ML20084T101

Forwards Minutes of Recent Safety Audit & Review Board Meeting Re Consideration of Addl Analyses Concerning ex-core Flux Detector Signal Oscillations

ML20084T101
Person / Time
Site:	Palisades, 05000000
Issue date:	07/20/1973
From:	Sewell R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Oleary J US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20084T038	List: ML20084T096 ML20084T101
References
NUDOCS 8306210232
Download: ML20084T101 (10)

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/ by K* s/ /w g*,syl e li {bi.E s s } s. General Offices: 212 West Michigan Avenue. Jackson. Michigan 49201. Ares Code 517788-0550 July 20,157/3 6 Mr. John F. O' Leary, Director Re: Docket No 50-255 Directorate of Licensing License No DPR-20 United States Atomic Energy Commission Washington, DC 205145

Dear Mr. O' Leary:

Preliminary results of investigations that have been con-l ducted at the Palisades Plant with respect to. excore flir.: detector signal oscillations were reported by letter on June 12, 1973 Since that time, additional test data analyses have been completed and the Safety Audit and Review Board (SARB) has met to consider these addi-tional analyses. In order to keep the regulatory staff abreast of the analy-tical results that have become available since our June 12, 1973 letter, I am attaching a copy of the minutes of the most recent SARB meeting concerning excore flux detector signal oscillations. Yours very truly, ga42L4 RBS/pb Ralph B. Sewell Nuclear Licensing Administrator e <v/ // C0CKETED Jut.231G73 "p$

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Ie [ /f' RLHiue( ) LMHausicr RBDetlitt ur ERCoop r To GSKerley GBMatheney RESewell JZReynolds LSGibson GJWalke TEMcElroy DMNoble FRon WJBeckius, P21-123 ~ x f. CODSumBIS ' 9 DATE July 11, 1973 D P0rl8r l C0mpBHy s s suesEcr SARB Minutes N L~7 / lNTERNAL CORRESPONDENCE INTERNAL CORRESPONDENCE Beck 14-73 cc RCYoungdahl KASwarts RALamley FEFlynn AVHume JFSimpson CJHartman Attendees Guests RLHaueter RBSewell RWDeVane GSKeeley GBMatheney JCalvin GJWalke TEMcElroy JMullooly WJBeckius IMiausler LSGibson GBSlade A special SARB meeting was held on June 28, 1973 to discuss the flux oscillations that had been observed in the excore nuclear instrumentation at the Palisades Plant. As a result of discussions with the Directorate of Regulatory Operations (DRO) and the Directorate of Licensing on June 7 and 8, 1973 CPCo had promised, among other things, to monitor flux signals on a shift basis and to notify DRO of any significant change. An inspector from Region III of DRO was on site most of the week of June 11 to June 15, 1973. During that time the definition of "Significant Change" was discussed with the Plant Staff. Members of the Plant Staff along with General Office personnel had, during this week, been working independently on a quantitative method of monitoring to augment the qualitative method that had been used since March 1973 This method l consisted of averaging the five largest peaks observed in a two-minute sample l strip chart trace to provide a " figure of merit". It had been evolved and was ( in the late stages of testing for applicability and statistical variation on l June 15. It was agreed by site personnel and the Region III inspector, that deviations greater than 14 millivolts around an average value of 74 millivolts (as the average a=plitude of the 5 highest peaks in 2-minutes) were to be defined as "significant" and therefore reportable. This level was determined by the statistics of the data (statistical significance) rather than on judgement of the consequences of a change of this magnitude. On Monday, June 25, 1973 the data exceeded the 88 millivolt "Reportability Threshold' and Region III DRO was notified by telephone. Since the AEC had been notified that the signals had changed the SARB meeting was called to satisfy the requir cuents set forth on Page 8 of SARB Minutes for the May 16 meeting (SARB would meet again if any significant change was seen during Plant operation or tests). / /, / /

f^ (&' f\\ ~ O SARB Members ( U 2 SARB Minutes July 11, 1973 The meeting was opened with two broad objectives defined: 1, To review the safety aspects of operation of the reactor with out-of-core flux variations of the currently observed amplitude and frequency. 2, To define a limit on flux oscillations above which further action would be taken. Representatives of the Plant Staff presented the Board with charts, time traces and data showing the recent upward trend. Attached Figures 1 and 2 effectively su=marize the data that was presented. The Plant representatives stated that the trend lines indicate that "The Figure of Merit" (average peak to peak amplitude of the five highest peaks observed in a two-minute test interval) had increased over 15% above the data presented to the DRO inspector on June 15 Combustion Engineering representatives noted that current values of the oscillation magnitude were still less than those observed in January 1973 before steam generator tube plugging. Possible causes and effects of various operating parameters on the excore oscillations were discussed by the Board. None of the items brought up by Board. Members could be related with any certainty to observed changes in the oscillations, One of the items discussed was the effect of crud levels in the primary coolant system. The Plant Superintendent stated that system suspended crud levels have been observed to be on a down trend during the entire period in question. This is believed to be a good indication that crud is not being " shaken" from the fuel. It was pointed out that declining suspended crud levels may indicate more is plating out. Following steam generator tube plugging, a two to three percent flow reduction apparently resulted 1-approximately 30 percent reduction in flux oscillation maplitude. This seems to indicate that the oscillations are very sensitive to slight flow changes and, in fact, are probably more sensitive to flow changes than Plant flow instrumentation. It would therefore appear possible that changes in crud laydown patterns could cause slight flow changes wh?.ch are undetectable by Plant flow instrumentation but which could still have significant effect on flux oscillation amplitudes. The Board concluded, that ;heoretical possibilities not withstanding, there is no data available to date that would tie observed changes in flux oscillations to system crud lesels. Another item discussed was the expected effect of changes in core power' distribution on flux oscillation amplitudes. It was pointed out that core axial shape has not changed appreciably during the period in question. The radial distribution has flattened a little. However, the radial effect is l believed to be compensated in existing oscillation data plots by dividing the l AC component of the signal by the DC level i.e. plots are in. terms of percent . of the signal that is noise (also referred to as power level " normalized"). The effect of soluble boron in the coolant (and reflector) declining with core burnup was also discussed. It is believed that to a first order approximation the amplitude of excore oscillation would not change for a given _a.. u

SARB Mem)ers T 3 SARB Minutes , July 11, 1973 movement of the core due to reduction in soluble boron. The cross section of the boron itself is small in the energy range of interest. The shift in the energy spectrum over which the water cross sections are averaged may have some slight effect. The calculation that converts percent excore oscillation to equivalent mils of motion is subject to some uncertainty in itself thereby making second order effects difficult to evaluate. Differing trends in the data would indicate this is not the cause for the currently observed variation. Combustion Engineering (CE) made a presentation to the Board outlining work CE had done thus far on analysis of Palisades flux oscillation data. The data was obtained by recording select signals (selection based on analysis of "on-line"SAICORwaveanalyzerdataand/orprevious"off-line" analysis)on14 channel FM magnetic tape at the Plant. The tapes were then taken back to CE Laboratories in Windsor where they are read, digitalized, and analyzed by an appropriately programmed digital computer. This method is laborious and time consuming. However, because of the low frequency and random nature of the signal variations, it is believed to be superior to all other techniques available at this time. CE has taken data on two occasions: Once in March 1973 and again in late May 1973. At the time of the meeting most of the analysis on the March data was complete. Only preliminary analysis of a small fraction of the May data had been ecmpleted. Their discussion th2refore was based primarily on t'he March data. CE's analysis techniques reduce the data in terms of power spectral density plots, amplitude probability density curves, coherence / phase angle plots, and RMS signal level in percent. Data obtained from excore detectors, prompt incore detectors, three accelerometers on the outside of the vessel, pressurizer pressure instrumentation and steam generator JP (reactor flow) instrumentation, were analyzed by these means. Power spectral density (PSD) plots show that all excores are similar in that they exhibit a broad band random process from zero (really the low frequency cut-off) out to 6 Hz. The 0 5 Hz peak previously reported was probably due to the low frequency cut-off characteristic of the SAICOR analyzer used in that determination. The energy content is inversely dependent upon the frequancy such that 80% of the energy appears in the O to 1.5 Hz band. There appears to be a tendency toward a secondary broad band peak in the 1.5 to 3 Hz region. Most of the remaining 20% of the energy content is expended in this band. There is no sharp response indicative of a resonant condition in the O to 6 Hz region. Although the amplitudes are at least a factor of 5 lower, PSD's for the incores resemble those for the excores including the secondary response in the 1.5 to 3 Hz region. Incores also see peaks at 14, 14.2 and 15.6 Hz which are believed to be hydraulic in origin. PSD plots of accelerometer data exhibit peaks at 6, 13.8, 14.8 and-16.5 Hz. A peak at 0.5 Hz.was observed but is believed to be the result of instrument noise. The 14.8 Hz peak is identified with the 888 rpm rotational speed of the pumps. The others are believed to be natural frequencies of the YW)R,e g 7m. _ wa,& ?. - ~.""'M'^ ^&~ sv ,,x.,

O fl SARB Members pU pV 5-SAR$ Minui;es d 4 - July 11, 1973 Coherence analysis of two accelerometers located 900 apart in the snubber region of the reactor vessel over the O to 20 Hz band show no coherence in the area of interest below 6 Hz. This tends to confirm that the 0.5 Hz peaks in the PSD plots are intrinsic noise. Significant coherence exists for the 6 and 13.8 Hz peaks. This tends to confirm that the coherence analysis is valid although these" hydraulic" peaks are not believed to be significant to the investigation. Coherence analysis of top and bottom halves of a given channel indicates perfect correlation out to 2 Hz, at 0 phase angle. Correlation exhibits rapid decline past 4 Hz. Correlation of channels across the core is good out to phase angle and declines steadily from 1.5 to 6 Hz. 0 1 5 Hz at 180 The west vessel accerlerometer, pressurizer pressure and loop flows were correlated with NI-005 lower excore. No coherence was found for any of the signals in the O to 20 Hz band. Flow signals from opposite coolant loops were correlated. No coherenec - was present at the strong 0.5 Hz peak. This indicates a non-linear relationship in the 0.5 Hz component in opposite loops. This non-linearity suggests a mechanism for hydraulically induced internals motion in that it appears the large 0.5 Hz co=ponent can be out of phase in opposite loops. Signals from two prompt response cobalt incores (6-5 and 7-5) located near the edge of the core on opposite sides of an axis through the outlet nozzles were recorded. Unlike the excores these two detectors show no coherence at low frequency. This climinates radial flux oscillations as a direct source of excore oscillations. Each of these incores were correlated with NI-005 lower excore. Incore 7-5 nearest to NI-005 excore shows no pronounced low frequency correlation. Sharp decreases at 14 and 14.8 Hz indicate the excores do not see the hydraulic peaks seen by the incores and suggests incore response at thest, frequencies does not involve power oscillations. Incore 6-5 which is across the core from NI-005 exhibits no correlation in the O to 1.5 Hz range, but shows strong coherence in the 1.5 to 3 Hz range. This coherence is thought to be significant and because it appears in range corresponding to a fuel bundle natural vibration mode it is believed to be symptomatic of fuel motion. Additional data was taken in Miy in hopes of further refining this aspect of the analysis. This data has not yet been analyzed at this time, however. CE stated that the driving force for the oscillation is believed to be hydraulic pulsations of the order of -"i psi maximum amplitude acting in the downcomer annulus and in the upper plenum region. The pulsations originate from a conbination of flow turbulence, inlet coolant impact on the core barrel andexcitationofthe0.4to0.6Hznaturalfrequencyofthepressurizer/ fluid system. The primary response is the barrel, support structure and core moving as a unit in the cantilever mode with a peak to peak value of 40 mils (t20 mils). This value is consistent with expected clearances at the snubbers as well as with the expected pressure differentials available to move the barrel. As o result of force from the barrel acting through the core support plate exciting the core at its own natural frequency the core is also moving with respect to the barre 3/ shroud. Because of reflector effects and possible changes in core edge flux resulting therefrom the physics of relating a certain core motion to i %. ]w - ] ms. 'r--m

O O O O SARB Memliers 6 SARB Minutes July 11, 1973 a certain excore oscillation amplitude is complicated and somewhat uncertain. Vibration analysis, however, indicated that if the core support plate is driven 120 mils in the fundamental frequency range of the fuel the fuel will deflect as much as 2.60 mils at its centerline. The O to 6 Hz range of the random core barrel motion encompasses the 2 to 3 Hz natural frequency. The amplitude of the fuel motion is not inconsistent with what might be required to explain the remainder of the excore amplitude over and above that contributed by core barrel j motion. This model is believed to be consistent with all data observed to date. The specific evidence supporting core barrel motion include: l 1. The similarity and phase relationships of'the excore detector ] response. ] 2. The truncated behavior and the fact that it occurs, on the high side on one side of the core and on the low side on the other. 3. The orientation with respect to the outlet nozzles for the two loops. 4 The fact that ouscrvations at Palisades and Maine Yankee indicate that the required pressure pulcations exist and are large enough 4 to cause the h0 mils peak to peak motion. 5. The required clearances are believed to be available in the snubber region. 6. 80dp of the energy is in the O to 1.5 Hz band. Specific evidence supporting fuel motion include: 1 Secondary response in the PSD plots for both incores and excores in the 1.5 to 3 Hz region. 2. Across the core coherence of incore and excore signals in the 1.5 to 3 Hz band and lack of coherence in the O to 1.5 Hz region. t 3. Similarity of incore and excore PSD plots with the AFD's being truncated for excores while the incores are Gaussian, 4. Excitation occurs over the O to 6 Hz band which includes the fuel natural frequency of 2 to 3 Hz. Vibration analysis indicates sufficient driving force is available to produce significant amplitudes. CE provided the Board with a review of reactor internal structures attachment mechanismo, and assembly procedures including the various tolerances and freedom of the various members to move relative to one another. i -~ -m- - - c-- - " ~-

m ,b 3 SARB Members 7 SARB Minutes July 11, 1973 CE stated that they had done a detailed analysis to assess the impli-cations of the reactor internals and fuel being subjected to flow induced vibrations sufficient to result in the reported power oscil htions. Specifically, the review showed that considering as-built dimensions, tolerances, and inspections performed to date, they conclude that there is adequate clamping force available at the core barrel flange, to hold the core barrel down and adequate frictional force to preclude movement in the translational mode under existing hydraulic forces. Lack of clearance would also prohibit movement in the rocking mode at the core barrel flange. They concluded that the most likely displacement is in the cantilever mode. The shell mode is also possible but is not consistent with observed excore oscillation patterns. Snubber design is such that the lower end of the core support barrel is restrained by design to a theoretical maximum of 124 uils in the hot condition. Calculations indicate the resulting stress of 2,760 psi in the core barrel is an order of magnitude below ASME Section III allowable for cyclic stress. Beam motion of the barrel will induce compatible movement in the ccre shroud. Resulting stresses in the core shroud attachment bolts would be well within acceptable fatigue limits. Movement of the upper guide structure (UGS) would similarily be in the beam mode. Random hydraulic froces in the upper plenum acting on the control rod shroud assemblies may induce some motion. Keys in the upper fuel bundle alignment plate would limit motion to M mils relative to the core barrel. Motion either by itself or with the core barrel may contribute to the excore oscillations. The contribution is not likely to be large in view of the small available clearances. Although hydraulic forces are not sufficient to lift a fuel assembly (500 lbs net downward force is maintained) lateral movement in translation, rocking, and pinned beam mode can be caused by lateral hydraulic forces. Lateral hydraulic forces are unlikely to be coherent over the whole core and therefore this mechanism is inconsistent with observed excore behavior. Inertial forces imparted to the fuel from movement of the core support plate can also cause fuel motion. This motion as discussed previously could be as great as 160 mils at the core centerline and is consistent with observed oscillations. Forced vibration autoclave tests on single tubes and hot loop tests on full assemblies have experimentally demonstrated that elastic motions of this magnitude are insufficient to cause significant fretting. The Board asked a number of questions relative to possible tie-in between the previous ring shim bolt failure and the present problem. CE reiterated that the observed failures were by high cycle low stress fatigue. Movement of the type suggested by current excore observations would result in high stress low cycle fatigue. Furthermore inspections at the time indicated that the clamping arrangement at the flange was essentially normal as designed. CE indicated that failure of the clamping arrangement, e.g. disfunctional ring shim, should result in significant generation of impact energy both at the flange and in the snubber region. Accelerometers in both areas are very quiet and free from indications of impacting. r

n O. (J 8 SARB Member = SARB Minutes July 11, 1973 1 The Board agreed that CE's model appeared to be adequately supported by the data and therefore continued operation at present levels, which were not significantly greater than those observed in March, presented no previously unreviewed safety problem. It was suggested that an amplitude approximately double the present amplitude might be adopted as an action level. The oscillations in January 1973 were believed to be approximately O.85% rms. Fifteen (15) percent above this would represent a statistically significant increase to indicate that possible degradation was taking place. This would correspond to about 1% rms as compared to a reported value of 0.56% on the day of the meeting. CE stated this was also essentially consistent with the point where first fatigue in excess of code allowable would occur under a very conservative model whereby all of the observed amplitude was assumed to be due to core barrel deflection. This model very conservatively assumes that the snubbers are disfunctional and that 53 psi is available across the sides of the core barrel although normal total reactor vessel differential pressure is only about 26 psi. The component failed under this model would be the lower UGS alignment pins, the loss of which is not an obvious safety problem. SARB agreed that 1% rms as calculated by CE's method would represent c reasonable limit. It was suggested that upon reaching the limit the Plant should revert to three pump operation thereby eliminating the oscillation problem and the consequential cyclic stresses. SARB asked to be updated if significant new data becomes available from CE's on-going analysis. The Board expects a final report from CE as soon as possible after analysis is complete. The Board considered the possibility of routine operating tests addressedtothefluxoscillation/coreinternal.sproblembutnomeaningful test could be suggested. The Board requested that the Plant look into installa-tion of a speaker to read accelerometer signals in the audible range on a routinebasistomonitorforpossibleloosepartsand/orinternalsimpacting. 0 heighu - <- y

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