ML19340A522
| ML19340A522 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/28/1972 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8008070711 | |
| Download: ML19340A522 (35) | |
Text
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Ilipp n.< 3 UNIT #1 ANNUAL REPORT YEAR, 1971 l
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COW.ONb'EALTH EDISON COMPANY DRESDEN NUCLEAR P0b'ER STATION UNIT #1 A 0"JAL REPORT CF STATION OPERATION FOR THE YEAR 1971 January 28, 1972
TABLE OF CONTENTS Page I.
Introduction..
1 II.
Summary of Operatians 1
A.
Scope of Operations 1
]
B.
Shutdowns 1
l C.
Load Restrictions 2
III. Discussion.
2 A.
Operating Experience.
2 1.
Generation.
2 2.
2 3.
Incidents 3
4.
Control Rod Drive 7
a.
Control Rod Drive Tests 7
b.
Control Rod Drive Inspection.
7 5.
9 a.
Blade Follow Checks 9
b.
Control Blade Inspection....
9 6.
Changes in Facility 10 a.
Turning Vane Installation 10 b.
Co're Spray Installation 10 c.
Vessel level sensor installation.
10 d.
Relocation of Motor Operators 10 7.
Personnel Radiation Exposure.
10 8.
Liquid Poison System.
11 9.
Radioactive Waste Disposal..........
11 10.
Fuel Assembly Cleaning and Testing.
11 a.
Fuel orifice cleaning 11 b.
Bow checking.
12 c.
Orifice Changes 12 11.
New Fuel - Cycle 8..............
12 12.
Inspections 12 a.
Irradiated Fuel Inspection.
12 b.
Metal Surveillance 13 1.
New Specimens inserrol.
15 3e 3
c.
Emergency Condenser Internal Inspection 15 d.
Reactor Vessel Inspection 15 e.
Primary System Weld Inspection..
16
- 1) Reactor Flange Inspection 16
- 2) Reactor Thimble Welds 16
- 3) Primary Piping..
16 a) Examinations.
16 b) Hydrostatic test.
17 f.
H.P. Turbine Inspection 17
- 13. Feedwater Heater Replacement.
17 14.
Piping Failures and Repairs 18
- 15. Tests 18 s.
Sphere Integrity Test Program 18 b.
Primary Steam Drum Safety Valves.
18 c.
Temperature Coefficient of Reactivity Check 18 d.
Fuel Sipping........
18 e.
Control Blade Depletion Test.
19 f.
Shutdown Margin Checks...
19 B.
License DPR-2 19 Correspondence References 1971..........
20 Appendix A...
21
~
i.
i
_DRESDEN NUCLEAR POWER STATION ANNUAL REPORT I.
Introduction This tenth annual report is submitted in compliance with paragraph 3.c (2) of the Utilization Facility License DPR-2, as amended, and covers operation of Dresden Nuclear Power Station Unit #1 during the year 1971.
II.
Sumnarv of Operations A.
Scope of Operations Operation of Dresden Unit di continued from the preceding year at various loads throughout the cycle.nd experienced a total of 8 outages. The unit experienced all primary heaters out of service from December 1970 to late April 1171. Operation continued with three secondary loops until June 23, 1971, when "A" secondary steam generator loop was returned to service, returning the unit to four loop operation. The last outage, the seventh partial refueling, inspeation and maintenance outage, began on September 10, 1971 and continued into 1972. The outage included a considerably augmented in-service inspection program based on the " Interim Acceptance Criteria for Emergency Core Cooling bystems for Light-Water Power Reactors", dated June 19, 1971; installation of a new turning vane and core spray system; replacement of tube bundles in two feedwater heaters ar.d one drain cooler; installation of new vessel level sen-sors and replacement of five control rod drives; inspection of H.P.
j turbine; refueling of 120 fuel assemblies; cleaning and flow testing of all fuel returning to the core for cycle 8 operation; and general maintenance and inspections made available by the shutde'sn.
During the year, additions to, and changes in, facility design were made by:
Installation of a new turning vane and core spray system; installation of new reactor low water level sensors, and relocation q
of the motor operators on va'.ves MO-101 and MO-109 in the condensate return lines of the emergency condenser.
One irradiated fuel shipment. consisting of one poison rod and six-teen fuel rods from varioug assemblies was made in September t o General Electric Ccmpany, Vallecitos, California for analy ais.
B.
Shutdowns The plant was shutdown eight times during 1971 as shown in Table 1 t
and Figure 1.
Six of the shutdowns were forced: one was caused from
1
. a failure of an " actuator amplifier" in the incore monitoring system; two by steam leaks ("A" cleanup line and extraction piping); one by a spurious signal on the nuclear instrumentation when a contractor broke a static wire in the 138 KV yard; one due to an operator error; and one due to a rapid over pressurization in the off-gas holdup line.
Two were scheduled shutdowns: one for repair of feedwater heater leaks, steam line drain leak, friction and scram testing and other mainten-ance; and one for the seventh partial refueling, core spray installa-tion, new turning vane installation, installation of new Reactor low water level sensors, replacement of five control rod drives, inservice inspection, and other maintenance. The total unit outage time from these eight outages was 3237 hours0.0375 days <br />0.899 hours <br />0.00535 weeks <br />0.00123 months <br /> 22 minutes.
C.
Load Restrictions The load restrictions imposed during the year are listed in table 2.
The restrictions were due to defective fuel, feedwater heater out-ages, and fuel burnup limitations.
III. Discussion A.
Operating Experience 1.
Generation The total reactor operating (critical) time during the year was 5620.38 hrs, and the total power for the period was 99,078.0 MWD. The gross electrical generation during the year was t
649,667.90 MWHe; net generation was 602,597.1 MWHe. As of December 31, 1971, the total gross generation since commencement of power operation on April 15, 1960, was 11,372,900.67 MWHe.
2.
Scrams a.
On January 14 at 1715 the reactor scrammed due to a spurious signal from the incore system.
The unit was operating at 140 MWe when an actuator amplifier in the incore monitoring system failed, scramming the reactor.
b.
On April 15 at 0557 reactor scrammed.
Tae reactor was' brought critical at 0453 and a positive 100 second period was established at 0550 as part of a ten.per-ature coefficient test. At 0557 a spurious signal occurred on Channel 7, indicating a short period, and subsequently scrammed the reactor. A detailed check found all systems related to Channel 7 functioning properly.
c.
On April 15 at 0645 reactor scrammed.
The reactor was brought critical and while the instrumentalton.
= _ _
3 was on the 7 MW range, the reactor scrammed due to up scale readings on the power range instrumentation. Channels 1 and 6 reached their trip points before the operator could range the instruments.
d.
On April 17 at 1534, the reactor scraemed due to a high incore flux.
While the reactor was in operation, a spurious signal occur-red when contractors working in the 138 KV yard broke a static wire on line 1210. The static wire fell across lines 1210 and 1207 tripping both lines causing enough voltage surge to affect the incores, e.
On May 2 at 1050 reactor scrammed.
The reactor was at approximately 85 MWe, when preparations were underway to increase power to 100 MWe. An operator closed B & C Primary Feedwater heater bypass valve rather than the "C" drain cooler bypags valve as directed. This resulted in a loss of feedwater supply and low primary drum water level which resulted in actuation of the primary drum low water level scram sensors.
3.
Incidents a.
Crack in "A" Cleanup Loop Suction Line.
During the latter part of January, air activity sampling showed an increase in airborne activity by a factor of 30 in the sphere east pipeway. This increase in activity initiated visual inspeccion that led to the detection of steam blowing through the insulation on a 4 inch stainless steel line, that is part of the primary piping system, which supplies the "A" cleanup demineralizer system. The unit was shutdown to allow further examination.
A crack was found downstream of the point where the 4 inch line is connected into the downcomer-header equalizer line of the primary piping system. This supply line had been valved out-of-service downstream of the crack at the cleanup heat exchangers since 1965. The crack was located in a section of pipe in the heat affected zone of a weld connect-ing a replacement pipe section installed in 1967. The out-side surface length of the crack was 3/16 inch and the inside
,ch was approximately 1 to 1 1/4 inches. The crack appeared run circumferential1y around the pipe in the heat affected zone of the weld.
The repair was initiated and welding was performed in accordance with procedures approved and qualified for Unit 2 and 3.
Documentation of the weld repair, materials and inspections was obtained.
We believe this instance once again demonstrates the ade-quacy of our leak detection system to detect leaks of small magnitude. It also reinforces our conclusion that small diameter 300 series stainless steel pipe is tough and ductile and if it does fail, will crack progressively until a leaking crack is developed and will not fail in a brittle or catas-trophic manner.
b.
Failure of Primary Drum Level Sensor Relay.
On March 11, 1971, Unit I was opera;.ing at 140 MWe with steady-state conditions. At 9:30 a.m. a fuse connected to the primary drum level scram rela: blew resulting in a trip of the "A" sa fe ty sys tem.
No abnormal conditions were noted af ter the fuse was replaced and the safety system was reset.
At 10:30 a.m. a burning odor was detected by the reactor operator and an investigation of the safety system relay board revealed tha t the same primary drum level scram relay fuse had blown again. The relay contacts were still closed and the protective glass was coated with a brownish film.
Because the "A" safety system had not tripped, as it should have, the fuse in the second primary drum low icvel scram relay in the "A" safety system was pulled at 11:00 a.m.
to trip the "A" safety system. This was accomplished to place the unit in the most conservative operating condition until the faulty relay could be replaced.
The failed relay was replaced at 1:10 p.m. - subsequent eval-uation of the event and examination of the relay indicated tha t the relay coil had shorted and the insulating material, having become sof t and tacky, had fallen on the relay contacts holding them in the closed position; preventing the safety system from tripping when the fuse blew.
This is the first known failure of this type in the cen years of operation of Dresden 1 and it is felt that the probability of this type of failure is small. The circuit design antici-pates the possibility of a relay failure by providing redundant relays. No further action is planned.
c.
Failure of MO-lCl on the Unit inEmergency Condenser System.
On April 13, 1971 during the pre-startup checks, the emergency condenser condensate return valves, MO-101 and MO-109, were closed. About one hour af ter they were closed, a 125 V DC
-S-ground appeared. Upon investigation of the 125 V DC ground, it was found that }0-101, the north emergency condenser condensate valve motor operator had failed. Inspection of the motor revealed that the insulation had flowed and the windings had low resistance to ground.
The motor was replaced and during testing, the motor failed to de-energize when the valve closed.
Af ter waiting approximately 20 seconds, the operator tripped the breaker manually. The torque switch was reset to a lower value and the valve was operated satisfactorily with the torque switch de-energizing the motor when the valve closed.
Investigation of the records showed that the torque switch settings had been increased in December 1969 to reduce leak-age through the valve. It was concluded that the high torque setting prevented the motor from tripping as required, thus causing a continuous supply of current to the motor and sub-sequent overheating of the varnish and insulation on the motor windings.
Corrective action entailed additional insulation installed on the valve to reduce operational temperatures, ventilation was directed to the motor for additional cooling, and a thermocouple has been placed on the motor to monitor operating temperatures during startup. As added assurance, we are relocating both valve operators outside the area during the 1971 Refueling Outage so that the motors will be accessible during operation, and in a cooler location.
d.
Malfunction of No. 4 Micromicroammeter.
On July 26, 1971 Unit I was operating at 450 MW, steady-state t
conditions. The operator noticed a red trip indication on micremicroammeter channel 4 without the corresponding Reactor Protection Channel Trip.
The trip indication was reset by the operator. The micro-microammeter was tripped manually by depressing the tri'p test button. The trip indication came on but the Reactor Protec-tion Channel Trip did,not occur. The auxiliary plug, which connects the micromicroammeter to the Reactor Protection Channel, was then unplugged causing the scram actuation relay to open giving the expected Reactor Protection Channel Trip.
This pinpointed the problem as internal to the micromicroam-meter. Further investigation found that the micromicroammeter trip relay would not open when de-energized. Sticky armature pivots prevented the contacts from opening. Apparently poor soldering techniques had allowed moisture to get into the
' f hermetically sealed relay. The detective relay was replaced and the micromicroammeter was returned to service. All six micromicroammeters were checked to insure that all micro-microammeter relays would open and give a Reactor Protection Channel Trip - all were satisfactory. The routine surveil-lance procedure for checking the micromicroammeters, has been modified so that each micromicroammeter will be checked to veri fy that a trip of the micromicroammeter will also give a Reactor Protection Channel Trip.
e.
Unplanned Release of Radioactive Liquid to the Rivar.
On Friday, September 10, 1971, Unit I was shutdown in pre-paration for the seventh partial refueling outage. The IB unloading heat exchanger, which removes decay heat from the reactor water and transfers it directly to the service water system, was placed in service on Saturday, September 11, 1971, at 0658. On Monday, September 13, 1971, at approximately 1515, analysis of a sample collected from the IB unloading heat exchanger service water outlet showed an activity of approximately 60,000 p ci/1, indicating a heat exchanger tube leak. The heat exchanger was secured at 1540.
Gross be.ca activity analysis of grab samples collected from the Unit 1 intake and discharge canals at the time of the l
Jelease, indicated on activity of approximately 195 p ci/l
]
above background. A gamma scan of the sample from the unload-ing heat exchanger service water outlet showed that the activity was typical of reactor water, as expected, and that l
Iodine-131 constituted approximately 28.77. of the total activity. The Iodine-131 activity, which is not included in the gross. beta analysis, was conservatively estimated at 86 p ci/1, so that calculated total activity in the canal, attributable to the tube leak is 281 p ci/1, or 281% of the 100 p ci/l unidentified activity permitted by 10CFR20. An isotopic analysis of the activity released indicates that the release was 29.77, of that permitted by 10CFR20 for identified activities. The isotopes, other than Iodine-131, contributed i
only an additional 17. of the identified limit.
A review of the recorder traces from the service water ef fluent process monitor indicated that the release began as soon as the unloading heat exchanger was placed in service. The trace indi.
cated an increase from the background reading of 0.8 - 1.1 x 104 CPS to 1.4 - 2.0 x 100 CPS at the time the heat exchanger was cut in.
The operator on shift at the time apparently failed to note the increase. From that point on, the recorder trace increased at a very gradual rate, until at approximately 0615 on Sunday, Sep' ember 12, 1971, it reached the alarm point of t
4 x 104 CPS (40% on scale). The operater looked at the monitor chart, but because the rise was at such a slow rate, the trend J
l l
t 7
was not obvious, so he up-ranged the monitor to cicar the ala rm.
The monitor had increased to approximately 4.2 -
5.8 x 104 CPS at the time the release was terminated. The grab sample used to calculate discharge activity was col-1ected at the end of the discharge period and is therefore believed to be conservative, since the leakage would be expected to remain constant oc increase with time.
To minimize the potential for reoccurrence we have modified the operating procedures to include a check of the prccess radiation monitor by the operators and the pulling of a grab sample of the service water effluent following the placement in service of an unloading heat exchanger. The release had been discussed with personnel involved and the tmportance of eliminating unplanned releases of radioactive materials to the environment has been re-emphasized. Also an engineering review of the system has been requested from our Mechanical and Structural Engineering Department.
4 Control Rod Drives a.
Control Rod Drive Tests 1
On April 10, 1971, all control rod drives were scram and friction tested, latched and timed for normal insertion and withdrawal.
1 As a routine check of drives to be overhauled during the fall refueling outage, all drives except F-9, J-8, H-5, H-9, and G-9 were again tested on September 14-15, 1971. Based on*this and previous tests, five drives were selected for replacement. These drives were replaced in November, 1971.
Scram and friction tests were not conducted before the end of the year.
b.
Control Rod Drive Inspection The Technical Specifications to Dresden's License DPR-2, as amended by change No. 12, dated November 23, 1966, states that during major outages, "Not less than two control rod drives mechanisms shall be removed, disassembled, and thor-oughly inspected at intervals not to exceed 24 months."
In addition to license requirements, drives are removed for inspection on the basis of drive test results and malfunctions experienced during operation.
Prior to shutdown-for the seventh partial refueling outage on September 10, 1971, five control rod drives were selected for removal and overhaul.
i Four spare drives (1229, 1291, 1253, 1235) plus one (1245) of the five removed from the core, were inspecteo during the outage and satisfactorily tested in the control rod drive test facility following overhaul. The four spare drives were used to replace four drives removed for over-haul (1307, 1233, 1280, 1251). Drive 1245 was removed, inspected, overhauled, and relocated back in the core.
The four drives that were replaced will be overhauled at a future date.
Af ter the removal of all fuel elements and all control rod blades, each of the five control rod drives was removed with its index tube fully withdrawn. The drives were then transported to the drive shop area at sphere elevation 565' for future disassembly and inspection.
1.
Visual Inspection Af ter drive disassembly, all parts were visually inspected as closely as radiation levels would permit. A k" plexi-glass face shield was used to protect against beta emission while reviewing the parts. All moving parts on the roller mount were actually moved to verify freedom of movement.
Visual inspection of the roller mount assembly was done at about a 3 foot distance or viewed underwater.
All canned magnets were placed in hot water and boiled to ensure the integrity of the magnet housing.
2.
Dye Penetrant Inspection A dye penetrant examination was conducted on drive com-ponents. A cleaner (SKC-S), penetrant (SKL-S), and developer (SKD-S) was used.
Af ter ultrasonic cleaning, parts were sprayed with pene-trant and allowed to sit 15 minutes to allow capillary action time to draw penetrant into any cracks. The dye was then wiped off with lint free rags removing all dye except that in small scratches and cracks. The part was sprayed with developer in order to draw dye from any cracks. The dye will then readily stand out against the white background and indicate if any cracks or scratches are present.
3.
Results of Inspection j
Nitrided guide roller pins were used to replace all a.
chrome plated pins in the roller mount assemblics.
. i b.
All rollers were checked with a go-no-go gage.
This gage has an outside diameter of 0.260".
The maximum tolerance for the hole of a new roller is 0.259", thus only one thousandth of an inch was acceptable for roller wear.
None l
of the twenty rollers inspected were replaced.
i c.
Dye checking revealed cracking or flaking chrome on one collet assembly, one guide plug, and two piston springs, d.
Figure 2 summarizes the locations of the drives i
removed for overhaul while Table 3 summarizes
.j the inspection resd es.
5.
Blade Following Checks During periods of operations, control rods have been verified for blade following on a weekly basis. During each startup, control rod patterns for criticality have been predicted and a
I all blade following verified.
I b.
Control Blade Inspection On November 2 and 3, 1971, six control blades scheduled for service in Cycle 8 were inspected. The blades were checked i
for bowing, twisting, angularity, and thickness by passing
- l formed patterns (go-no-go gage) over the blade. Visual examinations were made of all six blades to check for crudding, i
tubc integrity, excessive roller wear, and any surface damage.
All six blades were found to'be acceptable and returned to the j
reactor. The blades that were examined and their locations in the core are exhibited below:
7.
- c c e w c _ Yi I: Z:
l l
l
- n 3
l c
Bl' O
7 B-45 k'
h6 s
' 'I B
4 tfL g
B-19__
u i
1 Control Blades Inspected i
4 6.
Changes in Facility Design a.
Turning Vane Installation A new turning vane was installed in the reactor in December, 1971, during the seventh partial refueling outage. The new turning vane houses the ring header for the new core spray system. The description and design was covered in proposed change No.17 to License DPR-2 dated October 10, 1968.
b.
Core Sprav Installation A core spray syste, is in the process of being installed on Unit 1.
This system is part of Commonwealth Edison's program to meet the AEC's adopted interim acceptance criteria for the performance of emergency core cooling systems (ECCS) for light water reactors with which we will have to comply by July 1, 1974. Completion of the system and operational checkout is not expected to be until early 1972 The description and design was covered in proposed change No. 17 to License DPR-2 dated October 10, 1968, c.
Vessel level sensor installation.
The old float-type Magnetrol reactor low water level sensors have been replaced by Yarway (indication type) level sensors.
The magnetrol devices are being replaced because they have a history of unreliability due to crud accumulation which impairs the switch operation. Also, the Yarway's have an additional set of contacts which will be used to initiate the core spray system that is being installed during-the seventh partial refueling outage, d.
Relocation of Motor Operators From a past history of motor failure due to excessive heat, the motor operators on the emergency condenser condensate return lines, valves MO-101 and MO-109, were removed from the steam drum compartment to the Emergency Condenser (649') level.
The movement of these motor operators does in no way affect the safety function of these valves.
It does allow access to the operators during upit operation.
7.
Personnel Radiation Exposure Personnel exposeres to radiation during 1971 were within the limite specified in 10CFR20.
l 1
8.
Liquid Poison System The liquid poison system was operative at all times during the year. The boron poison was sampled on April 12 and October 7.
There were no conditions which would indicate a loss of boron from the solution tank.
Boron concentrations in the reactor water remained below detectable limits throughout the year.
9.
Radioactive Waste Disposa_1_
Release of radioactive liquid waste was accomplished in batch quantities at controlled release flow rates according to estab-lished procedures. The contribution to the activity of dilution water was always maintained within the limits specified in the applicable federal regulations.
The average contribution to the unidentified activity in the water utilized for radioactive liquid waste dilution during the year was calculated to be 0.209 x 10-7 uCi/ml (20.9 uuCi/1)
(river background not included) compared to an average limit of 1.00 x 10-7 uCL/ml (100 uuCi/1) for unidentified mixtures contain-ing no Ra-226, Ra-228, or I-129 as specified in 10CFR20.
Solid radioactive wastes were stored on-site pursuant to AEC License DPR-2.
Table 4 shows the radioactive waste shipments made during the year. A total of 18 radioactive waste ship-ments were made in 1971.
The concentration of noble fission gases in the chimney discharge to the atmosphere was maintained well within license limits. The annual average noble gas release rate from the chimney was approx-imately 23,850 uCi/sec.
There was one shipment of privately owned spent fuel in 1971. Cn September 2,1971, sixteen (16) fuel rods and one (1) poison rad (Gd 0 ) were removed from bundles discharged in 1969 and shipped 23 by motor freight to General Electric Company, Vallecitos Nuclear Center, Pleasonton, California 94566. Table 5 is a breakdown of Commonwealth Edison Company spent fuel shipments for reprocessing since initiation of fuel shipping in June, 1965. There were no Commonwealth Edison Company owned spent fuel shipments in 1971.
10.
Fuel Assembly Cleaning and Testing a.
Fuel orifice cleaning Fuel assembly inlet orifice cleaning was initiated on October 15, 1971, on all irradiated fuel assemblies scheduled for use in Cycle 8.
All orifices were manually cleaned and the ef fect-iveness of the cleaning determined by flow testing.
,. b.
Bow Check.ng All fuel assembly channels on fuel scheduled for use in Cycle 8 were bow checked. Five assemblies were found to have bowed channels, i.e., UN082, UN145, DU97, G5, and E109. The bowed channels were removed and replaced with spare channels and bow checked a second time. All five assemblies passed the check.
c.
Orifice Changes Various orifice changes were made in accordance with Gulf United Nuclear's refueling plan. These changes were neces-sary to facilitate moving of the higher exposed fuel to the periphery and the lower exposed fuel to the center of the Core.
11.
New Fuel - Cycle 8 One hundred twerty (120) new Gulf United Type VIII fuel assem-blies have been placed in the reactor. These fuel assemblies are sLmilar in nuclear and thermal hydraulic characteristics to the Type VI and VII fuel previously loaded in the core which was purchased from United Nuclear Corporation.
The Cycle 8 core reload consists of one hundred seventeen (117) assemblies being loaded in a 3 in 8 scatter in the central region of the core. The other three (3) assemblies were loaded in the periphery of the core.
Ninety-five (95) new assemblies were loaded with high flow type IB orifices and the remaining twenty-five (25) assemblies were fitted wit, the more restrictive type 2B orifices.
h Of the one hundred twenty new assemblies, there are eight (8) special assemblies. Four (4) are instrument assemblies (UN301, UN302, UN303, and UN304). The other four (4) assemblies (UN227, UN234, UN253, and UN257) were sent back to Gulf United where i
part of the rods in these assemblies were vacuum outgassed using j
a new process.
12.
Inspections a.
Irradiated Fuel Inspection A fuel inspection on Unit 1 Cycle 6 fuel was conducted by General Electric around the middle of the year.
Sixteen (16) fuel rods and one Gd 0; rod from nine fuel assemblies were 2
removed and shipped to Vallecitos for further analysis. The rods and bundles involved are listed in Table 6.
The fuel was shipped on September 2, 1971.
, Inspection of irradiated fuel at the end of Cycle 7 was performed in the fuel building by both Gulf United Nuclear Corporation, and General Electric Company representatives.
During the course of Gulf United Nuclear's inspection of sound bundles, UNO30 and UN064 fuel assemblies suffered spacer damage. Upon further inspection of these damaged assemblies, UN064 was rechanneled and placed in the reactor, but UN030 was damaged beyond repair and returned to storage.
b.
Metal Surveillance During the month of October, seven metal surveillance con-tainers from the reactor core periphery, three specimen l
containers from the reactor thermal shield, and seven spe-cimen containers from the turning vane were removed to the fuel building for inspection, photographing, and analysis of individual specimens by General Electric.
The following are the dispositions of the metal surveillance samples:
Core-pressure Vessel Removed from Returned to Item Description Core Position Core Position 1
Dummy Fuel Assy No.4 67-25 67-25 2
Dummy Fuel Assy No.5 58-25 58-25 3
Dummy Fuel Assy No.6 (1) 58-02 58-02 4
Dummy Fuel Assy No.7 67-02 67-02 5
Dummy Fuel Assy No.3 51-18 51-18 6
Dummy Fuel Assy No.9 74-18 74-18 7
Long Rack No. 23 (2) 75-13 Wall-pressure Vessel Removed from Returned to Item Description Thermal Shield Thermal Shield 1
Base Metal No.3(2)
@ 154 Core Position
% 51-07 2
Weld & Haz No.4-2(2)
@ 3340 Core Position
% 75-20 3
Part No. 5(3)
@ 26
@26 Core Position Core Position
^375-08
$ 75-08
Turning Vane Removed from Returned to Item Description T.V.
(old) position T. V.
(new) positten 1
Saddle Bag No.2 Rsp 168-169 Saddle Bag Holder @ NW Position of T.V.
2 Saddle Bag No.3 Rsp 177-178 Saddle Bag Holder @ NW Position of T.V.
3 Saddle Bag No.4 Rsp 162 Saddle Bag Holder @ NW Position of T. %
4 Pigeon Ladder (4) 163S 5
Pigion Ladder (4) 1739
()
6 VBWR S.B.
172E 7
T. V. Hanger (2) 175N (1) The old dummy fuel assy No. 6 was scraped - its samples were removed; some will be returned to vallecitos, the remainder being installed into dummy fuel assy No. 5.
A new dummy fuel assy #6 with new samples was inserted into the reactor.
(2) These parts will not be returned to the reactor, their disposition as to when they will be tested will be decided in March.
(3) This part was originally assumed to be part No.
2"., but upon further investigation revealed that the sample was part No. 5 and returned to the reactor. Part No. 22 had been removed during the 1967 oatage and is currently in the fuel pool.
(4) Some samples were removed and transferred to saddle bag No. 2; the remaining samples hava the same disposition as (2).
. i 1.
_New specimens inserted Two racks of stress corrosion capsules with new bourdon I
tube attachment were inserted.
'; e rack was placed in saddle bag No. 3 and one in saua z No. 4.
There are a i
total of 35 capsules of this type and consist of inconel -
600 and 347 stainless steel.
Four racks of stress corrosion capsules of the older type without the bourdon tubes were placed in saddle bag No. 3.
f There are a total of 51 of these capsules which also con-sist of inconel - 600 and 347 stainless steel.
One rack contining a variety of U-bends, tensile corrosion and ring corrosion specimens were placed in dummy fuel assy No. 7.
These specimens consist of Zircaloys.
The following metal surveillance containers were not re-moved during the outage:
- 1) Part No. 4-1 @ 32 on thermal shield
- 2) Part No. 2 @ 2060 en thermal shield
- 3) Part No. 10 R (G4), R (G5), and R (09) right side of steam drum.
As a result, the total quantity of metal specimen containers in the reactor are as follows:
Six specimen containers in the core periphery, Three specimen containers on the thermal shield, Three specimen containers on the turning vane, and One specimen container in the steam drum, c.
Emergency Condenser Internal Inspection On December 20, 1971, an internal visual inspection of the Unit i emergency condenser was made, l
All paint on all surfaces of the condenser was inspected and found to be in excellent condition.
No evidence of paint l
deterioration was observed.
The tubes were also visually inspected, and no abnormal con-ditions were noted.
d.
Reactor Vessel Inspection On November 13, a visual inspection was made of the reactor vessel internals with all fuel assemblies and control rods removed from the core.
A
. Typical areas observed during the inspection included; bottom core support grid, upper core grid hold-down lugs, upper core grid support ring, two riser nozzles, and two patches of vessel cladding. From the conditions observed, no problems exist with rhe integrity of the vessel internals.
e.
Primary System Weld Inspection
- 1) Reactor Flange Inspection The reactor vessel to flange weld, reactor vessel head to flange weld, studs, nuts, and ligaments were ultra-sonically inspected by Pittsburgh Testing Laboratories on November 13, 1971. The IIW2 block and G.E.
specifi-cations were used for instrument calibration and indica-tion criteria. Testing was performed under the super-vision of Jim Ford of G.E. and Wesley Witt of the company's Operational Analysis Department. The tests were witnessed by Mr. Dale Anderson of Travelers Insurance Company.
The ultrasonic inspection of the reactor vessel to flange weld was performed on 180 at the inside diameter on the top face. Twenty-eight ligaments were also tested on the top face between vessel stud holes. The entire reactor vessel head flange weld and all ligaments were inspected on the top face. Twenty-nine studs and 28 nuts were also inspected. No indication of defects were found.
- 2) Reactor Thimble Welds No reactor thimble welds were ultrasonically examined during the 1971 inservice inspection.
The normal makeup water for the Dresden 1 plant allows an orderly shutdown and cooldown without loss of reactor coolant for water line breaks up to 2.026 square inches (nominal pipe size of 1.6 inches inside diameter) and for a steam line break up to 8.1 square inches (3.2 inches I. D. ).
A failure of a CRD housing weld would result in a 1.1 square i
inch break and therefore is included in the exclusion criteria I
of IS-121,Section XI of the ASME B & PV Code.
- 3) Primary Piping a) Examinations Th? 1971 In-Service Inspection program has been consid-eretly augmented over those conducted in the past.
This year the in-service program has been based on the " Interim Acceptance Criteria for Emergency Core Cooling Systems 4
r
for Light-Water Power Reactors", dated June 19, 1971, Items IV.C.1 (b) 3 and 4 which states "An augmented in-service inspection program shall be inaugurated coverfig those portions of the system piping, pumps, and valves with a nominal diameter of four inches or greater, and for whose postulated failure, the per-formance of the installed emergency core cooling system would not be in compliance with the criteria.
The augmented program shall be based on the ASNE B &
PV Code,Section XI, except that the frequency of inspection of piping, pumps, and valves shall be tripled.
The 1971 primary system piping weld inspection on Dresden Unit #1 was completed during the seventh partial refueling, maintensnee, and inspection out-age.
The inspection uttilzed ultrasonic and dye penetrant techniques on the reactor vessel and steam drum bi-metallic welds.
119 circumferential piping welds were inspected. Wherever possible, one foot of the longitudinal weld from the intersection with the circumferential weld selected for the examination was also inspected. Seven indications were noted. Six of the indications were due to crown-of-the-weld bead or misma tch. The other indication was a scratch on the surface of a decontamination flange face. No repairs were necessary. Welds inspected in the reactor annulus included two 10" (capped) instrument nozzles, one 10" unloading heat exchanger nozzle, three 16" steam riser nozzles, and one 22" recirculation inlet nozzle, b) Hydrostatic Test A hydrosta tic test of the vessel had not been conducted by the end of the year, f.
H. P. Turbine Inspection An overhaul of the high pressure turbine and a;.sociated com-ponents was performed during the 7th partial refueling outage.
Radiation levels encountered were considerably higher than those found during the 1967 inspection. The inspection indi-cated that the H.P.
turbine was in satisfactory condition. No major repnirs were necessary at this time.
13.
Feedwater Heater Replacement "E" primary feedwater heater tube ;.undle was replaced during the refueling outage. The original beadle was a Foster Wheeler bundle with 556-7/8 inch monel "U" tubes The new tube bundle was manu-factured a,
authwestern and consists of 809-5/4 inch 304L stain-less sts.i "U" tubes.
-l
- J "D" primary feedwater heater tube bundle was replaced during the refueling outage. The original bundle was a Foster Wheeler bundle with 590-5/8 inch, 263-7/8 inch monel "U" tubes. The new tube bundle was manufactured by Southwestern and consists of 800-3/4 inch 304L stainless steel "U" tubes.
"A" drain cooler tube bundle was replaced during the refueling outage. The original bundle was a Foster Wheeler bundle with 992-5/8 inch 707 copper 30% nickel alloy "U" tubes. The new tube bundle was manufactured by Southwestern and consists uf 992-5/8 inch 304L stainless steel "U" tubes.
Installation of the three heater tube bundles was not complete l
by the end of the year.
14.
Piping Failures and Repairs 1
In November of last year, a crack was located in the 4" decestar-ination stub tube on "A" secondary steam generator loop. The loop was taken out of service at that time.
Repairs were rot completed i
until March of this year. On June 23 af ter repairs to the recircu-lation pump thrust bearing plate, motor installation, and discharge i
valve MO-ll3 the loop was returned to service.
- 15. Tests 3
2 a.
Sphere Integrity Test Program
)
As part of the sphere integrity test program, the sphere venti-lation valves, the 16' equipment hatch, and the sphere access
)
air locks were leak tested satisfactorily during the year.
i b.
Primary Steam Drum Safetv Valves Five (5) spare primary steam drum safety valves were installed during September 1971, replacing five of the safety valves pre-viously in service. A sixth safety valve (SS97t 9) was removed, tested, and replaced on the drum in the same position. The valves installed were cleaned, set and leak-checked in the shop facility.
i c.
Temperature Coefficient of Reactivity Check The moderator temperature coefficient was checked on August 15, 1971. It was found by extrapolating the curve obtained tha t the AEC limit of $1.00 reactivity insertion wou'd be approached at the end of a Cycle 7 exposure of 295,000 MWD. - 'e temperature t
coefficient is plotted versus moderator temperature n Figure 3.
d.
Fuel Sipping Fuel sipping in the reactor s ssel commenced September 17, 1971.
All Cycle 7 fuel were ' sipped in-core.
One hundred three (103)
19-fuel assemblies were sipped out-of-core in the Unit 1 fuel transfer pool.
Inspection of the data obtained from the sipping program indicated that 35 fuel assemblies were defective, e.
Control Blade Depletion Test After completion of 11 core sipping (September 30, 1971), a control blade depletion test was performed to compare actual blade depletion to calculated depletion. To perform this test, a new blade (B4) and the highest exposure blade (B76)
~
were placed in position F-6 under similar controlled conditions.
By means of a central core critical, these blades were cali-brated for notch worth. Analysis revealed that actual b1:ce worth depletion was much smaller than calculated. This indi-estes that there is no nuclear need to order new blades, the probable reason for blade replacement would come as a result of mechanical limits.
f.
Shutdown Margin Checks On December 2 and 3, 1971, reactor shutdown margins were con-ducted to demonstrate that the refueled Cycle 8 core met license DPR-2 requirements with regard to " stuck rod" criteria and that the margin is in excess of one percent throughout the core. Th1 core was shutdown in excess of 0.93% for 3 rods out on the peri-phery and in excess of 0.87% in the central region for 3 rods out.
g.
Initial Critical - Cycle 8 On December 3, 1971, the reactor was pulled critical on sequence SC to verify the worth of the cycle 8 loading. With a moderator temperature of 68 F the reactor went critical on 15 rods 7 notches. (See Figure 4).
B.
License DPR-2 Table 7 lists the amendments to our license requested and/or authorized during the year. Pertinent correspondence pertaining to these requests are listed in the Correspondence References.
-w a
. r CORRESPONDENCE REFERENCE 1971 (1) Letter to AEC, dated January 18, 1971, applying for an amendment i
to DPR-2 on data pertaining to airborne waste.
(Change No. 21)
(2) Letter to AEC, dated March 24, 1971, applying for an amendment to DPR-2 to permit Isading of up to 120 Type VIII fuel assemblies.
(Change No. 22)
(3) Letter from AEC, dated April 5,1971 authorizing Change No. 22 d
to the Technical Specifications of Facility License No. DPR-2.
(4) Letter to AEC, dated June 1,1971, applying for an amendment to DPR-2 in the use of respiratory protective equipment.(Change No. 23)
(5) Letter from AEC, dated September 8, 1971 authorizing Change No. 23 to the Technical Specifications of Facility License DPR-2.
f i
E I
1 e
.e
4 APPENDIX A 4
4
TABLE I OPERATING PERFORMANCE - 1971 Out ge Off System On System Outage Period No.
Date Time Date Time HR.
MIN.
Reason For Outage 121 1/14/71 1715 1/16/71 0147 32:32 Forced - Reactor (19 hrs. 56 min.) Scram resulted from failure of an " Actuator Amp-lifier" in the incore monitoring system.
122 2/3/71 0625 2/11/71 0918 194:53 Forced - Reactor (183 hrs. 45 min.) Shutdown due to a steam leak in the "A" cleanup line.
123 4/8/71 2155 4/16/71 0831 178:36 Scheduled - Reactor (150 hrs. 53 min.) Shut-down for CRD friction and scram testing, repair of feedwater heater leaks, steam line drain leak, and other maintenance.
124 4/17/71 1534 4/18/71 0740 16:06 Forced - Reactor (5 hrs. 31 min.) Scram due to high incare f. lux.
A spurious signal occurred when contractors broke a static wire on line 1210 in 138 KV yard, 125 4/30/71 2257 5/2/71 0707 32:10 Forced - Turbine (32 hrs. 10 min.). Reactor shutdown due to steam leak in the extraction piping.
126 5/2/71 1050 5/2/71 1646 5:56 Forced - Reactor (1 hr. 19 min.) Scram due to operator error - closed B & C Feedwater heater Bypass valve instead of C Drain cooler bypass valve.
127 8/12/71 1244 8/16/71 0318 86:34 Forced - Reactor.(63 hr. 50 min.) Shutdown due to a rapid ov erprassurization in the off-gas holdup lia.
TABLE I (con' t.)
OPERATIPC PERFORMANCE - 1971 Outogn Off System On System Outage Period No.
Date Time Date Time H R.
MIN.
Reason For Outage 128 9/10/71 2125 2690:35 Scheduled - Plant (2690 hrs. 35 min.)
Seventh partial refueling H.P. Turbine Inspection, Core Spray installation, replacement of three feedwater tube bundles, primary system piping inspection, installation of new turning vane, instal-lation of new vessel level sensors, and replacement of five control rod drives.
4 Total Outage Time 3237 hrs. 22 min.
i
E TABLE 2 LOAD RESTRICTIONS FOR 1971 Reduction From Maximum Date
__ Capability of 210 MW _
Condition January 1 - January 14 70 Minimize Off-Cas Activity, feed-water heater outage.
January 16 - February 3 70 Minimize Off-Cas Activity, feed-water heater outage.
February 11
'Ascch 25 70 Minimize off-Cas Activity, feed-water heater outage.
March 26 - March 29 110 Steam leaks.
March 30 70 Minimize Off-Cas Activity, feed-water heater outage.
March 31 - April 8 ilu To limit buroop of fuel to allow extension of cycle VII through the summer.
April 16 - April 17 110 Fuel conservation.
April 18 - April 30 110 Fuel conservation.
May 2 - June 7 110 Fuel conservation.
June 8 - August 4 90 Re-evaluation of core reactivity and a scheduled refueling outage in September.
August 5 - August 12 96 Decay from gradual fuel depletion.
August 16 95 Fuel conservation.
August 17 - September 10 110 Fuel conservation.
1 l
)
Table 3 1971 Control Rod Drive Overhaul and Inspection Summary Core SN Drive SN Drive Position Removed Installed Abnormal Symptoms of Drive Removed
_ Inspection Results B-2 1307 1229 a) Increased Friction Pressure Inspection will be performed b) Long Insert Time at a later date.
E-6 1233 1291 a) Short Buffer Time Inspection will Se performed a r. a latet - ta te.
F-6 1280 1253 a) Short Buffer Time Inspection will be performe/.
at a later date.
J-2 1251 1245 a) Long Insert Time Inspection will be performed at a later date.
D-10 1245 1235 a) Long Insert Time Main piston outer seals were stuck and severely scratched; stro piston seals were stuck at one broken.
Table 4 Solid Radioactive Uaste Shipments From Unit 1 Radwaste Date Destina tion
_ Content Activity (curies)_
Volume (ft3) 4/27/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 1.28 x 10~1 1215.0 4/28/71 Nuclear Engineering Co.
Sheffield, 11' aois Dry Active Waste 9.20 x 10-2 1552.0 4/29/71 Nuclear Engineering Co.
~
~1 Sheffield, Illinois Dry Active Waste 1.10 x 10 1584.0 4/30/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 1.00 x 10-2 340.4 5/01/71 Nuclea" Engineering Co.
Sheffie3.d, Illinois Dry Active Waste 3.20 x 10-2 985.5 5/27/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 8.30 x 10-2 1758.0 5/28/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 6.00 x 10-2 1122.5 6/sl/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 1.41 x 10-1 1029.7 6/03/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 5.30 x 10-2 207.0 9/13/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 1.0^ x 10~1 1545.0 9/14/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 4.11 x 10-1 1464.5 9/30/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 8.20 x 10-2 207.0 i
10/01/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 2.00 x 10~3 496.0 11/02/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 2.00 x 10-3 496.0 11/12/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 2.54 x 10~1 1308.0
Table 4 (coni)
Solid Radioactive Waste Shipments From Unit 1 Radwaste 3
Date Destination Content Activity (curies) folume (ft )
11/13/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 2.26 x 10-1 1365.0 12/09/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Active Waste 9.80 x 10-2 192.0 12/31/71 Nuclear Engineering Co.
Sheffield, Illinois Dry Accive Waste 1.92 207.0 m.
Table 5 SPENT FUEL SHIPMEffr
SUMMARY
Number of Assemblies or Containers Total Shipment Number Date Batch To Rail Truck Shipped 1
2 3
4 5
Rail Truck Date 1
6/11/65 24 0
0 0
0 24 24 2
6/30/65 24 0
0 0
0 24 48 3
7/16/65 24 0
0 0
0 24 72 4
8/ 3/65 24 0
0 0
0 24 96 5
8/16/65 24 0
0 0
0 24 130 6
9/ 2/65 24 0
0 0
0 24 154 7
9/22/65 24 0
0 0
0 24 168 1
8/ 1/66 0
0 4
0 0
4 172 8
8/ 5/66 16 0
6 0
0 22 194 2
8/15/66 0
0 4
0 0
4 198 3
8/24/66 0
0 4
0 0
4 202 4
8/28/66 0
0 4
0 0
4 206 9
8/31/66 0
0 12 8
0 20 226 5
9/ 5/66 0
0 4
0 0
4 230 6
9/12/66 0
0 4
0 0
4 234 7
9/14/66 0
0 4
0 0
4 238 10 9/16/66 0
0 0
20 0
20 L
258 8
9/1S/66 0
0 4
0 0
4 262 9
9/21/66 0
0 4
0 0
4 266 10 9/25/66 0
0 4
0 0
4 270 11 9/26/66 0
0 4
0 0
4 274 12 9/28/66 0
0 4
0 0
4 278 13 10/ 2/66 0
0 4
0 0
4 282 14 10/ 3/66 0
0 4
0 0
4 286 11 10/ 7/66 0
0 0
24 0
24 310 15 10/11/66 0
0 4
0 0
4 314 16 10/12/66 0
0 4
0 0
4 318 17 10/20/66 0
0 4
0 0
4 322 18 10/23/66 0
0 4
0 0
4 326 19 10/26/66 0
0 4
0 0
4 330 12 10/28/66 0
0 3
16 0
19 349 20 10/30/66 0
0 4
0 0
4 353 21 11/ 1/66 0
0 4
0 0
4 357 22 11/ 6/66 0
0 4
0 0
4 361 23 11/ 8/66 0
0 4
0 0
4 265 24 11/10/66 0
0 4
0 0
4 369 25 11/13/66 0
0 4
0 0
4 373 13 11/14/66 0
0 0
23 0
23 396 26 11/15/66 0
0 4
0 0
4 400 27 11/17/66 0
0 4
0 0
4 404 28 11/20/66 0
0 4
0 0
4 408 29 11/27/66 0
0 4
0 0
4 412 l
30 11/29/66 0
0 4
0 0
4 416
Table 5 (cont.)
SPENT FUEL SHIPMENT
SUMMARY
Number of Assemblies or Containers Total Shipment Number Date Batch To Rail Truck Shipped 2
3 4
5 6
7 Rail Truck Date 14 12/ 2/66 4
19 0
0 0
23 439 31 12/ 6/66 4
0 0
0 0
4 443 32 12/11/66 4
0 0
0 0
4 447 33 12/15/66 4
0 0
0 0
4 451 34 12/18/66 4
0 0
0 0
4 455 35 12/20/66 4
0 0
0 0
4 459 36 12/27/66 4
0 0
0 0
4 463 37 1/ 3/67 4
0 0
0 0
4 467 38 1/ 5/67 4
0 0
0 0
4 471 39 1/ 8/67 4
0 0
0 0
0 475 40 1/10/67 4
0 0
0 0
4 479 41 1/15/67 4
0 0
0 0
4 483 42 1/22/67 1
0 0
0 0
1 484 15 7/15/68 0
0 0
24 0
24 508 16 8/ 9/68 0
0 0
24 0
24 532 17 8/29/68 0
0 0
21 0
21 553 18 9/19/68 0
0 0
21 0
21 574 19 10/ 9/68 0
0 0
24 0
24 598 20 10/30/68 0
0 0
24 0
24 622 21 11/20/68 0
0 0
24 0
24 646 22 12/11/68 0
0 0
22 0
22 668 23 1/ 3/69 0
0 0
19 0
19 687 24 1/29/69 0
0 24 0
0 24 711 25 2/26/69 8
0 0
3 0
0 11 722
TABLE 6 i
FUEL RODS SHIPPED TO G. E.
FOR ANALYSIS ON SEPTEMBER 2, 1971 Assembly No.
Number of Rods DU-81 1
DU-93 2
G-11 1
G-20 2
G-28 1
G-42 3*
i G-73 3
G-79 3
XE-45 1
- One of these rods was a Gd 023doisonrod.
1 F
o TABLE 7
_SUPNARY OF LICENSE AMENDMENTS PENDING DURING 1971 Date Date Requested Authorized Request to amend License DPR-2 Still on data pertaining to airborne Pending at waste (Change No. 21) 1/18/71 end of 1971.
Request to amend License DPR-2 to permit loading of up to 120 Type VIII fuel assemblies.
(Change No. 22) 3/24/71 4/05/71 Request to amend License DPR-2 to establish compatibility between License DPR-2 and 10CFR20 in the use of respiratory protec-tive equipment. (Change No. 23) 6/01/71 9/08/71 l
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