IR 05000483/1986018

Insp Rept 50-483/86-18 on 860617-26,0708 & 14-16.Violation Noted:Failure to Maintain Reactor Power Level within Licensed Limits & to Notify NRC of Condition,Per Ol,Item F

ML20204E411
Person / Time
Site:	Callaway
Issue date:	07/29/1986
From:	Mccormickbarge, Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20204E401	List: IR 05000483/1986018 ML20204E397 ML20204E405
References
50-483-86-18, NUDOCS 8608010072
Download: ML20204E411 (9)

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. l U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-483/86018(DRS)

Docket No. 50-483 License No. NPF-30 Licensee: Union Electric Company

Post Office Box 149 St. Louis, MO 63166 Facility Name

Callaway, Unit 1 Inspection At: Callaway Site, Callaway County, M0 Inspection Conducted: June 17-26 and July 8, 14-16, 1986 Inspector:

h M M. L. kCormick-Barger YZ'//h(

Date Approved By Chief Test Programs Section 7/zf[f Date

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Inspection Summary Inspection on June 17-26 and July 8, 14-16, 1986 (Report No. 50-483/86018(DRS))

Areas Inspected: Routine announced inspection of control rod worth measurements

(61710), moderator temperature coefficient measurements (61708), shutdown margin (61707), and core thermal power evaluation (61706).

Results: Of the four areas inspected, no violations or deviations were identified in three areas; one violation was identified in the remaining area (failure to maintain reactor power level within licensed limits and failure to notify the NRC in accordance with the license - Paragraph 5).

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DETAILS 1. Persons Contacted S. Miltenberger, General Manager, Callaway Plant G. Randolph, Manager, Callaway Plant

• W. Campbell, Assistant Manager, Nuclear Engineering

• J. Blosser, Assistant Manager, Operations and Maintenance

• R. Affolter, Superintendent, Systems Engineering

• J. Gearhart, Superintendent, Operations Support Quality Assurance

• K. Bryant, Supervising Reactor Engineer

• T. Sharkey, Supervisor, Compliance

• T. Robertson, Quality Assurance Engineer

, *S. Petzel, Quality Assurance Engineer

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P. Mory, Operations Shift Supervisor J. Patterson, Operations Shift Supervisor K. Evans, Supervising Instrumentation and Control Engineer J. Knaup, Compliance Engineer B. Stanfield, Quality Assurance Engineer

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Additional station, technical, and administrative personnel were contacted by the inspector during the course of the inspectio * Denotes those personnel present at the exit intervie . Control Rod Worth Measurements The inspector reviewed licensee procedures and results to verify that prerequisites, precautions, and plant conditions were met, that values obtained were within acceptance criteria and consistent with technical specifications, and that any discrepancies were properly reviewed. The inspector utilized the following documents during the review:

• Engineering Technical Procedure ETP-ZZ-ST005, " Bank Reactivity Worth

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Measurement," Revision 1, File No. E70.10.01, dated April 8, 1986,

! and performed for Cycle 2 - Control Bank A on April 17, 198 * Test results for ETP-ZZ-ST005, " Bank Reactivity Worth Measurement,"

Revision 1, File No. E70.10.01, performed for Control Banks B, C, and D on April 17, 198 * Test Summary Sheet for ETP-ZZ-ST003, Revision 0, " Determination of Low Power Physics Testing Range," dated December 16, 1985,and performed for Cycle 2 on April 15, 198 * Startup Report for the Callaway Nuclear Power Plant Cycle 2, Westinghouse Electric Corporation, March 1986. Transmitted to Union Electric via: Letter 86EU*-G-0021, D. J. Petrarca, Westinghouse Nuclear Fuel Division Project Engineer to K. R. Bryant Callaway Supervising Reactor Engineer, " Union Electric Company, Callaway Nuclear Plant, Cycle 2 Startup Package," dated March 17, 1986.

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(Note: The Startup Report was used for approximately the first two months of Cycle 2 Operation until the following report was received by Union Electric.)

• Westinghouse Report WCAP-11120, "The Nuclear Design of the Callaway Unit 1, Nuclear Power Plant, Cycle 2," dated May 198 No violations or deviations were identifie . Moderator Temperature Coefficient Measurement The inspector reviewed licensee procedures and results related to the Cycle 2 zero power moderator temperature coefficient measurement to determine that prerequisites, plant conditions, and precautions were met, that results obtained were within acceptance criteria and technical specification limits and that any discrepancies were properly evaluate The inspector utilized the following documents during the review:

• Engineering Surveillance Procedure ESP-ZZ-00009, Revision 3, File No. GZZ.00.05, " Moderator Temperature Coefficient Measurement at Zero Power," dated April 7,1986, and performed for Cycle 2 on April 15, 198 .

• Engineering Technical Procedure ETP-ZZ-00014, Revision 3, " Negative Moderator Temperature Coefficient Maintenance Calculation," dated April 11, 1986, and performed for Cycle 2 on April 16 and 17, 198 * Special Report No. 86-02 transmitted to NRC Region III via letter from G. L. Randolph, Callaway Plant Manager to J. G. Keppler, NRC

- Region III Regional Administrator, "Callaway Plant Positive Moderator Temperature Coefficient, Special Report No. 86-02," dated April 17, 198 * Curve Book, Master File Copy, File No. E52.01, Figure 2-13, dated April 22,1986, " Control Rod Withdrawal Limits to Maintain MTC < 0, Effective Up to 4000 MWD /MTU Burnup, Cycle 2."

• Engineering Technical Procedure ETP-ZZ-00015, Revision 1,

" Preparation, Review, Approval, and Control of the Curve Book,"

dated October 23, 198 * Startup Report for the Callaway Nuclear Power Plant Cycle 2, Westinghouse Electric Corporation, March 198 * Westinghouse Report WCAP-11120, "The Nuclear Design of the Callaway Unit 1 Nuclear Power Plant, Cycle 2," dated May 198 No violations or deviations were identifie _- _ _ _ _ _ _ _ _

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4. Shutdown Margin The inspector reviewed licensee procedures and results related to shutdown margi3, estimated critical position calculations and reactivity anomaly checks. The inspector utilized the following documents during the review:

• Operations Surveillance Procedure OSP-SF-00001, Revision 5, File No. GSF.00.05, " Shutdown Margin Calculation," dated April 9, 1986, and performed on April 16, 1986, and May 19-25, 198 * Operations Surveillance Procedure OSP-SF-00005, Revision 2, File No. GSF.00.05, " Estimated Critical Position Calculation," dated April 9, 1986, and performed on April 17, 1986, at 8:15 * General Operating Proceaure OTG-ZZ-00002, Revision 9, File No. GZZ.00.05, " Reactor Startup," dated July 31, 1985, and performed April 17, 198 Steps 4.1.9, 4.2.10 and 4.2.11 related to estimated critical position calculatio * Engineering Surveillance Procedure ESP-ZZ-00013, Revision 3,

, File No. GZZ.00.05, " Core Reactivity Balance Comparison," dated November 14, 1985, and performed for Cycle 2 on May 16, 198 * Operations Surveillance Procedure OSP-ZZ-00001, File.No. G52.05, Attachment 1, " Control Room Shift and Daily Log Readings and Channel Checks - Modes 1 through 4," - Group Rod Position Indication Results

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for April 18 through May 31, 198 * Reactor Operator Daily Logs, File No. 052.03, dated April 17, 198 * Curve Book, Master File Copy, File No. E52.0 With respect to Procedure OSP-SF-00005, " Estimated Critical Position Calculation," the inspector noted that the calculation could be performed by hand in accordance with Procedure Attachment 1 or by using a compute At the end of Attachment 1, spaces were provided for three signatures:

" Calculated By," Checked By, and " Approved By." Spaces for these signatures were not supplied on the computer printout sheet for the April 17, 1986, calculation of estimated critical condition. In discussions between the inspector and the licensee, the licensee stated that the procedure would be changed to ensure that these signatures are provided for the computer calculation as well as the hand calculation. This issue is' viewed as an isolated occurrence with no immediate safety significance, and based on the licensee's proposed corrective action, the inspector has no further concerns in this are No violations or deviations were identifie l

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5. Core Thermal Power Evaluation The inspector reviewed licensee procedures and results for technical adet;uacy and to verify that core thermal power was maintained within prescribed limits. In addition, the inspector began a review of two licensee identified incidents related to core thermal power. The inspector utilized the following documents during this review:

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• Operations Surveillance Procedure OSP-SE-00004, Revision 5, File No. GSE.00.05, " Nuclear Instrumentation System (NIS) Power Range Heat Balance," dated May 10, 1985, and performed on the following dates: December 1-4 and April 19 through May 31, 198 * Temporary Procedure Change TCH 86-033, for Procedure OSP-SE-00004, Revision 5, Effective from January 17 to April 1, 198 * SNUPPS Periodic Logs, File No. 052.04.14, " Nuclear Power - %: Last One Minute Averages," for April 28 and May 7-8, 198 * Incident Report IR 85-486, related to a heat balance calculation performed in accordance with Procedure OSP-SE-00004 on December 3, 1985, which indicated that reactor thermal power was calculated to be 101.5% of rated thermal powe * Incident Report IR 84-877, related to a heat balance calculation performed in accordance with Procedure OSP-SE-00004 on December 3, 1984, which indicated that reactor thermal power was calculated to be 102.28% of rated thermal powe During the review of the power range heat balances from April 19 through May 31, 1986, the inspector observed that the results were above 100% power on two dates: April 28, 1986, 8:55 p.m. - 100.24%

power and May 8, 1986, 2:50 a.m. - 100.23% power. By reviewing the hourly printout of the nuclear power values from the periodic logs dated April 28 and May 7-8, 1986, the inspector determined that the average power over an eight hour shift had not exceeded 100% and therefore, the inspector had no concerns based on this revie Incident Report 85-486 At the request of the Callaway Resident Inspectors, the inspector reviewed Incident Report 85-486 which was written by the licensee to address the fact that the heat balance result obtained at 4:52 on December 3, 1985, was 101.5% of rated thermal power (101.5% of the licensed thermal power limit).

The immediate actions taken by the licensee were to reperform the heat balance to verify the calculation results and then to recalibrate all four power range nuclear instruments which had been reading approxi-mately 100% power to read 101.6% power and subsequently to reduce power to below 100% based on the recalibrated power range instrument __

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In the evaluation that followed these immediate actions, the licensee-t-

initially thought that errant behavior of a Loop 2 feedwater flow instrument (AE-FT-521A) might have been the cause of the high, power level result obtained from the heat balance calculatio During an

instrument calibration performed at 12.00 p.m. on December 3, 1985, AE-FT-521A was found to-be reading out of tolerance on the high side, but this out of tolerance condition was only sufficient to account for

a .4 to .5% power conservatism in the heat balance calculation. Since the heat balance result earlier that day had been in excess of rated thermal power by 1.5% power, this left approximately one percent power unaccounted for. Further evaluation by the licensee revealed that Loop 4 feedwater flow had been reading low (nonconservatively) for l approximately a month prior to December 3,1986, and then, for unknown

! reasons, corrected itself between 9:00 p.m. and 10:00 p.m. on

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December 2, 1985,as indicated on the hourly periodic computer log The licensee's evaluation revealed that the Loop 4 feedwater flow began reading low following a reactor trip on November 2, 1986. The licensee stated that comparison of data at 100% power before and after

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the November 2, 1986, trip indicated that following the trip, Loop ^Ts

! (which are )roportional to reactor power) were approximately one percent higler and that Loop 4 feedwater flow comprised approximately

0.8% less of the total feedwater_ flow. In response to a question

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posed by the inspector regardin

was operated above 100% power, g the length the licensee of time reviewed that hourly the reactor logs

during the period that Loop 4 feedwater flow was low (November 3 to December 3, 1986) and determined the length of time that these logs

, were indicating at or above 99% power was about 504 hours0.00583 days <br />0.14 hours <br />8.333333e-4 weeks <br />1.91772e-4 months <br /> (about

21 days).

Safety Significance The maximum reactor power level experienced during this incident was t approximately 101% (plus or minus a two percent error inherent in the heat balance calculation). Callaway's safety analyses (for which maximum power was a limiting case) were performed at 104.5% power, and

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therefore, operation at 101 i two percent power was bounded by the safety analyses. Hence, this incident had no immediate safety

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significanc Assessment of Reportability Callaway operated at up to one percent power in excess of the licensed power limit for approximately 504 hours0.00583 days <br />0.14 hours <br />8.333333e-4 weeks <br />1.91772e-4 months <br /> (21 days) which is a violation of Callaway Facility Operating License No. NPF-30, Item C.'1),( and as such, should have been reported to the NRC in accordance with Callaway Facility Operating License No. NPF-30, Item F which required that the NRC be notified in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> followed by a written notification within 30 days. The licensee did not report this to the NRC because of the misconception that, if the results of a daily heat balance were between 100% power and 102% power, that was within the error band of

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two percent power inherent in the heat balance calculation, and therefore, did not constitute a license violation as long as power

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level was immediately reduced in accordance with the following .

criteria: within 15 minutes if the power level had been found at 102%

power, within 30 minutes if the power level had been found at 101%

power, etc. The inspector explained to the licensee that although the NRC does generally take the position that violations will not be issued for minor power excursions of short duration (that is, 102%

power for 15 minutes, 101% power for 30 minutes, etc., not to exceed 100% power when averaged over an eight hour shift), that position applies, not from the moment at'which the licensee realizes that an excessive power condition exists, but rather is applied to the entire duration of the operation in excess of the licensed power limit. In addition, the inspector explained that it should not be assumed that a heat balance result in excess of 100% power but less than 102% power need not be reported solely on the basis that a i two percent power error is assumed for the heat balance as documented in Callaway's Final Safety Analysis Report. Based on these clarifications, the licensee reassessed the December 3, 1985, incident, informed NRC Region III personnel via a July 8, 1986, telephone call that their reassessment indicated the incident was reportable, and stated their intention to report the incident in a Licensee Event Report. During the July 8, 1986, telephone call, the licensee asked whether or not the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> telephone notification was necessary in this instance and Region III management replied that it would not be necessar Exceeding the licensed power level and failure to report this in accordance with license requirements is considered a violation of Callaway Facility Operating License No. NPF-30 (483/86018-01(DRS)).

Licensee's Corrective Actions to Prevent Recurrence Included:

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• Calibration of all eight feedwater flow transmitter * Increasing the calibration frequency of all feedwater flow transmitters from 18 months to quarterl * Programming the NSSS process computer to provide weighted one minute averages of feedwater flow for use in heat balance calculation * Programming the computer so that, for each steam generator, the two feedwater flow instrument channels are compared and a difference of greater than 2.5% causes the readout point on the control room computer alarm screen to turn from green to re (The licensee implemented an interim action via Temporary Change Notice 86-033 for Procedure OSP-SE-0004 which was in effect from ,

January 7 to April 1, 1986, and which required operators to

. perform a manual (non-computer) feedwater flow channel check for heat balances at greater than 95% reactor power and to direct the instrumentation and control group to resolve any differences greater than or equal to 2.5%).

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• Initiating a Bill of Materials to replace the Barton feedwater flow transmitters with Rosemont (or equivalent) feedwater flow instrument All of the above actions were documented in Incident Report 85-486 and had been completed prior to this inspection. Regarding the last item mentioned above, the Bill of Materials had been initiated but had not been issued prior to this inspection. The licensee intended to issue the procurement order by the end of June 198 Additionally, the licensee stated that their long range goal was to develop a validated heat balance computer program on the NSSS process computer, and in the interim, to take measures such as trending parameters that are independent indicators of thermal power, on a daily basis and forwarding daily heat balances to an engineering group for review. The licensee also established internal guidelines for engineering review of heat balance result Incident Report 84-877 During the review of Incident Report 85-486, the licensee volunteered the information that a similar incident had occurred on December 3, 1984 in which the heat balance result taken at 7:00 p.m. was 102.28%

of rated thermal power. The licensee provided a copy of the incident report (IR 84-877) to the inspector. As with the incident discussed in the previous paragraph (Paragraph 5.b.), the reactor power level, plus or minus two percent power for heat balance error, was bounded by Callaway's safety analyses (that is, less than 104.5% power), and therefore, this incident had no immediate safety significance. The licensee intends to reassess this incident for reportability based on their new understanding of the NRC position on exceeding licensed reactor thermal power (see Paragraph 5.b.).

This will be followed as an unresolved item (483/86018-02(DRS))

pending the licensee's reassessment of the event for reportability and subsequent NRC revie One violation and one unresolved item requiring further review and evaluation were identified during the inspection of this program are . Unresolved Items Unresolved items are matters about which information is required in order to ascertain whether they are acceptable items, open items, deviations, or violations. An unresolved item disclosed during the inspection is discussed in Paragraph . - --, .

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! Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)

on June 26, 1986, to discuss the scope and findings of this inspection and the likely informational content of the forthcoming inspection repor The licensee acknowledged the statements made by the inspector and stated that the Cycle 2 Startup Report and the Cycle 2 Nuclear Design Report, referenced within this inspection report, were considered proprietary, but references to thace documents would not be considered proprietar Followup telephone conversations with the Callaway Plant Manager and other Callaway personnel regarding Incident Reports85-486 and 84-877, discussed in Paragraph 5 of this report, took place on July 8,14-16, 198 .

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