ML20030A496
| ML20030A496 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/24/1968 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090763 | |
| Download: ML20030A496 (26) | |
Text
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CONSUMERS PC'n'ER COMPANY Docket No 50-155 Regulatory Supp! Fife Cy.7 Report of Operation of Big Rock Point Nuclear Plant License No DPR-6 November 1,1967 Through April 30, 1968 3
This report, submitted in accordance with Paragraph 3.D.(3)
.of Operating License No'DPR-6 (effective May 1, 1964), covers the eighth six-month operating period for the Big Rock Point Nuclear Plant (Plant).
I.
SU'C%RY OF OPERATIONS A.
Power Operation The Plant was on the line at 73 Mwe (gross) at the beginning of this period and was shut down on November 25, 1967 for T hours to replace the-off-gas filter. The filter was replaced due to a high dif-ferential pressure.
Plant load was. reduced to 59 Mwe (gross) in early December after off-gas activity had increased from an average value of 13,000 pC1/see to a maximum of 21,200 pCi/sec. This load reduction was made to preserve fuel integrity and to increase core life since the scheduled refueling was ' set back a month. ' The Plant was removed from service on December 9 for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> to perfom control rod tests (as required by the Technical Specifications). On two occasions in December, the Plant load' was reduced to 10 Mwe -(gross), :each for a short period of time to make temporary repairs and to stop steam leaks on th'e turbine bleeder trip valve to' the high-pressure heater.
The Plant was in service.for the entire month of January, carrying a load of between 59 to 65 Mwe (gross). One short-outage of
~
15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> wasL scheduled to allow for AEC Operator, License examinations.
Plant load was reduced on three separate occasions to repack both No-1 and No.2 reactor feed pumps.
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2 Starting in January, the Plant was operated with an "all-rods-out" core configuration with the exception of Drive No F-5, which was inoperative in Position (notch) 11.
(See reference letter to Dr. Peter A. Morris dated January 8, 1968.)
Load sagt;ed to approximately 56 Mwe (gross) due to the "all-rods-out" operating condition until the fourth refueling outage on February 11.
Before the outage in February, seal pressures for the No 2 reactor recirculating pump began a slow and steady decline. The inner seal filter and pressure reducer were both bypassed. Pressure to the seal was controlled by a pressure reducer in the bypass line. This pump was disassembled during the outage to install a new cartridge seal assembly and stub shaft.
The Plant resumed operation on March 15, following a five-week refueling. Due to problems with the lower bearing in the No 2 recire pwop during test operation, it was necessary to resume Plant operation wit? one pump while modifying repair parts.
(See reference letter to Dr. Peter A. Morris dated April 23,1968.)
The Plant was shut down on April 4 to reinstall the No 2 recire pump. After-completing this work on April 6, the reactor was made critical and Plant heat-up started.
Inadequate seal leak-off flow from the newly installed seal was noted. This flow did not improve even when full operating pressure was reached. A decision was made to shut down the reactor and install a spare shaft seal cartridge in the pump.
During the approach to critical, following seal replacement, Control Rod Drive B-4 could not be withdrawn from the fully inserted position. After various systematic checks were made, the Plant was shut down.
(See reference letter to Dr. Peter A. Morris dated April 23, 1968 explaining this drive malfunction.) A newly rebuilt drive was in-stalled in the B-4 slot.
Plant operation was resumed on April 9, with a load of 60 Mwe
( gross). The reduction (derate) in load was due to the many old fuel bundles used to reconstitute a full 84-bundle core after the "C" fuel
(
failed prematurely.
(See Refueling Outage Section for details.)
f 1
3 On April 23, while the Plant was carrying a 60 Mwe (gross) load, the reactor was scra=med, inadvertently, when the high-sphere-pressure sensors were accidently bumped. Plant operation was resumed after approximately 6 hourc and continued operating without incident for the remainder of the month.
B.
Refueling Outage Items of interest during the refueling outage are:
1.
F-5 Control Rod Drive Inspection (See reference letter to DRL dated January 8,1968.)
2.
Fuel Inspection
-Seventy-eight of the 84 fuel bundles in the core were trans-ferred to the spent fuel pool for one of the following reasons:
a.
Visual Inspection b.
Dry C T *q c.
Fuel Rod Removal for Profiling Twenty-nine out of the 33 Reload-2 "C" fuel bundles in the core indicated a positive leaking signal after being dry sipped. Tne
C" fuel rods are 0.449" OD, 0.034" vall, Zr-2 clad, UO vibratorily 2
compacted powder to about 85% theoretical density. The prerefueling core contained 33 "C" fuel bundles, or about 40% of the total core.. The averagebundleexposurefor.this"C"fuelwasabout7500 Mwd /T.
Inspection of the "C" fuel at the site (by periscope and TV) s showed that the cladding had been breached, during service, in local l
l blisters at isolated spots. Examination of this type blister on other
[
rods indicated that li_tpurities, either mixed with or adsorbed on the powder particles, ha$ reacted chemically with the clad at isolated spots l
to form local blisters of massive zirconium hydride. It is expected that this source of compacted powder fuel rod failure will be eliminated by improved processing methods, thus. removing deleterious impurities from the fuel rod interior. Zircaloy-clad vibratorily compacted powder fuelrodshaveoperatedsatisfactorilyinthePRTRandto20,000 Mud /T
-(with centermelting) at GETR for the AEC High Perfomance UO R&D Pro-2 l.
gram. A review of the processes used in making the PRTR and GETR rods (3
indicates that' impurities of the character which caused the. "C" fuel -
b
E rod failures were nct present. The three ;cvder fuel bundles, fcr the AEC High Perfc =ance UO Center elt Progran at the Plant, vere reworked 2
using this i= proved process 'cefore they were inserted inte the reactcr at cur last refueling outage.
Seven unirradiated "C" bundles vill be sitilarly reverked with the inproved process and leaded into the core at the next refueling.
Also, sc=e slight possibility is suspected that there vas =echanical da-age to lever end plug velds during vibraticn fcr pcvder cenpaction i
l vhich subsequently eculd result in leaks at these velds in service.
Vater, which would have entered the rods thrcugh end plug veld leaks, right also attack the clad inside surface. Precautiena:/ end plug veld re; airs vill also be nade in the seven "C" bundles being reverhed for
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1ater use. Methods of i=preved rod end suppcrt during vibration fer pcvder ec=; action which have been developed vin eli=icate this pcssi-
. bility in the future.
t All ED bundles (TE-C!, and D) were d:/ sipped. One lev-level leak siEnal was ncted in hdle D-2, with an ex;csure of 16,h73 Wd/T.
D-2 is a Phase II, _ RO, 30-nil, Zr-2 clad, penet-type fuel bundle.
Ten "A" (standard) and 12 "3" (Relcad-1) fuel bundles were dry i
- sipped with no positive leak signal noted.-
The new core leading consisted of the foncving:
a.
2h Standard "A" Brdles
- b.. 3 Instr =ented "A" Bmdles-c.
15 Develogental P.:niles (9 ED Phase IID and 6 Center-4
=elt D) d.
30 "B" (Relcad-1) emiles e.
12 "C"-(Relead-2) R:ndles 7
Note that the new core was relcaded with 1h additicnal standard tundles, l
1 3 'instru=ented bundles and k deveic;cental bundles that had been re=:ved frc= the core during previous cutages. Sc=e "C" fuel bundles steving
=inor leak sipsis vere instaned to ec=plete the Bh-bundle core This type core leading was necessary due to the pmture failures of the 4
"C" fuel.
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Core Internal Insrection Sufficient core internals were re=oved froc the reactor vessel to facilitate a visual inspection of a =ajority of the core fasteners and co=ponents. Fuel handling poles were used to deter =ine if any vessel fasteners or other co=ponents were structurally locse.
There vere no loose fasteners nor vere there any co=ponents that had lost integrity.
C.
Statistics The reactor was brought critical 15 ti=es and was critical for 3,h5h.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during this reporting period. Total heat produced by the reactor was 655,98k INht. The turbine generator was en the line for 3,h01.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and the electrical generation was 206,6Sk !?-h (gress).
Net Plant electrical' generation for this Ieriod was 195,161.1 !&h.
D.
P&D Progra Six center:elt fuel bundles were inserted in the core to gain
' experiences v!.th various types of center-tilting in different fuel ty;es.
(See A=end ent No 1 to Facility Operating Lietnse No DPR-6 dated March 12, 1968.)
II.
ROUTINE BWASES,' DISCHARGES AND SHIFIC;T OF RADICACTIVE 1&TERIAL is it-1 fc A total of approximately M IX d, y
-,m A.
curies of activation and fission gases was released to the environs via the stack. This figure is based on 2,733 EFFH of operation at an average release rate ~ of h
1.0 x 10 pC1/sec.
Off-gas activity increased to a maxi =u of 21,200 pCi/see in Dece=ber. This increase.in activity indicated new fuel failures. These were verified by dry sipping during the refueling outage.
B.
Liquid radioactivity released to Iake Michigan, by way of the circulating water discharge canal, numbered 6h batches, with a total activity of 1.82 curies. Fou-of the batches were released on a par-tially identified basis whereas.at least 90% of the activity was deter-l 1
lo3 58, g,95 r.ined to be a.co=bination of Ba
, Ba
, Ia
, Ru 2
kl and Zn'5 All other batches were released under Nb95, Ce
, Ce unidentified li=its.
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6 C.
The following is a tabulation of shiptents of radioactive materials:
Shipcent Transfer No Date From Transfer to Radioactive Material 1
1-9-68 DPR-6 Isotopes, Inc Two bottles (shoo el capa-29-55-6 city each) 0.c03 mci con-taining radioactive liquid.
2 1-13-68 DPR-6 Nuclear Engg 135 cu ft radioactive spent 16NSF1 resins - L08 C1.
3 1-17-68 DPR-6 Nuclear Engg 56 cu ft radioactive spent 16NSF1 resins - 84 Ci and 16 drums of low-level dry vaste.
4 1-18-68 DPR-6 Nuclear Engg 183 cu ft radioactive spent 16NSF1 resins - 366 C1.
5 1-19-68 DPR-6 Nuclear Engg 56 cu ft radioactive resins 16NSF1 (spent) - 67 2 Ci and 20 drums of low-level dry vaste.
6 1-20-68 DPR-6 Nuclear Engg 183 cu ft radioactive spent 16NSF1 resins - 450 C1.
7 1-23-68 DPR-6 Nuclear Engs 56 cu ft radioactive spent 16NSF1 resins - 56 Ci and 20 drucs of dry low-level vaste.
8 1-24-68 DPR-6 Nuclear Engg 16h cu ft radioactive spent 16NSF1 resins - 296 C1.
9 1-25-68
-DPR-6 Nuclear Engg Equipment used to pump 16NSF1 resin - 0.2 mci.
10 1-29-68 DPR Nuclear Engg '
waste - 0.4 C1.
97 drucs of low-level 16NSF1 11
.2-9-68 DPR-6 Nuclear Engg 102 cu ft of dry. solid' 16NSF1 radioactive vaste. - 7.67 C1.
12 3-20-68 DPR-6 US Naval Iab Reactor vessel surveillance 8-1393-2(A-66)' specimen - 12.k Ci.
13 3-26-68 DPR-6 Dresden Power Winch and motor - fuel pool Sta No 1 equipment - 0.01 mci.
f 12-5650-1 14 h-18-68 DPR NPI 32-rods of Co 459,000 19-12667-01 C1.
1
T D.
Environmental Monitoring Program Appendix A shows site and region maps as well as plotted data of selected air, water and film station sample analysis results. The charts have been updated to include all of 1967 and have been changed as follows: Reactor power level, off-gas activity and environmental film results appear on one chart. Off-gas activity and air particulate activity appear together on a separate plot. Finally, releases to Lake Michigan and radioactivity concentration in both condenser inlet and canal discharge waters make up the final chart. These changes will allow a more direct comparison to be made between environmental moni-toring results and plant vaste discharges.
III. RADIOACTIVITY LEV? 3 IN PRINCIPAL FLUID SYSTBi (FOR SIX MONTAS)
A.
Primary Coolant 1
Minimum Average Maximum Reactor Water Filtrate "
-3
-1 pCi/cc 1.1 x 10 5.5 x 10 1.7
' Reactor Water Crud
-3
-1 pCi/ce/ Turbidity 2.9 x 10 2.9 x 10 1.4 Iodine Activity
-5
-1
-1 pCi/cc 2.0 x 10 3.0 x'10 6.7 x 10 B.
Reactor Cooling Water System The principal radionuclides in the reactor cooling water system were K-h2 and Cr-51, which resulted from the activation of the potassiu=-
chromate inhibitor.
-Reactor Cooling Water "
-2
-1 pCi/cc 1.0 x 10 2.9 x 10 1.5 x 10 C.
Spent Fuel Pool Radioactivity in the spent fuel pool is principally activated corrosion products going into solution.from stored fuel and core components.
-2
-2 Fuel Storage Pool "} pCi/cc 1.; x lo 2 9 x 10 8.7 x 10
~
-3 Fuel Pool Iodine pCi/cc 1 5 x 10 5 0 x-10 ' 2.0 x 10
- A counter efficiency based on a ganna energy of 0.662 Mev and one gamma photon per disintegration. Decay scheme is assumed.to convert count rate j
to microcuries. All count. rates were teken at two hours after sampling.
Based on efficiency of Iodine-131 two hours after sampling.
- Based on APHA' turbidity units and 500 ml of filtered samples.
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8 IV.
PRINCIPAL MAINTENANCE PERFOPED A.
The No 2 reactor recirculating pu=p was removed from its case during the refueling cutage and codified for installation of a new cartridge type shaft seal. This modification required extensive cachining of the pu=p shaft and a major rework of the seal and cooling water piping syste=s. The special shield tank mentioned in the last report was used for this rebuild work. The carbon faces from the removed seals were in excellent condition, but the inner and middle rotating faces shoved siEns of deterioration, although fully intact.
After the No 2 recire pu=p sTs reassembled, it was test operated.
During heat-up, danage occurred to the pump. The plant was shut down and the pu=p disassembled to check on the extent of da-age and to deter =ine the cause. The following damage was discovered:
1.
Three labyrinths grooved and wiped.
2.
Carbon bearing chipped and cracked.
3 Bearing cavity damaged.
h.
KinEsbury thrust shoes (in totor) damaged.
5 Auxiliary cooling water impeller damaged.
To effect repairs, a redesigned bearing retainer was fabricated by Byron-Jackson. The suction cover was machined by Plant Maintenance to insure the proper fit of the cartridge bearing.
B.
No 2 control rod drive pu=p was completely overhauled. Excessive wear was experienced with the old pistons and crosshead stubs. New ni-trided replacements were installed. The chrote-plated pistons in the No 1 control rod drive pump were replaced with nitrided stainless steel pistons. New guide bushings and packings were also installed.
C.
The motor operator on the leaking d-c outlet valve (MO-TC63) was checked and found to be operating satisfactorily. The valve was disas-se bled and one side of the gate and seat was found to be marred. Repairs consisted of honing both faces of the gate valve.
D.
New stainless steel tube bundles were installed in each of the three feed-water heaters (ES, IPandHP). Previous experience with these heaters during operation indicated the following:
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1.
Exfoliation and oxygen attack occurred on the 70%-30%
copper-nickel tube material in the HP heater. The tube bundle was physically removed from service and bypassed for months after many tubes ruptured.
2.
Chemical tests showed oxygen attack in all heaters, thus causing many of the corrosion products noted in the feed-water system as well as the reactor.
V.
CHANGES, TESTS AND EXPONE!.TS PERFORMED PUPSUANT TO 10 CFR 50 59(a)
'This section describes the changes made to the facility within the six-month period without prior Commission approval pursuant to Sec-tion So.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 Hazards S m ry Report (FHSR).
It also included tests and experiments carried out at the Plant without prior Ccm=ission approval pursuant to Section 50 59(a).
Each change, test or experiment described was authorized only after a finding by Consumers Power Company that it did not involve a change in the Technical Specifications incorporated in Operating License DPR-6 (effective May 1, 1964) or an unreviewed safety question.
A.
Facility Chane;es 1.
A time delay was installed on the anion low dilution water flow alarm in the anion slow rinse step. This was to improve operation of the regeneration facility.
2.
An auxiliary relay (AR X) and circuitry were installed to provide actuation of the 386B turbine bypass auxiliary when the 199 OCB is tripped open manually by the console control switch. This auxiliary relay will provide an opening signal to the bypass valve.
i In the past, the opening signal was generated only on the.
loss of a tone relay signal to the 199 OCB between Emet Substation and Big Rock Point.
3 Piping was changed to the sphere high-pressure sensors to provide a more reliable protection system. Two pipes instead of one now feed these sensors to insure a redundant system.
4.
A1 pressure _ sensing line was added to monitor pressure in the core spray line to detect a piping rupture between the reactor vessel and motor-operated valve _(MO-7061).
(See reference letter to Dr. Peter A.
Morris dated _ March' 7,:1968.)
10 1
5 Thirty-two General Electric speed control valve canifolds were installed (two of them during the last cutage). Thirty-two cooling water stop check valves were modified so they would operate as shutoff valves only. A shutoff valve was also installed between the manifold and selector valve to facilitate maintenance.
These changes were made to insure better operating perfomance and to reduce maintenance.
6.
A nev 46 kv transmission line was installed to provide a backup source for station power. The changes to the Plant Substation included two new 2.h kv OCBs and a new 5000 kva, 46 kv/2.h kv transfomer.
Associated with this equiInent is a throv-over scheme to provide automatic transfer to the 46 kv source upon loss of nomal voltage on the 2.h kv station bus.
7 A new d-c motor-operated isolation valve was installed between the main steam line and the turbine bypass valve. This valve vill insure a pcsitive shutoff in the turbine bypass valve line in case of an emergency when the bypass valve vill not shut off.
8.
A vacuum interlock was installed on the new d-c operated turbine bypass isolation valve. This interlock vill keep the valve closed on-loss of condanser vacuum.
(See reference letter to Dr. Peter A.
Morris dated June 23,1967.)
B.
Tests 1.
Temperature Coefficient Test This test was conducted in March prior to power operation with the newly loaded core. Test data indicated that the coefficient turned negative at 157 F after adding 13 7 cents of reactivity.
2.
Fuel Sipping Test Many fuel bundles were tested for clad material integrity, utilizing the." dry sipping" technique. This procedure was discussed in previous semiannual reports.
3 Test of 46 Kv Throvover On March 12, 1968, the automatic transfer scheme was given an.
overall test after the installation of the new 46 kv line. Prior to the test, controls and timers for tripping and reclosing each of the motor
. (
circuits involved in the power return sequence were tested individually.
The overall test was a success with only two minor discrepancies noted.
The discrepancies are now being corrected.
11 4.
Reactor Vessel Expansion Test Four instruments were placed under the reactor vessel to measure downward displacement of the bottom of the reactor vessel during heat-up.
These measurements indicated a movement of 0.66" to 0 71".
This informa-tion vill be used for design of the control rod drive support fixture.
VI.
PERIODIC TESTING PERFORMED AS REQUIRED BY THE TECHNICAL SPECIFICATIONS The following tabulation shows the required frequency of testing, plus the testing date of the systems or functions, which may be periodically tested per Technical Specifications:
System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives Continuous withdrawal and insertion Each major refueling and 12/10 /67
. of each drive over its stroke with at least once every six 3/ 3/68 normal hydraulic system pressure.
months during periods of Minimum withdrawal time shall be 23 power operation.
seconds.
Withdrawal of each drive, stopping Each major refueling and 12/10/67 at each locking position to check at least once every six 3/ 2/68 latching and unlatching operations months during periods of
_ and the functioning of the position pcVer operation.
indication system.~
Scram of each drive from the-fully Each major refueling and 12/10/67 withdrawn position. Maximum scram.
at least once every six 3/ 2/68 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 3/ 3/68 entire stroke with reduced hydrau-not less frequently than
.lic system pressure to determine once a year.
that drive friction is normal.
Control Rod-Interlocks Rod withdraval blocked when any two Each major refueling but 3/.3/68 accumulators are at a pressure not less frequently than below 700 peig.
once every 12 months.
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t
t 12 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 3/ 3/68 three power range channels read be-not lesa frequently than low 5% on 0% - 125% scales (or below once every 12 months.
2%ontheir0%-40% scales)when reactor power is above the minimum operating range of these channels.
Rod withdrawal blocked when scram Each major refueling but 3/3/68 dump tank is bypassed, not less frequently than once every 12 months.
t Rod withdrawal blocked when mode Each major refueling but 3/ 3/68 selector-switch is in shutdown not less frequently than position.
once every 12 months.
Other Liquid poison system component Two months or less, 10/31/67 check.
116/68 3 3/68 5 3/68 Post-incident spray. system auto-At each ma.jor refueling 3/ 2/68 matic control operation.
shutdown but not less frequently than once a year.
Core spray system trip circuit.
Not less frequently than 3/ 2/68
.once every 12 months.
Bsercency condenser trip circuits.
Not less frequently than 3/12/68 once every 12' months.
' Containment Containment sphere access air
'Once every six months or.
3/ 3/68 locks and vent valves, leakage less.
rate.
Isolation valve operability and
.At least once every 12.
6 /13 /67 leak-tests.'
months.
Isolation valve controls and in-Approximately quarterly.
11 5/67
- strumentation tests.
3 12/68-
[
PendtrationLinspection.
At,least once every 12 6 /10 /67 months.
L ntegrated leakitest.
..Once every two years.
' h/30/66 I
s
13 The following instrument checks and calibrations were performed
.at least once a mor" Reactor safety system checks on circuita not re-quiring plant shutdown to check; air ejector off-gas monitor calibration; stack-gas monitor calibration; emergency condenser vent monitor calibra-tion; process monitor calibration; and the area monitoring system cali-I-
bration.
VII.: PERSONNEL TRAINING Three operaters received an AEC Operator License during January.
This is a continuation of our efforts to provide additional depth in anti-cipation of staffing the Palisades Plant.
By Robert L. Haueter (Signed)
Robert L. Haueter Assistant' Electric Production Superintendent - Nuclear Consumers Power Company Jackson, Michigan Date: June 24, 1968 Sworn and' subscribed to before me this 24th day of June 1%8.
t
.(SEAL)
Grace Warner (Signed)
. Notary Public, Jackson County, Michigan
-My commission expires January-15, 1972 1
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