IR 05000424/1986099
| ML20214K498 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/17/1986 |
| From: | Burnett P, Jape F, Long A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214K487 | List: |
| References | |
| 50-424-86-99, NUDOCS 8612020324 | |
| Download: ML20214K498 (6) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISslON
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REGION 11
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Report No.:
50-424/86-99 Licensee:
Georgia Power Company P. O. Box'4545
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Atlanta, GA 30302 I
, Docket No.:
50-424 License No.:
CPPR-108 Facility Name:
Vogtle 1 Inspection-Conducted: Oct er 14-17, 1986 Inspectors:
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Date Signed
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Dats Signed Approved y:
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E. Jape, ChMf, Test Progralut Se(tion Date~ Signed Engineering Branch I
Division of Reactor Safety SUMMARY LScope: This routine, unannounced inspection was conducted in the areas of review of startup testing procedures.
Results:
No violations or deviations were identified.
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8612O20324 861119 PDR ADOCK 05000424 G
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REPORT DETAILS 11.
Persons Contacted-Licensee Employees.
- C. Belflower, Quality Assu'rance (QA) Site Manager, Operations
- R.' M.:Bellamy,. Plant Support Manager
- W. L..Burmeister, Operations Supervisor
- R. J. Florian,~ Reactor Engineering Supervisor
- G. R. Frederick, Engineering Support Group Supervisor
- W. F. Kitchens, Manager of Operations H. P. Walker, Manager of Unit Operations -
NRC Resident Inspectors'
H. H. Livermore, Senior Resident Inspector, Construction J. F. Rogge, Senior Resident Inspector, Operations
- R.'J. Schepens, Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and f.indings were summarized on October 17, 1986, with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection findings, including the new followup items listed below.
No dissenting comments were received from the licensee.
Unresolved Item (UNR) 424/86-99-01:- Natural Circulation Test is Scheduled to be Performed After Operation at 75% Power, rather than
.before Power Ascension (Paragraph 5).
Inspector Followup Item (IFI) 424/86-99-02:
Sample and Control Refueling Canal Boron Concentration (Paragraph 5).
Proprietary material was reviewed during this inspection, but is not included in this report.
3.
Licensee Action on Previous Enforcement Matters
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Thissubjectwasnotaddressedintheinspection.
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Unresolved Items One unresolved item was identified during this inspection paragraph 5.
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.ReviewoflStartupTestProcedures.(72400,72500,73566,73570,73572).
The: following startup test procedures were reviewed for conformance to the Chapter 14 requirements of the FSAR, appropriate Regulatory Guides, and-
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current; industry practice:
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1-500-01~(Revision 1), Initial Fuel Load Test Sequence
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_1-500-02 (Revision-2), Inverse Count Rate Ratio Monitoring for ~ Core
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1-500-03 (Revision 1), Determination. of Core Power Range for-Physics Testing d.
1-500-04 (Revision 1), Precritical Test Sequence e.
1-5BB-01 (Revision 1), RCS Final Leak Rate.
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- 1-5BB-02 (Revision 1), Pressurizer Heater; and Spray Capability and Continuous Spray Flow Verification
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1-5BB-04 (Revision 1), Reactor Coolant System Flow Measurement h.
1-5B8-05 (Revision 1)', Resistance Temperature Detector Bypass Loop Flow Verification i.
1-5BB-07 (Revision 0), RCS Heat loss Measurement
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-l-SSB-01(Revision 1),ReactorProtectionTest
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1-55C-01 (Revision 1), Core Loading Instrumentation and Neutron Source
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Requirements.
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1-SSE-01 (Revision 1), Movable Incore Detectors i
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1-5SE-02 (Revision 1), Operational Alignment of Nuclear Instrumentation System Precritical n.
1-SSF-02 (Revision 1), Rod Control System o.
1-5SF-03, (Revision 1), CRDM ' Operational Test and Rod Position Indication Test
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1-5SF-06 (Revision 1), Operational Alignment of Process Temperature Instrumentation q.
- 1-SSF-07 (Revision 1), Startup Adjustir,ents of the Reactor Control System
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1-5SJ-01 (Revision 1), Reactor Systems Sampling for Core Loading
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1-55Q-01 (Revision 1), Metal Impact Monitoring System Test t.
1-600-01 (Revision 1), Inverse Count Rate Ratio Monitoring for Approach to Initial Criticality u.
1-600-02 (Revision 1), Initial Criticality v.
1-600-04 (Revision 1), Initial Criticality and Low Power Test Sequence w.
1-600-10 (Revision 1), Natural Circulation Demonstration x.
1-6SF-06 (Revision 1), RCCA Bank Worth Measurement at Zero Power The following three areas were identified as requiring additional action:
a.
(0 pen) UNR 424/86-99-01:
Natural Circulation Test is Scheduled after Operation at 75% Power, rather than prior to Power Ascension b.
(0 pen) IFI 424/86-99-02:
Sample and Control Refueling Canal Boron Concentration Procedures 1-588-06 '(Revision 0), Reactor Coolant Flow Coastdown, and
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1-5SF-04 (Revision 1), Rod Drop Time, were reviewed relative to FSAR acceptance criteria, but the licensee plans to revise these procedures to use different test methods to perform the tests.
The Natural Circulation Test, as described in procedure 1-600-10, is scheduled to be performed after 75% power operation in conjunction with the Loss of Offsite Power Test.
This does not match the test descriptions currently in the FSAR.
Also, operation at 75% power prior to demonstrating natural circulation capability violates the FSAR requirement that tests be sequenced to insure that the safety of the plant does not depend on the performance of untested systems.
Regulatory Guide 1.68 includes the Natural Circulation Demonstration as a low power test to be performed prior to power ascension.
It also describes the Loss of Offsite Power Test as performed between 10 and 25% power.
Usually this is one of the first tests performed during power escalation.
The licensee's reason for scheduling the test this way is to insure enough decay heat is available to drive natural circulation.
Traditionally, this test is performed at 3-5% full power.
This question has been brought to the attention of NRR, and will be tracked as:
(0 pen) UNR 424/86-99-01:
Scheduling Natural Circulation and Loss of Offsite Power Tests After Operation at 75% Power, rather than prior to Power Ascension
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4 The proposed Technical Specifications require that during refueling the baron concentration in the refueling canal be periodically sampled to insure it is maintained at required levels.
Neither procedure 1-500-01, Initial Fuel Load Test Sequence, nor 1-5SJ-01, Reactor Systems Sampling for Core Loading, addresses this requirement.
This will be tracked as:
(0 pen) IFI 424/86-99-02:
Sample and control refueling canal boron concentration The following inspector followup items in the area of startup testing were closed:
a.
(Closed) IFI 424/86-73-01:
Use the statistical reliability factor test to confirm source range nuclear instrumentation capability during fuel loading, initial criticality of a new core, and lowered reactor vessel water level Frequent and appropriate use was made of statistical reliability factor tests in procedures 1-500-02, Inverse Count Rate Ratio, and 1-600-01, Inverse Count Rate Ratio Monitoring for Approach to Initial Criticality b.
(Closed) IFI 424/86-73-02:
In RCS leakage measurements eliminate the effect of a change in T-average by either limiting the change to
+/- 0.1 degrees F or making a mathematical correction for the change in density.
Procedure 1-5BB-01 (Revision 1), RCS Final Leak Rate, now implements surveillance procedure 14905-1, RCS Leakage Calculation (Inventory Balance), which contains provisions for accounting for temperature changes c.
(Closed) IFI 424/86-73-04:
Prevent over-dilution of the VCT by closing the spray valve while in alternate dilute during the approach to initial criticality Procedure 1-600-02 has been revised to adequately address this item.
Prior to the start of dilution, the VCT spray supply valve from the alternate dilute source is isolated to prevent ovar-dilution of the VCT during the lengthy dilution process.
The initial dilution rate of 60 gpm is expected to lead to a reactivity change of 1000 pcm/hr.
Once inverse multiplication (ICRR) reaches 0.2, the dilution rate is reduced to 30 gpm, and the source range monitors are renormalized.
When the ICRR reduces to 0.3, dilution is stopped and the RCS is allowed to mix for at least 30 minutes.
If criticality is not achieved during mixing, reactivity is increased either by rod withdrawal, from bank D at 160 steps or by batch dilution.
The procedure describes an acceptably conservativeapproachtoinitialcriticality.
The inspectors reviewed the status of IFI 424/86-73-02:
Eliminate makeup to the VCT during RCS leak rate tests.
Procedure 1-5BB-01 (Revision 1), RCS Final Leak Rate, implements surveillance procedure 14905-1, RCS Leakage Calculation (Inventory Balance).
Step 3.1 of procedure 14905-01 permits
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additions to and sampling from the RCS if an accurate account of amounts is maintained.
No definition of accurate is given.
Better and more common
. practice is to forbid sampling and additions during.the test.
Procedure 1-588-01 does prohibit-additions when leakrate is measured during startup testing, but because sampling and additions are still allowed in procedure
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'14905-01 for leakrate surveillance, the IFI will remain open.
A procedure
. improvement noted during the inspection increased the test duration to a minimum of two hours.
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