of the cycle 1B fuel (Section 4.0) revealed a number of anomalies including fuel leakers, bowed fuel rods, and an appreciable number of fuel rods with collapsed sections. Ginna cycle 2 fuel performance is discussed in Section 6.0 based on the observed anomalies, the predictability of the anomalies (Section 5.0) and the administrative controls to be applied to cycle 2 (Section 2.0).
Section 3.0 contains pertinent fuel rod design and operation information.
O V
2.0 ADMINISTRATIVE CONTROLS The ddininistrative controls given in Table 1 have been instituted to either minimize further fuel damage or to mitigate the radiation release to the environment should fuel damage occur in spite of the steps taken for prevention.
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i.() I 3.0 DESIGN AND OPERATIONS
SUMMARY
l The maximum operating power level has been reduced, on an interim basis, j
from the uprated 1520 MWt to 1266 MWt (1520/1.20). This power level is considered 100% power when interpreting safety limits, limiting safety I
system settings, and limiting conditions of operation. The reduced oower
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level is necessary, utilizing the very conservative information and analysis available on a short term basis, to maintain thermal-hydraulic l
margins and to minimize fuel clad interaction.
The Ginna core consists of 121 fuel assemblies. The number of assemblies l'
in the core by fuel region is given in Table 2 along with the enrichment, density, and pressurization of each region.
Figure 2 gives the cycle 1B I'
fuel loading pattern and Figure 3 gives the cycle 2 fuel loading pattern.
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Table 3 gives partinent fuel rod mechanical design parameters.
3 The core average burnup for cycles lA and 1B is given in Table 4 along with the end of cycle region average burnup. Also given in Table 4 is
.the time and volume averaged linear power for the entire core and by i
region for cycles lA and 1B.
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4.0 FUELEXAMINATiON An underwater television viewing system was used to conduct visual examinations on 54 Ginna fuel assemblies. Twenty Region 1 assemblies, twenty-two Region 2 assemblies, eleven Region 3 assemblies and one Region 4A assembly were included in the visual examination sampling pl an.
Each examined fuel assembly was visually scanned on all four faces. The majority of the peripheral fuel rods examined (90%)
showed no abnormalities. The remaining 10% of peripheral rods showed a series of abnormal conditions including failures, bowed rods, and collapsed cladding. Table 5 summarizes the total number of peripheral fuel rods examined, the type of anomalies observed and the frequency with which these anomalies were observed in each region of fuel.
Each type of observed anomaly, and the frequency of occurrence are discussed in the following paragraphs.
Leaking Fuel Rods All remaining Region 3 fuel assemblies were replaced during the cycle 1B-2 refueling. Region 3 was originally intended to remain in the core for 3 cycles.
Fuel examination in 1971 indicated that the majority of Region 3 fuel assemblies contained leaking foal rods. Local hydriding with subsequent cladding breach is believed to be the cause of the majority of the, leaking fuel rods.
Fuel rods either leaking or suspected of leaking were also found in Ginna Regions 1 and 2 but to a much lesser extent than in Region 3.
Failed Fuel Rods (Visually confirmed breached cladding)
A total of 4 visually confirmed failed fuel rods were observed. One failure (a split in the cladding) occur. red in a Region 1 fuel assembly.
The remaining three failures (apparent holes in the cladding) were observed in 3 different Region 3 fuel assemblies. No failed fuel rods were observed on the periphery of Region 2 or Region 4A assemblies. The failed rods constitute.14% of the peripheral rods viewed.
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Collapsed Cladding rh A collapse is a short segment of unsupported cladding, which due to high differential pressure has crept down to the point of appearing partially or totally flat in profile.
Collapsed rods were the most frequently observed anomaly.
Collapsed rods were observed in Regions 1, 2, and 3 fuel. Observed collapsed areas were distributed over the upper 40% of the fuel rod length with the highest frequency of collapse occurring between 120 and 140 inches from the bottom of the rod. The lengths of collapse vary from.s to 4.0 inches.
In general, the length of collapse increases as the observed axial location approaches the top of the fuel rod.
Collapsed cladding was also observed in Beznau Unit No. 1 Region 2 (approximately 2%) but not in Regions 1 and 3.
No collapsed rods have been observed at Zorita which is in its third cycle and no collapsed rods have been observed in pre-pressurized fuel.
Fuel Pellet Gaps O
Fue, densification wili occur (discussed in section s) irrespective of whether the cladding collapses. This has been verified by hot cell exam'ination (x-ray and gamma scanning) of Beznau Unit No.1 fuel rods discharged after cycle 1.
As a result, since the fuel cladding does not shrink axially, either the fuel pellet stack is shortened or gaps occur between the fuel pellets. These gaps exist where the fuel claading collapses but they may also exist in uncollapsed fuel.
Fuel Rod to Top Nozzle Gaps Rod to nozzle gaps were estimated for the peripheral rods in the visually examined assemblies.
No instances of rod interference were observed in the Region 3 or Region 4A assemblies.
Rod to nozzle gaps are on the average greater for collapsed rods than for non-collapsed rods.
Two rods were observed in contact at Ginna lA cycle shutdown. This assembly was visually examined at the recent shutdown. Additional rods are now in contact but the assembly continues to perform satisfactorily.
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O s.o eRemCuBitm oF ANomues Fuel Rod Collaose-Creep'down of fuel cladding onto the fuel is a predictable result of long-term operation in a LWR.
Recognition of this phenomenon has led Westinghouse to pre-pressurize fuel regions starting in 1969.
In Zorita, Beznau No.1 and Ginna, the initial core loading of Zircaloy-clad fuel was not pre-pressurized.* Given a significant length of unsupported (by fuel pellets) clad, inward creep results in increased ovality and eventual instability leading to collapse of the clad into the axial gap. This phenomenon was observed at-Beznau No. 1 during the first refueling and at Ginna during the recent cycle 1-B refueling.
On-site and post-irradiation evaluations of Beznau No.1 fuel have indicated that the axial fuel gap is not the result of missing fuel pellets.
Rather, the fuel stack has densified and incomplete settle-O ment hes resulted in,ocei end distributed geps eiong the fuel rod length. This description has been verified by gamma-scanning, x-ray, rod we'ighing and density determinations made on Beznau No.1 fuel, Westinghouse test rods irradiated in Saxton and other Westinghouse l
post-irradiation studies.
Observations of fuel following Beznau No.1 cycle 1 and Ginna cycles lA and 1B have been evaluated and provide the basis for a consistent model of creep-induced clad collapse. Using the model and other analytical studies of tube collapse, an estimate has been made of the extent of additional collapses anticipated during subsequent operation. Region 1 l
rods exhibit significantly fewer collapses, generally of smaller size.
i These observations are consistent - small gaps lead to long collapse times.
i Region 1 rod coH apse frequency will also increase during cycle 2, however, the already observed tendency for fewer and smaller collapses than in Region 2 is expected to continue.
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Cores fabricated since 1970 are pre-pressurized in order to inhibit clad creep.
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Local Abnormal Flux Peaking Local power spikes resulting from air gaps and water gaps have been analyzed with transport codes and confirmed by experiments in the Saxton reactor. A model has been developed for combining multiple gaps and/or collapses in order to predict the maximum power spike which could occur.
The model produces the probability of not exceeding a given power spike.
Nearly all flux traces taken at Westinghouse plants utilizing the in-core movable detector system show local flux peaks, the vast majority of the peaks being less than 1%.
Flux peaks in the range of 4 to 9% have been detected on four traces (out of several thousand). This local flux peaking can be noted on the flux traces after a few months of operation. No collapsed rods were noted during the Ginna cycle 1A-1B refueling several months after the start of operation. This provides confinnation that gaps exist between fuel pellets even though the cladding has not collapsed.
The models developed to predict local flux peaking is very conservative when compared to the flux peaks noted on flux traces. Continued surveillance O
or the in-core riux 9eakin9 is niaaoed eau the comnarison or the "ew data with the models will provide confirmation of the conservatism.
Zircaloy Growth Westinghouse has developed a Zircaloy fuel rod gorwth correlation based on observations and measurements made at Saxton, Beznau No.1, Mihama 1, Zorita and Ginna. A total of appioximately 2000 measurements are included in the available Westinghouse data. At. Zorita and at Ginna, data have been obtained after each of two cycles providing a good test of the predictive ability of the design models. The growth' observed after two cycle exposure has in all cases been within the design uncertainty band and in most cases has been below the mean prediction.
In addition, as a measure of overall system predictions, the number of rods touching the top nozzle in the cold condition correlates well.
Based on these data, design evaluations have been performed to provide assurance of satisfactory performance during cycle 2 operation and the Q
subsequent cold shutdown.
Since the stainless thimble thermally expands to a greater extent than the Zircaloy clad fuel rods, the clearance between c
the top of the fuel rod and the top nozzle is a minimum in the cold,
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shutdown condition.
During operation, the rod exhibiting the highest growth will not contact the top nozzle until approximately 2/3 of the full second cycle exposure has been accumulated.
However, contact will not result in rod bowing until a substantial additional interference is experienced. This level of interference will not occur by the end of cycle 2.
Some rods are currently in contact at the cold shutdown condition.
Upon return to power, the differential thermal expansion between rods and thimbles will provide an additional 0.4 inch relative expansion of the assembly and provide a gap for all rods.
At subsequent cold shutdowns additional rods are expected to contact the top nozzle. Analyses have been performed which demonstrate that loads imposed on the fuel rods during cold shutdowns remain in the elastic range for all rods which do not exhibit collapse.
Some number of collapsed rods are expected to be in significant interference cold. These rods may be stressed beyond yielf in the cold condition and some could crack and release their fission gas inventory as a result.
Use of a controlled
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approach to the cold shutdown condition should cause the few failures so anticipated to occur over a sufficient time span to limit the activity rel e& sed.
d The rods impose a load on the assembly in the cooled down condition which has been analyzed and found to be acceptable. A conservative assessment of tolerance factors was made to quantify the risk of parting of the bottom nozzle. This risk has been determined to be considerably less than one chance in one hundred for the lead growth assembly.
Should such parting occur, the thimbles would be retained in their current positions by guide i
pins, scram action would not be compromised and the assembly could sustain additional subsequent irradiation.
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.6.0 PERFORMANCE EVALUATION-Fuel Clad Interaction When the fuel pellets are tightly constrained by the fuel cladding, anc 'a power increase takes place, the greater thermal expansion of the uranium oxide can potentially result in clad stresses above the Zircaloy yield strength. Although some local clad ductility may exist, it is difficult to take quantitative advantage of it.
Hence, clad yield stress is used as the design limit - exceeding this limit is then defined as clad failure.
Such failure is not however to be characterized as an abrupt or massive blowout or loss of integrity.
Failure here will likely take the form of local cracking, with some being through wall cracks.
Two conditions exist in this fuel which give rise to this constraint:
(1) clad creeps down on the pellet, providing radial constraint; and (2) clad collapses down into a pellet axial gap, providing axial constraint.
The potential for clad failure in both cases is considered low because Q
of the reduced power operation and considering the revised, more restric-tive, control rod insertion limits. The exception to this may be certain overpower transients which would cause an increment in rod average power. The administrative restrictions listed in Table 1 are intended to minimize fuel clad interaction and these modes of failure.
In addition, technical specification limit coolant activity and potential releases.
Finally, the coolant activity monitor provides a means to continuously evaluate clad failures and enables action to be taken in a timely manner should a large number of failures occur'.
O
DNBR Evalua' tion
/
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With respect to D"B under normal operation and expected transients in-(-)
Ginna Cycle 2, the following offsetting considerations exist:
s 1)
Core power will be reduced by a factor of 1/1.20; this will increase DNB ratios by 34%.
2)
If fuel stack height and local power spikes were to increase the local heat flux by 20%, the DNB ratio is calculated to be reduced by 10%.
Halving the poder spike effect is a consequence of the integrating "F" factor in the W-3 correlation, which relates uniform and non-uniform D"3 heat fluxes.
3)
Flow patterns will be altered in the vicinity of a collapsed section; however, sudden contraction and expansion effects are not detrimental.
In fact, increased turbulence and mixing effects tend to improve heat transfer.
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4)
Rod spacing may be altered slightly as a consequence of s
the collapsed section on some fuel rods; this effect will
- ' reduce D"B ratios by less than 10%.
5)
The fuel in the vicinity of collapsed rods will run at power level more than 15% below the core hot channel; this leads t'o DNB ratios increased by more than 255.
6)
If a collapsed section is in contact with an adjacent rod, the DNB ratio is not necessarily reduced (as evidenced by the contact of the support grids and fuel rods). Elec-trically heated DNB tests show a penalty for full length contact of 7%, which is clearly an upper limit.
It can be seen that those effects increasing DNB ratios for Cycle 2 are far in excess of the detrimental effects.
Thus, it is concluded that for Cycle 2 crerction, '":B ratics during neral =eration and
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in expected transients are at least as high as those reported during previous safety analyses. ~
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Fuel Temperature ek)g Fuel densification increases the beginning of life fuel pellet /
cladding gap which in turn leads to an increase in fuel temperature.
This temperature increase is less than 200 F.
The maximum fuel central temperature at overpower remains less than the minimum UO2 melting point of about 4800 F (45,000 MWD /MTU burnup). Unirradi-ated UO melting point is about 5100 F.
2 The amount of energy stored in the fuel is larger, for a given linear power rating, due to the increased fuel temperature. This stored energy affects the loss of flow (LOF) and loss of coolant (LOCA) accidents.
Analysis of a 3 ft.2 LOCA showed a 175 F peak cladding temp-erature increase to 2210 F assuming the same maximum linear power (core power decrease offsets power spike) and no reduction in total core power. Decreasing the total core power enhances cooling during blow-down. The densification and higher fuel temperatures result in a slower LOF heat flux decrease following reactor trip.
The slower heat flux decrease is expected to have less effect than the reduced operating power level.
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TABLE 1 O
ADMINISTRATIVE CONTROLS TO BE APPLIED TO GINNA CYCLE 2 1.
Control Rod Insertion Limits as given in Figure 1 under normal operating. conditions (except for physics tests). During load losses or automatic turbine runbacks,this limit may be exceeded by automatic (or manual) rod control but the limits will be re-established once power has stabilized.
2.
No part-length control rods in the core when reactor is at power, (except for special tests).
3.
Cycle 2 initial startup:
a) 3% power / hour from 0% to 75% power b) 3% power / day from 75% to 100% power c) continuous chemistry follow.
4.
Cycle 2 operation other than initial startup:
O a) 10% power / hour increase U
b) revert to initial startup procedures should low power (less than 95%) be i
maintained continuously for more than 24 days.
5.
Cycle 2 cold shutdown (anytime): Cool down slowly and monitor primary coolant chemistry continuously.
Hold at a constant temperature if a greater than expected increase in activity occurs until the situation can be cvaluated.
6.
Shutdown if overpower transient occurs and monitor coolant.
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.i TABLE 2 GINNA CORE LOADINGS FUEL DESCRIPTION ASSEMBLIES LOADED ENRICH DENSITY PRE-CYCLE CYCLE CYCLE FUEL BATCH (w/o)
(%)
PRESSURIZED 1A 18 2
Region 1 2.44 94 No 41 41 38 Region 2 2.78 92 No 40 40 10 Region 3 3.48 90 No 40 28 Region 4A 3.20 92 Yes 12 12 Region 4 3.14 94 Yes 40 Region 4B 2.90 94 Yes 21 TOTAL 121 121 121 t
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()
TABLE 3 FUEL R0D MECHANICAL DESCRIPTION (Beginning of Life)
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Fuel Rod 0.D. = 0.422 in.
Cladding Thickness = 0.0243 in.
Cladding Material = Zircaloy-4 Fuel Pellet 0.D. (Regions 1, 2, 3) = 0.3669 in.
Fuel Pellet 0.D. (Region 4A) = 0.3649 in.
Fuel Pellet 0.D. (Region 4, 4B) = 0.3659 in.
Fuel Pellet Height = 0.600 in.
()
Fuel Pellet Stack Height (Regions 1, 2, 3) = 144 in.
Fuel, Pellet Stack Height (Region 4A) = 142.8 in.
Fuel Pellet Stack Height (Regions 4, 4B) = 142 in.
Guide Thimble Material = 304 SS O
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O TABLE 4 GINNA FUEL OPERATING HISTORY CCRE REGION 1 REGION 2 REGION 3 REGION 4A Average Burnup-End of Cycle 1A (MWD /MTV) 7700 9,200 8,300 5,600 Average Burnup-End of Cycle 1B (MWD /MTU) 8830 18,800 18,000 10,000 9,500 (Cy.lB)
Time / Volume Averaged Linear Power -
4.2 4.8 4.4 3.2 Cycle 1 A (kw/ft)
O iime/ volume Avere9ed Lineer eower -
4.8 s.3 s.2 3.4 s.,
Cycle 1B (kw/ft)
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TABLE 5
SUMMARY
OF RGE CYCLE 1B VISUAL EXAMINATIONS NUMBER OF EXAMINATIONS NUMBER OF PERIPHEPAL PERIPHERAL REGION ASSEM.
RODS RODS COLLAPSED 1
20 1040 21 2
22 1144 83 3
11 572 20 4A 1
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