ML20044F988
| ML20044F988 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/24/1993 |
| From: | Fenech R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9306010288 | |
| Download: ML20044F988 (5) | |
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.c, HA Tennenee Vahey AuthWy, Post offee Box 2%C. Scody-Daisy.Tenreste 37379-2000 f
i Robert A Fenech Vce Pre 9 dent, Searyah Nuclea+ Plant May 24, 1993 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk i
Washington, D.C. 20555 I
Gentlemen
- In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SQN) - NRC INSPECTION REPORT NOS. 50-327, 328/93 REPLY TO NOTICE OF VIOLATION (NOV) 50-327, 328/93-11-01' Enclosed is IVA's response to Thomas A. Peebles' 1.etter to i
Mark O. Medford dated April. 23, 1993, which transmitted the subject NOV.
The alleged violation is associated with inadequate procedures for evaluating setpoints in excess of technical specification allowable values..As described in the enclosure to this letter, TVA denies this violation.
If you have any questions concerning this submittal,'please telephone l
Russell R. Thompson at (615) 843-7470.
Sincerely,
- f. W Robert A. Fenech i
Enclosure cc: See page 2 i
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U.S. Nuclear Regulatory Commission Page 2 May 24, 1993 i
cc (Enclosure):
Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission One White _ Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector
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Sequoyah Nuclear. Plant 2600 Igou Ferry Road-Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 l
Atlanta, Georgia 30323-2711 D
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ENCLOSURE 1 f
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RESPONSE TO NRC INSPECTION REPORT NOS. 50-327/93-11 AND 50-328/93-11
. i THOMAS A. PEEBLES' LETTER TO MARK 0. MEDFORD i
DATED APRIL 23, 1993 l
"10 CFR 50, Appendix B, Criterion V, requires in part that activities affecting quality be prescribed by documented instructions or procedures
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of the type appropriate to the circumstances, such as evaluating possible violations of Technical cpecifications and the reportability of related I
events.
" Contrary to the above, procedures and management direct" c-were not available to direct the evaluators to the correct conclus,ns for the i
event of May 16, 1992. On May 16, 1992, the trip setpoint for one Unit 2 i
intermediate range neutron monitor exceeded the allowable value of the LIMITING SAFETY 8YSTEM SETTING of Technical Specification 2.2.1.
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evaluators made the erroneous conclusions that the specification had not been exceeded and that the event was not reportable.
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" Itis is a Severity Level IV violation (Supplement I)."
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Rasis fotDisputine Violation 3D-327. 328/93-11-01 i
The procedure used to determine whether the trip setpoint for the subject instrument was actually outside allowable limits was Site Standard Practice (SSP) 12.9, " Incident Investigations and Root Cause Analysis."
I This procedure provides adequate guidance for conducting evaluations / investigations of events to determine their causes and the i
appropriate corrective actions. Each incident investigation team is i
headed by an event manager that has been trained on the procedure and has been certified in one or more root cause analysis techniques. Upon completion of the incident investigation, the team prepares a report stating the facts surrounding the eveut/ incident, a description of the analysis performed, the causes identified, and the recommended corrective actions. This report is then reviewed and approved by the Plant Event-l Review Panel, which is composed of plant management, prior to closure of the investigation and implementation of the corrective actions. This process, involving specifically trained team leaders with plant i
management review and approval, ensures thorough and consistent
. i implementation of the event investigation process at Sequoyah Nuclear Plant (SQN). The implementation of the SSP-12.9 process for determining whether the trip setpoints for the Unit 2 intermediate-range neutron i
monitor exceeded the allowable value of Technical Specification (TS) 2.2.1 was adequate to reasonably lead an evaluator to the correct conclusion. This process is described in detail below.
j Upon discovery that intermediate range (IR) nuclear instrumentation Channel N36 was indicating nonconservatively relative to the acceptance i
criteria of 0-PI-NUC-092-082.0, "Poststartup NIS Calibration Following Core Load," appropriate actions were taken to notifify Operations and to recalibrate IR Channel N36. As described in the inspection report, an investigation conducted in accordance with SSP-12.9 concluded that the i
primary root cause of the condition was inaccurate weighting factors.
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The safety implications of the condition were assessed by the investigation as required by SSP-12.9.
The condition was concluded to represent no adverse safety concequences. This conclusion was based on the fact that reactor power was not increased above approximately 4 percent until N36 was recalibrated and tested. Additionally, though the IR channels provide a diverse trip function to the power range (PR)
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channels, the operation of the IR trip is not assumed or considered in the accident analyses of the SQN Updated Final Safety Analysis Report (UFSAR). The four PR channels, which are modeled in the UFSAR analyses,
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were operable during the timeframe that N36 was incorrectly calibrated.
Section 15.2.1.1 of the SQN UFSAR states that the reactor trip for a i
postulated, uncontrolled, rod cluster control assembly bank withdrawal from a suberitical condition is assumed to be initiated from the low setting of the PR high neutron flux trip. A 10 percent increase is added to the nominal setting for conservatism, modeling the trip to be initiated at 35 percent reactor power. Even with the calibration error, N36 would have trapped before reaching this power level.
Reportability of the condition under 10 CFR 50.73 criteria was also assessed during the SSP-12.9 investigation. As described in the inspection report, an engineering evaluation determined that approximately two-thirds of the error in the N36 calibration was the result of the inaccurate weighting factors, which was classified as process error. The remaining one-third of the calibration error was the result of nonconservative rack drift, classified as rack error.
The difference between the TS reactor protection system nominal' trip setpoints and allowable values is based on rack error terms'as described in WCAPs 11239 and 11626, " Westinghouse Setpoint Methodology for Protection Systems, Sequoyah Units 1 and 2, Eagle 21 Version," and shown in the attached figure from WCAP 11626. As indicated, process errors are i
accounted for in the margin between the allowable value and applicable safety analysis limits. WCAPs 11239 and 11626, Revision 4, were submitted to NRC by letter dated April 23, 1990.
In the NRC safety evaluation supporting SQN Unit 1 License Amendment 141 referenced in the i
inspection report, it was stated:
"The staff has reviewed the Sequoyah instrument setpoint methodology document WCAP-11239 and 11626 (Reference 4), and finds that the allowable values of Tables 2.2-1 and 3.3-4 are consistent with the data in the setpoint methodology..." With this understanding of the basis for the TS allowable values, the actual rack error term for N36 was compared with that allowed by the TSs and was l
found to be bounded. Consequently, the condition was determined to not be reportable under 10 CFR 50.73.
In summary, it is TVA's position that site procedures provide adequate guidance to individuals to allow them to reach correct conclusions during performance of investigations of plant events. As described above, the condition associated with the incorrect calibration of N36 was properly investigated in accordance with plant procedures.
This investigation correctly assessed the safety significance of the condition and logically evaluated the reportability of the condition based on documented setpoint methodology.
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t WESTINGHOUSE CLASS 3 4-3 Safety Analysis Limit
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Process Measurement Accuracy I
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5ensor Measurement & Test Equipment i
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5ensor Pressure Effects
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Environmental Allowance 1
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STS Allowable Value
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Rack Measurement & Test Equipment I
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Rack Drift STS Trip Setpoint I
L Figure 4 3 Westinghouse STS Setpoint Error Breakdown - Digital Process Racks
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