IR 05000335/1986022
| ML17308A274 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/21/1986 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17308A273 | List: |
| References | |
| 50-335-86-22, 50-389-86-21, NUDOCS 8612030698 | |
| Download: ML17308A274 (19) | |
Text
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+y*y4 UNITEO STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900 ATLANTA,GEORGIA 30323 Report Nos.:
50-335/86-22 and 50-389/86-21 Licensee:
Fl ori da Power and Light Company 9250 Mest Flagler Street Miami, FL 33102 Docket Nos.:
50-335 and 50-389 License Nos.:
DPP,-67 and NPF-16 Facility Name:
St. Lucie 1 and
Inspection Conducted:
November 3 - 7, 1986 Inspector:
P.
. Burne t Approved by:
F. Jape, Section Chief Engineering Branch Division of Reactor Safety Date Signe
/i ~y Date Signed SUMMARY Scope:
This routine, unannounced inspection addressed the areas of reactor coolant system leakrate testing, thermal power monitoring, and followup of previous outstanding items.
Results:
No violations or deviations were identified.
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REPORT DETAILS Persons Contacted Licensee Employees
- J. H. Barrow, Operations Superintendent J.
A. Dyer, guality Control
- R. J. Frechette, Chemistry K. N. Harris, Vice President C.
F. Leppla, ( I&C) Supervisor E. Ordway, I
&
C Engineer
- L. W. Pearce, Operations Supervisor
- N. G. Roos, guality Control Supervisor D. A. Sager, Plant Manager
- D. M. Stewart, Technical Staff
- D. H. West, Technical Staff
- C. L. Wilson, Mechanical Maintenance
- B. Winnard, Independent Safety Engineering Group
- E. Wunderlich, Reactor Engineering Other licensee employees contacted included nuclear plant supervisors, assistant nuclear plants supervisors, operators, and office personnel.
NRC Resident Inspectors R. V. Crlenjak, Senior Resident Inspector H. T. Bibb, Resident Inspector
- Attended exit interview Exit Interview The inspection scope and findings were summarized on November 7, 1986, with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection findings.
No dis-senting comments were received from the licensee.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.
Unresolved Items No unresolved item was identified during this inspectio Reactor Coolant System Leakrate Measurement
- Unit 1 (61728)
Data were collected at 15 minute intervals over a 3.25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> period for analysis of the Unit 1 reactor coolant system gross and identified. leak rates.
This period encompassed the time the licensee was performing their routine surveillance to satisfy the requirements of Technical Specification 4.7.2.1.d.
The licensee's procedure calls for a minimum two-hour test.
The time was extended to facilitate comparison with the inspector's calcula-tions, which were performed using the microcomputer program RCSLK9.
That program is described in NUREG-1107, RCSLK9: Reactor Coolant System Leakage Determination for PWRs.
The plant-specific data used in the analysis are shown in attachment 1 and the result of the 3.25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> calculation in attach-ment 2.
The 13 sets of data collected allowed calculation of ten one-hour duration tests and six two-hour duration tests.
The licensee's procedure allows addition of makeup water during the test, but for a 2.25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> period no water was added.
That period was separately analyzed as, well as the total period.
Finally, the program was used to obtain the temporal increase in mass deficit and collected mass at the end of each set of observations.
These two data sets were each reduced to best fit straight lines using least-squares analysis.
The slopes of the lines were taken as gross and identified mass leakage.
The process was repeated for the ten observations of gross leakage without water addition.
The results are summarized in the table below:
MEAN GROSS LEAKAGE (GPM)
1-HOUR TESTS
2-HOUR TESTS
2.25-HOUR TEST
3.25-HOUR TEST FIT TO 13 OBSERVATIONS FIT TO 10 OBSERVATIONS 0.99 +/- 0.434 1.10 +/- 0.126 1.04 0.96 1.06 +/- 0.074 1.08 +/- 0. 147 (no water addition)
The slope of the identified leakage line for 13 observations was 0.71 +/-
0.015 gpm.
The two-thirteen observation fits are shown in attachment 3 and the one-ten observation fit is shown in attachment 4.
Both the regression analysis and the plotting of the results were performed with the SUPERCALC3 spreadsheet proaram.
One conclusion derived from these analyses was that water addition had no adverse effect on the quality of the results.
This was surprising in light of results at other facilities.
However, the makeup flow sensor, which inputs to the flow integrator, is based upon a calibrated orifice rather than a rotating vane or other less precise device, and calibration records showed the flow integrator to have good long term stability.
(The calibra-tion records of all instruments used in the measurement were inspected and found satisfactory.)
Another conclusion, based upon the standard deviations of the sets of one-and two-hour tests, was tlat the routine surveillance test duration should not be reduced below two hours.
The data collected here as well as those used by the licensee were obtained by visual observation of control board instruments, and the tests are nothing more than comparison of endpoints of the observables.
Errors in reading the endpoints become less significant as test duration increases.
Finally, the licensee's method of determining reactor coolant system leak-rate is acceptable.
Their result was bracketed by those obtained by the inspector.
6.
Thermal Power Measurements (61706)
For each unit the hourly surveillance of thermal power is performed by the unit's digital data processing system (DDPS).
The hourly log in addition to showing the result also echoes the variable data input to the calculation.
That set of data when augmented by observation of steam generator and pressurizer levels is sufficient to perform an independent analysis of thermal power using the microcomputer program TPDCER2 (Thermal Power Deter-mination in Combustion Engineering Reactors).
That program has yet to be formally documented, but it is similiar to the program documented in NUREG-1167, TPDWR2: Thermal Power Determination in Westinghouse Reactors, Version 2.
To configure TPDCER2 for use with each unit, unit-specific parameters are required.
To obtain the required data, the following references were reviewed:
a.
Updated Final Safety Analysis Report, St. Lucie Unit 1, January 22, 1986 b.
Updated Final Safety Analysis Report, St. Lucie Unit 2, April, 1986, c.
C.E.
Book No. 74267 (12/72),
Steam Generators, St. Lucie Plant Unit No. 1, January 22, 1986, and d.
C.E.
Book No. 71272 (10/77),
Steam Generators, St. Lucie Plant UniC No. 2.
Subsequently, it was determined that the FSAR descriptions of reactor coolant pump power and efficiency were not adequate, and those parameters were replaced by values determined during preoperational testing and used in the DDPS.
Values for insulation surface area and conductivity were manipu-lated to force agreement with the heat losses determined during preopera-tional testing and used in the DDPS, hence the warning messages on attachments 5 and 6.
The plant parameters used in the final analyses are shown in attachments 5 and 6 for Units 1 and 2, respectivel With the cooperation of the licensee, the inspector was able to obtain four sets of data at 15 minute intervals from each DDPS.
At the same time, the steam generator and pressurizer levels were recorded by hand from main control board instrumentation.
Each analysis performed using TPDCER2 used two sets of data taken 15 minutes apart.
Thus, two analyses were performed on each reactor resulting in four power determinations for each.
The data from the DDPS were, in most cases, in different units from those required by TPDCER2, and it was necessary to set up a
SUPERCALC 3 spreadsheet to make the required conversions reliably.
Clearly, the versatility and flexibility of TPDCER2 would be much enhanced if it could accept a variety of units for the input.
The results of the eight individual power calculations from TPDCER2 were consistently higher than those reported by the DDPS, by from 5.8 to 9. 1 megawatts thermal, but, in perspective, the worst disagreement was only 0.345 of the calculation; A review of the outputs did not reveal any obvious source of the disagreement.
The results of a typical set of calcu-lations by TPDCER2 are given in attachment 7.
It was concluded that the licensee's method of calculating thermal power to assure conformance to the license limit is acceptable.
7.
Followup of Outstanding Items (92701)
(Closed)
Inspector followup item 335/85-28-01:
Discuss the monotonic'hange in the Unit 1 reactivity deviation.
A review of the licensee's correspon-dence files confirmed that there had been continuing discussion with the fuel vendor on the issue of reactivity deviation, and that the vendor had submitted new prediction curves during the cycle.
The licensee ascribed part of the vendor's problem in prediction with lack of experience in analysing the performance of boron carbide burnable poison rods.
This item is closed.
(Closed)
Unresolved item 335/389/85-28-03:
The adequacy of the at-power moderator temperature coefficeient procedure as written and approved i s in question.
The procedure originally reviewed was OP 320051 (Revision 0).
Since that time the licensee has revised the procedure and issued Revision
on March 13, 1986.
The revised procedure contains adequate guidance on the collection and analysis of the required data.
Based upon this observation and discussi ons with the licensee.,
this item is closed.
Attachments:
RCSLK9 Parameter List - Unit
RCSLK9 Results
- Unit
Fit of 13 Observations Fit of 10 Observations TPDCER2 Plant Parameters
- Unit
TPDCER2 Plant Parameters
- Unit 2 TPDCER2 Heat Balance Data - Unit 2 (4 pages)
ATTACHMENT 1 PARAMETER LIST Unit Identification:
Plant Name Unit Number, Docket Number Nuclear Steam System Supplier Vessel and Piping:
Volume Pressurizer:
Level Units Temperature Compensated Calibration Curve Slope Uppex Level Limit Lower level Limit Relief Volume Control Tank:
Level Units Calibration Curve Slope Uppex Level Limit Lower level limit Geometric Method Available Drain Tank:
Level Units Calibration Curve Slope Upper Level Limit Lower level limit Geometric Method Available ST.
LUCIE
50-335 Combustion Engineering 9218 cubic feet No 559.05 pounds per
%
96.5 10.1 Quench Tank 281.5 pounds per
%
100
No 145 pounds per
20 No Quench Tank:
Level Units Calibration Curve Slope Upper Level Limit Lower level limit Geometric Method Available 137.44 pounds per
60
No
ATTACHMENT 2 NRC INDEPENDENT MEASUREMENTS PROGRAM REACTOR COOLING SYSTEM LEAK RATES STATION: ST.
LUCIE UNIT
DOCKET
- 50-335 TEST DATE
4 November 1986 START TIME: 0800 DURATION 3.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> TEST DATA System Parameters Pressure, psia T Ave, degrees F
Water Levels Pressurizer, Quench Tank, Volume Control Tank, Drain Tank, Water Charged
= 210 gal Initial 2250 571.75
53.5 50.5
Water Drained Final 2250 571.5 66.8 60.5
28.5 0 gal TEST RESULTS Change in Water Inventory in pounds:
Vessel
& Piping Pressurizer Volume Control Tank Less:
Water Charged Plus:
Water Drained Cooling System 166 Quench Tank (1)
-112 Drain Tank (1)
141 1748 Collected Leakage
-1553 962 218 1180 Leak Rates in gpm (3):
Gross Identified Unidentified 0.96 0.73 0.23 (1)
(2)
(3)
Determined from tank calibration curve.
Determined from tank. dimensions.
The density used for converting inventory change to leak rate was 62.31 pounds/cubic foot based on standard condition HHENT 3 ST. LUCIE, HCS Leakage Data and Least Squares Fit
ElData GL R69 n GL
~ Data IL Rag'n IL 1000
CL LQ EA I
lf)
lQ lQ P4
~
lC)
lQ tQ lA t~
C4 Test Duration hours
ATTAC
ST. LUCIE, HCS Leakage Data and Least Squares Fit 1.25
'1.5 1.75
2.25 2.5 2. 75 Test Duration hours
ATTACHMENT 5 Plant Parameters Identification Plant Name Unit Number ST.
LUCIE 1 Reactor Coolant System Piping and Components Number of coolant loops Pump power Pump efficiency Pressurizer inner diameter
(2/3)
4.53 MW each 93.3 percent 95.3125 inches Reflective Thermal Insulation Surface area Heat loss coefficient Nonreflective Thermal Insulation 13350 sq ft W 240 BTUs/hr-sq ft Surface area Thickness Thermal conductivity 8800 sq ft
inches
.39 BTUs/hr-ft-F Steam Generators Moisture carry-over Dome inside diameter Low level downcomer O.D.
High level downcomer O.D.
Low water level High water level
.2 232.5 156.25 230.125 79.656 155.656 percent inches inches inches inches inches W Warning, Data or Parameters preceded by W are suspicious or in error.
I
ATTACHMENT 6 Plant Parameters Identification Plant Name Unit Number ST.
LUCIE Reactor Coolant System Piping and Components Number of coolant loops Pump power Pump efficiency Pressurizer inner diameter
(2/3)
4.25 MW each
percent 95.5625 inches Reflective Thermal Insulation Surface area Heat loss coefficient Nonreflective Thermal Insulation 13350 sq ft W 240 BTUs/hr-'sq ft Surface area Thickness Thermal conductivity 8800 sq ft
inches
.39 BTUs/hr-ft-F Steam Generators Moisture carry-over Dome inside diameter Low level downcomer O.D.
High level downcomer O.D.
Low water level High water level
.2 232.5 156.375 231.18 79.656 155.656 percent inches inches inches inches inches W Warning, Data or Parameters preceded by W are suspicious or in erro ATTACHMENT 7 Heat Balance Data, Data Set
Time 951 0000-2400 hours Letdown Line Letdown Flow Letdown Temperature 34.27 195.6 gpm deg.
F Charging Line Charging Flow Charging Temperature 43.07 110.1 gpm deg.
F Pressurizer Pressure Water Level 2250 psia 201 inches Reactor T ave T cold 573.88 deg.
F 549 deg.
F Steam Generator A
Steam pressure Feedwater Flow Feedwater Temperature Surface Blowdown Bottom Blowdown Water level Moisture carry-over 859.8 5.869 437.1
24.48 123
.2 psia E6 lb/hr deg.
F gpm gpm inches percent Steam Generator B
Steam pressure Feedwater Flow Feedwater Temperature Surface Blowdown Bottom Blowdown Water level Moisture carry-over 868.8 6.04 435.9
20.21 117
.2 psia E6 lb/hr deg.
F gpm gpm inches percent
ATTACHHENT 7 Heat Balance Data, Data Set
Time 1005 0000-2400 hours Letdown Line Letdown Flow Letdown Temperature 35.09 gpm 195.3 deg.
F Charging Line Charging Flow Charging Temperature Pressurizer 43.47 gpm 110.1 deg.
F Pressure Water Level 2250 psia 201 inches Reactor T ave T cold 573.88 deg.
F 548.95 deg.
F Steam Generator A
Steam pressure Feedwater Flow Feedwater Temperature Surface Blowdown Bottom Blowdown Water level Moisture carry-over 859.8 5.868 437
25.17 123
.2 psia E6 lb/hr deg.
F gpm gpm inches percent Steam Generator B
Steam pressure Feedwater Flow Feedwater Temperature Surface Blowdown Bottom Blowdown Water level Moisture carry-over 868.4 6.06 435.8
20.43 117
.2 psia E6 lb/hr deg.
F gpm gpm inches percent
ATTACHMENT 7 TPDCER2 HEAT BALANCE Plant Name
ST.
LLCIE Docket No.
50-389 Unit No.:
Date
11/06/86 DATA SET
0951 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.618555e-4 months <br /> ENTHALPY FLOW POWER POWER (BTUs/lb)
(E6 lb/hr)
(E9 BTUH)
(MWt)
STEAM GENERATOR A Steam Feedwater Surface Blowdown Bottom Blowdown Power Dissipated STEAM GENERATOR B
Steam Feedwater Surface Blowdown Bottom Blowdown Power Dissipated OTHER COMPONENTS 1196.3 416.1 520.1 466.6 1196.0 414.8 521.6 466.6 5.859-5.869 0.00000 0.00983 6.032-6.040 0.00000 0.00811 C
7.009-2.442 0.00000 0.00458 4.5715 1339.8 7.214-2.506 0.00000 0.00379 4.7125 1381.1 Letdown Line Charging Line Pressurizer Pumps Insulation Losses Power Dissipated REACTOR POWER 168.8 83.9 611.5 0.01667 0.00281-0.02152-0.00181-0.00002-0.00001-0.05742 0.00971-0.04672-13.7 270 ATTACHMENT 7 DATA SET
1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br /> ENTHALPY FLOW POWER POWER (BTUs/lb)
(E6 jb/hr)
(E9 BTUH)
(MWt)
STEAM GENERATOR A Steam Feedwater Surface Blowdown Bottom Blowdown Power Dissipated STEAM GENERATOR B
1196. 3 416.0 520.1 466.5 5.858-5.868 0.00000 0.01010 7.008-2.441 0.00000 0.00471 4.5712 1339.7 Steam Feedwater Surface Blowdown Bottom Blowdown Power Dissipated OTHER COMPONENTS 1196.0 414.7 521.5 466.5 6.052-6.060 0.00000 0.00820 7.238-2.513 0.00000 0.00383 4.7288 1385.9 Letdown Line Charging Line Pressurizer Pumps Insulation Losses Power Dissipated REACTOR POWER 168.5 83.9 611.5 0.01707 0.00288-0.02172-0.00182-0.00002-0.00001-0.05742 0.00971-0.04667 1 3 ~ 7 2712.0