IR 05000424/1987024

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Insp Rept 50-424/87-24 on 870306-20.Violations Noted: Failure to Perform Adequate Surveillance of RCS Leakage
ML20214P370
Person / Time
Site: Vogtle Southern Nuclear icon.png
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:
References
TASK-2.E.1.2, TASK-TM 50-424-87-24, NUDOCS 8706030341
Preceding documents:
Download: ML20214P370 (7)


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UNITE 3 STATES Af hp NUCLEAR RESULATORY COMMISSION -

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

P. O. Box 4545

~ Atlanta, GA ~30302

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

NPF-68

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Facility Name: Vogtle 1

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Inspection Conducted: March 7 20, 1987 Inspectors -

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(If ML P. T. Bur Date Signed

O. ft. L m a S//7/87 A. R. Long

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J. 'L. Mathis Date Signed Approved by:

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F. Jape, Chief f/

Date Signed Engineering Branc Division of Reactor Safety 1.

i SUMMARY 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

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tests and surveillance procedures..

Results:

One _ violation was identified, Failure to perform an adequate c

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surveillance of reactor coolant system leakage - paragraph 5.d.

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REPORT DETAILS 1.

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
  • W. 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 personnel.

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 2.

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 findings.

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dissenting comments were received from the licensee.

Proprietary

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information was reviewed during the course of the inspection but is not incorporated in this report.

Violation 425/87-24-01:

Failure to perform an adequate surveillance of reactor coolant system leakage - paragraph 5.d.

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Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio., -.

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Unresolved Items Unresolved items were not identified during this: inspection.

5.

Review of Precritical Testing -(72596, 61728)-

The following completed precritical test procedures were reviewed:

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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

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significantly less than the 2.2 second < limit of - technical

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specifications.

Drop time was measured from the beginning of decay

of stationary gripper coil voltage to dashpot entry.

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

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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

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seconds.

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

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1-588-02, Pressurizer Heater and Spray. Capability 1and Continuous

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

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pressurizer pressure response to opening of _ both pressurizer spray.

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-criteria were satisfied.

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1-5BB-06, Reactor Coolant Flow Coastdown, was performed on March 5, 1987. There were three TERs written against the procedure. The flow

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coastdown time. constant measured following - the loss 'of-all four 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:

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determined during-the test equalled 0.936 seconds.

All ~ test -

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

i d.

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

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

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

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

coolant drain tank (RCDT), the licensee took further credit' for other

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identified leakage of 4.56 gpm, which came primarily from leakage

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measured through a valve in the seal water system., The' net 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 report.

-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

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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 the.

required surveillance. This has been identified as VIO 424/87-24-01:

Failure to perform an adequate surveillance of reactor coolant system leakage.

No other violations or deviations were identified.

6.

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, 1987.

In addition, the obtaining and analysis of pressurizer and reactor coolant system boron samples were witnessed.

Independent statistical analyses (chi squared tests) were performed on the source range nuclear instruments to confirm proper functioning.

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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 procedure.

No violations or deviations were identified.

7.

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, a.

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 D.

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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.

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b.

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.

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licensee initiated the action statement to establish rod withdrawal

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limits to assure that the operating moderator coefficient would be-negative.

Implementation and enforcement of those limits will be inspected.

during a later inspection.

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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'B.

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

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j-predicted values.

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No violations or deviations were identified 8.

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 -

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

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

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identified in preliminary - tests.

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 ft.

All acceptance criteria were met, and all test exceptions were resolved without' invalidating the test. TMI Action Item II.E.1.2 is closed.

No violations or deviations were identified.

ATTACHMENTS:

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Parameter List 2.

Reactor Cooling System Leak Rates

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ATTACHMENT 1

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PARAMETER LIST

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Unit Identification:

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Plant Name.

VOGTLE

Unit Number

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Docket Number 50-424 Nuclear Steam System Supplier Westinghouse

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Vessel and Piping:

Volume 10662 cubic feet

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Pressurizer:

Level Units

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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

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Volume Control Tank:

I Level Units

%

Calibration Curve

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Slope

'159.4 pounds per %

Upper Level Limit 100 %

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Lower level limit O%

Geometric Method Available No

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Drain Tank:

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Level Units

%

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Calibration Curve Slope 27.97 pounds per %

Upper Level Limit 75 %

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Lower level limit 20 %

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Geometric Method Available No

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Relief Tank:

Level Units

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Calibration Curve

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Slope 1249 pounds per %

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Upper Level Limit 64.74 %.

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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 2251.3 2259.1 T Ave, degrees F 556 556 Water Levels Pressurizer, %

25.83 23.16 Relief Tank, %

61.5 61.8 Volume Control Tank, %

56.52 36.5 Drain Tank, %

72.1 Water Charged = 0 gal Water Drained = 0 gal TEST RESULTS Change in Water Inventory in pounds:

Vessel & Piping

Relief Tank (1)

375 Pressurizer-1649 Drain Tank (1)

Volume Control Tank (1) -3191


Less: Water Charged

Collected Leakage 377 Plus: Water Drained

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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 curve.

(2)

Determined from tank dimensions.

(3)

The density used for converting inventory change to leak rate was 62.31 pounds / cubic foot based on standard conditions.

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