IR 05000338/1980005
| ML19309F806 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/14/1980 |
| From: | Burnett P, Upright C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19309F802 | List: |
| References | |
| 50-338-80-05, 50-338-80-5, 50-339-80-04, 50-339-80-4, NUDOCS 8005010277 | |
| Download: ML19309F806 (6) | |
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UNITED STATES 8006010277 8'
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NUCLEAR REGULATORY COMMISSION n
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E REGION 11
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101 MARIETTA ST N.W., SUITE 3100 d'
ATLANTA, GEORGIA 30303 Report Nos. 50-338/80-05 and 50-339/80-04 Licensee: Virginia Electric and Power Company Richmond, Virginia 23261 Facility Name: North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4, CPPR-78
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Inspection at North Anna site near Mineral, Virginia Inspector: [// f // w /
89O P. "r. Burn ~ett gateGigned
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Approved by: [ [ f m
/84 C. M. Uprig Ac g Section Chief, RONS Branch Yate,61gned SUMMARY Inspection on February 11-15, 1980 Areas Inspected This routine, unannounced inspection involved 31 inspector-hours on site in the areas of core performance monitoring (unit 1), post-refueling startup tests (unit 1) and initial startup tests (unit 2).
Results Of the three areas inspected, no items of noncompliance or deviations were identified in two areas; one item of noncompliance was found in one area (Deficiency - Failure to document surveillance tests, paragraph 5).
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DETAILS 1.
Persons Contacted Licensee Personnel
- W. R. Cartwright, Station Manager
- E. W. Harrell, Superintendent of Maintenance
- E. R. Smith, Jr., Superintendent of Technical Services
- G. A. Kahn, Engineering Supervisor
- W. R. Madison, NRC Coordinator
- D. L. Smith, Resident QC Engineer
- J. W. Ogren, Supervisor Administrative Services
- J. P. Smith, Reactor Engineer
- M. Stokes, Fire Marshall NRC Personnel
- M. S. Kidd, Senior Resident Inspector
- A. Tattersall, Resident Inspector
- W. H. Miller, Inspector, Region II
- Attended exit interveiw.
2.
Exit Interview A meeting was held with those persons indicated in paragraph 1 on February 15, 1980. The scope and findings of the inspection were presented. The findings of a deficiency for failures to adequately document the surveillance of reactivity anomally was received without substantial comment. The inspector noted that station surveillance procedures and post refueling startup tests related to core performance were generally limited to data acquisition, and that the methods of data reduction and analysis were not inspectable on site. He stated that an inspection to address these concerns would be directed toward the corporate offices in the future.
3.
Licensee Action on Previous Inspection Findings
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l Not inspected.
4.
Unresolved Items No unresolved items were identified during this inspection.
5.
Review of Unit 1, Cycle 1, Core Performance a.
Core Reactivity Balance Technical Specification 4.1.1.1.2 requires in part that the overall
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core reactivity balance be compared to predicted values to demonstrate
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agreement within plus or minus I percent delta k at least once per 31 effective full power days (EFPD).
The licensee has written the procedure 1-PT-11, " Core Reactivity Salance" to accomplish the required surveillance.
The inspector reviewed completed copies of this procedure covering the period of June 1978 to September 1979. The core burnup in megawatt days per metric ton is a datum of each completed procedure.
In comparing the span between sucessive surveillances it was noted that on two occasions the span exceeded the equivalent of 31 effective full power days including a 25% grace period. However, the burnup span seemed large for the corresponding span in calendar days. Further review of
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other licensee data led to the conclusion that the burnup reported on the 1-PT-11 data sheet bore no relationship to the date on the data sheet and, further, that it did not provide a record of adequate surveillance of the reactivity anomally. Other licensee provided information including the monthly summaries of fuel performance characteristics prepared by the Nuclear Fuels Organization at Corporate Headquarters using plant data and extended discussions with the station reactor engineer convinced the inspector that the surveillance for reactivity anomallies was carried out in a timely and technically correct fashion with acceptable results. Nevertheless, no acceptable formal record of the surveillance exists. Retention of such records is required by technical specification 6.10.1.d.
This failure to conform to the technical specifications has been identified as a deficiency (80-05-01).
b.
Late-In-Cycle Moderator Coefficient Technical Specification 4.1.1.4.b requires in part that the moderator temperature coefficient shall be measured within seven effective full power days af ter reaching an equilibrium boron concentration of 300 parts per million and compared with the limiting value for the all-rods-out full power condition. The licensee's procedure to meet this surveillance requirement is 1-PT-13, " Moderator Temperature Coefficient and Power".
The inspector reviewed a copy of this procedure completed on May 22, 1979 with a measured boron concentration of 295 parts per million. The procedure was completed sucessfully in that data for two types of temperatur coefficient measurements were obtained. One method was based on introducing a boron dilution transient and the second used control rod motion to initiate the transient. Neither this nor any other station procedure addresses the method by which the experimental data are analyzed and reduced to a moderator temperature coefficient. That analysis was performed off site by the Nuclear Fuels Organization, which reported a successful result without providing details of the analysis. The subject of the method of data analysis used in determining the moderator temperature coefficient after 300 parts per million boron concentration has been identified for inspector followup action (80-05-02).
The reported measured value of the moderator temperature coefficient was -27.1 pcm/*F.
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c.
Core Power Distribution Monitoring The records of the following surveillance tests related to core power distribution monitoring wert reviewed by the inspector for the period covered by cycle 1.
Satisfactory frequencies of performance and completed performance were confirmed in all cases. The specific procedures were:
(1)
1-PT-20.2, " Axial Flux Difference Every 31 Effective Full Power Days",
(2)
1-PT-21.1, " Reactor Core Flux Mapping",
(3)
1-PT-21.2, " Hot Channel Factors",
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(4)
1-PT-22, "APDMS Update",
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(5)
1-PT-22.1, "In-Core versus Er-Core Offset Evaluation",
(6)
1-PT-22.2, "In-Core verus Ex-Core Axial Offset", and (7)
1-PT-23, " Quadrant Power Tilt Ratio".
The tests named above are for the most part dependent upon core flux mapping using the movable in-core probe system. The data generated rm site are analyzed and evaluated by the offsite organization, which then issues a letter report. Completion of many of the pts above by on-site personnel amounts in many cases to to little more than transcribing numbers from the letter report to a periodic' test data sheet. Again, the methods of data analysis and results evaluation are not inspectable onsite. This and related questions have been identified for inspector followup (80-05-03).
6.
Review of Unit 1 Cycle 2 Startup Tests a.
Preparatory Activities The inspector reviewed the following completed surveillance tests:
(1)
1-PT-25, " Rod Position Indication System Functional Tests", which was completed on December 10, 1979; (2)
1-PT-17.2, " Rod Drop Time", which was completed on January 12, 1980, and (3)
1-0P-5.3, " Reactor Coolant System Leak Test", which was completed on January 15, 1980.
No questions arose from review and discussion of these tests.
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b.
Initial Criticality Normal unit operating procedures were used to bring unit from the refueling mode to initial criticality. Procedure 1-OP-1.4 was used in going from the hot shutdown to hot standby mode. This procedure provides for withdrawal of the shutdown rod banks and subsequent to that and exceeding 530 degrees Fahrenheit initates dilution of the coolant boron concentration to that predicted for the all-rods-out critical condition. Next procedure 1-0P-1.5 guided the approach to criticality in which inverse multiplication measurements were made every fifty steps of control rod motion as the control banks were withdrawn in normal sequence and normal 100 step overlap.
Criticality was obtained with 108 steps of D bank withdrawn. The inverse multi-plication plot attached to the procedure displayed the expected the shape of such a plot. Over the next two week period this proced'ure and the inverse multiplication approach was used on eight ocassions to get critical configurations with a variety of rod patterns and boron concentrations. Teview of the critical endpoints as recorded in the procedures indicated, in some cases, significant differences in control rod position for relatively small changes in boron concentration. The inspector questioned whether the recorded boron concentrations were in fact correct and if there was some time delay in obtaining the correct boron determination to go with the rod pattern. These questions were brought to the attention of the licensee and will be followed up by the inspector at a later inspection (80-05-04).
c.
Zero Power Physics Tests This surveillance testing was guided by procedure 1-PT-94 " Refueling Nuclear Design Check Test".
The inspector reviewed the completed tests and the appendices thereto, which were used to guide the specific measurements. Appendix B addressed the operational check out of the reactivity computer. Review of the completed work confirmed that the reactivity computer had been calibrated for both positive and negative reactivity insertions, and later review of the application of the computer in control rod calibration and other activities confirmed that the reactivities measured were within the calibrated range.
Appendix C addressed boron endpoint determinations for various control rod and temperature configurations. The inspector had no questions related to the work shown on the data sheets for these determinations.
However, these results were also used to determining the boron worth coefficient also known as the differential boron worth. This work was performed by the off-site organization and only the results were reported to the station.
Review of the procedure used to determine the boron worth coefficient has been deferred until a later inspection (80-05-05).
The inspector had no questions arising from his review of Appendix D which was used to measure isothermal temperature efficient or appendices E and F,which were used for control-rod-worth measurement r
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Appendix G was used to measure control-rod-bank worth using the rod-swap technique. The methods of predicting results and the results themselves will be discussed further with the licensee, as a point of information rather than a point of concern (80-05-06).
7.
Startup Test Program for Unit 2 Changes in the planned startup test prorgram as they relate to core perfor-mance, measurements were discussed briefly with licensee personnel.
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