IR 05000425/1988065
| ML20196E372 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/30/1988 |
| From: | Jape J, Whitener H, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20196E360 | List: |
| References | |
| 50-425-88-65, IEB-84-03, IEB-84-3, IEIN-88-073, IEIN-88-73, NUDOCS 8812090327 | |
| Download: ML20196E372 (8) | |
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b UNITED STATES j,
,j NUCLEAR REGULATORY COMMISSION
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REGION 11 o,
101 NARIETTA ST., N.W.
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ATLANTA. GEORGIA 30323 Report No.: 50-425/88-65
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Licensee: Georgia Power Company P. O. Box 1295 Birminghcm, AL 35201
$t No.: 50-425 Licensa No.: CPPR-109 3-ity Name: Vogtle 2
.5 tion Conducted:
Oc tober 24 28, 1988 YW i spectors'
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Date Signed
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Date Signed Accompanying Personnel:
S. Shaeffer bG,#w
/ //3 C#/cf Approved by:
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Datd Signed F. Jape, Chief Test Programs Section Engineering Branch Division of Reactor Safety
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SUMMARY i
Scope:
This routine, unannounced inspection was conducted in the areas
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of containment local leak rate preoperational test program review,
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reactor protection system preoserational test procedure review and test witnessing, preoperationa'
test results evaluation, and review of licensee action on previous inspection findings.
Results: The licensee's preoperational local leak rate testing program was adequate and met the testing reoutrements of 10 CFR 50, Appendix J and the Final Safety Analysis Report.
Program effectiveness was evidenced by the low sumation of total leakage for Type B and Type C leak rate tests.
However, minor weaknesses in the local leak rate test procedure were discussed with the If eensee concerning the possible nonconservative leak testing of containment p'
,_ valves (see paragraph 2.a.1) and inadequate containment penetration vent and drain instructions (see paragraph 2.a.5).
Further, a concern was identified involving the nonconservative alignment of the Post Accident Hydrogen Monitor System for the upcoming integrated leak rate test.
In the areas inspected, violations or deviations were not identified.
8912090327 001201 PDR ADOCK 050004Ds O
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REPORT DETAILS 1.
Persons Contacted Licensee Employees T. Forehand, lead Test Supervisor
- E. D. Groover, Quality Assurance (QA) Site Manager - Construction
- J. C. Guimbellot, Nuclear Operations lead Test Supervisor
- S. M. Hall, Procedures Superintendent
- H. N. Handfinger, Project Startup Manager
- C. W. Hayes, Vogtle Quality Assurance Manager
- G. A. Johnson, Test Supervisor
- R. H. Pinson, Vice President, Project Construction
- K. Pointer, Senior Plant Engineer Other licensee employees contacted during this inspection included craftsmen, engineers, operators, and administrative personnel.
Other Organizations Speer Consultants J. Phelps, Work Planning Supervisor R. L. Scott, Test Supervisor TER Services T. E. Renton, Leak Rate Consultant NRC Resident Inspector
- C. Burger, Resident Inspector 2.
Preoperational Test Procedures Review (61720, 70300, 70305)
a.
2-300-04, Containment Local Leak Rate Testing The inspectors reviewed the preoperational leak rate test program in detail to verify that procedures are developed which include the requirements of Appendix J to 10 CFR 50, ANSI N45.4, the proposed Technical Specifications and the applicable sections of the Final Safety Analysis Report (FSAR), Chapter 14.2.
Procedure 2-300-04 includes both Type 8 and Type C local leak rate tests.
This proce-dure was reviewed for approvals; assignment of responsibilities; inclusion of all leakage barriers described in FSAR Table 6.2.4-1;
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adequate test instructions, prerequisites and cautions; control of
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i test activity; appropriate test parameters; epproved test methods; adequate acceptance criteria; and, summation of total leakage.
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detailed walk through of all containment penetrations to be leak rate i
tested was performed to verify proper venting, draining, valve f
identification, valve alignment and system restoration.
Some concerns identified during review of this procedure are identified and discussed below along with the resolution or current status.
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i (1)
Purge Valves
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IE Notice 88-73 dated September 8,1988, identifies a possible i
leakage problem related to the direction of pressurization for the Fisher Series 9200 butterfly valves.
The licensee is using
Fisher 9280 butterfly valves in the containment purge lines.
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The licensee will verify the leak tight integrity of the inboard
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purge valves by installing a blind flange and performing the
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local leak rate test with the valve pressurized in the accident i
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direction. This will confirm the capability of the purge valves
to seal against accident pressure.
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The inspectors concluded that the purge valve issue is resolved
relative to the preoperational test.
Orientation of the
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butterfly valves in the purge lines is being reviewed by the
licensee relative to the method of performing a conservative leak rate test during plant operation.
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Post Accident Hydrogen Monitor l
i From a review of the Post Accident Hydrogen monitoring system,
the inspectors found that both the inboard and outboard isola-tion valves on the supply and return lines were properly Type C
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tested.
It was also determined that the licensee intends to l
close these valves during the integrated leak rate (Type A)
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This means that the portion of this system located i
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outside contair, ment between the supply line outboard isolation (
valve and the return line outboard isolation valve will not be
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tested for leakage.
Since the system may be pressurized post I
accident to detemine the hydrogen content of the containment
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atmosphere, it is the NRC position that the isolation valves be j
open and the system pressurized during the Type A test. Licensee j
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leak rate personnel agreed to review this matter.
This matter i
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will be followed up in the upcoming inspection of the integrated
leak rate test.
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l (3) Liner Weld Lesk Chase Channels The inspectors discussed the need to vent the containment liner l
weld leak chase channels to containment pressure during the Type
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A test.
The licensee stated that the vent valve 3 have been i
idertified and the channels will be vented during the Type A
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test. This matter is resolved.
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(4) Test, Vent and Drain Valves During the penetration reviews, the inspectors noticed that a number of nominal.75 inch test, vent or drain valves located
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between the inboard / outboard isolation valves were included in the FSAR Table 6.2.4-1 as requiring a Type C test.
Twenty-one of these valves were identified.
Eight of the 21 have been Type C tested due to the way the main isolation valves were L
tested.
Thirteen valves have not been Type C tested.
The licensee believes these valves were inadvertently included in the FSAR Table and intends to submit an amendment to the FSAR to eliminate the requirement for a Type C test.
This matter will be tracked pending NRC review and acceptance of the licensee's submittal to eliminate the Type C test requirement fi-these valves as Inspector Followup Item 425/88-65-01.
(5)
Penetration Draining The inspector found that a step by step instruction for draining
the penetrations prior to the local leak rate test was not included in the procedural instructions.
However, the procedure does require that each leak rate test be performed in sequence and requires sign off of each step. Step 6.1.2 specifies that the test initial boundary conditions are established.
Step 6.1.3 requires verification that the system is drained within the test
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boundary.
This matter will be reviewed further at a future inspection.
The licensee's program to identify conditions which could affect the leakage test and to maintain control of leakage barriers which
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have been successfully tested was reviewed. Review of the construc-tion acceptance tests, system deficiency punch lists, drawing change notices and current plant drawings prior to the leak rate test is required by the procedure.
This document review is performed to identify any system or component deficiency which can invalidate the test result.
Also any unsatisfactory leak rate test is identi-fied and tracked through a Test Exception Report (TER), Operating Deficiency Report (00R) and Maintenance Work Order (MWO).
In addition to these documents, the test data sheets for failed tests are retained in the test procedure and marked "Invalid" along with the data sheets for the successful test marked as "Official Retest".
The inspectors reviewed examples of the abnve documents and tracked two examples of
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valve repair and retest through the work planning system.
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inspectors concluded that the licensee's controls to establish and maintain the integrity of leak tight barriers are functional and
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effective.
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b.
2-3SB-01 - Reactor Protection Preop 2-3SB-02 - Reactor Protection Safeguards Test The above procedures were reviewed to verify that the technical content in the procedures was consistent with applicable sections of the FSAR Chapter 14.2 and Regulatory Guide 1.68, Revision ?
requirements.
The procedure had proper management review and approval and the test objectives were clearly stated. The procedures included references to FSAR sections, drawings, vendor data, speci-fications, and administrative controls of test performance, review, and approval.
Test equipment was specified along with equipment calibration controls.
Notes and precautions were provided to alert test personnel of important measures to br taken to protect equipment and systems and to avoid abnormal situations.
Prerequisites and initial test conditions were specified, including initial equipment and system lineup, electrical power and control requirements, and test performance approvals.
Step by step instructions appeared to be complete to the extent necessary to assure the test objectives were met.
Provisions were available for the verification and sign-off of test items and for recording details of the conduct of the tests.
Acceptance criteria were identified and provisions for test results revfew and approval were identified.
Within the areas inspected, no violations or deviations were identified.
3.
Preoperational and Local Leak Rate Test Witnessing (61720, 70317)
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Local Leak Rate Test Witnessing - 2-300-04, Containment Local Leak Rate Testing The inspectors witnessed the licensee's attempt to perform the Type C leak rate test for the outboard isolation valve, HV-27901 found in containment penetration No. 40, Fire Protection.
This valve had failed the initial local leak rate test and was being ratested after repairs.
The inspectors verified the documentation of ;he initial test results and reviewed the procedure TER and ODR which described the valve failure.
The inspectors verified that the test personnel were using the latest revision of the test procedure and containment penetration drawings.
The inspectors also verified that appropriate prerequisites were met, that test equipment was calibrated as required, and that valves were positioned as required by the test procedure.
Testing was interrupted when the licensee observed that water continued to leak from a test drain connection located in the test boundary.
The test supervisor took appropriate action to suspend testing and write a maintenance work request to investigate the likelihood that block valves, located upsteam, and serving as the outer test boundary, were leaking water back to the test connection.
The inspectors also reviewed test data sheets of all Type B and C leak tests and a preliminary summary sheet of measured leakages.
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The data sheets indicated that leakage is well below conservative acceptance guidelines and as a result, the summation of leakage is below the 0.6 La regulatory limit.
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Other Preoperational Tests Witnessed-The inspectors witnessed portions of the p:rformance of preopera-tional tests 2-3S8-01, Reactor Protection Preop and 2-300-02, Reactor
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Trip System and ESFAS Process Channel and Logic Response Time Test.
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The portions of the tests witnessed by the inspector included the verification of the following:
Appropriate revisions of the test procedures were available and
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in use by the test personnel
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Test prerequisites were met
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Tests were perfonned as required by the approved procedures
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Criteria for interruption of testing and continuation of an
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interrupted test were adhered to Significant events, unusual conditions, test discrepancies, and
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interruptions to testing were documented.
The inspectors witnessed portions of Subsection 6.8.3 of preopera-tional test 2-3SB-01 which demonstrate the operation of the pressurizer low pressure safety injection and reactor trip logic.
The specific steps witnessed were 6.8.3.7 through 6.8.3.19.11.
These steps verified the capability of the feedwater isolation valves safety injection logic, main feedwater control valve safety injection, and main feedwater bypass valves safety injection logic to perform their intenced function.
Testing was interrupted when a logic problem was discovered in the main feedwater isolation bypass valve logic.
The test supervisor suspended testing in order to investigate and review the associated system logic drawings.
The inspectors also witnessed portions of Subsection 6.9.3, specif-ically steps 6.9.3.68 through 6.9.3.356. These steps verified the operation of the steam generator high-high level safety features logic.
This logic verified the proper response of the steam generator high-high level alert alarms; the high-high level status light indications; and the high-high level turbine trip alarms. The licensee encountered no problems during testing.
The inspectors witnessed portions of Section 6.11.2 of preoperational teat 2-300-02 which tests the 4.16 KV emergency bus undervoltage safety feature sensor and logic response time.
The specific steps witnessed verified the proper sensor and logic response time for the A Train of the Reactor Protection System.
No problems were
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encountered during testing and preliminary analysis of the results
indicated that response times were within acceptance criteria.
Within the areas inspected, no violations or deviations were identified.
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4.
Preoperational Test Results Evaluation (70400)
The inspectors reviewed the following completed preoperational test procedures:
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2-3EG-01, Component Cooling Water (CCW) System i
2-3SF-02, Control Rod Drive Mechanism Motor Generator Set Preop The above two test procedures were reviewed to verify that:
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Test changes were approved in accordance with administrative procedures
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i Test changes did not affect the basis objective of the tests
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Test steps and data sheets were initialed and dated as ' required
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Test data were within acceptance criteria specified
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Test deficiencies had been resolved including retesting where
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required
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Management had evaluated the test results as required by
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administrative controls Within the areas inspected, no violations or deviations were identified.
5.
Followup on IEB 84-03, Refueling Cavity Water Seal (92703)
(Closed) By letters dated February 12 and March 7,1985, the licensee provided a summary of an evaluation of the potential and consequences of a refueling cavity water seal failure for the Vogtle plant. The licensee
considered the design differences in the permanently installed refueling
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cavity seal used at Vogtle and the cavity seal described in IEB 84-03,
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and concluded that a catastrophic failure of the cavity seal is not a credible event at Vogtle.
I The cavity seal design at Vogtle is not the typical temporarily installed
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reinforced steel rino with the double 0-ring inner and outer concentric inflatable rubber bladder used by many plants such as that of the Haddam Neck Nuclear Plant.
Haddam Neck experienced a gross seal failure on August 21, 1984. Vogtle's design is a permanently installed seal assembly of the passive mechanical type and consists of an angle plate welded to
the cavity liner plate on the bottom and the reactor vessel seal ledge on the top.
The two plates form a seam which serves to provide a bellows function to correct for normal reactor vessel expansion and contraction r
movement.
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The licensee contends that any leckage of the seal resulting from undetected failure of the angle plate or welds would be directed to the reactor cavity sump.
The cavity sump is monitored by redundant nonsafety grade level switches, which actuate the cavity sump pumps, and a separate safety grade level transmitter.
Seal leakage in excess of the sump pumps
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capacity of 100 gpm would be alarmed in the control room when the sump reaches the high level.
The inspectors also reviewed the refueling cavity makeup capacity in the event of a postulated loss-of-cavity-water accident. The refueling canal is normally filled by using an Residual Heat Removal (RHR) pump, which is aligned to the Refueling Water Storage Tank (RWST), and has a minimum capacity of 3000 gpm.
Therefore, makeup capacity would exceed any credible leakage during such an accident.
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6.
Exit Interview The inspection scope and results were summarized on October 28, 1988,
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j with those persons indicated in paragraph 1.
The inspectors described
the areas inspected and discussed in detail the inspection results listed
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below.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
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IFI 425/88-65-01 Review amendment to the Vogtle FSAR to eliminate small vent and drain valves from the Type C leak rate test.