ML19330A498
| ML19330A498 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/22/1980 |
| From: | Kann G VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML19330A497 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-660 NUDOCS 8007280400 | |
| Download: ML19330A498 (30) | |
Text
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s VIRGINIA ELECTRIC AND POWER COMPANY .
NORTH ANNA POWER STATION UNIT 2 NATURAL CIRCULATION TEST REPORT Conducted July 3, 1980 Through July 16, 1980 G. A. Kann Engineering Supervisor 1
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VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 2 -
NATURAL CIRCULATION TEST REPORT TABLE OF CONTENTS i LIST OF FIGURES ii 1
I. INTRODUCTION 1 II. NATURAL CIRCULATION TEST CHRONOLOGY 3 III. TEST RESULTS A. C00LD0WN CAPABILITY OF THE CVCS (2-ST-6) 6 B. CORE POWER MONITORING AND NUCLEAR INSTRUMENTATION 7 CALIBRATION (2-ST-7) s C. NATURAL CIRCULATION VERIFICATION (2-ST-8) 9 D. NATURAL CIRCULATION WITH SIMULATED LOSS OF POWER CONDITIONS (2-ST-9) 11 E. EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATION ON NATURAL CIRCULATION (2-ST-ll) 12 IV. CONCLUSIONS 14 V. FIGURES 18 i
4 LIST OF FIGURES FIG 1 REACTOR COOLANT TEMPERATURE VS. TIME i8 DURING C00LDOWN CAPABILITY OF THE CVCS (2-ST-6)
REACTOR C0OLANT TEMPERATURE VS. TIME 19 FIG 2 DURING NATURAL CIRCULATION VERIFICATION (2-ST-8) 20 FIG 3 REACTOR COOLANT PRESSURE VS. TIME DURING NATURAL CIRCULATION VERIFICATION (2-ST-8)
PRESSURIZER TEMPERATURE VS. TIME DURING 21 FIG 4 NATURAL CIRCULATION VERIFICATION (2-ST-8)
FIG 5 PRESSURIZER PRESSURE VS. TIME DURING 22 NATURAL CIRCULATION VERIFICATION (2-ST-8)
FIG 6 SATURATION MARGIN VS. TIME DURING NATURAL 23 CIRCULATION VERIFICATION (2-ST-8)
FIG 7 REACTOR COOLANT TEMPERATURE VS. TIME DURING 24 NATURAL CIRCULATION WITH SIMULATED LOSS OF. POWER CONDITIONS (2-ST-9)
FIG 8 REACTOR COOLANT PRESSURE VS. TIME DURING 25 NATURAL CIRCULATION WITH SIMULATED LOSS OF POWER CONDITIONS (2-ST-9)
FIG 9 REACTOR COOLANT TEMPERATURE VS. TIME 26 DURING NATURAL CIRCULATION WHILE ISOLATING A STEAM GENERATOR (2-ST-11)
FIG 10 REACTOR COOLANT TEMPERATURE VS. TIME DURING 27 NATURAL CIRCULATION WHILE UNIS0LATING A STEAM GENERATOR (2-ST-11) ii
a VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 2 ;
NATURAL CIRCULATION TEST REPORT I. INTRODUCTION This report presents a synopsis of the Natural Circulation Testing success-fully completed at North Anna Unit 2 from July 3 through July 16,1980 at power levels up to 3 percent of full power. This program consisted of 4 tests as follows:I 2-ST-6 Cooldown Capability of the Chemical and Volume Control System 2-ST-8 Natural Circulation Verification 2-ST-9 Natural Circulation with Simulated Loss of Power Conditions 2-ST-ll Effect of Steam Generator Secondary Side Isolation on Natural Circulation In addition, 2-ST-7, Core Power Monitoring and Nuclear Instrumentation (NI)
Calibration, was performed to monitor core power during the testing and to determine the relationship between cold leg RC temperature and NI detector signals at a constant core power.
The license to perform this special low power test program was issued by the NRC on July 3, 1980. The first series of tests (2-ST-6) was started on July 3, at 1910, and the last test was completed on July 16, at 2125. In order to satisfy the requirements of participation by each licensed operator in at least 2-ST-9 and observation of two other tests, 2-ST-8 and 2-ST-9 were repeated five and four times, respectively.
These series of tests demonstrated the plant's capability in several simulat-ed degraded modes of operation and proved to be very beneficial from the stand-point of operator training. It provided meaningful data beyond that obtained in the normal startup test program.
The following sections describe the chronology of events which occurred during this program and the test results including significant events.
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II. NATURAL CIRCULATION TEST CHRONOLOGY '
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Time Event -
Date .
Amendment No. 1 to Facility License No. NPR-7 issued.
7/3/80 --
i 7/3/80 1910 Started 2-ST-6 4
2145 Completed 2-ST-6 7/4/80 0410 N1 power range high flux trip setpoint set at 7% power.
Control TAVG (T-409) rescaled in preparation for 2-ST-7.
1509 Started first portion of 2-ST-7 to obtain NI data at 3% power.
2000 Completed first portion of 2-ST-7.
7/5/80 0139 NI calibrated to normalized values obtained in 2-ST-7.
T 0437 Engineered Safety Featuras and Reactor Protection Modi-
)
fication required for ST-8, ST-9, and ST-il, completed and tested.
All temporary recorders required for ST-8, ST-9, and ST-ll connected.
1200 Started 'rst run of 2-ST-8.
7/6/80 0045 Completed first run of 2-ST-8.
0519 Started first run of 2-ST-9.
0725 Completed first run of 2-ST-9.
1217 Reactor tripped while escalating to 3% FP in preparation for conducting second portion of 2-ST-7. Trip caused by SG "C" low level coincident with main steam /feedwater flow mismatch.
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Date M Event 7/6/80 1922 Started 2-ST-7, second portion.
7/7/80 0055 Completed 2-ST-7, second portion.
0134 Started 2-ST-11 .
0830 Completed 2-ST-11.
1525 Started second run of 2-ST-9.
1645 Completed second run of 2-ST-9.
7/8/80 0049 Started second run of 2-ST-8.
0415 Completed second run of 2-ST-8.
0920 Started third run of 2-ST-8.
1447 Reactor tripped on NIS Intermediate Range Hi Flux Setpoint. At time of trip, Instrument Technicians were performing weekly Periodic Test 2-PT-30.1 on intermediate range NI.
1530 Completed third run of 2-ST-8.
2105 Started third run of 2-ST-9.
2300 Completed third run of 2-ST-9.
7/9/80 0150 Started fourth run of 2-ST-8.
0445 Completed fourth run of 2-ST-8. Engineered Safety Features and Reactor Protection returned to normal.
7/16/80 1200 Engineered Safety Features and Reactor Protection modified for special test.
1339 Started fourth run of 2-ST-9.
1457 Completed fourth run of 2-ST-9.
1652 Started fifth run of 2-ST-8.
~1750 Reactor trip on low-low pressure. Failure to block low pressure SI trip.
Date Time Event 7/16/80 2125 Completed fifth run of 2-ST-8. Engineered Safety Features and Reactor Protection returned to normal.
Testing completed.
III. TEST RESULTS This section discusses the details of each Natural Circulation Test results. .
A. C00LD0WN CAPABILITY OF THE CHEMICAL AND VOLUME CONTROL SYSTEM (2-ST The objective of this test was to determine the capability of the charging and letdown system to cool down the reactor coolant system with the steam generators isolated and one Reactor Coolant Pump (RCP) running. With the reactor shutdown and approximately 535 F and 2235 psig, two of three RC pumps were tripped and all three steam generators were isolated. The rate of change of temperature was easily controlled and slow enough to allow easy reduction of data.
The letdown and charging flow were first established at maximum letdown flow for the first 18 minute period. At this time, the RCS had cooled down from 537.55 F to 536.95 F. The cooldown appeared to have been stopped by reducing charging flow from about 106 gpm to about 93 gpm (See Fig. I for time period 18 min. - 26 min.). When charging flow was increased to 100 gpm, the cooldown resumed. The measured T AVG change during this 30 minute period was 0.7 F. The cooldown rate was calculated to be 1.4 F/hr.
Af ter 30 minutes, a minimum letdown flow of 53 gpm was established.
F Referring to Figure 1, the average temperature increase was 1.1 for a 30 minute period. The heatup rate was calculated to be 2.2 F/hr.
The " knees" evident in Figure 1 were caused by manual adjustment of charging rate during the test.
i B. CORE POWER MONITORING AND NUCLEAR INSTRUMENTATION CALIBRATION (2-ST-7)
This procedure was prepared and conducted to achieve two purposes:
- 1) To monitor core power during low power operation under forced circula-tion, and 2) To acquire data to determine the relationship between cold leg RC temperatures and the power and intermediate range N1 detector signals at a constant core power (s 1%) during on RCS cooldown to s 515* F.
To monitor core power the control channel AT instrumentation was monitored using an expanded scale three pen recorder. This recorder was scaled such that full scale would indicate a AT equivalent to 5% FP or 3.1 F. This information was based on Unit 1 data which has an identical NSSS and had a similar Cycle 1 core load. To accomodate a cooldown lower than 530* F, the instrumentation on control AT was rescaled. The normal scaling for control AT is to indicate accurately in the range of 530 to 630 F.
This range was rescaled to read accurately in the range of 500 to 600 F.
This was acceptable during the Natural Circulation Test Program because the control AT functions were not required for plant nperation.
To obtain the data on power and intermediate range currents, two systems were used: 1) Indicated currents from control board indicators were manually logged, and 2) a 5 min trend was initiated on the plant computer to automatically log the corresponding voltage signals for the power range
< channels and currents for intermediate range channels. To facilitate the data analysis, control AT and cold leg temperature data was also trended roincident with the detector data.
The test was conducted in two phases. In the first phase, data was acquired during operation at 547 F s 3% FP from intemediate and power range detectors. This data was analyzed to detennine a more appropriate scaling for power range detectors than that used during a normal plant startup. This was necessary to achieve a range of adjustment on the power range scaling pot such that any slight change in detector response during the cooldown could be accounted for by front panel adjustment. The inter-mediate range currents were analyzed to provide confidence that the inter-mediate range trip setpoint was adjusted to < 7% FP and would remain conserva-tive during a cooldown where the indicated current to power ratio was expected to decrease.
In the second phase, data was acquired during the cooldown from 547 to 515 F with reactor power at approximately 1% FP. During the cooldown, the indicate.d % FP reading from power range NI's remained fairly constant and was determintd to remain within the accuracy of adjustment. Upon reaching
-8 515 F, the cooldown was tenninated and reactor power was reduced to s 10 amps on tb: reactivity computer. At this point, the reactivity computer XY plotter was employed for a further cooldown to s 505 F to acquire a reactivity vs. temperature plot. This plot revealed an isothermal temperature coefficient of -1.14 PCM/ F.
At anproximately 505 F, reactor power was increased to approximately li FP. Based on review of data collected during the cooldown, the power range scaling pots were adjusted to agree with indicated calorimetric AT power.
Sufficient range on the scaling pots was available tc make these adjustments.
Intermediate range trip setpoints were estimated to be set at < 7% full power (trip setpoint of 80 pa) based on intermediate range control board indications.
C. NATURAL CIRCULAT109 VERIFICATION (2-ST-8)
On July 5,1980, the plant was placed in natural circulation condi-tions for the first time. After achieving steady-state, condilions were held constant long enough to review the results of a T/C map and a full core flux map. No anomalies were noted, and the loss of pressurizer heaters and operation at reduced saturation margin tests were conducted. The total duration was 11 1/2 hours. On July 8, 1980, the procedure was conducted twice, perfoming only the natural circulation and loss of heaters parts. Both perfomances were initiated at a reduced RCS pressure of about 2100 psig. The first perfomance took 31/2 hours and the second perfonnance took 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.2 On July 9, 1980, the procedure did not involve performance of the loss of pressurizer heaters section. Initial system pressure was normal, but Auxiliary Spray was initiated imediately to limit the pressure excursion. The duration of the test was about three hours. The last execution of the procedure on July 16, 1980 also did not involve performance of the loss of pressurizer heaters section. The duration of the test was about 4 1/2 hours.
Each perfomance of this procedure was initiated as planned by tripping the RCP's while holding the reactor at very close to 3% RTP.
Plant response to single-step movements of D Bank is readily apparent in the data.
I The initial pressure transient in each case was slow and easily controllable, although somewhat higher than predicted. When initiating the transient from normal system pressure without using Auxiliary Spray, the maximum observed pressure on any protection channel was about 2310 l
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psig. This peak occurred at about +4 minutes, but remained off setpoint long enough for the integral portion of the controller to cause the PORV to lift. This was not anticipated. Subsequent repetitions of the procedure were modified to allow reduced i' .tial pressure (2135 psig) or imediate initiation of Auxiliary Spray. This effectively precluded lifting of a PORV. It was demonstrated that, if desired, Auxiliary Spray could almost completely stop the pressure increase.
The entire transient took 15 to 20 minutes to stabilize at a AT of 36 to 40 F. Initially, cold leg temperatures would drop up to 6 F, requiring adjustment of the single operating steam dump to recover heat.
Primary plant temperatures were found to be most sensitive to steam generator feedwater flow rate, which was quite f'! cult to stabilize at such low steaming rates. The effect of the transient on S/G levels was not observable; in each case some small S/G level transient was in progress when the RCP's were tripped.
Pressurizer level behaved almost exactly as predicted, dropping with cold leg temperature initially, and then increasing to the predicted level. I The huxiliary Spray Test section was completely successful in all i
respects. This mode of spray control was used in all subsequent tests.
It was demonstrated that pressure could easily be turned without undue sensitivity under all conditions observed.
Pressurizer pressure and temperature decay rate with loss of heaters was measured three different times and was not reproducible. Data was collected for a sufficient length of time to eliminate the effects of feedwater induced perturbations. The cause for the discrepancy has yet 4
to be detennined.
The training aspects of operation at reduced saturatior. margin was highly successful. The operators easily obse.ved an increasing margin when charging flow was increased only 10 to 20 gpm. With other para-meters held constant, a 4 F increase in margin was observed 6y increasing pressurizer level only 2%. Manipulation of the steam dump was demonstrated to be ineffective in restoring saturation margin when charging flow was held constant.
D. NATURAL CIRCULATION WITH SIMULATED LOSS OF POWER CONDITIONS (2ST9)
This procedure provided the mechanism for operator training during natural circulation with simulated loss of power conditions. The test was conducted four times (July 6, 7, 8 and 16). No problems were encountered during the tests except for a slight overfeed during the third test.
The overfeed resulted in a small delay in the test to restore temperature.
To achieve the test conditions of natural circulation at 1% RTP with loss of off-site power, auxiliary feedwater control valves MOV-FW200D, MOV-FW-200B, and HCV-200C were closed, steam dump controls were placed in manual and closed, non-emergency powered pressurizer heaters were tripped and locked out, and all three reactor coolant pumps and the main feedwater pump were simultaneously tripped. In the second stage of the test, loss of all AC power was simulated by securing the Auxiliary Feed PumpHouse Turbine Pump Room Exhaust Fan, Motor Pump Room Exhaust Fans, Motor Driven Auxiliary Feed Pumps 2-FW-P-3A and B, and pressurizer heater groups 1 and 4. Steam generators were fed by the turbine driven auxiliary feed pump. The test sequence required approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
During initiation of the natural circulation transient for the
first test, a PORV was actuated. For subsequent tests, auxiliary pressurizer spray was initiated to preclude lifting of the PORV. Steady state natural circulation conditions were achieved in approxi.mately 15 minutes for each test.
E. EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATION OF NATURAL CIRCULATION (2-ST-ll)
This procedure was conducted only once, on July 7,1980. Some data was lost because of loss of the plant computer coincident with tripping of the RCP's. Since the initial transient was no different from that observed during the conduct of 2-ST-8 and 9, the plant was stabilized, the computer was reinitialized, and testing was continued. The duration of the test was 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
Following stabilization of natural circulation conditions at 1%
RTP, 1B steam generator secondary siae was isolated and a very slow transient ensued. At about +1 hour into the transient, it became apparent that the steady increase in IB steam generator level was in excess of that expected due to temperature increase alone. B loop AT had dropped to 10 F and appeared to be holding.
Attempts at isolating the valve leakage into B S/G were only marginal-ly successful, each attempt producing a small reduction in AT until j
! about 6 F had been achieved. At this time feedwater discharge header pressure was reduced to a value sufficient to maintain feed to 1A and 1C steam generator, but less than the pressure in 1B S/G. 1B level was 5
reduced to 37% NR via blowdown and corrected B loop AT then dropped to 1 F within 30 minutes.
I' Restoration of flow in the idle loop, initiated by steaming and feeding, was as expected and uneventful. 3-loop, balanced, steady-state natural circulation conditions were confinned within 30 minut,es.
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IV. CONCLUSIONS The special low power test program at North Anna Unit 2 was completed satisfactorily and met all of its intended purposes. The test results demonstrated the plant's capability in several degraded modes of operation and provided meaningful data beyond that obtained in the normal startup test pro 5 ram.
The effects on RCS TAVG of variations in charging and letdown flowrates were readily apparent, and the rate of change in temperature was easily controlled. Steady state natural circulation conditions were easily establish-ed with no anomalies observed. Saturation margin was maintained without re.ctor coolant pump flow. Adequate saturation margin was also confirmed with-out pressurizer heaters and reactor coolant flow. The installed saturation meters indicated close agreement with the ASME steam tables and were always more conservative than the saturated margin derived from the steam tables.
The capability to maintain natural circulation cooling with sumuiated loss of offsite AC power was verified while maintaining steam generator level with all auxiliary feedwater pumps. Natural Circulation was also verified under conditions of simulated loss of offsite and onsite AC power while maintaining steam generator levels with only the steam driven auxiliary feedwater pump. The transient resulting from a partial loss of heat sink (one steam generator) with sufficient natural circulation flow to remove 1 percent reactor power was demonstrated, as well as establishing balanced natural l circulation in the primary loops after the isolated steam generator was returned to service. At no time during this test program did the plant 1
approach a condition where a reactor trip or test termination became necessary. No Operational Safety Criteria were met. A plant trip did occur twice during the Natural Circulation Test Program. At no time was plant safety compromised; plant behavior was as expected. -
The most beneficial aspect of this program was the experience gained by the plant operators in responding to degraded reactor operations and critical system alignments. This became evident during the subsequent runs e
of the natural circulation tests where the transition from forced flow to natural circulation was observed to be smoother due to the operators' increased awareness of plant behavior under these adverse conditions.
On the basis of the above conclusions, we have determined that the specie.1 test program was successful in all respects.
FOOTNOTES
- 1. Some of the North Anna Test Procedures include more than 6he TVA test.
NAPS TVA 2-ST-6 Test No. 6 2-ST-8 Test No.1 Test No. 3 Test No. 5 2-ST-9 Test No. 2 Test No. 7 2-ST-11 Test No. 4 TVA Tests 8 and 9 were not performed at North Anna Power Station.
- 2. Testing was inadvertently interrupted by a reactor trip unassociated with the conditions of natural circulation being maintained.
- 3. Investigations into the amount of noncondensible gasses in the pressurizer vapor space are being conducted at North Anna Unit 1.
, 4. Charging flowrate was held constant at about 50 gpm and auxiliary spray was stopped during each measurement. RCP seal injection flowrate was seen to be very sensitive to changes in RCS pressure, but the technique of the-individual operators in maintaining this flow was not recorded.
Varying amounts of PORV leakage were oberved.
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- 5. The minimum Wide Range TH0T - TCOLD indication on the CRT was 1.9 Imediately following the test, after forced circulation tad been established with the reactor shut down and under isothermal conditions,
,B loop AT indicated 0.9 . ~ .
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2-ST-6 .
REACTOR COOLANT TEM?EPATURE VS. TIME DURING C00LDOWN CAPABILITY OF THE CVCS C
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5 538.2 g538.0 MINIMUf1 g
o LETDOWN w 537.6 537.4 537.2 i
MAXIMUM l 537.0 LETDOWN x
I 536.8 m i
i 0 4 8 12 16 20 24 28 32 i TIME (MINUTES)
(2-ST-8)
REACTOR COOLANT TEMPERATURE VERSUS TIME DURING NATURAL CIRCULATION VERIFICATION 600 1 2 3 4
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KEY NOTES I '
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--: Average T hot 1: Tripped All RCP's , ,
520 -
- Average Tcold 2: Increased RCS Charging , ,
: Loop "B" Tcold 3: Conmence Bank "D" Insertion , ,
4: Restart RCP "C" ,
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/f 500 0 1 2 3 4 5 y 10 11 12 TIME (HOURS)
FIGURE 2: AVERAGE HOT AND COLD LEG WIDE RANGE VS. Tif1E
- . .. _. . . = _ . -
(2-ST-8)
REACTOR COOLANT PRESSURE VS. TIME DURING NATURAL CIRCULATION VERIFICATION.
2350 3 4 5 6 7 8
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' 1: Tr'ipped All RCP'S 2050 -
2: PORV Lifts (3 Times) 3: Pressurizer Heaters Secured p 4: Pressurizer Heaters Energized %
5: Pressurizer Heaters Secured Auxiliary Spray Commenced 6: Auxiliary Spray Secured 1850 RCS Charging Increased 7: Pressurizer Heaters Energized
. 8: Started RCP "C" ,,
bl N 1750 p 1 i 5 ( 7 8 9 10 11 12 TIME (HOURS) i FIGURE 3: REACTOR COOLANT PRESSURE VS. TIME I
_______.-____.____._______._________________m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
2-ST-8, STEP 4.18 PRESSURIZER C00LDOWN RATE PRESSURIZER LIQUID TEMPERATURE T0480A VS. TIME 652 g 650 L
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5 5.24 F/HR 4 $
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2 640 E
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20 40 60 80 100 120 140 160 0
TIME (MINUTES)
PRESSURIZER PRESSURE VS. TIME -
DURif1G f1ATURAL CIRCULATI0f4 VERIFICATI0f1(2-ST-8) 2250 2225 2
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w 30 60 90 120 150 180 210 240 270 TIME (Mif1UTES)
FIGURE 6 SATURATION MARGIN VS. TIME DURING NATURAL CIRCULATIOi, 70 - VERIFICATION (2-ST-8)
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- 4 4 os - - ACTUAL MARGIN (ASME TABLES) 4 4
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- s 1 20 10-2140 2150 220U 22iG 2220 2230 2240 2250 2300 2310 2320 2330 TIME (CLOCK TIME) 2400 R.C. SYSTEM PRESSURE VS. TIME 2-ST-8 (P0 499A)
INITIATE PZR SPRAY 2300~
$2200 2 c.
W 52100 0
START CHARGING E
2000 PUMP 1900 INCREASE STEAM i
DUMP FLOW 1800 s 1700 10 20 30 40 50 60 70 80 90 100 120 130 l
TIME (MINUTES) l
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REACTOR COOLANT TEMPERATURE VS. TIME DURING NATURAL CIRCULATION CONDITIONS WITH SIMULATED LOSS OF POWER CONDITIONS 2-ST-9 AVERAGE HOT LEG AND COLD LEG WIDE RANGE VS. TIME C
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-AUX. FEEDWATER PUMPS W
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A NJ 530 0 20 40 60 80 100 120 140 TIME (MINUTES)
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2-ST-9 REACTOR COOLANT PRESSURE VS. TIME DURING NATURAL CIRCULATION WITH SIMULATED LOSS OF POWER 2320 2300 E
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0 20 40 60 80 100 120 140 160 TIME (MINUTES)
FIGURE 9
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9 2-ST-11 REACTOR COOLANT T2MPERATURE VS. TIME DURING NATURAL CIRCULATION WHILE UNISOLATINL A STEAM GENERATOR H0T LEG AND COLD LEG VS. TIME C
L E 550 3
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20 40 60 80 100 120 140 0 ,
TIME (MINUTES)
___-_ _ _ _ ____ -___ _ ______________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - __-___ _ __ _ _ _ _ __ -_ . -_ __ _