IR 05000261/1982008

IE Insp Rept 50-261/82-08 on 820222-26.Noncompliance Noted: Procedure CPL-PT-8 Not Adequate for Evaluation of RCS Leakage Test Results

ML20052G264
Person / Time
Site:	Robinson
Issue date:	03/24/1982
From:	Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20052G233	List: IR 05000261/1982008 ML20052G245 ML20052G255 ML20052G258
References
50-261-82-08, 50-261-82-8, NUDOCS 8205140464
Download: ML20052G264 (4)

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UNITED STATES P

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Report No. 50-261/82-08 Licensee: Carolina Power and Light 411 Fayetteville Street Raleigh, North Carolina Facility Name:

H. B. Robinson, Unit 2 Docket No. 50-261 License No. DPR-23 Inspection at H. B. Ro inson plant site near Hartsville, South Carolina

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Inspe tor:

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Date Signed P. T. Bu n

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Approved by:

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Date Signed F. J e, Section Chief Engineering Inspection E anch Division of Engineering and Technical Programs SUMMARY Inspection on February 22-26, 1982 Areas Inspected This routine, unannounced inspection involved 32 inspector-hours on site in the areas of reactor coolant system leak-ro+.e measurements and control of heavy loads.

Results l

No violations or deviations were identi fi ed in one area and one violation (procedure not adequate for the evalution of test results paragraph 5) was found in the other area.

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t REPORT DETAILS 1.

Persons Contacted Licensee-Employees R. B. Starkey, Plant Manager

• W. Crawford, Manager, Operations and Maintenance

• J. Curley, Manager, Technical Support

• F. Lowery, Operating Supervisor

• D. Bain, GAQC Project Specialist

• C. Wright, Specialist, Regulatory Compliance Other. licensee employees contacted included two technicians, six operators, and two shift foremen.

NRC Resident Inspector S. Weise

• Attended exit interview 2.

Exit Interview The inspection scope and findings were summarized on February 26, 1982, with'

those persons indicated in paragraph I above. The licensee acknowledged the-findings without significant comment.

3.

Licensee Action on Previous Inspection Findings Not inspected.

4.

Unresolved Items Unresolved items were not identified during this inspection.

5.

Evaluation of Reactor Coolant System Leakage The inspector obtained parametric data, tank capacities and calibration factors, necessary to characterize the fluid inputs to and directly mea-sureable releases from the reactor coolant systems (RCS). These data become constant parameters of a generalized computer program for RCS leakage calculations in specific application to the Robinson Unit 2 plant. This computer program, RCSLK7, is described in the February 16, 1982 memorandum by R. L.

Baer, OIE, subject:

" PLANT DATA FOR REACTOR COOLANT LEAKAGE CALCULATIONS," and in OIE temporary instruction TI2515/48 issued on March 2, 198.s

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The inspector also collected operational data, tank volumes and flows as a function of time, for use with RCSLK7. Five sets of observations were made over periods of one to two hours. The time periods selected were those during which no samples were being drawn from the volume control tank (VCT),

power was steady and no dilution or boration of the VCT was in progress.

The system changes observed for all five sets of observations were so small that it was possible to tell by inspection, without recourse to RCSLK7, that the RCS leak rate was appreciably less than the one gpm limit addressed in technical specification 3.1.5.1.

All observations were made at a nominal 81 percent of rated thermal power.

The inspector reviewed the licensee's procedure CPL-PT-8, " Reactor Coolant System Leakage Evaluation" for content and spot checked the results obtained

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in January and February 1982.

In revievir[p the derivation of coefficients used in PT-8 the inspector found that in the term for moderator temperature correction, the wrong RCS capa-city had been used. The value used included the solid volume of the pres-surizer.

That volume should not have been included. The pressurizer was accounted for separately in the procedure as a saturated, partially-filled volume with a different characteristic temperature than that reflected by the average RCS temperature. Consequently, the magnitude of the coefficient (multiplier of the change in average temperature (T-AVG) during the measure-ment) was in error by about 16 percent.

The error was conservative for increasing temperature and non-conservative for decreasing temperature.

Further, the coefficient as derived was appropriate only to T-AVG in the range of 570 F to 580 F.

During steady-state power operations the effect of the errors in the multiplier of the change in T-AVG during the measurement was small in practice. Among the completed procedures (PT-8) reviewed the largest recorded change in T-AVG was 0.1 F over one hour. Any induced error in the RCS leak rate was then less than one-tenth gallon per minute.

Technical specification table 4.1-3, item 9, requires that primary system leakage be evaluated daily when the reactor coolant system is above the cold shutdown condition.

PT-8 was not adequate to satisfy that specification over the necessary range of application including current power operating conditions of a reduced T-AVG of 537.9 F.

The procedure was not adequate for the evaluation of test results, and hence is a violation of ter.hnical specification 6.8.1 on procedures VIO (50-261/82-08-01).

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Response to the Generic Letter on Heavy Loads (81-07)

The inspector reviewed the following documents and procedures:

a.

" Response to Request by NRC for Additional Information," Part I, G. G. Ward (CPL) 7/27/81, b.

FT-9.4, " Reactor Vessel Stud Removal and Replacement", Revision 8,

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c.

FT-9.6, " Head Lif ting Rig Operating Instructions", Revision 8, d.

FT-9.9,

" Upper Internals Lifting Rig Operating Instruction",

Revision 8, FT-10, " Refueling Outage Operations and Activities", Revision 44, e.

f.

FT-13, " Operating the Spent Euel Cask Handling Crane", Revision 3, g.

FT-15.2, " Spent Fuel Cask Handling Instructions for Loading and Ship-ping of PWR Fuel", Revision 12, h.

Maintenance Instructions:

(1) MP1-5, " Operation, Testing and Inspection of Cranes and Materials Handling Equipment", Revis'on 0, (2) MP1-5A, " Overhead and Gantry Crane Maintenance", Revision 0.

Procedures b through g each addressed safe load paths appropriate to the procedure and included drawings showing the areas for safe and prohibited paths for the load or loads being addressed. The procedures conformed to

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the licensee's response (item a) to the generic letter and NUREG-0612

" Control of Heavy Loads at Nuclear Power Plants."

NUREG-0612 states crane operators should be trained and qualified in accord-ance with Chapter 2-3 of ANSI B30.2-1976, " Overhead and Gantry Cranes." The licensee has instituted a crane operator training program to implement that requirement and procedures b, e, and f make specific reference to qualified operators. Neither the training program nor the list of qualified operators was audited in this inspection.

NUREG-0612 also requires that cranes be inspected, tested and maintained in accordance with Chapter 2-2 of ANSI B30.2.

The licensee's program as

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described in the procedures listed above fully addresses the requirements of Chapter 2-2 of those cranes within the purview of generic letter 81-07.

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