IR 05000324/1986017
| ML20205A959 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/09/1986 |
| From: | Burnett P, Jape F, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20205A929 | List: |
| References | |
| 50-324-86-17, 50-325-86-16, NUDOCS 8608110489 | |
| Download: ML20205A959 (4) | |
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p KEcq UNITED STATES
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NUCLEAR REGULATORY COMMISSION o.
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101 MARIETTA STREET,N.W.
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ATLANT A, GEORGI A 30323
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Report Nos.: 50-325/86-16 and 50-324/86-17 Licensee: Carolina Power and Light Company
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P. O. Box 1551 Raleigh, NC 27602 i
Docket Nos.:
50-325 and 50-324 License Nos.:
DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: June 9-13, 1986 Inspectors:
_.k Noh 7!9f66 J. L. Mathis Date Signed b V NsaJs
,h /84 P '.i T. 13u rn e tt Date Signed Approved by: N I /J Nu w/L 7/9/d F. Jape, Section Chief, T4st Programs (late Signed
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Section, Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection was conducted in the areas of post-refueling initial criticality and tests (Unit 2) and routine core perform-ance surveillance (Unit 1).
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Results: No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
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C. R. Dietz, General Manager
- E. A. Bishop, Manager-0peration
- E. B. Wilson, Nuclear Engineering Supervisor
- W. M. Hogle, System Engineering Supervisor
- R. M. Poulk, Sr. Specialist - Regulatory Compliance
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K. A. Scott, Nuclear Engineer M. A. Alford, Nuclear Engineer G. Wertz, Nuclear Engineer
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L. Johnson, Shift Operation Supervisor
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P. T. McNeil, Shift Operation Supervisor Other licensee employees contacted included shift supervisors, engineers, operators, and office personnel.
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NRC Resident Inspectors
- W. Ruland, Senior Resident Inspector i
L. W. Garner, Resident Inspector i
- Attended exit interview I
2.
Exit Interview The inspection scope and findings were summarized on June 13, 1986, with
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those persons indicated in paragraph I above. The inspectors described the areas inspected and discussed in detail the inspection findings. No dis-senting comments were received from the licensee. One inspector followup
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item (IFI) was identified.
I IFI 324/86-17-01 and 325/86-16-01:
Establishment of a periodic test
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requirement to perform and document control rod coupling checks prior to criticality following a refueling outage, paragraph 5.
The licensee did not identify as proprietary any of the materials provided i
to or reviewed by the inspectors during this inspection.
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Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.
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Unresolved Items Unresolved items were not identified during the inspection.
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5.
Unit 2 Post-Refueling Startup Testing (72700)
a.
Precritical Testing In preparation for Unit 2, Cycle 7 startup following refueling, the inspectors reviewed PT-50.0, Post Refueling Outage Startup Testing (a
sequential plan ~for those tests determined to be required for plant startup following a major refueling outage). The following precritical
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testing procedures were reviewed:
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Periodic Test (PT) 14.2.1, Single Rod Scram Insertion Times Test (Rev. 16)
The purpose of this PT is to measure scram insertion times for each control rod in the reactor.
Scram timing must be performed
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while the reactor vessel is at rated pressure and may be performed during the reactor vessel hydro prior to startup.
PT-14.2.1 was performed during the post-refueling Beginning of Cycle (BOC)
hydro-static test at a pressure greater - than 930 psig.
The licensee identified 15 control rods as needing scram time retesting.
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PT-50.1, Low Power Range Monitor (LPRM)/ Average Power Range Monitor (APRM) Initial Sensitivities The purpose of this PT is to set all LPRM gains so that they will read conservatively upon reactor startup. Gains on new LPRMs are set so that they will indicate 100% power for c fixed current input. The remaining LPRMs and APRMs are calibrated using data taken at the end of the previous operating cycle.
The above procedures were analyzed for embodiment of necessary test prerequisites, preparations, instructions, acceptance criteria and sufficiency of technical content.
In addition, the inspectors veri-fied that current written approved procedures were available and in use, that test equipment in use was calibrated, that test prerequisites were met, system restoration completed, and test results were adequate.
Other documents reviewed included Unit 2 Outage Log Book for the period 10-27-85 to 6-12-86, and the Brunswick Unit 2, Cycle 7, Cycle Manage-ment Report.
b.
Initial Criticality The post-refueling initial criticality was conducted in accordance with the procedure GP-02 (Revision 10), Approach to Criticality and Pressur-ization of the Reactor. During the first attempt on June 11, 1986 to take Unit 2, Cycle 7 critical in the A-sequence rod pattern, the
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licensee identified control rod 30-19 as being uncoupled. An unsuc-cessful attempt was made by the licensee to recouple the control rod.
PT-1.6.3, Rod Worth Minimizer Verification, was completed to implement 82 sequence for Unit 2 approach to criticality.
Technical Specification surveillance requirement 4.1.3.6.a, requires that each control rod shall be demonstrated to be coupled to its drive mechanism by withdrawing the control rod to the fully withdrawn posi-tion and then verifying that the control rod drive does not go to the overtravel position prior to reactor criticality after completing core alterations that could have affected the control rod drive coupling integrity.
The licensee took credit for the coupling integrity check during the single rod scram insertion times test (PT-14.2.1). However, there were no sign-off steps included in the PT for the coupling integrity check.
At the exit interview, the licensee confirmed a commitment to add steps to a PT referencing Technical Specitication 4.1.3.6 to assure and document coupling integrity checks in the future.
The concern that rod coupling was checked during this and previous outages on both units was not in question. PT-90.2, Friction Testing, has been administratively required following all outages, and OP-7, the guide =for all control rod operation, requires a coupling check whenever the rods are withdrawn to notch 48 (full out), and operator training emphasizes the need for coupling checks whenever rods are fully with-drawn. The issue is the necessity of proper documentation of a sur-
veillance required by Technical Specification 4.1.3.6.
This item will be tracked as inspector followup items 50-324/86-17-01 and 50-325/86-16-01, Establishment of a periodic test requirement to perform and document control rod coupling checks prior to criticality following a refueling outage.
Criticality was achieved in the 2B2X sequence on June 11, 1986 at 2011,
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rod 42-35 at notch 16. Technical specifications states that the core shall be subcritical by at least 0.38 percent AK/K with the highest worth control rod withdrawn, at cold xenon-free conditions, at any point in the cycle.
The BOC Shutdown Margin (SDM) Test demonstrates that this is the case by verifying the critical zero power K-effective used in the licensing analysis.
The SDM achieved at criticality was calculated to be 0.947 percent AK/K.
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No violations or deviations were identified.
6.
Unit 1 Power Distribution Monitoring (61702)
PT-01.11, Core Performance Parameter Check, accomplishes the surveillances required by Technical Specifications 4.2.1 Average Planar Linear Heat Generation Rate 4.2.2 Maximum Total Peaking Factor 4.2.3.1 Minimum Critical Power Ratio, and 4.2.4 LPRM. The inspectors reviewed nine of the procedures completed in the period January - April 1986, and found all parameters to be within specifications.
No violations or deviations were identified.
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