ML22238A087

NRC-2022-000208 - Resp 1 - Final, Sixth Report Complete Compressed

ML22238A087
Person / Time
Issue date:	08/24/2022
From:	NRC/OCIO
To:
Shared Package
ML22238A081	List: ML22238A082 ML22238A084 ML22238A085 ML22238A086 ML22238A087 ML22238A088
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.. 5 Shipments

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 78.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 1, 1970)

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 (125 VDC Inverter Power Supply)

Rectifier (Auxiliary Primary Pump Power Supply)

Volt-pac cores (Sodium System Heater Circuits)

Volt-pac cores (Freon Compressors)

Shaft Seals (Freon Systems)

Leaks * (Auxiliary Na/Air Rx Fan)

Motor Contactor (480 Volt Transformer)

Fan Motor (Argon Vent Vacuum System)

Vacuum Pump

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 Total 1 353

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 Number of Soii Samples Analyzed 6 Number of Water Samples Analyzed 7

(1) Results of Vegetation Anal;tses Average Radioactivit;t Content (pC. I J. gm-ash)

Month Gross AlEha Gross Beta

Au gust <15 964 September 16.6 1471 October <15 1395

Recheck Level 50 1820 Pre-operational Average 13 987

No evidence of Co-60, I-131, or Na-24 was ~bserved.

(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) Results of Water Anal:ises Average Radioactivit;i Contents(\\JC/m2.)

~ Gross Aleha Gross Beta August <l X 10- 8 2.3 X 10- 8 September <l 10- 8 -1* X 1.2 X 10

October <l X 10- 8 3 X 10-S Recheck Level 3 X 10- 8 1.5 X 10-1 Pre-oper9.tional Average <2 X 10- 9 6.1 X 10-8 No Co-60 or Cs-137 were observed above detection limits.

b. Envirorunen ta 1 Film Moni tor=ic:.:n._.g~(.:;.;A:.::u:gg..::u:.::s~t:......::tc:.:h:.::r~o~u;sag~h:......::S~e;.i;:e~t~em=b~e:.::r..!..)

Number of Stations 17 Total Films Analyzed 50* Maximum Radiation Level Reported O millirad/month

Maximum Radiation Level Reported.

during Pre-operational Survey 8 millirad/month

---

* One station destroyed by vanrialism.

c. Personnel Monitoring (1) Number of film badges issued: 53 August September 49 October 59

(2) Personnel Maximum Whole Body Radiation 130 mrem Received (Quarter ending Sept. 30, 1970)* 1' (3) Personnel Maximum Whole Body Radiation 10 mrem Received during October, 1970 (4) Number of Exposures to Radioactivity none Concentrations in Air in Excess of that specified in 10 CFR 20 (5) Number of Radiological Spills or Con l***

tamination Incidents

• • Reported on a Calendar Quarterly basis for July, August, September as per 10 CFR 20, Section 20.39 (ii).

• • • Minor hand contamination during extension rod transfer.

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 Gross Alpha: -8 <l x 10 µci/ml Beta -s µCi/ml* : 1.6 X 10

(4) Volume Discharged (gallons) 1773

• Identifi e d a~ 8Tritium and C-14. No gamma emitters observed above l x 10 µCi/ml.

b. Gaseous Radio a ctive 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

b. Long-lived Gross Beta: 1 X 10-lO µCi/cc

c. Noble Gas Concentrati~n: <l X 10- 9 µCi/cc

d. Halogen Activity: None observed

(4) Volume Discharged 390,600 ft 3

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, Refueling Cell to Shipping Cask Shielding

Shielding was installed which connects the vertical transfer port of the refueling cell to he fuel shipping cask,

i, Freon Unit Test Panel

A test panel was installe!d with instrumentation for routine monitoring of freon unit compressor suction and discharge pressures, load control signals and other parameters,

10. No changes have been made in the plant operating procedures related to transient tests,

11. The planned transient experiments are scheduled to begin in January, 1971.

C. Other Reportable Items

1. Main Primary Pump Flow Trans:i.ent

On October 28, with the reactor operating at 10 MWt, the Manual Balanced Oscillator Tests we:re in progress. When the main primary pump flow controller was shifted from "cascade" to "remote" setpoint, the main primary flow increased quickly from 2000 GPM to 3300 GPM and more slowly to 3700 GPM. The operator switched the controller back to "cascade", to* establish manual control of the flow and re established the 2000 GPM flow. The inlet and outlet sodium tempera ture changes of about 30°F over the approximate six minute duration of the transient were well within the allowable transient limits, The reactor was secured and the problem traced to an improperly wired switch which was intended to select a signal from either the Manual Balanced Oscillator equipment or the Automatic Balanced Oscillator equipment. 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 controller to react as it did when switched to "remote" and seeing zero output from the Balanced Oscillator equipment.

The switch-was removed from the system and the equipment checked out thoroughly. Flow contrc,ller response was normal in both "remote" and "ca::cade".

A Modification Request is being processed for installation of a two position switch to alloi.r selection of either the Manual or Automatic input signal.

Manual Balanced Oscillator j'.ests at 10 MWt were successfully 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 satisfactory. However, these tests demonstrated the normal functioning of the systems and did not provide informallon relative to the behavior with a very short duration scram pulse. Diagnostic tests were conducted after con firmation that all Safety System relays were properly adjusted to eliminate the potential for a recurrence of the "partial scram".

In a series of 72 recorded, 45 to 75 milliseconds 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 sufficient duration to break the voltage to the scram solenoids, that the scram contactor opened and completed its dropout.

3. Auxiliary Primary Pump Power Supply Failure

At 0030 on September 20, 1970 while the reactor was operating at 10 MW, cor:e average temperature of 610°, main primary flow of 4710 gpm, the auxiliary primary flow decreased to zero. The operators immediately initiated reactor shutdown. Investigation revealed smoke in the vicinity of the auxiliary primary pump power supply (volt-pac).

One core of the volt-pac was badly damaged at the*point of contact between the coil and brushes. The core was replaced (as well as another) in the volt-pac and the auxiliary primary was restored to service. The Site Safety Committee reviewed the occurrence and recommended that the volt-pac vendor applications engineer be requested 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 temperature near the volt-pacs. As 3 result of the review by the Applicat:l.ons Engineer, new brush holders are being supplied, increased cooling provided for the volt-pacs, a common neutral installed, and a preventive maintenance program instituted.

4. Reactor Heat Balance - Wide Range Monitor Discrepancy

See Appendix I.

5. +26.5 Volt Battery System Charger Failure

On September 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 sufficient 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 batteries showed that they were fully charged. A means of providing annunciation upon loss of charger current or upon battery discharge is being investigated for all battery systems.

6. Reactor Cover Gas Inlet Line Blockage Oa September 28, 1970, a "High Vacuum Reactor Vessel" alarm was re ceived. Inspection of the pressure recorder on the reactor cover gas ohowed that the cover gas pressure had remained positive. Observ~

ation of the vacuum breaker valve revealed that it did cycle open coincident with opening of the reactor cover gas vent valve. The reactor was scrammed and cover gas pressure increased in 1 psi steps to determine what pressure was necessary to vent the cover gas with out causing an opening of the vacuum breaker valve. This pressure was found to be 9 psig and at this point a change in cover gas venting rate was observed.

It was initially assumed that some obstruction existed in the vent line vapor trap, probably a sodium vapor deposit which was dislodged when cover gas pressure was raised. The vacuum break.er valve was tested and functioned as it should. On September 30, 1970, while the reactor was shutdown and cover gas at approximately 1 psig, the conditions recurred. This time the actual cover gas pressure was measured by placing a gauge on the vent line. The vent line pres-*

sure showed a negative pressure while the normal.cover gas supply pressure showed a positive pressure. (The cover gas supply pressure signal tap is on the raactor cover gas inlet line). The cover gas p~essure was increased and at an indicated 9 psig, a break through nd the indicated pressure rapidly dropped to zero. This occurred a.

experience indicated that the problem was a partial blockage on the cover gas inlet line, not on the outlet.

The inlet line heaters were energized and the line heated to maxi mum temperature to clear the sodium from the line. Heating of the lin e did not eliminate the problem. The refueling cell atmosphere was changed to air and the inlet line was removed for cleaning.

The blockage was determined to be in the inlet line near the lower surface of the outer head. The blockage was cleared using a plumber's snake and the systems returned to normal.

At the completion of the work, a pur of the reactor cov~r gas was performed to obtain an estimate,£ the flow rate through the inlet supply line for comparison purposes at later dates.

The cover gas monitor loop was leak checked, since it provided a possible source of 02 addition to the cover gas inlet line. (At low cover gas pressure, portions of the cover gas monitor loop could be slightly negative). A small leak was located at the inlet side of filter #953 and repaired. Ro tine leak check of this cover gas loop has been established to assure that it remains leak free.

7, Main Secondary Pump Induction Voltage Regulator Binding

To improve the main secondary pump characteristics 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 satisfactorily. (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 "proportional 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 Approximately 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 eposit which was dislodged gallons per minute to 1950 gpm for 10 minutes and then rapidly in acuum breaker valve was creased to 2700 gallons per minute. The flow then slowly returned ptember 30, 1970, while to normal. The reactor was secured and investigation of the erratic approximately l psig, the behavior was initiated, The drive motor was realigned and the cover gas pressure was system returned to service to observe main secondacy pump performance.

ine. The vent line pres-* Two shifts later, while changing main secondary flow in manual, normal.cover gas supply the induction regulator again failed to properly respond to a e cover gas supply pressure flow change signal. The new gears were removed and replaced with let line). The cover gas the original set. The matchup of the new gears appeared to cause d 9 psig, a break through binding. The drive motor was also changed out at this time, ly dropped Lo zero. This a partial blockage on No further problems have been encountered with the main secondary et. pump control. Proper functioning wa~ nbserved durine subsequent the line heated to maxi testing.

the line. - Heating of the 8. Man Access Suit Exhaust Hose refueling cell atmosphere removed for cleaning. On August 25, 1970, during a refueling cell entry to remove the inlet line near the lower positioner motor for repairs, the operator in the cell experienced was cleared using a plumber's an overinflation of the man access suit. The overinflation 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 onl) by pinching off the exhaust hose. Further checks of the could be identified. The overinflation condition could be repeated ver gas inlet line. (At exhaust hoses were conducted and the vacuum exhaust hoses for cover gas monitor loop both suits were found to be worn to the extent that a 180° bend ne leak check of this cover ak was located at the inlet 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

8/5/70 Personnel error - failure to switch ranges on WRM 83 Spurious Low Flow Main Secondary Pump noise spike 84 8/6/70 on controller

8/8/70 Low Flow Main Primary System - during flow 85 oscillation test

8/9/70 Personnel error - Low Flow Main Primary System -

86 secured oscillator power prior to return pump controller to normal 8/11/70 Manual - spurious Primary Sodium Leak Alarm 87 8/30/70 Personnel error - malfunction IRM when drawer 88 pulled to calibrate instrument

9/9/70 High Flux - WRM Ul - spurious transient when 89 switching ranges

9/11/70 Low Flow Main Primary - noise spike on controller 90 when changing flow

9/12/70 Low Flow Main Secondary - noise spike on con 91 troller when changing flow

9/16/70 Spurious low level - reactor sodium. Dirty 92 contacts in power supply

9/19/70 Low Flow Main Secondary - noise spike on controller 93 94 9/26/70 Low Level Rx Sodium - (while venting Main IHX)

95 9/28/70 Manual - Vacuum breaker valve opened

10/16/70 Personnel error - shorted terminals on core outlet 96 lower region T/C while taking manual mV readings

97 10/26/70 Site power loss during thunder-storm

10/27/70 Personnel error - low flow main pri (chunging 98 mode on P-1 controller from cascade to remote)

DEFINITIONS

ABC Air Blast Cooler APS Auxiliary Primary System ARM Area Radiation Monitor

, switch ranges on WRM ASS Auxiliary Secondary System

,dary Pump noise spike Aux. Auxiliary BRDO Breeder Reactor Development Operation CP Corrective Procedure 1 - during flow EM Electro-Magnetic

EP Emergency Procedure Iain Primary System - FCV Flow Control Valve Lor to return pump Fast Reactivity Excursion Device FRED odium Leak Alarm IFA Instrumented Fuel Assembly 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

>r sodium, Dirty PCV Pressure Control Valve PM Preventive Maintenance P'IP Provisional Test Procedure noise spike on controller PVT Primary Vent Tank

le venting Main IHX) Rx Reactor lve opened SRM Source Range Monitor TOP Temporary Operating Procedure terminals on core outlet TP Test Procedure

ing manual mV readings 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 observed that the rela tionship between the neutron instrumentation and reactor he~t balance data is not constant over the entire operating range of che reactor.

The initial evaluation of test data showed that the relationship depends on the reactor coolant temperature and on the reflector segment pattern.

This information was reviewed by the Site Safety Committee, and they concluded that the effect is not an operation anomaly, but an inherent characteristic of the neutron monitor system. They also concluded that the effect does not reduce plant safety margins, since the trip level on the Wide Range Monitor (Wfili) can be adjusted to meet the requirements*

of the Limiting Safety System SettiJg (LSSS)(l) for any planned o~erating conditions.

Additional reactor tests and systems analyses were performed to obtain a better definition of the parameters affecting the ratio of WRN reading to heat balance data (WRM/Q). These tests and analyses showed that the value of WRM/Q will remain constant within.+/-1 1/2% for reactor power levels of 15 MWt and above when the reactor coolant inlet temperature is a nominal 700°F. However, the value of WRM/Q will decrease up to 10%

when the coolant temperature is lowered to 400°F and reactor power is reduced below 15 MWt. This reduction is caused by: (1) changes in the sodium density; (2) changes in reflector segment position and pattern; and (3) thermal expansion of the core support shroud. The thermal expan sion effect is small (less than 1%). Other effects, such as temperature gradients and non-symmetrical flow distribution were considered, but were found to be of no significance.

The relationship between the WR.\\1 calibration and reactor operating con ditions was established. Knowledge of this relationship assures that the requirements of the Technical Specifications with respect to the LSSS and maximum power level can be met for all reactor operating conditions, and the planned test program was resumed. This relationship and the data used to justify it are presented 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:

a. The reactor conditions used to establish the calibration of

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 reactor power accordin~ to heat balance data, (l, 2 ) 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 the shield plug about five feet below the core. ( 3, 4 ) 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 re f 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:

(1) the reference configuration and composition;( 5 )

(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 pro u lems, 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 incre a se in sodium density and the input of the convergeJ flux,,.

from pro t 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 e duces the thermal flux (and boron activation) at the WRM' s by 18.4%.

4. Comparison of Observed and Predicted Effects the best cstimato of taken to be 2% per s The calculated sodium density effect should be reasonably accurate. to the nearest pcrce However, the effect of lowering one or more reflectors is difficult to estimate from calculations based on lowering all ten reflectors. The table below comp A comparison of WRM reading to in-core detector reading during the effects. Column 1 s initial loading to critical shows that lowering all ten reflectors loading, based on th would have a 28% effect on WRM/Q, whereas the calculated effect is the effect due to se 18.4%. The difference between calculated and observed effects may diction based on 2.4

be due to the approximation required for diffusion theory calcula tions (such as using 15% dense steel to mock up void regions) and SODIUM DENSITY ANI to the use of diffusion theory rather than transport theory to describe Compariso r the geometrical flux relationships. to Calif

The calculated effect of lowering one reflector segment would be 1.87., based on the calculated value of 18.4% for ten reflectors.

However, the effect may be as large as 2,4% per segment if it is assumed that the effect is proportional to the change in reflector Sodium density cf worth, since the last segment raised is worth more than the first one raised. 1-1/2 reflectors lowered for tempe Because of these uncertainties, a special test was performed in change at constan which the core excess reactivity was reduced by approximately 1.5$.

Reactor operation after this change permitted comparison of test 1/2 reflector sea data at identical power (5 MWt) and temperature conditions (4~0 to lowered for powcx at constant tempe 750°F) with the reflector segment configuration different by slightly more than one segment. The results of these tests are shown on Total predicted f Figure 4. Due to the reduced loading, the test had to be performed at 5 MWt, which can result in as much as +/-_6% variation due to re Maximum observed peatability of heat balance data at low power levels. (See Section 375°F (Figure 3) 5, below.) However, the results verify that raising one segment will increase the value Jf WRM/Q at given conditions of power and temperature, as predicted by calculations described above. The magnitud~ of the observed effect is larger than predicted, but this difference can be explained on the basis of test accuracy at low power levels and uncertainties in the analytical methods. Other data for changes in power level at constant coolant temperature (see Figure 2) show a smaller effect than predicted. Therefore,

the best estimate of the effect of changes in segment position is taken to be 2% per segment, based on the calculated values, rounded to the nearest percent.

The table below compares the predicted effects to the observed effects. Column 1 shows predicted effects for the normal core loading, based on the calculated value of 1.8% per segment for the effect due to segment position. Column 2 shows the same pre diction based on 2.4% per segment.

SODIUM DENSITY AND REF1.ECTOR POSITION EFFECTS ON WRM/Q:

Comparison of Test Data at 5 MWt, 375°F to Calibration Data at 15 MWt, 7O0°F

Normal Core Loading

Sodium density effect -5.9% -5.9%

1-1/2 reflector segments -2.8 % -3. 6%

lowered for temperature change at constant power

1/2 reflector segment -.9% - 1.2%

lowered for power change at constant temperature

Total predicted effect -9. 6% -10. 7%

Maximum observed effect at 5 MW, -12% -12%

375°F (Figure 3)

S. Estimated Repeatability

Data have been recorded for many combinations of reactor power, temperature level, and coolant flow rate. The repeatability of these data has been determined to be +300 KWt, assuming the WRM reading is correct. Therefore, the observed value of WRM/Q can be expected to fall within a +/-1. 5% band at 20 MWt, and this band would increase to

+6% at 5 MWt. These bands are indicated by the broken lines on Figures 2, 3, 4. These figures illustrate the reason for the large spread in data obtained at low power, and they also show that the high power data can be determined with good precision.

6, Method of Selecting the High Flux Trip Leve l

The high flux trip must be set so that the wide r~nge monitor wi l l be at the trip level when the reactor power is equal to or less than 21 MWt, as determined by heat balance data. (l, 2 ) - Thus, if the value of WRM/Q is 1. O, the trip level cannot exceed 105% on the WRM, (WRM/Q = 1.00 implies that 100% = 20 MWt.)

If the value of WRM/Q is less than 1. 0, the maximum allowable trip leve ~ust be reduced proportionately so that the trip level will not exceed the LSSS. Therefore, the maximu:n allowable WRM trip level will depend on the reactor operating conditions.

D. Safety Analysis

The Wide Range Monitor provides two important functions:

(1) a safety system trip in the event that reactor power exceeds a pre-selected value; (2) an *indication of the operating power level.

The calibration factor of the WRM is significant only when the trip set ting is close to the Limiting Safety System Setting or when the operating power level is close to the maximum allowable value.

The WRM trip setting will be adjusted as necessary to observe the LSSS of 21 MWt at all times, as required by the Technical Specifications.

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 to the design values used for the coolant system. ( 6) 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.