IR 05000424/1989029
| ML19332C062 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/14/1989 |
| From: | Belisle G, Burnett P, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19332C061 | List: |
| References | |
| 50-424-89-29, 50-425-89-33, NUDOCS 8911220262 | |
| Download: ML19332C062 (11) | |
Text
{{#Wiki_filter:y (;; l C ~ ..;., ' ' ' , . , '* , - - A ltig, . UNITED STATES ' , a'd, NUCLEAR REGULATORY COMMISSION
. h . REGION il - 101 MARIETTA STREET,N.W.
ATLANTA, GEORGI A 30323 , [ ...... ^ Report Nos.: 50-424/89-29~and 50-425/89-33 Licensee:' Georgia Power Company ' P._0. Box 1295 , Birmingham, AL 35201 . -r Docket Nos.:.50-424 and 50-425 License Nos.: NPF-68 and NPF-81 Facility Name: Vogtle 1 and 2 Inspection Conducted: October'16 to 20,1989 //[/4//M Inspector: Cu s r P. T. Burnett H ~~ ( 'Date ' Signed M
l ' - l L/ - U Inspector:' , J. Zeil - y ) ate Signed // ////N Approved by: % X G. A.LBelisle, Ch1 F V Da~te Sigped Test Programs Section Engineering Branch ~ Division of Reactor Safety > SUMMARY ! Scope: I; This routine,. unannounced inspection addressed the areas of core power L distribution monitoring; nuclear instrument calibrations; thermal power moni- ~ toring;c determination of ~ reactor shutdown margin, reactivity balances, and I estimated critical conditions; proposed measurements of' the moderator - temperature coefficient. at end of life; and followup on previously identified L items.
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Results: The required surveillance procedures for reactivity control systems and core power distribution limits were found to be technically adequate and performed-l at the required frequency. Review of core physics testing conducted during the current operating cycle for both units identified a few minor errors in abstracting data from the Plant ~ Technical Data Book.
This indicated the need for: closer attention to detail during the licensee's independent review of these procedures.
(Paragraph 2)
i A minor ' weakness was identified in the nuclear instrument channel calibration 18-month surveillance procedure.
Review of ion chamber current-voltage curves for acceptability was required, but no mechanism was established by Reactor , l.
Engineering personnel to feedback the results of their review into the Dbhk{{ffff , PDb . ._ . . .
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procedure.
At the exit meeting, management committed to revise the procedures to document and respond to the review.
The licensee is considering changing . the method of incore/excore nuclear instrument correlation to one not reviewed by the NRC or supported by the reactor vendor.
This method will be inspected further if implementation goes forward. (Paragraph 3) Minor weaknesses were identified in the procedure for measuring the moderator temperature coefficient near the end of the operating cycle.
At the exit-interview the licensee committed to improve the internal precision required for boron concentration measurements and to-establish an acceptance criterion for agreement between measurements made during boration and deboration.
(Paragraph 6) No violations or deviations were identified.
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> AfPORT DETAILS
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Persons Contacted Licensee Employees
- J. G. Aufdenkampe, Technical Support Manager
- H..E. Beacher, Senior Plant Engineer
- G. B. Bockhold, Jr., General Manager, Vogtle Nuclear Operations-i
- S. A. Bradley, Reactor Engineering Supervisor
, M. S. Driney. Instrumentation and Controls Supervisor.
- D..M. Christiansen, Reactor Engineer
- G. R. Frederick, Quality Assurance Site Manager - Operations D..C. Garlington, Reactor Engineer
- J. W. Godsey, Plant Engineer
M. L. Hobbs, Instrumentation and Controls Superintendent R. L. Mansfield, Plant Engineer A.LL. Mosbaugh, Plant Support Manager
- R. M. Odom, Supervisor, Nuclear Safety and Compliance
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- J. E. Swartzwelder, Operations Manager T. L. Wendt, Reactor Engineer Other. licensee employees contacted included engineers, technicians, and
office personnel.
-NRC Resident inspectors L
- R. F. Aiello, Resident Inspector
- J. F. Rogge. Senior Resident Inspector R. D. Starkey, Resident Inspector.
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- Attended exit interview on October 20, 1989, Acronyms and initialisms used throughout this report are listed in the i
' last paragraph.
2.
Core Power Distribution Monitoring (61702) The inspectors reviewed surveillance procedure 88007-C (Revision 0), Hot Channel Factor Determination, for technical adequacy and determined it to be acceptable.
This procedure is used by the licensee for verifying that the hot channel factors are within the limits as specified by TS 4.2.2.2 and 4.2.2.3.
These TSs require the determination of hot channel factors every 31 EFPD.
For Unit 1, Cycle 2, the inspectors examined the last . . . - - - . -.
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' five completed surveillances and determined that they were conducted at ' . the correct frequencies and with acceptable results.
Unit 2 Cycle 1, surveillances were reviewed from cycle startup to the date of the inspec-tion.
These surveillances were also conducted at the correct frequency i and with acceptable results. The inspectors reviewed the latest full flux ! map computer printouts for each unit and verified that no anomalies existed in the measured neutron flux data for each incore instrunent thimble location.
Normalized flux data was taken from 49 thimble locations for the Unit 1 map and 50 locations for the Unit 2 map.
A minimum of 44 flux thimbles is required for a full core flux map.
The inspectors reviewed surveillance procedure 88005-C (Revision 0), Target Axial Flux Difference Determination, for technical adequacy and it i was determine <1 to be acceptable. This procedure determines the target AFD l of each operable excore power range channel by measurement of the flux difference for each channel. TS 4.2.1.3 requires target AFD detennination by direct measurement every 92 EFPD and TS 4.2.1.4 requires the target AFD to be updated once every 31 EFPD by either direct measurement or ] interpolation. The inspectors reviewed all completed surveillance records for the current operating cycle for both units and verified that target i AFD limits were detennined at the required frequency.
Target AFD curves j were accurately generated after measurement or interpolation of the Target i AFD for all surveillances reviewed. All the surveillances were conducted i at or near full power and equilibrium xenon conditions.
No violations or deviations were identified in this area, l l 3.
NuclearInstrumentationCalibrations(61705) i a.
Revi u of Maintenance Procedures The follouing maintenance procedures were reviewed for content, clarity, and implementation of TS surveillance requirments:
(1) 24700-1(Revision 20,issueo8/28/89),NuclearInstrumentSystem Power Range Channel IN41 Channel Calibration, is functionally identical to similar procedures for the three other PRNI chan-nels on Unit I and the Unit 2 channels.
Section 4.2, Plateau l Voltage Verification, is performed at or above 20 percent RTP.
Current from each chamber is recorded as voltage is adjusted from 300 to 1500 yde in 50 ydc steps.
The data and graphs of l current versus voltage are then forwarded to Reactor Engineering l for evaluation.
In the interim, the voltage is adjusted to 800 vdc regardless of the results of the current-voltage graph or the as-found voltage for the channel.
This procedure could leave the channel at a non-optimum voltage until Reactor Engineering completes evaluation of the current-voltage data. A mechanism for Reactor Engineering feedback to set the optimum vt>ltage or to otherwise evaluate the acceptability of chamber perfonnance has not been identified.
This is a refueling interval test (TS Table 4.3-1 item 2.a note 5).
This concern . ! . -en-a- , ,, ,, _. _ _ - --.w- _ _ _ _ _. _ _ _ _ _. _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _, _ _ _ _ _ _ _. _
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3 t , i was discussed with I&C personnel.
They made a commitment to ~ document the review of the curves in this procedure and to make the procedure responsive to the results of that review.
(Inspectorfollowupitem 50-424/89-29-01and50-425/89-33-01) i , Delay time is measured and adjusted in section 4.4.18.
Data sheet 19 eecepts values from 2.0 to 2.2 seconds.
The negative
rate trip bistable calibration (NC301) is performed in section 4.4.19.
It must trip on a 5 percent decrease and must not trip.
on a 4.5 percent decrease.
The positive rate trip bistable calibration (NC303) is performed in section 4.4.20.
It must ' , trip on a 5 percent increase and must not trip on a 4.5 percent - increase.
The method of calibration avoids the miscalibration , errors observed at other facilities a few years ago.
, (2) 24697-1 (Revision 1, issued 8/28/89), Nuclear Instrumentation System (NIS) Intermediate Range Channel 1N35 Channel Calibra-
tion, is typical of procedures used for other channels on Unit I and on Unit 2.
Plateau voltage verification is performed in , section 4.2.
A power level of at least 20 percent RTP is ' required, and the current-voltage relationship is measured at 50 l , volt intervals.
The data are plotted and the plot and table ' forwarded to Reactor Engineering for review.
As was found in the review of the PRNI calibrations, there is no mechanism in , the procedure for the results and conclusions of the review to
feedback into the calibration, and the chamber. excitation !
voltage is set at 800 vdc.
The licensee made a similar
' commitment to revise the procedure with provisions for Reactor
Engineering feedback. This will be tracked by the NRC using the , above inspector followup item 50-424/89-29-01 and , 50-425/89-33-01.
The procedure was acceptable in all other respects.
. (3) 24695-1 (Revision 3, issued 3/1/89), Nuc; ear Instrument System i (NIS) Source Range Channel IN31 Channel Caitoration, is func-tionally identical to the procedures for the other source range - channels in both units.
Following discussion of this procedure with I&C personnel, all of the inspectors' questions were
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resolved.
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Calibre,tions Against Thermal Power
Completed procedures 14030-1/2, Power Range Calorimetric Channel Calibration, for Units 1/2 were reviewed for selected monthly peri-ods: August 1989 for Unit 1 and August 10 to September 12, 1989 for Unit 2.
In all cases, the required daily frequency, when in mode 1, was met or exceeded.
Control room personnel appear to be doing a good job of keeping the PRNIs in close agreement with thermal power.
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Incore/Excore Nuclear Instrunent Correlations { t Procedure 88023-C (formerly 55003-C). Incore/Excore Detector Calibra-
tions, is required to be performed every 92 EFPD by TS Table 4.3-1.
> ' Review of completed procedures confirmed that the surveillances have been performed with the required frequency and acceptable results , during the current cycle of each unit.
Recent improvements in the ' procedure are addressed in paragraph 7 under followup activities.
Recently completed procedures show good' correlation coefficients (at l least 0.98) for all channels.
The procedure does not contain an ! acceptance criterion requiring a minimum value of the correlation coefficient to assure tMt the data used were acceptable for the , purpose.
The licensee stated that they are considering changing the ! entire procedure to use a single point calibration, which will make ! the correlation coefficient issue moot.
According to the licensee, j the single point method is in use in facilities in other NRC Regions, , although, it has not been endorsed by Westinghouse for use in West-inghouse plants.
The method has not been reviewed by NRR, but will
be reviewed by Region II if implementation at Vogtle proceeds, d.
Negative Rate Trip on October 11, 1989 , ' A negative rate trip occurred on Unit 2 on October 11, 1989.
The inspector discussed the investigation of the trip with plant engi-t neering personnel.
The cause has been traced to the open failure of ' a diode, in both directions, in the stationary gripper coil circuit s of centrol rod K-2 in control rod bank B.
The concomitant failure of the gripper to hold the rod led to a negative reactivity insertion and the trip.
Westinghouse has confirmed that the reactivity worth i of K-2 was sufficient to cause a negative-rate trip in a properly , L calibrated trip circuit. Thus, the immediate response to the dropped
! rod was proper and expected.
The diode was sent to Westinghouse for i evaluation. The failure mode was found to be the same as three other recent cases of failure of the same kind of diode.
Prior to restart, all rods were functionally checked, and no other problems in rod i l drive performance were found.
Many similar diodes are in use > throughout this plant and others.
Three failures do not appear to l indicate a significant reliability problem with this diode.
No violations or deviations were identified in these areas.
4.
Thermal Power Monitoring (61706) ' Improvements in procedure 88075-C, Precision Heat Balance, are addressed in paragraph 7 under followup activities.
I On June 15, 1989, the licensee reported that Unit 2 may have operated slightly overpower due to a miscalibrated feedwater flow transmitter.
. That event and the corrective actions were discussed with plant . - -- -- . .. _, _ - _ _ - --.
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l i engineering personnel. The licensee's resolution of the issue appeared to be-adequate.
Conclusions from earlier inspections that the licensee's measurements of parameters used in thermal power analysis and the method of analysis were acceptable were reexamined and found to be still valid.
) However, neither the licensee's nor the NRC's methods of thermal power j analysis are proof against small errors in instrument calibratioa.
The i proter" on against such errors must come from the instrument installation j and c-ration procedures.
The Unit 2 response to a turbine load transient was addressed in NRC Inspection Report 50-425/89-20.
During the current inspection, two event , reports related to that transient were reviewed.
Event Report 2-89-011 l and Event Report 2-89-012 collectively address the problems with the turbine control circuitry that led to the transient on May-14, 1989, and i to a. plant trip on May 22, 1989.
The inspector concluded that the ! licensee's investigations were thorough, within the limits of the i available information, and properly identified the root causes of the events.
Analysis of the transient on May 14 was hampered by the lack of plant data.
Records from the fault recorder and the ERF computer that are , automatically saved following a trip were not saved during the transient.
l As corrective action, the licensee issued a night order to all shifts directing them to manually initiate the computer and recorder following any plant transient, which did not lead to automatic initiation.
Further.
the transient and the need to save transient data were subjects in lesson j plan RQ-LP-63103-01 in the licensed operator requalification program.
No violations or deviations were identified in this area.
5.
Determination of Reactor Shutdown Margin, Reactivity Balances, and ECCs (61707) The following procedures used in surveillance of core shutdown margin and
i I overall core reactivity balances were reviewed and determined to be - l technically adequate.
The procedures adequately implemented TS require-
ments for shutdown margin determination and control.
' L 14915-1 Revision 12), Special Conditions Surveillance Logs L 14005-1 Revision 6), Shutdown Margin Calculations l-14000-1 Revision 21), Operations Shift and Daily Surveillance Logs l 88003-C Revision 0), Shutdown Margin By Minimum Bank Height 88013-C Revision 0), Overall Core Reactivity Balance ' , Procedure 14005-1 is used to calculate the shutdown margin for all modes of operation.
Core reactivity conditions are determined from curves and tables from the PTDB.
The inspectors reviewed the last 11 completed surveillances for Unit 1 and the last 6 surveillances for Unit 2 to ensure that the calculations were performed correctly.
Three of these surveillances for each unit were reviewed in detail to assure that correct data was taken from the PTDB tables.
From this review, three minor discrepancies were identified in the Unit 2 shutdown margin calculation performed on November 8,1989.
Two discrepancies involved errors in - _ . ..
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) l abstracting an ARI boron worth value and a critical boron worth value from I tables in the PTDB. The other discrepancy was a missing data sheet review I sign-off. The inspectors verified that the calculated shutdown margin was within applicable limits using the correct reactivity values determined by j the inspectors from the PIDB.
The problems identified in this ' surveillance indicated a weakness in the independent reviews the procedures receive before the calculations are accepted.
l l Procedure 88013-C is used to calculate the overall core reactivity and ) compare this with the design prediction to demonstrate agreemerft within i 1000 pcm. This comparison is required every 31 EFPD in accordance with TS 4.1.1.1.2.
Five completed. surveillances for Unit 1. Cycle 1, were
examined; all completed surveillances for the current Units 1 and 2 j operating cycles were examined.
All records reviewed were found to be
complete and were conducted at the required frequency.
The largest ' difference between the calculated and predicted reactivity was 157.5 pcm , l for Unit 1, Cycle 1, 262 pcm for Unit 1, Cycle 2, and only 55 for Unit 2, , ' i Cycle 1.
The inspectors also reviewed completed surveillance records of procedure ' l 14940. Estimated Critical Condition Calculation.
This procedure is used , I to calculate the critical rod position for a preset boron concentration or critical boron concentration for a preset rod pattern.
The int.pectors I reviewed, in detail, seven recently completed procedures to ensure completeness, accuracy in using the PTDB tables and curves, arithmetical ] correctness, and reasonable agreement between actual and predicted criti- , cal conditions.
For all seven of the records reviewed, the licensee determined ECPs for a preset boron concentration.
Reasonable agreement
between ECPs and actual critical rod positions were found; 78 steps was i l the largest difference noted.
One minor discrepancy was identified in a Unit 2 ECP calculation perfonned on May 20, 1989, regarding an incorrect , ECP value abstracted from the PTDB.
Using the table of integral rod worth versus control rod step withdrawal, for a calculated rod worth of 1237.5 i i i pcm, the licensee abstracted a value of 109.5 steps withdrawn on control ' rod bank C.
The inspectors detennined the correct value from the table to be 89.3 steps.
For 109.5 steps withdrawn, the equivalent integral rod
worth is actually 1048.41 pcm.
The actual critical position for this startup was 92 steps withdrawn on control rod bank C.
Using 89.3 as the
correct ECP, the actual critical position is still well within the ! l allowable, t750 pcm margin for agreement between actual and predicted i ( criticality as specified by the procedure. The inspectors discussed with > the licensee the need for greater efficiency in abstracting data from the . PTDB and the need for more thorough procedure reviews.
No violations or deviations were identified in this area.
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End-of-LifeModeratorTemperatureCoefficientMeasurement(61708) Procedure 88009-C (Revision 0), Moderator Temperature Coefficient Determi-l nation (E0L), was issued on 9/29/89.
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l Novecber 1989, when Unit 1 CB, drops below 300 ppm 3.
In steps 8.1.3, 8.1.4, 6.1.5, and 8.1.18, the procedure accepts successive measurements of i equilibrium conditions in the RCS or PZR if the neasurements agree within j ' 10 ppmB and the average concentrations in the RCS and PZR agree within 20
ppmB. These allowable variations appear tuo broad to obtain the arecision i , necessary for the measurement. Other utilities use 2 ppmB for tie inter-l nel agreement of equilibrium measurements.
One way to help assure the j required internal precision of the C measurements is to have all of the n measurements done by the same chemistty technician.
The thermal power measurement is derived from the loop delta-Ts only.
No use is maoe of U1118, which is probably the best measurement of thermal power.
Data Sheet 4 assumes that the initial Co for the dilution step is the
final for the boration step.
If there is a significant delay between the
two steps, that may not be the case, and provision should be made to
obtain independent initial values for the dilution.
There is no acceptance criterion for agreement between the ITCs meatured , during boration and dilution.
Common industry practice is to require agreement within 10 percent of the mean.
Following discursion of these procedural weaknesses, the licensee made a commitment to revise the procedure to require d2 ppmB internal agreement among C measurements and 10 percent agreement between ITC measurements.
(Inspechorfollowupitem 50-424/89-29-02and50-425/89-33-02)
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Followup of Previously Identified 1 toms (92701) (Closed) Inspector followup item 50-425/89-22-02: Assure surveillance procedures for flow and thermal power require that plant instrumentation produce no less conservative results than do the precision instruments.
Procedure 54075-C has been replaced by procedure 88075-C (Revision 0, " issued 8/22/89), Precision Heat Balance.
Section 8.11 of that procedure
requires that any plant flow transmitter be within +2.5 percent or -0.0 percent of the precision instruments or that a MWO be issued to correct the indication within those limits.
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(Closed) Inspector followup item 50-424 and 425/89-02-01: Revise l 14940-1/2 (Revision 6), Estimated Critical Condition Calculation, and ' 12003-C (Revision 12), Reactor Startup Mode 3 to Mode 2, to provide better integration between the procedures and better organization of activities within the procedures.
Both procedures were revised.
The most signifi-cant changes were made to 12003-C and included adding a requirement to reperform 14940-1/2 if reactivity conditions changed by more than 250 pcm prior to control bank withdrawal and changing the responsibility for performing the the ICRR calculation and plot to the operating crew.
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, ' (Closed) Inspector followup item 50-425/89-22-01: Revise procedure 50003-C to normalize chamber currents to thermal power vice nuclear instrument channel power indication.
Procedure 55003-C has been replaced by procedure 88023-C (Revision 1), Incore/Excore Detector Calibration, which normalizes chatber current to thermal power as measured by on-line computer program U1118 on the plant computer.
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Exit Interview The inspection scope and findings were summarized on October 20, 1989 with those persons indicated in paragraph I above.
The inspectors described the areas inspected and discussed in detail the inspection findings.
Dissenting comments were not received from the licensee.
Proprietary ! information was reviewed, but is not contained in this report.
The commitments listed below were confirmed.
Inspector followup item 50-424/89-29-01 and 50-425/89-33-01: Document the review of the ion chamber I-V curves in the calibration procedures and ' make the procedure responsive to the results of that review.
(Paragraph 3.a(1))
Inspector followup item 50-424/89-29-02 and 50-425/89-33-02: Revise the ! MTC at EOL procedure to require i2 ppmB internal agreement among C mea-a surements and 10 percent agreement between measurements obtained dbring boration and dilution.
(Paragraph 6.)
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Acronyms and Initialisms Used in This Report axial flux difference AFD - ARI all rods in - boron concentration C - dNlta-T-temperature difference , estimated critical condition ECC - estimated critical position ECP - effective full power days EFPD - end of life E0L - emergency response facility ERF - I&C instrument and control - inverse count rate ratio ICRR - isothermal temperature coefficient ITC - moderator temperature coefficient MTC - maintenance work order MWO - percent millirho pcm - parts per million boron ppmB - power range nuclear instruments PRNI - PTDB Plant Technical Data Book - pressurizer PZR - . - - - - - - . - . .
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- reactor coolant system RCS - rated thermal power RTP - technicalspecification(s) ! TS - " volts direct current yde - j.. h . i ( l > , t , -$ P r
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