ML20056B823
| ML20056B823 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/11/1971 |
| From: | Duncanson R NORTHERN STATES POWER CO. |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9102110414 | |
| Download: ML20056B823 (3) | |
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NORTHERN STATES POWER COMPANY Minneapolis, Minnesot J "'
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Dr. Peter A. florris, Di rector
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United States Atomic Energy Commission 9 Y
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Dear Dr. !!crris:
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MONTICELLO T1JCLEAR GEtE ' s PLANT Docket No. 50-263 License No. DPR-22 IRM Safety System Seitmgs Exceeding 15% of Rated Neutron Flux A condition occurred at the f/onticello Nuclear Generating Plant on March 3,1971 which is reportable to your office in accordance with the provisions of Appendix A, Technical Specifications, of the Provisional Operatine License DPR-22. The occurrence was classi fied as an Abnormal Occurrence os defined in Section 1.n.1 of the Technical Specifications and j
requires reporting in accordance with Section 6.6.8.3 of the S eci fications.
p The Region 111 Compliance Office has been notified in accordance with the re-quirecents of Section 6.6. A of the Technical Specification.
Sum,ry Descrintion of ihe Occurrerice On March 3,1971, heat balance data taken at approximately 12% power showed that for the existing rod pattern that all of the Intercediate Range Vonitor (IRM) scram settings were above the Technical Spec-ification limit of 15% of rated neutron flux.
The' scram settings of the eight IRMs on the highest selectable range (Range 10) varied from 24%
to 90%. The IRf/s were immediately recalibrated such that the scram trip settings were 12% of rated power on Range 9 and operations were administrative 1y limited to prevent operations beyond Range 9.
A physi cal stop was installed on March 8th preventing the selection of Range 10 on the IRMs.
_D_is cus s i on As part of the startup test program, two calibrations of the IRMs were performed prior to the calibration determined from heat balance data taken at approximately 12% power. The first calibration, which was performed at the tire of initial criticality, was based on a comparison of SRM and IRM readings in the overlap region. The average SRM count rate was related to core thermal power using an empirical forcula develop' by General Electric Co.
The second calibration was performed durine ibe initial heatup.
Core thermal power for the second calibration was determined by a heat balance calculation which was based on the measured heatup rate. The calibration based on the SRM count rate method, k
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when linearly extrapolated to Range 10 of the IRI/c, gave an average scram setting approximately eight times lower than the final heat balance data.
The calibration based on the reactor heatup rate, when linearly extrapolated to Range 10 of the IWs, gave an average scram setting approximately four tires lower than the final heat balance data.
The IRM calibration data obtained during the heatop showed that two IRM channels had extracolated scram settings on Range 10 that exceeded the 15%powerlevel. khis was not considered unusual because the rod pattern existing at that time skewed the flux away from the two channels.
it is of interest to note that the rod pattern affects were such that the two IRM channels with the highest extrapolated scram settings during the heatup were found to have the lowest extrapolated scram settings at the time of the final heat balance.
Reactor operations were conducted at power levels below 15% of rated power prior to the final calibration of the IRMs.
During the reactor heatup and testing phase at pressures below 1000 psig, reactor power was no higher than approximately E4 of rated.
On one occasion prior to the final calibration of the IRJs, a high flux scram occurred at 1000 psig when feed. vater flow to the core was increased accidently.
Based on the final heat balance data, which may not be vtlid for the rod pattern existing at the time of the scram, the scram occurred when an IRM reached
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a scram setting of 190 y
a scram setting of 10$ power (Trip S stem B) af ter a preceding IRM reached power (Trip S stem A).
Analysis of computer data y
indicates that the neutron flux remained above the 15% power level for approximately 1.25 seconds.
Technical Speci fication 2.1.C.1 states that the neutron flux shall not exceed the scram setting established (15% power) for longer than 0.95 seconds.
Cenclusions Changes in the control rod pattern which occur from cold criticality
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to a% proximately 10 percent power (STARTUP mode of operation) affect the I RM calibration in two ways:
1.
The calibration of a particular IRM channel is significantly af fected by rod motion in the vicinity of the channel, and 2.
An overall "decali brati on" of P e IRM channels occurs as the core is heated up to rated condi ions and the control rod density in the core is significantly reduced.
Once an accurate calibration of the IRM channels is determined for the i
highest range of operation (Range 9),
the established scram settings should be conservative for operations at lower power and temperature condi t i ons.
However, as core burnup proceeds an additional decalibration l
affect on the IR!!s may arise due to the decreased control rod density l
in the core.
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An on-site investigation of the occurrence has been completed and a report has been written.
This information will be available to the Region lli l
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.* Compliance Inspector for review during his next site visit.
Yoursveryfruly, f
[WM[
L R.O. Duncanson Gen. Supt. of Power Plants flechanical Chairman - Monticello Safety Audit Comittee ROD /caf
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