ML20030A494
| ML20030A494 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/27/1968 |
| From: | Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090757 | |
| Download: ML20030A494 (13) | |
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Docket No 50-155
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License No DPR-6 May 1, 1968 Through October 31, 1968 This report, suositted in accordance with Paragraph 3.D.(3) of Operating License No DPR-6 (effective May 1, 1964), covers the ninth six-month operating period for the Big Rock Point Nuclear Plant (Dlant).
I.
SUMMARY
OF OPERATIONS A.
POWER OPERATION The plant was on the line for the entire month of May with
. load limited to between 60 Mwe and 6h Mwe (gross). This derate was necessary due to the significant amount of stainless steel that had to be reinserted in the reactor following the significant "C" fuel failures experienced during the previous run. With plant load at 6h Mwe (gross) on May 2h, corresponding to 200 Mvt, the maximum steady-state heat flux was h70,000 Btu /ft -hr on the "centermelt" assemblies, with an MCHFR of 1.54.
On June 3, the plant was removed from service to repair two steam leaks located in unions adjacent to the explosive valves on the reactor poison system. Plant operation resumed on June k.
Off-gas activity increased from approximately 3,400 pCi/see toL14,000 pCi/see on June 9, thus indicating the failure of fuel cladding in the core, f
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.s Plant load was reduced June 12 and 13 to 57 Mwe (gross) and 52 Mve i
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- s (gross), respectively, so as to not exceed MCHFR at 122% full power
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.and to reduce off-gas activity.,0ff-gas activity dropped to 5,300 pCi/see by June 16. Thermal power was increased to 186 Mvt on June 21, C
corresponding to 60 Mwe (gross) to satisfy the requirements specified in the centermelt fuel program. The increase in core power was to es-tablish equilibrium conditions in the UO2 pri r to shutdown so that hot.
y cell examination of the rods would be meaningful.
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Off-gas activity increased to 11,000 pCi/sec during the thermal increase. The plant was removed from service at 2237 hours0.0259 days <br />0.621 hours <br />0.0037 weeks <br />8.511785e-4 months <br /> on June 21, for the fifth refueling outage.
The plant resumed operation July 16, following the 25-day outage. On July 21, a 30 Mwe load rejection test was successfully per-formed. See Section V-B for test details. The plant load was limited to 60 Mwe (gross) while obtaining core data to increase reactor power.
'On July 23, load was raised to 72 Mwe (gross) and remained at this level until a readjustment was made to allow the turbine control valves to fully open..On July 29, the plant lead was-increased to Th Mwe (gross).
The unit was removed from service to repack the No 2 reactor recirculation pump butterfly valve on September 10 after completing a day continuous operating period. A short scheduled outage occurred on September 21 to change the off-gas filter in the stack due to a high AP and to conduct _ operator training on the control console.
A. scheduled outage occurred October 13 to repair two steam leaks and to replace the high-pressure heater drain valve diaphragm.
-While returning the plant to power, control rod B-5 could not be moved from-Notch 15 using normal operating pressure.- The reactor was shut' down, and during cooldown, the drive was exercised.using increased hy-l>
draulic. pressure until it worked freely. The drive was then removed from the reactor.and rep 1' aced with a spare one. The drive was disas-I sembled and inspected, but no obvious reason for the malfunction was
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evident...
I Two short-duration outages were scheduled late in October to inspect:and replace 0-rings in three control rod drive flanges.
Off--
- gas activity increased from_3,700 pci/see to-12,500 pCi/sec during the I
-last:veek of October, indicating cladding failure in the new core.: Plant I
~ outputivas steady at 72 Mwe (gross) at the end of the month with off-gas activity at approximately 13,000 pci/sec.
i iB. REFUELING OUTAGE
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Items of interest occurring during the refueling outage are:
- l ~. : Fuel Inspection-
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JFifty-seven fuel _ bundles were transferred.from the core to the-
. spent' fuel pool.. Dry fuel sipping begani on June 26 and.vas completed on p-July 1..
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'No of the 2k standard "A" bundles in the core indicated a t -
positive leak signal after being dry sipped. The "A" fuel rods have 19 mil vall, stainless steel cladding, containing UO fuel pellets. The 2
two leaking bundles had exposures of 9,000 Mvd/T and 8,800 Mvd/T, zeepec-tively. No leak suspects were noted in the remaining 25 "A" bundles.
Three standard "B" (Reload 1) bundles (out of 30 in the core) vere dry sipped with no leak signals noted. The "B" fuel rods have 3b and 31 mil valls, Zirealoy-2 cladding, containing UO fuel pellets.
2 These bundles had exposures ranging between 8,000 Mwd /T and 12,300 Mvd/T.
Five of the 12 "C" (Reload 2) bundles in the core were dry sipped with one leak and one suspect signal noted. The "C" fuel rods have 34 and 31 mil valls, Zircaloy-2 cladding, containing UO P "d*#-
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' The leaking bundle had an exposure of 8,h00 Mwd /T and the suspect 7,700 Mad /T. Only five "C" bundles were sipped as the other "C" bundles were detected as leakers during the fourth refueling outage.
The leaking "C" fuel found above provided no new information
- regarding the failure mode.
.All six centermelt fuel bundles and nine of the 16 develop-mental bundles in the core were dry sipped. Results appear in the R&D section. (I-D) of this ' report.
The new core consisted of the following bundles:
(a) ' 3 developmental bundles (two R&D Phase II "D" and one
'centermelt "D" intermediate performance containing pellet fuel).
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-(b) 30 "B"'(Reload 1) bundles.
i (c) 10 "C" (Reload-2) bundles.
(d)) hl "E" (Reload 3) bundles.
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-- Forty-one bundles of "E" fuel vere loaded in the core during refueling. These bundles have a 9 x 9 fuel rod array, ho mil vall, Zircaloy-2 cladding, containing UO fuel. pellets enriched as follove:
2 23T 2.35% U
- Lov 235
-Middle -. 2.'93% U 23F 3.55% U High' Note the "E" fuel design was' changed from powder UO to pellet 2
UO after our
'.'C." fuel bundles prematurely failed as noted in our eighth 2
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. semiannual report. dated June 24,;1968.
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Seven powder UO "C" fuel bundles that were returned to General 2
Electric Company for reworking and outgassing are included in the core loading above.
2.
Core Internal Inspection No loose components (nuts, bolts, etc) were found during our surveillance inspection of core internals for the fifth refueling outage with the exception of a grid bar beam clamp.
'A replacement of this clamp is anticipated in the next refueling outage.
During the fourth refueling outage, one wedge was found lying in the vessel. The wedge (one of six) was used to disable the lower seal ring of the thermal shield during the thermal shield fix in 196h-1965.
Recently, the General Electric Company stated (per telegra= dated August 16, 1968) that the Big Rock Plant can operate safely even with all core vedges missing from the seal ring.
One segment of the thermal shield seal at 180 orientation was removed and reinstalled to:
(a) Remove a reactor vessel vall surveillance coupon located behind the thermal shield, and (b) Inspect the vessel cladding where the seal segment bar contacts the essel vall. No evidence of year was noted. Measurements for the seal segnent fitting tolerances showed no significant chanEe from previous measurements.
C.
STATISTICS The reactor vas brought critical 11 times. The reactor was critical for 3,698 hours0.00808 days <br />0.194 hours <br />0.00115 weeks <br />2.65589e-4 months <br />. The plant was on the line.for 3,632 hours0.00731 days <br />0.176 hours <br />0.00104 weeks <br />2.40476e-4 months <br /> with electric generation of 245,361 Mv hours (gross) or-232,963 Mv hours n
(net). The thermal output of the reactor was 77k,606 Mv hours.
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D.-
R&D PROGRAM Six centermelt fuel bundles were inserted in the core during the fourth refueling outage.to gain experience with centermelting of different fuel types.- (See Amendment No 1 to Facility Operating License No DFB-6 dated March 12, 1968.) ' Two of the six centermelt fuel bundles indicated a positive leik signal.after being dry sipped. The failures-c vere in the-advanced performance, powder UO fuel bundles with h0 mil 2
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Zircaloy-2 cladding. All the inter =ediate and advanced performance pellet fuel in the center =elt program showed no leak signals during sipping. The one intermediate performance powder UO bundle with 35 mil Zirealoy-2 clad-2 ding was a suspected leaker after the first sip; however, after a resip, no signal was indicated.
Three develop = ental bundles of the nine bundles (R&D Phase II) dry sipped were suspected leakers. Exposures of the three suspects are as follows:
1.
D-1, 30 mil vall, Zircaloy-2 clad, pellet fuel at 16,800 Mwd /T.
2.
D-5, 19 mil vall, Incoloy 800 clad, pellet fuel at 12,800 Mwd /T.
3 D-7, 19 =il vall, Inconel 600 clad, pellet fuel at 13,000 Mad /T.
II.
ROUTINE RELEASES, DISCHARGES AND SHIPMD.~1' 0F RADIOACTIVE MATERIAL g, ggy Q :, ' d A.
A total of approxi=ately '_El - 10' curies of activation and fis-sion gases was released to the environs via the stack. This figure is based on 3,209 EFFH of operation at an average release rate of h,000 uCi/sec.
Off-gas activity increased to a maximum of 12,000 pCi/see in the last week of October. The off-gas mixture changed from a recoil mixture to a diffusion mixture, thus indicating a failure of fuel cladding.
3.
LIQUID DISCHARGE During this reporting period, the liquid radioactivity, released to Lake Michigan by way of the circulating water discharge canal, numbered 30 batches, with a total activity of 5.72 curies. - Two of the batches were released on a partially identified basis where at least 90% of all the activity was determined to be a combination of Zn65, Co58 140 1h0 and Ba
-La All other batches were released under unidentified limits.
C.
SHIPMENTS A total of fourteen off-site shipments of radioactive caterials was made as follows:
Ship-ment Transfer-No Date From Transfer to Radioactive Material 60 1
5/-9/68 DPR-6 NPI 19-12667-hk Rods of Co h06,000 01 Curies
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2 5/13/68 DPR-6 ATCOR, Inc Low-an1 Intermediate-Pressure 31-116LO-2
' Heater Tube Bundles - 1h =C1
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i Ship-ment Transfer L
No Date From Transfer to
- Radioactive Material 3 Solid 3
5/15/68 DPR-6 ATCOR, Inc Approx 110 Ft 31-ll6h0-2 Radwaste - 2.55 Curies k
5/27/68 DPR-6 Univ of Mich 4 Liters of Reactor Water 21-215-h 2 mci 5
6/25/68 DPR-6 GE-Vallecitos 1200 ml Reactor Water, 3 Doz 00lT 60(Cal)
Off-Gas Vials, Reactor Water Crud Samples - 0.1 uCi 6
7/ 6/68 DPR-6 GE-Vallecitos Profilometer Equipment and 0017 60(Cal)
Hand Tools - 2.0 mci i
7 7/26/68 DPR-6 NPI 19-12667-h8 Rods of Co60 h00,000 01 Curies 8
-7/29/68
.DPR-6 US. Naval Lab Reactor Vessel Surveillance 8-1393-2(A 66) Specimen - 5.33 Curies 9
7/29/68 DPR US Naval Lab Reactor Vessel Surveillance 8-1393-2(A-66) Specimen - 13.h6 Curies 10
.8/ 7/68-
~ DPR16 GE-Vallecitos Profilometer Equipment and 0017-60(Cal)
Hand Tools - 2.0 mci 1
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.8/15/68 DPR-6 GE-Vallecitos Crud Samples - Heater Drainc
'0017-60(Cal).
1.6 mci 12
-9/24/68
.NPI 19-12667-2h Rods of Co60 - 325,000
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-Curies 13 9/24/68-DPR GE-Vallecitos'. Crud Samples - Heater Drains
.0017-60(Cal)
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10/29/68'
- DPR LGE-Vallecitos Crud Samples - Heater Drains 0017-60(Cal)~
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7 III. RADIOACTIVITY LEVELS IN PRINCIPAL FLUID SYSTD4 (FOR SIX MONTRE)
A.
PRIMARY C00IANT Minimum Average Maximum Reactor Water Filtrate "
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pCi/cc 6.5 x 10 7.4 x 10 1.2
- Reactor Water Crud
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pCi/cc Turbidity 1.0 x 10 h.h x 10 2.33 Iodine Activity
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2.2 x 10 6.0 x 10~
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pCi/cc 1.0 x 10 B.
REACTOR COOLlHG WATER SYSTD4 The principal radionuclides in the reactor cooling water syste=
vere E and Cr which resulted from the activation of the potassium chromate inhibitor.
Minimum Average Maximu-Reactor Cooling Water (a)
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-1 pCi /cc 1.3 x 10 1,5 x 1g 6.0 x 10 C.
SPENT FUEL POOL Radioactivity in the spent fuel pool is principally activated corrosion products that have gone into solution from stored fuel and core components.
Minimum \\
Average Maximum Fuel Storage Pool (a) pCi/cc 1.5 x 10~
1.2 x 10-1 1.0
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Fuel Pool Iodine (b'
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-2 pCi/cc 3.0 x 10 8.0 x 10-5 3.5 x 10 i
IV.. PRINCIPAL MAINTENANCE PERFORMED A. - A spare reactor feed pu=p barrel was inspected and chemically cleaned to eliminate crud deposits from previous operation. All' pump com-ponent parts were checked and no replacements were required. Clearances were set'as recommended by the manufacturer. This barrel assembly was installed in the No 2 reactor feed pu=p during refueling outage in July.
- Based on APHA turbidity units and 500 ml of filtered sample..
- A counter efficiency based on a gamma energy.of 0.662 Mev and one gn=ma
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photon per disintegration.. Decay scheme is assumed to' convert count rate to microcuries. LAll count rates were taken at two hours after sarpling.
131 Based on efficiency of Iodine two hours.after sampling.
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During the cutage cn October 13, an elbow in the scisture separatcr E.
loop seal was replaced and a flange was replaced in the steen seal drain line. The air-operated diaphrag in the dump valve frcs the high-presstre heater was also replaced.
C.
Centrol rod drive B-5 was re=cved frc the core for inspection after it calfunctioned in service. I;ev seals and guide bushings were in-stalled en the actuating pisten and two sets of internal pisten seal rings 1
and 0-rings were replaced. The block valve screens in the hydraulic syntes vere changed.
In the centh of October, two outages vere caused by water leaking D.
Both flange surfaces of each drive frc three centrol rod drive flanges.
!;e.* 0-rings vere inspected and honed though no irregularities were noted.
The leak-were installed. The 0-ring spacers were inspected and cleaned.
age water vcs chemically analyzed and confirmed that the leakag was red drive cooling vater rather than reactor water.
CHA';GES, TESTS A';D EXPO ;E';TS PERF0PRED PURSUA';T TO 10 CF? 50.5M a)
V.
This section describes the changes made to the facility within the six-senth period without prior Cc=ission approval pursuant to Secticn 50.59(a) of Title 10, Code of Federal Regulations, to the extent that such changes constitute changes in the facility, as described in the Final Eazards Su==ary Eeport (FESR). It also included tests and experiments carried out at the Plant withcut prior Cc=ission apprcval pursuant to Section 50.59(a).
Each chan6e, test er experiment is described as authcrized only after a finding by Consumers Power Company that it did not involve a change in the Technical Specificaticns inecrporated in Operating License DPR-6 (effective May 1, 196h) or.an unreviewed safety questien.
A.
FACILITY CEA?IGES After installing the d-c operated main steam bypass isolation 1.
valve in the fourth refueling outage, the main steam bypass valve CV h106 the manual to automatic mode of cperation while operating was changed frc:
The d-c valve vill give positive shutoff actuation for the at full power.
turbine bypass line if the main bypass valve fails open in the autcratic
= ode at full power or remains cpen after a manual signal for closing is given.
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2.
The time delays (tripping relays) in the electrical schtte for the condenser circulating vater purps, which close on less of voltage, were changed from two seconds to four seconds. This is to allow both pu=ps to ride thrcugh the voltage sving after station power is restored following a =acentary station pcver trip.
3.
The control switches and circaitry for both circulating water vacuu= pumps were modified to provide for automatic return of one punp to power upon return of staticn power. One pump is necessary to insure air re=cval cn the water box side of the tain steam condenser.
h.
Two frequency recorders vere installed in the control ro =
with the ranges being 50-70 Hz and 58-61 Hz.
Installation cf the frequency recorder on the front control panel required relocation of the in-core flux a=plifier recorder to the back control panel. The flux reccrder is ncv located in the panel adjacent to the flux amplifiers.
5 The location of the off-gas flow transmitter was coved cut of a high radiation area into the condensate pump room to facilitate
=aintenance during power operation. Erratic operation of the transmitter prior to the outage was traced to dirty restricting crifices in the charter diaphrag= capsule. The orifices have teen rencved, which results in a vider recorder trace frc= increased sensitivity to flav variaticns.
In addition, a direct reading AP gauge was installed in parallel with the flev trans=itter to verify the recorder readings.
6.
New detent assemblies were installed on the picoa==eter range switches. The detent assemblies are an addition to the range switch and vill provide = ore positive switch indexing.
T.
The high-pre.sure heater drain valve is incapable of handling the flov at full lead operation since the retubing of the heaters early this year. The portien of the flow that is not handled by the valve is passed directly to the condenser via the heater du=p valve. This results in improper heater operation and erratic off-gas f1cv neasurements te-cause of the unstable situation in the hot well. An auxiliary level svitch was added to the du=p control valve to open at a 10-15% increase in heater level, thus preventing the flooding of the heater. Furtharecre, I
a 1" bypass valve and piping vere installed around the heater drain valve to control flow and reduce cycling in the system.
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B.
TESTS 1.
Checks and Calibrations Reactor protection sensor checks and calibrations were completed prior to No 5 refueling and return to power operation. The check-out of neutron monitoring instrumentation was delayed due to a decrease in the insulation values of the high-voltage and signal cables. This was attrib-uted to high humidity conditions on the connectors at the external sphere penetrations. The connectors were cleaned and a heat lamp applied to drive out excessive moisture. Subsequent problems with high humidity conditions in the same area caused considerable difficulty during the month of August.
A dehumidifier was installed in the cable penetration area to reduce the humidity and an increase in airflow to the ion chamber guide tubes cor-rected the problem.
2.
Containment Leak Rate Test The containment leak rate test was conducted during the fifth refueling outage. Leakage from containment was within allovable ltnits.
See Special Report SR-10 on Containment Leak Rate Test 1968, Docket No 50-155, License DPR-6 dated December 4, 1968.
3.
Load Rejection Test The 30 Mwe (gross) load rejection test was conducted on July 21.
All controls performed as expected to "save" the reactor, but analysis of pressure, valve position and flow data indicates shortcomings which vill not permit. successful tests at loads above approximately 50 Mwe (gross).
.The following data vere noted from recorder information. The neutron flux (Uc 2 pico) showed a transient flux increase from h9% to 69%. The-pressure at the bypass valve showed a slow transient increase from 1,250 to 1,290 psig which followed the short duration spike of 1,305 psig. The short spike occurred during the valve da=pening cycle during the first one half. second following the anticipatory ' opening of 60%. The valve
. position then stabilized near 40% open after one half minute..
h.
Temperature Coefficient Test This-test was conducted in July prior to power operation with
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the newly loaded core. Test data indicated that the coefficient turned negative at 150 F after. adding 'lk.6 cents' of reactivity.
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Fuel Sipping Test Many fuel bundles were tested for clad material integrity, uti-lizing the " dry sipping" technique. This procedure was discussed in pre-vious semiannual reports.
6.
Whole Body Counting Sixty-six personnel were whole body counted by Helgesen Nuclear Service, Inc on September 23 acd 2h.
All radionuclides found were lov.
The maximu= value of Zn found was less than 0.3% of the maximum permis-sible body burden. The maximum amount of Co and Co observed in a.av personnel was 0.8% of the maximum parmissible body burden.
VI.
PERIODIC TESTING PERFORMED AS REQUIRED BY THE TECIUiICAL SPECIFICATIONS The following tabulation shows the required frequency of testing, plus the testing date of the syste=s or functions, which may be periodically tested per Technical Spe 'Cications:
System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives
. Continuous withdrawal and insertion Each major refueling and 7/ 9/68 of each drive over its stroke with at least once every six 11/17/68 normal hydraulic system pressure.
months during periods of Minimum withdrawal time shall be 23 pover operation.
seconds.
Withdrawal of each drive, stopping Each major refueling and 7/ 6/66
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L at each locking position to check at least once every six 11/17/68 7
' latching and unlatching operations months
-Miug. periods of and the functioning of the position power o - r. tion.
indication system.
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Scram of each drive fran the fully Each major refueling and 7/ 6/68 l
vithdrawn position. Maximum scram at least once every six 11/17/68 time from system trip to 90% of-
' months during periods of insertion shall not. exceed 2.5 sec.
power operation.
Insertion of each drive' over its Each major refueling but 7/ 7/68 entire stroke with reduced hydrau-
.not less frequently than-
-lic system pressure to. determine once a year, that drive' friction is normal._
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Control Rod Interlocks 1
Rod withdrawal blocked when any two Each major refueling but 7/ 6/66 accumulators are at a pressure not less frequently than belov 700 psig.'
once every 12 months.
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System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Interlocks (Contd)
Rod withdrawal blocked when two of Each major refueling but 7/ 6/66 three power range channels read be-not less frequently than low 5% on 0% - 125% scales (or belov once every 12 months.
2% on their 0%
40% scales) when reactor power is above the minimum operating range of these channels.
Rod withdrawal blocked when scram Each major refueling but 7/ 6/68 dump tank is bypassed.
not less frequently than once every 12 months.
Rod withdrawal blocked when mode Each ma,ior refueling but 7/ 6/68 selector switch is in shutdown not less frequently than position.
once every 12 months.
Other Liquid poison system component Two months or less.
7/ 5/68 check.
9/ 6/68 ll/1L/68 Post-incident spray system At each major refueling 7/ 6/68 automatic control operation.
shutdown but no'. less frequently than once a year.
Core spray system trip circuit.
Not less frequently than 7/15/68 once every 12 months.
E=ergency condenser trip circuits.
Not less frequently than 7/15/68 once every 12 months.
Containment Containment sphere access air locks Once every six months or 9/ 6/68 and vent valves, leakage rate.
less.
Isolation valve operability and At least once every 12 7/ 2/73 leak tests, months.
Isolation valve controls and Approximately quarterly.
6/ L/68 instrumentation tests.
7/15/68 8/30/68 9/21/68 Penetration inspection.
At least once every 12 7/ 6/:18 months.
g Integrated leak test.
Once every two years.
7/1?/68
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13 The following instrument checks and calibrations were performed at least once a month:
1.
Reactor safety system checks not requiring plant shutdown.
2.
Air ejector of gas monitor.
3.
Stack-gas monitor calibration.
k.
Emergency condenser vent monitor.
5 Process monitor.
6.
Area monitoring system.
VII. PERSONNEL TRAINING Two AEC Operator Licenses and four AEC Senior Licenses were received at the Plant this period.
By s '.
i Gerald J. Walke Supervisory Engineer - Nuclear Consumers Power Co=pany
-Jackson, Michigan i
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'Date:
I Sworn and_ subscribed to before me this-day of i n s oea i A Ar,,
Notary Public, Jackson County, Michigan My commission expires January 15. 1972 i'
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