ML20085D228

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Reactor Plant Operating Experience Rept Apr 1962-1963
ML20085D228
Person / Time
Site: Saxton File:GPU Nuclear icon.png
Issue date: 04/30/1963
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SAXTON NUCLEAR EXPERIMENTAL CORP.
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ML20083L048 List: ... further results
References
FOIA-91-17 NUDOCS 9110150357
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This Operating Experience Report for the Saxton Reactor Plant has been prepared in accordance with subparagraph D(h) of the Saxton Provisional Operating License DPR-h. This subparagraph was added to the Saxton License by Amendment No.1, issued on October 9, 1962. This report is also being used as a basis for requesting that the(Saxtod' Provistonal" Lice Qeconvertedtoa

[p full-term license. _

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1 May 29, 1963 1

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J TABLE OF CONTENTS Page No.

SIM4ARY ................................................... 1 CORE LOADIN3 AND INITIAL CRITICALITY . . . . . . . . . . . . . . . . . . . . . . 3 REA CT OR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . h EXPERIMENTAL PROGRAM J Control Rod Driv e and Scram Tes ts . . . . . . . . . . . . . . . . . . . . . 10 Open Reactor Vessel Tests ............................. 13 Core Parameter Measurement Tests with R eac tor V es s el Clos ed . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 Power Operational Tests ............................... 16 Loss of Lead Iransient Tests . . . . . . . . . . . . . . . . . . . . . . . . . . 17 1

Loss of Coolant Flow and Natural Circulation Tests . . . . 18 Power Coef ficient Meas urements . . . . . . . . . . . . . . . . . . . . . . . . 19 Thermal and Hydraulic Meas ur ements . . . . . . . . . . . . . . . . . . . . 20 I

, Reactor Ves sel Meas ur ements . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 CONOLUSION ............................................... 21 TABLE I - Operating Statistics ....................... 5 TABLE II - Comparison of Reactor Design and 4

Analytical Values Versus Measured and Calculated Data ....................... 11 l TABLE III - Comparison of Assumed and Measured Core Reactivity Coefficiente Used for Ac cide nt Analys es . . . . . . . . . . . . . . . . . . . . . . . . 12 FIGURE 1 - Saxton 21 Assembly Core ................... lh I

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SUnt%i:.

Construction of the Saxton reactor plant was essentially completed in April cf 1962 Fuel loading was started on April 6 after a complete operational check of the control rod drives, main coolant system, and other reactor auxiliary systems and after a finding by the AEC that those portions j of the facility required for operation with the reactor vessol head off at power levels up to 200 KW were complete. Core loading was completed on April 12 and the reactor was taken critical for the first time on April 13.

j Electric power was generated for the first time on November 16, 1962 and a reactor power level of approximately 90% of the design power level was schieved on January 21, 1963. After recalibrating the main coolant flow rater the design reactor power level of 20 Mdt was achieved for the first i time on April 22, 1963 From April 13 until June lb, the reactor was operated on a two-shif t basis for the purpose cf training and licensing reactor plant super-viscrs and operators and for the purpose of calibrating nuclear instrumenta-tion and controls and obtaining core parameter measurements under ambient conditions with the reactor vessel head removed.

During the period between June lb and August 15, the reactor was shut, down, primarily for the purpose of making repairs and modifications to

! the safety inaection system and safety injection pumps, installing the reacter veseel head, assembling in-core instrumentation connections, and i m.du: ting t ests en the pressuricer level control instrumentation. The radioactive waste disposal system was also completed during this period ard other miner modifications and repairs were completed. A Finding of Completion was issued by the AEC on August 15,1962, permitting further

operation of the reactor with the reactor vessel head installed.

The reactor was started up again on August 15 and from this time

. un:11 November 7, with the exception of brief periods when the reactor was shut down for maintenance and modification, the reactor was operated at low peser levels of less than 200 KW for the purpose of carrying out the core parameter measurement tests required by Item 16 on page 12 of the Sexton Technical Specifications. Certain minor changes in the facility, as described in the Saxton Final Safeguards Report, have previously been l

reported to the AEC in accordance with the Saxton operating license and are therefore omitted from this report. Also, approximately one week of this period was devoted to the training and licensing of additional reactor plant supervisers and operators. During this period the reactor was operated on a two-shif t, five-day per week basis; however, the reactor was manned l i areund the clock, seven days per week; and operations, such as checking out the nuclear instrumentation and varying the boron concentration or main coolant temperature, were done on the third shift or on weekends.

'Ihe core parameter measurement tests were essentially completed on Ncycmber 7, and the period between November 7 and November 16 was used to calltrata the power range nuclear instrumentation and make a final check on l

the secondary system equipment and instrumentation, preparatory to operating l the reacter at the higher power levels required for electric power generation.

Electric power was generated for the first time on November 16 at reacter power levels up to approximately 7 6 MWt.

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m Following the initial operation of the turbine-generator with nuclear steam on Novenber 16, the reactor was shut down in accordance with a previously arranged schedule for a period of about three weeks in order to correct a number of nicor equipment deficiencies and leaks that had l

developed and to make some minor modifications and additions. Following this shutdown, the reactor was operated at power levels of 6.9,12.h5,15.15, and 17.6 Mdt for the purpose of conducting the power operational tests required by Iten 17 on page 12 of the Saxton Technical Specifications. The reactor was taken up to 17.8 Mdt on January 21, 1963, and was operated i

satisfactorily at this power level until January 28 when it was shut down following completien of the power operational tests.

The reactor was shut down cold for approximately six weeks during all of the menth of February 1963 and part of March 1963, primarily for the purpose of installing radiation shielding around the purification surge tank and for removing test subassemblies from the core and installing the new test subassemblies described in Addendum No.1 to the Safeguards Report for l Phase 1 of the Sexton Five-Year Research and Development Program. This addendum, dated December 20, 1962, was filed with the AEC as Supplement No. 2 to Amendment No.10 to the application for license. Other modifications to the secondary system and maintenance work in the reactor plant were carried out daring this period. A higher than desirable concentration of suspended matter was observed in the primary coolant system during this period and a new filter was installed in the purification system for cleaning up the system.

The ructor was started up again during the latter part of March and was operated up until April lh at a power level of approximately 17.8 Kdt.

t This was a test of approxinate3y 21 days' duration for the purpose of obtaining burnui. reactivity data and equilibrium Samarium data. An irrtestigation made during this period revealed that an error had been made in the original I calibration of the nain coolant flow meter. The reactor power level readings

' based on the primary system calcrimetry which was higher than the secondary I syst n calorimety were found tc be consmatively on the high side by approximatcly 11%. The actual reactor wer levels based on the correct alibrat on of the main ecolant flow meter have been used throughout this repert. This test was completed at the .7.8 Kdt power level and then the flow met:r was recalibrated during the period from April 17 to April 22 during which time the plant was shut down both for the purpose of giving reactor operator license examinations and for the purpose of repairing a i

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steam generator safety valve flange leak. The reactor was brought up to l

' 20 Edt for the first tine on April 26 and was operated at this power level during the remainder of April as part of a second 21-day test made for the purpose of obtaining a further check on burnup reactivity and equilibrium i

Samariam prior to starting the chemical shim control tests.

All of the core parameter measurement tests, power operattonal tests, and other instrument and control tests required by the Technical Specificatiens have been completed and, in fact, these tests and the data obtaini d were far more extensive than would be considered normal for a reactor

' built solely for power production. The results of all of these tests were l favorable frem a safety standpoint and confirmed the safety analyses that were made and presented in the Saxton Final Safeguards Report. Flux wire measurements combined with assumed engineering hot channel factors indicate that the ever-all power density hot channel factor is approximately 18% less than thc design power density hot channel factor. Flux wire measu ements ha"c also indicated that the over-all enthalpy hot channel factor is appr ai:rately jh% below the design value.

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3 1 CORE LOADING AND INITIAL CRITICALITY l Following the completion of functional testing of the control rod I mechanisms and the satisfactory completion of other inspections and tests, the AEC issued an authorization on April 3,1962, providing for loading the first coro and fer operating the reactor with the reacter vessel head off at power levels not exceeding-200 KWt. Delivery of all the fuel elements including the five 9-rod subassemblies, the six control rods and control rod fuel followers, and the three neutron sources was made by April 5 and the centainment vessel equipment access opening was c1ceed and sealed preparatory to core loading.

.la All of the rix control rods were conrected to the rod drive mechanisms in the reactor vessel in the " rods-in" position. The water in the reactor and fuel storage well, au well as the main coolant water, was

> borated to approximately 1600 ppm, which was calculated to give a shutdown margin in excess of 10%. The No. 2 and No. 5 control rods were withdrawn 12 inches during the loading operation. Periodically these rods were scrammed and were completely withdrawn to check their effective. ness.

Loading of the core was carried out without difficulty or incidents in I accordance with observed inverse count rate data taken during the loading >

operation. The five 9-rod subassemblies were loaded last on Thureday, April 12, and the reactor was taken critical for the first time at lih0 a.m.

1 on Friday, April 13, 1962, with the rods withdrawn approximately 19 inches, The only difficulty er.perienced during the core loading and initial criticality was that the source range BF-3 detectors did not pick up a suf-ficient number of counts from the neutron source in order to comply with one of the requirements of the operating license technical specifications. In order to obtain a sufficient number of counts, it was _ necessary to tempo-I rarily install the source range detectors inside the reactor vessel. The first approach to criticality with banked rods was terminated because of a misleading inverse count rate plot. This trouble was due to some extent I to erratic operation of the source range nuclear instrumentation. The location of the neut on detectors inside the vessel and a' possible shielding -

effect of the rod pattern being used were also believed to be factors. ~

After some adjustments and cleaning of the contacts 1n the rod bleck relay, 1 the reactor was taken critical on the second attempt with control rods 2 and 5 withdrawn approximately 20 inches and control rods 1, 3, h and 6

- withdrawn approximately 19 inches. ' Neutron " windows" consisting of stain- .

l less steel . water-tight boxes were interposed-between the reactor vessel support structure and the source range BF-3 detectors in their normal locatien.- This modification vas made .to displace borated well_ water in order to obtain F counts per second as required by the; Technical Specifica-I tions, and criticality was achieved again on April 16 with all rods banked at 18.9 -inches.

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I RU.CTOR OPERATION From initial criticality on April 13, 1962 to April 30, 1963, a period of approximately one year, the reactor was critical a total of 2397.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Tne reacter was shut down cold on six different occasions during this period for a total of approximately 22 weeks for the purpose j of completing the plant, making modifications to plant equipment and systems, and performing minor maintenance on plant deficiencies that developed.

Table I shows a list of pertinent reactor plant operating statistics as of April 30. W 3 T eing the monthe of April, May and June of 1962, the reactor was total of 10h hours in the startup range. The reacter was operated critiral a on a ho-shift basis for the purpose of training supervisors and operators and D r calibrating and testing the nuclear instrumentation and controls g

and carrying out some of the core parameter measurement tests at ambient 3

temperature with the reacter vessel head removed. Th:so tests and the results are discussed in this report under "Dcperinental Program."

r The reactor was shut down cold most of the time during the period l between June Ib and August 15 to repair a leak in the reactor and fuel storage well, to install the reactor vessel head and to assemble tho in-core instrumentation connections, and to complete several items of constnation g and testing required in order to obtain AEC approval for operation of the a

reactor with the reactor vessel head installed. In order to obtain data for evaluating the pressurizer level indication erratic behavior which had been observed during previous operation of the main coolant system, hot functional tests were conducted during this period with the main coolant system at operating pressure and temperature. One of the brated in-core instrumentation seals on top of the reat tor vessel head failed during this period. It was later learned that the pressurizer level indication effect was a minor calibration problem that was minimized by the addition of insulation to the pressurizer level columns. The in-ecre le instrumentation seal leak was later corrected by making a new seal above the leaking seal.

During the period from August 15 to September 7, the reactor was operated for 190 hours0.0022 days <br />0.0528 hours <br />3.141534e-4 weeks <br />7.2295e-5 months <br /> for the purpose of conducting the core parameter measurement tests at power levels below 200 KWt at ambient temperature and

' varying boren concentrations. Following these tests, the reactor was down I ec3d frem September 7 to September 23 to make a new in-core instrumentation st w to install vent valves on the pressarizer level columns. Further testug pre ed that venting cf these columns was not the solution to the erratic pressurizer level indication. Other minor maintenance items, such as leaks that had developed in the in-core instrumentation and process e

instrumentation differential pressure cell connections in the containment vessel, were also taken care of during this period.

l During the balance of September, October and up until November 15, the reactor was critical a total of 128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />, most of which was at power

, .I levels below 200 KWt, for the purpose of completing the core parameter

! metsurement tests. Nearly all of this operation was at operating preuure l and temperature and varying boron concentrations. Several days in October l

l ] the reactor was operated on two shifts for training operators preparatory l to their taking the AEC reactor operator license examinations which were conducted on October 9, 10 and 11.

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TABLE I OPF.RATIN3 STATISTICS AS OF APRIL 30, 1963 I

REACTOR Times started up No. 210 Hours critical Hrs. 2397.9 Times scrarmed (Scheduled Shutdown) No. 166

Times scrammed (inadvertent) No. 12 Therms 1 power generation MdH 19,220 Average burnup MdD/MTU 900 Control rod positions at 20 Mdt- Rod out (Equilibrium Xenon)

. Rod No. 1 In. 16.25 Rod No. 2 In. O Rod No. 3 In. 27 Rod No h In. 27 Rod No. 5 In. O Rod No. 6 In. 18.3 i

3 MAIN COOLANT AT 20 MWt Gross activity Ac/cc 0.2h3 - 0.906 Impurities ppm 0.050 - 0.061 0xygen ppm less than 0.005 Chlorides ppm less than 0.1 Hydrogen ec/kg 15 - h2

} WASTE DISPOSAL Mskeup to primary plant Gals. 76,282 Water discharged to river Gals. 261,096 A:tivity discharged to river Ac 20,202

- Drums of warte processed No. 9 l No.

l Drums of waste stored 9 Gar de sy tank releases Cu.ft. 65,000

Gas activity released Juc 369,679 RADIATION l

Ruimam accumulated dose (individual) nrem lho Average accumulated dose mrc:: 1.55 Maximum radiation level at 20 M4t outside contaiment vessel -

Northeast pipe tunnel mrem /hr. 130 Charging pump room miem/hr. 21 1

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During the period from November 7 to November 16, additional operational checks were made on the secondary system equipnent and instru-ments and controls, using boiler steam fron the Pennsylvania Electric Company power plant. The power range nuclear instrumentation was calibrated at hi herC power levels up to approximately one megawatt by means of the purification 'iystem heat removal equipment and by means of controlled dis-charge of steam to the secondary system. On November 16, the Unit No. 2 turbine-generator in the Pennsybrania Electric Company plant was operated for the first time usin;' nuclear steam. The unit was operated for approxi-mately three hours in order to obtain a check on the feedwater controls, as well as the over-all operability of the nuclear plant and the secondary )

bystem, including the turbine-generator. The reactor was operated at a power level of approximately 7.55 Kdt during this test.

$ The reactor was shut down cold from November 17 to December 12 in order to replace the in-core instrumentation and process instrumentation l iifferential pressure cell shut-off valves with a new type of valvo having l compression fitting connections rather than threaded connections. Con-siderable leakage had been experienced with the original valves that used threaded connections. This period was also used to insulete the pressurizer level columns and to carry out miscellaneous maintenance work required prior l to extended power operation. Prior to resuming power operation on December 16, a control rod worth measurement test was carried out for one additional l rod configuration and an additional hot functional test was conducted in F order to evaluate modifications that had been made to the pressurizer level instr.umentation.

On December 16, the Unit No. 2 turbine-generator was started again with nuclear stear. and was operated at a reactor power level of approximately 7.5 Kdt for three days, during which time some of the power operational tests required by Item 17 on page 12 of the Saxton Technical Specifications I were carried out. The reactor was shut down cold on December 22 in order to make repairs to a leaking flanged joint on one of the steam generator safety valves. Reactor heat-up started on December 2h and the reactor was I brought up to a power level of 12.h5 MWt on December 26. The reacter was operated at this power level until January 3 for the purpose of conducting -

the power operational tests. The reactor was shut down hot for several days during the week of January 6 to 12 for the purpose of recalibrating the in-I core instrumentation pitot = tubes and temperature sensing elements. The reactor was brought back up to power on January 11 and after evaluating additional data'taken at the 12.h5 Edt puwer level, the reactor power was I increased to a power level of 15,1 Kdt on January 1h. The reactor was operated et this power level during the remainder of the week of January 13-19 for the purpose of taking additional power operational test dath.

1 On January 21, the reactor was taken to a power level of 17.8 MWt and was

. operated at this power level until January 28 for the purpose of conducting-additional power operational tests.

The reactor was shut down cold during the entire month of February ard part of the month of March, primarily for the prpose of removing several of the 9-rod subassemblies and installing new 9-rod special sub-I. assemblies and a hollow tube assembly, all of which were described in Addendum No.1 to the Safeguards Report for Phase 1 of the Saxton Research and Development Program. It was also necessary to install a close-fitting I lead brick and cement block radiation shield around the purification surge tank and this job required more tbne than expected and actually established t

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. length of this shutdewn. Fission product gases accumulated in the top

' the purification surge tank and caused higher than desirable radiation

.<vels in the adjacent charging pump room and .n other areas of the Centrcel and Auxiliary Building. This shield has reduced the radiation levels in M1 areas to satisfactory levels. Installation of this shield was reported to the AEC Division of Licensing and Regulation on March lb,1963 as Change Repert No. 7.

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The 9-rod subacsembly No. 1 and the 9-rod subassembly No. 2, descrioed in Addendum No.1 to the Phase 1 Safeguards deport, were installed g

in the prescribed locations ta the core after two of the uniformly enriched

' 9-rod subasse .blies were removed from these locations. The b-rod tubassembly No. 3, described in Addendum No.1, was not completed in time for installation during this shutdown. The hollow tube subassembly No, h, described in Addendum No.1, was damal-2d during fabrication and as installed in the core in Location II in place ei one of the standard uniformly enriched 9-rod sub-assemblies, has only one 1/2-inch 1.D. thimble in which a rod oscillator

' was installed during this shutdown. The removal of the standard subassemblies and the installation of the new special subassemblics was carried out expeditiously with the exception of some small amount of welding required on one of the subassemblies in order to be absolutely sure that the sub-i assenbly would be properly located in the reactor core.

1 In removing the standard subassemblies it was observed that they

) vere pcrtially covered with a dark brown film that flaked off as the sub-assemblies were moved through the water to the storage rack. Samples of this material were collected and after a few days it was observed that this material expanded and became fluffy. Examination of the elements with an underwater periscope revealed that the material remaining on the elements also had a fluffy appearance. Microscopic examination of these samples showed that they contained cotton and paper fibers coated with corrosion product oxides. An analysis of these samples showed that the major con-stituents were combustible 62.7% 1ron lb.1% nickel 5.85% and chromium l '

less than 2L This material is believed to have gotten into the main coolant l l tvste . via the boric acid system, since prior to the cold shutdown the f impurities in the main coolant system were in the parts per billion range j

and well below specification. The cotton nr.d paper fibers that were revealed under the microscope would indicate that this material could be pieces of the paper bag used for the boric acid, or other foreign material that might actually have dropped into the boric acid tank. During this l l investigation it was also discovered that the Technical Grade boric acid l

that was being used had an impurity level of about 10 to 12 times that of l

the Special Quality boric acid that is presently being used, l

Af ter the reactor vessel port Conoseals and in-core instrumentation fer the new special subassemblies were assembled, the main coolant pump and l

' purification system were put into operation in an effort to clean up the main coolant system. The sintered stainless steel post-filter in this system plegged up imnediately and had to be taken out of service. Examination ef a sa .ple removed from this post-filter showed that it had a dark brown, very l fluffy appearance when dispersed in water and was slightly ferromagnetic.

I l Study of this sample under a microscope showed corrosion product oxides ,

containing some fibrous material and other granular material. Analysis of this sample showed the major constituents to be combustible 10.1%, iron 33.8%,

nickel 13.8%, chromium 2.3% and silicon 5.8%. While it aay have been possible I to clean up the main coolant system by means of the bleed and feed system and

a discharging the water te the tn terground waste disposal tanks, it was decided I

that it would u preferAle to install a high-capacity-type filter having replaceable cari idges sized for 25 micron particles in series with the existin; sr te e j stainless steel pcet-filter. This new filter was instaned during the firtt ween in l' arch and af ter having been in service approximately two days, the crud level in the main coolant systen was reduced from 2.7 ppm to 0.02S p , whj ch is less than the specificaticn f or such impurities. A change repr c e. ring the installation of this filter is being prepared and will be filed vith the AEC Division of Licensing and Regulation in the near future. hoplaceable cartridge-type filters have also been installed in the discharge 2:ne from the boric acid makeup punp and a change repert fer these filters is being prepared and will be filed in the near future.

Reactor heat-up was etarte, 'n March 12 and the reactor war up to jerating temperature en Earch lb. The reacter was operated at zero power over the weekend of March 16 and 17 for the purpose of training SELNI (Societa Elettrooucleare Italiana) operators and after additional low power

- physics tests and safety system and control tests were completed, tha turbine was put on the line and the rearter was taken up to a power level of 17.8 Mdt en Friday, March 22. The weck of March 2L-30 was used zo train Santon reactor operaters in recovery orocedures from a reactor scram and a turbine trip.

The reactor uns operated at the power level of 17.8 MWt until April 15 for the purpose of following Xenon and Samarium buildup and fuel j burnup with rod control as part of the prechemical shim control series of tests. The reactor power level was reduced to zero on April 15 and a Xenon follow test was cenducted until April 16 at which time the reactor was shut l

down in order to repair a steam generator safety valve flanced joint leak.

' This repair required a cold shutdown and reactor operator license examina-tions that had been scheduled for this week for Saxton,SELMI and Westinghouse I personnel had to be rescheduled on April 25 and 26.

I An investigation made just prior to the reactor shutdown on April 16 revealed that an error had been made in calibrsting the main coolant flow meter and that the meter was rcading approximately 11% high, Disagreement l

between the primary and secondary calorimetry and the low electrical output were the ret.uons for making an investigation which led to this discovery.

The flow meter was recalibrated during the shutdown in April. The reactor heat-up was started on April 18 and the reat. tor was brought up to the rated power level of M Mdt on April 22. On April 23 and 2h, a loss of load test aM a natural circulation test were carried out at the ?O Mdt power level.

After the reactor operator license examinations were completed on April 25 an'i ?b, the reactor was operated at 75% of full power for three days for uperimental purposes and was then taken up to full po9er level of 20 MWt an April 29. The reactor wao operated at 20 Kdt on April 29 and 30 as a part of the second three-week full load test to obtain Xenon and Samarium bui"; dup and fuel burnup data prior to starting the chemical shim control tes te .

With the exception of minor instrumentation and equipment problems aM other mirscellaneous problems that ave normally experienced during the ca % ' tion and initial testing of a power reactor, the Saxton reactor has

! ope s ,d very sat,isfactorily. The control rod drive mechanisms have per-l r formed as expect c ' and the n'aclear instrumentation and controls have also performed satis!actorily, with the exception of some trouble that has been l

t l experienced wit' the nuclear ins'srnentation. Several of the source range 1

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Pf4 de*ectors have had te be replated. The trouble with there detectors ta relieved to have been caused by several ft.cters such or (1) a mobture i problem due to their Ic stion directly in the neuti e. 6hield water, (?) a confiderable ancunt of operatien in the source rany during the core prrav tcr rehsuver:er.t t ests, en! (3) operation of the detectors at too high a neutron fin due

  • o the antH everlap intween the source range instrumentatien an!.

the interr.ediate rance itstrumentation. All of the above deficiencies have be en corrected or nc longer retent a proticm and the BF-3 detecters have been giving reliable perfe macce. Internittent " Rod Stop Fast burnup hate" and "High Start uJ bte s rm *inrms occurred during the apptca:h te ft01 I p:wer on April P? 9-1 the 'ctero'.ttent acram alarms have occurred ori power d range :hannel> 1 and "C'. L.a drcp tinct and scram circuit firing tiner hm all teer. we'.1 within the design values used in the ac:ident studier.

I a "M M et! has 14en %rL~.med manually a tott.1 ef if4 tivt f '-

scheduled shutdowne and has been scramed inadvertently 12 times. A few cf the it.edurtert strant. were due to operator error and others were due to in. proper scram eettings and to electronic circuit failures designed to fail safe.

l During the period of operation covered by this report eight s licenae Technical Specification Ctange Requests have been filed with tre AEO. Seven of these change requests have been approved by the AEC and Change hoy.est No. 8 is still being reviewed by the AEC. Eight Change l D.rpert e have been filed with the AEC during this period.

E - n ir.

  • o*'.a m' hemistry har been maintained withir spe:1ficat;;".e F' aH
  • im? dtrin.; th priod covered by this report. Afte* initin'. p?w?r l m.~rctir r it. Nmeerd er 196L rr.diation levele in the sampling room and Wr r pcr.p r::m ndi*,ated that there were fiesion produ::tr ir. the msir c : :' W . St Tler tcken and ar.aly:ed during an apprcach to power in Janary I IW ' M m d. that the fStion products are coming from a pinhola leak r a' m - hr ~ic p aranita.. The main coolant activity has fluctuated due t< iv emitt e-~ geratinr Of the purification rystem in an nffort to redute tu rnntwrea cf the enacg:.ng pump pa: king. The grors activitv at pow I has vr. rad fra 0.?id u:/ec to 0.006 pc/c and has varied fror 6.h X 10 4" /e:

C .*_ .? . M uher *,he reac tor is shut down. The I-133 hat verie.i frcr 1.76x.'[";',::/c: tc 1.0$ j:c/00 at full power.

The long time interval between initial criticality and electrical p:wer operaticn for the Saxten reactor war due to a large extent to the i more extensive data that has been taken during the core parameter tests and power operational tests than would be taken .for a production power raaetor. This data te, of course, needed in order to intelligently and asf ely plan and ualuate the Saxton five-year experimental pregram. The.

I wealth of knowledge and experience already obtained on the Saxton reactor will be a great aid to the operatore in conducting future experir.ents that are planned as a part of the five-yecr research and development program.

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_DJ UC/1NTAL PRO 3 HAM Darinr, the period between initial critienlity on April 13, 1962 and April 2h,1963, all of the series of core parameter measurement tests and power operatienal tests listed as Item 16 and Item 17 of the Saxten Technical 3pecifications were completed. Table II gives a conparison of J the experimental and calculated reactor characteristics with the design and analytical values that were used for the accident analyres that were made fer the Saxten Final Safeguarus heport. Table III also gives a compariren of the assumed and measured coefficients of reactivity for the reactivity accidents analyr.ed in the final Safeguards Report. A summary of tho t ests that were made, including some of the results, is given in

l. this rection of the report.

CONTROL ROD 1 RIVE AND SCRAM TESTS Prior to loading the fuel into the reactor vessel, the centrol reds re.d :entrol rod fellowers were installed in the vessel and opsrated with the upper core support barrel installed in order to check rod with-l drawal speed, red drop times, operability of the rod drive and scram me:hanisms and io che:k the rods for binding or rubbing. Also, the rod positien ind.is.ators in the main control room and the experimental equipment I room were ce.librated by optically observing the position of the control rod.

Lt. ring there testo, the rod speed was found to be approximately 1.8 incher per muute whit.h it faster than than the 1.5 inches per minuto used for the l reartivity accident ctudies. This rod withdrawal speed was reduced by 3

approximately 50% by changing the gearing in the control mechnnisms. The rea drop times varied fron a maximum of 0.763 second for rod No. 2 to a minimum of 0.663 second for rod No. 5. During these tests the red drive mechanirms, rod position indication circuite and indicators, and the control rod shock absorber mechanisanall functioned satisfactorily. He binding or rubbing of the control rods was observed during these tests.

The scram circuit firing times were che:ked prior to leading the core and were checked again on September 16, 1962 and on February 27, 1963 I in compliance with Item lh of the Technical Specifications. The results of these tests are shown in the table below:

Initial Scram Scram Times (ms)

I Times (ms) Sept.1962 Feb. 1963 Manual ecram 21 LO 30 100 i Pump breaker 25 150 270 225 Power Range Channel A h07 Power Range Channel B 66h 260 225 Power Range Channel C 395 270 225 I Interrediete Range Channel A 216 290 250 Intemediate Range Channel B 212 280 250 Im s of flow 176 230 190 I Low pressure (pressurizer) 216 250 205 Lew preesure (main coolant) 171 170 200 Low level (pressurizer) 359 225 280

, l 3ho 195 l High temperature 360 I

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-,-pe-- g ..,,,,m,

j 11 TABLE II, COMPARISON OF REACTOR 4

D1310N AND ANALYTICAL VALUES VIRSUS

}^.JASUltED AND CALCULATED DATA i

Item ,

Design Experimental *

1. Control red operating opeed, in/ min 15 1.026(Max.)

6 F

. 2. Main coolant flow, lbs/hr 2.8 x 10 2.95 x 10' 6 6

3. Effective flow for heat transfer, lbs/hr 2.52 x 10 2.h8 x 10 l
h. llent transfer at 20 Mdt

- DNB ratio 2.91 3 70 Peakpowerdensity,L~d/ftofrod 13 3 10 75 Maximum flux, Iitu/hr - ft2 hhh,000 363,000 l

5. I!ct charnel factors at 20 Mdt

! Feat flux - F q 3 2h 2.65 I Coolant rice - F AH 2.30 1. %

6. A.erage coolant rise in vessel at 20 Mdt - F 20.00 19 35
7. Contrcl characteristics Ke .t;, cold and clean 1.250 1.26 l 1.168 1. '. B T.pff. hot and clean 1.12h 1. ~.',

Kegg, full power and equilibrium Xc and Sm ,

All rods out boron concentrations Cold chutdown (k = 0.97) 2700 2h30 Het shutdown (k = 0.97) 2500 2360

6. Rod drop time and scram circuit firing time, sec. 1.5 1.20(Max.)

1

  • All experimental values cre nominal values and do not include uncertainties.

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TABIZ III f

C0!! PARIS 0110F AOSU}ZD AllD IG%SURED CORE REA0TIVITY COEFFICIEllTS USED FOR ACCIDINT AIMLYSIS Accidents Assumed Measured Startup from Source Unborated at 530 F Reactivity insertion A k/sec 5 x 10

~b

  • 2.h x 10-h lioderator temperature coefficient A k/ F -3 x 10 -h.3 x 10~b

-6 +b.B x 10-6 Moderator pressure coefficient Ak/ psi +3 x 10 Borated at 80 F i Reactivity insertion A k/sec 5 x 10~b *2.1 x 10~b Moderator tenperature coefficient Ak/F +1 x 10*b -0.2 x 10~b Modcrator pressure coefficient Ak/ psi -1 x 10-6 40.3 x 10-6 Borated at 530 F Reactivity insertion A k/sec 5 x 10~b *2.h x 10~b Moderator temperature coefficient 4k/ F -1.6 x 10*b -3 2 x 10~b

-6 Moderator pressure coefficient A k/ psi +1.6 x 10 +3 0 x 10~0 Rod Withdrawal at Power Unborated at 530 F Reactivity insertion A k/sec 5 x 10~b *2.h x 10 Moderator temperature coefficient Ak/F -3 0 x 10~b -h.3 x 10~b Moderator pressure coefficient Ak/ psi +3 0 x 10-6 +h.8 x 10-6 I

  • Highest value observed with six rode banked.

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It will be noted that 4provemente tere made in some of the scratri circuit times that were cbserved during the initial test. These inprovements were made by minor adjustments and chanres in the electronic circuits. 'ihe sum I of the maximum scram circuit time and maxinum rcd drop time is less than the 1.5 seconds used for the accident analyses that were made for the Final Safeguards hepert.

0FG REACT 0it V13SFL TILTS Luring April, by and June 1962 when the ri.acter was cperated for apprcxtr.tte3y 105 houre with the reactor vessel head removed, some preliminary ecre para .eter nearuremtntr wtre made. Just critical rod positions for banked redt and a nu9er of rod configurations showed that the reactor core and ccntrol reds had good synnetry. Shutdown margin, boren worth, and differential red werthe were calculated frem startup rate meneuremente and power decay measurenerte following scram taken at boron concentrations of approximately 1600 rpm, 3 500 ppm and 1h00 ppm. The inverse boron worth was found to vary frem arproximately 130 ppm /To k/k at 15h0 ppm concentration to approximate'ly 90 ppmMa k/n at lbhD ppm concentrttion. The shutdown margin varied from lb.hT4 k/x at approxinately 1600 ppm concentration to approximately 10.5% A k/k at a concentration of approxinately lh00 ppm. These shutdown margins were in excess of the 10% shutdown margin required by the Technical Specifications for operation with the reactor vessel head removed. Differ-ential control rod worths were measured for banked rods and three rod onfigurations for each of the three boren concentrations mentioned above.

Figare 1 i? a cross section of the Saxton core showing the location of the nuntered c ontrcl rods. The differential rod worth varied from a maximum of (6.9V . 0.6) x 10-3Ak/in. for a 6-rod bank confi uratienC with lh00 ppm boron cer.centation to a minimum of(2.61 10.95)x 10->4k/in. for rods 1, 3 and 5 withdrawn 100T and rods 2, h and 6 banked with 1600 ppm beron concentration.

Veld ccefficient measurements were made at the end of this period by taking just criticel banked centrol red position measurements with the

.' x i fus3 tubassembly shown in Figure 1 removed and tha came measurements wm a'.e made with a hollow aluminum box inserted in the ccre ir, the ser.e poeitien. Calculations based on the resulte of these experiments showed the w:rth of the 3 x 3 fuel subassembly referenced to the core configuration with ths arterMy rencved to be approximately +(6.2010.9) x 10-34k/k. The verth of the aluminum veid referenced to the core configuration with tb 3 x ' stbaerently removed was calculated to be approxintely

+(3.lb 1 0.e.) x 10-34k/k.

All ef the mearurenents taken during this period were prelininary 7

and the resalts presented above, with the exception of the void worths, were determined with better precision during subsequent e.xperimental work.

Fer example, the shutdown margin was found to be more than the numbers given 3

l abeve.

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I 11 A Il C D E t' O W

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15 CCRE PAkt METER MEASUid. melit TESTS WITH REACTOR VESSEL CICSED i

After the reactvr vessel head war installed in June 1962 and the head closure and in-core instrumentstion ports and connections had been 4

l hydrottatically tested satisfactorily, another series of control rod tests was cenducted with the nain coolant system at 530 F and 2000 psi. Average rod withdrawal speed varied from a maximum of 1.026 in./ min, for rod !!o. $

to a minimum of 0.'/0h in./ min. for rod lio, b. Rods flos. 1, 2, 3 and 5 were drcpped five times, rod lio. 6 was dropped 20 times and rod lio. L war drcppcd 30 t k.cs. Rod dro}> tests were made again on December 11, 1962 and scain on P. arch 15, 1963 in accordance with lien lh of the Techt.ical

' S;' etific at ions . The maximum drop time was 0.827 second on December 11 and 0.903 c,n P. arch 15. This maxinum delay time from scram initiation to ccntrol rods fully incerted is 1.195 seconds which is considerally 3 re r a

than the 15 seconds value used for accident analyses that were made for the final Safecunrde Report.

I Crbtrni rod vorth and boren worth tests at ambient temperature

' (apprcximately lh00F) were carried out from August 22 to September 5 for six differerit controi rod configurations with primary emphasis placed on the bard'ed control rod configuration. Doron concentration was varied from a

. all-rods-cut concentration (just critical) of 1958 ppm down to 612 ppm, which was higher than the lowest safe limit (giving in excess of 1% shutdown

! ma pin with a stuck control rod) as required by paragraph (c)(2) under item 19 of the Technical $poeif$Mtions. The calculated maximum differ-entia3 rod worth based on these experiments was found to be approximately Ib.0 x 10-34 k/ih. for the all-rods-banked configuration. This differential rod worth is ecuivalent to a maximum reactivity insertion of 2.h x 10-h Ak/sec.

for the maxhum rod speed of 1.026 in./ min, and is less than the maxinum insertion rate of 5.0 x 1044k/sec. used for the accident analyses.

I 2cron 4crths were determined from data obtained during banked con-trol rod measurements with the exception of two instances when both direct and indirect measurements were made uti31 zing a single control rod for reactivity control. The inverse boren worths obtained from there tests were found to vary frem approximately 91 ppm /%4 k/k at 860 ppm to approxi-metely160 ppm /%4k/kat1hh0 ppm.

I Based en the boron worth measurements that have been made, the madnum reactivity insertion by means of boric said dilution from the bleed l

and feed systen operating at 30 gpm is 17 x 10*54k/sec. This compares

, a with a value of L.7 x 10-54 k/sec, assumed in the Final Safeguards Report.

On September 2h, the re&ctor plant heatup was started and the moderator te .perature coefficient vat, measured over a temperature range of 1730F to $650F for a boren concentration of approximately 1hhD ppm. The coefficient varied frem -0.42 x 10-h4k/0F at 1730 F to -2.8 x 10-hAk/0F l et 5977. The moderator temperature coefficient at zero boren concentration and 537 F wat -L.3 x 10*ha k/0F. luring this series of tests it was learned that a tempe sture limit for a 3% shutdown margin with a clean core was appreximately hBOUF rather than the h300F as predicted in the Final Safeguarde Repo~t. All of the various operating instructions have been corre:ted to show the temperature limit of h800F rather than the original teaperature limit of h300F.

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16 I).e pressure coefficients at normal operating pressure and 520"F were deurr. iud at six beron concentrations ranging frem 0 ppn to If72 ppm.

Ecaturements were made by observing the change in startup rate for tw:

dif f(rer.+. r.ain coolant prescures with the nederator tenperature, boron uration held constant. Th ceefficient varied fron L.B x 10-6concentro+

4k ion, and control red04k/

cenfig/

psi psi at 0 ppm to 3 0 x 10 y at 1572 ppm.

I While the reacter was ahut dcwn cold durin;; the pcried of September 7 to f.eptenber 23 for modifications and maintenance work previously discussed, the six-month periodic test of scram settings and deram circuit response tir.es required by Iten lh of the Technical Specifications was c onduc t ed. The results of this test have already been reported in this I reper* vnder 'Tcntrol hed Drive and Scran Tests.t i

During this period when the boren concentratien was being reduced, a close check was kept on the shutdown margin and the rod withdrawal limi.t I devices were set so as to provide the stuck rod shutdown margins given in the Tcchnical Specificatione. In order to provide 2% ehutdown with a stuck rod at operating presture and temperature and zere beren concentratien, these lirits were set at approximately 16 inches for rods Nos. 2 and 5 and approxinately 2B inches for rods Nos.1, 3, h and 6.

POWEh OPERATIONAL TESTS T% p:wer operational tests required by Item 17 of the Technical Spenficatiene and ether tests required by the operating licente were carried out d.ri g tre months of December 1962 and January 1963. In December 1962, the rea:ter .ar operated in the power range for 172 hours0.00199 days <br />0.0478 hours <br />2.843915e-4 weeks <br />6.5446e-5 months <br /> and a total of 1918 negawatt-hours of heat were generated. In January 1963, the reactor was Operated in the power range fer 362 hours0.00419 days <br />0.101 hours <br />5.98545e-4 weeks <br />1.37741e-4 months <br /> and Fh07 megawatt-hours of heat were generated.

The power range nuclear instrumentation was calibrated in the 200 E.' to 1 M4 range in November, prior to initial electric operatien, by usir.g the 'cleed and feed system nonregenerative heater and the steam-turbine driven boiler feed pumps and other auxiliary eteam load as a heat sink.

During the power operation in December 1962 and January 1963, the reactor was operated at progressively higher power . levels from approxinately 8.9 Mdt up to a power level of 17.8 Kdt. Further calibration of the nuclear power range instrumentation was made at each of the various power levels by comparing the product of the main coolant flow and enthalpy rise across the reacter vessel to the steam flew and enthalpy rise acroce the steam generater.

Due to the error made in calibrating the main coolant flow meter, tne nuclear power range inetrumentation and the reacter power level recorder were set apprcximately 11% higher than the power that was actually being generated.

The main coolant ficw meter was recalibrated during April 1963 and the nuclear iretrumentation and reacter power level recorder were also adjusted after the reaetrr wue started and operated in the power range.

Radiation surveys were made at various points around the fa:ility and containment vessel at each power level. Some neutron streaming was observed around the centainment vessel during the first power runs at 8.9 Kdt.

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, s hadis' ion leven in the range of 30 to ho mre:& in the area of the equip-ment a:een opening tc the contsiraent vessel were observed and radiation ic"elt c high as 200 n n./h- vere observed in the undercrcund pipe tunnel 4 adjac ent to t% containnent vessel teneath the equipment access cper.ing. l J

Tnir rm; tron strctairc and the radiation levels in these areas were con-eicerably reduced by raising the reactor and fuel storage well water 3evel one foct, and by placing stacks of plywood ever the ventilating openings in the containment vessel operating floor. The stacke of plywcod used over the ventilatirc epenirge to the reactor and fuel stcrage well will bs ured at permanent shelding during operation. The ventilating ducts l mntioned are not required during reactor operation and the stacks of ply- ,

l wood can be remcved for ventilation of the reacter and fuel et.:"cge wt .L prirr te pere nn:1 entry irto this area. The radiation level t the equip-c+a- m 1rt-ing :: the P rimer+ venel war ret ed te ' et thr 25 mren/hr and the radiation levels at other peints arcurri the contair. ment vessel at grade level and on tep of the containment vessel are less than

, 2 5 mrem /hr. The additional shielding provided reduced the radiation level in the underground tunnell however, this radiation level during operation at 20 Mdt is still in excess of 100 mrem /hr and will therefore have to be treated as a high radiation area in accordance with AEC Regulation 10 CFR 20.

A locked t;ste has been provided to restrict access to this area during

, operation. A high radiation area also existed in the charging pamp room in the Contrcl and Auxiliary building during power operation and operation of the bleed and feed system, due to the presence of finion product gases in the tcp of tha purification surge tank that to located in one end of the cherring pump roem. Radiation levele in the vicinity of the charging pump? have teen obeerved as h1C h at h00 mrem /hr; however, levels immediately 94l e : et - to the cutside walir cf the charging pump room have been found to M te2 m t he tolerance levels specified in AEC Regulation 10 CFR 20.

Additlend shalding in the form cf a permanent eonerete block and lead

' brick va'.1 har been provided in the area adjacer.t to the purification surge tank and radiation levels in the area of the charging pumps hevo been redu'ed te 21 mrem /hr when the purification system is ir. ope ~ation at 10 gym.

LOSS OF LOAD TRANSIENT TESTS i

5 Tests of plant response to a complete less of ttrbine load were perf erned without automatic reactor control, starting from approximately

8.9,12.h5 and 17.8 Kdt; and with control, from 12.h5,17.8 and 20 Mat.

The unecr* olled tes ts at B.9 and 12.h5 Kdt demonstrated that the pressure reducing statier. between the steam generator and the turbine would not react f ast erough to prevent opening the low pressure header safety valve l

when the load was lost. Thus a loss of turbine load becans an increare in load on the reactor, and in these tests the reactor power output rose significantly before the pressu e reducing valve was closed. In t hs I remaiaing tests the pressure reducing valve was manually switched to the

< "close" setting at the time of loss of load, so that the safety valve did not epen. This action has been made automatic on turbine trip or generator trip, and a full load trip from 20 Edt was made on April 23,1C63 Every-thing performed satisfactorily. The hot leg temperature increared 120F and the rods were automatically inserted 2 33 inches. The 20 Ndt lead loss a

tests showed that the secondary pressure rise was not excesrive without any mr.naal a: tion for several minutes, but these tests were run without

[

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18 chmical shim, that it, with a high moderater er efficient. With the reduced mcderator coefficient accompanying the maximum full power beton concentration, the stem gencater stfety valve prenure wculd still P0t % reached vitbin abcut five minates after turbine trip, if it were reacned at all.

Tne reactor control system worked satisfactorily, except that the autonatic control system did not cut out when the pcwer level was drcppec below 10f power. This malfunction was due to a f aulty relay which has since been replaced. In these tests the control system was turt ed cff by the f operater at low power.

6, i LOES OF C00hANT FLOW A?D NATlHAL CIRCULATION TFSTS l e Loss of coolant flow tests with and without the inertia of the variable frequency generator were performed at zero power, and 3ces of flew i

and natural circulation tests were performed at 8.9,12.h5 and 17.6 Mdt. l In the 17.8 Kdt loss of flow test, scram was actuated by the purp breaker, (

and in other tests the reac'or was manually scrammed at various tiw before the breaker was opened. A .oss of flow test at 20 Kdt was made ou April 2h, I I 1963 by tripping the main coolant pump. This test showed the natural

- circulation was approximately 2-h% of full flow and the maximum clad temperature recorded was 5650F.

l- The main coolant flow coast-down, as indicated by a venturi in the pipe, agreed very well with the predicted curve over the first seven seconds of the transient, except for a lac of about 0.3 seconds. This lag 9 may perribly be in the instrumentation, which is unlikely, but neverthelese it is eteentitlly unexpla4.ned. If the lag exists, it would mean that an added measure of conservatism is inherent in the Itse of flow accident I studier which have been made. The indicated flow seven seconds after

.! t-Appirg the brearam was about 12% (zero power and 1h HWt cu.-ves agree withir ab ut 1%) compared to a prediction of 85; however, the venturi flow indi:.atier. at + hic ficw rate is not cor.tidered reliabic. Time to low fir v ( 2. x 105 lbr/h-) scram c: curred in 0.72 second and zero flow vat reached in 6.3 seconde. It is ceneluded that the flew coast-down is move f averable than was predicted.

The lose of flow following opening of the coupling between the varicble frequer.cy generator and its motcr is much slower than that caused i

j by epening the pump supply breaker due to the inertia of the generator i as expected. For example, two seconds after opening the coupling the flow was 75% of the initial flow. compared to 32% observed two seconds after opening the breaker, and 25% predicted for the latter case.

Natural circulation tests have indicated that substantial :3atural

( circulation is established af ter the loss of pump power. Indications include the fact th t there was no undue rise in clad temperature at ary of twelve points monitored in the core, a reasonable rise in core outlet temperature, and the transmission of temperature changes around the loop.

Freliminary analyses of two cases lead to an estimate of 2% to h% f.hw following operation at 12.h5 Mdt.

l In the first case, the reactor was scrammed three reinuter before the pump war tripped, thus an essentially uniform temperature was attained.

Following pump trip, it took about two minutes to establish circr.lation,

c . - _

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I daring unith time temperatures rose five to tight degrees at both the cere inlet and core outlet. Af ter t ro minut es, the core temperatures slowly de;reased, and a very reugh ertimte of 2% to Li fitw stas obtained, with 8 70 to 25'T temperature drop thr ough the steam generater and appvently I less then 50F rise through the core.

In a second case, the pump was tripped one second after scram.

Cor t- inlet temperatures remained constant, while the rice through the core went frta 100F tr 360F and back to 80F ia one minute, then remained roughly constant. The temperature drop through the steam generator variet between 20CF and f 00F dunts the first five minutes. An analysis of the temperature trantmission between the core and the steam cencrator indicated between 25 an? 3T flow desing this time.

I I It is concluded that although circulation is apparently not as great se the LT to h-1/2% predicted for a steady driving head of 200F temocrature difference, it is sufficient to remove decay heat without any excessive temperature rise in the core.

PUdER COEFFIC1ENT MEASlTREMENTS Power coefficient measuremente were made in December 1962 and l January 1963 during the approach to full power. These coefficients were j obtained both by comparing rod positions at the various power levels and by comparing the main coolant temperature differential at different power levelt with the rede in a fixed position. Those two different methodt gave seme variation in the coefficients obtained and also it wrr4 difficult tc ottsin a true coefficient in the tero to 50% power range because of the tine required to roll the turbine. The power coefficients obtained dr. ring 1 these tests are shown below:

Power Range Powerpoefficient i 10-4AL/%

0-hh.5 -1.13 to -1.28 0-62.25 -1.01 to -1.1h I

hb.5-66.75 -0.8h to -0.91 75 6-S9.0 -0.60 to -0.85 A power coefficient measurament made at 55.75% power 1 vel af ter the power The level instrumentation temperature was coefficient at 37.h Mdtrecalibrated with Xenon waswas-3.8 x-1.36 10- x 10- Ak/f.hok/ F .

i 5300 F as comptred with a temperature coefficient of -h.3 x 10-hAk/0F under these same conditions with a clean core. Equilibrium Xenon measured at the 20 Kdt power level amounted to 2 x 10-2A k and occurred 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after ree hing the 20 Mdt power level. The mcxirau:n Xoaon af ter a shutdewn from 17.h Edt occurred 6-1/2 houro af ter shutdown and was mearured as 5.7 x 10-34 k above squ111brium Xenen.

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f THIMML AND '!YTa&l0 MElSlTRLMLh'TS Flt.x wire data and thermal and hydraulie data were taken during l pewer runr made in Iecencu 196? and Anuary 1963 at 6.y, l?.Z,15.1 4 and 17.h Kit power levels. Some of the first ter.perature and flow meastee-nente taken indicated tha L there might be a flow tilt and as a result the reactor was shut down for t.pproximately a week in order to calibrate and j check both the cere inlet and core outlet pitot tubes and temperature detectcr instrw'.ettation. Furthsr rtudy revealed that thert actually is a

, power or thermal t lt in the directier. of the eleven dummy assemblies located on twe ridu of the m e. Westinghouse reports that this power

' tilt war not f redicted in the original PD; analyses which treated the du'ny astenblitt as a hortagenecas mix *ure of steel and water. Subsequent PD' rtedlet, uring a nere refi*d representation of the dumy astenbli m

. give good agreement between mearured and predicted results. Based on this finding, PD2's wer e made for tht3 reference rod configuration fcr the core g

with standard subaesemblies and lith the special fuel subassemblics that I were installed in the core during the month of Hmh 1963 This evaluation

' also showed that the thernal tilt immediately after Martup and prior to Xenon buildup is greater t, hart the thermal tilt after equilibrium lenon l buildup, which indicates a flattening of the power tilt frcm st&rtup to 1 an equilibrium condition.

Additional flur wire measurements and thermal and hydraulic measurements were taken 3e*4 in March at power l a els of 8.9 and 17.h Kdt.

Evaluation of these mtastrements, based on the latest PDQ's, indicates that g

the heat flux factor 'i ts 2.65 as ccmpared with a design value of 3 2h and g

the enthalpy rise far. tor FA}j in the hot channel is 1.98 as compared with a design italue of 2.30. Westinghouse reports that these values have un-certaintiea as hich as 10% ard therefor

  • these numbers could be as high as 2.92 for Fg and 2.18 for Tgij Preliminary evaluation of flux virs measure-j

. ments and themal and hydraulic r.easurements taken recently at the 20 Mdt

power level show that the het channel factors are essentially the same as l those shown above.

REACTOR VI3SEL MT.AStrREMPRTS l

l Thermocouples installed on the outside of the reactor vessel were eboer/ed during this test period. 16t the 20 Kdt power level the me.rimum tenperature rise across the vessel wall was 120F which ie about two thirdt l

< of the temperature rise that had been calculated. Strain gauges mounted i en the outeide of the versel showed a continuous trift after the rea: tor started operating at power. This drif t is belit:.ved to be due t,o rsdiation effects on the strain gauges. These strain gauges were civing data con-l sistent with the calculated vessel stresses durxng the non-nuclear and zero i i power test period.

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