ML22238A084
| ML22238A084 | |
| Person / Time | |
|---|---|
| Issue date: | 08/24/2022 |
| From: | NRC/OCIO |
| To: | |
| Shared Package | |
| ML22238A081 | List: |
| References | |
| FOIA, NRC-2022-000208 | |
| Download: ML22238A084 (28) | |
Text
SOUTHWEST EXPERIMENTAL FAST OXIDE REACTOR EIGHTH QUARTERLY PLANT OPERATION REPORT Prepared by:
R. V. Myers W. P. Ku11kel C. E. Russell J. o. Arterburn Manager, SEFOR Facility Prepared for transmitt,L to the Division of Reactor Licensing, UnLed States Atomic Energy Conmission, as required by License DR-15.
BREEDER RF.ACTOR DEVELOPMENT OPERATION GENERAL ELECTRIC COMPANY SUNNYVALE, CALIFORNIA 94086
,....,."""'""""....-----~---------- -
A.
B.
- c.
D, TABLE OF CONTENTS Introduction Suumary of Plant Operations
- 1.
- 2.
3,
- 4.
- s.
- 6.
- 7.
- 8.
- 9.
Operating Data Plant Shutdowns Reactor Scrams Cover Gas Activity Major Items of Plant Maintenance, Instrumentation and Control Work Surveillance Testing Radiation Monitoring Program Off-Site Radioactivity Release and Shipments Significant Modifications Approved by Facility Manager and Completed During Report Period Other Reportable Items 1,
- 2.
- 3.
- 4.
- s.
- 6.
- 1.
- 8.
Safety System Relay K5 Malfunction Preon System Time Delay Malfunctions Auxiliary ABC Door Malfunction Safety System Relay Malfunctions Instrument Nitrogen Supply to Reactor Overflow Valve Violation of Limiting Condition for Operation -
Technical Specification 3.12.B.5 Reactor Vessel Head Bolt Stresses Awtiliary Intermediate Heat Exchanger Performance Safety Review and Audit Activities Table I - Reactor Scrams Definitions i
1 1
1 l
1 1
2 2
4 6
s 15 15 15 16 17 18 19 19 20 23 24 25-26
A.
B.
EIGlITH QUARTERLY PLANT OPERATION REPORT Introduction This report is submitted in fulfillment of the requirements of License DR-15 for the report period of February 1, 1971 thru April 30, 1971.
Summary of Plant Operations These data are the result of reactor operation for the period of February 1, 1971 thru April 30, 1971.
- 1.
Operating Data Reactor Critical Maximum Power Level Longest Continuous Run to Date (December 27, 1970 thru January 3, 1971) 263.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 20.0 MW 153.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
- 2.
Plant Shutdowns
- 3.
- 4.
The reactor was shutdown on February 7, 1971 to perform modifi-cations necessary to convert the plant for Fast Reactivity Excursion Device (FRED) testing, Operation resumed on February 25, 1971.
On '.iarch 14, 1971 the reactor was shutdown for an outage to evaluate reactor vessel head bolt stresses during heating and cooling cycles.
The reactor was restarted on April 2~,. 1971 to conclude the outage.
Reactor Scrams (See Table I)
Equipment Personnel Manual Other (Loss of Site Power)
Cover Gas Activity Total 2
l 13*
1 17 The Cover Gas Monitor was in service during the quarter, and indicated no anomalous fission gas activity.
Cover gas samples were obtained to quantitatively measure the isotopic constituents. These samples consisted of *:outine monthly cover gas analyses, special experiments to further refine sampling and identification techniques, and pre and post FRED transient samples.
These results indicate no fission gas levels other than those normally anticipated from tramp uranium and/or pin-hole cladding penetrations.
- Includes 11 planned as part of Test Program.
on
- 5.
Major Items of Plant Maintenance, Instrumentation and Control Work
- 6.
A total of 198 malfunctions were corrected, distributed as follows:
Mechanical Electrical Instrumentation Significant malfunctions included:
Auxiliary Inlet RTD Sodium Piping Heater Circuits Fuel Grapple Load Cell Multi-ton Load Cell 54 71 73 198 Safety Syetem Relays Main Primary Sodium Pump Generator Stator Exciter Regulator Sodium Leak Detectors Auxiliary ABC Doors Deep Well Water Pump Sodium Level Probe Freon Condenser Fan Motor Surveillance Testing
- a.
Compliance testing was conducted in accordance with the Technical Specifications using LTP's (License Test Procedures).
Weekly Tests Bi-Weekly Tests Monthly Tests Quarterly Tests Semi-Annual Tests Annual Tests Total 231 12 53 45 8
_3 352
- b.
Maintenance Calibration Testing was conducted in accordance with the Technical Specifications Monthly Calibrations Semi-Annual Calibrations Annual Total 6
8 0
14
(
' *i I
i
Work llows:
edures),
lance
- c.
Saaples of the sodium from the SEFOR primary loop were obtained March 17, 1971.
One set of samples was retained on site for radiochemical analysis, The other set of 3-sample cups were sent to Vallecitos Nuclear Laboratory, where they were analyzed for metallic impurities as well as radiochemically.
These results are summarized below, and deviations from the levels measured previously are considerec to be within normal limits.
Metallic Constituents of SEFOR Sodium in Sample Obtained March 17. 1971 Element Concentration Fe 6
Si 20 Ca 5
Mg 6
Cr 6
Mn 4
B 2
Mo, v, Be, Bi
<12 Al, Sn
<2 L
<20
< Represents lower limit of detection for instrument
~ppm) used.
The ~arbon content was 13 ppm, while the U-235 was 4 ppb.:t 2 ppb and the U-238 was 200 ppb.:t 50 ppb.
Radiochemical Data (as of April 101 1971)
Nuclide dpm/16 8!!! sample Na-22 l.9x 106 Aa-110 4.9 X 104 Sb-124 Llx 10"'
These r~ults were obtained by counting the gross Slllllple with a Ce-Li detector. Following che111ical separation, an aliquot was counted for I-131 and results showed the level to be below the limits of detection.
- 7.
Radiation Monitoring Program
- a.
Environmental Sampling (February 1, 1971 through April 30, 1971)
- 1)
Number of Vegetation Samples Analyzed Number of Soil Samples Anclyzed Number*of Water Samples Analyzed Results of Vegetation Analyses Average Radioactivity Content (pCi/gm-ash)
Month AlJ:!ha February 32.6 March 30 April 20.7 Recheck Level 50 Pre-Operational Avg.
13 15 6
9 Beta 1497 1657 1640 1820 987 No evidence of Co-60, I-131, or Na-24 was observed in the vegetation samples before transmittal from the site.
- 2)
Results of Soil Analyses
- 3)
Average Radioactivity Content (pC1/gm)
Month AlEha Beta February 26 29 March
<15 55 April 20 23.3 Recheck Level 32 45 Pre-Operational Avg.
25 34 No evidence of Co-60 or Cs-137 was observed above detection limits.
Results of Water Analyses Average Radioactivity Content (~C/mi)
Month AlJ:!ha February March April Recheck Level Pre-Operational Avg.
<l x 10-8 1.9 X 10-8
<l x 10-8 3 X 10-8
<2 X 10-9 Beta 3.2 X 10-8
<3 X 10-8 3.2 X lQ-8 1.5 X l0-7 6.1 X 10-8 No Co-60 or Cs-137 were observed above detection limits.
- b.
- c.
Environmental Film Monitoring (February 1, 1971 thru April 30, 1971)
Number of Stations Total Films Analyzed 17 51 Maximum Radiation Level Reported Maximum Radiation Level Reported 12 millirad/quarter*
8 millirad/month during Pre-Operational Survey Personnel Monitoring
- 1)
Number of film badges issued:
- 2)
- 3)
- 4)
- 5)
February March April 48 48 50 Personnel Maximum Whole Body Radiation Received (Quarter ending March 31, 1971)**
Personnel Maximum Whole Body Radiation Received during April, 1971 Number of Exposures to Radioactivity Concentrations in Air in Excess of that specified in 10 CFR 20 Number of Radiological Spills or Con-tamination Incidents 130 mrem 250 mrem None None The report for April, 1971, showed 12 millirad at 2 stations. However, the control badge showed 4 millirad, indicating possible irradiation of the film in transit or storage.
Reported on a Calendar Quarterly basis for January, February and March as per 10 CFR 20, Section 20.39 (ii).
- 8.
Off-Site Radioactivity Release and Shipments (February 1 1 1971 thru April 301 1971)
- a.
Liquids
- 1)
Fission and Activation Products (except tritium) a)
Total curie activity released (Ci) b)
Total volume of liquid waste discharged (gallons) c)
Total volume of dilution water (gallons) d)
Volume average concentration at discharge point ~
x 264 (µCi/m.e.)
(Le) e)
MPC used (µCi/m.e.)
f)
Percent of limit (%)
g)
Maximum concentration released, averaged over not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (µCi/m.e.)
- 2)
- 3) a)
Total curie activity released (Ci) b)
Volume average concentration at dis-charge point: lb& x 264
(µCi/m.e.)
(Le) c)
Percent of limit(%)
Estimated Carbon 14 release (Ci) 4.6 X 10-7 12 X 103 62 X 103
<1.9 X 10-9
-7 1.0 X 10
<1.9
<5 X 10-8
-2 9.4 X 10 4.0 X 10-4 13.4
-4 3.5 X 10 Note: All liquids are released to a tile field. Measured con-centrations refer to values at the point of discharge into the tile.field.
- b.
Gaseous
- 1)
Noble and Activation a) b)
c) d)
e) f)
g) h)
i)
Total curie activity released (Ci)
Total volume of gas released (ft3)
Time average release rate (b.l.a) (µCi/sec) 7.9 MPC used (µCi/m.e.)
Licensed limit for annual average (µCi/sec)
Percent of annual limit(%)
Maximum hourly average release rate (µCi/sec)
Licensed limit for hourly average (µCi/sec)
Percent of hourly limit(%) 1.37 155,700 0.174 2 X lQ-S 800 2.2 X 18 3400 0.53
ruary 1 1 1971 thru tritium) 1rged llons)
- charge eraged n
4.6 X 10-7 12 X 103 62 X 103
<1.9 X 10-9
-7 1.0 X 10
<1.9
<5 X 10-8
-2 9.4 X 10 4.0 X 10-4 13.4
-4 3.5 X 10 eld. Measured con-iischarge into the
(µCi/sec)
(µCi/sec) e (µCi/sec)
(µCi/sec) 1.37 155,700 0.174 2 X 10-8 800
-2 2.2 X 10 18 3400 0.53
- c.
- d.
- e.
- 2)
Halogens with half-lives >8-days and particulates with half-lives >8-days.
a) b)
c) d)
e) f)
g) h)
Total curie activity released (Ci)
Time average release rate (b.2.a) (µCi/sec) 7.9 MPC used* (µCi/mi)
Licensed limit for annual average (µCi/sec)
Percent of limit (%)
Maximum hourly average release rate (µCi/sec)
Licensed limit for hourly average (µCi/sec)
Percent of hourly limit(%)
- 7
<44 X 10 _8
<5,6 X 10 1.0 X 10-~0
- 5. 6 X 10-
<l.0 X 10- 3
<9.4 X 10-7 5,6 X 10-2
<l. 7 X 10-3 Number of Samples analyzed during quarter ending April 30, 1971 Liquid Gaseous 11 10 Number of Radwaste Discharges Liquid Gaseous Radioactive Shipments 7
10 Quarterly Summary of Radioactive Shipments February 1 1 1971 thru April 30 1 1971 Date Description 3/18/71 23 Boxes Low-Level Solid Waste 10 Steel Rods (C0-58, MN-54
& Other Activation Isotopes) 3/16/71 Polyethylene Flask Containing
<5µCi of Tritium 3/19/71 3/25/71 4/2/71 4/19/71 Three Activated Head Bolts (SS Activation Isotopes)
Approximately 60 gm Na (Primarily Na-22)
One Irradiated FRED Slug (SS Activation Isotopes)
Compensated Fission Chamber (Depleted Uranium)
To Nucl. Eng. Co.
Walnut Creek, California GE(Vallecitos)
GE(San Jose)
GE(Vallecitos)
GE(San Jose)
GE(San Jose)
Radioactive Content
<25 mCi
<20 mCi
<5µCi
<0.5 mCi
<l mCi
<1 mCi
<l mCi Based on the possible presence of I-131.
- 9.
Significant Modifications Approved by Facility Manager and Completed During Report Period.
- a.
Modifications Associated with Fast Reactivity Excursion Testing.
- 1)
The fuel rod disassembly station in the refueling cell was modified to provide a means of remotely installing or removing the FRED poison slugs from the FRED actuator rod.
The original function of the disassembly station was retained making it a dual purpose device.
- 2)
Signals from the lift-off switches and the proximity switches on the FRED device were provided to the Data Acquisition System for inclusion in the data collected during the transient tests. The Data Acquisition System circuits are isolated from the associated relays by diodes to prevent reverse current flow and are fused to prevent possible loss of interlock protection for positioner.
- 3)
An Excursion Mode was integrated into the Reactor Safety System which provided the following:
a)
An interlock to prevent firing of FRED by preventing voltage to relays K4 and KS (and FRED latch switches) unless: Main Primary flow >4300 gpm, Switch is in "Excursion" Mode.
b)
An automatic reactor scram when FRED fire switch is moved from "off" to "fire".
c)
An additional time delay ~400 milliseconds) for reactor scram accomplished by an RC circuit in parallel with the coil of each of the six scram contactor relays (Kl) when switch is in "Excursion" mode, only.
When switch is in the "Normal" or "Oscillator" mode K4 and KS contacts open to disconnect the RC circuit.
d)
A block reflector raise accomplished by a pair of normally open contacts in block reflector raise circuit when switch is in the "Excursion" mode.
e)
A scram delay time test circuit which involves a series resistor net-work with a 1.5V battery acti-vated by a set of scram contactor relay contacts.
l
)
l
- 4)
After the installation the following tests were performed to verify proper operation of the modified system:
- 5) a)
Reflector drop measurements in which the delay from drop out of relay K2 (scram contactor hold-in) and the initiation of reflector movement was measured for reflectors No. 3 and No. 8 in the "Normal" and "Oscillator" modes, and in the "Excursion" mode.
b)
Pre-operation Test No. 59, Para. 7.4.4 (2/20/71)
This pre-operational test involved a series of measurements to determine the scram delay time from initiation of FRED fire to scram contactor drop-out for each of the six scram contactors in the "Normal" and "Oscillator" modes and in the "Excursion" mode.
c)
License Test Procedures:
M-0 Monthly Test of Manual Scram Buttons, Manual Rod Block and Manual Contain-ment Isolation M-0 Monthly TE>..st of Reactor Safety System Q-0 Reactor Control System The photo-electric cells for the FRED were repositioned due to interference with access to the thru-head re-fueling ports in their original location. The new locations were verified to be free of interference and the proper functioning of the photo cells was verified following the modification.
- 6)
Following the modification to the disassembly station and approximately 50 test firings. the FRED actuator
- 7) rod and poison slug were disassembled.
Reassembly could not be accomplished due to the 81Dall ('li().002 11 diametrical) clearance between the matin~ parts. The clearance was increased to 0.004" by enlarging the hole in the adaptor
- piece, The gauge originally provided on the FRED Transfer Chamber (1/4% accuracy) was relocated to the accumulator charging line downstream of the charging valve. This gauge required frequent recalibration when installed on the transfer chamber due to zero shifts caused by bumping or jarring.
b, C,
The transfer chamber was equipped with a more rugged gauge calibrated at normal operating pressure (2000 psig),
- 8)
The overall length of the FRED poison slug connector piu was reduced by 1/16".
During the assembly/disassembly of the FRED poison rod from the extension rod in the Disassembly Station, it was observed that the tines on the connector pin were too long.
After bending, the tines extended out of the chamber and beyond the diameter of the adapter piece and created interference with the Disassembly Station.
In addition, the tines were bending near the root, resulting in a deformation of the tines which required excessive force to extract the pin.
- 9)
A shielded container has been designed and fabricated to provide for storage of irradiated poison slugs external to the Refueling Cell, Drain System From Moisture Collection Tanks to Liquid Radwaste.
A system by which the condensate from the nitrogen dryer, instrument nitrogen dryer and the reactor building air con-ditioner can be drained to the liquid radwaste system has been installed. Previously the condensate had to be carried out of the reactor building in containers.
The system uses a 60 gallon batch tank into which the con-densate from each of the three sources can be drained. After isolating the batch tank from the three sources, pressure supplied from the plant nitrogen system is used to force the condensate to the liquid radwaste system.
The system includes locked closed valves on the drain line for each moisture collection tank and on the inlet, outlet and drain lines to the condensate batch tank to assure maintenance of containment integrity.
Contact Protection, Safety System Chassis Selenium contact protectors were added across the coil of each scram contactor in the safety system. Originally each safety system bus was supplied 125 V DC thru a set of series contacts of several mercury-wetted relays and thru the coil of two large DC scram contactor relays.
No contact protection
,ly leB 1al tn
- d.
- e.
was provided to the mercury-wetted relay contacts against the large induced current that occurred when the scram con-tactor relays were de-energized. The potential existed for damage to the mercury-wetted contacts by the induced current, including welding of the contacts. Durini bench test~ of a mockup cfrcuit arcing between mercury-wetted relay connector pins had been observed.
With selenium contact protectors added across the scr-un relay coils, the high current pulse is eliminated.
Thru-Head Rod Storage Well A thru-head rod storage well has been established in floor-well F-3 in the refueling cell. This provided temporary storage for the thru-head rods while removed from the reactor without requiring disassembly from the extension rods, Pro-visions were made to provide an argon atmosphere within the floorwell.
Reactor Vessel Vacuum Breaker Argon Supply Isolation Valve A manual valve (normally locked open) has been installed in the argon supply to the valve operator on the Reactor Vessel Vacuum Breaker valve.
The valve provides a means to isolate the argon supply to the vacuum breaker and provide assurance that the vacuum breaker will not open during those occasions when the Re-fueling Cell is on air. Previously, valve 8056-24 was used for this purpose, but closing 8056-24 prevented the operation of other equipment which was required for specific needs with the cell on air.
- f.
Installation of Additional 125 V DC Inverter An additional Inverter was inst~lled to improve the reliability of the 120 V AC power for nuclear instrumentation recorders (and other equipment) obtained from the 125 V DC battery system, By the addition of another larger Inverter (1000 vs 750 watts) delays in reactor operation from malfunction of existing unit can be eliminated. The output of only one inverter will be connected at any time by a manually operated switch. Each unit will function independent of the other.
- g.
h,
- 1.
- j.
Gas Sample Station for Refueling Cell and Nitrogen Zone An additional sample box equipped with connections to the gas sample panel and gaseous radwaste system has been installed for obtaining bomb samples of the refueling cell and the nitrogen-zone. The modification involves no new containment penetrations and no cross connections with the cover gas sample station.
480 V Bus 2A Overcurrent Lockout Relay Section C.8, pages 19, 20 and 21 of the Seventh Quarterly Operations Report discussed a condition which was found to exist on the Bus 2A Overcurrent Lockout Relay.
The B233 relay was converted to a standard B234 relay.
A wiring rearrangement was made, wiring labeled, and the relay name plate changed.
Sketches were provided to identify terminal board connections.
This modification restores the intended function with an alarm for overcurrent lockout added.
Refueling Cell Penetration N-10 Penetration N-10 was changed from a two connector co-axial penetration to a four connector penetration for compatibility with Oak Ridge National Laboratory instrumentation for reactor noise measurements.
Stud Tensioner Modification The reactor vessel headbolt (stud) tensioners were modified to eliminate or reduce the probability for occurrence of certain identifiable events, created either by the design of the tensioner or by act1ons of the operator.
The following changes were made:
- 1)
A pin was installed to prevent the socket from unscrewing from the puller bar.
- 2)
A short length of 1-1/ 4" schedule 40 pipe was installed around the puller bar and above the socket. The pipe serves two purposes, one to limit the upward travel of the puller bar when the tensioner is lifted by the hoist, and the other, to provide an energy absorption mechanism should the socket disengage from the stud, as occurred March 18,
- k.
- 3)
A short length of 1-1/4" schedule 40 pipe was installed over the puller bar between the holding nut and the winged nut on the top of the puller bar.
With the holding nut screwed up against this length of pipe, the length of the puller bar is fixed and consequently the distance the* puller bar socket screws onto the stud is fixed.
- 4)
One stud tensioner is adjusted for outer head stud use only, and the other for inner head only. Each stud tensioner is labelled for its correct usage.
With the change made in item 3, above, the procedure requires that the holding nut will be merely a specified fraction of an inch above the spring plate when the socket is properly threaded onto the stud, thus providing a visual verification of adequate engagement of threads.
- 5)
A metal guard was installed around the grapple fitting on the upper end of the puller bar to prevent operator injury should he fall across the top of the tensioner.
Programmed Flow Reduction Rate and Time Delay The time delay prior to flow reduction was changed to 10
(+/-1) seconds and the rate was changed to 57 (-0 + 10) gpm/sec.
To prevent too rapid cooling and yet provide adequate cooling following a scram from the various modes of reactor operation dictated by the experimental program, (including transient tests as well as steady-state), the time delay before programmed flow rate reduction of the primary coolant and the rate of reduction are adjusted to meet the needs of the particular test to be performed. The primary coolant flow rate is varied by a motor driven potentiometer that adjusts the exciter voltage on the motor-generator (MG~ sets. The one potentiometer controls the voltage for both MG sets. The speed of the motor determines the rate of change of exciter voltage or the primary flow rate.
The initial flow rate reduction (100 gpm/sec) and delay (2 seconds) were established for 20 MW steady state.
The delay time was subsequently changed to 6 seconds for the Familiarization Transient tests to prevent the flow rate interlock from inserting the latches prior to the FRED piston returning to original position after firing. Calculations of these values for the sub-prompt critical transient tests from 1, 5, 10 and 15 MW indicate that a delay of 10 seconds and a flow rate reduction of 57 gpm/sec provide an optimum match of heat removal with the heat fluxes generated.
1, IFA Transport Fixture and Grapple Modification Lanyards were added to the IFA transport fixture to facilitate release of the upper clamp and release of the grapple when the IFA is to be removed from the fixture or the grapple released following insertion of the IFA into the core.
m, Man Access Suit Exhaust System
- n.
The exhaust system vacuum was reduced to a value of 5 to 10 inches of Hg to limit the possible suit vacuum to an acceptable value in the event of multiple failures in the suit air supply system.
The vacuum system control valve, which bleeds air into the exhaust system, was changed from a 1/2" valve to a 1-1/2" valve and pipe section to provide adequate capacity for regulation at the reduced exhaust system vacuum, A
gauge was installed to provide vacuum indication.
Primary Drain Tank Vent Line Modifications A short length of 1/4 inch schedule 80 pipe was installed downstream of isolation valve 8036-49 in the Primary Drain Tank normal vent line. The new pipe section was intended to limit flow in the normal vent path such that throttling of 8036-49 would not be necessary.
(Previously throttling of 8036-49 had resulted in plugging at the valve seat due to sodium carryover during venting.) The 1/4 inch schedule 80 pipe did not sufficiently limit flow, so a section of 1/4 inch stainless tubing with a 0.035 inch wall was sub-stituted. This provided the desired flow with 8036-49 fully open.
I I
- c.
Other Reportable Items
- 1.
- 2.
Safety System Relay K5 Malfunction During the performance of License Test Procedure, (LTP) M-0-9, Monthly Test of Reactor Safety System on 2/21/71, a scram contactor dro~out failed to occur w'..en the scram chassis test switch was actuated for one of the three High Temperature Core Outlet {Upper Region) Thermocouple circuits. The scram logic is one-out-of-three. The other two circuits tested satisfactorily.
LTP M-0-9 was performed with satisfactory results last on 1/17/71.
The cause of the malfunction was identified as effectively shorted contacts on the mercury-wetted relay K5 (as if the con-tacts were stuck closed), These contacts are in the 12 V DC bus circuit. Other contacts on this relay functioned properly to provide an alarm, lighted light on scram chassis and a signal to the sequence recorder. The thermocouple reading was normal
(~500°).
The relay was replaced and performance tests were repeated to verify proper operation.
The most probable cause of the malfunction was a momentary overcurrent condition in the 125 V DC system, which occurred on 2/13/71, during checkout of system modifications to provide a scram delay for the transient test program.
The cause of the overcurrent condition was the inadvertent installation of capacitors with a low rated voltage in the time delay circuit for the scram contactors. Properly rated capacitors were subsequently installed and the system modifications were satisfactorily checked-out.
The possibility of contact damage due to induced currents when the scram bus is de-energized existed at the time of this malfunction, although no failures due to this cause have occurred at SEFOR.
This possibility has been essentially eliminated by the installation of selenium contact protectors on all 18 of the scram and containment isolation relays.
Satisfactory performance of the safety system was verified after the modifications were completed.
Freon System Time Delay Malfunctions On February 20, 1971 during performance of the semi-annual License Test Procedure, SA- 0-4, Nitrogen Cooling Refrigerant Isolation Valves, a failure to close freon header isolation valves and to trip off nitrogen blowers No. 1 and No. 2 occurred when the pressure at the freon header pressure switch was reduced below 60 psi. The scram trip function from each of these same pressure switches operated satisfactorily.
The surveillance test was performed with satisfactory results last on 8/20/70.
- 3.
The cause ot the malfunctions was identified a~ tlcW1aged circuit components (burned capacitors and resistors) in the time delay relays in parallel with the freon header pressure switch contacts.
The damage to the time delay relays created an effective short circuit around the contacts.
The time delay relays bypass the low freon header pressure trip for 2-1/2 minutes after system start up to permit the pressure in the freon header to be established.
The Site Safety Committee concluded that the failure of the relays was probably enhanced by the repeated current pulses associated with the many starts of the nitrogen blower - freon systems.
Measurements of voltage and currents revealed no abnormally high values.
The Committee recommended that the relays be replaced on an annual frequency.
Auxiliary ABC Door Malfunction On March 5, 1971 reactor power was increased to l MW in preparation for Familiarization Transient No. 4 (0.5$ from 1 MW).
When an attempt was made to open the Auxiliary ABC doors, the south door (bottom) would not open.
The reactor was shutdown.
By use of lubrication and external force, door operation was restored.
No internal interference with door operation was detected.
'fhe "bearings" in which the door support shafts rotate are merely holes drilled thru steel plates. Small holes were drilled into the "bearing" area, grease fittings installed and grease applied.
After satisfactory door operation was demonstrated thru several cycles, reactor operation was resumed.
No bending of the shaft was observed.
The Site Safety Committee concluded that:
a)
The probable cause of the door malfunction was the exposure of "bearing" and shaft to high temperature and moisture, b) for plant safety, the potential need for the Auxiliary ABC is after the reactor has operated at high power levels, and not on reactor startup, and c) in the event of a need for reducing the cooling by the Auxiliary ABC after plant power is reduced, sufficient time is available to manually close the doors if necessary, to maintain sodium temperature above a desired minimum level.
The Site Safety Committee recommended that a replacement door shaft bearing system be desibned and installed.
- 4.
Safety System Relay Malfunctions At 0030 on 4/10/71 performance of surveillance test LTP Q-0-7 was completed and reactor sodium level was returned to the operating level. At 0310 it was observed that the reactor vessel cover gas pressure was at 18 psig and the controller on Panel 5 was signaling for the reactor cover gas supply valve to open fully.
The valve lineup was verified to be correct, and the reactor sodium was at the norm~l level. To regain cover gas pressure control, a jumper was installed on the block cover gas supply valve contacts of the safety system chassis for low level reactor sodium uutil the malfunction could be corrected.
Investigation of the malfunction revealed that the block cover gas supply valve contacts on relay Kll in safety system chassis Al were open and would not close. In addition, the contacts on relay K3 in chassis C were closed and would not open.
The contacts on relay Kll of chassis Bl appeared to function properly.
The insulation on the interconnecting wiring between the contacts in the three chassis indicated that melting had occurred. These three contacts in series provide a one-out-of-three logic for trip on reactor vessel low sodium level. Also in series with these three contacts are contact on relays K4 and K5 of chassis Z which function on an Auxiliary Primary System leak and the coil of relay AVPL (Automatic Vent Primary Loop) whose contacts provide the current to the solenoid for the cover gas supply valve. All
.re mercury-wetted contact relays.
During the performance of LTP Q-0-7 the reactor vessel sodium level
- is lowered to permit verification of proper functioning of the reactor overflow check valve. In accordance with the procedure, a jumper is installed to supply 26.5 V DC thru the contacts in the Z chassis and the AVPL relay to maintain control of the cover gas supply valve (and consequently the cover gas pressure) during the time the contacts in chassis Al, Bl, and Care open because of the low sodium level.
Since control of the cover gas pressure was normal prior to the per-formance of Q-0-7, it was concluded that the cause of the daillage to the contacts in chassis Al and C and the malfunction was the inadvertent application of a higher voltage than 26.5 Vora shorting to ground which may have occurred during the installation or removal of the jumper.
LTP Q-0-7 is the quarterly surveillance functional tests of:
Primary Drain Tank Emergency Valve Reactor Overflow Valve Pump-Around-Pump Discharge Valve Reactor Overflow Check Valve (PCV 8059-44)
(FCV 393-21)
(FCV 393-4)
(216)
The five relays were replaced with new relays, the damaged wiring was replaced, each chassis was checked for any other damage ~nd surveillance tests were performed to verify proper function.
- 5.
The Site Safety Co11111.ittee concluded that:
a) a mis-attachment of jumper or inadvertent shorting to ground was the most probable cause of the malfunction, and b) the present system for identification of terminal locations and attachment of jumpers appears to be susceptible for the occurrence of this type of incident, As a result of Site Safety CollDllittee recommendations a new design is being prepared for attachment of jumpers, and in the interim all operations and maintenance personnel have been cautioned regarding the need for carefulness in identification of locations for jumpers and in their attachment.
Instrument Nitrogen Supply to Reactor Overflow Valve On 4/9/71 while performing the surveillance test LTP M-0-6, (Monthly Channel Test of Primary Drain Tank - Reactor Cover Gas Differential Pressure Monitor and Primary Drain Tank and Reactor Cover Gas Argon Pressure Control Valves Operational Test), the reactor overflow valve (FCV 393-21) failed to close when a trip was created by increasing the primary drain tank pressure greater than 10 psi above the reactor cover gas pressure.
LTP M-0-6 was performed last on 3/13/71 with satisfactory results.
Entry was made into the Primary Sodium Area to determine the cause of the malfunction.
The instrument nitrogen supply line (1/4" copper tubing) had broken at the swage-lock fitting on the valve operator. The valve fails open on loss of pneumatic power.
A new swage-lock fitting was installed and the tubing inspected for adequate length for flexibility since the valve operator moves.
The valve operated normally following the repair.
The Site Safety Committee concluded that:
a) the cause of the failure appeared to be fatigue, perhaps enhanced by a necking-down of the tubing during tightening of the fitting, b) the time of occurrence of the failure cannot be determined definitely, but occurred sometime after the last test on 3/13/71, and c) closing of the valve is a third order safety function in the event of a pipe break accident required if the overflow line check valve failed to seat and the Primary Drain Tank emergency vent failed to vent when Frimary Drain Tank pressure exceeded the cover gas pressure by more than 10 psi.
- 6.
The Site Safety Committee recommended that during the next major outage, (when the Primary Sodium Area atmosphere is on air} the tubing to each of the other valve operators be inspected for adequate length for flexing, for protection against vibration, and extent of possible work-hardening if vibration has occurred.
Violation of Limiting Condition for Operation - Technical Specification 3.12.B.5 Limiting condition for Operation 3.12.B.5 of the Technical Specifi-cation requires that, "whenever a poison slug worth more than $1 is lowered into the core by means of the FRED, containment integrity shall be maintained and the isolation valves on the outer containment ventilation lines shall be closed."
During the first calibration of the sub-prompt transient slug ($0.95 to $0.98} containment integrity was maintained but the outer containment ventilation lines were not closed.
The Transfer Chamber was not attached, the Accumulator pressure was zero, the Latches were engaged and the mode switch was in the "Normal" position. Evaluation of data subsequent to the calibration indicated the worth of the slug to be $1.01.
The Site Safety Committee reviewed the event and concluded that a slug worth in excess of $1 was not anticipated prior to calibration and that the procedure for the calibration was inadequate in that a pro-vision for closing the ventilation valves was not included which would have prevented this inadvertent violation of the Limiting Condition for Operation.
The procedures have been revised to include a requirement for closing the ventilation valves during the initial calibration of any slug.
- 7.
Reactor Vessel Head Bolt Stresses An investigation was conducted to determine the causes of apparent loss in pre-load of the reactor vessel outer head bolts. The results of this investigation wer2 discussed with members of the DRL staff on April 27, 1971, and were documented in a separate report sent to the DRL on May 28, 1971. Corrective actions taken to reduce the bolt stresses to acceptable values include the following:
a}
Reduction of the rate of change of sodium temperature from 20°F/hr to 10°F/hr for temperature changes greater than 125°F.
b}
Reduction of the pre-load for the outer head bolts from a stress level of 8500 psi to less than 4000 pai.
Changes have been proposed for the Technical Specifications to provide assurance that bolt stresses will remain within acceptable limits.
- 8.
Auxiliary Intermediate Heat Exchanger Performance The reduced heat transfer coefficient for this component was pre-viously reported on page 12 of the Sixth Quarterly Plant Operation Report.
An investigation of the design and manufacturing tolerances for this component was made, The results of this investigation led to the conclusion that the reduced coefficient is caused by shell side (primary loop) by-pass flow between the vertical side plates a1.1d track bars at the top of the shell.
(See Figure I-4. 2 in Supplement 17 to the FDSAR.)
The nominal spacing between these parts provides two passages for by-pass flow, 1/8" x 39" each.
Flow through these passages completely bypasses the tube bundle and mixes with the remainder of the. flow in the bottom of the IHX as it approaches the outlet nozzle.
The by-pass flow rate was estimated using an orifice coefficient for a slit taken from "Close-clearance Orifices" by Andrew Lenkei in Product Engineering, April 26, 1965. The flow through each of the parallel paths was determined by equating the two pressure drops (across the bundle and through the slits). This procedure predicts 100 GPM across the bundle and 150 GPM by-pass (60%).
A 60% by-pass is probably too high, since the gap may not necessarily be a constant 1/8" in width.
Any small variation would substantially change the calculated by-pass.
A second calculation was made, assuming that the by-pass flow is not cooled before mixing with the remainder of the flow at the outlet nozzle. This calculation indicates a by-pass flow of about 20%.
Therefore, it is reasonable to conclude that the reduction in effective heat transfer coefficient is due to a significant amount of by-pass flow in the am~iliary IHX.
The measured value for the overall heat transfer coefficient, U, of the auxiliary IHX was reported in the Sixth Quarterly Operation Report {page 12) and compared to the predicted value, as given below.
The design value from Table V-7 of the FDSAR {page 5-12) is also given here for comparison.
Measured:
Predicted:
Design:
u
= u
=
= u
=
525 BTU/Hr Ft2°F 1140 1060 This coefficient is related to the IHX parameters by the following equation:
Q = UA (MTD)
Q = Heat.transfer rate, BTU/Hr A= Heat transfer area, Ft2 MTD = Mean temperature d:ifference, °F Coolant temperatures and flcJw rates can be identified as shown below:
T4 Secondary Coolant Loop
(
w p
!i.
T2 Primary Coolant Loop
)
__ _)
W = Coolant flow rate, GPM T
- Coolant temperature, °F The design values for these parameters are given in Table V-7 of the FDSAR and are based on a heat load of 2.5 MWt (Q = 8,532,000 BTU/hr), since the IHX was 13ized for possible future expansion to a 50 MWt plant. The nominal auxiliary loop heat load for the 20 MWt plant is 1 MWt, since the nominal auxiliary flow rate is 5% of the main flow rate and the auxiliary reactor inlet and outlet coolant temperatures are nomin~lly equal to the main reactor inlet and outlet temperatures.
Using the above referenced value of U, typical values for the auxiliary loop at a heat load of l MWt would be:
Q, BTU/hr 6
3.413 X 106 3.413 X 10 U, BTU/hr ft2°F 1060 525 WP, GPM 247 247 W8, GPM 242 242 T1, OF 804 804 T2, op 700 700 T3, OF 649 597 T4, OF 753 701 As can be seen by comparison of the above,,alues, the only effect of the reduced heat t*ransfer coefficient is to reduce the values for the secondary sodium temperatures. This in turn causes some increase in the,air flow requirement for the auxiliary sodium/air heat exchanger, U:low~ver, since the l MW heat load is only 40% o! the 2. S MWt c,apacity of this heat exchanger, the installed air flow capacity,exceeds the required capacity for all operating conditions, and the system heat removal capability exceeds all normal and emergency requirements.
The estimated air flow required for the above l MWt condition is about 60% of the design value, This est~mate is based on test data which shows sodium temperatures and flow rates approximately equal to the above conditions with one of the two installed fans operating at 1/7. speed, the other fan operating at full speed, and with the air flow control louvres partially closed.
Operation of the system with unequal primary and secondary coolant flow rates would not have a significant effect on the above conclusions, since the loop average temperature tends to remain constant when the coolant flow rate is changed at a given heat load. For example, reducing the secondary coolant flow rate to 121 GPM would change the above estimated values of T3 and T4 to S42°F and 701°F, respectively, for U = 525.
Under emergency conditions r*equiring use of the auxiliary system to cool the reactor, the reactor would be shut down, Assuming that the main coolant system cannot be used, the maximum steady state heat load applied to t.he auxiliary system can be determined from Figure 2-4 of Suppleme111t 19 to the FDSAR, which snows the total core decay heat after 100 days operation. This curve yields a value of P/P0 = 4% at 1 mi.nute after scram, decreasing thereafter.
Thus, the maximum core decay heat load (baaed on 20 MWt) would be 800 KWt.
This heat load car.t be removed by the auxiliary system with no difficulty, as indic:ated above.
If conditions were such that system temperature red\\l1ction was required, the heat load removed by the auxiliary sys1tem could be increased as necessary by increasing the system air fl.ow up to the design value.
For emergency conditions im.rolving loss of plant power, the heat removal capability of the emergency system is maintained by the application of power from the emergency diesel generator. In the event that other failures occur, the reactor can be cooled by natural circulation as discussed in the report entjltled,.Natural Circulation Testing of the SEFOR Reactor Coolant Systen1.
11 (Previously sent to the DRL,)
D.
Safety Review and Audit Activities
- 1.
Twenty-one meetings of the Site Safety Committee were held during this quarter.
- 2.
One trip was made to the site to review plant safety.
- 3.
The evaluation of the reactor vessel head bolt stresses and the corrective measures taken to resolve the problem were reviewed and presented to the DRL.
Proposed Change No. 5 to the Technical Specifications was prepared and submitted to the
- DRL, This change provides the corrective actions end surveillance required to assure that the bolt stresses remain within acceptable values.
(See paragraph C.8 of this report.)
- --....... ----==--=--=---
~ - --*---
Date 2/4/71 2/25/71 2/25/71 2/25/71 2/26/71 2/28/71 3/4/71 3/4/71 3/5/71 3/5/71 3/5/71 3/6/71 3/6/71 3/12/71 3/13/71 3/13/71 4/4/71 Cause TABLE I REACTOR SCRAMS Loss of Site Power (during local thunderstorm)
FRED Test (PTP 46)
FRED TEST (PTP 46)
Manual Scram, spurious leak detector alarm Manual Scram, spurious leak detector alarm FRED Test (TP IV-1) 40¢@ 1 MW High Temperature Main Secondary Cold Leg FRED Test (TP IV-1) 40¢@ 1 MW FRED Test (TP IV-1) 50¢ @ 1 MW Personnel Error WRM 01 Failure to Switch Ranges Malfunction Containment Ventilation Radiation Monitors, spurious signal FRED Test (TP IV-1) 50¢@ 1 MW FRED Test (TP IV-1) 40¢ @ 10 MW FRED Test (TP IV-1) 40¢ @ 10 MW FRED Test (TP IV-1) 50¢@ 10 MW FRED Test (TP IV-1) 50¢@ 10 MW Manual Scram (TOP 117)
No, 107 108 109 110 111 112 113*
114 115 116 117*
118 119 120 121 122 123
- 2 out of 3 trip logic, in which one trip signal combined with the existing trip for one freon unit shut down during low power operation to cause reactor scram. 1*
BRDO CP EM EP FCV FRED IFA IFST IHX IRM LTP MPS MSS NFSV PAP PCV DEFINITIONS Air Blast Cooler Auxiliary Primary System Area Radiation Monitor Auxiliary Secondary System Auxiliary Breeder Reactor Development Operation Corrective Procedure Electro-Magnetic Emergency Procedure Flow Control Valve Fast Reactivity Excursion Device Instrumented Fuel Assembly Irradiated Fuel Storage Tauk Intermediate Heat Exchanger Intermediate Range Monitor License Test Procedure Main Primary System Main Secondary System New Fuel Storage Vault Pump-Around-Pump Pressure Control Valve Definitions: (continued)
PM PTP PVT Rx SRM TOP TP WRM Preventive Maintenance Provisional Test Procedure Primary Vent Tank Reactor Source Range Monitor Temporary Operating Procedure Test Procedure Wide Range Monitor