ML22238A087
| ML22238A087 | |
| Person / Time | |
|---|---|
| Issue date: | 08/24/2022 |
| From: | NRC/OCIO |
| To: | |
| Shared Package | |
| ML22238A081 | List: |
| References | |
| FOIA, NRC-2022-000208 | |
| Download: ML22238A087 (28) | |
Text
'!:ABLE OF CONTENTS Page A. Introduction l B. Summary of Plant Operations l
- 1. Operating Data l
- 2. Plant Shutdow~ 1
- 3. Reactor Scran:~ 1
- 4. Cover Gas Activity 1
- 5. Major Items o,f Plant Maintenance, 2 Instrumenta,tion and Control Work
- 6. Surveillance Testing 2
- 7. Radiation Monitoring Program 3
- 8. Off-Site Radioactivity Release and Shipments
.. 5
- 9. Significant Modifications Approved by 6 Facility Ma.nager and Completed During Report Period
- 10. Transient Test Procedures 7
- 11. Schedule for Transient Tests 7 C, Other Reportable Items
- 1. Main Primary Pump Flow Transient 7
- 2. Scram Relays Auxiliary Contacts 8
- 3. Auxiliary Primary Pump Power Supply Failure 9
- 4. Reactor Heat Balance - Wide Range Monitor 9 Discrepancy
- 5. +26.5 Volt Battery System Charger Failure 9
- 6. Reactor Cover Gas Inlet Line Blockage 10
- 7. Main Secondary Pump Induction Voltage 11 Regulator Binding
- 8. Man Access Suit Exhaust Hose 11
- 9. Reactor Vessel Vacuum Breaker Valve 12
- 10. Intermediate Heat Exchanger Performance 12 D. Safety Review and Audit Activities 13 Table I - Reactor Scrams 14 Definitions 15 Appendix I - Report for Item C.4 i
SIXTH QUARTERLY PLANT OPERATION REPORT A. Introduction This report is submitted in fulfillment of the requirements of License DR-15 for the report period of August 1 through October 31, 1970.
B. Summary of Plant Operations These data are the result of reactor operation for the period August 1, 1970 through October 31, 1970.
- 1. Operating Data Reactor Critical 906.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Maximum Power Level 17.5 MW (nominal)
Longest Continuous Run to Date (October 28, 29, 30, 31, November 1, 1970) 78.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />
- 2. Plant Shutdown No major outages were sustained during the report period.
- 3. Reactor Scrams (See Table I)
Equipment 9 Personnel s Manual 2 Total 16
- 4. Cover Gas Activity The cover gas monitor operated during the quarter and data were obtained as a function of reactor power level. These results were supplemented with a cover gas sampling technique which used a SO ml charcoal filter and a millipore filter in addition to the 400 ml cylinder gas sample which was useJ on previous gas samplings.
With the addition of these filters, the sensitivity of the sampling technique was increased sufficiently to observe a background level of Xenon radioactivity. The magnitude of the Xenon corresponded to that anticipated from the low level "tramp" uranium.
On-line spectrometric studies were conducted also which revealed the presence of Neon-23, a short lived (T1 / 2~38 seconds) isotope resulting from fast neutron activation of sodium-23. This isotope has been observed at EBR-Il and RHAPSODIE.
- 5. Major Items of Plant Maintenance, Instrumentation and Control Work A total of 131 equipment malfunctions were corrected, distributed as follows:
Mechanical 43 Electrical 47 Instrumentation 41 131 The following is a tabulation of the significant malfunctions:
Flexible Coupling (Oscillator Angular position servo-trans-mitter)
Motor reductor (Oscillator positioner)
Level Instrumentation (Reactor Vessel)
Fuel Cleaning, hanQling and Inspection Equip. (Refueling Cell Fuel Inspection Equipment)
Contactors (Safety System)
Rectifier (26.5 VDC Battery charger)
Silicon controlled Rectifier (125 VDC Inverter Power Supply)
Volt-pac cores (Auxiliary Primary Pump Power Supply)
Volt-pac cores (Sodium System Heater Circuits)
Shaft Seals (Freon Compressors)
Leaks * (Freon Systems)
Motor Contactor (Auxiliary Na/Air Rx Fan)
Fan Motor (480 Volt Transformer)
Vacuum Pump (Argon Vent Vacuum System)
- 6. Surveillance Testing
- a. Compliance testing was conducted in accordance with the Technical Specifications using LTP's (License Test Procedures) .
Weekly Tests 234 Bi-Weekly Tests 12 Monthly Tests 55 Quarterly Tests 45 Semi-Annual Tests 6 Annual Tests 1 Total 353
- b. Maintenance Calibration Testing was conducted in accordance with the Technical Specifications.
Monthly Calibrations 12 Semi-Annual Calibrations 8 Annual 0 Total 20
- 7. Radiation Monitoring Program
- a. Environmental Sampling (August l through October 31, 1970)
Number of Vegetation Sall'~les Analyzed 15 6
Number of Soii Samples Analyzed Number of Water Samples Analyzed 7 (1) Results of Vegetation Anal;tses Average Radioactivit;t Content (pC.J. I gm-ash)
Gross AlEha Gross Beta Month August <15 964 September 16.6 1471 October <15 1395 Recheck Level 50 1820 Pre-operational Average 13 987
~bserved.
No evidence of Co-60, I-131, or Na-24 was (2) Results of Soil Analyses Average Radioactivity Content (pC./gm)
J.
Gross AlEha Gross Beta
~
August <15 48 September 15.4 43 October 20 22 Recheck Level 32 45 Pre-operational Average 25 34 No evidence of Co-60 or Cs-137 was observed above detec-tion limits.
(3) Resu lts of Wate r Anal :ises Contents(\JC/m2.)
Aver age Rad ioac tivit ;i Gros s Aleh a Gros s Beta
~ 8
<l X 10-8 2.3 X 10-Augu st 8 -1*
<l X 10-1.2 X 10 Sept embe r 8 3 X 10-S Octo ber <l X 10-1 3 X 10-8 1.5 X 10-Rech eck Leve l Pre- oper 9.tio nal 9 10-8
<2 X 10- 6.1 X Aver age ctio n limi ts.
No Co-60 or Cs-137 were obse rved abov e dete
- sag~h
- ......::S~e;.i;:e~t~em=b~e:.::r..!..)
- b. Envirorunenta 1 Film Moni tor=ic:.:n._.g~(.:;.;A:.::u:gg..::u:.::s~t:......::tc:.:h:.::r~o~u 17 Number of Stat ions 50*
Tota l Film s Anal yzed O mill irad/ mon th Maximum Rad iatio n Leve l Repo rted Maximum Rad iatio n Leve l Repo rted 8 mill irad/ mon th duri ng Pre- oper atio nal Surv ey .
-- One stat ion dest roye d by vanr ialis m.
- c. Pers onne l Mon itori ng (1) Number of film badg es issu ed:
53 Augu st 49 Sept embe r 59 Octo ber (2) Pers onne l Maximum Whole Body Rad iatio n 1 130 mrem Rece ived (Qua rter endi ng Sept . 30, 1970 )*
(3) Pers onne l Maximum Whole Body Radi ation 10 mrem Rece ived duri ng Octo ber, 1970 (4) Number of Expo sures to Rad ioac tivit y Con cent ratio ns in Air in Exce ss of that none spec ified in 10 CFR 20 (5) Number of Rad iolo gica l Spil ls or Con- l***
tami natio n Inci dent s st, on a Cale ndar Qua rterl y basi s for July , Augu
- Repo rted (ii) .
Sept embe r as per 10 CFR 20, Sect ion 20.3 9 n rod tran sfer .
- Mino r hand cont amin ation duri ng exte nsio
- 8. Off-Site Radioactivity Release and Shipments
- a. Liquid Radioactive Waste Discharge (1) Number of samples analyzed during q1;arter ending October 31, 1970 8 (2) Number of Liquid Radwaste Discharges 4 (3) Maximum Radioactivity Level Measured
-8 Gross Alpha: <l x 10 µci/ml Beta : 1.6 X 10-s µCi/ml*
(4) Volume Discharged (gallons) 1773
above l x 10 µCi/ml.
- b. Gaseous Radioactive Waste Discharged
{1) Number of samples analyzed during quarter ending October 31, 1970 22 (2) Number of Gaseous Radwaste Discharges 22 (3) Maximum Radioactivity Measured
- a. Long-lived Gross Alpha: <l X 10-12 µCi/cc 1 X 10-lO
- b. Long-lived Gross Beta: µCi/cc 9
- c. Noble Gas Concentrati~n: <l X 10- µCi/cc
- d. Halogen Activity: None observed 3
(4) Volume Discharged 390,600 ft
- c. Radioactivity Shipments -
Date Quantity Amount August 3, 1970 22 fuel rods 12501.20 gm Pu-239 + 241 118.28 gm U-238
- 9. Significant Modifications Approved by Facility Manager and Completed During Report Period
- a. Freon System High Suction Pressure Unloading The number of cylinders which unload on high suction pressure signal was reduced from 6 out of 8 to 2 out of 8. This provides better system recovery from the high suction conditions while providing adequate protection for the compressor motor.
- b. Locked Access Barrier to Reactor Building Air Shaft A locked gate was added to control access to the reactor building air shaft during reactor operations due to radiation dose rate at the primary sodium area door.
- c. Locked Access Barrier to Crane Bay Catwalk A locked gate was added to control access to the area around the Crane Bay Marine Hatch during reactor operation due to radiation dose rate levels.
- d. Auxiliary Primary Pump Power Supply C-ipacitors An additional capacitor was added to the Auxiliary Primary Pump Power Supply to improve the AC Power Factor. This allows operation of the Auxiliary Primary Pump at full rated flow (250 gpm).
Previous operation at reduced flow provided adequate capacity for the power levels at which the reactor had been operated.
- e. Locked Barrier to Gaseous Radwaste Vault A locked barrier was installed to create a shielded controlled access storage for radioactive solid waste.
- f. Level Probes 128-1 and 128-3 Modifications Two replacement level probes were modified to increase the clearance between the probe and the housing. This reduces the possibility of sodium bridging between the probe and surrounding components. Probe 128-1 was modified to lengthen the bottom end cap to provide additional clearance for internal wiring.
- g. Nitrogen Supply to Outdoor Pneumatic Operators To prevent malfunction of outdoor equipment (containment vent-ilation valves and air blast cooler doors) from freezing of moisture in air supply lines, the air supply was replaced with nitrogen from the liquid nitrogen supply.
h, Refuelin g Cell to Shipping Cask Shieldin g Shieldin g was installe d which connects the vertical transfer port of the refuelin g cell to he fuel shipping cask, i, Freon Unit Test Panel A test panel was installe! d with instrume ntation for routine monitori ng of freon unit compress or suction and discharg e pressure s, load control signals and other paramete rs,
- 10. No changes have been made in the plant operatin g procedur es related to transien t tests,
- 11. The planned transien t experime nts are schedule d to begin in January, 1971.
C. Other Reportab le Items
- 1. Main Primary Pump Flow Trans:i.en t On October 28, with the reactor operatin g at 10 MWt, the Manual Balanced Oscillat or Tests we:re in progress . When the main primary pump flow controll er was shifted from "cascade " to "remote" setpoint ,
the main primary flow increase d quickly from 2000 GPM to 3300 GPM and more slowly to 3700 GPM. The operator switched the controll er back to "cascade ", to* establis h manual control of the flow and re-establish ed the 2000 GPM flow. The inlet and outlet sodium tempera-ture changes of about 30°F over the approxim ate six minute duration of the transien t were well within the allowabl e transien t limits, The reactor was secured and the problem traced to an improper ly wired switch which was intended to select a signal from either the Manual Balanced Oscillat or equipmen t or the Automati c Balanced Oscillat or equipmen t. The switch had a third position labeled "OFF" and the wiring error resulted in the "Manual" position actually being the "OFF" position . This caused the flow controll er to react as it did when switched to "remote" and seeing zero output from the Balanced Oscillat or equipme nt.
The switch-w as removed from the system and the equipmen t checked out thorough ly. Flow contrc,ll er response was normal in both "remote" and "ca::cade ".
A Modifica tion Request is being processe d for installa tion of a two position switch to alloi.r selectio n of either the Manual or Automati c input signal.
Manual Balanced Oscillat or j'.ests at 10 MWt were successf ully per-formed without further problems .
- 2. Scram Relays Auxiliary Contacts ask Shieldin Reactor operation at 5 MW for Test Procedure Group III, Static Tests was in progress. Tests with main primary and main secondary h connects the ve. tical transfer flow rates of 800 gpm had been completed and the main secondary c the fuel shipping cask. flow rate was being increased by movement of the flow controller setpoint when a rea.ctor scram occurred at 1355 on September 12, 1970.
Main secondary flow at the time of scram was approximately 1400 gpm.
Flow fluctuations w*ere al>out +/-25 gpm. The reflectors dropped 1ith instrumentation for routine approximately 5 cm, carriage separation occurred on the fine drives, 1pressor suction and discharge power dropped to ap,proximately 3 MW, when an automatic scram reset 1als and other parameters. occurred. An annu~1ciator alarm and the scram event recorder f.ndicated "Low Flo"' Main Secondary". No other event was recorded
, plant operating procedures related or observed before or during the scram, The operator immedi.ately pushed the manual scram button and the ts are scheduled to begin in January, scram was completed. A low pressure freon header trip had been inserted previously in the safety system since one freon unit was not required for the 5 MW operation. The short duration trip from the low flow-main i,econdary completed the two-out-of-three logic for scram. Subsequent investigation revealed that the main con-tacts on the Kl (scram solenoid) contactor were opening a noticeable time before the "Hold-in" (or auxiliary) contacts (through which the operating at 10 MWt, the Manual contactor coil curi:ent flows). With this relath*e opening of con-in progress. When the main primary tact;, on the contac:tor, if a trip signal consisting of a short d from 11cascade 0 to "remote" setpoint, duratior, pulse werEi received by the scram relay (mercury wetted quickly from 2000 GPM to 3300 GPM contacts vith time to open of 3 to 4 milliseconds) the voltage he operator switched the controller could be removed f1:om the scram bus, the main contacts could open, manual control of the flow and re- the sccam bus voltage restored and the main contacts reclosed The inlet and outlet sodium tempera- before the auxiliary contacts opened. Measurements on the 12 the approximate six minute duration contactors in the Safety System with an ohm!!ter reve;;,led that in 9 in the allowable transient limits. of the contactors 1:he auxiliary contacts opened after the main contacts.
problem traced to an improperly to select a signal from either the On September 13, the contacts on all contactors were adjusted to pment or the Automatic Balanced permit the auxilia1ry contacts to open simultaneously with or before ch had a third position labeled the main contacts so that the phenomenon described above would not 1lted in the "Manual" position actually recur. Following these adjustments, the surveillance tests associated
, caused the flow controller to react with the safety sy,,tem were performed to demonstrate the normal 1ote" and seeing zero output from the functioning of the safety system. These tests include:
Q-0-5, Quarte*i:ly Test of the Reactor Control System
, system and the equipment checked
,r response was normal in both M-0-5, Monthly Channel Test Right ,and Left Manual Scram Buttons Manual Containment Isolation 1g processed for installation of a Manual Block Raise of Reflectors
,lection of either the Manual or M-0-9, Monthly Sub-Channel Test and Channel Test Scram Protection Logic ts at 10 MWt were successfully per- Containment Isolation Reflector Block Raise Action 7-
The results from all tests were satisfac tory. However, these tests demonstr ated the normal function ing of the systems and did not provide informal lon relative to the behavior with a very short duration scram pulse. Diagnost ic tests were conducte d after con-firmatio n that all Safety System relays were properly adjusted to eliminat e the potentia l for a recurren ce of the "partial scram".
In a series of 72 recorded , 45 to 75 milliseco nds short duration trips fed into the six Safety System trip chasses, it was demon-strated that in every case in which the pulse width was of sufficie nt duration to break the voltage to the scram solenoid s, that the scram contacto r opened and complete d its dropout.
- 3. Auxiliar y Primary Pump Power Supply Failure At 0030 on Septembe r 20, 1970 while the reactor was operatin g at 10 MW, cor:e average temperat ure of 610°, main primary flow of 4710 gpm, the auxiliar y primary flow decrease d to zero. The operator s immediat ely initiated reactor shutdown . Investig ation revealed smoke in the vicinity of the auxiliar y primary pump power supply (volt-pa c).
One core of the volt-pac was badly damaged at the*poin t of contact between the coil and brushes. The core was replaced (as well as another) in the volt-pac and the auxiliar y primary was restored to service. The Site Safety Committe e reviewed the occurren ce and recommended that the volt-pac vendor applicat ions engineer be requeste d to visit the site in an attempt to correct the cause of the volL-pac failures , and that actions be taken to reduce the ambient temperat ure near the volt-pac s. As 3 result of the review by the Applicat: l.ons Engineer , new brush holders are being supplied ,
increase d cooling provided for the volt-pac s, a common neutral installe d, and a preventi ve maintena nce program institute d.
- 4. Reactor Heat Balance - Wide Range Monitor Discrepa ncy See Appendix I.
- 5. +26.5 Volt Battery System Charger Failure On Septembe r 17, 1970 the battery charger on the +26.5 volt battery system was observed to have failed. The reserve charger was placed on the line, but it does not have sufficie nt output to supply the normal bus load and provide charging current at the same time. The reactor was shutdown until the charger was repaired and the charging current to the batterie s showed that they were fully charged. A means of providin g annuncia tion upon loss of charger current or upon battery discharg e is being investig ated for all battery systems.
e
- 6. Rea ctor Cov er Gas Inl et Lin e Blo ckag um Rea ctor Ves sel" alar m was re-Oa Sep tem ber 28, 197 0, a "Hig h Vacu re reco rde r on the reac tor cov er ceiv ed. Insp ecti on of the pre ssu ssu re had rem aine d pos itiv e. Obs erv~
gas ohowed tha t the cov er gas pre reve aled tha t it did cyc le open atio n of the vacuum brea ker valv e cto r cov er gas ven t valv e. The coin cide nt with ope ning of the rea pre ssu re incr ease d in 1 psi step s rea cto r was scrammed and cov er gas essa ry to ven t the cov er gas wit h-to dete rmi ne wha t pre ssu re was nec um brea ker valv e. Thi s pres sure out cau sing an ope ning of the vacu poi nt a chan ge in cov er gas was foun d to be 9 psig and at this ven ting rate was obs erve d .
e obs truc tion exis ted in the ven t It was ini tial ly assu med tha t som vap or dep osit whi ch was disl odg ed line vap or trap , prob ably a sodi um The vacuum brea k.er valv e was when cov er gas pre ssu re was rais ed.
On Sep tem ber 30, 197 0, whi le test ed and fun ctio ned as it sho uld.
gas at app roxi mat ely 1 psig , the the rea cto r was shut dow n and cov er actu al cov er gas pres sure was con diti ons recu rred . Thi s time the ven t line . The ven t line pre s-*
mea sure d by plac ing a gaug e on the le the norm al.c ove r gas sup ply sure showed a neg ativ e pre ssu re whi re re . (The cov er gas sup ply pre ssu pre ssu re showed a pos itiv e pre ssu gas inle t line ). The cov er gas sign al tap is on the raa cto r cov er icat ed 9 psig , a brea k thro ugh ind p~e ssur e was incr ease d and at an ed pre ssu re rap idly drop ped to zero . Thi s occ urre d a.nd the ind icat lem was a par tial bloc kag e on exp erie nce ind icat ed tha t the prob the out let.
the cov er gas inle t line , not on ed and the line hea ted to max i-The inle t line hea ters wer e ene rgiz from the line . Hea ting of the um mum tem pera ture to clea r the sodi elim inat e the prob lem . The refu elin g cel l atm osph ere line did not line was removed for clea ning .
was chan ged to air and the inle t in the inle t line nea r the low er The bloc kag e was dete rmi ned to be kag e was clea red usin g a plum ber' s
surf ace of the out er hea d. The bloc al.
norm snak e and the syst ems retu rne d to pur of the rea cto r cov~ r gas At the com plet ion of the wor k, a e ,£ the flow rate thro ugh the was perf orm ed to obt ain an esti mat purp oses at late r date s.
inle t sup ply line for com pari son che cke d, sinc e it prov ided a The cov er gas mon itor loop was leak the cov er gas inle t line . (At pos sibl e sou rce of 02 add itio n to of the cov er gas mon itor loop low cov er gas pre ssu re, por tion s ll leak was loca ted at the inle t cou ld be slig htly neg ativ e). A smaRo tine leak chec k of this cov er side of filt er #953 and rep aire d.
assu re tha t it rem ains leak free .
gas loop has bee n esta blis hed to 7, Main Secondary Pump Induction Voltage Regulator Binding To improve the main secondary pump characterist ics for the flow or Vessel" alarm was re- oscillation mode, a gear ratio change on the induction voltage r on the reactor cover regulator was recommended by BRDO Engineering. The new gears remained positive. Observ- were installed and performance checked out satisfactori ly. (A that it did cycle open secondary flow control response test was performed by the Instru-er gas vent valve. The ment Engineer and the results reviewed by program and analysis).
e increased in 1 psi steps Reactor operation was resumed on October 23, 1970 and preparations o vent the cover gas with- were made for performing the first balanced oscillator tests. The er valve. This pressure "reset" and "proportiona l band" settings were changed to values change in cover gas previously established by the Instrument Engineer to improve the response of the main secondary pump controller for operation in oscillator mode.
tion existed in the vent eposit which was dislodged Approximatel y 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later the main secondary pump flow dropped 50 acuum breaker valve was gallons per minute to 1950 gpm for 10 minutes and then rapidly in-ptember 30, 1970, while creased to 2700 gallons per minute. The flow then slowly returned approximatel y l psig, the to normal. The reactor was secured and investigatio n of the erratic cover gas pressure was behavior was initiated, The drive motor was realigned and the ine. The vent line pres-* system returned to service to observe main secondacy pump performance.
normal.cover gas supply Two shifts later, while changing main secondary flow in manual, e cover gas supply pressure the induction regulator again failed to properly respond to a let line). The cover gas flow change signal. The new gears were removed and replaced with d 9 psig, a break through the original set. The matchup of the new gears appeared to cause ly dropped Lo zero. This binding. The drive motor was also changed out at this time, a partial blockage on et. No further problems have been encountered with the main secondary pump control. Proper functioning wa~ nbserved durine subsequent the line heated to maxi- testing.
the line. - Heating of the refueling cell atmosphere 8. Man Access Suit Exhaust Hose removed for cleaning.
inlet line near the lower On August 25, 1970, during a refueling cell entry to remove the was cleared using a plumber's positioner motor for repairs, the operator in the cell experienced an overinflatio n of the man access suit. The overinflatio n caused the man to lose his balance and fall across the reactor head, f the reactor cov~r gas r,*Jncturing the suit. The standby crew immediately entered the he flow rate through the cell and assisted the suited man from the cell. No injuries or sat later dates. ill effects were sustained by the man in cell. Subsequent checks of th~ suit and suit systems were conducted but no malfunctions ked, since it provided a could be identified. The overinflatio n condition could be repeated ver gas inlet line. (At onl) by pinching off the exhaust hose. Further checks of the cover gas monitor loop exhaust hoses were conducted and the vacuum exhaust hoses for ak was located at the inlet both suits were found to be worn to the extent that a 180° bend ne leak check of this cover produced a flattening of the hose with a resulting restriction of that it remains leak free. flow through the hose. This condition existed only at the point immediately adjacent to the fitting which joins the hose to the suit back plate. further downstream, a 180° bend did not cause significant flattening of the hose.
The Site Safety Committee reviewed the event and the findings on the hose condition and recommended that new hoses be obtained and, if possible, strain relief be provided to reduce the bending of the hose at this attachment to the back plate.
- 9. Reactor Vessel Vacuum Breaker Valve Last quarter's report contained an account of the failure of the reactor vessel vacuum breaker valve to function properly. Tests at monthly intervals were proposed to gain assurance that the valve (normally tested quarterly) remained functional. Satis-factory tests of the valve were conducted on August 19 1 September 29 and October 20, 1970.
- 10. Intermediate Heat Exchanger Performance During power operation heat balances were obtained for the maj(>r heat exchange equipment. A comparison was made of the heat tr,ans-fer *coefficients for the intermediate heat exchangers with the following results:
Predicted U, Measured U, BTU/Hr Ft 2 °F BTU/Hr Ft 2 °F Main IHX 1180 1340 4500 GPM 9.8 MW Auxiliary IHX 1140 525 245 GPM LO MW The overall heat transfer coefficient of the main IHX exceeds the design value, but that of the auxiliary IHX is significantly be,low its design value. However, the auxiliary IHX was sized for a 2.5 MWt heat load (for possible future expansion to a 50 MWt plant), while the present normal heat load is 1 MWt or less. The reduced heat transfer has been tentatively attributed to bypass flow around the internal baffles. The performance has not changed since the lo,w value was first observed and does not indicate a progressive problem.
No operational problem exists since the rate of design heat loaid to required heat load is greater than the ratio of design heat transfer coefficient to measured heat transfer coefficient, and system temper-atures can therefore be maintained at their normal values.
D. Safety Review and Audit Activities
- 1. The Safety Review Committee was convened at the SEFOR site on October 21 and 22, 1970.
- 2. Sixteen meetings of the Site Safety Committee were held during this quarter.
- 3. Proposed changes to the Technical Specifications were submitted to the DRL to incorporate limits based on experience gained in operating SEFOR up to 10 MWt. These changes are required by Technical Specifications 3.10.E and 6.6.B.3.
- 4. Three trips were made to the site by members of th~ Safety and Quality Assurance Subsection of BRDO to review plant safety, operating experience, and compliance with the Technical Speci-fications.
TABLE I REACTOR SCRAMS es on WRM Pers onne l erro r - fail ure to swit ch rang 83 8/5/ 70 nois e spik e Spur ious Low Flow Main Seco ndar y Pump 84 8/6/ 70 on con trol ler flow Low Flow Main Prim ary Syst em - duri ng 85 8/8/ 70 osci llati on test System -
Pers onne l erro r - Low Flow Main Prim ary rn pump 86 8/9/ 70 secu red osc illat or power prio r to retu con trol ler to norm al Alarm Man ual - spur ious Prim ary Sodium Leak 87 8/11 /70 draw er Pers onne l erro r - malf unct ion IRM when 88 8/30 /70 pull ed to cali brat e instr ume nt t when High Flux - WRM Ul - spur ious tran sien 89 9/9/ 70 swit chin g rang es con troll er Low Flow Main Prim ary - nois e spik e on 90 9/11 /70 when chan ging flow on con-Low Flow Main Seco ndar y - nois e spik e 91 9/12 /70 trol ler when chan ging flow Dirt y Spur ious low leve l - reac tor sodi um.
92 9/16 /70 cont acts in power supp ly on con trol ler Low Flow Main Seco ndar y - nois e spik e 93 9/19 /70 IHX)
Low Leve l Rx Sodium - (whi le vent ing Main 94 9/26 /70 Man ual - Vacuum brea ker valv e open ed 95 9/28 /70 core outl et Pers onne l erro r - shor ted term inal s on ual mV read ings 96 10/1 6/70 lowe r regi on T/C whil e takin g man Site power loss duri ng thun der- storm 97 10/2 6/70 ngin g Pers onne l erro r - low flow main pri (churemo te) 98 10/2 7/70 mode on P-1 con trol ler from casc ade to DEFINITIONS ABC Air Blast Cooler APS Auxiliary Primary System ARM Area Radiation Monitor ASS Auxiliary Secondary System
, switch ranges on WRM Aux. Auxiliary
,dary Pump noise spike BRDO Breeder Reactor Development Operation CP Corrective Procedure 1 - during flow EM Electro-Magnetic EP Emergency Procedure Iain Primary System -
Lor to return pump FCV Flow Control Valve FRED Fast Reactivity Excursion Device IFA Instrumented Fuel Assembly odium Leak Alarm IFST Irradiated Fuel Storage Tank on IRM when drawer IHX Intermediate Heat Exchanger ent IRM lntermP.diate Range Monitor us transient when LTP License Test Procedure MPS Main Primary System
.se spike on controller MSS Main Secondary System NFSV New Fuel Storage Vault 1oise spike on con- PAP Pump-Around-Pump PCV Pressure Control Valve
>r sodium, Dirty PM Preventive Maintenance P'IP Provisional Test Procedure noise spike on controller PVT Primary Vent Tank le venting Main IHX) Rx Reactor SRM Source Range Monitor lve opened TOP Temporary Operating Procedure terminals on core outlet Test Procedure
- ing manual mV readings TP WRM Wide Range Monitor inder-storm
, main pri (changing
>ID cascade to remote)
APPENDIX I EFFECT OF OPERATING CONDITIONS ON THE NEUTRON FLUX INST~UMENT CALIBRATION A. Summary During the planned test program at SEFOR, it was observe d that the rela-tionshi p between the neutron instrum entation and reactor he~t balance data is not constan t over the entire operati ng range of che reactor .
The initial evaluat ion of test data showed that the relation ship depends on the reactor coolant tempera ture and on the reflect or segment pattern This informa tion was reviewe d by the Site Safety Commit tee, and they t
conclud ed that the effect is not an operati on anomaly , but an inheren charac teristic of the neutron monitor system. They also conclud ed that the effect does not reduce plant safety margins , since the trip level on the Wide Range Monitor (Wfili) can be adjuste d to meet the requirem ents*
ng of the Limitin g Safety System SettiJg (LSSS)( l) for any planned o~erati conditi ons.
Additio nal reactor tests and systems analyse s were perform ed to obtain a better definit ion of the paramet ers affecti ng the ratio of WRN reading to heat balance data (WRM/Q). These tests and analyse s showed that the value of WRM/Q will remain constan t within .+/-1 1/2% for reactor power levels of 15 MWt and above when the reactor coolant inlet tempera ture 10%
is a nominal 700°F. However, the value of WRM/Q will decreas e up to when the coolant tempera ture is lowered to 400°F and reactor power is reduced below 15 MWt. This reducti on is caused by: (1) changes in the sodium density ; (2) changes in reflect or segment positio n and pattern and (3) thermal expansi on of the core support shroud. The thermal expan-sion effect is small (less than 1%). Other effects , such as tempera ture were gradien ts and non-sym metrica l flow distrib ution were conside red, but found to be of no signific ance.
The relation ship between the WR.\1 calibra tion and reactor operati ng con-ditions was establis hed. Knowledge of this relation ship assures that the requirem ents of the Technic al Specifi cations with respect to the LSSS and maximum power level can be met for all reactor operati ng conditi ons, data and the planned test program was resumed . This relation ship and the used to justify it are present ed in this report.
B. Conclusions Analyses of the test data and calculatlons which were made led to the con.cl us ions presented below *
.1, The ratio, WRM/Q, depends on three parameters:
The reactor conditions used to establish the calibration of a.
the WRM.
- b. Reactor coolant inlet temperature.
- c. Reflector segment position.
2, The value of WRM/Q can be determined from Figure 1 for all reactor.
operating conditions. (Segment position is directly related to reactor power and coolant temperature,)
- 3. The observed effects agree reasonably well with the -predicted effects, as shown by Figures 2 and 3.
4, The Limitlng _Safety System Setting of 21 MWt, which is based on 2
reactor power accordin~ to heat balance data, (l, ) can be observed for all reactor operating conditions by proper adjustment of the WRM trip level.
C. Discussion
- 1. Location of WRM Detector~
The neutron detectors for tne Wide Range Monitors are located in 3 4 the shield plug about five feet below the core. ( , ) This location provides the thermal flux level (with a near zero gradient) which is required for the detectors. This location results in a linear response with power level, except for the effects of changes in sodium density and reflector position.
2, Thermal Expansion Effects The distance between the core centerline and the WRM detectors will decrease by about 0.3 inch when the coolant temperature is increased from 400°F at zero power to 760°F average temperature at rated power.
This would cause the normalized WRM reading to change by less than 1%, The effect is small and since it cannot be measured separately, it is included with the effects due to changes in coolant inlet temperature.
- 3. Calculated Effects 0 f Sodium Density and Reflect6r Position Calculations were performed to determine the influence of sodium density and ref lector segment position on the WRM readings. Using two- dimensional diffusion theory in four energy groups with a fixed fission source distribution, the neutron flux at the SEFOR wide range monitors (WRM's) was computed for the following problems:
5 (1) the reference configuration and composition;( )
(2) problem (1) with sodium density increased 6 percent in all regions, to represent sodium density changes going from 760°F to 400°F;
-(3) problem (1) with the radial reflectors in the down position, In all three proulems, the power level remained fixed, bu_t the source distribution in problem (3) was modified to correspond with the re-flector down situation .
With the exception of the adjustment to the cross sections to reflect the increa se in sodium density and the input of the convergeJ flux ,,.
from prot lem (1) as the initial flux guess, the conditions for problem (2) were identical to those of problem (1).
In problem (3), the radial reflectors were lowered such that the top of the reflectors were 3.572 cm (1,4 inch) above the bottom of the core, The region above the reflectors adjacent to the core was replaced by the void material (15 percent SS) used throughout the void regions of the reference problem. To maintain the mesh descrip-tion of the reference problem and yet preserve the effect of the reflector le~gth of 99,06 cm (39 inches), the density of the reflec-tor in the interval at the bottom of the reflector was reduced so as to preserve the proper number of reflector atoms.
These calculations showed that a 6% increase in sodium density (which corresponds to a reduction in sodium temperature from 760°F to 400°F) causes a 6.5% decrease in the thermal flux (and boron activation) at the WRM's. For a constant total fission source, the reflector down configuration r educes the thermal flux (and boron activation) at the WRM' s by 18.4%.
ts the best cstim ato of
- 4. Comp arison of Obser ved and Pred icted Effec taken to be 2% per s ld be reaso nably accu rate.
The calcu lated sodiu m dens ity effec t shou to the neare st pcrce more refle ctors is diffi cult However, the effec t of lowe ring one or lowe ring all ten refle ctors . The table below comp to estim ate from calcu latio ns based on detec tor readi ng durin g the effec ts. Column 1 s A comp arison of WRM readi ng to in-co re lowe ring all ten refle ctors loadi ng, based on th initi al loadi ng to criti cal shows that as the calcu lated effec t is the effec t due to se would have a 28% effec t on WRM/Q, where and obser ved effec ts may dicti on based on 2.4 18.4% . The diffe rence betwe en calcu lated diffu sion theor y calcu la-be due to the appro xima tion requi red for mock up void regio ns) and tions (such as using 15% dense stee l to SODIUM DENSITY ANI than trans port theor y to desc ribe to the use of diffu sion theor y rathe r Comp arisor the geom etrica l flux relat ionsh ips. to Calif refle ctor segm ent would be The calcu lated effec t of lowe ring one 18.4% for ten refle ctors .
1.87. , based on the calcu lated value of 2,4% per segm ent if it is However, the effec t may be as large as to the chang e in refle ctor assum ed that the effec t is prop ortio nal Sodium dens ity cf worth more than the first worth , since the last segm ent raise d is one raise d. 1-1/2 refle ctor s lower ed for tempe ial test was perfo rmed in Becau se of these unce rtain ties, a spec chang e at const an ed by appro xima tely 1.5$.
which the core exces s reac tivit y was reduc itted comp arison of test Reac tor opera tion after this chang e perm 1/2 refle ctor sea eratu re cond ition s (4~0 to lower ed for powcx data at iden tical powe r (5 MWt) and temp at cons tant tempe gurat ion diffe rent by sligh tly 750°F ) with the refle ctor segm ent confi these tests are shown on more than one segm ent. The resu lts of Total predi cted f the test had to be perfo rmed Figur e 4. Due to the reduc ed loadi ng, as +/-_6% varia tion due to re-at 5 MWt, which can resu lt in as much Maximum obser ved power level s. (See Secti on 375°F (Figu re 3) peat abili ty of heat balan ce data at low that raisi ng one segm ent 5, below .) However, the resu lts verif y cond ition s of power and will incre ase the value Jf WRM/Q at given ns descr ibed above . The temp eratu re, as pred icted by calcu latio r than pred icted , but this magn itud~ of the obser ved effec t is large of test accur acy at low diffe rence can be expla ined on the basis anal ytica l metho ds. Othe r power level s and unce rtain ties in the tant coola nt temp eratu re data for chang es in powe r leve l at cons pred icted . Ther efore ,
(see Figur e 2) show a smal ler effec t than the best estim ate of the effec t of chang es in segme nt posit ion is value s , round ed taken to be 2% per segme nt , based on the calcu lated to the neare st perce nt.
the obser ved The table below compa res the predi cted effec ts to norma l core effec ts. Column 1 shows predi cted effec ts for the segme nt for loadi ng, based on the calcu lated value of 1.8% per the same pre-the effec t due to segme nt posit ion. Column 2 shows dictio n based on 2.4% per segme nt .
SODIUM DENSITY AND REF1.ECTOR POSITION EFFECTS ON WRM/Q Comp arison of Test Data at 5 MWt, 375°F to Calib ration Data at 15 MWt, 7O0°F Normal Core Loadi ng
-5.9% -5.9%
Sodium densi ty effec t
-2.8% -3. 6%
1-1/2 refle ctor segme nts lower ed for tempe rature chang e at const ant power
- .9% - 1.2%
1/2 refle ctor segme nt lower ed for power chang e at const ant tempe rature
-9. 6% -10. 7%
Total predi cted effec t
-12% -12%
Maximum obser ved effec t at 5 MW, 375°F (Figu re 3)
S. Estim ated Repe atabi lity of react or powe r, Data have been recor ded for many comb inatio ns repe atab ility of these temp eratu re level , and coola nt flow rate. The ing the WRM readi ng is data has been deter mine d to be +300 KWt, assum can be expe cted to corre ct . Ther efore , the obser ved value of WRM/Q band would incre ase to fall with in a +/-1. 5% band at 20 MWt , and this the broke n lines on
+6% at 5 MWt. These bands are indic ated by reaso n for the large Figur es 2, 3, 4. These figur es illus trate the also show that the sprea d in data obtai ned at low powe r, and they prec ision .
high powe r data can be deter mine d with good 6, Method of Selec ting the High Flux Trip Level r~nge moni tor wil l The high flux trip must be set so that the wide is equa l to or less than be at the trip leve l when the react or powe r 2 - Thus , if the value
)
21 MWt , as deter mine d by heat balan ce data. (l, d 105% on the WRM, of WRM/Q is 1. O, the trip leve l canno t excee (WRM/Q = 1.00 impl ies that 100% = 20 MWt.)
um allow able trip If the value of WRM/Q is less than 1. 0, the maxim the trip leve l will leve ~ust be reduc ed prop ortio natel y so that allow able WRM trip not excee d the LSSS. Ther efore , the maximu:n ition s .
leve l will depen d on the reac tor oper ating cond D. Safet y Anal ysis funct ions:
The Wide Range Moni tor provi des two impo rtant or powe r excee ds (1) a safet y syste m trip in the even t that react a pre-s elect ed value ;
(2) an *indi catio n of the oper ating powe r level nt only when the trip set-The calib ratio n facto r of the WRM is sign ifica Setti ng or when the oper ating ting is close to the Limi ting Safe ty Syste m value .
powe r level is close to the maximum allow able sary to obser ve the LSSS The WRM trip setti ng will be adjus ted as neces nical Spec ifica tions .
of 21 MWt at all times , as requi red by the Tech Operation at power levels in the range of 17, 5 MWt to 20 MWt will require coolant inlet temperatures of a nominal 700°F value, due 6
to the design values used for the coolant system. ( ) WRl1 calibration data has been obtained at power levels to 17.S MWt with 700°F coolant temperatures, and additional data will be obtained up to 20 MWt. Thus the value of WRM/Q will be known for the conditions at the maximum allowable power.
Therefore, it is concluded that the requirements of the LSSS and maximum power limit can be met and that the WRM variation with operating conditions does not adversely affect plant safety.
References:
- l. SEFOR Technical Specifications, Pages 2.2-2,3.
- 2. SEFOR Technical Specifications, Definitions 1. 5 and- 1. 6.
- 3. SEFOR FDSAR, Paragraph 10.2.3.2.2, Page 10-9.
- 4. SEFOR FDSAR, Supplement 9, Figure 1- 4.
- 5. SEFOR FDSAR, Supplement 9, Appendix A.
- 6. SEFOR FDSAR, Section 5.2.
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