IR 05000482/1984039
| ML20133G450 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/08/1985 |
| From: | Martin L, Tapia J, Whittlesey K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133G437 | List: |
| References | |
| 50-482-84-39, NUDOCS 8508080687 | |
| Download: ML20133G450 (6) | |
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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-482/84-39 Construction Permit:
CPPR-147 Docket:
50-482 Category:
C Licensee:
Kansas Gas and Electric Compan.v P. O. Box 208 Wichita, Kansas 67201 Facility Name: Wolf Creek Generating Station Inspection At:
Burlington, Kansas Inspection Conducted:
December 16-20, 1984 l
Inspectors:
ai lo/2.o/B5 J.
api a\\, React @ Inspector, Project Section B Date
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Re tor Project B' ranch 2 I
7/8/16
- . A. Whittlesey,' Reactor Inspector, Project K
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(fat 6 Section A, R6 actor Project Branch 2 l
Approved:
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//f/AS E E. Martin, C ef, Project Section A fTatd'
Reactor Pro ect Branch 2
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Inspection Summary Inspection Conducted December 16-20, 1984 (Report STN 50-482/84-39)
Areas Inspected:
Special, unannounced inspection of the containment structural integrity test and the integrated leakage rate test.
The inspection involved 64 inspector-hours onsite by two NRC inspectors.
Results: Within the areas inspected, no violations or deviations were identified.
8508080687 850726 PDR ADOCK 05000482 PDR G
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-2-DETAILS 1.
Persons Contacted Principal Licensee Employees
- H. K. Chernoff, Licensing Engineer
- K. Ellison, Startup Technical Support Supervisor
- F. R. Faist, Startup Assistant Lead System Test Supervisor
- R. L. Hoyt, Emergency Plan Administrator
- W. M. Lindsay, Quality Services Supervisor
- K. R. Peterson, Licensing Engineer
- L. Stevens, Lead Engineer
- M. A. Stewart, Manager, Nuclear Safety D. Thuet, Test Director
- J. R. Vaux, System Test Supervisor Bechtel Power Corporation D. Blackhetter, Startup Engineer R. Blum, Startup D. Chapman, Startup Engineer F. Garriel, Startup Engineer H. Hill, Startup Engineer B. Patel, Test Director R. Riggs, Senior Startup Engineer
- Denotec those persons who attended the exit meeting on December 20, 1984.
2.
Structural Integrity Test (SIT)
The SIT or initial construction pneumatic pressure test, conducted at
.115 percent of design pressure to demonstrate the capability of the containment structure to withstand specified internal loads postulated to result from the design basis accident pressure was addressed during this inspection.
Preoperational Test Procedure No. SU3-GP02, Revision 0,
" Structural Integrity Test," was reviewed and determined to satisfy NRC requircients and licensee commitments listed in the Final Safety Analysis Report (FSAR). The NRC inspectors ascertained that the test program, procedures, and instructions pertaining to the SIT had been reviewed and approved for use by authorized licensee personnel.
The FSAR describes the licensee's compliance with Regulatory Guide 1.18, Revision 1, " Structural Acceptance Test for Concrete Primary Reactor Containments." The review of the preoperational test procedure provided verification that the following attributes, considered essential to the conduct at a successful test, were correctly addressed:
a.
Placement and number of displacement transducers ensured an accurate determination of the containment's overall deflection pattern.
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The maximum test pressure was 1.15 times the design pressure of 60 psig.
c.
The specified containment pressurization and depressurization rates satisfied the guidelines specified in Bechtel Power Corporation Interoffice Memorandum No. 836, September 17, 1980.
d.
Data collection during the continuous increase in pressure conducted such that the change in pressure during data collection did not cause a change in structural response greater than 5 percent of the total anticipated deflection.
e.
The specified concrete surface locations selected for work pattern mapping represented the entire structure.
f.
The crack mapping inspections were conducted at 0, 40, 69, and again 0 psig and documented crack widths which exceeded 0.01 inches, g.
The procedure in case of unacceptable structural response was clearly delineated.
h.
Data collection points were conducted within a specified pressure tolerance.
i.
The procedure was reviewed against system operating procedures, and open items pertaining to the test were listed in the procedure.
j.
Test changes were controlled in accordance with Startup Administrative Procedure No. ADM-14-200, Revision 8, "Preoperational
Test Program Implementation."
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Test personnel requirements and responsibilities were clearly j
delineated.
The NRC inspectors performed direct observations of the data collection, instrumentation placement and ongoing crack mapping to determine technical adequacy and conformance with the requirements of the SIT preoperational test procedures.
Deflection measurement instrumentation was inspected during a tour of the containment interior.
The documentation and engineering evaluation in Startup Field Report No.1-GP-13 for modifying the anchor point location of transducers E6 and R9 were reviewed.
A total of 38 transducers were used, 18 measuring radial deflections, 8 in the vertical direction, and 12 located around the equipment hatch.
Four equipment hatch transducers exceeded their allowable value of 0.25 inches by 11, 16, 12, and 1 percent at the peak test pressure of 69.2 psig.
The NRC inspectors verified that an engineering evaluation was initiated at 60 psig when the maximum displacement approached 98 percent of the allowable.
Startup Field Report No. 1-GP-14 documents engineering approval to proceed to the maximum test pressure contingent upon a post test analysis of the area that exceeded expansion criteria.
All four
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Although allowable limits listed in the test procedure were exceeded, deflection magnitudes remained within the tolerance reserved by the design engineer.
Additional verification of tne design stress analysis as indicated by the structural response at full test pressure was obtained through exterior containment crack measurements witnessed and independently taken by the NRC inspectors at the equipment hatch.
The maximum measured containment crack width increase remained below the allowable value of 0.06 inches.
The track width and length measurements taken by the NRC inspectors were compared to the data being recorded by test personnel on the crack mapping data sheet and confirmed the accuracy of data being recorded.
As a result of this inspection, the NRC inspectors determined that the structural response of the containment at 115 percent of design pressure demonstrated that the containment structure can withstand the postulated design basis accident pressure and that structure did not sustain any structural damage.
3.
Integrated Leak Rate Test (ILRT)
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The preoperational containment ILRT conducted using the Absolute Method as described in ANSI N45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Power Plants," and ANSI /ANS-56.8-1981, " Containment System Leakage Testing Requirements," was addressed during this
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inspection.
The inspection included procedure and records review, test witnessing and independent calculations by the NRC inspectors.
This
. inspection effort was performed in order to ascertain whether the testing was conducted in accordance with approved procedures and satisfied the specified acceptance criteria contained in 10 CFR Part 50, Appendix J,
" Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors," Wolf Creek Generating Station Technical Specifications, and the FSAR.
Preoperational Test Procedures No. SU3-GP01, " Primary Reactor Containment Integrated Leakage Rate Test," and No. SU8-GP01, " Local Leakage Rate Test," incorporate all referenced requirements and criteria. These procedures were reviewed by the NRC inspectors and no discrepancies from the specified requirements and criteria were noted.
The review provided verification that the following test attributes were correctly addressed:
a.
Containment interior and exterior inspection requirements specified.
b.
Instrument locations justified by area surveys.
c.
Instrument calibration requirements specified.
d.
Instrument loss / test abort criteria delineated.
e.
Instrument error analysis performed.
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Type B and'C test results correction to Type A test results specified.
g.
Venting of internal isolated volumes required.
h.
Isolation valve closing mode specified to be the normal mode.
i.
Proper postaccident system alignment to prevent creation of artificial leakage barriers specified.
j.. _ Quality control hold points specified.
k.
Test log entries required for repairs needed to complete test.
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Acceptance criteria specified.
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Data acquisition requirements specified.
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. Data analysis technique specified.
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Maximum rate of depressurization specified.
Following the completion of the SIT, an offgasing time period was observed to allow for the decay of.any transient effects resulting from changes in temperature regimes and slow migration of air into porous surfaces and minute openings.
During this period of neutralizing the pressurization affects on the stability of the containment atmosphere, the NRC inspectors conducted a tour of the containment interior.
The selected components of test methodology observed included the distribution of instrumentation and the valve position verification of 16 randomly selected valves.
After a 4-hour temperature -stabilization period, the reduced pressure ILRT
conducted at 26 psig was witnessed by the NRC inspectors.
Containment test parameters were monitored during the conduct of the test.
The
' calculated mass point. leakage rate at a 95 percent confidence limit was
.n.008 percent per day with the acceptance criteria being 0.108 percent per day.
Test changes and test deficiencies resulting from local leak surveys
.were reviewed and found to conform with procedural requirements.
Valves isolated as a result of the local leak surveys were identified in the test
. log and are required to be subsequently tested with the result having to be included in the computation of the overall leakage rate results.
The test log also provided documentation of the inadvertent automatic start of the hydrogen mixing-fans near the end of the 24-hour test..This override of desired test status resulted from ongoing work on essential
~ lectrical bus no. NN.
The potential of.this event to. influence the e
leakage rate computation as a result of an induced temperature variation in'the stabilized containment atmosphere remained inconsequential due to its late occurrence.
Through discussions with licensee representatives, the NRC inspectors were informed that additional procedural controls were I
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The NRC inspectors witnessed portions of the verification test conducted by imposing a known leakage on the containment. ~The measured verification leak rate was 0.147 percent per day while the allowable lower and upper limits were 0.112 and 0.184.
Subsequent to the performance of the peak pressure test, the NRC inspectors obtained the new data and computed the leakage rate in accordance with the Mass, Point Data Analysis technique.
The computations performed were compared with the licensee's results for the purpose of verifying the calculational procedure and confirming the results.
This analytical technique confirmed the acceptability of the results obtained by the licensee.
4.
Exit Interview The NRC inspectors met with the licensee representatives denoted in paragraph 1 at the conclusion of the inspection.
The NRC inspectors summarized the scope and findings of the inspection.
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