IR 05000269/1973010
| ML19322A870 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/28/1973 |
| From: | Jape F, Kelley W, Murphey C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19322A867 | List: |
| References | |
| 50-269-73-10, NUDOCS 7911270628 | |
| Download: ML19322A870 (13) | |
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UNITED STATES
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ATOMIC ENERGY COMMISSION o
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i DIRECTORATE OF REGULATORY OPERATICNS
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RO Inspection Report No. 50-269/73-10 Licensee: Duke Power Company Power Building 422 South Church Street Charlotte, North Carolina 28201 Facility: Oconee Unit 1 Docket No.: 50-269 License No.: DPR-38 Category: B2 Location: Seneca, South Carolina Type of License: B&W, PWR, 2568 Mw(t)
Type of Inspection: Routine, Unannounced Dates of Inspection: August 29, 1973 September 17-21, 1973 Dates of Previous Inspection: August 13-17, 1973
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Principal Inspector:
F. Jape, Reactor Inspector Facilities Test and Startup Branch Accompanying Inspectors:
W. D. Kelley, Reactor Inspector l
Engineering Section Facilities Construction Branch C. M. Campbell, Radiation Specialist Radiological and Environmental Protection Branch Principal Inspector:
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~E7'23 F. Jape, Reactor Jnspdctor Date Facilities Test and Startup Branch Reviewed By!/
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C. E.' Murphy, Chief
' Date Facilities Test and Startup Branch
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RO Rpt. No. 50-269/73-10-2-
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SUMMARY OF FINDINGS I.
Enforcement Action A.
Violations None B.
Safety Items
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None
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II.
Licensee Action on Previously Identified Enforcement Matters A.
Violations 1.
Station Modifications (R0 Report No. 50-269/73-6)
DPC's response, dated July 20, 1973, was verified by the inspector. There are no further questions or comments at this time on this item.
(Details I, paragraph 9)
2.
Audit of Operations (R0 Report No. 50-269/73-7)
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DPC's response, dated September 7,1973, has been received and is_ currently being evaluated.
B.
Safety Items None III. New Unresolved Items 73-10/1 Operability of RC-4
RC-4 would not open without bypassing the motor thermal overload circuit.
(Details I, paragraph 2)
IV.
Status of Previously Reported Unresolved Items 73-9/1 ROB 73-3, " Defective Hydraulic Shock Suppressors and Restraints" DPC's response, dated August 31, 1973, has been received and has been submitted on September 11, 1973, to RO:HQ for evaluation.
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R0 Rpt. No. 50-269/73-10-3-
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73-9/2 Administrative Procedures The licensee has provided the three administrative procedures that were identified as required to be in agreement with Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operations)." This previously identified unresolved item is closed.
(Details I, paragraph 8)
73-4/1 Reactor Coolant Pump Flow A report has been received from the licensee, dated August 23, 1973, entitled " Reactor Coolant Flow Evaluation." RO is currently reviewing the report.
71-10/1 Flow Meter Error Analysis and Tests This item is related to 73-4/1, " Reactor Coolant Pump Flow,"
above, and is also being evaluated.
71-7/1 Thin Walled Valves (R0 Report No. 50-269/71-5,Section II, paragraph 3)
The engineering analysis of 1-RV-67 (also labeled 1-RC-66)
was reviewed. The inspector asked if the analysis considered the thinnest wall sections. The licensee will determine the wall dimensions and redo the stress analysis using the thinnest section.
(Details III, paragraph 2)
V.
Unusual Occurrences A.
Boron Analysis During Zero Power Physics Tests Corrective actions described in DPC's report, dated July 17, 1973, were verified during this inspection, and RO:II has no further questions on this item.
(Details I, paragraph 3)
B.
A0-269/73-5, " Failure to Verify Operability of Reactor Building Spray Pump 1B After Maintenance" Corrective actions described in DPC's report, dated September 7, 1973, were verified during this inspection and RO:II has no further questions on this item.
(Details I, paragraph 4)
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R'0 Rpt. No. 50-269/73-10-4-
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VI. Other Significant Findings A.
RDB 72-2, " Simultaneous Action of the Safety Injection Signal on Both Units of a Dual Unit Facility" DPC's response to ROB 72-2, dated November 22, 1972, and March 30, 1973, were reviewed by the inspector. Action described in the response letters was verified and RO:II has no further questions.
(Details I, paragraph 5)
B.
ROB 72-3, "Limitorque Valve Operator Failures" DPC's response to ROB 72-3, dated January 26, 1973, was reviewed by the inspector. RO:II has no further questions on this item.
(Details I, paragraph 6)
C.
ROB 73-2, 'Halfunction of Containment Purge Supply Valve Switch" DPC's response to ROB 73-2, dated August 27, 1973, was reviewed by the inspector, and there are no further questions on this item.
(Details I, paragraph 7)
VII. Management Interview A management interview was held on September 21, 1973, with the following in attendance:
Duke Power Company (DPC)
J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent S. E. Nabow - System Project Engineer T. L. Cotton - Junior Engineer R. F. Wardell - Corporate QA Manager J. W. Cox - Assistant Plant Engineer The following items were discussed:
A.
Operability of RC-4 The inspector stated that the failure of RC-4 to open may be reportable as an unusual event. This item will be carried as an unresolved item pending further review.
(Details I, paragraph 2)
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RO Rpt. N. 50-269/73-10-5-o
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B.
Previously Identified Enforcement Matters
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The status of previously identified enforcement matters, as i
i described in Section II of the Summary of Findinge, was discussed.
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(Details I, paragraph 9)
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C.
Previously Reported Unresolved Items The status of previously identified unresolved items, as described in'Section IV of the Summary of Findings, was discussed.
(Details I, paragraph 8; Details III, paragraph 2)
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D.
Boron Analysis Errors The inspector indicated that he had verified corrective actions to prevent recurrence presented in DPC's report dated July 17, 1973.
(Details I, paragraph 3)
E.
A0-269/73-5, " Failure to Verify Operability of Reactor Building
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Spray Pump 1B After Maintenance" The inspector stated that he had verified actions to prevent a recurrence presented in DPC's report dated September 7,1973.
(Details I, paragraph 4)
F.
Regulatory Operations Bulletins
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The inspector stated that DPC's responses to ROB 's 72-2, 72-3
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and 73-2 have been reviewed and that there were no questions or j
comments on these items.
(Details I, paragraphs 5, 6 and 7)
G.
Power Escalation Tests The inspector stated that he had reviewed the test results of three test procedures and had no' major questions or comments, j
.(Details I, paragraph 10)
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Health Physics Practices During Maintenance Outage on August 14-29, 1973 The inspector stated that an audit was conducted of health physics practices during the maintenance outage on August 14-29, 1973. There were no major questions or comments.
(Details II, paragraphs 3 and 4)
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'RO Rpt. No. 50-269/73-10 I-1
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DETAILS I Prepared by:
AUb v 4ks<_,.
F." Jape,ReactorfInspector Date Facilities Test and Startup Branch Dates of Inspection: Septe,mb r 17-21, 1973 Reviewed by:/
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'Dat6 Facilities Test and Startup Branch 1.
Individuals contacted Duke Power Company (DPC)
J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent R. M. Koehler - Technical Support Engineer W. A. Brown - Chenist J. W. Cox - Assistant Plant Engineer D. J. Rains - Assistant Plant Engineer R. L. Wilson - Performance Engineer D. R. Bradshaw - Shift Supervisor R. J. Brackett - Junior Engineer
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L. E. S:hmid - Operating Engineer
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2.
Operability of RC-4 RC-4, power relief block valve, was closed to isolate RV-67, (also labeled RC-66) power relief valve, following discussions with the licensee's representative concerning the analysis of valve wall thickness for RV-67.
It could not be determined with certainty that valve wall thickness calculations for RV-67 had been made in accord-ance with applicable codes.
(This question is being carried as unresolved item 71-7/1. See Details III, paragraph 2.)
The licensee stated that the valve manufacturer would be contacted to determine the status of RV-67.
This was done and information received by telephone indicated that the original valve calculations had not considered the thinnest wall section, but a reanalysis, using the thinnest wall thickness, indicated that the valve would meet applicable codes. Upon receipt of this information, the licensee
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'IM) Rpt. No. 50-269/73-10 1-2
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attempted to reopen RC-4.
The valve would not open on initial attempts.
The licensee reported that the valve was opened when the motor thermal overload circuit was bypassed.
Since RC-4 is a reactor coolant boundry valve, its failure to operate may be reportable as an unusual event.
Further review and discussions are planned on this item.
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The item will be carried as an unresolved item 73-10/1 pending final disposition.
3.
Boron Analysis During Zero Power Physics Tests The corrective actions described in the licensee's report, dated July 17, 1973, were reviewed by the inspector. The batches of sodium hydroxide solution are being prepared in small quantities and are dated.
Each batch is retained for a two-week period and any solution remaining at the end of the period is discarded.
standard of known boron concentration is used on a regular basis A
to check the accuracy of the titrator.
twice to ensure precision.
Also each sample is analyzed The boronometer is calibrated on a periodic basis as described in PT 210/10.
Any significant difference between the boronometer indication and the lab anal cause for the discrepancy. ysia is investigated to determine the The inspector had no further questions on this item.
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4.
A0-269/73-5, " Failure to Verify Operability of Reactor Building Spray Pump 1B After Maintenance" The corrective actions described in the licensee's report, September 7,1973, were verified by the inspector.
dated A performance test was run on reactor building spray pump 1B on August 30
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Results of this test indicated that the pump was serviceable.
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i addition, the station superintendent has issued two interstation In memoranda to the maintenance supervisor, dated September 14, 1973 and September 18, 1973, which states instructions related to proper
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completion of a plant votk order and to documentation requirements of work on safety related systems.
i The inspector has no further questions on this item.
5.
ROB 72-2. " Simultaneous Action of the Safety Injection Signal on
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Both Units of a Dual Unit Facility" DPC's responses to ROB 72-2, " Simultaneous Action of the Safety Inspection Signal on Both Units of a Dual Unit Facility," dated
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RO Rpt. No. 50-269/73-10 I-3 T
w November 22, 1972, and March 30, 1973, were verified by the inspector.
The licensee has incorporated appropriate limits and precautions in OP/1/A/1107/02, " Normal Power," dated July 26, 1973, to assure transformer ratings are not exceeded.
There are no further questions on this item.
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6.
ROB 72-3, "Limitorque Valve Operator Failures"
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s The licensee's response to ROB 72-3, "Limitorque Valve Operator Failures," dated January 19, 1973, was verified by the inspector.
The torque switch torsion spring was replaced on 171 operators for Unit 1.
The replacement program was completed late in 1972.
The inspector had no further questions on this item.
7.
ROB 73-2, " Malfunction of Containment Purge Supply Valve Switch" DPC's response to RDB 73-2, " Malfunction of Containment Purge Supply Valve Switch," dated August 27, 1973, was verified by the inspector.
As stated in the report, the reactor building inlet and outlet purge valves are controlled by separate switches. These are labeled PR-1, 2, 4, 5 and 6.
The inspector had no questions regarding this item.
8.
Administrative Procedures During a previous inspection,1/ he licensee indicated that three t
additional administrative procedures would be provided in order to be in agreement with Regulatory Guide 1.33, " Quality Assurance Program requirements (Operations)." These have been prepared and were reviewed by the inspector. The findings of this review are summarized below:
a.
Shift and Relief Turnover The licensee has issued Administrative Procedure No. 15, " Shift and Relief Turnover," that specifies what must be done by an oncoming shif t prior to assuming control of the plant. This procedure also states that each oncoming shif t shall patrol the plant to verify correct equipment parameters and status.
The inspector had no further questions on this item, b.
Access to Containment Access to the containment building during reactor operation and shutdown has been covered by Administrative Procedure No.14, 1/ RO Inspection Report No. 50-269/73-9, Details I, paragraph 12.
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i R0 Rpt. No. 50-269/73-10 1-4
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" Access to Containment." The inspector had no comments or questions regarding this procedure.
c.
Recall of Standby Personnel to Plant The licensee has provided Administrative Procedure No.13, " Recall of Station Personnel," to establish a call list of plant personnel.
Personnel designated on call are required to notify the station security guard of his name and telephone number where he can be reached. The inspector had no comments or questions on this procedure.
This previously identified unresolved item, 73-9/2, is now resolved.
9.
Station Modifications The method for handling station modifications, as described in DPC's response dated July 20, 1973, in regards to the violation contained in RO Inspection Report No. 50-269/73-6, was reviewed with licensee representatives. The method is currently being revised. During the interim period, Administrative Procedure No.10 is being used along with a track sheet to ensure all reviews and appropriate approvals are received.
The inspector reviewed several modifications currently being processed. Each modification has received the reviews and approvals required by the procedure. There were no deficiencies nor comments on the handling procedure.
10. Power Escalation Tests The results of the following power escalation tests were reviewed by the inspector:
TP 800/31, " Pseudo Rod Ejection Tert" TP 800/32, " Loss of Off Site Power" TP 800/33, " Dropped Control Rod Test" The inspector had no questions or comments regarding these test results.
Power escalation testing is continuing at 95% reactor power.
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R0 Report No. 50-269/73-10 II-1
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s DETAILS II Prepared by:
I J l Radiation Specialist Q/ M. Campbell
'Datd 7/
Radiological and Environmental Protection Branch Dates of Inspection: August 29, 1973
ff Reviewed by:
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// T. Sutherland
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' Acting Chief Radiological and Environmental Protection Branch 1.
Individuals contacted Duke Power Company (DPC)
J. E. Smith - Plant Superintendent R. M. Koehler - Technical Support Supervisor C. L. Thames - Health Physics Supervisor C. T. Yongue - Assistant Health Physics Supervisor R. I4onard - Labman G. Iten - Labman J. Long - Laboratory Technician 2.
Organizational Changes None 3.
Review of Health Physics Practices During the Maintenance Outage of August 14-29, 1973.
Discussions with licensee representatives, review of selected Radiation Work Permits (RWP's), review of survey and sampling data associated with these RWP's, and review of other pertinent records showed that RWP's were being issued in accordance with established licensee procedures, appropriate surveys and sampling were done prior to and during these jobs, protective clothing requirements were appropriate to the radiological conditions identified at the work locations, and
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the required health physics records were being kept. The inspector pointed out, to a licensee representative, some areas in which the record-
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keeping could be improved and where more consistency is needed in the manner in which some records are being kept. A licensee representative agreed and informed the inspector that this would be done. Management agreed.
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RC Rpt. No. 50-269/73-10 II-2
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4.
Review of the Internal Exposure of Two Men on August 22, 1973 This occurrence was discussed with licensee representatives and the RWP and associated surveys and sample data for this job were examined by the inspector. Although this occurrence is not reportable to the Directorate of Regulatory Operations, the licensee is doing an evaluation and prepa' ring an internal eport. A licensee representative informed the inspector that their evaluation was not complete at the time of this inspection. The uptake of airborne activity by the individuals involved has been evaluated both by the licensee and by an outside contractor. The available preliminary data, from whole body counting, indicated both reasonable agreement between the two evaluations and agreement that the internal exposure was a small percentage of the applicable maximum permissable body burden. The
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inspector stated that the completed licensee evaluation would be reviewed during a subsequent inspection. Review of subsequent RWP's (and associated air sampling data) used for the completion of this job showed that once the presence of airborne activity had been detected, the licensee had taken appropriate action to preclude any additional internal exposures.
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'R0 Rpt. No. 50-269/73-10 III-1
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DETAILS III Prepared by:
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W. D. Kelley, Re' actor'Iryr5ector tate'
Engineering Section Facilities Construction Branch Dates of Inspection: September 19-21, 1973 Reviewed by:
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/ C B'yaht, Senior Inspector
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gineering Section Facilities Construction Branch 1.
Individuals Contacted Duke Power Company (DPC)
- W. H. Owen - Vice President - Design Engineering D. G. Beam - Construction Manager
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L. R. Davison - Associate Field Engineer, NDT A. R. Hollins - Associate Field Engineer, Welding i
- By telecon 2.
Thin Walled Valves The licensee has identified and measured the wall thickness of all reactor pressure boundary valves in response to RO:II letters of June 30, 1972, and February 16, 1973. The documentation of the measurements and the engineering evaluation of valves that did not r.eet the original purchase specifications are onsite.
The inspector reviewed these and has no question of these meeting the requirements of the RO letters with the exception of the pressurizer pilot operated relief valve.(1-RV-67).
The licensee had previously presented a stress analysis for this valve.
After discussion with the inspector, this analysis had been discarded and a new set of calculations was made. This second analysis had been reviewed and approved for input, results, and
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application by Babcock and Wilcox Company (B&W) on August 15, 1973.
The inspector was informed by the licensee that DPC engineers had also checked the stress analysis.
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,RO Rpt. No. 50-269/73-10 III-2
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A review, by the inspector, of the calculations and the vendor's drawing revealed that the thinnest wall sections of the valve had apparently not been used in the stress analysis.
A visual inspection of an identical valve, using fiber optics, confirmed that the valve -
was built in accordance with the drawing used by the inspector in his review of the stress analysis.
The licensee then' attempted to measure the thinnest wall section ultrasonically, using the shear wave technique, but was unable to do ao with the crystal he had onsite.
The licensee also requested B&W to supply them with the dimension of the thinnest wall section.
If the dimension for the valve's thinnest wall section is less than the dimension used in the stress analysis, the licensee will return the analysis to B&W for corrective action.
This will remain an open item.
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