IR 05000424/1987024

Insp Rept 50-424/87-24 on 870306-20.Violations Noted: Failure to Perform Adequate Surveillance of RCS Leakage

ML20214P370
Person / Time
Site:	Vogtle
Issue date:	05/15/1987
From:	Burnett P, Jape F, Long A, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214P355	List: IR 05000424/1987024 ML20214P352 ML20214P362 ML20214P370
References
TASK-2.E.1.2, TASK-TM 50-424-87-24, NUDOCS 8706030341
Preceding documents:	ML20235Y217 (21 July 1987)
Download: ML20214P370 (7)

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Report No.: 50-424/87-24 e Licensee: Georgia Power Company

P. O. Box 4545 1 ~ Atlanta, GA ~30302 ,

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

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Facility Name: Vogtle 1

) Inspection Conducted: March 7 20, 1987 Inspectors - 44,- (If ML Date Signed 4 P. T. Bur O. ft . L m a S//7/87 A. R. Long Datt Signed

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k.f. Wh J. 'L. Mathis sh4187 Date Signed Approved by: > f[rM7 p F. Jape, Chief f/

Engineering Branc Date Signed Division of Reactor Safety 1.

i SUMMARY

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Scope: This routine, unannounced inspection was performed to witness initial

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criticality, performance of zero power physics tests, and to review completed tests and surveillance procedure c Results: One _ violation was identified, Failure to perform an adequate surveillance of reactor coolant system leakage - paragraph '

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REPORT DETAILS Persons Contacted Licensee Employees

• C. E. Belflower, Quality Assurance Site Manager

• W. L. Burmeister, Operations Supervisor

• J. F. D'Amico, Nuclear Safety and Compliance Manager J. A. Edwards, Senior Nuclear Specialist - Operations R. J. Florian, Reactor Engineering Supervisor G. R. Frederick, Senior QA Engineer

• C. Gabbard, Regulatory Specialist T. Greene, Plant Manager T. S. Hargis, Operations Shift Supervisor C. W. Hayes, Vogtle QA Manager W. F. Kitchens, Manager of Operations C. F. Meyer, Operations Superintendent M. J. Rowe, Operations Superintendent J. Schwartzwelder, Operations Technical Assistant Other licensee employees contacted included engineers, technicians, operators, and office personne Other Organizations NRC Resident Inspectors H. H. Livermore, Senior Resident Inspector, Construction R. F. Rogge, Senior Resident Inspector, Operations R. J. Schepens, Resident Inspector

• Attended exit interview Exit Interview The inspection scope and findings were summarized on March 20, 1987, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection finding No ,

dissenting comments were received from the licensee. Proprietary '

information was reviewed during the course of the inspection but is not incorporated in this repor Violation 425/87-24-01: Failure to perform an adequate surveillance of l reactor coolant system leakage - paragraph j Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio l

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Unresolved Items Unresolved items were not identified during this

inspectio . Review of Precritical Testing -(72596, 61728)-

The following completed precritical test procedures were reviewed:

t 1-5SF-04, Rod Drop Time, was performed at full flow, four - reactor -

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1- coolant purr.ps running hot condition, -557 F. and' 2235 psig, on

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j March 1, 1987. The drop times of ~ all 53 control-and safety rods'were significantly less than the 2.2 second < limit of - technical

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specification Drop time was measured from the beginning of decay

of stationary gripper coil voltage to dashpot entr Test Evaluation . Report (TER) 1-5SF-04-01 was 'writte'n ' beca'use two control rods, H02 in control bank C group 1 and H08 in control bank D:

group 2, were outside the plus or minus- two standard- deviation

. (sigma) limit specified in acceptance criterion.9.3. Each rod was -

retested by six additional drops. Each failed the retest criterion that the span of the six drops for each rod be less - than 0.02 -

second Subsequent evaluation by the~ licersee confirmed that the distribution of drop times for each rod was statistically expected.-

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I , Pressurizer Heater and Spray . Capability 1and Continuous

Spray Flow Verification Test, was performed on March 5,- 1987. The

pressurizer pressure response to opening of _ both pressurizer spra valves was within the allowable range, as-was the pressure response to activation of all pressurizer heaters. - The pressurizer power operated relief valves open in two seconds or less.-- All acceptance-

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, criteria were satisfie BB-06, Reactor Coolant Flow Coastdown, was performed on March 5,

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1987. There were three TERs written against the procedure. The flow coastdown time . constant measured following - the loss 'of- all four

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reactor coolant pumps was 13.175 seconds, which.is-greater than the

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design value. The reactor coolant low flow trip loop time response:

determined during- the test equalled 0.936 seconds. : All ~ test -

exceptions appeared reasonable and did not invalidate.the test.-

i Surveillance Procedure No. 14905-1 (Revision 5), RCS : Leakage .

Calculations (Inventory Balance),. was performed on March -7,1987 to satisfy Technical Specification surveillance requirement 4.4.6.2.1 '

The reported total leakage was 5.67 gpm.- After correcting for the --

identified leakage-to the pressurizer relief tank (PRT) and reactor.

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coolant drain tank (RCDT), the licensee took further credit' for other identified leakage of 4.56 gpm, which came primarily from leakage ,

measured through a valve in the seal water system., The' net l l unidentified leakage of 0.76 gpm was - acceptable. To. verify the l acceptability of the procedure proper and the constants used in it, i the same data recorded by the licensee were analyzed using RCSLK9. -

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l That microcomputer program is fully described in NUREG-1107, "RCSLK9; Reactor Coolant System Leak Rate Determination for PWRs." .The plant-specific data necessary to customize the program for use on Vogtle Unit' I were obtained from the FSAR and the Plant Technical Data Book. Those plant parameters are given in Attachment 1 to this repor The results of the RCSLK9 calculation (Attachment 2) were significantly different, except for good agreement with the amount of leakage into the PRT and the RCDT. - The total or gross leakages-differed by an unacceptability large 1.16 gpm. . Use of RCSLK9 at ,

other facilities has demonstrated consistent agreement within 0.2 gpm for acceptable procedures. A detailed review of procedure 14905-1 revealed significant errors in the constant used to correct for changes in pressurizer level and in the equation used to adjust for changes in reactor coolant system average temperature. .Although in the instance examined the error could .be considered conservative, that would not be the case for all expected variations in level or temperature. Hence, the procedure is not. adequate to perform th required surveillance. This has been identified as VIO 424/87-24-01:

Failure to perform an adequate surveillance of reactor coolant system leakag No other violations or deviations were identifie . Initial Criticality Witnessing (72592)

The inspectors witnessed the withdrawal of. the control banks, initiation of dilution to criticality, and were in the control room through most of the dilution process including the attainment of criticality at 8:35 am on March 9, 198 In addition, the obtaining and analysis of pressurizer and reactor coolant system boron samples were witnesse Independent statistical analyses (chi squared tests) were performed on the source

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range nuclear instruments to confirm proper functionin The initial critical configuration of control bank D and boron concentra-tion was in good agreement with the predicted configuration. Criticality was achieved in a well-controlled manner and in full adherence to procedur No violations or deviations were identifie . Zero Power Physics Tests (72572)

Portions of the following tests were witnessed in the control room.- .The completed test procedures were reviewed for completeness and calculations within the procedures were spot-checked, Boron Endpoint Measurements Boron endpoint measurements were performed in accordance with '

procedure for the following control rod configurations: all rods out

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~4 (AR0), control; bank D fully- inserted, control' banks D and C fully i inserted, control banks D, C, and B fully inserted, control ~ banks _

C, B and A fully-inserted. For each configuration, the measured and'

predicted values agreed within the ' acceptance criterion on +/-- 10% of'

j the predicted-values.

a l Isothermal Temperature Coefficients (61708)

The isothermal temperature coefficient was. measured in accordance with procedure for the AR0 and D-bank-in_ configurations. At.ARO the i corresponding moderator temperature coefficient :.was slightly positive, which is contrary to Technical Specifications. The

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licensee initiated the action statement to establish rod withdrawal limits to assure that the operating moderator coefficient would be-negativ Implementation and enforcement of those limits will be inspecte during a later inspectio ; Control Rod Worth Meas'urements (61710)

Control rod worth measurements were -performed in; accordance with procedure for each control bank-in succession starting'with control-

bank _D and proceeding to shutdown bank' Each

bank measurement satisfied the acceptance criterion of +/- 10% agreement with '

j- predicted value l

} No violations or deviations were identified Followup to Open Items (92701)

(Closed) TMI Action Item II.E.1- 2, " Auxiliary Feedwater System Automatic

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j Initiation and- Flow Indication." The review of diagrams and drawings documented in inspection report 50-424/86-90 confirmed that the auxiliary feedwater (AFW) system automatic initiation and flowrate indication were ~

designed in accordance with requirements. However, testing of the system -

was not complete when the report was . issued because ' of required -

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modifications to the motor operated discharge valves- and flow orifices -

identified in preliminary - test Retesting was completed on March 1, 1987.

The inspector reviewed the section of the completed test procedure that-involved retesting of the motor operated discharge valves ' and flow

orifices. .The motor driven AFW pumps each delivered a minimum of-630 gpm and 1175 gpm in concert at a total discharge head of 3500.+ 105--O f All acceptance criteria were met, and all test exceptions were resolved without' invalidating the test. TMI Action Item II.E.1.2 is close No violations or deviations were identifie ATTACHMENTS:

Parameter List

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~ . Reactor Cooling System Leak Rates

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ATTACHMENT 1 -

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PARAMETER LIST

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Unit Identification:

Plant Nam VOGTLE  ;

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Unit Number 1 Docket Number 50-424

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Nuclear Steam System Supplier Westinghouse

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Vessel and Piping:

Volume 10662 cubic feet

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Pressurizer:

Level Units  %.

Temperature Compensated No e Calibration Curve

Slope 617.52 pounds-per %

Upper Level Limit 100 %

Lower level Limit 0-%

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Relief Relief Tank j

d i Volume Control Tank:

I Level Units  %

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Calibration Curve Slope '159.4 pounds per %

, Upper Level Limit 100 %

Lower level limit O%

, Geometric Method Available No

. Drain Tank:

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Level Units  %

Calibration Curve Slope 27.97 pounds per %

Upper Level Limit 75 %

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Lower level limit 20 %

Geometric Method Available No

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Relief Tank:

Level Units  %

Calibration Curve '

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Slope 1249 pounds per %

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Upper Level Limit 64.74 %.

Lower level limit 24.74 %

Geometric Method Available No

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ATTACHMENT 2

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NRC INDEPENDENT MEASUREMENTS PROGRAM REACTOR COOLING SYSTEM LEAK RATES STATION: VOGTLE TEST DATE : March 7, 1987 UNIT : 1 START TIME: 0451 DOCKET : 50-424 DURATION 2.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> TEST DATA Initial Final System Parameters Pressure, psia 225 .1 T Ave, degrees F 556 556 Water Levels Pressurizer, % 25.83 23.16 Relief Tank, % 6 .8 Volume Control Tank, % 56.52 3 Drain Tank, % 72 7 Water Charged = 0 gal Water Drained = 0 gal TEST RESULTS Change in Water Inventory in pounds:

Vessel & Piping 60 Relief Tank (1) 375 Pressurizer -1649 Drain Tank (1) 3 Volume Control Tank (1) -3191 ------

Less: Water Charged 0 Collected Leakage 377 Plus: Water Drained 0

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Cooling System -4780 Leak Rates in gpm (3):

Gross 4.51 Identified 0.36 Unidentified 4.16 (1) Determined from tank calibration curv (2) Determined from tank dimension (3) The density used for converting inventory change to leak rate was 62.31 pounds / cubic foot based on standard conditions.