ML20206G232
| ML20206G232 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/14/1988 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| LIC-88-952, NUDOCS 8811220159 | |
| Download: ML20206G232 (9) | |
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Omaha Pubilt Power District 1623 Harney Omaha, Nebraska 68102 2247 402/536 4000 November 14, 1988 LIC-88-952 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station PI-137 Washington, DC 20555
References:
1.
Docket No. 50-285 2.
Letter EA-88-201 from R. D. Hartin (NRC) to K. J. Morris (OPPC), dated October 12, 1988 3.
Letter LIC-88-577 from K. J. Morris (OPPD) to Document Control Desk (NRC), dated July 1, 1988 4.
Letter LIC 88-620 from K. J. Morris (0 PPD) to Docume:,c Control Desk (NRC) dated July 25, 1988
SUBJECT:
Response to Notice of Violation 50-285/88-22-01, 88 22-02 Omaha Public Power District (0 PPD) recently received Reference 2 containing Notice of Violation 88 22-01 and 88-22-02.
The violations concerned 1)
Errors in the Cycle 11 Setpoint Analysis and 2)
Incorrect information submitted in a response. OPPD's response to these violations is attached to this letter.
If you have questions concerning this matter, please do not hesitate to contact us.
Sincerely,
. Morris Division Manager Nuclear Operations KJH/sa Attachment c:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Ave., N.W.
Washington, DC 20036 R. D. Martin, NRC egional Aaministrator P. D. Milano, NRC Project Managt P. H. Harrell, NRC Senior Resider.' Inspector G911220159 891114 PDR ADJCK 00000TA5 a
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Document Control Desk LIC-88-952 Page 1 l
ATTACHMENT l
During an NRC inspection conducted during the period June 29 through July 18, 1988, violations of NRC requirements were identified.
In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1988) the violations are listed below:
A.
10 CFR Part 50, A)pendix B, Criterion III states, in part, that e M.sures shall be establis1ed to assure that applicable regulatory requirements and the design basis, as defined in Section 50.2, for those structures, systems and components to which Appendix B applies are correctly translated into
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specifications, drawings, procedures and instructions and that the design control measures shall provide for verifying or checking the adequacy of l
design.
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The approved "Quality Assurance Program," Section A.4, "Design Control," of i
Appendix A to the Updated Safety Analysis Report for the Fort Calhoun Station states, in part, that the verification of engileering and design adequacy of the contractors' design documents is performed in accordance with an OPPD approved quality program and procedures.
Paragraph 4.3.5 of Section 5.1 of the licensee's Quality Assurance Manual, "Control of Plant Design and Modifications," states, in part, that design analyses such as physics, stress, thermal, hydraulic, and accident analyses, shall be performed in a planned, controlled, and correct manner.
Contrary to the above, OPPD failed to verify the adequacy of design and failed to perform design analyses in a correct manner in that OPPO utilized a computer program supplied by a contractor to perform setpoint analyses for Cycle 11 reactor operations without verifying the adequacy and correct
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use of the computer code.
The licensee failed to determine all factors l
required to be used in computer-derived calculations of the setpoints for the thermal margin / low pressure reactor tri) function of the Reactor Protection System, resulting in setpoints t1at reduced the margin of safety.
This condition existed from June 7, 1987, when Cycle 11 operations f
began, until June 29, 1988, when this condition was corrected.
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t This is a Severity Level III violation (Supplement 1)(285/8822-01).
j OPP 0's Resoontq:
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1.
The Reason for the Violation if Admitted OPPD admits to the violation as stated.
The errors in the Cycle 11 set >oint analysis were identified by OPPD and promptly reported to the NRC in.ER 88 16, "Cycle 11 Setpoint Errors" dated July 29, 1988.
i Identification of the errors occurred during preparation for the 1
performance of the Cycle 12 setpoint analysis as a result of the review of l
the Cycle 11 setpoint analysis. When conducting a reload analysis, it has been a standard OPPD practice to review the previous cycle's analysis prior i
to performing the same analysis for the cycle being analyzed. As part of L
the review of the Cycle 11 setpoin+. analysis, identification of two potential errors in the Cycle 11 analysis affecting Technical i
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Document Control Desk LIC-88-952 Page 2 SpecificationFigure1-3(ThermalMargin/LowPressureLSSS-4PumpOperation) and Technical Specification Figure 2-6 (Limiting Condition for Operation for Excore Monitoring of LHR), was made at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> en June 28, 1988.
At this time, a preliminary evaluation of the potential impact of the errors was performed.
It was estimated that approximately a two to three percent error existed in the TM/LP trip equation.
This corresponds to a non-conservatism of approximately 100 psia.
Significant offsetting operating margins were then identified and evaluated. The margins identified were:
(FrT) peration with a maximum total integrated radial peaking factor (1) Oless than 1.70. This is significantly less than the Technical Specification limit of 1.80.
This margin was estimated to be much greater than 100 psic.
(2) Operation with a maximum core inlet temperature of 541' F versus the Technical Specification limit of 545' F.
This conservatism was estimated to be 88 psia.
The combination of these two margins was determined to result in an offsetting conservatism greater than the potential error. A similar margin for the operational total planar radial peaking factor (F T) versus the Technical Specificationlimitof1.85wasidentifiedtoexisEforTechnicalSpecification x
i Figure 2-6.
Thus the evaluation group consisting of the Reactor Engineer, the l
OPPD personnel responsible for the setpoint analysis, and a setpoint expert from Combustion Engineering concluded that operation with the two errors identified did not present a safety issue for potental operation outside the bounds of the design limits. Analysis of the Cycle 11 setpoints was initiated imediately following this evaluation including better quantification of the l
errors.
l Following the meeting on June 28, 1988, interim corrective actions were initiated to ensure conservative plant operation during the period in which the impact of these errors was being quantified.
As a precautionary measure, the Reactor Engineer contacted the afternoon Shift Supervisor and discussed with him the conservative actions to be taken until the errors could be quantified. The Shift Supervisor was instructed that if I
any situation arose that necessitated the use of the excore linear heat rate Limiting Condition for Operation (LCO), the unit would be brought to 80% power and not the 90% power level as identified in Technical Specification Figure 2-6.
GiventheactionstatementinTechnicalSpecification2.10.4(1)(b),
Figure 2.6 would not have been entered under any circumstances that night.
In
- addition, instructions were provided that the unit was to be held at 90% power and the core inlet temperature maintained at 541' F until the situation was resolved, in order to prevent any reactivity changes which might have the potential for initiating a transient event which could challenge the RPS.
l Document Control Desk LIC-88-952 Page 3 l
After this discussion, the Reactor Engineer called the Manager Fort Calhoun Station and apprised him of the above described events. They discussed the applicability of entering an LCO.
Entry into the LC0 was not made because the error in the TM/LP trip had not been quantified. After this conversation, the Plant Manager called the Shift Supervisor to confirm that upon entry into Figure 2-6, an imediate power reduction to 80% would be initiated.
Based on the Manager Fort Calhoun Station's recomendation, the Reactor Engineer alsa called the Manager-Administrative & Training Services to discuss the situation.
The Manager-Administrative & Training Services comunicated with other management personnei.
The Reactor Engineer called the night Shift Supervisor to fully brief him on the circumstances related to the information provided to the previous shift. Additional guidance was provided that the Variable High Power Trip was not to be reset in order to limit and minimize the effects of a transient associated with an unplanned power increase.
On June 29, 1988, the Reactor Engineer was notified that the Thermal Margin Pvar equation used in the RPS trip units was non-conservative by approximately 80 psi based on preliminary calculations which remained to be independently reviewed.
It was identified that it would take several additional hours to perform the analysis verification and confirm the exact value needed to place the TM/LP trip units back in a conservacive condition. The Plant Manager and his staff were imediately brought up to date on this latest development.
On June 29,1988 at 0920 hours0.0106 days <br />0.256 hours <br />0.00152 weeks <br />3.5006e-4 months <br />, although the exact degree of non-conservatism had not been identified, the station was conservatively placed in a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LC0 per Technical Specification 2.15(3). The TM/LP trip units were declared inoperable and rec &libration was initiated. As an additional conservative measure the Plant Manager also stipulated that, at four hours into the LCO, if significant progress had not been made on recalibration of the trip units, a st.utdown would be initiated. A conservative value of 100 psia was chosen such that it would bound the analytical result (i.e. 79 psia) determined by the Engineering Staff.
Per 10 CFR 50.72, at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, a four hour report was made to the NRC. A meeting was held after four hours into the LCO to determine if adequate progress had been made.
By 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, the recalibration and the operability surveillance tests were completed and the four trip units had again been declared operable.
Technical Data Book Figures were prepared for the TM/LP Trip / Pre Trip and excore LHR LC0 functions.
These revised figures were approved by the Plant Review Comittee on July 1,1988, and an Operations Memorandum (0PS Memo 88 05)
I was issued the same day instructing operators to utilize the Technical Data Book Figures in lieu of the corresponding Technical Specifications.
Following correction of the setpoint analysis errors, an evaluation was conducted by both the department responsible for the analysis and by a Management Investigative Safety Team (MIST) to determine the root cause(s) of why the errors occurred.
The MIST, which is a mechanism for formally mobilizing a team to organize and direct the evaluation of an event, was initiated by the Plant Manager on 61y 13, 1988 in accordance with Nuclear Operations Division Quality Procedure N00-QP-18, "Management Investigative Safety Team.
Requirements for evaluation include collection of event related information, evaluation of the significance of the event, identification of the cause, determination of the necessary corrective actions required to fully
Document Control Desk LIC-88-952 Page 4 mitigate the event and prevent recurrence, and a review of corrective actions to assure effectiveness. The MIST review for the TM/LP error concurred with internal technical responses to the event and therefore centered on determining the root causes and corrective actions in the manap went area. The MIST report, issued September 12, 1988, contained the results, conclusions and recomendations from the investigation.
Both the MIST and the responsible department evaluations independently determined the same three root causes of why the errors occurred.
The root causes were:
(a) Lack of previous experience for both the analyst and reviewer.
Personnel performing the analysis were qualified in accordance with Technical Services procedures to perform the work, but had little previous experience in this discipline. Management failed to take appropriate actions to compensate for this lack of previous experience nor did procedures require an experience level for analysis reviewers.
(b) The training program did not adequately define basic setpoint methodology training to support conversion from a hand calculation method to use of a setpoint utility code.
Training which consisted of two parts was fragmented with methodology training in July, 1986 and applications (i.e. setpoint utility) training in November 1986.
Delays between the two sessions without the use of the knowledge gained in July, 1986 resulted in lack of retention of details.
The training, which was consistent with procedures, was also not performance-based.
In addition, the training requirements were not properly defined.
There was no validation testing of employees trained to ensure that they fully understood the theory end were able to correctly apply it through the use of the setpoint utility code.
Personnel trained did not, nor were they procedurally required to, perform a verification and validation procesc to demonstrate that they could successfully produce the same results as a standardized case, such as the Cycle 10 analysis. The setpoint utility code and methodology training failed to either specify or account for the level of experience of the trainees.
(c) Although the setpoint utility code was verified and validated by the vendor the setpoint utility code / users manual combination was not verified and validated by either the vendor or OPPD prior to analysis application. A deficiency of the set)oint utility manual was that it did not describe the application of tie radial peaking scaling factor and the selection of transient / Loss of Coolant Accident (LOCA)
RequiredOverpowerMargin(ROPM). Consequently, application of the setpoint utility code along with verbatim compliance to tne user's manual resulted in not applying the scaling factor and LOCA R0PM.
a It should be noted that had any one of the three root causes above not existed, the analysis errors probably would not have occurred.
Oscument Control Desk l
LIC-88-952 Page 5 2.
The Corrective Steos Taken and Results Achieved i
The Cycle 11 setpoint analysis errors were found during performance of the Cycle 12 setpoint analysis.
Review of the previous cycle's analysis prior to performing the next cycle's analysis is a standard practice to ensure continuity from cycle to cycle. Upon confirmation of the Cycle 11 errors, Combustion Engineering was utilized to analyze the Cycle 11 setpoints and to determine if any other setpoint analysis errors existed. The analysis produced revised Thermal Margin / Low Pressure setpoints that were incorporated into the Reactor Protective System.
It was also determined that no additional errors i
j existed beyond the two already identified.
In addition, the Cycle 10 setpoint analysis was reviewed and correct application of the radial peaking scaling factor and LOCA R0PM were verified to ensure only the Cycle 11 instance of misapplication. An optimized setpoint analysis was later performed by l
Combustion Engineering under OPPD's direction which confirmed a lesser i
non conservatism of 47 psia versus the original analysis non-conservatism of 79 psia.
q An evaluation was conducted to assess and quantify the safety impact of the setpoint errors. This evaluation confirmed that no design liraits were or would have been violated under transient or steady state conditions with respect to Technical Specification Figures 1-3 and 2 6.
It was noted that the Thermal l
l Margin trip function is only credited in two safety analysis events, the Excess 1
Load ',USAR 14.11) and the RCS Depressurization (USAR 14.22).
In addition, the 4
Thermal Margin Trip function reaches the trip floor at approximately 72% power (for the most adverse axial power distribution).
Below this power level, the low pressure trip would act in place of thermal margin trip.
Thus, the j
l operability range required for the thermal margin trip was 72% to 100% power.
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The evaluation concluded that, as a result of conservative fuel cycle design and plant operation below Technical Specification limits, sufficient margins existed to offset the 47 psia non conservatism in the fechnical Specification Figure 1-3 Pvar equation.
Tnese margins consisted of:
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Integrated radial peaking factor margin of 6.67 percent (approximately l
200 psia)
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2.
Inlet temperature margin of 4' F (88 psia) j For the 4 percent power non conservatism in Technical Specification Figure i
j 2 6, the following conservatisms or mitigating factors were relevant:
1.
A 7.03 percent margin between the total unrodded radial peaking factor l
Technical Specification limit of 1.85 and the maximum operation value of 1.72 existed.
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2.
Technical Specification /igure 2 6 had not been entered since the very i
early fuel cycles at Fort Calhoun.
Entry into this LCO (i.e.,
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Technical Specificatioa Figure 2 6) would not have occurred for another three days when the previous core power distribution would i
have expired.
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D'ocument Control Desk i
LIC-88-952 Page 6
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To minimize the probability of a similar error, an experience requirement was i
established for core reload analysis reviewers which stipulates that the t
i reviewer must have previously and successfully completed an analysis of the type to be reviewed. This requirement was established on an interim basis through a departmental memorandum issued August 10, 1988. A proceduralized requirement has been prepared and is being incorporated into the Production
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Engineering Division (PED) Quality Procedure (PED-QP-28). These instructions j
require that the reviewer be (1) properly qualified to perform the analvr;s in accordance with PED QP-16 (PED Training Program), (2) have previously and i
successfully performed the analysis to be reviewed, and (3) be cognizant of the PED QP-5 (Engineering Analysis Preparation, Review and Approval) review requirements.
I The setpoint utility code vendor was contacted to revise and expand their code user's manual to correC.
.ie deficiencies identified during this investigation.
A format conducive to minimizing the probability of errors when utilized by i
first time / inexperienced users was also suggested. This work was completed by
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the vendor and receipt of the revised documents by OPPD occurred September 1, 1988.
l Combustion Engineering is performing independent technical reviews of Cycle 12 core reload analyses to ensure proper application by 0 PPD personnel of CE f
reload analysis methodologies. This broad scope examination is currently
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underway and consists of reviewing the sixteen analyses performed and L
i previously reviewed by OPPD personnel.
These reviews will further reduce the i
l probability of the existence of potential oversights in the other reload analysis disciplines.
In addition, Combustion Engineering is independently
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reviewing the Cycle 12 Core Reload Application previously transmitted to the NRC. Depending upon these reviews, Combustion Engineering may be contracted to i
review future reload analyses performed by ; PPD.
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i Implementation of changes to the improved training pro; ram with benchmarking of j
standard cases as well as testing of personnel occurred in 1987 after the l
1 occurrence of the Cycle 11 setpoint analysis errors, but prior to their detection in June 1988.
Implementation of these changes at an earlier date i
would have reduced the probability of the occurrence of the setpoint analyris errors.
The revised training requirements consist of methodology and j
i applications training with formal testing and personnel performance l
j evaluations.
The training also includes verification and validation testti. of i
personnel code use by benchmarking to standard cases.
These implement (d i
i training practices are being incorporated into a revision to a PED Quality
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l Procedure.
3.
The Corrective Steos that will be Taken to Avoid Further Violations In addition to the actions completed and ongoing as described above, OPPD f
j will or a broad scope basis:
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a)
Implement procedural requirements to provide experienced and qualified personnel to conduct independent reviews of analyses.
Quality Procejure PED-QP 28 will fulfill these requirements for review of i
reload analyses.
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Document Centrol Desk LIC-88-952 Page 7 I
b)
Develop and implement program requirements which assure training programs match incumbent skills with job requiraments for retpoint analysis work.
1 c)
Implement a verification and validation program for all sa'~ety related software used by 0 PPD.
These actions are expected to be completed by May 30, 1989.
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4.
The Date When Full Comoliance will be Achieved Full compliance with implementation of conservative and valid setpoints was achieved on June 29, 1988.
Formal implementation of an alternate Technical Specification Figure 2-6 was achieved on IJ1y 1,1988 through issuance of i
Operations Memorandum 88-05.
Informal implementation of Technical Specifi-l cation Figure 2 6 had previously occurred on June 28,1988 (the day of dis-covery).
t Implementation of the above "Corrective Steps to be saken" is expected to be completed L May 30, 1989.
Interim correctn e actions were provided in
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a memorandum from the Division Manager - Production Engineering to all PED I
managers and supervisors in which instructions wert, provided in making work assignments and authorizing use of computer c. odes in design and analysis i
work. This guidance is deamed to be sufficient until the three corrective i
4 steps listed above have been completed.
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Document Control Desk LIC-88-952 Page 8 B.
Section 50.9 of 10 CFR Part 50 states, in part, that information provided to the Comission by a licensee shall be complete and accurate in all material respects.
Contrary to the above, the licensee submitted a letter, dated July 1,1988, that stated the shift supervisor had placed instructions in his log to reduce reactor power to 80 percent if the excore detectors were used to measura linear heat rate.
The NRC inspector noted following NRC's receipt of the July I letter that the shift supervisor's log did not contain instructions to reduce reactor power to 80 percent.
This is a Severity Level IV vio'..; ion.
(Supplement 1) (285/8822-02)
Of PD RESPONSE 1.
The Reason for the Violation if Admitted OPPD aomits to the violation as stated. At the time of discovery of the potential for error in the analysis, OPP dok several corrective actions as detailed in the response to Violation A.
Included in the corrective actions were contacts to the NRR and regional offices to explain proposed interim actions, actions planned, and reasons for the error. After these conversations, OPPD comitted, verbally, to sumarize the corrective actions and interim comitments in a letter to the NRC.
The potential error in the setpoint methodology was discovered on June 28, 1988, and the letter was submitted on July 1, 1988.
This short time frame did not allow for the usual internal review process.
In addition to the short time frame, several informational telephone calls were held late in the evening to assure that the operations staff was cware of the concern.
The information which was placed in the letter was believed to be true by the person writing the response; in the short time frame concerned no verification of that information was made.
In short, the individual believed that the shift supervisor had written the information in the logi he did not verify that this had in f.ct been done.
The cause of this violation is the failure to achieve effective comunication.
2.
The Corrective Steos That Have Been Taken and Results Achieved Reference 4 was submitted to the NRC, correcting the information contained in the July 1,1988 letter.
Procedures for processing information for 1
submittal to the NRC cxist and are utilized to reduce the likelihood of submittal of incorrect information.
These procedures are considered to be adequate; in the short time involved in the response, the usual verificatiom etc., was not performed.
3.
The Corrective Steos That Will be Taken to Avoid Further Violations The existing procedures and policies which govern submittal of information to the NRC are believed to be adequate to preclude recurrence of this concern. Attention to detail and the irrportance of following procedures are being stressed in daily activities and will contribute to reducing the likelihood of recurrence.
4.
Iht,Date When Full Comoliance Will be Achieved OPPD is currently in fell compliance.