IR 05000416/1990001

Insp Rept 50-416/90-01 on 900108-12.No Violations or Deviations Noted.Major Areas Inspected:Core Power Distribution & Thermal Limits,Calibrs of Nuclear Instrumentation & Core Thermal Power

ML20006C223
Person / Time
Site:	Grand Gulf
Issue date:	01/23/1990
From:	Belisle G, Burnett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20006C222	List: IR 05000416/1990001 ML20006C221
References
50-416-90-01, 50-416-90-1, NUDOCS 9002070137
Download: ML20006C223 (7)

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o NUCLEAR REGULATORY COMMISSION.

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ATLANT A, GEORGI A 30323

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w; %.R_eport No'. : - 50-416/01

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A Licensee:. System Energy Resource, Inc.

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iDocket No.:

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License No.:

NPF-29-

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Facility Name: Grand Gulf.

' Ibspect'idn ' Conducted: January 8-12, 1990

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Approved by:

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G. A. Belisle, Chief VV Date Signed Test. Programs Section t

Engineering Branch.

Division'of. Reactor Safety

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- SUMMARY

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Scope:

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'This routine,; unannounced inspection addressed the surveillance's of core power

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distribution and thermal _ limits, calibrations of nuclear instrumentation,

surveillance:of core thermal power and-the. instrumentation. used in - thermal

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power measurement, and surveillance of the core reactivity anomaly.

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Results:

. The procedures to perform these surveillances were adequate and had been performed with the required frequency during.the current operating cycle.

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One case.of a computer point used in the reactor heat balance and, thermal power

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determination not being calibrated was identified. The licensee had made a

I similar finding and. initiated corrective action prior to the inspection.

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.(Paragraph 4.c)

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No' violations or' deviations were identified.

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9002070137 900129

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REPORT DETAILS

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Persons Contacted

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Licensee Employees W. T. Cottle, Vice President, Nuclear Operations

• J. P. Dimmette, Manager, Plant Maintenance

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• W. C. Eiff, Principal Quality Engineer

• R. T. Errington, Reactor Engineering Superintendent W. M. Harrell, Engineer,' Systems Engineering

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• C. R. Hutchinson, General Manager

• R. A. Martin, Reactor Engineer

• W. R. Patterson, Technical Assistant to the General Manager r

• J. C. Roberts, Manager, Performance and System Engineering

• J. S. Summers, Compliance Coordinator

• S. F. Tanner, Manager, Quality Services J. W. Yelverton, Manager, Plant Operations

• G. G. Zinke, Plant Licensing Superintendent

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Other licensee employees contacted included engineers, operators, security force members, and office personnel.

NRC' Resident Inspectors

• H. O. Christensen, Senior Resident Inspector

• J. L. Mathis, Resident Inspector

• Attended exit interview on January 12, 1990.

Acronyms and initialisms used throughout this report are listed in the final' paragraph.

2.

Power Distribution Monitoring (61702)

i a.

Documents Reviewed (1) ANF-89-079(P), Grand Gulf Unit 1, Cycle 4, Startup and Opera-tions Report,

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(2) ANF 88-149, Grand Gulf Unit 1, Cycle 4, Reload Analysis.

(3) ANF 88-150, Grand Gulf Unit 1,

Cycle 4,

Plant Transient Analysis.

l (4) -XN-NF833(P), POWERPLEX(R) Core Monitoring Software System, User's Manual for The Grand Gulf Nuclear Station.

(5) 06-RE-1J11-V-0001 (Revision 33), Power Distribution Limits

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Verification (addresses Technical Specifications 4.2.1, 4.2.3, 4.2.4, Table 4.3.1.1-1 items 2b and 2c, and Table 4.3.6-1 item 2.a).

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.(6).17-5-02-500 -(Revision 0), Core Monitoring System Verification, t

is perfomd at the start of a fuel cycle to assure that the

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data. decks. for use with the PowerPlex system are correct for monitoring fuel burnup, control rod exposure and LPRM sensitivi-

-ty and exposure.

Test case calculations are performed to

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confirm proper operation of the computer programs. used in

s PowerPlex, b.

Application of Software and Hardware All of the present core is composed of fuel bundles manufactured by ANF, which also provided the PowerPlex software system for core monitoring.

The software is described in document 2.a(4) and is installed on a PRIME computer, which is backed up by another PRIME.

• Plant instrumentation, including the LPRMs and TIPS remain interfaced to the original plant computer, which now serves as a data link

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rather than a data analyzer. With the original plant computer down, it is possible to manually enter plant data into PowerPlex.

This software is more fully described in the topical report XN-NF-80-19, Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology.

g for Boiling Water Reactors-Neutronic Methods for design and Analysis (March 1983).

This topical report was approved in an NRC letter dated April 7, 1982.

The PRIME computer and software provide the opportunity tc. perform three-dimensional core power distribution analysis and prediction in a manner of minutes and whenever desired. This analytical capability is available to the operations staff around-the-clock. The reactor-engineering staff has two portable terminals, which the on-call engineers keep in their homes. Without returning.to the site, it is possible for them to monitor, analyze, and predict core performance.

Therefore it is never necessary for operations to move a rod at any time without the benefit of reactor engineering judgement based upon review of the predicted effect calculated by a core-design quality code.

A new software system using the CASMO III and MICROBVRN computer programs is currently under development. A topical report will be submitted for review by ONRR, whose approval will be required before implementation on site.

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I c.

Completed Tests and Procedures Surveillance of thermal limits is performed using an approved

procedure, document 2.a(5). The inspector reviewed completed proce-dures and confirmed that the surveillances had been perform with the proper frequency and sati s f actory results during September and October 1989.

It was noted that virtually all of the LPRMs were l

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operable and in use for the power distribution measurements. 'The o-records showed that only-2 to 4 of 168 LPRMs were.out.of service when 7'

any of the measurements was made.

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The inspector also reviewed the copief of document 2.a(6) which were completed before the - start of the current operating cycle and confirmed that the results were satisfactory.

No violations or deviations were identified.

3.

Calibration of Nuclear Instruments (61705)

a.

Documents Reviewed (1) 06-RE-1C51-0-0001 (Revision 27), Local Power Range Monitor Calibration, (addresses Technical Specifications Table 4.3.1,1-1 item 2, 4.3.7.7, 4.3.6, and 4.3.1,1).

(2) 17-5-02-18 (Revision 0), APRM Calibration to Core Thermal Power, i

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is used whenever recalibration is necessary during routine operation, b.

Completed Tests and Procedures Reviewed Calibration of LPRMs by use of the TIPS is accomplished by approved-procedure, document 3.a(1).

Review of completed procedures and surveillance records confirmed that the calibrations had been performed at intervals of 1000 Mwd /T throughout the current cycle.

Procedures for calibration of the ApRMs are contained in document 3.a(1), in which case the APRMs are recalibrated immediately after the LPRM calibrations, and in document 3.a(2), which assures recal-i

.ibration to thermal power every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Review of over 20 procedures completed during the current cycle showed no instances of unacceptable gain adjustment factors.

No violations or deviations were identified.

4.

Thermal Power Monitoring and Instrumentation (61706)

a.

Documents Reviewed

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(1) 09-S-02-1 (Revision 7), Manual Core Heat Balance, is the manual analog of the heat balance program installed on the PRIME

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computers and is used when the PRIMES are unavailable, The procedure uses the same process computer points for data input as the computer calculation. The equation used in the procedure is sufficient for calculation of heat balance of a BWR.

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(2) 17-S-02-2 (Revision 0), Core Thermal Power Verification, is used to monitor the performance of the the computer calculation of

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thermal power and to confirm that the inputs to the calculation

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=are reasonable. Measured steam flow 1s not used in the calcula-it tion of thermal power and is independent of the calculation.

The. licensee has developed a correlation (a third order polyno-mial) between ' steam flow and thermal power, which was derived i

from paired observations of steam flow and thermal power.over a

broad range of power levels.

For a'given thermal power calcula-

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tion, the correlation is used to predict the expected steam

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Dr flow, which is then compared with the measured steam flow.

• Agreement with in two percent of the mean ic considered accept-

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a able.; Larger disagreement triggers ar, investigation of 'the inputs to the thermal power calculation. This test is scheduled to be performed.on a weekly frequency, b.

Completed Tests and Procedures Reviewed

- No example of performance of document 4.a(1) was found for the

current cycle, which speaks well of computer reliability. A records review confirmed that decument 4.a(2) was performed with weekly frequency' during September 1989.

The percentage difference between predicted and observed steam flows ranged from 0.2 to 0.6 percent.

c.

Calibration of Computer Points Used in Thermal Power Analysis

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The computer points used for either the computer or manual calcula-tion of thermal power are identified in document 4.a(1).

LCIs for

.the instruments used in the measurements required for the calculation were reviewed to determine if the computer points were calibrated as part of the process. All points were being calibrated as part of the LCIs with one exception.

The points for recirculation pump. powers-had not been calibrated routinely. The licensee had identified this omission prior to this inspection and had draf ted LCI 07-S-52-B33-1, Reactor Recirc Motor Watt Transducer, to remedy the oversight. The procedure will be performed with refueling outage frequency, and implementation will begin with the next refueling outage.

The schedule for implementation is adequate.

Pump heat is a relatively small contributor to the reactor heat balance, and the indicated pump powers are close to expected values.

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5.

Shutdown Margin and Reactivity Anomaly (61707)

a.

Documents Reviewed (1) ANF-89-079(P), Grand Gulf Unit 1, Cycle 4, Startup and Opera-tions Report.

(2) 06-RE-SB13-V-0401 (Revision 24), Shutdown Margin Demonstration, (addresses Technical Specification 4.1.1.a b).

(3) 06-RE-SB13-V-0017 (Revision 26), Reactivity Anomalies (addresses Technical Specifications 4.1.2.1 and 4.1.2.b).

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b.

. Measurement of Cycle 4 Shutdown Margin The beginning-of-cycle. measurement of shutdown margin is described in

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document 5.a(1). The predicted R factor for increase in core reac-tivity during the cycle from consumption of burnable poisons.in the

't fuel was 0.0 percent rho, which led to a minimum acceptable shutdown?

margin' of 0.38 percent rho for the measurement.

The measured shutdown margin using in-sequence rod withdrawal was 1.42 percent rho, when measured on April 27, 1989, using document 5.a(2).

The

inspector reviewed the data analysis and agreed with the results.

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Measurement ofLReactivity Anomalies

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Surveillance of reactivity anomalies is adequately addressed in document 5.a(3).

Review of procedures completed during the current

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cycle confirmed that the-surveillance was being performed at an acceptable frequency with satisfactory results.

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No violations or deviations were identified.

6.

Observation of Safety-Relief Valve Operation (73756)

On January 1,1990, under work order #00002827, which included special instructions, the licensee opened and closed 12 SRVs in sequence in an

. attempt to. reseat them and reduce heat input to the suppression pool and 3'

. airborne radioactivity to the containment building.

The inspector observed the operation from the vantage point of the STA's console. The plant parameters displayed on the console were adequate to. monitor and evaluate the plant response to the changing heat sink. The responses appeared to be stable and within the expected range of change.

7.

Followup of Open Items (92701)

(Closed) Inspector followup item 50-416/87-27-02: Human factors concerns identified in the PRA-based inspection.

The thirteen items identified in the report were placed on an action items tracking list (PMI 87/8257) and evaluated by the responsible plant organi-zations. Where necessary, corrective actions were identified and imple-mented or are currently in progress.

(Closed) Inspector followup item 50-416/88-13-01: Inspect test results of fuel pool cooling heat exchangers, RHR steam condensing mode, floor drain evaporator performance and heat load, and chemical waste evaporator performance and heat load.

The heat removal performance of fuel pool heat exchanger A was measured on February 26, 1988, and evaluated using the STER heat exchanger performance calculation program.

The estimated maximum heat transfer rate was 4 million Btu /hr, which was close to the expected performance level.

However, with the increased spent fuel storage capacity, the design basis

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capacity of the fuel -pool capacity will not be ~ sufficient over the long term. The licensee is currently considering' options for system modifica-tion. Fuel pool cooling capability and modifications are being discussed

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with and will be reviewed by ONRR prior to implementation; hence, no further followup action by Region II is necessary.

The auxiliary boiler has been removed.

Since it-was the intended heat source for the floor drain and chemical waste evaporators; those systems will not be placed in service.

The.RHR steam condensing mode of operation would require extensive plant i

modifications to implement. There are no plans to perform the modifica-tions, and the system valves are currently locked out.

8.

Exit Interview

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' The inspection scope and findings were summarized on January 12, 1990, with those persons indicated in paragraph 1 above.

The inspector described the areas inspected and discussed in detail the inspection findings.

No dissenting comments were received from the licensee.

Proprietary material' was reviewed in the course of the inspection, but is not included in this report.

9.

Acronyms and Initialisms Used in This Report ANF. -

Advanced-Nuclear Fuels, Inc.

.APRM -

average power range monitor

. Btu /hr-British thermal unit per hour-boiling water reactor BWR

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FSAR -

Final Safety Analysis Report-LCI loop calibration instruction

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LPRM -

local power range monitor

' Mwd megawatt days

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ONRR -

U.S.N.R.C. Office of Nuclear Reactor Regulation reactor heat removal'(system)

RHR

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RTP rated thermal power

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SRV safety relief valve

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STA shift technical advisor

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TIP traveling incore probe

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