ML20154P442
| ML20154P442 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/19/1988 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| LIC-88-823, NUDOCS 8810030064 | |
| Download: ML20154P442 (29) | |
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l Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 2247 402/536-4000 September 19, 1988 LIC-88-823 U.S. Nuclear Regulatory Commission ATTH: Document Control Desk Mail Station P1-137 Washington, DC 20555
References:
1.
Docket No. 50-285 2.
NRC Safety Evaluation from NRC (P. D. Milano) to OPPD (K. J.
Morris) dated July 14, 1988 Gentlemen:
SUBJECT:
Clarification of Information Relating to the Operation of the Internals Vibration Monitoring System The purpose of this submittal from the Omaha Public Power District (OPPD) is to provide clarification of information relating to the operation of the Internals Vibration Monitoring (IVM) system at the Fort Calhoun Station. The information was requested in the July 14, 1988 Safety Evaluation relating to the threshold levels for the IVM system (Reference 2). The attached report is the formal doc-umentation of the discussion held during a telephone conversation on June 27, 1988 between Mr. J. Fisicaro and other OPPD staff members and Messrs. P. Milano and L. Lois of the NRC.
It is OPPD's belief that the proposed thresholds for the IVM system are valid based on an adequate inspection of the reactor internals and the evaluation of experiences at similar Combustion Engineering facilities, and that OPPD's staff has sufficient expertise to use the IVM system.
The attached report provides supporting documentation to resolve concerns raised in the SER and the June 27, 1988 telephone conversation.
OPPD believes this information, in conjunction with previous OPPD submittals relating to the thermal shield deferral, addresses the concerns raised in the Safety Evaluation.
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l Document Control Desk LIC-88-823 Page 2 OPPD believes, along with Combustion Engineering, that by using the proposed threshold levels OPPD is able to detect early signs of degradation of the ther-mal shield support mechanism.
It is requested that the NRC perform a reevalua-tion of the proposed threshold levels using the attached report for supporting information.
If you have further questions concerning this matter, please do not hesitate to contact us.
Sincerely, nV,<D$w g,, sfbbs>tku, K. J. Morris Division Manager L
Nuclear Operations i
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LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, DC 20036 R. D. Martin, NRC Regional Administrator, Region IV P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector l
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CLARIFICATION OF INFORMATION RELATING TO THE OPERATION OF THE INTERNALS VIBRATION MONITORING SYSTEM This report is in response to the Safety Evaluation performed by the office of Nuclear Reactor Regulation relating to the threshold levels developed for the FortCalhounStationInternalsVibrationMonitoring(IVM) system.
The purpose of this report is to provide additional information on internals support degrad-ation events at other Combustion Engineering facilities, the methodology used for developing IVM threshold levels, and the analysis of this data. This addi-tional information was requested in the July 14, 1988 NRC Safety Evaluation of the proposed threshold levels.
The body of the report addresses the questions raised in the NRC Safety Evaluation.
This information was discussed in a telephone conversation held between OPPD and the NRC on June 27, 1988. Attachment 1 is included to 3rovide additional information on the low power IVM data acquisition at Fort Calioun. is included to provide the basis of the Fort Calhoun IVM threshold levels.
OPPD believes that the submitted threshold levels are able to detect early signs of degradation of the thermal shield support mechanism.
It is requested that the NRC perform a reevaluation of the proposed threshold levels using the following report for supporting information.
INTRODUCTION TheOmahaPublicPowerDistrict(0 PPD)submittedtheFortCalhounThermal Shield Support System Inspection Deferral report to the NRC on August 28, 1986 (Reference 1).
The purpose of the deferral report was to replace the commit-ment for a 1987 thermal shield inspection with an internals inspection to be performed during the 1993 outage.
The NRC safety evaluation of the deferral report, dated February 12, 1987, accepted the deferral request contingent upon OPPD supplying additional information on the Internals Vibration Monitoring (IVM) system and developing IVM threshold levels (Reference 2).
On October 13, 1987, OPPD submitted the additional information (Reference 3) and the February 25, 1988 submittal (Reference 4) proposed the Fort Calhoun IVM threshold levels.
The NRC Safety Evaluation of the threshcid levels, dated July 14, 1988 (Reference 5) determined that the proposed threshold levels were not accept-able.
It is the intent of this report to provide clarification and additional supporting information in order to support OPPD's position that the proposed threshold levels are valid and accepta)1e.
Two questions were posed in the Safety Evaluation of the IVM threshold levels, they werer 1.
Is the method on which the Fort Calhoun IVM is based reasonable?
2.
Are the proposed threshold levels for detecting thermal shield support degradation reasonable?
OPPD believes that both the methodology and the proposed threshold levels are valid.
This report will explain the basis for OPPD's position on these ques-tions.
p ITEMS OF CLARIFICATION Before directly addressing the two cuestions, OPPD would like to clarify some important items that were referencec to in the subject Safety Evaluation.
O ST. LUCIE VISUAL INSPECTION l
One of the concerns raised by the NRC in Section 2.1 and Section 2.3 of the threshold level evaluation relates to the statement that the St. Lucie experience indicates that visual ins)ections may not be adequate to determine the condition of the tiermal shield supports. The NRC evaluation states, "Neither visual inspection detected degradation of the thermal shield supports."
The inspections referred to at St. Lucie were )erformed in 1978 and in 1981 with the core support barrel and tiermal shield in place inside the reactor vessel. The two examinations performed included a remote visual examination of the up>er guide structure and only accessible parts of the core support )arrel.
The St.
Lucie examinations were performed in an attempt to locate loose parts within the reactor vessel. As discussed in Reference 7, the thermal shield was not included in the scope of the inspection and thus degradation of the supports was not detected.
O FORT CALHOUN VISUAL INSPECTION Section 2.3 of the safety evaluation compares the St. Lucie visual examination to the Fort Calhoun 1983 inservice inspection of the reactor internals. As discussed above and presented in Reference 7, the St. Lucie visual inspection was not a 10 year inservice inspection, and the Thermal Shield was not listed in the scope of their inspection.
The emphasis of the 10 year 151 at Fort Calhoun was placed on inspection of the thermal shield positioning pins, locking collars, and lock welds. The core support barrel and the thermal shield were placed in a lower portion of the refueling cavity for this inspection. As a result of this inspection of the core support barrel and thermal shield, OPPD and the inservice ins)ection contractor concluded that these components were in the as-)uilt condition.
Section 6.2.2 of the August 28, 1986 OPPDsubmittal(Reference 1) provides a detailed description of the thermal shield support in-spection performed in 1983 and the results which support OPPD's conclusions.
O FORT CALHOUN CURRENT INTERNALS CONDIT*0N OPPD believes that the Fort Calhoun Station reactor internals are i
currently in the as-built condition.
This statement is based on i
the following three items:
1 1) 1983 ISI examination of the reactor internals.
j 2) long term frequency behavior of the 12.5 Hz resonance.
3)
Low power data analysis.
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FORT CAf340UN CURRENT INTERNALS CONDITION (Contiaued)
The first supporting point in assessing the present condition of the reactor internals is the 1983 ISI examination of tha reactor t
internals.
This examination concluded that the core support bar-rel and the thermal shield were in the as built condition.
The previous section on the Fort Calhoun Visual Inspection should be referenced for additional information.
The second supporting point is the long term frequency behavior of the 12.5 Hz resonance.
The 12.5 Hz resonance is the primary indi-cator of any change in the adequacy of the reactor internals sup-port mechanism.
2 The long term frequency behavior of the 12.5 Hz resonance was dis.
cussed in detail in Section 6.3 of the initial deferral report (Reference 1). Neutron noise data was analyzed and presented from the period of 1974 through 1986. As stated in the Reference (1) report, Section 6.3.2 Conclusions, "Therefore, based on excore neu.
tron signal evaluations, thermal shield support system conditions have not degraded at full power conditions."
l The final item supporting the reactor internals current condition is the low power neutron noise data collected at Fort Calhoun.
The low power data identifies any changes in the mechanical pre-load of the reactor supporting mechanism and provides the earliest warning of a possible change in the effectiveness of the position-ing pins.
A number of factors influence the effectiveness of the positioning pins.
The positive factors, those which increase the coupling be-tween the core support barrel and the thermal shield, are tha ini-tial mechanical preload and the differential thermal expansion force which increases with reactor power levels.
If the reactor approaches an isothermal condition (i.e., low power), the differen-tial thermal expension force approaches zero and the initial me-chanical preload would be the sole force saintaining positive load on the positioning pins.
By comparing the low power data to the full power excore data, an evaluation can be made as to the ade-quacy of the initial mechanical preload and any changes in the pre-load from previous cycles.
The use of this technique at Fort Call: pun is further described in, the SHORN V paper "The Use of Excore Neutron Noise at Near Zero Reactor Power to Monitor Thermal Shield Support Sys-tem Integrity." OPPD periodically performs low power IVM measure-ments when returning to power operation following a refueling out-age. As described in the SHORN paper, evaluation of low power data at Fort Calhoun has shown no indication of degradation of the thermal shield support system.
Based on the previously discussed items, OPPD believes that ther.
mal shield support degradation has not taken place and that the thermal shield supports are currently in the as built condition.
O ORNL POST-FAILURE ANALYSIS The NRC Safety Evaluation of the proposed threshold levels used the results of the post-failure analysis performed by Oak Ridge National Laboratory (ORNL) of neutron noise data taken from St.
Lucie.
The ORNL post failure report indicated that the degrada-tion of the thermal shield supports was manifested by small shifts (less than 2 Hz) in the frequency of several resonances in the neu-tron noise and the largest frequency shifts occurred during the second cycle.
It is believed that this led ORNL to conclude that thermal shield support degradation occurred early in plant life at St. Lucie.
The ORNL analysis, reported in ORNL/NRC/LTR 85/24, was based on changes viewed in IVM Power Spectral Density plots over time.
The ORNL analysis identified a small frequency change (loss than 2 Hz) in a frequency as being the identifier of support degradation.
OPPD, along with Combustion Engineering, believes that the fre-i quency selected by ORNL is not related to core barrel motion, i
r OPPD believes that the ORNL report was used as the basis for the NRC conclusions of the threshold levels of the Fort Calhoun IVM program and does not believe that this is the correct conclusion.
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The following is a discussion of what OPPD and CE believe is the correct post failure analysis of the St. Lucie event.
Combustion Engineering (CE) report CEN 272(F).P (Reference 6) util.
ized calculated natural frequencies and mode shapes of the coupled core support barrel and thermal shield system for St. Lucie to determine changes in the vibration characteristics with assumed I
changes in the condition of the thermal shield support system. A comprehensive finite element model was developed to analyze differ-ent support cases such as; as built conditions, with positioning l
pins removed, and with certain support lugs removed.
Significant changes in the cos 28 shell mode of vibration were calculated j
vith sinulated support system degradation.
i In the as built condition, the supports fully couple the thermal l
shield to the core support barrel causing the system to achieve j
approximately the characteristics of the core support barrel
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For the no positioning pins condition, the bottom of the t
l thermal shield is uncoupled from the core support barrel, allowing l
the partially coupled thermal shield to tend towards its own char.
6 acteristics. With lug damage the two components uncouple, allow-i ing the thermal shield to achieve approximately the characteris-
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ties of the thermal shield alone.
The frequencies for the cos 2f i
shell mode for these three support cases are 7.6 Hz, 5.1 Hz, and l
3.1 Hz, respectively.
For the same three support condition cases, l
i the cantilever beam mode of the core support barrel and the ther-mal shield remains essentially constant:
6.8 Hz, 6.7 Hz, and 6.7 Hz.
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By comparing the cos 2f shell mode frequencies calculated using the finite element model to the actual operational data taken from j
i St. Lucie, the different types of support mechanism degradation states can be traced.
The results indicate that the positioning
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pins had not lost their effectiveness until the beginning of Cycle 4 and that support lug wear was first evident in the middle of l
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O ORNL POST-FAILURE ANALYSIS (Continued) i i
Cycle 5.
The cos 2# shell mode frequency did not change in the second cycle as reported by ORNL. Also, the frequency shifts t
of the cos 2# mode associated with degradation of the thermal shield supports was manifested by rather large shifts (between 2.5 and 4.5 Hz) in frequency.
For the above reasons, OPPD believes that the ORNL report conclu.
sions should not be used in the evaluations of the proposed thres-hold levels. OPPD utilizes the same finite element analysis that was performed for St. Lucie in developing the IVM threshold levels for the Fort Calhoun Station which is discussed in detail in.
O IVM AND LPM FNALUATION The NRC Safety Evaluation pointed out two items of concern about I
IVM and LPM evaluations performed at the Fort Calhoun Station.
The first item deals with the RMS threshold values calculated in the IVM analysis while the sacond item decis with the frequency of j
loose parts monitoring data analysis.
t One item under Section 2.3 of the Safety Evaluation discusses the derivation of RMS threshold values. This section of the evalua-tion implies that OPPD only performs RMS level determinations of the neutron noise data.
OPPD performs a complete analysis of each set of neutron noise data that is acquired. This data analysis l
i includes but is not limited to RMS threshold levels.
Power Spec-tral Densities, Cross Power Spectral Densities, Coherence, and Phase plots are used as other methods of analysis along with the Phase Separated Power Spectral Densities from which the RMS levels j
are determined. Additional information on the data analysis and i
the RMS threshold value derivation can be found in Section II of I
the OPPD report cited in Reference 3.
I A second related item that warrants clarification is the frequency j
of loose parts monitoring data analysis.
The NRC evaluation
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stated, "In the proposed surveillance of neutron noise, LPM data l
is utilized only after six successive months of monitoring indi-cate a loss of effectiveness of positioning pins."
This statement t
i is not true and can be clarified by reviewing the previous OPPD
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submittals to the NRC.
The original August 28, 1986 submittal i
(Reference 1) and the February 25, 1988 threst.old level submittal i
(Reference 4) clearly states that LPM data is analyzed on the same j
quarterly schedule as IVM data.
OPPD utilizes LPH data on the same schedule and in conjunction with IVM data to obtain the neces-sary information to access the current internals condition.
l ITEMS OF CIARIFICATION
SUMMARY
The previous five items were clarified due to the items recurring I
in the NRC Safety Evaluation.
The clarification will now allow OPPD to directly address the two questions raised by the NRC.
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ILI)g METHOD ON VHICH THE FORT CAU10UN IVM IS BASED REASONARQ1 The NRC Safety Evaluation questions five key areas in assessing the effective-ness of detecting internals degradation.
Each of the five areas will be ad-draased in detati in the following discussion. OPPD believes that there is solid evidence that Fort Calhoun has the required expertise to complement the IVM hardware and software to effectively diagnose thermal shield support degrad-ation. This statement is based on the following five areas.
O THE TRAINING AND EXPERIENCE OF THE OPERATOR Ths NRC safety evaluation questions the training and experience of the personnel responsible for the Fort Calhoun Station LPM and IVM programs. Combustion Engineering is responsible for completing the analyses with OPPD personnel evaluating the results.
The individual from OPPD responsible for the LPH and IVM programs at Fort Calhoun has over four years of nuclear plant experience and has attended the previous three U.S. reactor diagnostic con.
forences.
The theme of these conferences is new LPM and IVM tech-nologies.
The individual has completed par;*cipation in the Com-bustion Engineering Owners Group program for LPM and IVM educa.
tion.
The Owners Group developed good practice manuals for both programs and also provided in depth training on both technologies.
The individual coauthored the attached technical paper presented at the 1987 SMORN conference entitled, "The Use of Excore Meutron Noise at Near Zero Reactor Power to Monitor Thermal Shield Support System Integrity."
As part of the recent reorganization of the nuclear operations de-partments, OPPD has assigned this individual as the System Engi-neer responelble for the IVM and LPH systems.
This assignment is an assurance that OPPD is committed to providing the time and re-l sources necessary to maintain and improve the LFM and IVM programa at Fort Calhoun.
IEE_ EXISTENCE OF A SUITABLE "BASELINE" LIBRARY OF SPECTRA The "baseline" spectra library for Fort Calhoun was presented in the August 28, 1986 OPPD submittal (Reference 1).
Section 6.3 of this report examines neutron noise data acquired at Fort Celhoun during the period from 1974 through 1986.
The SER states that Maine Yankee recorded data monthly for only one fuel cycle. OPFD believes that the "baseline" library of spectra for Fort Calhoun is more extensive and thst the August 28, 1986 report should be referenced for additional information on the subject.
ADFOUATE ADMINISTRATIVE CONTROLS The Fort Calhoun Station uses auequate administrative controls to ensure high quality data, periodic engineering staff review, and transmittal of results to plant management.
The LPH and IVM pro-cedures are administered by the Preventive Maintenance program at Fort Calhoun.
The procedures are routinely performed by qualified technicians at the station to assure high quality of the data.
r 4
O M EQUATE ADMINISTRATIVE CONTROLS (Continued)
The procedure also requires that the results of the data analysis be transmitted to the plant managems,c for review.
O DIAGNOSTIC PROCEDURES The diagnostic procedures used for investigating changes in the noise signals that are not described by the baseline library are stated in the threshold level submittal (Reference 4).
The diag-nostic procedures propose to increase monitoring of IVM data from a quarterly basis to a monthly basis This increased surveillance allows for adequate trending of the threshold levels and provides a status of the degree of support mechanism degradation.
The levels of support mechanism degradation are directly related to the RMS threshold levels to be discussed in detail in Attachment 2.
The diagnostic procedures also allow for increased monitoring of loose parts monitoring data. The LPM data is analyzed on the same schedule as IVM data.
O EIANT MANAGEMENT ACTION The administrative controls require that the results of the IVM and LPM analysis be transmitted to the plant management. Any changes in the IVM or LPM data will be noted and appropriate actions recommended.
The appropriate actions, such as increased monitoring for diagnostic purposes or plant shutdown for a thermal shield support inspection, will be recommended as stated in the threshold level submittal (Reference 4).
ARE THE PROPOSED THRESHOLD LEVELS FOR DETECTING THERMAL SHIELD SUPPORT DECRADA.
TION RFASONABLE7 This section of the NRC safety evaluation raises several questions on the rea-sonab).eness of the proposed threshold levels for detecting possible thermal shield.upport degradation.
Several of these questions, such as the long term frequency behavior of the 12.5 Hz resonance and the lack of coordination be-tween the LFH and IVM systems, have already been answered in the Items of Clar-ification section of this report.
Other questions raised in the safety evalu-ation, such as the questioning of the 100% increase in RMS levels when dagrad-ation occurs and the teasoning behind basing the threshold levels on the 6 to 10 Hz range, will be answered in Attachment 2.
Attachment is a detailed de-scription of the methodology used to develop the Fort Calhoun threshold levels.
The purpose of this section is to provide a brief summary of the method used to develop the Fort Calhoun Station threshold levels.
The threshold level method used at Fort Calhoun is based on the experience gained during the failure mechanism analysis of the St. Lucie thermal shield.
The basic approach used for the threshold levels are to monitor the cos 2f shell mode of the core support barrel and thermal shield.
For the as built con-dition, the frequency of this mode is calculated and shown in actual operation-al data to be at 12.5 Hz.
For the condition of loss of effectiveness of the positioning pins, the frequency of this mode is calculated to be 7.9 Hz.
When the 12.5 Hz mode disappears, the power at that frequency mode will appear at 7.9 Hz.
The 7.9 Hz in phase mode would tend to be hidden by the out of-phase beam mode at 7 Hz in a standard Power Spectral Density (PSD) plot.
Ey eliminating the out of phase portion of the PSD using established phase separ-ating techniques, the cos 2f in phase mode can be easily tracked.
Looking for the disappearance of the cos 28 shell mode at one frequancy and its reappearance at a different frequency provides the most meaningful method for monitoring the condition of the thermal shield support system.
To provide a threshold value for aonitoring in response to the NRC request, the power asso-ciated with the as built cos 24 in phase mode was quantified using previous IVM operational data.
The frequency associated with degraded thermal shield support conditions for the cos 2d mode were calculated using a detailed fin.
ite element model.
The power associated with degraded thermal shield support conditions is assumed to remain constant and shift to the calculated frequen-cies.
Based on the comparison of the two cases, a percent of change in RMS levels for the 6 to 10 Hz range was devised. The acceptable and unacceptable percent of change levels are explicitly stated in the threshold level submittal (Reference 4) Fort Calhoun IVM Threshold Level bevelopment, should be refer-1 enced for additional information on the development of the threshold levels for the Fort Calhoun Station.
CONCLUSION Based on the cases presented in this report, the Fort Calhoun Station noise mon-itoring program would be able to detect degradation of the thermal shield sup-ports before any damage took place. OPPD possesses the expertise and adminis-trative controls necessary to ensure an effective IVM program. OPPD has ad-dressed the specific points raised in the NRC safety evaluation and requests the NRC acceptance rf the Fort Calhoun Station IVA threshold levels.
REFERENCES 1.
Letter from R. L. Andrews (OPPD) to USNRC, "Fort Calhoun Thermal Shield Support System Inspection Deferral," dated August 28,1986 (LIC-86 421).
2.
Letter from W. Paulsen (USNRC) to R. L. Andrews (OPPD), "Thermal Shield Support System Inspection Deferral, Fort Calhoun Station, Unit No.
1,"
dated February 12, 1987.
3.
Letter from R. L. Andrews (OPPD) to USNRC, "Additional Information on the Fort Calhoun Internals Vibration Monitoring System," dated October 13, 1987 (LIC 87 673).
4.
Letter from R. L. Andrews (OPPD) to USNRC, "Threshold Levels for the Fort Calhoun Internals Vibration Monitoring System," dated February 25, 1988 (LIC 88 091).
5.
Letter from P. D. Milano (USNRC) to K. J. Morris (OPPD), "Request for Additional Information Relating to the Operation of the Internal Vessel Monitoring (IVM) System," dated Jaly 14, 1988.
6.
CEN 272(F), "Final Report on the St. Lucie Unit 1 Post Cycle 5 Plant Recovery Program," dated February 1984.
7.
Combustion Engineering Report TR-FIS 017 "Floridt Power and Light, St.
Lucie Unit, Reactor Internals Inspection Report, dated March 25, 1982.
ATTACHMENT 1 The Use of Excore Neutron Noise at Near Zero Reactor Power to Monitor Thermal Shield Support System Integrity Joseph W. Quinn Combustion Engineering, Inc.
Christopher J. Sterba Omaha Public Power District Joel A. Stevens Combustion Engineering, Inc.
Keywords:
Themal shield, excore detectors, neutron noise, near isothemal conditions, phase separated PSD.
Abstract Several nuclear reactors which incorporate a thermal shield design have experienced degradation of the themal shield support structure.
Combustion Engineering, Inc. (C-E) and the Omaha Public Power District (OPPD) have developed both equipment and a surveillance program which monitors the performance of the themal shield support structure at the OPPD Fort Calhoun nuclear station.
Back.1round:
Both the ASME and the American National Standards Institute (ANSI) in their joint standard have accepted a consisten*, program of Internals Vibration Monitoring (IVM) using neutron noise to be a reliable reactor internals surveillance method (Ref. 1).
The U. S. utility, OPPD, has developed an administrative procedure to monitor reactor internals vibration levels using existing excore neutron detector signals at the Fort Calhoun station, located north of Omaha, Nebraska.
The Fort Calhoun reactor (Figure 1) is a 502 MWe Pressuri:ed Water Reactor (PWR) designed by C-E.
In operation since late 1973 the reactor internals include a Core Support Barrel (CSB) ano a Themal Shield (TS).
During the scheduled ten year In-service Insoection (ISI) perfomed in 1983 a visual examination was made of the CSB/TS support structures.
The examination indicated that all support components were in good condition.
OPPD has been actively involved in a regular program of IVM surveillance of reactor internals metion since 1985.
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Core Barrel Themal Shield Descriotion:
Figure 2 details the arrangement of the Ft. Calhoun reactor internals.
The themal shield is a right circular cylinder concentric with the core barrel. The thermal shield is attached to the core barrel and extends the length of the active core.
The themal shield is suspended from support lugs on the core support barrel.
Radial positioning is provided by two sets of positioning pins.
Figure 3 describes the CSB/TS connection scheme.
CSB/TS Dvnamic Behavior:
Because of the si;pport lugs and the radial positioning pin the CSB/TS form a closely coupled mechanical structure which exhibits identifiable beam
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The beam bending mode (Figure 4a) is a cantilever mode of vibration of the CSB, similar to a simole beam with one end free and one end clamped.
Ir. this mode the CSB cross section remains circular and translates.
The shell modes of vibration (Figure ab, ac) are vibration modes involving circumferential variation in the shape of the CSB.
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ces 44 Z2.8 Column #1 in Table 1 lists the calculated in-water modal frequencies of the Fort Calhoun CSB/TS (Ref. 2).
These frequencies can be identified in a typical Auto Power Spectral Density (APSD) taken from the Ft. Calhoun reactor at 100% power (Fig. 5).
Analysis of the CSB/TS support structure (Ref. 3) indicates that a loss of radial positioning pin effectiveness can initiate the decoupling of the CS8/TS.
This decoupling effects vibration modal frequencies. The detection of radial positioning pin condition is possible using the neutron noise portion of the existing linear power range detectors.
Column #2 of Table 1 lists the calculated in-water modal response frequencies for the Ft. Calhoun CSB/TS in the case of a postulated loss of effectiveness of all rad.al positioning pins.
Based on comparison between "Nominal" and "All pins removed" cases, it can be observed that only the first shell mode cf vibration (C05 20) changes significantly from its nominal 12.5 Hz down to 7.9 Hz with all pins removed.
By monitoring the frequency of the CSB/TS first shell mode one can infer changes in positioning pin effectiveness.
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Reactor Power and Positioning Pin loading:
A number of factors influence the effectiveness of the radial positioning pin including mechanical preload and differential thermal expansion, both of which increase the coupling force between the thermal shield, and the CSB.
The pressure differential across the thermal shield, support lug bending and radiation induced relaxation all decrease the CSB/TS coupling force.
The themal expansion force is the only positioning pin loading force which can be controlled during reactor operation.
Increased temperature differential between the core support barrel and the thermal shield increases the interference between the positioning pin and the CSB.
Power Level Selection:
In order to monitor positioning pin effectiveness it was desired to select plant operating conditions for which the CSB/TS would be most susceptible to any reduction in positioning pin effectiveness.
If the reactor is in an isothermal condition the themal expansion force is zero and the mechanical preload given the positioning pins during installation would be the sole force maintaining a positive load on the positioning pins.
Zero rector power produces ideal isothemal conditions between the CSB/TS components. However, zero reactor power does not provide an adequate neutron flux to allow the linear power range excore detectors to be used.
If a simple measurement scheme using existing instrumentation is desired then a near zero power level must be selected which:
1)
Produces a minimal thermal expansion force.
2)
Provides an adequate excore detector flux signal.
The themal expansion force exerted by the CSB on the positioning pins is:
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R = Core barrel outer radius expansion for the core Lo = Posit!oning pin length t,arrel. thermal shield, ai = Temperature difference between CBS and TS.
Since sT is nearly lirear with reactor power the thermal expansion force is also nearly linear with reactor power. Usingthislinearrelationthe themal expansion force at 5% reactor power wouid only be 1/20th the 100%
themal expansion force.
From a themal expansion force consideration the lower the reactor power the better.
Reactor power below 5% does not produce significant thermal expansion force.
i The other consideration in selecting a reactor power level was would there be sufficient vibration related neutron noise signal to allow a successful measurement using the plant linear power range detector signals.
The background noise. level is made up of uncorrelated noise sources, not related to the vibration induced fluctuations, found anywhere along the excore detector signal path. This background level has an APSD which is constant at all frequencies and has a constant voltage value. The vibration related fluctuations measured by the excere detectors are linear with reactor power.
Our measured signal is their sum of:
X (signal) = vibration (power) + BACXGROUND (constant)
Since the background level is constant it becomes a progressively larger percentage of the measured signal as the reactor power level decreases.
Figure 6 is a composite of APSD's for three different power levels (30%, 10%
and 5t).
In all three traces the vibration portion of the sional remains a relatively constant percentage of the O.C. power.
The background noise is constant regardless of reactor power level and therefore consistently becomes a larger percentage of the signal as the power level decreases.
A problem exists with measuring shell mode frequencies.
The shell mode vibration scale factor, (% neutron flux / mil of motion), is smaller by a factor of three (3) thar the scale factor for the beam mode (Ref. 4).
Coupled with this smaller shell mode scale factor is the smaller shell mode vibration amplitudes.
Together these two factors produce a detected shell mode neutron flux signal which is about a factor of one h*Jndred (100) smaller than the beam mode response in the APSD.
Examining Figure 6 at the beam and shell frequencies it is quite easy to identify the beam mode frequency at all three power levels. However, it becomes increasingly difficult to identify the first shell mode as the reactor power decreases.
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Five percent reactor power was chosen to be the plant power level for the positioning pin effectiveness measurement. The five percent power level is both the highest power level at which themal conditions are not appreciable and the lowest reactor power at which CSB/TS shell mode frequencies can be detected.
i Data Acquisition System:
The measurement of reactor internals vibration datt requires specialized signal conditioning equipment.
C-E has developed an Internals Vibration Monitoring (IVM) system which provides all of the necessary electronic equipment and analysis functions needed to perfom regular IVM measurements.
The IVM system contains a Signal Conditioning Module (SCM) which houses the processing electronics, a portable computer, which centrols the SCM and comDutes the analysis functions, and a graphics hard copy printer.
The Signal Conditioning Module accepts two channels of excore signals as inputs and outputs the bandlimited and amplified vibration fluctuations which make up less than 0.25% of the gross power level excere signal.
The C-E SCM (Fig. 7) provides:
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i The SCM is totally controlled in all its ranges and functions by the computer.
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The output of the SCM is then processed by the IVM analysis snftware package and analysis results are stored en both disk and hardcop,w grapnic
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printer output.
I The C-E IVM analysis software acoutres both time and frecuency domain data. The standard IVM analysis functions include: Auto and Cross Power j
Spectral Density functions, signal Coherence and Relative Phase functions as t
well as user specified band select RMS (root-mean-souare) energy calculations.
Also available is the implementation of spectral phase separation algorithms.
The tVM software can alsu recall previous computed analysis results to allow ccmparison between two analysis data sets.
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In June 1986 a near zero neutron noise measurement was made at Ft.
Calhoun plant. This measurement was made at nominally 5% reactor power using the C-E IVM system. Two linear excore detectors 180* apart were measured and i
analyzed. The signal obtained from the first shell mode exhibits an in-phase relationship between detectors which are 180' apart (Fig. 8).
In addition to l
the nonnal spectral analysis a special phase separated APSD (Ref. 5) was l
i calculated. This phase separation technique applies to the AP50 processing for which the measured APSD is the sum of two processes which are either in phase (0') or out-of-phase (180') relative to each other.
l For this special case the measured APSD can be separated based on phase into APSD (0*) and APSD (100*).
Since our interest was in the CSB/TS first shell mode frequency, the in-phase. APSD(0*) spectra, would be the most sensitive to change of that vibration mode.
Figure 9 is the 0*-phase APSD from the 5% reactor power measurement at Ft. Calhoun.
The frequency of this i
vibrational mode is still noted to be in the same 10-12 Hertz region as in the 100% power APSD (Fig. 5). The inference from this analysis was that the i
CSB/TS support system was effective.
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Ft, C thoun Q* Phase AP10 at $1 Reactor fewer.
Inspection Deferred Th'* ouch Monitorino l
In 1984, the U. S. Nuclear Regulatory Comission (USNRC) concern over CSB/TS support system problems led OPPD to comit to the USNRC to perfom an examination of the Ft. Calhoun CSB/TS during the 1987 maintenance outage.
f This examination would have been perfomed four years after the nomal ISI ten OPPD petitioned the USNRC in August 1986 for deferral of year inspection.
this examination until the next scheduled ISI inspection to be perfomed in 1993.
I The deferral effort consisted of both an analytical effort and a comitment to an IVM surveillance program.
Emphasis in the deferral effort was also given to the positive results of the near :ero power !YM measurement l
taken in June 1986.
l In February 1987 the USNRC. granted OPPO a deferral from CSB/TS examination until the scheduled 1993 ISI date. OPPD's commitment to the USNRC included the analysis of IVM data on a quarterly basis and collecting near
[
isothemal IVM data once per fuel cycle.
Conclusion The early detection of themal shield support system structural change is IVM measurements at the primary goal of the OPPD IVM program at Ft. Calhoun.
near :ero reactor power provide an early warning of a possible change in radial positioning pin effectiveness by monitoring changes in the CSS /TS
~r vibration characteristics.
The OPPD surveillance program provides a reliable I
early warning of potential structural problems with a minimal imoact on plant operations.
P l
I References 1.
ASME/ ANSI OM-5, "Inservice Monitoring of Core Barrel Axial Preload in Pressurized Water Reactors",1981.
2.
C-E Report, "Summary Report on Ft. Calhoun Thermal Shield Inspection Deferral Progra'a", CEN-334 (0) Rev. 01, 1986.
3.
Lubin, B., Longo et al, "Analysis of Vibration Monitoring and Loose Parts f
Data Related to the St. Lucie 1 Thermal Shield Failure", Presented SMORN Y 1987.
1 l-4 Kosaly, G., "Noise Investigations in Boiling Water and Pressurized Water Reactors", Progress in Nuclear Enerov, 1979.
sis of Pressurized Water Mayo Charles W., "Detailed Neutron Noise Analy(ATKE), 1977.
5.
Reactor Internals Vibration". Atomkernenergie I
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ATTACHMENT 2 FORT CALHOUN IVM THRESHOLD LEVEL DEVELOPMENT BACKGROUND The technical approach utilized to develop the Fort Calhoun IVM threshold levels is based on the experience gained during the failure mechanism analysis of the St. Lucie 1 thermal shield (Reference 1).
In the St. Lucie analysis program, vibration structural response calculations were performed on a detaileo finite element model of the core support barrel, thermal shield and thermal shield sup-ports as a coupled system. These calculations were performed for the system in its nominal as designed condition and in assumed degraded conditions. A signi-ficant change in frequency for a particular shell mode of vibration was noted from the calculations with the simulated loss of all positioning pin effective-ness and further change occurred when simulated support lug damage was intro-duced into the calculation.
St. Lucie Loose Parts Monitoting (LPM) and Internal Vibration Monitoring (IVM) data was reviewed and reanalyzed to determine if there were quantifiable changes in the data during plant operation prior to actual degradation occurring.
LPH data shows changes in both magnitude and frequency of the Japact signals and changes in location of the signals with operating time.
Phase separation analy-sis of the IVM data showed changes in an in phase shell mode frequency with time.
The initial frequency and the corresponding frequency reductions found in the degraded condition in the IVM data showed excellent correlation with the val-ues calculated for the nominal and assumed degraded support system conditions.
The available evidence (summarized graphically in Figure 1) indicates that the thermal shield support system degradation process is lengthy, and that the pro-cess is detectable by the methods of excore neutron noise monitoring and loose parts monitoring systems.
FORT CAlllOUN FREOUENCY CALCUIATIONS Experience gained from the St. Lucie Unit 1 program demonstrated that degrada.
tion in the thermal shield support system were manifested as frequency peak changes in the spectra of the excore detector noise signals.
Therefore, anal-ytic predictions of changes in frequencies and modes with assumed changes in the thermal shield support system are used to interpret the data acquired from the excore detector signal monitoring program.
A detailed three dimensional finite element model of the core support barrel, thermal shield and thermal shield support system has been developed for Fort calhoun (see Figure 2).
Natural frequencies of the coupled system were calcu.
lated by means of the SAP 4 computer program for the nominal support system case, for the loss of effectiveness of all positioning pins case, and the re-moval of the structural restraint provided by four support lugs.
The results of these frequency calculations are summarized in Table 1.
For Fort Calhoun, as was also the case for St. Lucie Unit 1, there are signifi-cant changes in frequency of the cos 28 mode of vibration of the thermal shield and core support barrel system when simulated damage of the thernal shield support systen is introduced into the analysis. This cos 28 mode, which results in in phase relationships between cross core pairs of detectors, provides the requisite identifier of the condition of the thermal shield support system which can then be tracked via the IVM system.
FORT CAIFOUN CYCLE 10 IVM DATA Four sets of 1004 power neutron noise data was analyzed for Fort Calhoun from Fuel Cycle 10.
Data from the cross core detector pairs B safety - C safety and B control A control were acquired on 4/30/86, 5/28/86, 7/28/86 and 10/18/86.
The four data sets were analyzed using phase separating techniques (Reference 2) over the frequency bands 0 4 Hz, 4 6 Hz, 6-10 Hz and 10-15 Hz.
These bands were chosen so as to contain the Power Spectral Density (PSD) peaks corresponding to the initial (nominal condition) frequency (12.5 Hz) and the reduced (degraded condition) frequencies (7.9 Hz and 5.4 Hz) associated with the cos 2f mode used to monitor the condition of the thermal shield supports.
The Root Hean Square (RMS) values of amplitude for the phase separated PSD's for the chosen frequency bands were calculated (see Table 2).
The values listed in the columns headed 10-15 Hz provide an estimate, as a function of time for the various detectors, of tne amplitude of the 12.5 Hz peak of the in phase (O') cos 2f mode associated with the nominal condition of the thermal shield support system.
It is well known (e.g., see References 3, 4, 5 and 6) that neutron noise not re-lated to internals vibration can vary throughout a fuel cycle and from cycle to cycle. These variations may be related to fuel burnup, soluble boron concentra-tion, temperature and modifications in fuel management of design.
The varia-tions in the Fort Calhoun data during cycle 10, calculated as a percent increase over the 4/30/86 data, are shown in Table 3.
Note that the largest increase over any three month period for the 6 10 Hz band was 14.5% and that the largest increase over any one month period for the 4 6 Hz band was 6.3% ((27.8 8.9) +
3).
ESTIMATE OF EXPECTED CHANGES IN IVM DATA VITH POSTULATED CHANGES IN THERMAL SHIELD SUPPORT SYSTEM CONDITIONS Natural frequency calculations performed for Fort Calhoun, summarized in Table I
1, show significant changes in the frequency of the cos 2# mode of vibration of the thermal shield and core support barrel with postulated damage to the ther-mal shield support system.
For the nominal, as designed condition of the ther-mal shield support system, the frequency is 12.5 Hz; for the case of loss of effectiveness of all positioning plus, the frequency is 7.9 Hz; and for the case of support lug / support pin wear, the frequency is 5.4 Hz.
With the postulated loss of effectiveness of all positioning pins, the power associated with the 12.5 Hz peak will move to a peak at 7.9 Hz.
With the post-ulated addition of support lug / support pin wear, the power originally associated with the 12.5 peak will decrease further to a peak at 5.4 Hz.
On this basis, estimates of expected changes in neutron noise data with the post-ulated changes in thermal shield support condition have been calculated.
The results are giver. in Table 4, the case of loss of effectiveness of positioning pins (6 10 Hz band) and the case of support lug / support pin wear (4 6 Hz band).
For example, for the C safety detector, the RMS amplitude of the 6 10 Hz band is estimated to increase from 100% to 1176, as a function of time, over its nominal value with postulated loss of effectiveness of all positioning pins.
Also, for the same detector, the RMS amplitude of the 4 6 Hz band is estimated to increase from 436 to 51% over its nominal value with
FORT CAIAOUN IVM THRESHOLD VALUES A threshold value is proposed based on the following suggested approach:
1.e.,
that a threshold value be chosen that reduces the likelihood of wear on the thermal shield support lugs / support pins.
Experience at St. Lucie 1 indicated that approximately two years had passed between the time that loss of effective-ness of the positioning pins was detectable by IVM and the time that support lug / support pin wear had occurred.
It is proposed, therefore, that a threshold value be selected that results in an inspection program about six months after detection of loss of effectiveness of all positioning pins.
This can be done as follows:
1.
Acquire, reduce, and evaluate 1004 power IVM data at the beginning, end, and at three month intervals for each fuel cycle.
Calculate the in phase (O') RMS amplitudes of the 0 4 Hz, 4 6 Hz, 6 10 Hz, and 10 15 Hz frequen-cy ranges.
2.
amplitude for the 6 10 Hz range for the present data to that of the previous quarterly data.
3.
If there is a definite 12.5 Hz in phase peak in the PSDs, and the in-creasu in the RMS amplitude for the 6 10 Hz range is less than 254, con-tinue to monitor at three month inteirvals.
If both of these conditions are met, the thetual shield is adequately supported.
(The 256 increase is based on an expected three month increase associated with fuel burnup of 154 plus a potential 104 IVM system amplitude uncertainty).
4.
If there is a greater than 254 but less than 1004 increase in the RMS amplitude for the 6-10 Hz in phase range from the previous quarterly 4
data, begin acquiring, reducing, and evaluating data at one month in-tervals.
The only action to be taken at enis time is increased monitor-ing for trending purposes.
No inspection is recommended at this time.
5.
If there is no definite 12.5 Hz in phase (O') peak in the PSDs, and if the increase in the RMS amplitude of ths 6 10 Hz range is greater than 1004, the changes in the neutron noise data indicate a loss of effective-ness of the positioning pins.
Begin to monitor IVM data at one month in-t i
tervals for trending purposes. An inspection of the thermal shield sup-port structure is recommended.
The inspection should occur within the next six months, i
6.
If six successive months of monitoring indicate a loss of effectiveness of positioning pins and if this is corroborated by evaluation of Loose Parts Monitoring (LPM) data, it will be recommended that the plant be
]
shut down and an inspection of the thermal shield support structure be performed as soon as practicable.
LPM data is acquired, reduced and evaluated on the same schedule as that for IVM.
Near zero power IVM data is acquired, reduced and evaluated once per fuel
- cycle, l
I
REFERENCES 1.
CEN 272(F), "Final Report on the St. Lucie Unit 1 Post Cycle 5 Plant Recovery Program," dated February 1984.
2.
"Detailed Neutron Noise Analysis of Pressurized Water Reactor Internals Vibration," Charles W. Mayo, Atomkernenergie 22 pp. 9 13 (1977).
3.
"Analysis of Changes with Operating Time in the Calvert Cliffs Unit 1 Neutron Noise Signals," J. P. Steelman and B. T. Lubin, SHORN II, September 1977.
4.
"Review of Borse11e PVR Noise Experiments, Analysis and Instrumenta-tion," E. Turkan, SMORN II, October 1981.
5.
"Use of Neotron Noise for Diagnosis of In. Vessel Abnomalies in Light.
Vater Reactors," D. N. Fry, et. al., NUREG/CR 33303 (ORNL/TH.8774)
January 1984, 6.
"Contribution of Fuel Vibrations to Ex Core Neutron Noise During the First and Second Cycles of the Sequoyah 1 Pressurized Water Reactor,"
F. J. Sweeney, et. al., SMORN IV, October 1984, i
l 1
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JSM057/RL TABLE 1 FORT CALHOUN THERMAL SHIELD, CORE SUPPORT BARREL AND THERMAL SHIELD SUPPORT SYSTEM IN-WATER MODAL FREQUENCIES (HERTZ) vs. SUPPORT SYSTEM CONDITION All Pins Removed &
Mode Nominal All Pins Removed 4 lugs Removed BEAM 7
7 7
COS 20 12.5 7.9 5.4 COS 30 16.3 14.9 14 C05 40 22.8 22 21.3
TABLE 2 RMS AMPLITUDE OF PHASE SEPARATED PSD's FOR CHOSEN FREQUENCY BANDS B SAFETY C SAFETY 3
3 DATE 0' PHASE PSD (RMSx10 )
DATE 0' PHASE PSD (RMSx10 )
(1985)
FOR THE FREQUENCY BANDS (1986)
FOR THE FREQUENCY BANDS 0-4Hz 4-6Hz 6-10Hz 10-15Hz 0 4Hz 4-6Hz 6-10Hz 10-15Hz 4-30 25.708 1.982 0.883 0.924 4-30 27.330 2.090 0.898 0.901 5-28 27.407 2.011 0.919 1.008 5-28 28.858 2.184 0.949 0.965 7-28 30.594 2.234 0.982 1.138 7-28 31.179 2.460 1.020 1.106 10-28 33.681 2.311 1.058 1.360 10-28 33.605 2.529 1.112 1.302 A CONTROL 8 CONTROL l
OATE 0' PHASE PSD (RMSx10 )
DATE 0' PHASE PSD (RMSx10 )
3 (1986)
FOR THE FREQUENCY BANDS (1986)
FOR THE FRE0VENCY CANDS j
0-4Hz 4-6Hz 6-10Hz 10-15Hz 0-4Hz 4-6Hz 6-10Hz 10-15Hz 4-30 27.530 2.013 0.907 1.411 4-30 26.975 2.055 1.151 1.299 5-28 30.806 2.096 0.981 1.546 5-28 28,888 2.140 1.218 1.412 7-28 33.617 2.193 1.002 1.545 7-28 31.003 2.372 1.318 1.532 10-28 37.048 2.573 1.018 1.765 10-28 36.275 2.565 1.414 1.744 l
TABLE 3 PERCENT INCREASE IN RMS AMPLITUDE OF PHASE SEPARATED PSD's FOR CHOSEN FREQUENCY BANDS _
8 SAFETY
% CHANGE W/ TIME C SAFETY t CHANGE W/ TIME DATE 0* PHASE PSD RMS INCREASE (1)
DATE 0' PHASE PSD RMS INCREASE (1)
(1986)
FOR THE FRE0VENCY BANDS (1986)
FOR THE FREQUENCY BANDS 0-4Hz 4-6Hz 6-10Hz 10-15Hz 0-4Hz 4-6Hz 6-10Hz 10-15Hz 4-30 4-30 5-28 6.6 0.9 4.1 9.1 5-28 5.6 4.5 5.7 7.1 7.28 19.0 I?.7 11.2 23.2 7-28 14.1 17.7 13.6 22.8 10-28 31.0 16.6 19.8 47.2 10 28 23.0 21.0 23.8 44.5 I
8 CONTROL
% CHANGE W/ TIME A CONTROL
% CHANGE W/ TIME DATE 0* PHASE PSD RMS INCREASE (1)
DATE 0* PHASE PSD RMS INCREASE (1)
(1986)
FOR THE FREQUENCY BANDS (1986)
FOR THE FREQUENCY BANDS 0-4Hz 4-6Hz 6-10Hz 10-15Hz
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