IR 05000270/1973007
| ML19317D336 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/26/1973 |
| From: | Epps T, Jape F, Murphy C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19317D333 | List: |
| References | |
| 50-270-73-07, 50-270-73-7, NUDOCS 7911270559 | |
| Download: ML19317D336 (12) | |
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Q,'t, UNITED STATES t
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ATOMIC ENERGY COMMISSION I f ', [e'
DIFICTCRATE OF REGULATCRY CPERATICN3
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RO Inspection Report No. 50-2,0/73-7 Licensee: Duke Power Company Power Building 422 South Church Street Charlotte, North Carolina 28201
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i Facility Name: Oconee Unit 2 l
Docket No.:
50-270 i
License No.:
CPPR-34 Category:
B1/A3
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Location:
Seneca, South Carolina
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Type of License:
I Type of Inspection:
Special, Announced
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Dates of Inspection: June 10-17, 19-21, 1973 i
Dates of Previous Inspection: May 14-17,1973 Principal Inspector:
F. Jape, Reactor Inspector Facilities Test and Startup Branch e
Accompanying Inspector:
T. N. Epps, Reactor Inspector Facilities Test and Startup Branch (June 16&l7, 1973)
Other Accompanying Personnel:
M. D. Fairtile (June 14, 1973)
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'i E. Draney (June 14, 1973)
h Principal Inspector:
M.O*M_i p:2<._,
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F. Jape, Reactor Inspector, Facilities Test Date an Startup Branch
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/ /bt, J Reviewed by:-
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Date C.E. Murphy,, Chief,FacilitiesTestandStartupBranch I,
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RO Rpt. No. 50-270/73-7-2-SUMMARY OF FINDINGS I.
Enforcement Action None II.
Licensee Action on Previously Identified Enforcement Matters A.
Violations The folleving item remains open:
Welding Program Deficiencies (RO:II Letter to DPC, dated March 8, 1972, Item 5)
The DPC consultant's report on welding deficiencies and documentation in the DPC welding program will be reviewed when available to RO:II.
B.
Safety Items l
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f-There are no previously identified safety items.
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III. New Unresolved Items
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None
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IV.
S'tatus of Previous 1v Reported Unresolved Items 72-8/l Receiving Inspection Documentation An audit of recently completed QC-31 forms revealed that j
the receiving inspection documentation contains the
information to satisfy the intent of the procedure.
This I
item is resolved.
(Details I, paragraph 4 )
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72-8/2 QA Documentation i
The licensee has revised the filing of QA documentation
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such that records are retrievable.
This item is resolved.
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73-2/1 Reactor Vessel Core Support Records The reactor vessel core support records are available at the Oconee Nuclear Station.
These records were reviewed
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by the inspector.
This item is resolved.
(Details I,
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paragraph 6)
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RO~Rpt. No. 50-270/73-7-3-
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i 73-1/1 Core Flooding System Testing Recuirements The results of TP 201/7 " Core Flooding System Flow Test,"
remain to be reviewed.
This item remains open.
73-6/1 Test Sequence for the Reactor Building Structural Integrity Test and the Integrated Leak Rate Test The licensee's test sequence differs from the guidance provided by R0 Headquarters.
This item remains open.
(Details I, paragraph 7)
V.
Design Changes None I
VI.
Unusual Occurrences None
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VII. Other Significant Findings
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None
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VIII. Management Interview A management interview was held on June 17, 1973, with J. E. Smith, Plant Superintendent at the conclusion of the inspection and by telephone with D. G. Beam, Project Manager, on June 20, 1973.
The following items were discussed:
A.
TP 150/2, " Reactor Buildine Structural Integrity Test" and TP 150/3, " Reactor Buildine Integrated Leak Rate Test" The inspector stated that he had reviewed the test procedures
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and prerequisite tests relevant to TP 150/2 and TP 150/3.
Comments were discussed with licensee personnel and all were resolved.
The inspector commented that he had witnessed the test work completed thus f ar and was aware of the dif ficulties i
encountered in performing the test.
(Details I, paragraphs 2 and 7 and Details II, paragraph 2)
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B.
Emergency Proce:dures All of the emergency proceder*.s C., for Unit 2 have been reviewed by the inspector.
Tae inspector stated that he had no comments or questions on the EP's that have been
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R0 Rpt. No. 50-270/73-7-4-
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I issueo, but that the three EP's identified during the previous inspection 1/ remain outstanding. 'The licensee i
representative concurred.
(Details I, paragraph 3)
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Followup on Outstanding Items The inspector stated that he had reviewed three outstanding items and had no further questions or comments on them.
(Details I, paragraphs 4, 5 & 6)
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1/ RO Inspection Report 50-270/73-6, Details, paragraph 11.
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RO Rpt. No. 50-270/73-7 I-l (~J h
s DETAILS I Prepared by:
A<'4tL
d ' M 'b F. Jape, Reactor'
Date Inspector, Facilities Test and Startup Branch f
Dates of Inspection: Jun,e 10-17, 1973
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Reviewed by,:-
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C'. E. Murphy, Chie'f Date '
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Facilities Test and
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Startup Branch i
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1.
Individuals Contacted Duke Power Company (Steam Production Department)
J. E. Smith - Plant Superintendent R. C. Collins - Unit 2 Performance Engineer
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M. Burgess - Junior Engineer D. J. Rains - Assistant Plant Engineer i
R. M. Koehler - Technical Support Engineer
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J. W. Hampton - Assistant Plant Superintendent H. B. Barron - Junior Engineer
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M. D. McIntosh - Operating Engineer (Construction)
D. G. Beam - Project Manager (by phone)
D. L. Freeze - Principal Field Engineer
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Wiss, Janney, Elstner, and Associates
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R. Krause - Engineer
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Bechtel Power Corporation B. F. Digiongio - Senior Engineer G. Cranston - Engineer D. Dundas - Engineer M. Moravan - Engineer J. Lysak - Engineer S. Elliott - Engineer
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R0 Rpt. No. 50-270/73-7 I-2 P
I 2.
Review of Test Procedures for the Reactor Building Structural Integrity t
Test and the Containment Leak Rate Test
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The following test procedures were reviewed and comments discussed
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with the licensee's representative:
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(a; P 150/2 Reactor Building Structural Integrity Test Inspector's Comment - Section 5.6.1.2 of the FSAR states i
that the test pressure will be held for at least an hour.
This requirement is not specifically stated in the test procedure.
Licensee's Response - The one hour requirement is stated in TP 150/3 "R3 Integrated Leak Rate Test," and the two tests are run concurrently.
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(b) TP 150/3 Reactor Building Integrated Leak Rate Test
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Inspector's Comments
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There is no indication that all of the test equipment will be calibrated prior to use.
Specifically the
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flowmeter and pressure gauges listed in Section 5.0 do-not have a prerequisite calibration test listed.
2.
Appendix J of 10 CFR 50.contains a statement regarding closure of containment isolation valves when performing a leakrate test.
This precaution or limitation is not covered by the test procedure.
Ecw will isolation valves be treated during perfor=ance of the leakrate test?
3.
In Section 7.0 " Required Unit Status," There is no mention of the status of the N, system for the core flooding tanks. Will this system be isolated?
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4.
What is intended by the item 7.13 which states in part,
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. visually inspected for seal tightness."?
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5.
Item 8.2 refers to the location of the RTD's as being the same cs for Unit 1.
Where is this information and will it be a part of the test records for Unit 2?
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6.
Section 5.6.1.2.2 of the FSAR states that R3 air temperature will be maintained between 60 and 100F during the test.
This limitation is not contained in the test pro cedure.
Step 12.25 states that electric heaters will be used to satisfy temperature requirements.
What are the temperature requirements?
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I RO Rpt. No. 50-270/73-7 I-3 h
Licensee's Response 1.
A statement will be added regarding the accuracy and calibration requirements of special test equipment.
2.
A precautionary statement regarding closure of isola-tion valves idll be added to the test procedure.
3.
The isolation of the core flooding system N2 supply line is covered in Enclosure 13.2 " Valve Checklist."
4.
This will be reviewed and clarified.
5.
A drawing showing the RID locations will be included in the test data package.
6.
The allowable te=perature range as given in the FSAR will be added to the test procedure.
(c) TP 150/11 Reactor Buildine Liner Plate Surveillance Program No comment.
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(d) TP 150/12 Concrete Crack Surveillance Test
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Inspector's Comment - Step 12.1.3 states that only typical cracks will be recorded.
In Section 5.6.1.2.2 of the FSAR it is stated that cracks <0.01 inch in width will be mapped.
Are these two statements consistent?
Licensee's Resconse - This step in the procedure has been revised to require capping of cracks consistent with the FSAR.
(e) TP 150/13 Prestressing Tendon Anchor Zone Surveillance Program No comment.
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3.
Review of Emergenev Procedures (EP)
The inspector reviewed the following EP's and had no major comment or question regarding then:
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EP 1800/1 Load Rejection l
EP 1800/2 Turbine Trip EP 1800/3 Reactor Trip EP 1800/4 Loss of Reactor Coolant EP 1800/5 Boron Dilution
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EP 1800/6 Loss of Reactor Coolant Flow
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R0 Rpt. No. 50-270/73-7 I-4
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EP 1800/7 Condenser Circulating Water System Failure EP 1800/8 Steam Supply System Rupture
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EP 1800/9 Earthquake i
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EP 1800/11 Loss of Low Pressure Injection System (Decay Heat Removal)
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EP 1800/12 Loss of Control Room EP 1800/13 Spent Fuel Damage EP 1800/16 Loss of Power EP 1800/17 Steam Generator Tube Rupture EP 1800/18 Emergency Procedure EP 1800/19 Loss of CCW Intake Canal EP 1800/20 Loss of Normal EPI Makeup or Letdown EP 1800/21 Inoperable Control Rods The review of EP's is complete except for the question regarding the lack of a procedure for:
i (a) loss of instrument air system, (b) malfunction of automatic reactivity control system, and (c) malfunction of pressure control system.
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The lack of these three EP's was discussed in a previous report-and will be reviewed in a subsequent inspection.
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4.
Receiving Inspections (Units 2 & 3)
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A sample of twelve recently completed QC-31, " Report of Receiving Inspection," forms was During a previous RO inspection 27eviewed by the inspector.
the inspector noted that the entry made under the " Extent of Inspection Made" was " visual" for most QC 31's.
The licensee representative indicated that = ore information would be included on the QC-31 form for documentation purposes.
The twelve QC-31 forms reviewed during this current inspection were found to contain more detail regarding the " Extent of
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Inspection Made." The inspector had no further comments or questions on this item.
1/ RO Inspection Report 50-270/73-6, Details, paragraph 11.
2,/ See RO Inspection Report 50-270/72-8, Details II, paragraph 2.
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R0 Rpt. No. 50-270/73-7 1-5
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I The QC-31 forms for the main steam relief valves, the i
letdown storage tank, and the B-1 reactor coolant pump
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were reviewed by the inspector.
These requestedduringapreviousinspectiongArecordswere i
but were not available at that time.
The inspector had no questions
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or comments on these records.
5.
Filing of QA Records (Units 2 & 3)
The files for the QA records have been rearranged to improve retrievability of these records.
For example, QC-31 forms are filed by system.
Thus if the QA records for a particular piece of equipment is desired, the file for the system can easily be pulled and the record for the particular piece of equipment can be located. This change is g result of discussions held earlier with the licensee management.2j
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6.
Reactor Vessel Core Support Records The inspection and vendor certification records for material and fabrication for the reactor vessel upper and lower core
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supports yere reviewed.
These records were previously
reported / as not available at the site.
The records describe
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the revised design of the reactor laternals and the repair
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procedures.
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During hot functional testing of the Oconee 1 Nuclear Steam System, several reactor internal components failed.
To correct these failures, the B&W company revised the design -
of the reactor internals.
Hot functional testing of the Oconee 1 Nuclear Steam System, with the revised internals, has been successfully completed. A similar testing program for the Oconee 2 Nuclear Steam System will be completed prior to initial fuel loading.
7.
TP 150/2 Reactor Building Structural Integrity Tent and TP 150/3 Reactor Building Integrated Leak Rate Test The inspector witnessed the initial part of the structural integrity test and the containment leak rate test.
Approved procedures were being used to conduct these tests and the five prerequisite periodic tests, (PT) and the six instrument procedures, (IP), were reviewed by the inspector.
(Results of the review of the 5 PT's are reported in Details II, paragraph 2.)
The inspector examined the following test equipment prior to the beginning of pressurization of the reactor building:
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1/ See RO Inspection Report 50-270/72-8, Details II, paragraph 2.
2/ See RO Inspection Report 50-270/72-8, Details II, paragraph 7.
3/ See RO Inspection Report 50-270/73-2, Details II, paragraph 2.
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RO Rpt. No. 50-270/73-7 I-6 a.
RTD's
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Precision Pressure Gauge c.
Barometer d.
Hygrometer e.
Flewmeter The inspector had no comment or question regarding the test equipment.
Pressurization of the reactor building began at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on June 14,19 73.
Data were recorded as prescribed in the test procedures. At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> on June 15, 1973, a pressure of approximately 30 psig was reached and a period of stabilization was begun.
Stabilization was not achieved and a change in the test sequence was initiated. The revised sequence was to comp.lete the structural test, then reduce pressure to the design accident pressure and complete a leak rate test followed by the reduced (surveillance) test pressure leak rate test.
The inspector informed the licensee that this sequence <as in
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violation of Appendix J of 10 CFR 50, and it was recommended that the licensee contact the Oconee Project Leader in AEC's Licensing Branch.
This wts done and the various testing alternatives were reviewed.
Final plans were not completed by the conclusion of this portion of the inspection on June 17, 1973.
Following discussions with Licensing, the test program was continued. A structural integrity test pressure of 67.9 psig was reached at 10:10 am hrs. on June 18, 1973, and held until 11:30 am on June 18, 1973.
Test pressure was reduced to 59.6
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psig at 1:08 pm on June 18, 1973, and held for 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br /> and 50 minutes.
During this period, stabilization was not achieved.
The licensee
.then decided to reduce pressure to enter the building for inspection
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and repair of leaks. Depressurization began at 4:58 am on June 21, 1973, and the building was at atmospheric pressure at 9:30 as on
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June 22, 1973.
l Plans are to reschedule the containment leakrate test at a
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later date.
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j RO Rpt. No. 50-270/73-7 II-1
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DETAILS II Prepared by: >
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T. N. Epps, Red'et~or Date
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Inspector,A$cilities e
Test and Startup Branch Dates of Inspection:
June 16-17, 1973 Reviewed by:_.
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Y C. E. Mudphy, Chief Dat e'
Facilities Test and Startup Branch 1.
Individuals Contacted Duke Power Company (DPC)
J. E. Smith - Plant Superintendent R. M. Koehler - Technical Support Enginee r L. E. Schmid - Assistant Operating Engin;er R. C. Collins - Performance Engineer
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D. J. Rains - Assistant Plant Engineer 2.
Containment Leak Rate Test Prerequisites
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Prerequisites The inspector reviewed five test results packages for tests that were prerequisites for the Unit 2 containment leak rate test. The results reviewed were for the following tests:
Electrical Penetration 0-ring Seal Leak Test (PT-150-5)
Personnel and Emergency Lock Leak Test (PT-150-8)
Equipment Hatch Leak
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Rate Test (PI-150-4)
Reactor Building Inspection Test (PI-150-16)
Reactor Building Pneumatic Leak Test (PI-150-6)
Results were found to be within specified limits, b.
Conduct of Test Pressurization of the Unit 2 containment building was
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started at 2:00 P.M. on June 14, 1973.
30 psig was
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R0 Rpt. No. 50-270/73-7 II-2 O-t l
attained at 4:00 P.M. on June 15, 1973.
The leak rate
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at this pressure was determined to be approximately
.2% per day at 29.5 psig. An investigation was conducted to determine the cause of the excessive leak rate (maximum
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allowable =.125% per day at 29.5 psig).
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DPC suspected that the reactor building coolers were
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isolate the coolers for the integrated leak rate test.
Justification was that these coolers are either isolated
or operating at a pressure equivalent to reactor building i
design pressure, i
The sequence of the test was also changed such that the structural test would be conducted before the 29.5 psig
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test.
DPC was informed that appendix J of 10 CFR 50 requires that the low pressure test (29.5 psig) be conducted prior to the design pressure (59 psig) test.
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DPC now plans to complete the structural test on the unit (
2 containment before conducting the leak rate tests. Air compressor problems slowed the rate of pressurization of s
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the containment for these tests.
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