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Insp Rept 50-424/88-06 on 880112-13 & 25-29.Violation Noted. Major Areas Inspected:Completed Startup Tests,Thermal Power Monitoring,Estimated Critical Position & Shutdown Margin Calculations & Information Notice Response
	ML20149M423
Person / Time
Site:	Vogtle
Issue date:	02/18/1988
From:	Burnett P, Jape F; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20149M417	List: ML20149M416; ML20149M418; ML20149M423;
References
	50-424-88-06, 50-424-88-6, IEIN-87-020, IEIN-87-20, NUDOCS 8802260039
	Download: ML20149M423 (8)
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-Report Nos.: 50-424/88-06 t

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Licensee: -Georgia Power Company (

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P. O. Box 4545

Atlanta, GA 30302

Docket No.: 50-424 License No.: NPF-68

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Facility Name: Vogtle 1 ;

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Inspection Conducted: January 12-13 and 25-29, 1982 i

Inspector:

E T. truMIEtt~

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Date Signed l

Approved by: /M M/M -

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F. Jape, Sectiori Chie" / i

Engineering Branch

! Division of Reactor Safety l

SUMMARY

Scope: This routine, unannounced inspection addressed the areas of completed

startup tests, thermal power monitoring ECP and shutdown margin calculations,

j response to an Information Notice, ard followup of open items.

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Results: One violation was identified - Inadequate program for review of l'

software used in surveillances - paragraph 6.

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REPORT DETA!LS

1. Persons Contacted

Licensee Employees

G. Bockhold, General Manager, Vogtle Nuclear Operations.

R. M. Bellamy, Plant Manager

W. L. Burmeister. Operations Superintendent

C, L. Cross, Senior Regulatory Specialist

R. J. Florian, Reactor Engineering Supervisar

W. Gabbard, Senior Regulatory Specialist

B.-Gover. Engineering Supervisor

W. F. Kitchens, Operations Manager

W. C. Marsh, Deputy Operations Manager

W. E. Mundy, Quality Assurance Supervisor

W. T. Nicklin, Regulatory Compliance Supervisor

K. Pointer, Senior Plant Engineer, Nuclear Safety and Compliance

P. Rushton, Plant Training Manager

D. H. Smith, Superintendent of Nuclear Operations

R. E. Spinnato, Independent Safety Engineering Group Supervisor

Other licensee employees contacted included engineers, technicians,

operators, and office personnel.

Other Organi:ations

C. B. Holland Westinghouse

W. C. Phoenix, Censul Tec

J. Willis, Westinghouse

NRC Resident Inspectors

J. F. Rogge, Senior Resident Inspector, Operations

R. J. Schepens, Senior Resident Inspector, Construction

C. W. Burger, Resident Inspector

Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized on January 29, 1988,

with those persons indicated in paragraph 1 above. The inspector de-

scribed the areas inspected and discussed in detail the inspection find-

ings. Licensee management made a commitment to continue using the

inverse-multiplication approach to criticality until a more reliable

method of xenon analysis, than the currently used power-block-averaging

method, is justified. Dissenting coments were not received from the

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licensee. Proprietary information is not contained in this report. The i

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findings included:

Violation 424/88-06-01: Failure to have an adequate program to

control computer sof tware used in surveillances - paragraph 6. l

3. Licensee Action on Previous Enforcement-Matters l

This subject was not addressed in the inspection.

4. Unresolved Items

No unresolved items were identified during this inspection. ,

5. Completed Startup Test Procedures (72400, 72616, 72624)

The completed startup test procedures listed below were reviewed for

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completeness, adherence to FSAR test descriptions, and conformance to

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Regulatory Guide 1,68. .

a. 1-55Q-01, Metal Impact Monitoring System Test, is described in FSAR !

14.2.8.2.19. It was started on February 3, 1987 and completed on -

June 30, 1987. All acceptance criteria were satisfied. ,

b. 1-6AE-01, Steam Generator Level Control Test, is described in FSAR

14.2.8.2.25. This test was performed on November 9,1987 in conjunc-

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tion with test 1-65C-02, and used data obtained from that test. The ,

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acceptance criteria were satisfied, but one performance criterion was '

not; steam generator level swings were greater than 10%. This

deficiency was evaluated in TER 16 for this test. The evaluation .

l stated the larger swings were acceptable because the power change was !

greater than the planned 10%, actually 12%. This correlation of !

power change with level seems to miss the point of the test and the ;

function of the level controller. However, see the evaluation ;

l performed under test 1-65C-02 below, j

c. 1-6SC-02 Load Swing Test, is described in FSAR 14.2.8.2 27. It was l

performed at 100% RTP on November 9,1987, by first decreasing power '

by 10% and then, after stabilizing the unit, by increasing power 10%.

All acceptance criteria were satisfied; neither the reactor nor ,

turbine tripped and safety injection was not initiated. No manual ;

intervention was necessary to maintain reactor power. RCS tempera-

ture, pressurizer pressure and level, steam generator levels and

pressures remained within acceptable ranges throughout the test. One

performance criterion was not satisfied; steam generator level swings

were greater than the 5% specified in step 6.3.26. The steam genera-

tor level variations were reviewed by Westinghouse, the NSSS vendor,

and documented in a letter, 5.0. No: GAE301, dated December 15, 1987

The vendor found the actual performance satisfactory in all respects.

Westinghouse did recommend specific actions to improve plant perfor-

mance. Those to reduce feedwater flow oscillations have been

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completed. Fine tuning the steam generator level controllers for

higher power operation has not been done. The licensee believes the

tuning completed for low-power operation is the best compromise;

since that will reduce the number of low-power steam generator level

trips, which have plagued the plant in the past. The test is com-

plete except for removal of test equipment, which was in progress

during this inspection. The test results are sumarized graphically

in Supplement I to the Startup Report.

d. 1-6SE-01, Axial Flux Difference Calibration Test, is described in

FSAR 14.2.8.2.29. The test was performed at 75% RTP (nominal) on May

9, 1987. The test was performed at 100% RTP on June 3, 1987 and

completed on June 30, 1987.

The inspector independently analyzed the data using a least-squares

analysis spreadsheet with the micro computer program

SUPERCALC3(VI.1). The inspector's values for zero-offset currents

and slope of current against offset agreed closely with the licens-

ee's at 75%. To analyze the data obtained at 100%, the inspector

merged them with those from 75%; since only three sets of data, all

at negative offsets, were obtained at 100%. The resulting slopes and

zero-offset currents were little-r anged from the earlier values, but

the variances of each parameter were increased. The correlation

coefficients of the fits were all decreased, but none dropped below

0.92. The licensee choose to analyze the three full-power observa-

tions separately and to recalibrate the PRNIs. Further inspection of

the licensee's routine surveillance results for incore-excore corre-

lations will be performed before reaching a conclusion on which is

the better practice,

e. 1-800-01, Plant Performance, is described in FSAR 14.2.8.2.55. It

was completed for 100% power operations on December 14, 1987. The

test had no acceptance criteria, but the data for evaluating plant

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performance were obtained, and from their review specific recommenda-

tions for future improvements in plant performance and reliability

were developed for consideration by management.

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f. 1-600-13, Power Ascension Test Sequence, is described in FSAR l

14.2.8.2.50. It is complete at 100% power except for the following

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steps:

6.13.9 Perform 1-700-03, Steam Generater Moisture Carryover Test,

(management has stated this test will not be perfomed), !

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6.13.15 Perform 1-6SD01, PERMS (an environmental monitoring system

test).

Nrne of the tests left to be performed are essential. Therefore, the

inspection program for Unit I startup tests is closed, l

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No violations or deviations were identified. .

6. Thermal Power Monitoring (61706)

The licensee has completed the review of the plant computer and manual :

procedure calculations of thermal power that was'first addressed in ,

inspection report 424/87-67. The PROTEUS (plant computer) calculation of l

thermal power accessed at point U1118, has been traced and validated. No

change in the programing proper was required. A data base of constants i

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is used in the calculation. Some of the constants required modification

to reflect the plant or to assure a degree of conservatism when bounding i

values were used. The review is documented in the internal report.-

"Reat.torEngineeringValidationofU1118"_(REV-1118).

In reviewing the default values entered into the database, the inspector

found one to be based upon a non-conservative assumption. If a blowdown -

flow element is faulty, the program turns to the database for substitute .

value. The substitute value assigned was 90 gpm. Since steam flow is i

obtained by subtracting blowdown flow from feedwater flow, power would be

under estimated if blowdown were less than 90 gpm. To be conservative, .

the substitute value should be zero. The licensee was informed of this ;

finding during the first phase of this inspection. Prior to the second t

phase of inspection, the licensee had confirmed the finding of non- !

conservatism and evaluated the maximum effect as +0.2% RTP per affected :

steam generator or 28Mwth over power if all four loops were affected.

By the end of the inspection, steps were underway to change the default ;

value of blowdown flow to 0 gpm prior to restart of the unit. ;

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Procedure 00410-C (Revision 2), Computer Software Control, defines soft- ;

ware as computer programs and data files containing prograriner-specified l

constants, flags and setpoints. The data base discussed above was classi- !

fied as a set of addressable or accessible constants and, hence, changes '

thereto were not subject to the review requirements of the procedure. The !

licensee recognized their vulnerability to unauthorized or unreviewed I

changes to the data base and changeo the access level to the highest, l

level 8, on December 18, 1587. From that time on, only the computer i

engineer has had access to the data base, and, presumably, future changes )'

to the database will be made only in conformance with the requirements of

00410-C, which requires proper review by qualified reviewers prior to

implementation (step 3.1.1).

However, no retrospective review of the changes to the database by a

qualificd re'iewer was performed, but the PRB did require that the values

in the dW ~ se be confimed to be those recommended in REV-1118. Al-

though d' ,i REV-1118 was discussed in detail with the PRB, it did not

function a qualified peer or interdisciplinary reviewer.

As written, 00410-C does not assure that a change in a database constant,

which could affect surveillances performed with the computer or using a

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computer-generated result, as in procedures 14030-1 and 12004-1, would be ;

l recognized and reviewed as a change to a surveillance procedure. l

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This programmatic weakness has been identified as Violation 424/88-06-01:

Failure to have an adequate program to control computer software used in l

surveillances. l

7. Shutdown Margin and Estimated Critical Position Calculations (61707)

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Procedure 14005-1 (Revision 3), Shutdown Margin Calculations, addresses

the surveillance requirements of Technical Specifications 4.1.1.1.1 and

4.1.1.2. The procedure is used in concert with curves and tables in the i

PTDB. Ten procedures completed in the period June 3 to November 11, 1987 *

were reviewed for completeness and arithmetical accuracy. Two were

compared in detail with the PTDB to assure that proper values had been i

abstracted from that document for use in the procedures. No discrepancies

were found. l

Procedure 14940-1 (Revision 5), Estimated Critical Condition Calculation,

is used to calculate critical control rod position for a preset boron

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concentration or critical boron concentration for a preset rod pattern.

It too uses curves and tables from the PTDB. Completed procedures for the

startups on November 12, 1987, November 6, 1987, and October 31,1987 were l

reviewed for completeness, accuracy in using the PTDB data, and comparison ;

of predicted and actual critical configurations. No discrepancies were :

found, and the agreement between predicted and measure configurations

ranged from 12 to 284 pcm, which were within the allowable limits speci- '

fled in the procedure. l

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One potential problem was identified with both procedures. For periods ;

when the reactor has not been at constant power for more than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, an !

"EquivalentXenonPowerWorksheet"(Tab 1.4.3ofthePTDB)isusedto l

calculate an equivalent, xenon-at-saturation power. Experience at another ,

facility has shown that a similar worksheet greatly over estimated xenon ,

concentration when there had been periods of no or low power operation i

included in the average. The xenon reactivity error contributed to the l

occurrence of a startup rate in excess of 16 decades per minute (see 1

Inspection Report 395/85-12-04). It appears that a similar potential l'

exists at Vogtle 1; since the review of shutdown margin calculations

revealed that in %e period 0300 6/6/87 to 0538 6/7/87 the calculated

xenon reactivity increased from -247pcm to -517pcm witit only zero power

operation in the interval.

The licensee is currently developing improved methods of ceiculating

xenon. Until a better method is justified, the licensee has made a '

comitment to compensate for the potential non-conservatism in the present

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method by continuing the practice of approaching criticality by inverse

multiplication monitoring. The licensee will also review the shutdown

margin procedure to assure errors in xenon calculations do not compromise

shutdown margin.

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No violations or deviations were identified.

8. Resolution of Open Items (92701)

(Closed) UNR 424/86-99-01: Natural circulation test and loss of offsite

power test were scheduled to be performed after achieving 75% power rather

than before power ascension and before exceeding 25% power respectively.

The test descriptions in the FSAR, 14.2.8.2.46 and 14.2.8.2.47, were

changed by amendment 33. The tests were successfully performed in accor-

dance with amended descriptions.

(Closed)IFI 424/86-99-02: Sample and control refueling canal boren

concentration during initial fuel loading. Technical Specification 3.9.1

was amended to exempt sampling the refueling canal during initial fuel

loading with the canal level below the vessel flange.

9. Followup of Information Notice (92703)

(Closed) Infonnation Notice 87-20: Hydrogen t.eak in Auxiliary Building.

From discussion with an engineering supervisor the inspector learned:

a. A poll of the industry revealed no consensus on the the valve type to

be used in hydrogen service. However, the licensee has judged it

prudent to replace all packing valves passing hydrogen in the

auxiliary building with diaphragm valves, except for one air operated

control valve to the VCT, for which no diaphragm equivalent was

found. A DCP was written for completion during the first refueling

outage.

b. The hydrogen lines meet seismic design criteria.

c. The low normal flow of hydrogen coupled with normal service tran-

sients make use of the excess flow check valves on the skid impracti-

cal; since they would have to be set so low.

d. The normal use of hydrogen is 600 SFC/ day. Recently it has risen to l

13,000 SCF/ day. This problem is being corrected during the current l

outage. The hydrogen lost during recent operatior did not enter the

auxiliary building atmosphere, but went directly to a stack which '

exited the turbine building roof. The hydrogen concentration in the j

stack was as high as 17%.

e. The HVAC system has been operation since the first event, which i

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generated the notice; so the potential for building up hydrogen

within the facility has been greatly reduced,

f. An inline flow meter is being added to the system. It is expected to

arrive on site on February 2, 1988.

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This item is closed.

Attachment: l

Acronp s and Initialisms l

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ATTACHMENT

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ACRONYMS AND INITIALISMS (

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, BD - blow down

-CFR - Code of Federal Regulations l

1 DCP - design change procedure i

1 ECP - estimated critical position j

EFPD- effective full power days

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FSAR- Final Safety Analysis Report !

gpm - gallon per minute i

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HVAC- heating, ventillating and air conditioning l

LER - licensee event report, required by 10 CFR 50.73 :

j MCB - main control board !

i hth- megawatts of thermal power ;

NSS - nuclear steam system l

j NSSS- nuclear steam supply system )

. pcm - percent milli-rho )

3 PRB - Plant Review Board !

! PRN!- power range nuclear instrument I

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] PTDB. Plant Technical Data Book i

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RCP - reactor coolant pump !

! RCS - reactor coolant system i

j RTP - rated thermal power, the licensee limit in Nth, f

TER - test evaluation report j

_ VCT - volume control tank i

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ML20149M423

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