ML20030A505
| ML20030A505 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/13/1971 |
| From: | Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090787 | |
| Download: ML20030A505 (20) | |
Text
.
Regulatory Fib Cy.
D3'h NW wu: n*H CONSUMERS POWER COMPANY 4
Docket No 50-155 Report of Operation of Big Rock Point Nuclear Plant License No DPR-6 January 1, 1971 Through June 30, 1971 I.
SUMMARY
OF OPERATION A.
Power Operation Due to premature failure of several "E" fuel bundles (see twelfth semiannual report), the Plant was operated at 53 MWe (g) during this report period'to limit the heat flux on the fuel clad-ding, thereby possibly reducing fuel cladding deterioration until modifications can be made to the Plant to reduce crud deposit levels on cladding surfaces.
On January 1,1971, the Plant was on the line at 53 MWe (g) with an off-gas release rate of approximately bl,000 pCi/sec. The Plant operated at 53 MWe (g) until January 23 when the unit was re-moved from service on a scheduled outage to repair turbine condenser tube leaks and the steam leak in the intermediate-pressure heater line.
'During the testing of the containment isolation valves (CV klO7), prior to returning the Plant to service, the main steam drain (outside) isolation valve failed to close. This failure to close was traced to the solenoid operator (SV h317). The defective solenoid valve was replaced and the control valve operated normally. The cause of misoperation was quite apparent upon disassembly of the unit.
4 Interior components were covered with corrosion, indicatire i
thr.t moisture had been present in the valve internals. This source of noisture was determined to be an improperly operating instrument air dryer. The instrument air dryer unit was replaced during the February-March refueling outage. In addition, four similar valves were replaced during the refueling outage so that the original valves could be in-spected. There were no signs of residue or moisture or excess wear.
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1 The Plant was returned to service after 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of outage on January 24,19T1. ~ Af ter reaching 53 MWe (g), the off gas stabilized at approximately 46,000 pCi/sec.
On February 3, it was no longer porsible to maintain accept-able seal temperatures on the cartridge seal for the No 2 recirculating pump. The-pump was shut down and isolated and power level reduce. to 48 MWe. Deterioration of this seal had been noted as early as July 1970; however, the pump continued to operate satisfactorily until February 3, 1971. With the decrease in power level, the off-gas release rate was dropped to approximately 18,000 UCi/sec.
The Plant operated with the No 2 recirculating pump isolated until February 12 when it was shut down on a scheduled outage for the eighth refueling as well as the installation of the redundant core spray cystem. On February 13, three candidates for AEC reactor licen-ses mac ? critical approaches and brought the reactor power to approxi-mately
.5 MW. On March 13, 1971, the Plant was returned to service t
following a 28-day, 8-hour and 40-minute outage at a power level of 53 MWe (g) (see Section I.B. and IV. for discussion of refueling activi-f ties). During 'the start-up, the minimem flow valves on the feed-water pump discharge operated erratically. Tnese had been two of the four solenoid valves that were replaced during the refueling outage as a i
portion of the investigation into the failure of the main steam drain (outside) isolation valve. The replacement kits were slightly modified kits supplied by the vendor.
Sh_s valve is used many places throughout the Plant, such as solenoid operators, for containment isolation valves.
These ' failures are being investigated with the vendor as well as other possible courses of alternate action to be taken to increase reliability of the valves in this type service.
The Plant continued to operate at 53 MWe (g) with the off-gas release rate of approximately 1,500 pCi/sec, until April 20,19T1 when it was removed from service on a scheduled outage to repair a steam leak from the packing of the butterfly valve located on the discharge
- side of the No 1 reactor recirculating pump. Then on April 20, 1971, the Plant was returned to service at 53 MWe (g) after a 23-1/2-hour
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3 The Plant was again removed from service on April 29, 1971 to make adjustments to the turbine initial regulator. These adjustments improved turbine control in the lower loading range (approximately 20 MWe). The unit was returned to service at 53 MWe (g) on April 30. The duration of the outage was 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 55 minutes.
On May 12, the members of the Company-wide union (Utility Work'ers of America) went out' on strike and remained on strike during the remainder of the reporting period at Big Rock Point. The union-represented personnel are the control operators, auxiliary operators, maintenance and janitcrs, totalling 29 men. Operation of the Plant has been continued utilizing supervisory personnel, engineers and techni-cians.
The Plant continued to operate at' 53 MWe (g) at an off-gas release rate of approximately 2,300 pCi/sec until May 12,
'^7'. when the reactor scrammed by high neutron flux on Channels 1, E.v:9 3 (During the refueling outage, the picoammeters were reset at c00%
corresponding to 200 MW due to plans to limit power level to 53 MWe (g) t because of fuel clad crudding.) The scram was caused by a load rejection
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due to a fault of the 139 kV transmission line. Subsequent inspection j
of the Plant revealed no abnormalities and the Plant was returned to service on May 13, 1971 following a 21-hour outage. The power line failure was traced to a corner strain pole which had fallen into the adjacent two phases of the power line. The failure of the pole was attributed to being cut at the base approximately half way through the thickness of the pole. In addition, the guy line to the pole had been cut. At the time of this load rejection, the Plant was being manned by supervisory personnel.
The outage was longer than a normal load rejection due to the necessity to reconstruct the corner strain pole. As soon as this was completed, and tested, the unit was returned to service. All actions taken by supervisory personnel during this abnormal operating occurrence were.both prompt and proper. The automatic plant station power trans-fer to the h6 kV line was normal. The emergency diesel generator op-erated in a satisfactory manner but was not used.
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h On May 26, during routine exercising of the control rod drive, Drive'C-3 would not withdraw after being inserted one notch.
Subse-quent attempts.to withdraw the drive were halted with the drive out of position at Notch 17 (C-3 is normally fully withdrawn). Operations have continued at 53 MWe (g) with C-3 at Notch 17 For further details, please refer to the letter from Mr. R. L. Haueter to Dr. P. A. Morris, dated June 7, 1971.
The Plant operated at 53 MWe (g) until June 2,1971. At this time, the unit was removed from service on a scheduled outage to re-pair a steam leak in the turbine stage drain piping to the high-pressure heater. During this outage, a defective solenoid operator was found on the dirty sump (outside) isolation valve. This solenoid valve, SV-h896, was also one of the four solenoid valves that had been replaced during the refueling outage as a portion of the investigation into the failurs of the main steam drain outside isolation valve. As stated before, an investigation is being conducted to determine the cause of failure of this type valve. The unit was returned to-service after 2h hours and h2 minutes on June 3,1971 and continued to operate at 53 MWe (s) until the end of the reporting period. At the end of the period, the off gas had increased gradually to approximately 2,700 PCi/sec.
B.
Refueling Outage The new loading after completing the seventh refueling outage consisted of the following:
1.
Reload "B" Bundles - 2 2.
Reload "C" Bundles - 4 3
Reload "E" Bundles - 52 (Includes 3 EEI-Pu Bundles, k MEG Bundles and 12 PEG Bundles) 4.
Reload "F" Bundles - 24 5
Reload "D" Bundles - 2 (Jersey Nuclear Corporation)
Fuel pin profilometry work conducted during and subsequent to the refueling outage revealed that although fuel crudding had de-creased, it did not decrease as much as would be expected if a source of crudding was predominntely a hideout source. However, these results t-
t 5
are still inconclusive because samples of the crud layer deposited on 1
fuel rods during the last run have not yet been chemically analyzed.
The continuing water chemistry program indicates that the tube leakage in the regenerative heat exchanger has increased over the past year allowing increased bypass flow of the clean-up system demineralizer.
These results, coupled with the fuel profilometry results, have indi-cated procurement of replacement heat exchangers. The new heat ex-changers are to be identical to the presently installed heat exchangers with the exception of the tube material. The new heat exchangers vill utilize stainless. steel tubing where the old utilized 70-30 Cu-Ni 4
tubing material.
Delivery for the new heat exchangers is scheduled prior to the 1972 refueling outage. Currently, plans include limiting power level to 53 MW until at least these heat exchangers are replaced.
Data taken on fuel ascembly extraction forces, during the core unloading phase of the refueling outage, indicate that it will be necessary to replace approximately 38 fuel channel support tube assem-blies during the next several years due to channel warpage. It is currently planned to procure for delivery prior to the 1972 refueling outage 42 of these assemblies with a redesigned orifice and support tube to channel transition. This redesign is based upon flow testing that showed slightly diminishef flow in the outer rods of the lower tier of a fuel assembly. This lata is in agreement with profilometry data which indicates preferential crudding of the outer rods in the lower tier of the fuel assemblies. A " Request for Change of the Technical Specifications" will be submitted when the flow testing of the redesign has been completed. It is felt that this change, together with the elimination of the copper source in the clean-up system heat exchangers, will significantly alter the fuel rod crud levels experienced over the past five years.
on March 2, 1971 during control rod drive system check-outs 1
prior to fuel loading, Drive B-6 could not be withdrawn from its fully inserted position. This problem was reported to Dr. P. A. Morris by i
letter dated March 26, 1971 from Mr. R. L. Haueter.
Since the
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6 writing of that letter,. roller and pin have been removed from a. control i
rod blade removed from the reactor during the 1971 refueling outage.
The roller and pin were examined and showed no signs of excessive wear.
However, the pin was slightly bent, probably during removal, and the roller was not free to rotate. It is planned to visually inspect the rollers on the 16 outer periphery blades during the next refueling outage in an attempt to locate where the roller came from that caused the difficulty in withdrawing Control Rod Drive B-6.
C.
Statistics The reactor was brought critical eight times during the report period. The reactor was critical for 3,597 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> with electries' gen-eration of 182,057 MWh (g) or 173,119 3 MW net.
The thermal output of the reactor was 563,836 MWh.
II.
ROUTINE RELEASES, DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIALS A.
Gaseous Releases - The gaseous radioactivity released to the environs from the stack during power operation is based upon turbine generating hours and is summarized below:
Month Curies Released January 104,000 February 37,200 March 2,2 60 April h,530 May 6,950 June 8,660 B.
Liquid Discharge - During this report period, the liquid radio-activity released to Lake Michigan by way of the circulating water dis-charge canal numbered 20 batches with total activity of 2 97 curies. All batches were identified for isotopic composition which showed that approx-imately 21% was Zn-65, 27 2% was Cs-134 and Cs-137 and 27 7% was I-131.
The remaining portion was composed of Co-58, Co-60, Mn-54, Fe-59, Ce-144, Cr-51, Zr Nb-95, Ce-lh1 and Ru-103 i
7 Batches Total Month Released Gallons mci January 3
15,090 1,313 0 February 6
27,325 707.4 March 8
39,375 850 3 April 2
10,280 64.0 May 0
0 0
June 1 __
4,365 36.6 Total 20 96,435 2,971 3 C.
Shipments - A total of 16 off-site shipments of radioactive material were made during this reporting period:
Ship-ment Transfer Transfer No Date From Tb Radiouctive Material 221 1/22/71 DPR-6 B&W In-Core Wire -
SNM-778 50 Ci Lynchburg, Va 222 1/26/71 DPR-6 GE-Val Feed-Water Crud and Filtrate -
0017-60 0.2 mci Calif 223 1/27/T1 DPR-6 NECO 2400 Gal of Low-Level Waste -
16-NSF-1(A-ll) 1.0 Ci Fbrehead, Ky 224 1/29/71 DPR-6 NECO 2400 Gal of Low-Level Waste -
16-NSF-1(A-ll) 1.0 Ci More!.ead, Ky 225 3/1/71 DPR-6 GE-Val Fuel Inspection Tools -
0017-60 1 9 mci Calif 226 3/17/71 DPR-6 NPI Irradiated Cobalt - 479,230 Ci 19-12667-01 DOT, SP 5364 227 3/23/71 DPR-6 NPI Irradiated Cobalt - 507,360 Ci 19-12667-01 DOT, SP 5364 228 3/24/71 DPR-6 GE-Val Dummy Fuel Rods - 100 Ci 0317-60 DOT, SP 5971 229 3/24/71 DPR-6 GE-Val Centermelt Fuel Rod - 1000 Ci 0017-60 DOT, SP 5607 230 3/26/71 DPR-6 NPI Irradiated Cobalt - 267,600 Ci I
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8 Ship-y ment' Transfer Transfer No Date From To Radioactive Material 231 3/29/Tl DPR-6 NECO 115 Gal Barrels Iow-Level 16-NSF-1( A-11)
Waste - 0 3 Ci Morehead, Ky 232 h/9/71 DPR-6 BI Pump Recirculating Pump Parts -
0743-74 0.2 mci Los Angeles 233 h/lk/71 DPR-6 GE-Val Feed-Water Crud and Filtrate -
0017-60 0.2 mci Calif 23h 4/27/71 DFR-6 GE-Val Fuel Inspection Tools -
0017-60 0.2 mci Calif 235 5/28/71 DPR-6 GE-Val Feed-Water Crud and Filtrate -
0017-60 0.2 mci Cali f 236-6/ 7/71 DPR-6 Palisades Steam Drum Crud Samples -
Nuclear Plant 3 mci 21-08606-05 Mich III. RADIOACTIVITY LEVELS IN PRINCIPLE FLUID SYSTD4S A.
Primary Coolant ~
Minimum Averare Maximum Reactor Water Filtrate (a)
-1
-1
-1 pCi/cc 1.31 x 10 2.8 x 10 1.45 x 10
- Reactor Water Crud 2.63 x 10 8.75 x 10
-1
-1 pCi/cc Turbidity 1.16 x 10 Iodine Activity (b)
-2
-2
-1 pC1/cc 5 x 10 9 x 10 2 x 10 I
(a)A counter efficiency based on a gamma energy of 0.662 MeV and one gamma photon per disintegration decay scheme is assumed to convert count rate to l
microcuries. All count rates were taken 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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Minimum Average Maximum Reactor Cooling Water (a) 2 2
-3 pCi/cc 5.8 x 10 2.9 x 10 h.32 x 10 C.
Spent Puel Pool
- Radioactivity in the spent fuel pool is principally activated corrosion' products from the stored spent fuel and core components.
Minimum Average Maximum Puel Storage Pool (a) pCi/cc 1.45 x 10 1.17 x 10 h.35 x 10-2
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-2 Puel Fool Iodine 3
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pCi/cc 2 x 10 8 x 10 6 x 10 IV.
PRINCIPLE MAINTENANCE PERFORMED A.
Reactor During the refueling outage in February, 2 in-core detector assemblies were removed from the reactor vessel and replaced with 2 new in-core detector assemblies and a total of 16 control rod blades were removed from the reactor and replaced with 16 new blades.
A redundant core spray system construction project which was begun last year was completed and the system put in service for reactor l
run No 9 The work was done by Bechtel Corporation under a construction general work order..The system is designed to provide a redondant source of emergency cooling water to the reactor core and the reactor containment building.
The water supply for the system is in parallel with the existing core spray system and can be remote-manually selected from either of the two existing fire system supply headt-7 or from the core spray recirculation t
" A counter efficiency based on a gamma energy of 0.662 MeV and one gamma photon per disintegration decay scheme is assumed to convert count rate to microcuries. All count rates were taken at two hours after sampling.
- b)Based on efficiency of Iodine-131 two hours after sampling.
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10 system. Operation of the core spray system is autome vica?.ly initiated by reactor low-weter level along with low reactor pressure switches.
The post-incident reactor containment building sprays are
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automatically controlled by high building pressure sensor switchee.
All piping, valves and instrumentation have been installed and checked out. The system was flushed and hydrostatically prescure-tested. A complete set of preoperational 1-sts were run prior to putting the system in service.
B.
Primary System Weld Inspection An ultrasonic examination of primary system piping velds was performed by Southwest Research Institute personnel. Tentative results were satisfactory with no defective welds noted.
The inspected welds were as follows :
1.
Thirty 3-inch pipe velas in the clean-up demineralizer system located in the recirculation pump room. Of these welds, lh were previously inspected by General Electric personnel in 1970 and six of these welds were newly installed welds on the clean-up system thermal tee assembly; 2.
Eight 20-inch velds in the main recirculation Loops 1 and 2 lo sted in the recirculation pump room. Four of these welds were previously inspected by General Electric personnel in 1970; 3
seven 6-inch welds including one 6-inch tee in the shut-down cooling system located in the recirculation pump room. The 6-inch tee and the velds connected to it were previously inspected by General Electric personnel in 1970; 4.
Twelve h-inch velds in the newly installed section of the backup core spray line located on the reactor deck; 5
The head-to-flange veld on the reactor vessel closure head; 6.
Six 14-inch riser nozzles and four 17-inch downcomer nozzles at the steam drum located in the steam drum cavity. The area inspected was the " safe-end" to pipe veld and the " safe-ends" on each nozzle. A dye-penetrant insi etion of the " safe-ends" indicated no visual surface defects; and
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Three "J" welds attaching control rod drive housings to-
' stub tubes for Drives B-2, D-6 and F-2.in the control rod drive room.
The. insulation removal and the installation of the snap-on insulation were performed by Consumers Power Company maintenance person-nel. With snap-on insulation installed on all steam drum downcomer and riser nozzles, reinspection of these nozzles should result in con-
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siderably lower personnel exposure in the future.
C.
Cortrol Rod Drive System l-Control Rod Drives B-2, D-6 and F-2 were removed and three spare reconditioned drives installed. The "J" welds on the control rod drive housings were examined as the drives were being changed.
The annual inspection of the drive support structure was 4
conducted during the month of March. Clearance measurements were.made e
and recorded. The support modules are within clearance specifications.
A-2, B-2, D-4, E-5 -and F-5 accumulator bladders and seals were replaced to correct for leakage.
D.
Steam Drum The No 4 position relief valve was removed and replaced with i
the spare and set for 1575 psig.
I Inspection of the steam drum internals was accomplished by removing the manhole cover and grinding off the weld on the stainless steel seal plate. After inspection, the seal plate was welding in po-sition ar.d dye penetrant tested before reinstalling the manhole cover, i
E.
No 2 Recireviation Pump Seal The No 2' recirculation pump seul which had failed and was replaced during the February 1971 refueling cutage was inspected under the supervision of Mr. Ernest Lindros, a Byron-Jackson field engineer.
Excessive crud and foreign material was found on the internal components of the seal assembly. The inner seal was found in good condition except for two breaks on the stationary carbon face - one each on the inner and i
outer radii. The outer seal exhibited moderate wear with erosion marks found about one-half-inch wide on the stationary carbon face. The outer seal split
,s showed excessive burning probably caused 'oy misalignment 2
and during initial start-up. Both U-cup assemblies and the cartridge shaft sleeve exhibited little wear and appeared to be in good condition.
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It is theorized that accumulation of foreign material causing 1-excessive wear on the carbon face was the probable cause for seal failure.
No unusual problems were incurred during the repair and reassembly.
F.
Clean-Up Thermal Tee Assembly Replacement Conflicting ultrasonic test reports from two different testing companies over five-3-inch welds in the clean-up system during the March 1970~ inspection led to the ultimate replacement of these welds in February 1971. The second testing company returnea in September 1970 and, veri-fied the existence of the defects and also confirmed that they had not grown in size. The different results of both companies were resolved to a question of geometry. The first company used a 45 transducer probe while the second company used a 45 and 70 transducer probe and was apparently more successful in "seeing" the roots of the welds. With
. this confirmation of defective velds and no growth in the defects, the repairs were planned for the annual refueling outage in February 1971.
All repairs were made to the standards set forth in USAS B31.7
" Nuclear Pbwer Piping." In preparation for this repair, five Consumers Power Company welders were qualified to the required stainless steel welding procedure and the necessary nuclea grade piping and fittings were ordered.
The repairs consisted of cutting a section of 3-inch piping and 3-inch valve out of the clean-up system in the recirculation pump room. Three cuts were made - at an elbow, tee and valve. This cutout section was then removed from the recirculation pump room to the shop where measurements were taken and a new assembly fabricated. The valve was cut from the assembly and both ends of the valve were veld-repaired and ground in preparation for welding.
The fabricated assembly consisted of a nev 3-inch elbow, spool piece, tee and existing 3-inch valve welded together with three butt welds. The short spool piece contained a thermal sleeve which was welded in place with a fillet weld. Upon completion of each butt weld, radiographs were taken and each veld was considered acceptable.
While the new assembly was being fabricated, the three pipe ends in the recirculation pump room were ground to a 3T-1/2 bevel and prepared for welding. No weld repairs of these pipes were necessary.
4 13 The assenbly was transported to the recirculation pump room p.
where it was fitte1, tacked in place and the final three butt welds made. Radiographs vere taken after the n at - passes for these three 1
velds were completed. When the radiographs proved that each root pass was acceptable, cover passes were laid on all three welds. The welds were ground flush with the piping ar ' final radiographs taken a ad found to be acceptable.
At the completion of the repairs, an ultrasonic inspection also confirmed that all velds were acceptt';'e. Finally, a dye pene-trant inspection of all six butt velds was conducted and all welds were considered acceptable to USAS B31.7 " Nuclear Power Piping."
The removed welds which were not destroyed in the process of cutting out the assembly were cut apart and inspected. Two of the welds were found to have circumferential cracks at the roots approximately 1/8-inch wide. A visual examination of the three-inch stainless steel tee revealed five cracks on the inner surface of the tee. These cracks and subsequent examination results were reported to Dr. F. A. Morris by letters dated June 17, 1971 and August 4, 1971 from Mr. R. L. Haueter.
G.
Core Spray Heat Exchansrer i
The new core spray tube bundle (90/10 Cu Ni B-395) was in -
stalled and leak rate tests were performed at 137 psig. Flow tests were i
run with a 1.1-inch diameter restricting orifice installed on the f 3-charge side, limiting the flow to approximately 210,000 lb/h. This is done safely with the maximum limit of 225,000 lb/hfortheheatex-changer shell side flow.
H.
Penetration Welds The annual inspection of all containment vessel penetration welds as required by the Technical Specifications was conducted. No discrepancies were noted.
j I.
Condensate and Feed-Water System A new Feed-Water Bypass Valve CV kO12 has been received and
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installed. Severe erosion had occurred between the valve seat and the body resulting in no control of flow through the valve. Deterioration of a 2-inch elbow and short section of pipe on the discharge side of
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lb the valve necessitated replacement; also, all socket welds were dye pene-trant checked for cracks and porosity upon completion of the installation.
Reactor Feed-Water Pump Motors No 1 and No 2 were inspected and cleaned of,11 and foreign material. Exceesive wear on the motor bearing shaft seals (labyrinth type) allowed for leakage of oil past the seals and into the motor windings. The seals were cleaned and refitted to the proper cleare.nce specification. The lube oil reservoirs on both pumps were drained and new oil added.
New discharge check valves have been installed ir both feed-water loops. Erosion and deterioration in the valve seats necessitated replacement.
J.
Turbine Generator The emergency governor exerciser reset solenoid valve was replaced during the February refueling outage. The defective valve solenoid coil was found operative, but valve internals showed exces-sive wear causing leakage.
Two Schedule 40, three-inch elbows, a tee and a short nipple on the high-pressure turbine extraction drain to the intermediate-pressure heater were replaced with Schedule 80 materials. Radiographs indicated reduced wall thickness on the outer radii of 75% and greater.
Visible inspection verified the severe erosion. Repairs were also made in the intermediate-pressure extraction drain to the intermediate-pressure heater as four Schedule h0, four-inch elbows and a nipple showed reduction in wall thickness by as much as 40%. Schedule 80 replacements were thought necessary to insure turbine reliability.
Water box manhole gaskets were replaced on the east side of the condenser correcting a leakage problem.
A deteriorated seat and disc on the six-inch hydrogen cooler discharge valve causing excessive leakage necessitated replacement of the valve. The two east end three-inch discharge valve seats on the hydrogen cooling unit were ref,cound and the valves returned to service.
Two four-inch valves were installed in the turbine lube oil cooler lines to isolate the lube oil coolers for maintenance repairs, if necessary.
1
~ 15 Prior to the refueling outage in February, the turbine condenser was inspected for. tube leaks. Two leaking tubes were found and the tubes were plugged. One leaking tube was the ninth tube from the bottom lower left on the south water box. The other -leaker was located on the north water box side, top low, first tube on the north side.
The tube leaks were discovered by maintaining sealing steam on the turbine and running the condenser mechanical vacuum pump which maintained a vacuum of approximately 20" Hg.
Strips of Saran Wrap were then laid over each end of the condenser tubes and the leaks were readily determined by the flexing of the Saran Wrap.
Inspection of the turbine condenser during the refueling outage revealed steam cutting on top rows of the tube bundles. Tb reduce wear on the tubes, stainless steel wire mesh (one-inch squares) was cut, supported and welded above the main condenser tube bundle. The mesh is expected to decrease the size of the water droplets and thus reduce surface erosion on the condenser tubes.
V.
CHANGES. TESTS AND EXPERIMENTS PERFORMED RIRSUANT TO 10 CFR 50.59(a)
A.
Facility Changes C-lh9 - Replacement of Solenoid Valve SV-h900 n Pbison System Recirculation Valve CV-4050 The dual coil solenoid valve controlling CV-h050 was replaced with a single coil solenoid valve of the type used extensively throughout the Plant.
This change was made for two reasons; namely, (1) to enable rapid replacement and/or repair of the solenoid valve should failure occur (repair parts could not be obtained for the original valve), and (2) to provide for closing of the control valve on electrical failure (previously, the control valve would not change position on electrical failure).
C-150 - Clean-Up Demineralizer Tbmperature Switch This change consisted of installing a backup temperature switch and associated circuitry to trip the clean-up pump, close the discharge valve and annunciate through the clean-up demineralizer trouble alarm if the pump inlet temperature exceeds 125 F.
This will provide redundant protection for the clean-up demineralizer high-temperature trip and alarm I
a g
16 switch located in the Plant temperature recorder (which would be in-(
operative should the recorder fail).
This addition will also prevent clean-up system startup snould reactor blowdown through the bypass exceed 125 F as the new sensor is located ahead of the bypass tee.
C-154 - Core Spray Heat Exchanger Cooling Water Flow Orifice A restricting orifice in the shell discharge line of the core sprr.y heat Exchanger was installed to limit the flow to approximately 225,000lb/h.
(See " Principle Maintenance Performed,"Section IV, for information relative to the installation of a new core spray heat ex-changer.) The restricting orifice limits the water flow to 210,000lb/h at 133 psig electric fire pump discharge pressure.
C-155 - Isolation Valve Relocation The clean and dirty sump (inside) isolation valve solenoid Valves SV 4869 and SV-4891 and the poison system recirculation Valve SV h900 were relocated from the recirculation pump rocm to Room hh2 (lo-cated above the control rod drive access room). This change was made to allow servicing of the valves during power operation since thia area is readily accessible and not a radiation area.
The relocation involved the installation of a new air supply line which connected into the instrument air header near the equipment lock and approximately 25 feet of 3/8-inch copper tubing positioned between the solenoid valves and the control valves.
The average closing times of the valves in their new con-figuration are as follows:
CV-4031 (SV 5869) 1.h5 seconds CV 4025 (SV h891) 2 30 Seconds CV4050(SV-h900)
T.h0 Seconds Maximum closing times allowable from receipt of signal:
CV h031 (SV-4869) 6.0 seconds CV h025 (SV-4891) 5.o Seconds CV h050 (SV h900) 15 0 Seconds l
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17 C-156 - Bnergency - Diesel Generatoc Lube Oil Pressure Trip g-An additional low lube oil pressure trip was installed and tested. This trip scheme will provide protection to the diesel engine if the ' lube oil pressure fails to build up during a start of the engine.
The original trip scheme is designed to trip the engine if a lube oil failure occurs while the engine is tunning and needs to be " cocked" by initial pressure buildup. The new scheme has a built-in delay of 15 seconds to allow lube oil pressure buildup after establishing generator output voltage. If the lube oil pressure is not greater than 415 psi (which will provide adequate' bearing protection), the unit will trip.
C-157 - Well Water Pump House Temperature Alarm.
This addition was made to the well water storage tank annun-cintion circuit to provide protection against freezing conditions in the pump house during cold weather.
A temperature alarm was added to monitor the pump house temperature and alarm if the temperature drops below 38 F.
An aux-iliary relay was included in the alarm circuitry to provide fail-safe action, the relay power coming from the pump house heater supply (loss of which will also be annunciated). This addition was brought about by freezing of the instrument reference lines in the pump house during abnormally low tempemtures in the area. Additional heating capacity was added to restore temperature to ntrmal.
4 a
C-160 - Air Compressor Cont = 1 The air compressor control scheme was modified to allow sim-ultaneous loading and unloading of the air compressors. This will allow running two or more compressors in parallel to improve compressor operation by providing a longer cooling cycle.
Previously, with two compressors on, each compressor unloader was controlled by its own pressure switch. If the pressure switch set-tings were slightly different, only one compressor would load while the other would be idle. This would cause excessive wear and high temper-atures to occur on the working compressor, as the air usage is about equal to the capacity of-one air compressor.
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18 With the control modification, the loading and unloading cycles of the compressors in service are approximately equal, resulting in a 4-greatly increased cooling time. The normal mode of operation of the air compressors has been changed to operate two compressors on " hand" control and one in " auto" for backup should the ' air pressure fall be-low 85 psig during peak demand periods.
B.
Tests A temperature coefficient test was conducted in March prior to power operation with the newly loaded core. Test data indicated
[
that the temperature coefficient turned negative at 128 F after adding 9 0/ of. reactivity.
VI.
PERIODIC TESTING PEPFORMED 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 j
be periodically tested per the Technical Cpecifications:
System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives Continuous withdrawal and insertion Each major refueling 3-3-71 of each drive over its stroke with and at least once normal hydraulic system pressure.
every six months during Minimum withdrawal time shall be periods of power 23 seconds.
operation.
Withdrawl. of each drive, stopping Each major refueling 3-3-71 at each locking position to check and at least once every latching and unlatching operations six months during per-and the functioning cf the position iods of power operation.
f indication system.
Scram of each drive from the fully Each major refueling 3-3-71 withdrawn position. Maximum scram and at least once time from syrtem trip to 90% of every six months during insertion shall not exceed 2 5 periods of power seconds.
operation.
Insertion of each drive over its Each major refueling 3-3-71 entire stroke with reduced hy-but not less than once draulic system pressure to deter-a year.
mine that drive friction is normal.
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19 System or Function Frequency of Dates t
Undergoing Test Routine Tests Tested Control Rod Interlocks 1
Rod withdrawal blocked when any Each major refueling 2-26-71
.two accumulators are at a pres--
but not less frequently 3-3-71 sure below 700 psig.
than once every twelve months.
Rod withdrawal. blocked when two of Each major refueling 3-3-71 three power range channels read but not less frequently below 5% on 0-125% scales (or below than onel every twelve 2% on their 0-40% scales) when re-months.
actor power is above the minimum operating range of these channels.
Rod withdrawal blocked when scram Each major refueling 3-3-71 dump tank is bypassed.
but not less than once every twelve months.
Rod withdrawal blocked when mode Each major refueling 3-3-71 selector switch is in shutdown but not less frequently positi on.
than once every twelve months.
Other Liquid poison system component Two months or less.
2-T-71 check.
h-6-71 6-6-71 Post-incident spray system auto-At each major refueling 3-3-71 matic control operation.
shutdown but not less frequently than once a year.
Core spray system trip circuit.
Not less frequently 3-10-71 than once every twelve months.
Emergency condenser trip circuits.
Not less frequently than 3-12-71 once every twelve months.
Containment 1
Containment sphere access air locks Once every six months or 4-10-71 and vent valves, leakage rate.
less.
4-11-71 Isolation valve operability and At least once every 3-h-71 leak tests.
twelve months.
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20 j
l System.or Function Frequency of Dates Undergoing Test Routine Tests Tested Containment (Contd)
Isolation valve controls and
.Approximately quarterly.
2-28-71 instrumentation tests.
3-11-71 5-13-71 6-2-71 Penetration Inspection.
At least once every 3-2-71 twelve months.
- ntegrated leak test.
Once every two years.
3-25-70 I
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 off-gas monitor.
3 stack-gas ronitor calibration.
4.
Emergency condenser vent monitor.
5 Process monitor.
6.
Area monitoring system.
O I
lY W
N Gerald J. @.e Nuclear Fuel Management Administrator Consumers Power Compar.y Jackson, Michigan Ihte: August 13, 1971 Sworn and subscribed to before me this 13th day of August 1971.
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L o,och)
Notary Public, Jackson County, Michi6an My commission expires January 15, 1972 0
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