ML20062D616
| ML20062D616 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/22/1978 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20062D614 | List: |
| References | |
| NUDOCS 7811270156 | |
| Download: ML20062D616 (104) | |
Text
{{#Wiki_filter:e , 0 1 e. THE FORT ST. VRAIN INITIAL APPROAGI TO POWER TESTS (B-SERIES) INTERIM REPORT 8 Report for Period Ending August 22, 1978 781127ciSL
() , TABLE OF CONTENTS Page In t ro d u c t io n - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 3 Acknowledgement --------------------------- 10 Historical Sucmary of Plant Operation ---------------- 11 Testing Sucmary --------------------------- 16 Steam System Perfor:ance Verification (B-1) ------------- 20 Chemical Impurities in the Primary Coolant ( B- 2) - - - - - - - - - - - 30 PC RV Perfo rmance Tes ts (B-3) - - - - - - - - - - - - - - - - - - - - - 35 Section A - PCRV Liner Cooling System - - - - - - - - - - - - - - 35 Section B - PCRV S tructural Performance - - - - - - - - - - - - - 35 Section C - PCRV Leak Tightness - - - - - - - - - - - - - - - - - 35 ! i . r Primary Sys tem Perfo rmance (B-4) - - - - - - - - - - - - - - - - - - - 38 ; Plant Instrumentation Performance (B-5) --------------- 43 l Plant Transient Perfor=ance (B-6) ------------------ 44 6 Plant Automatic Control System Performance (B-7) - - - - - - - - - - - 103 , Reactivity Coefficient FMasurements (B-8) - - - - - - - - - - - - - - 104 Differential Rod Worths (B-9) - - - - - - - - - - - - - - - - - - - - 107 i Xenon Buildup and Decay Ikasurements (B-10) - - - - - - - - - - - - - 10 8 , Xenon S tab ility Tes ts (B-11) - - - - - - - - - - - .- - - - - - - - - 109 , f Shielding Surveys (B-12) - - - - - - - - - - 2 - - - - - - - - - - - 110 l +' l Radiochemical AnalysLi sf the P.i=ary Coolant (B-13) - - - - - - - - - 111 l l
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s INITIAL APPROACII TO POUER TESTS (D SERIES STARTUP TESTS) The initial approach to power is accomplished in a series of dis-crete power level stages. At each power level, tests are made to measure the characteristics of the plant and to ensure that the plant is within its design limits and the power can be safely increased to the next stage. ; The initial phase of 'the approach to power program will increase j the reactor power and steam conditions in stages until approximately 28% power , when rated steam conditions are achieved. From this level to full power the reactor power is increased in stages maintaining rated steam conditions. i . The sequence for the.perfor=ance of these tests is given in Figure 1 together with the corresponding approxicate reactor power levels. The reactor power levels, helium flow rates, feeduater flow rates, steam temperatures, and steam , pressure given in the following description of the initial approach to power may di ffer somewhat from those in the actual approach to power due to change in test requirements or i= prove =ents in operating methods identified during t other tests. t In general, the initial approach to power will be accomplished in
- the following order:
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- 1. Feedwater flow wi.1 first be established through both steam ,
I generator loops and the bypass flash tank system using a boilcr feed pump. ; Helium flow through the core will be provided using one circulator in each loop. [
- 2. The reactor power will be increased to approximately 2%. .
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- 3. The reactor power, feedwater flow, and helium flow rate will be simultaneously increased to 5% power, 20% helium flow, and 25% feedwater i
flow using reactor generated steam from the bypass flash tank supplemented ! by the auxiliary boiler to power the circulator turbines, turbine driven boiler feed pump, and other plant steam requirements.
- 4. The reactor power will then be increased to approximately 8%
l concurrent with an increase in'feedwater flow to about 30%. The helium flow I will be maintained at about 20% during this power increase. At this condi- ! tion the second circulator in each loop will be started, maintaining constant helium flow and the main steam pressure will be increased to 2,400 psig.
- 5. The reactor power will be increased to about 11% and feedwater will,be reduced to 25% to initiate boiling.
([) 6. The reactor power will be increased to about 18% simultaneously with an increase in helium flow to about 33%, maintaining a 25% feedwater flow, followed by an increase in reactor power to about 26% with a helium flow of 49%. Nt this condition the main steam te=perature will ' e about ' 800*F. .
- 7. The helium flow will then be reduced to about 40% concurrent '
. with a slight adjustment of the reactor power to about 28%.
- 8. The reactor power will be increased in stages to about 40%,
50%, 60%, 80%, and finally to 100% of full power. During these power level increases, the helium flow rate through the core will be increased to main-tain full steam conditions. . This report covers teste performed at approximately 70%. O
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O Each power icvel was maintained for a period of time to perform one or more of the following tests. Preliminary analysis of these measure-ments as specified in the overall controlling test document was iompleted 9 prior to increasing the reactor power to the next stage. . Steam System Performance Tes ts (B-1) . Just prior to steaming and at subsequent power levels during the initial rise to pcwer, data will be accumulated and analyzed on the performance of the steam generators, the tur-bine and the steam plant auxiliaries. Measurements of the turbine _ performance will be made at the lower power levels and the turbine will be loaded at about 28% reactor power. - . Analysis of Chemical Impurities in the Primary Coolant (B-2) . As the reactor power level is increased to about 11% of rated, the core and reac-(h tor internale will experience temperatures in' excess of those reached during the core heatup for reactivity coefficient measurements. At these tempera-tures, additional impurities will be degassed. , Data on the performance of the helium purification system in removing these chemical impurities from the primary coolant will be tak:n and analyzed. PCRV Performance Tests (B-3) . As the reactor power level is in-creased to 28% power, the helium pressure and temperatures approach their quarter load values which results in a system heat load of approximately 80%. At each power level stage up to 28% power and at selected stages up to full power, data uill be taken and analyzed on the performance of the PCRV and its cooling system on the structural response of the PCRV to increased inter-nal pressure and on the primary system helium use rate. () 1 1
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' i Primary Coolant Systen Performance Tes ts (B-4) . At each power l
1evel data on the performance of the helium circulators and their auxiliaries
, will be taken and analyzed. Measurements of the radial power distribution (region peaking factors) will be made at approximately 2%, 5%, and 8% reactor l power. Data on the performance and calibration of the core helium flow ori- l fice valve will be obtained at approximately 28%, 50%, and 100% reactor power. ,
Plant Instrumentation Performance Tests (B-5) . In these tests the !
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performance of the portions of the plant instrumentation which could not be tested prior to power operation will be checked. The nuclear instrumentation ;
. r will be calibrated by means of heat balance measurements and analyses. The calibration of the condensate and feedwater flow instrumentation and the_
core region outlet thermocouples will be checked. The core region outlet () thermocouple test will be performed just prior to the first adjustment of i the helium flow orifices at approximately 8% power and again at approximately r i 100% power. " Plant Transient Performance Tests (B-6) . In these tes.s, the tran- I sient performance of the plant will be tested and analyzed. - The testing will , : a t include: a scram and turbine trip from approximately 28% reactor power with rated steam conditions, a turbine trip from approximately 40% reactor power, , i a main turbine generator load rejection from approximately 60% reactor power ! L l to houpe load,' sequential tripping of the two circulators in a loop from f i approximately 80% reactor power and resultant loop shutdown, and boiler feed pump start and stop transients. . r. e
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O Plant Automatic Control System performnce Tes ts (B-7) . The com-ponents of the autountic control system will be placed into service and tested as the controlled variables come into their controllable range. Dynamic veri- . fication tests of the control system will be performed at selected power levels during the power level increase of the initial approach to full power. .A demonstration of full load change from approximately 100% to approximately 25% turbine load will be made under full automatic control. Reactor Coefficient Measurements (B-8) . Measurements of changes in reactivity will be made during the approach to full power by measuring the change in control rod positions required to produce a core temperature and reactor power level change. _ Differential Control Rod Worth Measurements (B-9) . The reactivity O - worth of contro1 rods which are moved durins che initia1 rise to power v111 be measured using a reactivity computer to obtain the instantaneous reactivi-ty change produced by a control rod motion. Xenon Buildup and Decay Measurements (B-10) . The reactivity change produced by buildup, burnout, or decay of xenon poison follos-ing a pouer level change will be mer.sured by recording the change in the critical control rod positions following a change. Xenon Stability Test (B-ll) . In this test the absence of any sus-tained xenon oscillations is demonstrated. At 100% power a perturbation is produced f rom equilibrium xenon by inserting a control rod in one region and
' withdrawing a control rod in another position. The indicated power level and O
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o Xenon Stability Test (B-ll) (con tinued) . region outlet Eemperatures
- are recorded as a function of time and analyzed for the presence of any oscilla-tion produced by xenon. .
Shielding Surveys (B-12) . At approximately 28% reactor . power and * ,
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approximately 100% reactor power surveys of the radiation levels within the plant are performed. An addit'ional survey is taken during and following any regeneration of the helium purification system. These measured data are re-
, corded and analyzed to demonstrate the adequacy of the shielding design. !
Radiochemical Analysis of the Primary Coolant (B-13) . In this test the radioactive gaseous fission products in the primary coolant will be sam-pled and analyzed. These tests are used in the initial startup phase to de-fine fuel fission product release-to-birth ratio at zero burnup and wilJ yield () information on the fraction of failed fuel particle coatings. This test is performed at each major power level of the initial rise to power. 4 e I L O 1 -
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O - ACKNOWLEDCDENT The contents of this report on the results of B Series Startup testing at Fort St. Vrain, Unit No.1, have been taken from unpublished, in-ternal reports of General Atomic Company and Public Service Company of Colo-rado. J This is an interim report based on preliminary data and therefore i both data and results are subject to change. This report will be supple-mented periodically as further testing is completed. e 4 S I c-e.----rr q ewef --- --- -y-s- v- y - ~ - y
HISTORICAL
SUMMARY
OF PLANT OPERATION At the beginning of the report period, the plant was operating at approximately 65% pcNet. On May 24, 1978, an unplanned automatic Loop 1 shutdown occurred due to a failure of the feedwater flow controller. Operation on one loop at about . 30% reactor power continued until May 26, 1978, when the main turbine generator was shutdown to recover the shutdown loop. The main turbine was re-synchronized on May 28,1978, about 32 hours af ter it was taken off. A test was initiated on June 2,1978, to investigate the primary system temperature fluctuations. This test consisted of partial insertion of a ( number cf control rods to flatten the core power distribution and adjustment of the region helium flow orifice valves to minimize inter-regional variations in differential pressure. On June 3,1978, while making these adjustments at 50% reactor power, temperature fluctuations appeared. Reactor power was reduced to 35%, stopping the fluctuations and the remainder of the orifice adjustments completed on June 4,1978. The test then called for an increase from 40% to 45% reactor power at a rate of 5% per minute in an attempt to initiate fluctu-a tions . Fluctuations were observed which required a reactor power reduction to 34% to terminate. The test demonstrated that a reduction in inter-regional differential pressure did not prevent temperature fluctuations. It is planned to repeat this test with a different orifice valve adju3trant pattern to further evaluate the effect of minimizing inter-regional differential i pressure. The plant operated normally at about 65% reactor power (203 NWe) except as described above, until June 6,1978, when an unplanned automatic shutdown of Loop 2 occurred during surveillance testing. The plant was shutdown at this time and remained shutdown until June 10, 1978, for operator licensing examinations. l l l l l
The plant returned to power operation and the main turbine-generator synchronized to the system at 28% reactor power on June 12, 1978. Plant out-put was then increased to 65%. On June 14, 1978, an unplanned trip of the 1D circulator occurred i during surveillance testing. The shutdown circulator was restarted and the plant returned to 65% power within three hours. A test of the redesigned cold reheat attemperator nozzles was per-formed, with acceptable results. Resolution of attemperator flow control system problems is continuing. ( Slightly higher than acceptable primary system moisture and total oxidant levels were experienced during this period, and have been traced to leakage past the buffer helium dryer, air operated, bypass valve. A modification is planned to provide manual isolation valves and a bypass around the air operated valve to facilitate repair during operation. During the installation of the modification, the leaking valve will be repaired. The plant operated at 57% (170 We) to 67% (212 We) reactor power until June 29, 1978. During this period the plant load was limited by condenser ; vacuum because one of the circulating water pumps was out of service to repair the motor. . On June 29, 1978, power was lost to non-interruptable instrument bus two. This event resulted in a perturbation of the helium circulator auxiliaries, an automatic isolation and steam water dump of the Loop 2 steam generator, and a reactor scram. Two blown fuses were found in the inverter for the failed instru-men t b us . The loop shutdown and steam water dump were caused by high moisture in Loop 2 primary coolant, resulting from the circulator auxiliary system upset. The reactor was taken critical June 30, 1978, and returned to power as rapidly as the moisture concentration in the primary coolant permitted.
6 The turbine generator was resynchronized with the system on July 4,1978, ' at a reactor power of 28%. Plant operation was restricted by the availability of only one helium circulator in each loop. Helium circulator lA was out of service due to an inoperable operator on the inlet steam isolation valve, and circulator 1C in Loop 2 was indicating high buffer-mid-buffer differential pressure with the main drain controller in automatic. The buffer-mid-buffer problem on 1C circulator was cleared by in-creasing the circulator speed. Helium circulator LA inlet steam isolation valve was manually opened and the circulator returned to service on July 6,1978. The reactor power level was increased to 40% (110 MWe) on July 6,1978, and to 52% (162 MRe) on July 8, 1978. On July 7,1978, a localized fire occurred in the turbine building at a hydraulic valve operator that leaked hydraulic fluid on the hot reheat steam piping. The fire was readily extinguished using hand held fire extinguishers. No permanent damage was done by the fire. The reactor power was increased to 58% (175 MWe) on July 10,1978, and operated at this level until July 14, 1978, when an unplanned trip of circulator ID occurred during surveillance testing of the Plant Protective System. Failure s of an integrated circuit chip caused this trip. While trouble shooting to determina the cause of the ID circulator trip, . circulator 1C tripped causing loss of primary coolant flow throught the Loop 2 steam generators. The reactor was manually-scrammed following this occurrence. , The plant was .ceturned to power, the turbine generator resynchronized and power increased to 58% (175 MRe) on July 17, 1978. Two trips of instrument bus inverter -lc occurred during this period, (July 13, 1978, and July 14, 1978) caused by failed silicon controlled rectifiers. .~
, The plant operated between 45% (133 MWe) and 65% (195 MWe) until
. July 31, 1978. Beginning on July 26, 1978, during the evenings, with lowee system power demands, the power level was decreased to 45% to reduce the core outlet
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temperature below 1,200*F. This mode of operation reduces the total time of operation above the 1,200*F, which is the threshold for the 10 ppm total oxidants conditional limit on primary coolant impurity levels. I On Jul" 31, 1978, a winding failure and insulation fire occurred in 1 480 volt transfomr 1A and'the plant was manually scrammed. The loss of the 480 volt transformer caused an upset in d2e buffer helium system for Loop 1, 4
, resulting in ingress of an usknown amount of moisture into the. primary system.
4 The reactor was taken critical on August 5,1978, and operated between 2% and 6% of rated power. Further increase in power was restricted
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by high levels of moisture in the primary coolant. The moisture level in the primary coolant was between 230 ppm and 1 800 ppm. *
- Approximately 60 gallons of water were removed frcm the primary system j during August, 1978. Operations pu,rsued to increased the rate of cleanup of the primary system moisture included
; a) Operation at low power levels to increase the cold gas temperature within the PCRV.
4 b) Cycling of the reactor pressure. c) Operation with various combinations of circulators. . (operation of the 1A circulator increased the primary system moisture level). f d) Increasing the temperature of the PCRV Liner Cooling Sys tem. The reactor power was increased to 25% of rated on August 16, 1978, I resulting in an increase in primary system moisture level from 130 ppm to 190 ppm. The primary system moisture level subsequently decreased to 80 ppm. Reactor power was increased to 29% and the main turbine generator was synchro-nized on August 17, 1978. The turbine was taken of f line ' two hours later and . l the reactor power was reduced to 26% of. rated due to increased primary system moisture.
l i . i The main turbine generator was re-synchronized and loaded to 45 se on August 18, 1978. On August 20, 1978, an automatic reactor scram and steam . water dump of the Loop 2 steam generator occurred due to a noise spike on a high level moisture uenitor with the low level monitors previously tripped. The reactor was returned to 28% power and the main turbine generator reloaded t to 45 We on August 21, 1978. On August 22, 1978, a turbine trip occurred due ! t to an upset in the turbine first stage pressure and the main steam ta=perature , following a system frequency spike. The main turbine generator was back on line about one hour later. T f l I { i f i e e e
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. ., .. TESTING
SUMMARY
SITT B-7 Feedwater Flow Control Both feedwater flow control loops responded adequately to transient step changes on FC-2205 and FC-2206. Some noise was experienced while testing Loop 2. How-ever, the noise was determined not to be due to the controller gain. SUT B-7, Deaerator Level Control The deaerator level tuning consisted of tuning LIC-3175 first then tuning ETC-3175 with both controllers (LIC-3175, FIC-3175) in operation. In each case the system response appears adequate and less than 1/4 amplitude damping. SUT B-7, Feedwater Valve AP Control l
, The action from the feedwater valve AP controllers PDC-22127 and HC-31207 due to i
the transient testing appears to be adequate. The PDC-22127 action is somewhat slow due to its being tuned for loop trips. SUT B-7 Circulator Speed Control The circulator steam turbine speed control system response meets the 1/4 ampli-tude damping criteria.' Very rapid control from the speed controllers was sacri-ficed to prevent small oscillations of the speed valves, and thus increased wear. SUT B-7 Cold Reheat Pressure Ratio Control The previous settings on the two cold reheat bypass ratio controllers PC-2243 and PC-2244 were lef t unchanged due to the satisfactory control system response during
- i l this transient testing series. The step change in the pressure ' ratio on Loop 1 i
l did not cause any changes in circulator speed on Loop 2 and vice versa. This ( indicated excellent decoupling of pressure ratio from circulator speed. l
i l .- . i SUT B-7, Reheat Steam Desuperhea ters Reheat steam bypass to the condenser temperature control could only be achieved for this test when the reheat steam bypass desuperheater controller was adjusted so that its output signal would close the spray valve more than the minimum of , 15%. Temperature setpoint changes were made with a stable response. However, j to insure an adequate spray condition in the reheat steam desuperheaters the spray valves have a minimum,open position of 15%. (The spray valves will never fully close when the reheat bypass valves are open.) Because of this minimum open position, temperature control of this system will probably never occur--due to overspray. Only one of the reheat steam desuperheater controllers was tuned, and the other controller settings were changed to match the one. that was tested. The primary purpose of the reheat steam desuperheaters is to protect the con-denser, and the results of this test indicate that the system operates satis-factorily to perform this function. SUT B-7, Main S team Desuperheaters The main steam desuperheater control system exhibits satisfactory response to transient step changes. Prior to performing this test, and because of previously noted control instability at low steam flow, the valve positions of TCV-5208 and TCV-5207 were changed from a linear characteristic to a square characteristic by FCN-4139. When the steam flow through desuperheaters was reduced, no undamped oscillations were seen. SUT B-7, Main Steam Temperature Control The main steam temperature control system was tuned by making changes in the setpoint of the main steam temperature controllers (TC-2225, TC-2226) one loop at a time. The response of both loops was satisfactory and exhibits less than 1/4 amplitude damping. ,
SUT B-7, Reheat S team Temperature Tes t The response of the reheat steam temperature control system to step changes appears satisfactory. However, quarter amplitude damping response cannot be verified because the seccnd step change was made prior to the first step change stabilizing. Both the reheat steam temperature controller and the flux , controller have a small deadband which prevents excessive control rod movement. Thus due to the deadband the 1/4 amplitude damping criteria is not really applicable to this test. This item is considered closed. SUT B-7, Load Change Response In this test a load change was made (at approximately 1% per minute) from 69% power to 30% power. The plant remained at the 30% power level for one hour to stabilize, then the power was increased at 1% per minute to the previous power level of 68%. All the following plant limits as called out by B-7, Part 20F were acceptable: main steam temperature limits, reheat steam temperature limits, main-to-reheat steam temperature limit and turbine inlet steam pressure and temp-s erature limits. On Both the load decrease and the load increase, feedwater flow exhibited some small oscillations. The main steam and reheat steam appeared to droop the correct amount during the load change, but did not follow the ramp function. The cause for this appears . to be the inherent thermal lag of the plant during a load increase or decrease. The output of the main steam temperature controllers TC-2225 and TC-2226 showed a high response during this thermal lag to increase the mainsteam temperature, but their action on circulator speed is very small, as most of the action is f rom feedwater flow.
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SUT B-9, Control Rod Calibration Additional reasurements of the reactivity worth for control rod Group 4A were' s Performed at positions between 145" to 176" withdrawn. These measurements were consistent with previous measurements performed on this control rod group. The $ total integral worth of control rod Group 4A was measured. 1 9 + 1 1 4 l 4 i 4
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.li . 4 i STEAM SYSTEM PERFURMANCE VERIFICATION (B-1) 4 I This test was not scheduled during the report period. I 1 J 4-4 J d 4 i A I 4 J -t f i ). i 1 .j - b
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PCRV PERMRMANCE TESTS (B-3) This test was not scheduled during the report period. i
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24 - PRIMARY SYSTEM PERFURMANCE (B-4) ! l This test was not scheduled during the report period. f f a l i l i I I 1
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PLANT INSTRUMENT PERFURMED (B-5) This test was not scheduled during the report period. i
PLANT TRANSIENT PERFORMED (B-6) This test was not scheduled during the report period. 4 e f D O 6 t
Startup Test B-7 Plant Automatic Control System Performance Tests Part 2D - Data Feedwater Flow Control Data collected furing the performance of Part 2D is shown in Figures B-7.2D.1 through D-7.2D. 3. Part 3D - Deareator Level Control Data collected during the performance of Part 3D is shown in Figures B-7.3D.1 and D-7.3D. 2. Part 7D - Feedwater Valve Differential Pressure Control Data collected during the performance of Part 7D is shown in Figures B-7. 7D.1 through B-7. 7D.4. Part 8D - Circulator Speed Control Data collected during the performance of Part 8D is shown in Figures B-7.8D.1 through B-7.8D.4. Part 10D - Circulator Pressure Ratio Control Data collected during the performance of Part 10D is shown in Figures B-7.10D.1 through B-7.10D.4. # Part 12C - Reheat Steam Desuperheat Control Data collected furing performance of Part 12C is shown on Figure B-7.12C.l. Part 13A - Main Steam Desuperheater Temperature Control Data collected during performance of Part 13A is shown on Figurcs B-7.12A.1 through B-7.13A.4. Part 14E - Main Steam Temperature Control
- Data collected durios performance of Part 14E is shown on Figures B-7.14E.1 through B-7.14E.8.
Part 15E - Reheat S team Temperature Control - Data collected during performance of Part 15E is shown on Figures B-7.15E.1 through B-7.15E.S. l
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