ML20041F367
| ML20041F367 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/01/1982 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20041F361 | List: |
| References | |
| NUDOCS 8203160433 | |
| Download: ML20041F367 (56) | |
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i ia THE FORT ST. VRAIN A
! INITIAL APPROACH TO 1-j POWER TESTS (B-SERIES) .
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4 i i INTERIM REPORT.'1 i Report for Period Ending November 22, 1981 !
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TABLE OF CONTENTS Page Introduction............................................. ......... 3 8 Series Test Descriptions......................................... 6 Figure 1: Rise-to-Power Testing Sequence ........................ 11 Acknowledgement .................................................. 12 Hi storical Summary of Plant Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 T e s t i n g S u mm a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 Steam System Performance Tests (B-1 Series)
- 1. Steam Generator Steady State Performance (B1-1) ........ 16
- 2. Turbine Generator Steady State Performance (B1-6) . . . . . .16 Chemical Impurities in the Primary Coolant Tests (B-2 Series)
- 3. Primary Coolant System Impuri ties (B2-1) . . . . . . . . . . . . . . . . 17
- 4. Purification Chemical Impurities (B2-2) ................ 17
- 5. Calibration of Gas Chromatograph (B2-3) ................ 18 PCRV Performance Tests (B-3 Series)
- 6. PCRV Liner Cooling B3-10) .............................. 18
- 7. Liner Cooling Maximum Temperatures (B3-1R)...... ....... 18
- 8. Liner Cooling Adjustment (As Required) (83-15).......... 19
- 9. PCRV Internal Temperatures (B3-2R) . . . . . . . . . . . . . . . . . . . . . . 19
- 10. PCRV Data Scan (B3-20).................................. 19
- 11. PCRV Leak Tightnes s (B3-3A) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
- 12. Full Pressure PCRV Leakage (SR 5.2.16a-M)
(As Required) (83-38) .................................. 20 Primary Coolant System Performance Tests (B-4 Series)
- 13. Ci rcul ator Primary Coolant (B4-1(0)) . . . . . . . . . . . . . . . . . . . . 20 Plant Instrumentation Performance Tests (B-5 Series)
- 14. Nuclear Instrument Calibration (BS-1) .................. 21
- 15. Core Region Thermocouple Calibration (B5-2) ............ 22
- 16. Feedwater Flow Calibration (85-3) . . . . . . . . . . . . . . . . . . . . . 22 Radiochemical Analysisdo the Primary Coolant Tests (B-13 Series) .
- 17. Radioactive Gas Analysi s (813-1) . . . . . . . . . . . . . . . . . . . . . . 23
- 18. Iodine Probe Analysi s (B13-2) . . . . . . . . . . . . . . . . . . . . . . . . . . 23 Act i v i ty Schedul e pe r T-164 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
r Individual B Series Test Data 81-1 ......................................... ............. 26 B1-6 ........................................................ 32 B 2- 1, B2- 2 , B 2- 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 B3 10........................................................ 40 B3-1R, 83-1S ................................................ 41 B3 20........................................................ 42 B3-2R........................................................ 45 8 4 - 1 ( 0 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 -
B 5 - 1, 8 5 -2 , 8 5 - 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0 B12-2.......................... ............................. 51 B13-2........................................................ 52 S tatu s o f B-Se ri e s Start-up Te st. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53
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INITIAL APPROACH TO POWER TESTS (8 SERIES STARTUP TESTS)
The initial approach to power is accomplished in a series of discrete power level stages. At each power level, tests are made to measure .
the characteristics of the plant and to ensu o 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 incresse 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. The sequence for the performance of these tests is given in Figure 1, together with the corresponding approximate reactor power levels. The reactor power levels, helium flow rates, feedwater flow rates, steam temperatures, and steam pressure given in the following description of the initial approach-to power may differ somewhat from those in the actual approach to power due to change in test requirements or improvements in operating methods identified during other tests.
In general, the initial approach-to power will be accomplished in the following order:
- 1. Feedwater flow will first be established through both steam generator loops and the bypass flash tank system using a boiler
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_4 feedpump. Helium fiow through the core wil; be provided using one circulator in each loop.
- 2. The reactor power will be increased to approximately 2%.
- 3. The reactor power, feedwater flow, and helium flow rate will be simultaneously increased to 5% power, 20% helium flow, and 25%
feedwater flow using reactor generated steam from the bypass flash tank supplemented by the auxiliary boiler to power the circulating turbines, turbine driven boiler feedpump, and other plant steam requirements.
- 4. The reactor power will then be increased to approximately 8%,
concurrent with an increase in feedwater flow to about 30%. The helium flow will be maintained at about 20% during this power increase. At this condition, 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.
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- 6. The reactor power will be increased to about 18% simultaneously with an increase in helium flow to about 33%, maintaining 25%
feedwater flow, followed by an increase in reactor power to about 26% with a helium flow of 49%. At this condition, the main steam temperature will be about 800 degrees fahrenheit.
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- 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 maintain full steam conditions.
This report covers tests performed between 70% and 100% reactor power.
Each power level was maintained for a period of time to perform one or more of the following tests. Preliminary analysis of these measurements, as specified in the overall controlling test document, was completed prior to increasing the reactor power to the next stage.
r Steam System Performance Tests (B-1)
Just prior to steaming, and at subsequent power levels during the initial rise-to power, data will be accumulated and analyzed on the performance of the steam generators, the turbine, 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 reactor internals will experience temperatures in excess of those reached during the core heat-up for reactivity coefficient measurements. At these temperatures, 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 taken and analyzed.
PCRV Performance Tests (B-3)
As the reactor power level is increased to 28% power, t..a 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 will be taken and analyzed on the performance of the PCRV and its cooling system on the structural response of the PCRV to increased internal pressure and on the primary system helium use rate.
Primary Coolant System Performance Tests (B-4)
At each power level, data on the performance of the helium circulators and their auxiliaries will be taken and ar,alyzed.
Measurements of the radial power distribution (region peaking factors) will be made at approximately 2%, 5%, and 8% reactor power.
Data on the performance and calibration of the core helium flow orifice valve will be obtained at approximately 28%, 50%, and 100%
reactor power. ,
Plant Instrumentation Performance Tests (B-5)
In these tests, the performance of the portiens of the plant instrumentation, which could not be tested prior to power operation, will be checked. The nuclear instrumentation 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 the helium flow orifices at approximately 8% power and again at approximately 100% power.
Plant Transient Performance Tests (B-6)
In these tests, the transient performance of the plant will be tested and analyzed. The testing will include: a scram and turbine trip from approximately 28% reactor power with rated steam conditions, a turbine trip from approximately 40% reactor power, a main turbine generator load rejection from approximately 60% reactor power to house load, sequential tripping of the two circulators in a loop from
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approximately 80% reactor power and resultant icop shutdown, and boiler feedpump start and stop transients.
Plant Automatic Control System Performance Tests (B-7)
The components of the automatic control system will be placed into service and tested as the controlled variables come into their controllable range. Dynamic verification tests of the control system will be perform"ed 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 worth of control rods which are moved during the initial rise-to power will be measured using a reactivity computer to obtain the instantaneous reactivity change produced by a control rod motion.
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Xenon Builduo and Decay Measurements (B-10)
The reactivity change produced by buildup, burnout, or decay of xenon poison following a power level change will be measured by recording the change in the critical control rod positions following a change.
Xenon Stability Test (B-11)
In this test, the absence of any sustained xenon oscillations is demonstrated. At 100% power, a perturbation is produced from equilibrium xenon by inserting a control rod in one region and withdrawing a control rod in another region. The indicated power level and region outlet temperatures are recorded as a function of time and analyzed for the presence of any oscillation produced by xenon.
Shieldino Surveys (B-12)
At approximately 28% reactor power and approximately 100% reactor power, surveys of the radiation levels within the plant are perfo rmed. An additional survey is taken during and following any regeneration of the helium purification system. These measured data are recorded 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 sampled and analyzed. Tnese tests are used in the initial startup phase to define fuel fission product release-to-birth ratio at zero burnup and will 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.
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4 ACKNOWLEDGEMENT ,,,
t IS The contents of. this report on the results of 8 Series Startup. ,
Testing at Fort St. Vrain, Unit No. 1, have been taken from unpublished, internal ' reports of General Atomic Company and Public-Service Company of Colorado.
This is an interim report based on preliminary data; and therefore, both data and results are subject to change. This report will be supplemented periodically as further testing is. completed.
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HISTORICAL
SUMMARY
OF PLANT OPERATION i J
The last testing reported was covered in interim report number 8 for the period ending August 22, 1978. After this, the plant continued operation with the Nuclear Regulatory Commission limit of,70% power.
t In February, 1979, the shutdown for the first refueling was started ,
with the shutdown extending until June, 1979. After startup, normal operation was resumed until September 1, 1979, when the plant was.
shutdown due to inconsistencies found in the safety related piping ,
and hangers. This was found during an audit and analysis committed ,
to by the Company and was reported in Reportable Occurrence 50-267/79-35/01-T-0.
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On September 15, 1979, the reactor was started up and operation /
continued until October 26, 1979, when a maintenance ' shutdown began to install the region constraint devices (RCD). The RCD installation was completed, and the reactor taken critical on December 25, 1979. ,
The RCD's are metal clamps that join the tops of all the regions together to maintain nearly uniform gap flow areas between regions.
Following a reactor scram, 18 helium circulator static seal failed in i
. January,1980, requiring the circulator to be replaced with the spare.
On Februa ry 17, 1980, the reactor was taken critical again, and the
/ Y turbine generator was put o,r. line March 5,1980, c
Operation continued throughout_ the year and into 1981.
On March,16, 1981, Asehdment 23 was issued, which authorized
' fluctuation testing at greater than 70% power. Fluctuation testing Isas-startedonMarch 18, 1981, with power levels of 70% to 88% being l reached during the period April 17 to April 24, and on May 13, 1981.
OnMay13l,1981, the turbine generator tripped on high vibration, due to a loose shroud on the low pressure turbine blading and was follow'ed by a reactor scram. As the turbine repair was to take several weeks, the second refueling outage was started on I May 20, 1981. During this period, 1B circulator was replaced again i
due to another failed static seal.
s The refueling shutdown was finished in July, 1981, and the reactor was taken critical on July 13, 1981. Plant operation has continued throughout the rest of this report period at power levels up to 100%
' and 323 MWe.
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TESTING
SUMMARY
c' This , report covers the period from August 23, 1981, through November 22, 1981, and the analysis of the test results collected during the last period of May 23, 1981, to August 22, 1981.
Due to the installation of the region constraint devices and Nuclear .
Regulatory Commission requirements, the B Series Startup Tests, RT-500K (fluctuation testing at greater than 70% power) and RT-485 (control rod drive temperature data collection) were run concurrently, with RT-500K being the controlling document for testing under normal conditions from 40% to 100% power. The test schedule is called out in T-164 (coordination procedure for testing at greater than 70% power) and is included at the end of the test summary section. ,
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Testing' at greater than 70% power was initiated on four occasions.
The first three (April 17, 1981, April 24, 1981, and May 13, 1981) were to a maximum power level of 88%, at which time turbine vibrations necessitated a plant shutdown and the second refueling outage was begun. However, on November 5, 1981, Rt-500K testing again began, and on November 6, the plant attained 100% power. The plant remained at the 100% power leyel for approximately 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> until shutting down on November 9, 1981, to commence a major system modification outage.
3 Based on tests completed so far, the following conclusions can be drawn concerning plant behavior (for details refer to individual B.
Series Test Reports).
- 1. Steady-state performance above 70Y power was generally satisfactory. However, it was noted that the reheat outlet temperature of module B-2-5 read much lower than the remaining five modules in Loop 2. Although the 20'F range (referred to the average) was not quite met, it is believed that the operating limitations associated with the attendent performance of RT-500 (Fluctuation Testing above 70% Power) resulted in the observed off-normal temperatures. Following normal system operating procedures, expected reheat temperatures can be achieved. Similar situations at lower power levels have been resolved satisfactorily by judicious adjustment of primary t coolant flow orifices and feedwater trim valves. Special
- instrumentation outputs are being analyzed by General Atomic.
Preliminary results appear to be acceptable. Unless subsequent investigation reveals inconsistencies or data outsioe of 'B' i
series acceptance criteria, the Fort St. Vrain technical review
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1 committee considers SUT B1, Part 1, closed. (B1-1) 4 f
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- 2. Steady-state performance of the turbine generator was generally satisfactory. At 100% indicated reactor power, the generator was producing 323 MW(e) versus the expected value of 339 MW(e).
It is postulated that the low feedwater flow may be contributing to the less-than-design output. It is also noted that feedwater 4
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heater No. 5 was not in service, nor was a reheat steam temperature of 1000*F being developed. The reheat steam temperature was restricte'd to less than 1000*F due to limitations imposed by the RT-500 fluctuation testing. SUT B1, Part 6 is considered closed. Further investigation into possible feedwater flow discrepancies will be addressed as part of an analysis associated with the results of SUT B5, Part 3,
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Feedwater Flow Calibration (B1-6)
- 3. Prima ry coolant impurity levels monitored at 100% power were sati sf actory. The hydrogen getter was not in service; it is being modified by.a change notice. Hydrogen data was not available due to gas chromatograph equipment outage. Primary coolant impurities are monitored on a routine basis as part of normal plant operation. No unresolved issues remain from this startup test. SUT B2, Part 1 is closed. (B2-1)
- 4. Operation of the purification trains over the past several years have indicated satisfactory performance on a routine basis.
Train operation at 100% appears to be satisfactory in general, despite the fact that data recorded during the 100% power run indicated that the 'B' train (in-service for approximately three months) was showing signs of impurity break-through. However, overall train performance was satisfactory; SUT B2, Part 2 is considered closed. (B2-2)
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- 5. Gas chromatograph operation and periodic calibration are routine operatons. Results of calibrations performed prior to 100%
powe'r operation were consistent with previous calibrations.
SUT 82, Part 3 is considered closed. (B2-3)
- 6. Total liner cooling heat load was determined to be between 14 and 17 million BTU /hr. This is consistent with expected values.
Data analysis indicates that the turbine flow meter FE-46165 which measures flow to subheader 4T (PCRV top penetrations, Loop 2) may be indicating a higher flow rate than actually exists. Public Service Company Results Department will investigate. It is also possible that the individual flow meters for the high temperature filter absorber units are out of calibration as evidenced by inconsistent heat loads for the 1
separate cooling tubes. Investigation here is also unde rway.
Based on the results of data collection at 100% power, SUT B3, Part 1Q is considered closed. However, routine system analysis of system performance will be carried out on a regular basis.
(B3-10)
- 7. Hot spot tube temperature rises were consistent with expected values published in Public Service Company's letter to the Nuclear Regulatory Commission, P-78037. The entire liner cooling system will be monitored on a regular routine basis to assure that 'no significant changes have occurred within either the system or the PCRV. The individual tubes wrapped around the
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o Region 5 top penetration show markedly different temperature rises (21*F versus 9'F). This will be checked. SUT B3, Part IR is considered closed. (83-1R)
- 8. No subheader or tube valve positions required adjustment during the 100% power run. Subheader 2BS (46248) on the bottom head system should have its indicated flow increased from 109 to 145 gpm to bring the overall system temperature rises into better agreement. SUT B3, Part IS is considered closed.
(B3-15)
- 9. Two sets of data were collected while operating at 100% power.
The results differed by 30 to 50*F. The lower temperature data set was consistent with previous values collected at lower power. The second set more closely reflected expected temperatures. This data will continue to be collected at regular intervals to establish historical reference information and possible trends. SUT B3, Part 2R is considered closed.
(83-2R)
- 10. Information collected from the PCRV Data Acquisition System has been anhlyzed and found to be valid. Due to the lack of recent zero (low) PCRV Internal Pressure data, load cell responses could not be reconciled to the acceptance criteria. Public Service Company will, however, provide additional data to
General. Atomic Company for further analysis. The analytical results will be evaluated by Public Service Company. Two instrument channels (100 and 247) will be checked for discrepant response. SUT B3, Part 20 is considered closed pending the above load cell information. However, regular routine data analysis from the PCRV Data Acquisition System will be continued per Public Service Company's established procedures. (B3-20)
- 11. Helium leakage from the PCRV was negligible. Helium leakage from the PCRV auxiliary piping system was consistent with values observed over the past several years. A substantial purified helium leak into the reheat steam system by way of the Loop 2 steam generator interspaces is known to exist. Overall results of this test were satisfactory. SUT B3, Part 3A is considered closed. Monitoring of plant wide helium leakage will continue on a routine basis. (B3-3A)
- 12. The data for this test is routinely collected via Public Service Company Surveillance Test 5.2.lba-M, and was performed on November 7, 1981, at 100% power. No discrepant responses ere observed. SUT B3, Part 38 is considered closed. (B3-38)
- 13. The reactor / primary coolant system was being operated under conditions specified by RT-500 (fluctuation testing). As such, primary system conditions were different than those anticipated
by the SVT acceptance criteria. Primary coolant flow was about 10 to 13% greater than expected based on AP information from sensing points at the circulator inlets. Overall circulator AP was less than expected due to flow resistance values established by RT-500. Continuous operation at 100% reactor power was determined to be satisfactory, and the Fort St. Vrain technical 1 review committee judged that the system performance could be adjusted to meet the desired acceptance criteria if the RT-500 restrictions had not been governing. SUT 84, Part 1 is ,
considered closed. (B4-1(0))
- 14. Indicated reactor power equals 856 MW(t) (=102%). Actual reactor power depends on accurate measurement of feedwater flow and heat balance calculations plus data logger information.
Some of these data yield power levels of 814 MW(t) (=96.5).
Further effort to refine and otherwise produce more accurate power calculations is needed, especially in the area of feedwater flow accuracies. (See SUT B5, Part 3.) Control rod patterns expected at 100% steady state power were generally achieved. The observed reactivity discrepancy was 0.003aP.
There is a known bias of 0.003AP in the Base Reactivity Curve.
The bias will be corrected when the new Base Reactivity Curve is approved by the Nuclear Facility Safety Committee. This correction will result in a reactivity discrepancy that is essentially zero. SR-5.1.4-W-P testing confirmed the existence of a proper control rod pattern. (B5-1)
- 15. A broad range of results were observed during SUT B5, Part 2.
The nature of this instrumentation, plus existing recognized
- discrepant problems, all contribute to the uncertainties associated with data collected for this SVT. The recorded results were satisfactory, and further indicated that additional con firmatory information be obtained from reduction of fluctuation monitoring data. The Fort St. Vrain Technical Committee further established that acceptance criteria established for this test are believed to be too restrictive, and that routine annual surveillance tests performed recently indicate satisfactory results. Previous core outlet thermocouple traverses recorded in tests RT-509 and RT-524 will be reviewed to confirm adherence with the requirements of LCO 4.1.9. SUT B5, Part 2 is considered closed. (B5-2)
- 16. The results of SVT B5, Part 3, have concluded that further effort is needed to resolve apparent discrepancies. The SUT itself is declared closed, in that the 100% data wss satisfactorily recorded and the analysis successfully concluded.
The analytical results however, indicate that a more detailed plan must be developed to determine the predominant variables and ranges which contribute to the end result. A Public Service Company T test will be prepared and issued to continue data collection and analysis as required to ultimately resolve the feedwater flowrate issue. General Atomic, San Diego, has advised that certain sensitivity calculations are currently in 4
progress, the results of which may assist Public Service Company
in developing their approach to resolution of feedwater flow discrepancies. (85-3)
- 17. No fast gas samples were obtained at 100% power due to flow blockage in the sample line. With respect to activity values and radiation background values, recent data indicate a level of about 1/10th of the ar.ticipated values. Operation of the purification system has been fully satisfactory in removing radioactive gases as evidenced by purification train outlet samples. This SUT is considered closed, however, further analyses of the primary coolant will be obtained via routine surveillance programs augmented by Public Service Company T-test performance as required. (B13-1)
- 18. The iodine analysis test was in progress at the time of the 0155, November 9, 1981 plant shutdown, and no complete data was collected. This test will be repeated at 100% power at the next opportunity. This test is considered open. (13-2)
, , ,-f-' PUBLIC SERVICE COMPANY OF COLORADO FCET ST. VRAIN NUC'. EAR GENilAflNG STAT!dN I-164 E
E h'-dN > Page 9 of 20 STEADY STATE 3-0 CATA COLLECTICN" APPROXIP. ATE PC' DER LEVELS Sui et,LE ~50 ~70 '30 '90 'ICO 31-1 Steam Genera::r Steady State 0 0 0 0 R.
31-3 Steam Genera::r Steady State (Data with Plugged Tuce) R 31-4 Steam Generater Steacy Sta a (HP Feecwater Heater Sy;assec) R 31-6 Turbine Genera:ce Steady State Performance 0 0 0 0 R 32-1 Primary Cociant Systam m:urities 0 0 R C R 32-2 Purification Chemical Impurities 0 0 R 0 R 32-3 Calibration of Gas Chroma:: graph O O R 0 R 33-1R Liner Cccling Maximum Temperatures G 0 R 0 R 33-15 Liner Cccling A0J (As Required) - - - - -
33-10 PCRV Liner C cling 0 0 0 0 3 33-20 PCRV Oa:a Scan 0 0 0- 0 R 33-2R PCRV Internal Tem:eratures 0 0 R 0 R 33-3A PCRV Leak Tightness 0 0 R 0 R 33-33 Full Pressure PCRV Leakage SR 5.2.15a-M (as re;utrec) - - - -
R 34-1(0) Circulator Primar-/ Coclant 0 0 0 0 R 35-1 Nuclear Instrument Calibratica 0 0 0 0 R 35-2 Core Region TC Calibration - - - -
R 35-3 Peecwater Flew Calibra:!cn 0 0 0 0 R 3
311 Xenon Stability Tes:
312-2 Shielding Survey 0 0 0 0 R 313-1 Radicactive Gas Analysis 0 0 0 0 R 312-2 I: dine Pr be Analysis - - - -
R
- RT .35 C ntr:1 Red Drive :nternal Tem:eratures 0 - - - -
- (Jata to be taken at 6% Intervals Between 70 anc 10C%)
RP29C-M Genera::r Temperature Monit: ring 0 0 0 0 0 "0" 0::icnal Casa Taking "R" Required Casa Taking ,
'NCTE: Per further de: ails en incivicual 3-0 steady state da:a taking frecuency see : ages 10-15.
FIGURE 1
-25 STEAM SYSTEM PERFORMANCE VERIFICATION (B-1)
Data
- 1. Typical steam generator temperature performance data taken at 100*.' is shown in Table B-1.1 (page 26), and on pages 27 through 30,
- 2. Typical main turbine performance data taken at 100*. reactor thermal power (100*.' turbine load) are shown in Table B-1.2 (page 31), and on pages 32 through 36.
TABLE B-1.1 STEAM GENERATOR PERFORMANCE DATA
- Average Cold Reheat Inlet Helium Steam Temp- Reheat Steam Main Steam Module Temperature erature 'F Temaerature Temoerature B-1-1 1376 545 995 992 B-1-2 1348 537 992 993 (1)
B-1-3 1375 540 1003 997 B-1-4 1350 545 995 999 (1)
B-1-5 1364 540 991 990 B-1-6 1367 540 998 989 B-2-1 1317 552 979 988 B-2-2 1379 553 990 990 B-2-3 1384 553 987 991 B-2-4 1398 555 1000 990 (2)
B-2-5 1372 555 970 989 (3)
B-2-6 1397 552 1000 988 (2)
(1) Low helium temperature to B-1-2 and B-1-4.
(2) High helium temperature to B-2-4 and B-2-6.
(3) Possible high attemperation flow to B-2-5.
Low hot reheat steam at module outlet.
All values rounded to nearest whole unit.
l j
INSTRUMENT DATA Linear Reactor Power Level 100%
Wide Range Reactor Power Level 100%+
Helium Pressure 680 PSIA Helium Pressure 670 PSIA Helium Pressure 690 PSIA Average Circulator Inlet Temperature 717'F Circulator Speed, IA 9270 RPM Circulator Speed, 1B 9255 RPM Circulator Speed, 1C 9201 RPM Circulator Speed, 10 9240 RPM Circulator Helium Flow,1A 109%
Circulator Helium Flow,1B 109%
Circulator Helium Flow, 1C 111%
Circulator Helium Flow, 10 108%
Circulator Helium Differential Pressure, IA 8.0 PSI Circulator Helium Differential Pressure, 1B 7.6 PSI Circulator Helium Differential Pressure, 1C 7.8 PSI Circulator Helium Differential Pressure, 10 8.4 PSI Feedwater Temperature, Loop 1 380'F Feedwater Temperature, Loop 2 390'F Feedwater Flow, Loop 1 1,074,000 lbs/hr Feedwater Flow, Loop 2 1,091,200 lbs/hr -
Cold Reheat Pressure Before Circulators, Loop 1 870 PSIG Cold Reheat Pressure Before Circulators, Loop 2 870 PSIG Cold Reheat Pressure After Circulators, Loop 1 595 PSIG
INSTRUMENT DATA Cold Reheat Pressure After Circulators, Loop 2 595 PSIG Main Steam Pressure, Loop 1 2500 PSIG Main Steam Pressure, Loop 2 2450 PSIG Main Steam Temperature, Loop 1 999'F Main Steam Temperature, Loop 2 993'F Turbine First Stage Pressure 1500 PSIG Turbine Throttle Pressure 2420 PSIG Hot Reheat Steam Pressure, Loop 1 545 PSIG Hot Reheat Steam Pressure, Loop 2 545 PSIG Desuperheater Water Flow, Loop 1 76,000 lbs/hr Desuperheater Water Flow, Loop 2 72,000 lbs/hr Feedwater Valve Differential Pressure, Loop 1 100 PSID Feedwater Valve Differential Pressure, Loop 2 100 PSID Hot Reheat Pressure 530 PSIG Hot Reheat Temperature at Turbine Generator Inlet 970'F Steam Chest Pressure 2360 PSIG Flash Tank Pressure 880 PSIG Turbine Generator Load 324 MW Module Feedwater Flow, Channel 1 81.5%
Module Feedwater Flow, Channel 2 80.0%
Module Feedwater Flow, Channel 3 84.5%
Module Feedwater Flow, Channel 4 80.5%
Module Feedwater Flow, Channel 5 83.6%
Module Feedwater Flow, Channel 6 83.5%
Module Feedwater Flow, Channel 7 85.0%
INSTRUMENT DATA Module Feedwater Flow, Channel 8 80.7%
Module Feedwater Flow, Channel 9 82.0%
Module Feedwater Flow, Channel 10 85.0%
Module Feedwater Flow, Channel 11 80.3%
Module Feedwater Flow, Channel 12 88.0%
Module Main Steam Temperature, Channel 1 992'F Module Main Steam Temperature, Channel 2 993'F Module Main Steam Temperature, Channel 3 997'F Module Main Steam Temperature, Channel 4 999'F Module Main Steam Temperature, Channel 5 990'F Module Main Steam Temperature, Channel 6 989'F Module Main Steam Temperature, Channel 7 988'F Module Main Steam Temperature, Channel 8 990'F Module Main Steam Temperature, Channel 9 991'F Module Main Steam Temperature, Channel 10 990'F Module Main Steam Temperature, Channel 11 989'F Module Main Steam Temperature, Channel 12 988'F Circulator Outlet Steam Temperature, Loop 1 670'F Circulator Outlet Steam Temperature, Loop 2 675'F Reheater Inlet Temperature, Channel 1 545'F Reheater Inlet Temperature, Channel 2 537'F Reheater Inlet Temperature, Channel 3 540'F R heater Inlet Temperature, Channel 4 545'F Reheater Inlet Temperature, Channel 5 540'F Reheater Inlet Temperature, Channel 6 540'F
2NSTRUMENT DATA f Reheater Inlet Temperature, Channel 7 552'F Reheater Inlet Temperature, Channel 8 553'r Reheater Inlet Temperature, Cha9nel 9 553'F Reheater Inlet Temperature, Channel 10 555'F Reheater Inlet Temperature, Channel 11 555'F Reheater Inlet Temperature, Channel 12 552'F j
Rehaater Outlet Temperature, Channel 1 995'F Reheater Outlet Temperature, Channel 2 992'F Reheater Outlet Temperature, Channel 3 1003'F Reheater Outlet Temperature, Channel 4 995'F Reheater Outlet Temperature, Channel 5 991'F Reheater Outlet Temperature, Channel 6 998'F Reheater Outlet Temperature, Channel 7 979'F Reheater Outlet Temperature, Channel 8 990'F Reheater Outlet Temperature, Channel 9 987'F Reheater Outlet Temperature, Channel 10 1000*F Reheater Outlet Temperature, Channel 11 970'F Reheater Outlet Temperature, Channel 12 1000'F
TABLE B-1.2 MAIN TURBINE GENER/. TOR PERFORMANCE DATA
- Item Predicted Values Measured Data Load MW(e) 339 323 Main Steam Temperature 1,000 984 Reheat Steam Temperature 1,000 969 First Stage Pressure, PSIG 1,652 1,522 Main Steam Pressure 2,412 2,412 Cold Reheat Pressure 679 468 Feedwater Temperature 409 396 Condenser Pressure 2.5 2.45 Turbine Speed 3,600 3,600 ,
l Attemperation Flow 121,913 148,000 l
Maximum Vibration (Pt. 5) ---------
4.0 Mils All values rounded to nearest whole unit.
INSTRUMENT DATA Load 325 MW Number of alours at Load 8.75 Hours Reactive KVA 36 Mega Vars Hydraulic Piston Stroke (Control Valves) 100%
Hydraulic Piston Stroke (Control Valves) 100%
Hydraulic Piston Stroke (Control Valves) 69%
Hydraulic Piston Stroke (Control Valves) -
10%
9 Steam Chest Pressure 2400 PSIG Main Steam Temperature 984'F First Stage Pressure 1510 PSIG High Pressure Exhaust to Reheater 850 PSIG High Pressure Exhaust to Reheater 730'F Reheat Bowl Pressure 510 PSIG Hot Reheat Temperature 969'F Hot Reheat Steam to IP Turbine Pressure 530 PSIG Cold Reheat Steam PT #14 468'F Fifteenth Stage Pressure, Heater 1 7.2 PSIA
, Fourteenth Stage Pressure, Heater 2 12.5 PSIA Thirteenth Stage Pressure, Heater 3 24.0 PSIA Eleventh Stage Pressure, Heater 4 78 PSIA ,
Tenth Stage Pressure, Heater 5 148 PSIG Eighth Stage Pressure, Heater 6 238 PSIG Feedwater Temperature Leaving Top Heater 396 F Exhaust Pressure 23.13 Hg Barometer 24.98 Hg I
1 j
INSTRUMENT OATA I
Steam Flow 2.0 x 10' lbs/hr Condensate Flow 2.1 x 10' lbs/hr Quantity of Makeup Demineralizer Water 50 GPM Feedwater Heaters in Service All But #5 Quantit'y of A ttemperation to Reheat Loop 2 72 x 108 lbs/hr Quantity of Attemperation to Reheat Loop 1 76 x 102 lbs/hr Main Steam Pressure -
2500 PSIG Main Steam Pressure 2490 PSIG Operating Oil Pressure 225 Bearing Header Pressure 26.5 Main Pump Suction Pressure 23 Turbine Speed 3600 RPM Armature Current 8600 Amps Armature Current 8750 Amps Armature Current 8500 Amps Armature Voltage 21.7 KV Field Current 2100 Amps Field Voltage 245 Volts Field Temperature 64'C Hydrogen Pressure 19.5 PSIG Hydrogen Purity 97%
Differential Fan Pressure, Water 3.4 H 2O Exhaust Hood Temperature 110*F Exhaust Hood Temperature 112 F Cold Gas Temperature 32.7'C
~
INSTRUMENT DATA Cold Gas Temperature 35.0'C Cold Gas Temperature 30.5'C Cold Gas Temperature 34.0'C Stator Temperature, Liquid Out 56.5'C Stator Temperature, Liquid In 42.0'C Stator Temperature, Machine Gas Temperature 36'C .
Exciter Air In Temperature 28.6'C Exciter Air Dut Temperature 41.5'C Collector Air Out Temperature 41.5'C Turbine Temperature and Differential Temperature Data Point 1 990'F Turbine Temperature and Differential Temperature Data Point 2 940*F Turbine Temperature and Differential Temperature Data Point 3 980'F Turbine Temperature and Differential Temperature Data Point 4 930'F Turbine Temperature and Differential Temperature Data Point 5 880'F Turbine Temperature and Differential Temperature Data Point 6 780'F Turbine Temperature and Differential Temperature Data Point 7 960'F Turbine Temperature and Differential Temperature Data Point 8 835'F Turbine Temperature and Differential Temperature Data Point 9 130'F Turbine Temperature and Differential Temperature Data Point 10 460'F iurbine Temperature and Differential Temperature Data Point 11 230'F
INSTRUMENT DATA Turbine Temperature and Differential Temperature Data Point 12 140'F Turbine Temperature and Differential Temperature Data Point 13 730'F Turbine Temperature and Differential Temperature Data Point 14 730'F Turbine Temperature and Differgntial Temperature Data Point 15 660'F Turbine Temperature and Differential Temperature Data Point 16 660'F Turbine Temperature and Differential Temperature Data Point 17 845'F Turbine Temperature and Differential Temperature Data Point 18 820*F Turbine Temperature and Differential Temperature Data Point 19 130'F Turbine Temperature and Differential Temperature Data Point 20 10'F Turbine Temperature and Differential Temperature Data Point 21 0'F Turbine Temperature and Differential Temperature Data Point 22 60'F Turbine Vibration Data Point 1 0.0 Mils Turbine Vibration Data Point 2 0.0 Mils Turbine Vibration Data Point 3 3.0 Mils Turbine Vibration Data Point 4 0.75 Mils Turbine Vibration Data Point 5 4.0 Mils Turbine Vibration Data Point 6 1.2 Mils
i INSTRUMENT DATA l
i t J
Turbine Vibration "A", #1, Data Point 7 1.0 Mils Turbine Vibration "A", #2, Data Point 8 -1.2 Mils Turbine Vibration "C", #1, Data Point 9 0.75 Mils
{ ~
Turbine, Vibration "C", #2, Data Point 10 2.1 Mils-l Turbine Vibration Data Point 11 0.0 Mils !
i 1
l
-J
m CHEMICAL IMPURITIES IN THE PRIMARY COOLANT (B-2)
/
Data Table B-2.1, page 38, shows the data collected during operation at 100% power.
~
TABLE B-2.I (1) CORE OUTLET REACTOR 02 CO CO2 11 Cil4 N2 HO 2
DATE TIME TEMPERATURE *F POWER % PPMV PPMV PPMV PPkV PPMV PPMV PPMV 11-6-81 1800 1401 100 (2) 14.0 1.6 (2) 0.4 42.2 0.0 2000 1403 100 15.6 2.9 0.4 37.8 STD 2200 1405 100 14.6 2.5 0.4 00S 0.0_
11-7-81 0000 1413 100 15.0 2.3 0.3 40.0 0.0 0200 1414 100 14.8 2.5 0.3 39.2 OOS 0400 1414 100 16.6 2.0 0.4 43.9 00S 0600 1400 100 22.3 2.5 0.4 42.9 OOS 0900 1404 100 11.0 2.1 0.4 33.3 STD 1100 1413 100 10.8 2.2 0.3 33.3 0.0 1400 1412 100 13.3 2.3 0.4 71.7 0.0 1800 1414 100 13.8 2.3 0.4 46.7 0.0 2200 1416 100 13.9 2.5 0.4 46.7 0.0 11-8-81 0200 1422 100 12.0 2.2 0.4 33.3 00S 0600 1415 100 15.1 2.3 0.5 35.8 OOS 1100 1422 100 13.4 2.2 0.4 28.2 STD 1400 1420 100 10.7 2.3 0.4 24.3 OOS 1800 1421 100 14.4 2.5 0.4 29.4 00S 2200 1427 100 12.7 2.5 0.4 28.2 STD (1) Data times listed to nearest hour (2) Out of Service - No data taken ,
39 PCRV PERFORMANCE TESTS (B-3)
The heat loads for each subsection of the PCRV at 100% power are shown in Table B-3.1, page 40. The design allowable heat loads at 100 % power are also shown. (SUT-B3-10)
The " hot spots" noted on the PCRV liner / concrete are listed in Table B-3.2, page 41. (SUT-B3-IR,83-1S)
The PCRV data scan provided temperature, pressure, and strain values.
This data is listed in Table B-3.3, pages 42 through 44. (SUT-83-20)
The internal temperature data for the PCRV are listed in Tables B-3.4A and B-3.48, pages 45 and 46. (SVT-E3-2R)
J
TABLE B-3.1 HEAT LOAD (X 10' BTU /HR) 80*; POWER DESIGN ALLOWABLE PCRV SUBSECTION LOOP 1 LOOP 2 AT 100f.
Top Penetrations 0.97 0.98 2.270 Core Support Floor Top 1.61 1.51 ------
Core Support Floor Side 0.29 0.57 ------ .
_ Core Support Floor Bottom 1.23 1.12 ------
Total Core Support Floor 3.13 3.20 8.076 Upper Barrel and Top Head 1.44 1.44 6.292 Lower Barrel 1.25 1.19 3.350 Bottom Head and Bottom Penetrations 1.76 1.73 4.887 Total by Loop 8.55 8.54 12.435 TOTAL PCRV 17.09 24.87
r TABLE B-3.2 PCRV LINER / CONCRETE HOT Sp0TS Local Maximum Concrete Temperature Effective Concrete at the Temperature at the Liner Cooling Liner / Concrete Liner / Concrete Location / Area Tube AT Interface - F Interface *F Top Head 20 F @ 55% 201 F @ 100% 128 F @ 100% ,
Penetrations 30*F @ 100%
Core Outlet 20*F @ 80% 200*F @ 20% 120*F.0 100%
Thermocouple 21 F @ 100% 346 F @ 100%
Penetrations Core Barrel 40 F @ 60%2 250*F @ 27%2 120 F @ 100%
Seal Area 47*F @ 100% 326*F @ 100%2 Peripheral 20*F 0 40% 200*F @ 100% 120*F @ 100%
Seal Area 28 F @ 100%
Loop Divider 15 F @ 100% 200*F @ 20% '.20*F 0 100%
Baffle Area 354*F @ 100%
Steam 16 F @ 100% 200 F @ 40% 120 F @ 100%
Generator 220 F @ 100%
Penetrations Cross Over 3 F @ 100% 200*F @ 11% 110*F @ 100%
_ Pipe 330*F @ 100%
FSAR Stated 20 F2 150 Normal 130*F Value 200 Normal Hot Spot 2 2 Amendment 29 established 40*F cooling tube AT, 250*F local maximum concrete temperature at the liner / concrete interface for the core support floor. Amendment 29 also allows under certain circumstances, a local maximum concrete temperature of 360 F.
.m u
TABLE B-3.3 Temperatures Channel Thermocouple Temperature (*F) 144 TE-1170-28 114 535 TE-1170-17 115 l 432 TE-1170-46 -
114 444 TE-1170-47 109 445 TE-1170-48 110 414 TE-1170-65 113 1
^
416 TE-1170-54 114 418 TE-1170-56 118 L _ 422 TE-1170-78 121 r .i 506 TE-1170-72 116 f -
- f. '
l.031 TE-1170-127 111 i )
128 TE-1170-96 119 325 TE-1170-100 113 235 TE-1170-110 121
~
l 030 TE-1170-109 110
-j '- 343 TE-1170-111 121 331 TE-1170-114 101
,,/ !.
335 TE-1170-107 105
- w *
- r
+
43-TABLE B-3.3 (CONTINUED)
Internal Vessel Pressures Channel Load Cell Load (KIPS) 2 VM-3 1300.6 1 VI-10 1266.2 0 VM-17 1321.8 3 -
VI-24 1309.9 4 VM-31 1279.7 100 VI-38
- 101 TOR-U2 1276.3 102 TIR-M1 1076.8 103 BOR-L4 1359.4 j 104 BIR-M4 1226.8 200 C0-1.6 1246.0 201 CI-1.4 1259.5 202 CO-2.2 1290.4
_ 203 CO-2.1 1234.1 204 CO-3.2 1362.3 300 CO-3.1 1204.3
[
301 CI-7.4 1094.4 302 CO-8.3 1144.3 303 CI-10.1 1102.6 304 ;,, _. CO-11.2 1107.2 i
400 i CI-15.3 1138.2 Bad reating - at of range.
TABLE B-3.3 (CONTINUED)
Internal Vessel Pressures, Continued Channel Load Cell Load (KIPS) 401 CO-17.6 1188.2 402 CO-17.5 1210.8 403 CO-17.4 1281.6
- 404 CO-17.3 1225.4 500 CI-18.5 1221.2 501 CO-19.6 1355.3 Strain Channel Strain Gauge Strats 150 YE-1170-43 704.4 458 YE-1170-58 -30.7 247 YE-1170-74 742.5**
280 YE-1170-63 -361.8 366 YE-11205-16 -40.5
- Questionable reading.
M
L 1
e r .
u 9 5 6 6 8 8 3 0 8 8 1 3 5 6 7 1 5 t t 1 9 0 5 3 0 ea 2 2 5 1 1 1 2 2 4 1 7 4 0 7 1 3 4 5 5 4 4 4 4 4 4 2 1 1 2 3 2 4 l r 0 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 t e 1 up Om e
T e e cn 2 9 7 3 6 6 9 4 5 7 2 6 4 1 0 1 2 0 g i o . . d fi 2 5 6 2 r 5 5 3 6 0 8 8 0 5 9 6 5 3 6 i t ri 1 3 2 1 3 4 1 3 6 1 2 2 3 2 3 4 4 3 e v
Os A o
P d
o R 9 6 3 8 7 8 9 9 0
) 1 9 3 9 6 1 9 5 4 l . . . .
0 0 9 6 0 8 0 1 on 0 9
6 4
0 9
0 9
0 9
0 9
0 9
8 4 9 9 1
9 1
9 8 4 9 8 9 9 ri 1 1 1 1 1 1 1 1 1 1 1 t( 1 1 1 1 1 1 1 n
o C
n o 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 i 0 1 2 3 3 3 3 3 3 3 3 g 2 2 2 2 2 2 2 2 2 e
R A
4 F
3 *
- e B r u 5 6 8 7 8 7 7 3 6 7 9 8 e t t 0 4 5 4 6 9 1 3 6 0 l ea 7 7 7 1 0 9 5 1 2 0 8 1 2 6 1 1 l r 3 4 4 5 5 4 4 5 4 5 3 5 4 4 4 5 4 4 4 b 1 1 1 1 1 1 1 1 1 1 1 1 1 a t e up 1 1 1 1 1 1 T
Om e
T en 0 6 4 0 9 2 1 3 5 8 3 5 7 co 5 7 5 0 3 2 .
ii .
0 7 4 9 f t 6 4 5 8 2 5 4 5 8 1 6 3 0 2 1 2 2 7 6 2 2 5 1 5 2 3 2 4 3 2 2 2 1 ii 2 rs O o P
d o
R 3 0 6 0 7 0 6 0 2 0 7 0
) 2 9 0 3 1 8 1 l . 0 0 9 0 on 5 9 0 0 9 0 1 0 0 9 0 0 0 0 0 2 8 9 9 8 9 9 9 8 9 9 9 8 ri 1 1 1 1 1 1 1 1 t( 1 1 1 1 1 n
o C
n o 5 6 7 8 9 0 1 2 3 4 5 6 7 8 9 i 1 2 3 4 1 1 1 1 1 1 1 g 1 1 1 e
R e 1 m 8 i - 5 T 7 3
/ 0 8 e - 0 t 1 a 1 D
F e
r u
t t 1 9 4 9 8 9 1 2 4 3 2 6 0 5 8 3 7 8 2 ea 4 9 9 2 1 1 6 0 7 2 1 6 5 2
9 2 2 2
5 3
4 2
2 4
l r 0 4 3 4 5 5 4 5 4 4 5 4 1 1 1 1 1 1 1 1 1 t e 1 1 1 1 1 1 1 1 1 1 1 1 up Om e
T e e g
cn i o 2 2 8 4 5 4 9 7 4 7 7 6 5 2 1 0 4 9 a f i . r it 5 8 3 6 2 4 8 9 5 9 4 5 3 0 2 5 1 1 e ri 1 3 2 1 3 5 1 2 6 1 2 2 3 3 3 4 5 3 v
- Os A o
P d
o R
) 3 2 4 7 2 8 7 2 5 8 4 9 8 8 8 1 7 9 l . .
on 0 7 0 0 0 9 0 9 0 0 1 0 9 7 0 9 0 0 ri 9 6 9 9 9 8 9 6 9 9 9 9 8 6 9 8 9 9 t( 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 n
o C
n o
i 0 1 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 g 2 2 2 2 2 2 2 2 2 2 3 3 3 3 3 3 3 3 e
R F
B
- 4 e r
3 t u
- et 8 5 5 1 9 7 9 6 5 6 4 0 6 5 2 6 6 3 6 B l a 5 7 7 1 9 9 4 1 4 1 5 2 3 7 4 1 5 8 5 t r 3 4 4 5 4 4 4 5 4 5 4 5 4 4 4 5 4 4 4 e ue 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
_ l O p b m
_ a e T T e
cn 3 9 i o 5 7 6 0 2 1 0 8 3 0 7 0 1 3 2 0 4
_ fi . .
i t 6 4 5 8 2 5 4 8 8 1 1 5 0 2 8 9 1 4 7 ri 2 2 2 7 6 2 2 6 1 5 2 3 2 4 2 2 2 2 1 Os o
P d
o R
) 8 9 8 4 0 4 3 3 9 6 0 7 8 9 0 0 5 5 0 l .
on ri 0
2 9
8 9
8 0
9 9
8 0
9 1
9 0
9 2 9 8
0 0 9
0 0 9
0 0 9
2 9 8
0
_ t( 1 1 1 1 1 1 1 1 1 1 1 1 1
_ n o
C n
o i 1 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 8 9 g 1 1 1 1 1 1 1 1 1 1 e
R e 1
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v.
o PRIMARY SYSTEM PERFORMANCE (B-4) p Data o Table B-4.1, page 48, shows the data collected at 100*.' power for all four helium circulators.
TABLE B-4.1 .
p
)4 Reactor Power 100%
~
HELIUM CIRCULATOR 1A XXXXX Speed (RPM) 9,354 Flow (%) 112%
a/P (PSID) 7.2 Diffuser A/P (PSID) 6.7 Compressor A/P (PSID) 8.3 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.54 HELIUM CIRCULATOR IB XXXXX Speed (RPM) 9,315 Flow (%) 112%
a/P (PSID) 9.5 Diffuser A/P (PSID) 5.3 Compressor A/P (PSID) 7.8 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.54 f
4
_s
TABLE B-4.1 (CONTINUED) lE Reactor Power 100%
HELIUM CIRCULATOR IC XXXXX Speed (RPM) 9,386 Flow (%) 113%
a/P (PSID) 9.3 Diffuser A/P (PSID) 5.7 Compressor A/P (PSID) 7.8 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.51 HELIUM CIRCULATOR 10 XXXXX Speed (RPM) 9,420 Flow (%) 110%
a/P (PSID) 7.5 Diffuser A/P (PSID) 4.8 Compressor A/P (PSID) 8.5 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio- 1.51
7 50-PLANT INSTRUMENTATION PERFORMANCE (B-5) i F
Data .
Data taken indicated no major problems. A defined method of establishing power level from heat balances is needed in the future to accurately compare indicated. Data logger inputs are being verified.
Comparison of Actual and Predicted Data Indicated = 856 MWT = 102% power.*
Actual = 856.9 MWT.**
Agreement = 0.1%
- Taken from NIM-1133-3, etc.
e
- - , ..~. ~ .
, -r%.- -,,,---m, - ,-4i
o SHIELDING SURVEYS (B-12)
Data 100*,' power level .
Highest detected levels:
South wall hot service facility - 1.5 mr/ hour.
Floor of hot service facility above T-6301-1.5 Mr/ Hour All other locations were less than .2 mr/ hour. All locations were well within anticipated values.
~. >* _5g_
RADI0 CHEMICAL ANALYSIS OF -THE PRIMARY COOLANT (B-13)
. Data The iodine monitor xenon collection was underway at the time of reactor shutdown. Therefore,:no' iodine datawas collected at the 100*.' power level. Data will be collected in the future at high power levels.
e i
i
. . -- : - ~-. ._: ,, ._,_ _ . . - , _. __
. . i STATUS OF B-SERIES STEADY-STATE START-UP TESTS SE TITLE STATUS B1-1 hteamGeneratorSteadyState Closed 81-3' Steam Generator Steady State Closed 1 (Data with Plugged Tube)
B1-4 Steam Generator Steady State Closed 1 (HP Feedwater Heater Bypassed)
B1-6 Turbine Generator Steady State Closed Performance B2-1 Primary Coolant System Impurities Closed 82-2 Purification Chemical Impurities Closed 82-3 Calibration of Gas Chromatograph Closed 83-1R Liner Cooling Maximum Temperatures Closed B3-1S Liner Cocling ADJ (as required) Closed 83-1Q PCRV Liner Cooling Closed B3-2Q PCRV Data Scan Closed 83-2R PCRV Internal Temperatures Closed B3-3A PCRV Leak Tightness Closed 83-3B Full Pressure PCRV Leakage Closed SR 5.2.1.6a-M (as required)
B4-1(0) Circulator Primary Coolant Closed 85-1 Nuclear Instrument Calibration (100%) Open B5-2 Core Region TC Calibration Closed B5-3 Feedwater Flow Calibration Closed B11 Xenon Stability Test (100%) Open i
1 Steam Generator Steady State data was taken with HP feedwater heater #5 removed from the system. However, both of these SUT's are considered closed. Whenever all feedwater heaters are operational, data will be taken to analyze the complete system.
9
r-
, . -- O STATUS OF B-SERIES START-UP TESTS (CONTINUED)
SU_T TITLE STATUS 2
B12-2 Shielding Survey (100*.') Open B13-1 Radioactive Gas Analysis Closed B13-2 Iodine Probe Analysis (100*.') Open RT-485 Control Rod Drive Internal- Open Temperatures (data to be taken at 6*.' interval s between 70*.' and 100".)
2 Shielding Survey data was acquired at 100% power. However, this startup test is still considered open pending final review of results by General Atomic Company and Public Service Company of Colorado.
~ ~ ,
9 so s . p THE FORT ST. VRAIN INITIAL APPROACH TO POWER TESTS (B-SERIES)
INTERIM REPORT 21 Report for Period Ending November 22, 1981 m
r-s 6
., i TABLE OF CONTENTS Page Introduction.............................. ........................ 3 8 Series Test Descriptions.................................... .... 6 Figure 1: , Rise-to-Power Testing Sequence ........... ............ 11 Acknowledgement .................................................. 12 H i sto ri cal Summa ry o f Pl ant Ope rati o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Testing Summary .................................................. 15 Steam System Performance Tests (B-1 Series)
- 1. Steam Generator Steady State Performance (B1-1) ........ 16
- 2. Turbine Generator Steady State Performance (B1-6) ...... 16 Chemical Impurities in the Primary Coolant Tests (B-2 Series)
- 3. Primary Coolant System Impuri ties (B2-1) . . . . . . . . . . . . . . . . 17
- 4. Purification Chemical Impurities (82-2) . . . . . . . . . . . . . . . . 17
- 5. Calibration of Gas Chromatograph (B2-3) . . . . . . . . . . . . . . . . 18 PCRV Performance Tests (B-3 Series)
- 6. PCRV Li ne r Cool i ng 83-1Q) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
- 7. Liner Cooling Maximum Temperatures (B3-1R) . . . . . . . . . . . . . . 18
- 8. Liner Cooling Adjustment ( As Requir ed) (B3-1S). . . . . . . . . .19
- 9. PCRV Internal Temperatures (83-2RT .................... 19
- 10. PCRV Data Scan (B3-20).................................. 19
- 11. PCRV Lea k Tightne s s (B3-3A) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
- 12. Full Pressure PCRV Leakage (SR 5.2.16a-M)
(As Required) (83-38) .................................. 20 Primary Coolant System Performance Tests (B-4 Series)
- 13. Ci rculator Prima ry Coolant (B4-1(0)) . . . . . . . . . . . . . . . . . . . . 20 Plant Instrumentation Performance Tests (B-5 Series)
- 14. Nuclear Instrument Calibration (BS-1) .................. 21
- 15. Core Region Thermocouple Calibration (B5-2) . . . . . . . . . . . . 22
- 16. Feedwater Flow Calibration (B5-3) . . . . . . . . . . . . . . . . . . . . . . 22 Radiochemical Analysis fo the Primary Coolant Tests (B-13 Series)
- 17. Radioacti ve Gas Analysi s (B13-1) . . . . . . . . . . . . . . . . . . . . . . 23
- 18. Iodi ne Probe Analysi s (813-2) . . . . . . . . . . . . . . . . . . . . . . . . . . 23 l
Activity Schedule per T-164 ......... ........................... 24
r '
s
- - .. . i
't Individual B' Series Test Data B1-1 ........................................................ 26 B1-6 ........................................................ 32 8 2 - 1, B 2-2 , B 2 - 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 B3-10........................................................ 40 B3.IR, B3-IS ................................................ 41 B3 20........................................................ 42 B3 2R........................................................ 45 B41(0)......................................................48 B 5_1,- B 5 2 , B 5 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0
. n i DAao._(e..........................................................ci JA B13 2........................................................ 52 S tatu s o f B-Se'-t e s S tart-up Tes t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 1
y - m
,, . 3-INITIAL APPROACH TO POWER TESTS (B SERIES STARTUp TESTS)
The init1al approach to power is accomplished in a series of discrete 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 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. The sequence for the performance of these tests is given in Figure 1, together with the correspond!ng-approximate reactor power levels. The reactor power levels, helium flow rates, feedwater flow rates, steam temperatures, and steam presstre given in the following description of the initial approach-to power may differ somewhat from those in the actual approach to power due to change in test requirements or improvements in operating methods identified during other tests.
In general, the initial approach-to power will be accomplished in the following order:
- 1. Feedwater flow will first be established through both steam generator loops and the bypass flash tank system using a boiler
,, _4
. . i I . .
feedpump. Helium flow through the core will be provided using one circulator in each loop.
- 2. The reactor power will be increased to approximately 2%.
- 3. The reactor power, feedwater flow, and helium flow rate will be simultaneously increased to 5% power, 20% helium flow, and 25%
feedwater flow using reactor generated steam from the bypass flash tank supplemented by the auxiliary boiler to power the circulating turbines, turbine driven boiler feedpump, and other plant steam requirements.
- 4. The reactor power will then be increased to approximately 8%,
concurrent with an increase in feedwater flow to about 30%. The helium flow will be maintained at about 20% during this power increase. At this condition, 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.
9 t
t
- 6. The reactor power will be increased to about 185 simultaneously with an increase in helium flow to about 33%, maintaining 25%
feedwater flow, followed by an increase in reactor power to about 26% with a helium flow of 49%. At this condition, the
~
main steam temperature will be about 800 degrees fahrenheit.
1
- 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 maintain full steam conditions.
This report covers tests performed between 70% and 100% reactor power.
Each power level was maintained for a period of time to perform one or more of the following tests. Preliminary analysis of these measurements, as specified in the overall controlling test document, was completed prior to increasing the reactor power to the next i stage, s
s
,,, i . , ,
Steam System Performance Tests (B-1)
Just prior to steaming, and at subsequent power levels during the i
initial rise-to power, data will be accumulated and analyzed on the ,
performance of the steam generators, the turbine, 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 Imourities in the Primary Coolant (B-2)
As the reactor power level is increased to about 11% of rated, the core and reactor internals will experience temperatures in excess of those reached during the core heat-up for reactivity coefficient measurements. At these temperatures, 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 taken and analyzed.
PCRV performance Tests (B-3)
As the reactor power level is increased 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 will be taken and analyzed on the performance of the PCRV and its cooling system on the structural response of the PCRV to increased internal pressure and on the primary system helium use rate.
j
1 4
primary Coolant System Performance Tests (B-4) p At each power level, data on the performance of the he'ium circulators and their auxiliaries will be taken and analy ed.
Measurements of-~the radial power distribution (region peaking factors) will be made at approximately 2%, 5%, and 8% reactor power.
Data on the performance and calitration of the core helium flow orifice valve will be obtained at approximately 28%, 50%, and 100%
reactor power.
Plant Instrumentation Performance Tests (B-5)
In these tests, the. performance of the portions of the plant instrumentation, which could not be tested prior to power operation, will be checked. The nuclear' instrumentation 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 the helium flow orifices at approximately 8% power and again at approximately 100% power.
plant Transient Performance Tests (8-6)
In these tests, the transient performance of the plant will be tested and analyzed. The testing will include: a scram and turbine trip from approximately 28% reactor power with rated steam conditions, a turbine trip from approximately 40% reactor power, a main turbine generator load rejection from approximately 60% reactor power to house load, sequential tripping of the two circulators in a loop from
1 6
approximately 80% reactor power and resultant loop shutdown, and t
boiler feedpump start and stop transients.
plant Automatic Control System performance Tests (8-7)
The components of the automatic control system will be placed into service and tested as the controlled' variables come into their controllable range. Dynamic verification tests of the control system will be performed at selected power levels during the oower 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 worth of control rods which are moved during the initial rise-to-power will be measured using a reactivity computer to obtain the instantaneous reactivity change produced by a control rod motion.
,. . . .g.
Xenon Buildup and Decay Measurements (B-10) i The reactivity change produced by buildup, burnout, or decay of xenon poison following a power level change will be measured by recording the change in the critical control rod positions following a change.
Xenon Stability Test (B-11)
In this test, the absence of any sustained xenon oscillations is demonstrated. At 100% power, a perturbation is produced from equilibrium xenon by inserting a control rod in one region and withdrawing a control rod in another region. The indicated power level and region outlet temperatures are recorded as a function of time and analyzed for the presence of any oscillation produced by xanon.
Shielding Surveys (B-12)
At approximately 28% reactor power and approximately 100% reactor power, surveys of the radiation levels within the plant are performed. An additional survey is taken during and following any regeneration of the helium purification system. These measured data are recorded and analyzed to demonstrate the adequacy of the shielding design.
- N
%%m
. s 4
Radiochemical Analysis of the primary Coolant (8-13)
In this~ test, the radioactive gaseous fission products in the primary coolant will be scmplet and analyzed. These tests are used in the' initial'startup phase to define fuel fission product release-to-birth ratio at zero burnup and will 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.
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b ACKNOWLEDGEMENT The contents of this , 5 report on the results of B Series Startup Testing at -Fort St. Vrain, Unit No. 1, have -been taken from unpublished, internal reports of General Atomic Company and Public Service Company of Colorado.
This is an interim report based on preliminary data; and therefore, both data and results are subject to change. This report will be supplemented periodically as further testing is completed.
O w e
- o . ,
HISTORICAL
SUMMARY
OF PLANT OPERATION The last testing reported was covered in interim report number 8 for the period ending August 22, 1978. After this, the plant continued operation with the Nuclear Regulatory Commission limit of 70% power.
In February, 1979, the shutdown for the first refueling was started with the shutdown extending until June, 1979. After startup, normal operation was resumed until Septemear 1, 1979, when the plant was shutdown due to inconsistencies found in the safety related piping and hangers. This was found during an audit and analysis committed to by the Company and was reported in Reportable Occurrence 50-267/79-35/01-T-0.
On September 15, 1979, the reactor was started up and operation continued until October 26, 1979, when a maintenance shutdown began to install the region constraint devices (RCD). The RCD installation was completed, and the reactor taken critical on December 25, 1979.
The RCD's are metal clamps that join the tops of all the regions together to maintain nearly uniform gap flow areas between regions.
Following a reactor scram, 1B helium circulator static seal failed in January, 1980, requiring the circulator to be replaced with the spare.
l l
l
On February 17, 1980, the reactor was taken critical again, and the turbine generator was put on line March 5, 1980.
Operation continued throughout the year and into 1981.
On March 16, 1981, Amendment 23 was issued, which authorized fluctuation' testing at greater than 70% power. Fluctuation testing was started on March 18, 1981, with pcwer levels of 70% to 88% being reached during the period April 17 to April 24, and on May 13, 1981.
On May 13, 1981, the turbine generator tripped on high vibration, due to a loose shroud on the low pressure turbine blading and was followed by a reactor scram. As the turbine repair was to take several weeks, the second refueling outage was started on May 20, 1981. During this period, 1B circulator was replaced again due to another failed static seal.
The refueling shutdown was finished in July, 1981, and the reactor was taken critical on July 13, 1981. Plant operation has continued throughout the rest of this report period at power levels up to 100%
and 323 MWe.
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TESTING
SUMMARY
This report covers the period from August 23, 1981, through November 22, 1981, and the analysis of the test results collected during the last period of May 23, 1981, to August 22, 1981.
l Due to the installation of the region constraint devices and Nuclear ,
Regulatory Commission requirements, the B Series Startup Tests, RT-500K (fluctuation testing at greater than 70% power) and RT-485 (control rod drive temperature data collection) were run i concurrently, with RT-500K being the controlling document for testing under normal conditions from 40% to 100% power. The test schedule is i
called out in T-164 (coordination procedure for testing at greater l than 70% power) and is included at the end of the test summary section.
Testing at greater than 70% power was initiated on four occasions.
The first three (April 17, 1981, April 24, 1981, and May 13, 1981) were to a maximum power level of 88%, at which time turbine vibrations necessitated a plant shutdown and the second refueling outage was begun. However, on November 5, 1981, Rt-500K testing again began, and on November 6, the plant attained 100% power. The plant remained at the 100% power level for approximately 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> until shutting down on November 9, 1981, to commence a major system modification outage.
Based on tests completed so far, the following conclusions can be drawn concerning plant behavior (for details refer to individual B 4
Series Test Reports).
- 1. Steady-state performance above 70% power was generally satisfactory. However, it was noted that the reheat outlet temperature of module B-2-5 read much lower than the remaining five modules in Loop 2. Although the 20*F range (referred to the average) was not quite met, it is believed that the operating limitations associated with the attendent performance of RT-500 (Fluctuation Testing above 70% Power) resulted in the observed off-normal temperatures. Following normal system operating procedures, expected reheat temperatures can be achieved. Similar situations at lower power levels have been resolved satisfactorily by judicious adjustment of primary coolant flow orifices and feedwater trim valves. Special instrumentation outputs are being analyzed by General Atomic.
Preliminary results appear to be acceptable. Unless subsequent investigation reveals inconsistencies or data outside of 'B' series acceptance criteria, the Fort St. Vrain technical review committee considers SVT B1, Part 1, closed. (81-1)
- 2. Steady-state performance of the turbine generator was generally satisfactory. At 100% indicated reactor power, the generator was producing 323 MW(e) versus the expected value of 339 MW(e).
It is postulated that the low feedwater flow may be contributing to the less-than-design output. It is also noted that feedwater I
i
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,3 . . .
heater No. 5 was not in service, nor was a reheat steam
, temperature of 1000*F being developed. The reheat steam temperature was restricted to less than 1000 F $due to limitations imposed by the RT-500 fluctuation testing. SUT B1, Part 6 is considered closed. Further investigation into possible feedwater flow discrepancies will be addressed as part of an analysis associated with the results of SVT B5, Part 3, Feedwater Flow Calibration (B1-6)
- 3. Primary coolant impurity levels monitored at 100% power were satisfactory. The hydrogen getter was not in service; it is being modified by a change notice. Hydrogen data was not available due to gas chromatograph equipment outage. Primary coolant impurities are monitored on a routine basis as part of normal plant operation. No unresolved issues remain from this startup test. SUT B2, Part 1 is closed. (B2-1)
- 4. Operation of the purification trains over the past several years have indicated satisfactory performance on a routine basis.
Train operation at 100% appears to be satisfactory in general, despite the fact that data recorded during the 100% power run indicated that the 'B' train (in-service for approximately three months) was showing signs of impurity break-through. However, overall train performance was satisfactory; SUT B2, Part 2 is considered closed. (82-2)
N _ _ . .
- s. ._
, 5. Gas chromatograph operation and periodic calibration are routine l
operatons. Results of calibrations performed prior to 100*.
l
! power operation were consistent with previous calibrationsc SUT B2, Part 3 is considered closed. '(B2-3)
- 6. Total liner cooling heat load was determined to be between 14 and 17 millton BTU /hr. This is consistent with expected values.
1 Data analysis indicates that the turbine flow meter FE-46165 l
whic.h measures flow to subheader 4T (PCRV top penetrations,
! Loop 2) may be indicating a higher flow rate than actually l
l exists. Public Service Company Results Department will ,
investigate. It is also possible that the individual flow meters for the high temperature filter absorber units are out of calibration as evidenced by inconsistent heat loads for the l
separate cooling tubes. In"estigation here is also underway.
Based on the results of data collection at 100*. power, SUT B3, Part 10 is considered closed. However, routine system analysis of system performance will be carried out on a regular basis.
l (B3-10)
- 7. Hot spot tube temperature rises were consistent with expected values published in Public Service Company's letter to the
! Nuclear Regulatory Commission, P-78037. The entire liner cooling system will be monitored on a regular routine basis to i
l assure that 'no significant changes have occurred within either l
the system or the PCRV. The individual tubes wrapped around the l
- - - . - - - - - . - - - - - - - ~ ~ - -
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Region 5 top penetration show markedly different temperature rises (21 F versus 9'F). This will be checked. SVT B3, Part IR is considered closed. (83-1R)
- 8. No subheader or tube valve positions required adjustment during the 100% power run. Subheader 2BS (46248) on the bottom head system should have its indicated flow increased from 109 to 145 gpm to bring the overall system temperature rises into better agreement. SUT B3, Part IS is considered closed.
(83-15)
- 9. Two sets of data were collected while operating at 100% power.
The results differed by 30 to 50*F. The lower temperature data set was consistent with previous values collected at lower power. The second set more closely reflected expected temperatures. This data will continue to be collected at regular intervals to establish historical reference information and possible trends. SUT B3, Part 2R is considered closed.
(B3-2R)
- 10. Information collected from the PCRV Data Acquisition System has been analyzed and found to be valid. Due to the lack of recent zero (low) PCRV Internal Pressure data, load cell responses could not be reconciled to the acceptance criteria. Public Service Company will, however, provide additional data to m
+
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s General Atomic Company for further anal / sis. The analytical results will be evaluated by Public Service Company. Two instrument channels (100 and 247) will be checked for discrepant
.l response. SUT B3, Part 20 is considered closed pending the above load cell information. However, regular routine data analysis from the PCRV Data Acquisition System will be continued per Public Service Company's established procedures. (B3-20) 3
- 11. Helium leakage from the PCRV was negligible. Helium leakage from the PCRV auxiliary piping system was consistent with values observed over the past several years. A substantial purified helium leak into the reheat steam system by way of the Loop 2 steam generator interspaces is known to exist. Overall results of this test were satisfactory. SVT B3, Part 3A is considered closed. Monitoring of plant wide helium leakage will continue on a routine basis. (B3-3A)
- 12. The data for this test is routinely collected via Public Service Company Surveillance Test 5.2.lba-M, and was performed on November 7, 1981, at 100% power. No discrepant responses were observed. SUT B3, Part 3B is considered closed. (B3-38)
- 13. The reactor / primary coolant system was being operated under conditions specified by RT-500 (fluctuation testing). As such, primary system conditions were different than those anticipated I
- - - , , =,
s by the SVT acceptance criteria. Primary coolant flow was about 10 to 13% greater than expected based on AP information from sensing points at the circulator inlets. Overall circulator AP was less than expected due to flow resistance values established by RT-500. Continuous operation at 100% reactor power was determined to be satisfactory, and the Fort St. Vrain technical review committee judged that the system performance could be l
adjusted to meet the desired acceptance criteria if the RT-500 restrictions had not been governing. SVT 84, Part 1 is considered closed. (B4-1(0))
- 14. Indicated reactor power equals 856 MW(t) (=102%). Actual reactor power depends on accurate measurement of feedwater flow and heat balance calculations plus data logger information.
Some of these data yield power levels of 814 MW(t) (=96.5).
Further effort to refine and otherwise produce more accurate power calculations is needed, especially in the area of feedwater flow accuracies. (See SVT B5, Part 3.) Control rod patterns expected at 100% steady state power were generally achieved. The observed reactivity discrepancy was 0.003aP.
There is a known bias of 0.003aP in the Base Reactivity Curve.
The bias will be corrected when the new Base Reactivity Curve is approved by the Nuclear Facility Safety Committee. This correccion will result in a reactivity discrepancy that is essentially zero. SR-5.1.4-W-P testing confirmed the existence of a proper control rod pattern. (85-1)
i
- 15. A broad range of results were observed during SVT B5, Part 2.
The nature of this instrumentation, plus existing recognized discrepant problems, all contribute to the uncertainties associated with data collected for this SUT. The recorded results were satisfactory, and further indicated that additional confirmatory information be obtained from reduction of fluctuation monitoring data. The Fort St. Vrain Technical Committee further established that acceptance criteria established for this test are believed to be too restrictive, and that routine annual surveillance tests performed recently indicate satisfactory results. Previous core outlet thermocouple traverses recorded in tests RT-509 and RT-524 will be reviewed to confirm adherence with the requirements of LCO 4.1.9. SUT B5, Part 2 is considered closed. (BS-2)
- 16. The results of SUT B5, Part 3, have concluded that further effort is needed to resolve apparent discrepancies. The SUT itself is declared closed, in that the 100% data was satisfactorily recorded and the analysis successfully concluded.
The analytical results however, indicate that a more detailed plan must be developed to determine the predominant variables and ranges which contribute to the end result. A Public Service Company T test will be prepared and issued to continue data collection and analysis as required to ultimately resolve the feedwater flowrate issue. General Atomic, San Diego, has advised that certain sensitivity calculations are currently in progress, the results of which may assist Public Service Company
I in developing their approach to resolution of feedwater flow discrepancies. (85-3)
- 17. No fast gas samples were obtained at 100% power due to flow blockage in the sample line. With respect to activity values and radiation background values, recent data indicate a level of about 1/10th of the anticipated values. Operation of the purification system has been fully satisfactory in removing radioactive gases as evidenced by purification train outlet samples. This SUT is considered closed, however, further analyses of the primary coolant will be obtained via routine
! surveillance programs augmented by Public Service Company T-test i
performance as required. (B13-1)
- 18. The iodine analysis test was in progress at the time of the 0155, November 9, 1981 plant shutdown, and no complete data was collected. This test will be repeated at 100% power at the next opportunity. This test is considered open. (13-2)
. -24' -
. .- ,/-' PUBLIC SERVICE COf.1PANY OF COLORADO s .
[4
. ; FCRT St. VRMN NUC'.!AR GENERATING STAT!'jN I-16A t -
rage 9 of 20 STE:0Y STATE 3-0 CATA COLLECTICN" APPROXIMATE PC'4ER LEVELS SUT TITLE ~50 ~70 '30 '90 '100 31-1 Steam Genera:ce Steady 5:ste 0 0 0 0 R-31-3 Steam Genera:ce Steady State (Data witn Plugged Tube) R 31-4 Steam Generator 5:eacy Sta a (HP Feecwater Heater SyPassed) R 31-6 Turbine Genera:ce Steady State Performance 0 0 0 0 R 82-1 Primary Coolan System Im;urities C 0 R C R 32-2 Purification Chemical ImPurt-tes C 0 R 0 R 32-3 Calibration of 3as Chromatograph 0 0 R 0 A 33-1R Liner Cooling Maximum Temperatures G 0 R 0 R 33-15 Liner Cooling ADJ (As Required) - - - - -
33-10 :CRV Liner Cooling 0 0 0 0 R 33-20 PCRV Casa Scan 0 0 C. O R 33-2R PCRV Internal Temperatures 0 0 R 0 R 33-3A PCRV Leak Tign ness 0 0 R 0 R 33-33 Full Pressure PCRV Leakage SR 5.2.154-M (as requirec) - - - -
R 34-1(0) Circulator Primary Coolant 0 0 0 0 R 35-1 Nuclear Instrumen: Calibration 0 0 0 0 R 35-2 Core Region TC Calibration - - - -
R 35-3 Feedwater Flow Calibra-ion 0 0 0 0 R 311 Xenon Stability Tes:
- R 312-2 Shielding Survey 0 0 0 0 R 313-1 Radioactive Gas Analysis 0 0 0 0 R 313-2 Icdine Probe Analysis - - - -
R
- RT-485 Control Rod Drive Internal Temoeratures 0 - - - -
- (Data to be taken at 6% Intervals Between 70 anc 10C%)
RP290-M Generator Tem;erature Monitoring C 0 0 0 0 "0" Oc:ional Casa Taking "R" Required Ca:a Taking
- NOTE: For further details on indivicual 3-0 steady state da:a taking frecuency ses cages 10-15.
FIGURS 1
l l
STEAM SYSTEM PERFORMANCE VERIFICATION (B-1)
Data
- 1. Typical steam generator temperature performance data taken at 100?; is shown in Table B-1.1 (page 26), and on pages 27 through
- 30. .
- 2. Typical main turbine performance data taken at 100*; reactor thermal power (100P. turbine load) are shown in Table B-1.2 (page 31), and on pages 32 through 36.
6 k__.- _
,' ,l ,
TABLE B-1.1 STEAM GENERATOR PERFORMANCE DATA
- Average Cold Reheat Inlet Helium Steam Temp- Reheat Steam Main Steam Module Temperature erature 'F Temoerature Temperature B-1-1 1376 545 995 992 B-1-2 1348 537 992 993 (1)
B-1-3 1375 540 1003 997 B-1-4 1350 545 995 999 (1)
B-1-5 1364 540 991 990 B-1-6 1367 540 998 989 B-2-1 13/7 552 979 988 B-2-2 1379 553 990 990 B-2-3 1384 553 987 991 8-2-4 1398 555 1000 990 (2)
B-2-5 1372 555 970 989 (3)
B-2-6 1397 552 1000 988 (2)
(1) Low helium temperature to B-1-2 and B-1-4.
(2) High helium temperature to B-2-4 and B-2-6.
(3) Possible-high attemperation flow to B-2-5.
Low hot reheat steam at module outlet.
All values rounded to nearest whole unit.
- -i
o .,. ', . .
INSTRUMENT DATA
' inear Reactor Power Level
. 100%
W de Range Reactor Power Level 100%+
Helium Pressure 680 PSIA Helium Pressure 670 PSIA Heliam Pressure 690 PSIA Average Circulator Inlet Temperature 717*F Circulator Speed, 1A . 9270 RPM Circulator Speed, 18 9255 RPM Circulator Speed, 1C 9201 RPM Circulator Speed, ID 9240 RPM Circulator Helium Flow, 1A 109%
Circulator Helium Flow,1B 109%
Circulator Helium Flow,1C 111%
Circulator Helium Flow, 10 108%
Circulator Helium Differential Pressure, 1A 8.0 PSI Circulator Helium Differential Pressure, 18 7.6 PSI Circulator Helium Differential Pressure, 1C 7.8 PSI Circulator Helium Differential Pressure, ID 8.4 PSI Feedwater Temperature, Loop 1 380'F Feedwater Temperature, Loop 2 390'F Feedwater Flow, Loop 1 1,074,000 lbs/hr Feedwater Flow, Loop 2 1,091,200 lbs/hr Cold Reheat Pressure Before Circulators, Loop 1 870 PSIG Cold Reheat Pressure Before Circulators, Loop 2 870 PSIG Cold Reheat Pressure After Circulators, Loop 1 595 PSIG
t INSTRUMENT DATA Cold Reheat Pressure After Circulators, Loop 2 595 PSIG Main Steam Pressure, Loop 1 2500 PSIG Main Steam Pressure, Loop 2 2450 PSIG Main Steam Temperature, Loop 1 999'F Main Steam Temperature, Loop 2 993'F Turbine First Stage Pressure 1500 PSIG Turbine Throttle Pressure 2420 PSIG Hot Reheat Steam Pressure, Loop 1 545 PSIG Hot Reheat Steam Pressure, Loop 2 545 PSIG Desuperheater Water Flow, Loop 1 76,000 lbs/hr Desuperheater Water Flow, Loop 2 72,000 lbs/hr Feedwater Valve Differential Pressure, Loop 1 100 PSID Feedwater Valve Differential Pressure, Loop 2 100 PSID Hot Reheat Pressure 530 PSIG Hot Reheat Temperature at Turbine Generator Inlet 970*F Steam Chest Pressure 2360 PSIG Flash Tank Pressure 880 PSIG Turbine Generator Load 324 MW Module Feedwater Flow, Channel 1 81.5%
Module Feedwater Flow, Channel 2 80.0%
Module Feedwater Flow, Channel 3 84.5%
Module Feedwater Flow, Channel 4 80.5%
Module Feedwater Flow, Channel 5 83.6%
Module Feedwater Flow, Channel 6 83.5%
Module Feedwater Flow, Channel 7 85.0%
INSTRUMENT DATA Module Feedwater Flow, Channel 8 80.7%
Module Feedwater Flow, Channel 9 82.0%
Module Feedwater Flow, Channel 10 85.0%
Module Feedwater Flow, Channel 11 80.3%
Module Feedwater Flow, Channel 12 88.0%
Module Main Steam Temperature, Channel 1 992'F Module Main Steam Temperature, Channel 2 993'F Module Main Steam Temperature, Channel 3 997'F Module Main Steam Temperature, Channel 4 999'F Module Main Steam Temperature, Channel 5 990'F Module Main Steam Temperature, Channel 6 989'F Module Main Steam Temperature, Channel 7 988'F Module Main Steam Temperature, Channel 8 990'F Module Main Steam Temperature, Channel 9 991'F itodule Main Steam Temperature, Channel 10 990'F Module Main Steam Temperature, Channel 11 989'F Module Main Steam Temperature, Channel 12 988'F Circulator Outlet Steam Temperature, Loop 1 670'F Circulator Outlet Steam Temperature, Loop 2 675'F Reheater Inlet Temperature, Channel 1 545'F Reheater Inlet Temperature, Channel 2 537'F Reheater Inlet Temperature, Channel 3 540'F Reheater Inlet Temperature, Channel 4 545'F Reheater Inlet Temperature, Channel 5 540'F Rehaater Inlet Temperature, Channel 6 540'F
~_.
INSTRUMENT DATA Reheater Inlet Temperature, Channel 7 552'F Reheater Inlet Temperature, Channel 8 553'F Reheater Inlet Temperature, Channel 9 553'F Reheater Inlet Temperature, Channel 10 555'F Reheater Inlet Temperature, Channel 11 555'F Reheater Inlet Temperature, Channel 12 552'F Reheater Outlet Temperature, Channel 1 995'F Reheater Outlet Temperature, Channel 2 992'F Reheater Outlet Temperature, Channel 3 1003'F Reheater Outlet Temperature, Channel 4 995'F Reheater Outlet Temperature, Channel 5 991'F Reheater Outlet Temperature, Channel 6 998'F Reheater Outlet Temperature, Channel 7 979'F Reheater Outlet Temperature, Channel 8 990'F Reheater Outlet Temperature, Channel 9 987'F Reheater Outlet Temperature, Channel 10 1000'F Reheater Outlet Temperature, Channel 11 970'F Reheater Outlet Temperature, Channel 12 1000'F I
t . -
TABLE B-1.2 MAIN TURBINE GENERATOR PERFORMANCE DATA
- Item Predicted Values Measured Data Load MW(e) 339 323 Main Steam Temperature 1,000 984
, Reheat Steam Temperature 1,000 969 First Stage Pressure, PSIG 1,652 1,522 Main Steam Pressure 2,412 2,412 Cold Reheat Pressure 679 468 Feedwater Temperature 409 396 Condenser Pressure 2.5 2.45 Turbine Speed 3,600 3,600 Attemperation Flow 121,913 148,000 Maximum Vibration (Pt. 5) ---------
4.0 Mils All values rounded to nearest whole unit.
INSTRUMENT DATA Load 325 MW Number of Hours at Load 8.75 Hours Reactive KVA 36 Mega Vars Hydraulic Piston Stroke (Control Valves) 100%
' Hydraulic Piston Stroke (Control Valves) 100%
Hydraulic Piston Stroke (Control Valves) 69%
Hydraulic Piston Stroke (Control Valves) 10%
Steam Chest Pressure 2400 PSIG Main Steam Temperature 984'F First Stage Pressure 1510 PSIG High Pressure Exhaust to. Reheater 850 PSIG High Pressure Exhaust to Reheater 730'F Reheat Bowl Pressure , 510 PSIG Hot Reheat Temperature 969'F Hot Reheat Steam to IP Turbine Pressure 530 PSIG Cold Reheat Steam PT #14 468'F Fifteenth Stage Pressure, Heater 1 7.2 PSIA Fourteenth Stage Pressure, Heater 2 12.5 PSIA Thirteenth Stage Pressure, Heater 3 24.0 PSIA Eleventh Stage Pressure, Heater 4 78 PSIA Tenth Stage Pressure, Heater 5 148 PSIG Eighth Stage Pressure, Heater 6 238 PSIG Feedwater Temperature Leaving Top Heater 396 F Exhaust Pressure 23.13 Hg Barometer 24.98 Hg 9
33-INSTRUMENT DATA Steam Flow 2.0 x 10' lbs/hr Condensate Flow 2.1 x 10' lbs/hr Quantity of Makeup Demineralizer Water 50 GPM Feedwater Heaters in Service All But #5 Quantity of Attemperation to Reheat Loop 2 72 x 108 lbs/hr ,
j - Quantity of Attemperation to Reheat Loop 1 76 x 108 lbs/hr .
Main Steam Pressure 2500 PSIG Main Steam Pressure 2490 PSIG Operating 011 Pressure 225 4 Bearing Header Pressure 26.5 Main Pump Suction Pressure 23 I
Turbine Speed 3600 RPM Armature Current 8600 Amps Armature Current 8750 Amps Armature Current 8500 Amps Armature Voltage 21.7 KV Field Current 2100 Amps Field Voltage 245 Volts Field Temperature 64'C Hydrogen Pressure 19.5 PSIG Hydrogen Purity 97%
Differential Fan Pressure, Water 3.4 H 2O Exhaust Hood Temperature' 110 F 1
' Exhaust Hood Temperature 112 F Cold Gas Temperature 32.7'C
-% , -- - % ..w e , -~
34-INSTRUMENT DATA Cold Gas Temperature 35.0'C Cold Gas Temperature 30.5'C Cold Gas Temperature 34.0'C Stator Temperature, Liquid Out 56.5'C Stator Temperature, Liquid In 42.0'C Stator Temperature, Machine Gas Temperature 36'C Exciter Air In Temperature 28.6'C Exciter Air Out Temperature 41.5'C Collector Air Out Temperature 41.5'C Turbine Temperature and Differential Temperature Data Point 1 990'F
! Turbine Temperature and Differential Temperature Data Point 2 940'F Turbine Temperature and Differential Temperature Data Point 3 980'F Turbine Temperature and Differential Temperature Data Point 4 930*F Turbine Temperature and Differential Temperature Data Point 5 880'F Turbine Temperature and Differential Temperature Data Point 6 780*F 1 Turbine Temperature and Differential Temperature Data Point 7 960'F Turbine Temperature and Differential Temperature Data Point 8 835'F Turbine Temperature and Differential Temperature Data Point 9 130'F Turbine Temperature and Differential Temperature Data Point 10 460'F Turbine Temperature and Differential Temperature Data Point 11 230'F
-, ,. , - - - - - , . , , . - - . , - y
INSTRUMENT DATA Turbine Temperature and Differential Temperature Data Point 12 140'F Turbine Temperature and Differential Temperature Data Point 13 730'F Turbine Temperature and Differential Temperature Data Point 14 730'F Turbine Temperature and Differential Temperature Data Point 15 660'F Turbine Temperature and Differential Temperature Data Point 16 660'F Turbine Temperature and Differential Temperature Data Point 17 845'F Turbine Temperature and Differential Temperature Data Point 18 820'F Turbine Temperature and Differential Temperature Data Point 19 130'F Turbine Temperature and Differential Temperature Data Point 20 10'F Turbine Temperature and Differential Temperature Data Point 21 0'F Turbine Temperature and Differential Temperature Data Point 22 60'F '
Turbine Vibration Data Point 1 0.0 Mils Turbine Vibration Data Point 2 0.0 Mils Turbine Vibration Data Point 3 3.0 Mils Turbine Vibration Data Point 4 0.75 Mils Turbine Vibration Data Point 5 4.0 Mils Turbine Vibration Data Point 6 1.2 Mils
INSTRUMENT DATA Turbine Vibration "A", #1, Data Point 7 1.J Mils Turbine Vibration "A", #2, Data Point 8 1.2 Mils Turbir.3 Vibration "C", #1, Data Point 9 0.75 Mils Turbine Vibration "C", #2, Data Point 10 2.1 Mils Turbine Vibration Data Point 11 0.0. Mils
e CHEMICAL IMPURITIES IN THE PRIMARY COOLANT (B-2)
Data Table B-2.1, page 38, shows the data collected during operation at 100*.' power.
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TABLE B-2.1 (1) CORE OUTLET REACTOR 02 CO CO 2 11 Cll4 N2 HO2 DATE TIME TEMPERATURE *F POWER % PPMV PPMV PPMV PPkV PPMV PPMV PPMV 11-6-81 1800 1401 100 (2) 14.0 1.6 (2) 0.4 42.2 0.0 2000 1403 100 15.6 2.9 0.4 37.8 STD 2200 1405 100 14.6 2.5 0.4 OOS 0.0 11-7-81 0000 1413 100 15.0 2.3 0.3 40.0 0.0 0200 1414 100 14.8 2.5 0.3 39.2 00S 0400 1414 100 16.6 2.0 0.4 43.9 OOS 0600 1400 100 22.3 2.5 0.4 42.9 OOS 0900 1404 100 11.0 2.1 0.4 33.3 STD 1100 1413 100 10.8 2.2 0.3 33.3 0.0 1400 1412 100 13.3 2.3 0.4 71.7 0.0 1800 1414 .00 13.8 2.3 0.4 46.7 0.0 2200 1416 100 13.9 2.5 0.4 46.7 0.0 11-8-81 0200 1422 100 12.0 2.2 0.4 33.3 00S 0600 1415 100 15.1 2.3 0.5 35.8 00s 1100 1422 100 13.4 2.2 0.4 28.2 STD 1400 1420 100 10.7 2.3 0.4 24.3 OOS 1800 1421 100 14.4 2.5 0,4 29.4 00S 2200 1427 100 12.7 2.5 0.4 28.2 STD (1) Data times listed to nearest hour (2) Out of Service - No data taken
PCRV PERFORMANCE TESTS (B-3)
The heat ' loads for each subsection of the PCRV at 100% power are shown in Table B-3.1, page 40. The design allowable heat loads at
~
100 % power are also shown. (SUT-83-10)
The " hot spots" noted on the PCRV liner / concrete are listed in Table B-3.2, pagc 41. (SVT-83-1R,83-15)
The PCRV data scan provided temperature, pressure, and strain values.
This data is listed in Table B-3.3, pages 42 through 44. (SUT-83-2Q)
The internal temperature data for the PCRV are listed in Tables B-3.4A and B-3.48, pages 45 and 46. (SUT-83-2R) r- - c 3
TABLE B-3.1 HEAT LOAD (X 10' BTV/HR) 80*; POWER DESIGN ALLOWABLE PCRV SUBSECTION LOOP 1 LOOP 2 AT 100*;
Top Penetrations 0.97 0.98 2.270 Core Support Floor Top 1.61 1.51 ------
Core Support Floor Side 0.29 0.57 ------
Core Support Floor Bottom 1.23 1.12 ------
Total Core Support Floor 3.13 3.20 8.076 Upper Barrel and Top Head 1.44 1.44 6.292 Lower Barrel 1.25 1.19 3.350 Bottom Head and Bottom Penetrations 1.76 1.73 4.887 Total by 1. cop 8.55 8.54 12.435 TOTAL PCRV 17.09 24.87 u.
TABLE B-3.2
'PCRV LINER / CONCRETE HOT SPOTS Local Maximum Concrete Temperature Effective Concrete at the Temperature at the Liner Cooling Liner / Concrete Liner / Concrete Location / Area Tube AT Interface - F Interface *F Top Head 20*F @ 55". 201 F @ 100% 128*F @ 100".
Penetrations 30*F @ 100*4
' ~
Core Outlet 20*F @ 80". 200 F @ 20% 120*F @ 100%
Thermocouple 21 F 0 100". 346 F @ 100".
Penetrations Core Barrel 40 F @ 60%8 250*F @ 27".2 120*F @ 100%
Seal Area 47*F @ 100*; 326 F 0 100".2 Peripheral 20*F @ 40". 200 F @ 100*. 120*F @ 100".
Seal Area 28 F @ 100".
Loop Divider 15 F @ 100". 200*F @ 20". 120*F @ 100".
Baffle Area 354 F @ 100%
Steam . 16*F @ 100". 200*F @ 40*.' 120 F @ 100".
Generator 220 F @ 100%
Penetrations Cross Over 3*F @ 100% 200*F @ 11% 110*F @ 100%
Pipe 330*F @ 100%
FSAR Stated 20*F2 150 Normal 130*F Value 200 Normal Hot Spot *
- Amendment 29 established 40*F cooling tube AT, 250*F local maximum concrete temperature at the liner / concrete interface for the core support floor. Amendment 29 also allows under certain circumstances, a local maximum concrete temperature of 360*F.
r TABLE B-3.3 4 i
l' Temperatures
~
Channel Thermocouple Temperature (*F) 144 TE-1170-28 114 535 TE-1170-17 115 432 TE-1170-46 114 444 TE-1170-47 109 445 TE-1170-48 110 414 TE-1170-65 113 416 TE-1170-54 114 418 TE-1170-56 118 422 TE-1170-78 121 506 TE-1170-72 116 I
t 031 TE-1170-127 111 3
128 TE-1170-96 119 325 TE-1170-100 113 235 TE-1170-110 121 030 TE-1170-109 110 343 TE-1170-111 121 331 TE-1170-114 101 335 TE-1170-107 105 n-
.- .{ ,,
TABLE B-3.3 (CONTINUED) p I
Internal Vessel Pressures .
Channel Load Cell Load (KIPS) 2 VM-3 1300.6 1 VI-10 1266.2 0 VM-17 1321.8 3 VI-24 1309.9 4 VM-31 1279.7 100 VI-38 101 TOR-U2 1276.3 102 TIR-M1 1076.8 103 BOR-L4 1359.4 104 BIR-M4 1226.8 200 C0-1.6 1246.0 201 CI-1.4 1259.5 202 CO-2.2 1290.4 203 CO-2.1 1234.1 204 CO-3.2 1362.3 300- ' CO-3.1 1204.3 301 CI-7.4 1094.4 302 CO-8.3 1144.3 30'3 CI-10.1 1102.6 304 CO-11.2 1107.2 .
400 CI-15.3 1138.2 Bad reading - out of range.
,,.- - -4A-TABLE B-3.3 (CONTINUED)
Internal Vessel Pressures, Continued Channel load Cell Load (KIPS) 401 CO-17.6 1188.2 402 CO-17.5 1210.8 403 CO-17.4 1281.6 404 CO-17.3 1225.4 500 CI-18.5 1221.2 501 CO-19.6 1355.3 Strain
! Channel Strain Gauge Strats _
150 YE-1170-43 704.4 458 YE-1170-58 -30.7 247 YE-1170-74 747.5"*
280 YE-1170-63 -361.8 366 YE-11205-16 -40.5 Questionable reading.
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PRIMARY SYSTEM PERFORMANCE (B-4)
. Data i
Table B-4.1, page 48, shows the data collected at 100*.' power for all four heliu.m circulators.
k -
l TABLE B-4.1
\
Reactor Power 100*.'
HELIUM CIRCULATOR 1A XXXXX Speed (RPM) 9,354 F1ow (*.') 112%
A/P (PSI?) 7.2 Diffuser A/P (PSID) 6.7 Comoressor A/P (PSID) 8.3 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.54
_ HELIUM CIRCULATOR IB XXXXX Speed (RPM) 9,315 Flow (%) 112%
A/P (PSID) 9.5 Diffuser A/P (PSID) 5.3 Compressor A/P (PSID) 7.8 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.54
. ~m
.0, TABLE B-4.1 (CONTINUED)
Reactor Power 100%
HELIUM CIRCULATOR IC XXXXX Soeed (RPM) 9,386 Flow (%) 113%
A/P (PSID) 9.3 Diffuser A/P (PSID) 5.7 Compressor A/P (PSID) 7.8 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.51 HELIUM CIRCULATOR 10 XXXXX Speed (RPM) 9,420 Flow (%) 110%
A/P (PSID) 7.5 D1ffuser A/P (PSID) 4.8 Compressor A/P (PSID) 8.5 Steam Turbine Inlet Pressure (PSIG) 890 Steam Turbine Outlet Pressure (PSIG) 600 Steam Turbine Bypass Pressure Ratio 1.51
4 PLANT INSTRUMENTATION PERFORMANCE (B-5)
Data Data taken indicated no major problemt. A defined method of establishing power level from heat balances is needed in the future to accurately compare indicated. Data logger inputs are being verified.
Comparison of Actual and Predicted Data Indicated = 856 MWT = 102% power.*
Actual = 856.9 MWT."*
Agreement = 0.1%
Taken from NIM-1133-3, etc.
B5-3 Feedwater flow data conflicts with ' data logger at'#4 feedwater heater extraction conditions - if calculated directly from data logger = 814.2 MWT.
,1 i
SHIEL0 LNG SURVEYS (8-12)
Data 100*,' power level.
Highest detected levels:
South wall hot service facility - 1.5 mr/ hour.
Floor of hot service facility above T-6301-1.5 Mr/ Hour All other locations were less than .2 mr/ hour. All locations were well within anticipated values.
- - . . . . . . . . - . . ..~. .. -. . .. .. - . -
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4l i RADI0 CHEMICAL ANALYSIS OF THE PRIMARY COOLANT (B-13)
. .' Data. .
1
. The' iodine -monitor xenon collection -was underway at the time of-
~
4 reactor shutdown. -Therefore, no iodine data was collected ,at the i.
100*; power level . Data will be collected in the future at high power i levels.
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- . ' , ' . . me STATUS OF B-SERIES STEADY-STATE START-UP TESTS SUT TITLE STATUS 81-1 Steam Generator Steady State Closed 81-3 Steam Generator Steady State Closed 2 (Data with Plugged Tube) 81-4 Steam Generator Steady State Closed *
(HP Feedwater Heater Bypassed)
B1-6 Turbine Generator Steady State Closed Performance B2-1 Primary Coolant System Impurities Closed 82-2 Purification Chemical Impurities Closed 82-3 Calibration of Gas Chromatograph Closed 83-1R Liner Cooling Maximun Temperatures Closed B3-15 Liner Cooling ADJ (as required) Closed 83-1Q PCRV Liner Cooling Closed BT 20 PCRV Data Scan Closed 83-2R PCRV Internal Temperatures Closed B3-3A PCRV Leak Tightness Closed B3-3B Full Pressure PCRV Leakage Closed SR 5.2.1.6a-M (as required)
B4-1(0) Circulator Primary Coolant Closed 85-1 Nuclear Instrument Calibration (100%) Open 85-2 Core Region TC Calibration Closed B5-3 Feedwater Flow Calibration Closed 811 Xenon Stability Test (100%) Open 1
Steam Generator Steady State data was taken with HP feedwater heater #5 removed from the system. However, both of these SVT's are considered closed. Whenever all feedwater heaters are operational, data will be taken to analyze the complete system.
-~
- . . .s . ', , .
- STATUS OF B-SERIES START-UP TESTS (CONTINUED)
SE TITLE STATUS 2
B12-2 Shf elding Survey (100*.') Open B13-1 Radioactive Gas Analysis Closed 813-2 Icdine Probe Analysis (100*.') Open RT-485 Control Rod Drive Internal Open Temperatures (data to be taken at 6'.' intervals between 70% and 100%)
2 Shielding Survey data was acquired at 100*." power. However, this startup test is still considered open pending final review of results by General Atomic Company and Public Service Company of Colorado.
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