IR 05000324/1979032
| ML19256E760 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/27/1979 |
| From: | Burnett P, Julian C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19256E756 | List: |
| References | |
| 50-324-79-32, 50-325-79-33, NUDOCS 7911150151 | |
| Download: ML19256E760 (5) | |
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UNITED STATES
NUCLEAR REGULATORY COMMISSION o
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i REGION ll
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101 MARIETTA sT., N.W., SUITE 3100 o
ATLANTA, GEORGIA 303o3
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Report Nos. 50-324/79-32 and 50-325/79-33 Licensee:
Carolina Power and Light Company
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411 Fayetteville Street Raleigh, North Carolina 27602 Facility Name: Brunswick Units 1 and 2 Docket 'Jos. 50-324 and 50-325 License Nos. DPR-62 and DPR-71 Inspected at Brunswick uite near Southport, North Carolina Inspected by:
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W2 7[7f C. Julianu
' Date Signed I
Approved by:_
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'P. T. Burnett,' Acting Section Chief Date Sign'ed SUmfARY Inspected on August 28-31, 1979.
Areas Inspected This routine, unannounced inspection involved 28 inspector-hours onsite in the areas of investigation of a unit 1 fuel assembly misload incident and review of post refueling testing results.
Results Of the 2 areas inspected, no items of noncompliance or deviations were identified.
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DETAILS
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1.
Persons Contacted Licensee Employees
- A. C. To11ison, Jr., Plant Manager
- J. M. Brown, Ope- ' ions and Maintenance Superinteudent
- G. T. Milligan, ce.._neering Supervisor
- D. N. Allen, Quality Assurance Supervisor
- K. E. Enzor, I&C Maintenance Supervisor
- M. A. Jones, Project Engineer, Reactor
- R. M. Poulk Jr., NRC Coordinator E. B. Wilson, Engineer D. A. Brenner, Engineer E. Eagle, Engineer R. Beverage, Quality Assurance Other licensee employees contacted included various technicians, operators, security force members, and office personnel.
Other NRC Inspectors
- J. E. Outzs, Resident Inspectcr
- B. R. Messitt
- B. T. Moon
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on August 31, 1979 with those persons indicated in Paragraph 1 above. The inspector stated that items of discussion related to the rotated fuel assembly in Unit I would be unresolved pending receipt of the operating history of the rotated bundle during cycle 2 opera'aua The inspector presented the findings on Unit 1 post refueling tests and plant status as detailed in paragraphs 6 and 7.
The licensee representa-tive committed to take action to close the open items identified.
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Licensee Action en Previous Inspection Findings Not inspected.
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4.
Unresolved Items Unresolved items are matters about which more info mation is required to determine whether they are acceptable or may involve noncompliance or deviations. A new unresolved item identified during this inspection is discussed in paragraph 5.
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5.
Unit 1 Misloaded Fuel Assembly During an entrance interview with the plant manager at the start of the inspection, the inspector was informed that the licensee had that day discovered *. hat fuel assembly IJ 0197 occupying core position 29-10 in Unit I had been placed in the core rotated 180* about its longitudinal axis from the intended orientation. The licensee had promptly reported the situation to tae NRC resident inspector, and intends to make a followup written report to Region II.
Through direct observation and discussions wi'i various licensee personnel, the inspector confirmed that prompt corrective action had been taken. The reactor power level had been reduced to 88% of rated in order to limit the rotated bundle to an apparent lilear heat generation rate (IJIGR) no greater than 9.7 kilowatts / foot and an apparent minimum critical power ratio (MCPR)
not less than 1.38.
These operating limits for the rotated bundle were adopted by the licensee at the recommend :. ion of General Electric and are i
based on their calculated results of the " worst case rotated bundle loading error" reported in the Unit I cycle 2 reload submittal document NED0-24166.
The inspector reviewed computer program OD-6 printouts on Augu. 29,1979 and August 30, 1979 monitoring thermal data in the fuel assembly at core position 29-10 and verified that the more conservative indicated limits stated above on DIGR and MCPR were not being violated.
The inspector observed that administrative action had been taken to assure that the new limits on bundle 29-10 are observed.
The rotated fuel bundle is not directly monitored by an adjacent LPRM detector. Rather, its thermal performance parameters are determined by the process computer ar.suming core symmetry. Since the cell peaking factors used by the process computer are not appropriate for the case of a rotated bundle, there is uncertainty as *
'he accuracy of the thermal parameters of this bundle. Allowing an addA snal margin on the operating limit will prevent exceeding the technical specification limits in future operation, but the reactor had been operated at full rated power during cycle 2 before discovery of the rotated bundle.
After consultation with General Electric, the licensee has determined that the 180* rotation of bundle IJ 0197 in core position 29-10 could cause the linear heat generation rate to exceed its nominal full power rate by 2.24 kw/ft. The licensee's records show that for two separate periods of opera-tion during cycle 2, the process computer generated value for LHGR on bundle IJ 0197 reached a maximum of 11.26 kw/ft. Adding on the anticipated 2.24 kw/ft due to bundle misorientation, it appears that during the two periods in question, bundle IJ 0197 operrted at a IJIGR value of 13.5 kw/ft.
This exceeds the limiting condition for operation of 13.4 kw/ft specified
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In a telephone conversation on September 21, 1979, a licensee repre-sentative stated that they feel that the 2.24 kw/ft value stated above is extremely conservative and that further analysis by General Electric is 1337 013
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in p ogress. This matter of possibly exceeding the Technical Specification limit on LHGR will remain unrisolved pending the licensee and General Electric completing their analysis of the rotated bundle. The maximum LHGR experienced by bundle LJ 0197 during Unit I cycle 2 operation will then be determinea (Unresolved Item 50-325/79-33-01).
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Fu a handling procedure FH-11 calls for a core verification to be performed after refueling using an underwater television system. The inspector reviewed the documentation of the Unit 1 fuel handling evolution and tue core verification. Three people, an engineer, a senior reactor operator, and a gt.ality assurance representative, simultaneously observed the core verification on a television monitor placed on the refueling platform and concluded that the core was correctly loaded. A video tape was made of the core verification. The licensee found the rotated bundle during a later review of the tape in search of mislcaded bundles in other areas which have produced higher LHGR values than expected during cycle 2.
The inspector viewed the bundle orientation verification tapes for Unit I and Unit 2 for the current fuel cycle. No additional core loading errors were found. The rotated bundle LJ 0197 was clearly misorierted on two Unit I core verification tapes, and probably would have been detected had the tape been reviewed by the licensee prior to beginning of cycle 2 operation.
The plant manager stated that an independent review of the core verification tapes prior to reactor restart is a corrective action under consideration.
The inspector stated that the corrective action proposed by the licensee in his eveat report will be reviewed during a future inspection ( Inspector followup item 50-325/79-33-02).
6.
Review of Unit 1 Post Refueling Testing The inspector reviewed the results of Unit 1 post refueling testing as
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follows:
PT 50.0 " Post Refueling Outage Startup Testing",
PT 50.1 "LPRM and APRM Initial Sensitivities",
PT 50.2 "SRM/IRM/APRM Overlap Determinations",
PT 50.3 "TIP Reproducibility",
PT 50.3.1 " Total TIP Uncertainty",
PT 50.4 " Process Computer New Cycle Update and Verification",
PT 50.12 " Measurement of In-Sequence Critical Data", and PT 14.3 " Shutdown Margir?
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The data generated from procedures PT 50.3 and 50.3.1 consist of brge array of nunbers which must be mathematically manipulated for evr.n:n. ion of results. This was accomplished by means of the plant process computer using a program written by the plant staff. The inspector noted that the output was not labeled so as to allow audit of the data and that no test run of the programs had been documented to verify that the programs perform the calculations correctly. The licensee had identified previously one error in the program used to evaluate PT-50.3 data. The licensee repre-sentative agreed te correct these problems and the inspector stated that 1337 014
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this item would be reviewed during a future inspection (Inspector follow-up item 50-325/79-33- 03).
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Procedure PT 50.12, " Measurement of In-Sequence Critical Data", measures the initial, post-refueling crit.:a1 control blade configuration and com-pares it to the General Electric prediction. This comparison is made for information only, and no acceptance criterion is specified. The inspector questioned whether a suitable acceptance criterion should be stated for this test, as a large disagreement between observed and predicted critical control blade configuration could be indication of insufficient shutdown margin or a core reactivity anomaly. The licensee representative agreed to evalute the need for an acceptance criteria for PT 50.12 (Inspector followup item 50-325/79-33- 04).
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Plant Status The inspector noted during a control rocm tour that both units were oper-ating with an abnormally high number of annunciator alarms locked in.
Discussitc.
tth the operators revealed that there was an assortment of alarms due to equipment cut of service, alarms due to intermittent equip-ment problems and alarms that are normally in during operation. Licensee representatives stated that a program is rader way to make design changes as required to eliminate unnecessary alarms during normal operation.
The inspector t.xpressed particlular concern about the reactor building closed cooling water (RBCCW) and service water discharge radiation monitors having a common annunciator. On August 29, 1979 this annunciatcr was lit nu Unit 2 due to a high reading on the RBCCW radiation monitor channel. A licensee representative stated that this was due to a high background reading in the area of the RBCCW detector and that this condition had existed for about a month. Under these conditions, a legitimate high count
. ate alarm on the service water discharge channel would not annunciate on the control room board to alert the operator to a potentiial release of radioactivity from the plant.
The plant manager stated that this condition would be given priority for correction (Inspector followup item s50-324/79-32-01).
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Review of Unit 2 Post Refueling Testing The results of the following post refueling tests performed on Unit 2 were reviewed by the inspector.
PT 50.G, " Post Refueling Octage Startup Testing" PT 50.1, "LPRM and APRM Initial Sensitivities" PT 14.3, " Shutdown Margin" PT 50.2, "SRM/IRM/APRM Overlap Determinations" No items of noncompliance were identified in this review.
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