ML20214P370
| ML20214P370 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/15/1987 |
| From: | Burnett P, Jape F, Long A, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214P355 | List: |
| References | |
| TASK-2.E.1.2, TASK-TM 50-424-87-24, NUDOCS 8706030341 | |
| Download: ML20214P370 (7) | |
See also: IR 05000424/1987024
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Report No.: 50-424/87-24
Licensee: Georgia Power Company
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P. O. Box 4545
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~ Atlanta, GA ~30302
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Docket No.: 50-424
License No.:
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Facility Name: Vogtle 1
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Inspection Conducted: March 7 20, 1987
Inspectors -
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P. T. Bur
Date Signed
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A. R. Long
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J. 'L. Mathis
Date Signed
Approved by:
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F. Jape, Chief f/
Date Signed
Engineering Branc
Division of Reactor Safety
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SUMMARY
Scope:
This routine, unannounced inspection was performed to witness initial
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criticality, performance of zero power physics tests, and to review completed
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tests and surveillance procedures..
Results:
One _ violation was identified, Failure to perform an adequate
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surveillance of reactor coolant system leakage - paragraph 5.d.
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8706030341 870521
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ADOCK 05000424 32
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- C. E. Belflower, Quality Assurance Site Manager
- W. L. Burmeister, Operations Supervisor
- J. F. D'Amico, Nuclear Safety and Compliance Manager
J. A. Edwards, Senior Nuclear Specialist - Operations
R. J. Florian, Reactor Engineering Supervisor
G. R. Frederick, Senior QA Engineer
- W. C. Gabbard, Regulatory Specialist
T. Greene, Plant Manager
T. S. Hargis, Operations Shift Supervisor
C. W. Hayes, Vogtle QA Manager
W. F. Kitchens, Manager of Operations
C. F. Meyer, Operations Superintendent
M. J. Rowe, Operations Superintendent
J. Schwartzwelder, Operations Technical Assistant
Other licensee employees contacted included engineers, technicians,
operators, and office personnel.
Other Organizations
NRC Resident Inspectors
H. H. Livermore, Senior Resident Inspector, Construction
R. F. Rogge, Senior Resident Inspector, Operations
R. J. Schepens, Resident Inspector
- Attended exit interview
2.
Exit Interview
The inspection scope and findings were summarized on March 20, 1987, with
those persons indicated in paragraph 1 above. The inspector described the
areas inspected and discussed in detail the inspection findings.
No
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dissenting comments were received from the licensee.
Proprietary
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information was reviewed during the course of the inspection but is not
incorporated in this report.
Violation 425/87-24-01:
Failure to perform an adequate surveillance of
reactor coolant system leakage - paragraph 5.d.
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3.
Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
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4.
Unresolved Items
Unresolved items were not identified during this: inspection.
5.
Review of Precritical Testing -(72596, 61728)-
The following completed precritical test procedures were reviewed:
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a.
1-5SF-04, Rod Drop Time, was performed at full flow, four - reactor -
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coolant purr.ps running
hot condition, -557 F. and' 2235 psig, on
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March 1, 1987. The drop times of ~ all 53 control-and safety rods'were
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significantly less than the 2.2 second < limit of - technical
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specifications.
Drop time was measured from the beginning of decay
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of stationary gripper coil voltage to dashpot entry.
Test Evaluation . Report (TER) 1-5SF-04-01 was 'writte'n ' beca'use two
control rods, H02 in control bank C group 1 and H08 in control bank D:
group 2, were outside the plus or minus- two standard- deviation
(sigma) limit specified in acceptance criterion.9.3.
Each rod was
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retested by six additional drops.
Each failed the retest criterion
that the span of the six drops for each rod be less - than 0.02
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seconds.
Subsequent evaluation by the~ licersee confirmed that the
distribution of drop times for each rod was statistically expected.-
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b.
1-588-02, Pressurizer Heater and Spray . Capability 1and Continuous
Spray Flow Verification Test, was performed on March 5,- 1987.
The
pressurizer pressure response to opening of _ both pressurizer spray.
valves was within the allowable range, as-was the pressure response
to activation of all pressurizer heaters. - The pressurizer power
operated relief valves open in two seconds or less.-- All acceptance-
criteria were satisfied.
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1-5BB-06, Reactor Coolant Flow Coastdown, was performed on March 5,
1987. There were three TERs written against the procedure. The flow
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coastdown time . constant measured following - the loss 'of- all four
reactor coolant pumps was 13.175 seconds, which.is-greater than the
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design value.
The reactor coolant low flow trip loop time response:
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determined during- the test equalled 0.936 seconds.
- All ~ test -
exceptions appeared reasonable and did not invalidate.the test.-
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d.
Surveillance Procedure No. 14905-1 (Revision 5), RCS : Leakage .
Calculations (Inventory Balance),. was performed on March -7,1987 to
satisfy Technical Specification surveillance requirement 4.4.6.2.1c.
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The reported total leakage was 5.67 gpm.- After correcting for the --
identified leakage-to the pressurizer relief tank (PRT) and reactor.
coolant drain tank (RCDT), the licensee took further credit' for other
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identified leakage of 4.56 gpm, which came primarily from leakage
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measured through a valve in the seal water system., The' net
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unidentified leakage of 0.76 gpm was - acceptable.
To. verify the
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acceptability of the procedure proper and the constants used in it,
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the same data recorded by the licensee were analyzed using RCSLK9. -
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That microcomputer program is fully described in NUREG-1107, "RCSLK9;
Reactor Coolant System Leak Rate Determination for PWRs." .The
plant-specific data necessary to customize the program for use on
Vogtle Unit' I were obtained from the FSAR and the Plant Technical
Data Book.
Those plant parameters are given in Attachment 1 to this
report.
-The results of the RCSLK9 calculation (Attachment 2) were
significantly different, except for good agreement with the amount of
leakage into the PRT and the RCDT. - The total or gross leakages-
differed by an unacceptability large 1.16 gpm. . Use of RCSLK9 at
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other facilities has demonstrated consistent agreement within 0.2 gpm
for acceptable procedures.
A detailed review of procedure 14905-1
revealed significant errors in the constant used to correct for
changes in pressurizer level and in the equation used to adjust for
changes in reactor coolant system average temperature. .Although in
the instance examined the error could .be considered conservative,
that would not be the case for all expected variations in level or
temperature.
Hence, the procedure is not. adequate to perform the.
required surveillance. This has been identified as VIO 424/87-24-01:
Failure to perform an adequate surveillance of reactor coolant system
leakage.
No other violations or deviations were identified.
6.
Initial Criticality Witnessing (72592)
The inspectors witnessed the withdrawal of. the control banks, initiation
of dilution to criticality, and were in the control room through most of
the dilution process including the attainment of criticality at 8:35 am on
March 9, 1987.
In addition, the obtaining and analysis of pressurizer and
reactor coolant system boron samples were witnessed.
Independent
statistical analyses (chi squared tests) were performed on the source
range nuclear instruments to confirm proper functioning.
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The initial critical configuration of control bank D and boron concentra-
tion was in good agreement with the predicted configuration. Criticality
was achieved in a well-controlled manner and in full adherence to
procedure.
No violations or deviations were identified.
7.
Zero Power Physics Tests (72572)
Portions of the following tests were witnessed in the control room.- .The
completed test procedures were reviewed for completeness and calculations
within the procedures were spot-checked,
a.
Boron Endpoint Measurements
Boron endpoint measurements were performed in accordance with '
procedure for the following control rod configurations: all rods out
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(AR0), control; bank D fully- inserted, control' banks D and C fully
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inserted, control banks D, C, and B fully inserted, control ~ banks D.
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C, B and A fully-inserted. For each configuration, the measured and'
predicted values agreed within the ' acceptance criterion on +/-- 10% of'
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the predicted-values.
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b.
Isothermal Temperature Coefficients (61708)
The isothermal temperature coefficient was. measured in accordance
with procedure for the AR0 and D-bank-in_ configurations. At.ARO the
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corresponding moderator temperature coefficient :.was slightly
positive, which is contrary to Technical Specifications.
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licensee initiated the action statement to establish rod withdrawal
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limits to assure that the operating moderator coefficient would be-
negative.
Implementation and enforcement of those limits will be inspected.
during a later inspection.
c.
Control Rod Worth Meas'urements (61710)
Control rod worth measurements were -performed in; accordance with
procedure for each control bank-in succession starting'with control-
bank _D and proceeding to shutdown bank'B.
Each: bank measurement
satisfied the acceptance criterion of +/- 10% agreement with
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predicted values.
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No violations or deviations were identified
8.
Followup to Open Items (92701)
(Closed) TMI Action Item II.E.1- 2, " Auxiliary Feedwater System Automatic
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Initiation and- Flow Indication."
The review of diagrams and drawings
documented in inspection report 50-424/86-90 confirmed that the auxiliary
feedwater (AFW) system automatic initiation and flowrate indication were
designed in accordance with requirements.
However, testing of the system -
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was not complete when the report was . issued because ' of required -
modifications to the motor operated discharge valves- and flow orifices -
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identified in preliminary - tests.
Retesting was completed on March 1,
1987.
The inspector reviewed the section of the completed test procedure that-
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involved retesting of the motor operated discharge valves ' and flow
orifices. .The motor driven AFW pumps each delivered a minimum of-630 gpm
and 1175 gpm in concert at a total discharge head of 3500.+ 105--O ft.
All acceptance criteria were met, and all test exceptions were resolved
without' invalidating the test. TMI Action Item II.E.1.2 is closed.
No violations or deviations were identified.
ATTACHMENTS:
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Parameter List
2.
Reactor Cooling System Leak Rates
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ATTACHMENT 1
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PARAMETER LIST
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Unit Identification:
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Plant Name.
VOGTLE
Unit Number
1
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Docket Number
50-424
Nuclear Steam System Supplier
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Vessel and Piping:
Volume
10662 cubic feet
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Pressurizer:
Level Units
%.
2
Temperature Compensated
No
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Calibration Curve
Slope
617.52 pounds-per %
Upper Level Limit
100 %
Lower level Limit
0-%
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Relief
Relief Tank
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Volume Control Tank:
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Level Units
%
Calibration Curve
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Slope
'159.4 pounds per %
Upper Level Limit
100 %
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Lower level limit
O%
Geometric Method Available
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Drain Tank:
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Level Units
%
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Calibration Curve
Slope
27.97 pounds per %
Upper Level Limit
75 %
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Lower level limit
20 %
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Geometric Method Available
No
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Relief Tank:
Level Units
%
Calibration Curve
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Slope
1249 pounds per %
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Upper Level Limit
64.74 %.
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Lower level limit
24.74 %
Geometric Method Available
No
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ATTACHMENT 2
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NRC
INDEPENDENT MEASUREMENTS PROGRAM
REACTOR COOLING SYSTEM LEAK RATES
STATION: VOGTLE
TEST DATE : March 7,
1987
UNIT
- 1
START TIME: 0451
DOCKET : 50-424
DURATION
2.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
TEST DATA
Initial
Final
System Parameters
Pressure, psia
2251.3
2259.1
T Ave, degrees F
556
556
Water Levels
Pressurizer, %
25.83
23.16
Relief Tank, %
61.5
61.8
Volume Control Tank, %
56.52
36.5
Drain Tank, %
72
72.1
Water Charged = 0 gal
Water Drained = 0 gal
TEST RESULTS
Change in Water Inventory in pounds:
Vessel & Piping
60
Relief Tank (1)
375
Pressurizer
-1649
Drain Tank (1)
3
Volume Control Tank (1) -3191
Less: Water Charged
0
Collected Leakage
377
Plus: Water Drained
0
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Cooling System
-4780
Leak Rates in gpm (3):
Gross
4.51
Identified
0.36
Unidentified
4.16
(1)
Determined from tank calibration curve.
(2)
Determined from tank dimensions.
(3)
The density used for converting inventory change to leak
rate was 62.31 pounds / cubic foot based on standard
conditions.
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