ML20030A510
| ML20030A510 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/02/1970 |
| From: | Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090803 | |
| Download: ML20030A510 (17) | |
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i CONSUMERS POWER COMPANY Docket No 50-155 REPORT OF OPERATION OF BIG ROCK POINT NUCLFAR PLAUT License No DPR-6 November 1,1969 - June 30,1970 0
I.. SUWARY OF OPERA"'IOUS A.
Power Operation - Due to the premature failure of several E fuel bundles (see 10th semiannual report), the Plant was operated at 53 We (gross) during this period to conserve reactivity and to reduce heat flux on the fuel cladding in the core, thereby possibly reducing fuel cladding deteriora-tion until analysis can be made to detemine the mechanism of fuel failure. Water chemistry tests are now being per-fomed in conjunction with General Electric Company to study crud effect problems in respect.to the recent fuel failures.
(See Test Section V-b.)
h ere was one short outage in the month of Fovember 1969; on November 5, the unit was shut down to repair a steam leak in the turbine stage drain line to the intemediate pressure heater. On January 8,1970 the unit was removed from service on a schedued outage to replace the off-gas filter. Also, several steam Icaks were repaired during this outage. Following start-up on January 9, the Plant continued to operate at 53 We until January 14, 1970 when the "all rods out" position was reached. At this time, the coast-down commenced and continued until February 13, the start of the seventh refueling outage. At the time of this shutdown, Plant power generation was 48 We (gross).
During the six-veek refueling outage, a turbine inspection was conducted, a control rod drive support structure was installed in the control rod drive room, portions of the Redundant Core Spray System were installed, and a contain-ment vessel leak rate test was conducted.
(See Section IV for other maintenance perfomed.)
We unit was returned to service March 29, 1970. During the remainder of March two short outages occurred due to turbine problems. On April 1 the Plant was removed from service to make minor adjustments to the turbine initial
, Q-{%'., N pressure regulator. Again, on April 11, 1970, the unit was l7 f 4-'
l;.
Ny, removed from service on a scheduled outage to test and in-
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N' spect the emergency governor exerciser and conduct a 30
-g ' megawatt load injection test.
(See Section V-B for load f-N J, re.jecti n test results.)
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24, 1970 a leaking core spray heat exchanger tube g'y/
forced the unit out of service.
C W is tube was plugged, the
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unit returned to service and continued to operate at 53 We 4
1970 when an outage was scheduled to remove untilJune'4}remainingnitridedstainlesssteelindex the three (3 f(10 /O Wo3
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I.
SUMMARY
OF OPERATIONS (Contd)
'A. -Power Operation (Contd) tubes. These tubes were removed due to finding'a flaw
-during preliminary metallographic examinations by General Electric Company in the one nitrided stainless steel control rod drive index tube removed from service during the 1970 refueling outage _ for testing. Testing of these tubes chould be completed about Hovember 1970 and a report will_be sub-mitted at.this tire. We nitrided index tubes were replaced by 'lT h precipitation hardened steel index tubes of the type that were origic. ally installed.
Following start-up June-7, the Plant operated continuously until ' June 28 when a fault on the.138 kV transmission line due to severe storm conditions-in the area caused a load re-
,)ection from 53 FNe (gross).and the reactor _ tripped on high pressure. - Station power loads were automatically transferred to the 46 kV source. During this outage four (k) control rod drive flange "O" rings were _ found leaking and were changed.
We Plant was in the process of' start-up at the end of the report period June 30.
B.
Refueling Outage - The new core loading after completing the seventh refueling outage consisted of the following:
(1)' Reload ~ B bundles - 5; (2) Reload C bundles - 7; 1 5 (3) Reload E bundles - 68; (4) Reload E bundle - 1 l
(Canadian Westinghouse); (5) Reload E bundles - 3 (EEI-Pu).
Other items of interest that occurred during the refueling outage are sunraarized in Section IV.
i C.
Statistics - 2e reactor was brought critical nine times during the report period. The reactor was critical for k,624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> with electrical generation of 236,269 FNh (gross) or 224,712 MWh (net). Se thennal output of the reactor was 729,64h M4h.
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 Curie Released i
November 9,750 December 22,500 January 40,900 February 19,500 March
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April 3,440 uay u,060 June 3,870 Total
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,$yQ Addendum to Eighth, Ninth and Twelfth Reports of
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TWEIF1'H SEMIANNUAL BrPORT OF OPERATIONS F
The following information was inadvertently nmitted from Section I of the Twelfth Semiannual Report of Operations:
On December 12, 1909, high temperatures were experienced c
on Control Rod Drive D-5 while attempting to backflush a rod drive j'
filter. The filter was properly valved to perform the backflushing operation; however, when the valve was opened to backflush the filter, it was opened too wide, reducing the cooling water header pressure thus allowing hot water to flow frc= the reactor vessel to the control rod drives.
Centrol Rod Drive D-5 temperature in-creased to 458 F.
Other control rod nrives showed temperature in-j 1
creases, but did not exceed the 350 F temperature limit. The back-flushing operation was secured upon receiving indications of control o
rod drive system abnormalities and the system temperatures and pres-sures returned to normal. The backflushing procedure was replaced with a chemical cleaning procedure to prevent recurrence of this tyre problem. The overheated drive was removed, dismantled and visually inspected during the February-March 1970 refuelin;, outage.
No signs of damage due to overheating were detected.
[
Prior to plant start-up on January 9,1970, all containment isolation valves were test-operated as required by the Technical Specifications. The inside isolation valve (CV-4031) on the dis-charge of the enclosure clean sump was found to be erratic in operation.
p The coil on the associated solenoid valve (SV-4869) was immediately d
replaced and the valve operated properly. Investigation revealed the replaced coil had a short in the windings and also revealed evidence of overheating. A high temperature coil was installed during the E
February-Ma,rch 1970 refueling outage.
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Routine scram time testing prior to fuel loading during the February-March 1970 refueling outage revealed Control Rod Drive F-2 Tne problem was traced to erratic venting to have erratic ceram times.
lt of the NC27-B solenoid valve that controls the scram inlet and o This valve was replaced and Control valves for Control Rod Drive F-2.
An exten;ive invecti-Rod Drive F-2 tested properly and consistently.
It was con-gation was conducted to determine the mode of failure.
cluded that the most probable cause of failure was that the spring that returns the slug to the shut position upon de-energicing the The manufacturer also solenoid had become disengaged from the slug.
r f t agreed this was the cost probable cause although neither the manu ac ur-er's nor the plant's tecting programs could cause an identical failure.
This investigation was hampered because the plant repairman who per-formed the repairs discarded the slug because it showed minor damage Scram time test procedures have been revised to the saat on the slug.
ill be de-to insure that improper venting of a scram solenoid va:Ve w 3
Thic wgs the firct failure of ti.is type at Big Rock Point.
tected.
While the plant was shut down on June 6,1970, the cortain-The dirty sump inside isolation ment isolation valves were tested.
The failure was traced valve (CV h025) failed to operate properly.
to the solenoid pilot valve (SV LS91) and this valve was replaced prior The dirty sump inside isolation valve then operated to ntart-up.
The cause of the failure The failed unit was bench-tested.
properly.
was attributed to loss of resiliency of the seat material due to nor-I, a;al wear.
yMTII RIMESH PEMIAN?PJAIr RFPORTS49t 'E.'.'.HMS E
i L_eIIQ.rMas made in the total radioective materials. of act, -
A.:
Oc d r il-A W th ut. ion and.fissim co ws-rewarterdic e man m.
c Gr a... M is totals are EJ.ghth and Ninth Semiannun~1-Re puk serrected n9 fonus4 J
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Routine scram time testing prior to fuel loading durir.c the February-March 1970 refueling outage revealed Control Rod Drive F-2 The problem was trace 6 to erratic venting to have erratic scram tites.
of the NC27-B solenoid valve that controls the scram inlet an This valve was replaced and Control valves for Control Rod Drive F-2.
An extensive investi-Rod Drive F-2 tested properly and consistently.
It was con-Gation was conductc0 to determine the mode of failure.
i cluded that the most probable cause of failure was that the spr ng the that returns the slug to the shut position upon de-ener;izing e
The manufacturer also solenoid had become disengaged from the slug.
factur-agreed this was the most probable cause although neither the manu l failure.
er's nor the plant's tecting programs could cause an identica This investigation was hampered because the plant repairman who per-formed the repaire discarded the slug because it showed minor damage procefures have been revisei Scrum time test to the seat on the slug.
l ill be de-to insure that improper venting of a scram solenoid va ve w failurc cf thic type at Eig Roch Point.
This was the first.
teeted.
While the plant was shut down on June 6,1970, the contain-The cirty sump inside isolation ment isolation valves were tested.
The failure was traced valve (CV h025) failed to operate properly.
l ed prior to the solenoid pilot valve (SV L691) and this valve was rep ac The dirty sump inside isolation valve then operatei to start-up.
The cause of the failure The failed unit was bench-tested.
properly.
due to nor-was attributed to loss of resiliency of the seat material mal wear.
g,i?"il ld.%-HWrH SE"I ANNU air REPORTS -t 9-9MU.T~ NQ an_errar.was made in the total-radi-cact,1ve materials of acti-r x Cch u-A~of~th ukion_ _and fisc4 nr coser raertef ah chure,m c u s. m. M Yd^ totals sre Eighth.und Ninth.Semiannua-1-tc p b Jarrected-as m' % -4
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3 II. ROUTITTE RELEASES, DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIALS (Contd)
B.
Liquf d Discharge - During this reporting period, the liquid.
radioactivity released to Lake Michigan by way af the cir-culating water discharge canal numbered 33 batches with total activity of 1.91 curies. All batches were identified by isotopic compositions which showed that approximately 46%
of the activity was Zn-65 The remaining portion was com-posed of Cs-137, Cs-13h, and I-131.
Batches Total mci Month Released Gallons November 1
5,230 31.6
. December 6
29,240 348.5 January 3
14,885 92.9 Febniary 5
24,035 56.9 March 11 54,795 785.0 April 3
13,660 98.6 May 4
.18,850 199.2 June 5
25.580 296.2 TOTAL 38 186,325 1,908.9 C.
Shipments - A total of 47 off-site shipments of radioactive material was made during this reporting period.
Shipment Transfer No Date From Transfer To Radioactive Material 169 11/ 5/69 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 170 11/20/d9 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 171 11/20 69 DPR-6 GE Val, 0017-60 Feed-water Crud and
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(Calif)
Filtrate 0.1 mci 172 12/ 5/69 DPR-6 GE Val, 0017-60 Feed-water Crud and Filtrate 0.1 mci 173 1/ T/70 DPR-6 NFS, CFS-1 12 Spent Fuel Assemblies (New York) 1,187,000 C1 174 1/ 9/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif) yntrate 0.1 mci 175 1/ 9/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 4
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II...
ROUTINE RELEASES,_ DISCHARGES AND SHIPMENTS OF RADIOACTIVE K
MATERIALS (Contd)
C.
Shipnents (Contd)
Shipment Transfer No Date From _
Transfer To Radioactive Material 176
.1/ 9/70 DPR-6 GE SIN-960 Eight Irradiated Fuel (Calif)
Rods 19,300 C1 177 1/21/70 DFB-6 NFS.CFS-1 12 Spent Fuel Assemblics (NY) DOT 1,293,75h Ci SP 5907 e
1 173 2/2/70 DPR-6 NFS,CFS-1 12 Spent Fuel Assemblies (NY) D0T 1,264,209 Ci SP 5907 1
i 179 2/10/70 DPR-6 GE Val 0017-60 Feed-water Crud and (Calif Filtrate 0.1 mci 180 2/10/70 DPH-6 GE Val, 0017-60 Feed-water Crud and l
(Calif)
Filtrate 0.8 mci
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181 2/23/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.7 mci
,i 182 2/14/70 DPR-6 NFS CFS-1 12 Spent Fuel Assemblics
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(NYhDOT 1,180,755 Ci l
SF 5907 1.
183 2/26/70 DPR-6 GE Val 0017-60 Fuel Bundle Crud Chips (Calif 0 5 mci 16h 2/27/70 DPR-6 NFS CFS-1 12 Spent Fuel Assemblies f
j-(NY) DOT 602,878 Ci SP 5907 i
185 3/12/70 DPR-6 GE Val, 0017-60 Fuel Bundle Crud Chips (Calif) 3 0 mci i
186
-3/12/70 DPR-6 NFS,CFS-1 12 Spent Fuel Assemblics (In) DOT 566,597 Ci SP 5907 187 3/12/70 DPR-6 NPI 19-12667-01, Irradiated Cobalt DOT SP 5364' 592,210 Ci 188 3/16/70 DPR-6 Dresdon Nuclear Fuel Inspection Equipment
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II.'. ROITINE RELFASES, DISCHARGES AND SHIP!ENTS OF RADIOACTIVE ITATERIAIE (Contd)
I-C.
Shipments (Contd)
Shipment -
Transfer No Date From Transt 'r To Radioactive Materials 189 3/25/70 DPR-6 NFS-CFS-1 12 Spent Fuel Assemblies (IU) DOT 539,641 Ci SP 5907 190 h/7/70 DPR-6 NFS-CFS-1 12 Spent Fuel Assemblies (IU) DOT 562,1h8 Ci SP 5907 191 h/8/70 DPR 6 UPI. 19-12667-01 Irradiated Cobalt
-DOT SP 536h 580,000 C1 192 h/ 9/70 DPR-6 GE Val, 0017-60 Three Dummy Rods (Calif) 65,090 Ci DOT SP 5971 t
193 h/11/70 DPR-6 HPI, 19-12667-01 Irradiated Cobalt DOT SP 536h 368,100 ci 2
194 h/14/70 DPR-6 Isotopes, Inc 200 ml Henctor Water 29-55-6 200 ml Condensate 0.1 mci 195 h/17/70 DPR-6 NFS-CFS-1 12 Spent Fuel Assemblies j
(IU) DOT hh0,979 Ci i
SP 5907
'196 h/22/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 197 h/28/70 DPR-6 NFS-CFS-1 12 Spent Fuel Assembliec (IE) DOT 517,416 CL SP 5907 198 5/15/70 DPR-6 GE Val, 0017-60' Feed-water Crud and (Calif)
Filtrate 0.1 mci 199 5/15/70 DPR-6 GE Val, 0017-60 Turbine Extraction (Calif)
Piping 0.5 mci 200
- 5/15/70 DPR-6 NFS-CFS-1 10 Spent Fuel Assemblies (In) DOT h38,806 Ci SP 5907 i
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6 II. ROUTINE F,ELEASES, DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIAIE (Contd)
C.
Shiinents (Contd)
Shipment Transfer No Date From Transfer To Radioactive Material 201 3/20/70 DFB-6 NRL-8-1393 ( A-66) Irradiated Metallic BE "(27 Vessel Specimens h7.9 ct 202 5/28/70 DPR-6 NFS-CFS-1 12 Spent Fuel Ascer.bliec (NY) DOT 6hh,638 Ci SP 5907 203 5/29/70 DPR-6 B & W SNM-778 In-core Wire 1 mci I4/nchburg, Va 20h 6/2/70 DPR-6 NECO 16-NSF-1 Two Fuel Channelo, Misc (A-11)
Material 17h0 C1 (Morchead,Ky)
DOT SP 6058 205 6/h/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels, Mice (A-ll)
Material 1740 C1 (Morehead, Ky)
DOT SP 6058 206 6/5/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 207 6/ 8/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels, Misc (A-ll)
Material 1740 C1 (Morehead,Ky)
DOT SP 6058 208 6/10/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels (A-ll) 17h0 C1 (Morehead,Ky)
DOT SP 6058 209 6/12/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels, Six (A-11)
In-cores 29ho Ci D0T SP 6058 210 6/15/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels, Two (A-ll)
In-cores 22h0 C1 (Morchead,Ky)
7 II.
ROUTINE RELEASES, DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIALS (Contd)
C.
Shipnents (Contd)
Shipment Transfer
_ No Date From Transfer To Radioactive Material 211 6/15/70 DPR-6 GE Val, 0017-60 Two Control Rod Drive (Calif)
Index Tubes, one Drive Piston 2 mci 212 6/17/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels, "vo (A-11)
In-cores 22h0 Ci DOT SP 6058 213 6/19/70 DPR-6 NECO 16-NSF-1 Two Fuel Channels (A-11) 1740 Ci (Morehead,Ky)
DOT SP 6058 21h 6/24/70 DPR-6 GE Val, 0017-60 Feed-water Crud and (Calif)
Filtrate 0.1 mci 215 6/2h/70 DPR-6 NECO 16-NSF-1 Misc Waste Material in (A-ll) 55-Gallon Barrel 100 C1 (Morehead,Ky)
DOT SP 6058 III. RADIOACTIVITY LEVELS IN PRINCIPLE FLUID SYSTEM (FOR SIX MONTHS)
A.
Primary Coolant Minimum Average Maximum
(* Reactor Water Filtrate
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1.09 pCi/cc 1.h5 x 10" h.05 x 10
- Reactor Water Crud
~1 pCi/cc/ Turbidity 1.75 x 10 2.62 x 10" 8.74 x 10
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Iodine Activity
-2
-1
-1 pCi/cc 7.2 x 10 1.3 x 10 2.5 x 10
("}A counter efficiency based on a gamma ener6y 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 hevrs after sampling.
(b) Based on efficiency of iodine two hours after sampling.
1
- oused on APHA turbidity units and 500 ml of filtered
- c. ample.
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III. RADI0ACTIVITI LEVELS IN PRINCIPLE FLUID SYSTEM (FOR SIX IGM (Contd)
B.
Reactor Cooling Water System Minim 1m Average Maximum
-2
-2 k"} Reactor Cooling Water k.36 x 10-1 3 x 10 8.75 x 10 C.
Spent Puel Pool - Radioactivity in the spent fuel pool is principally activated corrosion products from stored fuel and core components.
Minimum Average Maximum I
(" Fuel Storage Pool
-3
-3
-1 pCi/cc 1.16 x 10 8.75 x 10 8.7 x 10 (b)itel Pool Iodine
-5
-4
-3 pCi/cc 2.0 x 10 1 5 x 10 2 x 10 i
i
(*)A counter efficiency based on a gamma energy of 0.662 MeV and one gamma i
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 two hours after sampling.
I 1
IV. PRINCIPLE MAINTENANCE PERFORMED A.
Upper bearings were replaced on the No 2 condensate pump motor due to excessive noise caused by a chipped ball in the bearings.
B.
A reactor vessel beam clamp lock assembly on the southwest grid bars was installed and welded in position on February 28 during the refueling outage. This task was performed by General Electric Company and plant maintenance personnel.
C.
Damaged reactor vessel flange thermocouples (temperature recorder Points 7 and 8) were replaced with units of a more rugged nature.
In addition, both units were encased in stainless steel tubing to provide additional protection to the thermocouple sheath attached to the walls of the reactor extension tank. In the past, the thermocouples were damaged during installation of the reactor head thermal shield. It is hoped that the additional protection will prevent damage in the future.
D.
Five inecore detector assemblies and one dry well were removed from the reactor vessel and replaced with six new in-core detector assemblies (No 11, 13, 14, 16, 17 and 18).
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IV. PRIIGPLE MAINTENANCE PERFORMED (Contd)
E.
During the refueling outage, Bechtel Corporation installed a control rod drive support structure in the control rod drive The support stru tture is designed to prevent a control room.
rod from dropping out of the reactor core if one of the velds joining the control rod drive mechanism to the reactor vessel i
should fail. The structure consists of 32 individual support modules, one for each control rod drive, and associated struc-tural members. Each of the individual modules is removable to pennit replacement of centrol rod drives and is aligned and shimmed to match the location and elevation of its associated drive. For further information refer to Consumers Power Companylettersdated1/14/69and1/23/?OtoAtomicEnergy Cannission.
Clearance dimensions between module and control rod drive vere measured during reactor vessel heat-up to ensure the reactor vessel growth would impose no problems.
Clearances proved satisfactory and have been recorded. Periodic inspection of these clearances will be made.
l F.
A redundant core spray system was installed by Bechtel Corporation under a construction general work onier. This core spray system is designed to provide a redundant source of emergency cooling water to the reactor core.
Water. supply to the new system, which is in parallel with the existing core spray system, can be remotely manually selected from either of the two existing fire system supply headers or from the core spray recirculation system. Operation of the new snray system is automatically initiated by reactor water level and pressure switches that have been added.
All piping, valves, and instrumentation have been installed for this system except for the reactor vessel spray nozzle and associated piping. The spray nozzle vill be installed during a future outage.
All piping velds were radiographed and quality assurance records i
made. The complete reactor water level sensor system and E-ll portions of the redundant core spray system piping within the reactor building, were hydrostatically tested to 2,550 psig and 210 psig respectively.
Further information was supplied by letter from Consumers Power CompanytotheAtomicEnergyCommissiondated2/9/70.
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10 IV. PRINCIPLE MAINTENANCE PERFORMED (Contd)
G.
The annual inspection of all containment vessel penetration velds, as required by Technical Specifications, was conducted. No l
discrepancies were noted.
H.
Repairs vert effected on No 1 stack fan by installing a new fan assembly, new bearings, and new stack fan dtmper. Investigation revealed that while in operation during the month of February, a section of the damper-operating arm dropped into the fan assembly shearing off the blades.
No 2 stack fan damper was replaced with a new damper. Ihe damper on this fan also showed excessive wear. Inspection revealed the fan blading to be in good condition and no further repairs were necessary.
I.
Ultrasonic examination of primary system piping velds was performed by General Electric Company personnel. Results were satisfactory and no defective velds were noted.
Piping velds examined were as follovo:
1.
Twenty 3-inch pipe vel?ds in the clean-up demineralizer system located in the recirculation pump room.
2.
Two 20-inch pipe velds in main recirculation pump loops, 1 and 2 located in the recirculatio: pump room.
3.
One 6-inch T and three velds attached to it in shutdown system in recirculation pump room.
h.
Four safe end to nozzle velds and safe end to pipe velds were examined on each of the following nozzle numbers on the steam drum:
a.
Downcomer No B-4.
b.
Riser No C-1.
c.
Feedvater No E-2.
d.
Steam outlet No D 4.
5 Three "J" velds attaching control rod drive housings to stub tubes in control rod drive room.
The insulation removal work and the installation of the snap-on insulation vere performed by Consumers Power Company maintenance personnel.
9
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IV. PRINCIPLE MAINI'ENANCE PERFOPMED (Contd)
J.
Control rod drives D-5, D-6, and E-2 were changed out and three i
spare reconditioned drives installed. Drive B-5 was lowered and new flange 0-rings installed. This work was performed during the refueling outage.
- On June h,1970 an outage was scheduled at the recommendation of General Electric Company for the removal and inspection of three control rod drive assemblies which had special components of stainless steel nitrided index tubes.
The request was prompted by the inspection of another nitrided index tube which was removed during the last refuelin6 outage and returned to General Electric for metallurgical testing.
General Electric Company personnel dye-penetrant checked and ultrasonic tested the three drive index tubes and piston' tubes from the drives that were removed. Two of the three index tubes and one piston have been returned to General Electric for metal-lurgical testing.. The one remaining nitrided index tube will be installed.in one of the spare drives and reinstalled in the reactor at a later date.
During an outage on June 29, 1970 control rod drives C-2, C-4, D-6, and F-5 were lowered and new flange 0-rings installed.
Excessive water leakage at the flange necessitated the replace-ment.-
K.
A defective seal cartridge on the No 2 reactor recirculation 1
pump was replaced with a rebuilt spare. Temperature limitations could not be met for proper pump operations so the spare seal was' installed.
L.
During the refueling outage, the core spray heat exchanger tub-bundic.vas inspected for leaks. Five leaking tubes were r.
and plugged.
During an outage-in April, one tube was found to be leaking and was plugged.
Due to the recent problems encountered with leaking tubes, a leak test is being performed on a weekly basis. A new tube bundle has been ordered and will be delivered in August.
M.
During the refueling outage in February and March, a complete turbine-generator inspection was performed. No deterioration 3
or damage of any consequence was found. All turbine blading and shrouds appeared to be in good repair. Deteriorated stage packing was replaced as necessary.
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IV. PRINCIPLE MAIfffENANCE PERFORMED (Contd)
M.
(Contd)
Generator and exciter internals were inspected; only damage of a minor nature was found. Generator resistance and hi-pot tests were satisfactory.
Rotor commutator rings were found grooved and were turned down; the exciter collector rings were turned down also.
J Turbine stage drain piping was radiographed and all piping and piping bends that showed deterioration were replaced. Namerous i-steam leaks in this piping the past two years prompted this repair. work.
The main generator oil circuit breaker (116) closing scheme was modified to provide an interlock with the 286 differential tripping auxiliary. If tripping of the 116 is a function of 1,
the operation of the 286, the 116 cannot be closed until the 286 is reset. 2 1s change was made as a result of the double closure experienced in November 1970 when the unit was being synchronized and is now identical to the scheme used with the transmission line oil circuit breaker (199).
I N.
Due to a resin leak in the makeup demineralizer, the unit was inspected and found to have a loose inlet diffuser and also two l
pipe plugs on the air sparger ring leaking air. Repairs were i
effected and the demineralizer returned to service.
I O.
See Section V (Facility Changes) for further principal maintenance perfomed.
I V.
CHANGES. TESTS AND EXPERIMENTS PERFORMED PURSUANT TO 10 CFR 50 59 (a)
I
-A.
Facility Changes 1.
C-127 Picoammeter Range Switch - he range switch gain plug was replaced with a gain adjust'nent module which will allow gain change steps (5) of lower magnitude than gain plug removal. Bis gives greater flexibility with regard to matching power range flux readings tc> calculated power j
levels.
In order to derive maximum benefit from this modification, the range switch sensitivity was decreased by a factor of 3.16%. Factory parts were on hand for this field gain change, E
but the change had not been completed previously as it was not deemed necessary. In the past, the range switch gain plac was l
used to short circuit the feed-back resistance and major changes I
in calibration accomplished by detector movement.
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i V.
CHAUGES. TESTS, AND EXPERIMENTS PERIORMED PURSUANT TO 10 CFR 50.59 (a) (Contd)
A.
Facility Channes'(Contd) 2.
C-lkh Reactor Protection System Low Reactor Level Censors -
The four sensors (RE 09 A-thru D) were replaced with constant
. head chambers and remote sensors to eliminate high radiation exposures on routine testing of same during shutdowns and to eliminate 16 valves in a high radiation area. These valves had required many packing repairs which resulted in high maintenance radiction exposures.
Following cold water calibration checkout of the sensors and prior to refueling,.cm additional test was held. The water level in the reactor uno dropped to trip point eleva-tion to verify switch operation on an actual low water level.
All units functioned satisfactorily.
3 C-133 Control Rod Position Indication Motor-Generator -
Circuit changes described by Technical Specification Change 20_(Atomic Energy Commission letter to Consumers Power Company dated 3/27/70) were made to the control scheme of the motor-generator set to use it as an emergency source of power instead of the normal source.
t.
position indication vill normr.lly be fed from Instrument and Control Transformer 1 Y.
In the event of loss of this power supply, the automatic throv-over scheme vill start the M-G set and connect the position indication i
to its output.
i Test switches were installed in the scheme to allow functional testing as well as provide operation for maintenance purposes.
h.
C-lh7 Clean-Up Demineralizer Controls' - A solenoid valve was added in the control air line between the discharge valve hand control switch and the discharge valve. The solenoid valve is controlled by the demineralizer feel pump control circuit and will insure automatic closure of the pump discharge valve whenever the pump is off. In the past, differential pressure across the system could cause flow, even though the pump was off.
5 C-lh6 Dnergency Diesel G *nerator Control - The annunciator circuit was modified to include annunciation of the " manual" position of the control switch as a substandard condition.
Previously, the " Emergency Generator Engine Trouble" alam was initiated only in the "off" position of the control switch (in additior o protective alarms) to indicate a sub-I standard condition. This modificatina vill provide alarm protection for every position of the control switch.
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14 1
V.
CHANGES. TESTS. AND EXPERIMENTS PERFORMED PURSUAlfr TO 10 CFR 50.59 (a) (Contd)
A.
Facility Changes (Contd) 6.
C-lh5 Turbine Moisture Separator Drain Control - The Drain control valve inlet piping was increased from 6-inch diameter piping to 12-inch to increase the loop seal volume. This was done to provide greater stability to the level control in an effort to eliminate fluctuations in the loop seal, low pressure heater, and condenser hotwell.
B.
Tbsts 1.
The water quality test program started during the last report period has been completed. This program was a joint effort between General Electric Company and Consumers Power Company and was a study on condensate and feed-water system corrosion products.
The results of the test shoaed relatively low corrosion product input to the core during Puel Cycle 7 It has been estimated that approximately 80 pounds of copper were deposited on the fuel during Cycle 7 Since the total input of copper to the primary system via the feedvater during Cycle 7 was estimated to be less than 10 pounds, the presence of a copper source within the primary system was demonstrated.
A test program is progressing at this writing on corrosioa and leak rate studies on the clean-up demineralizer heat exchanger loop.
2.
A temperature coefficient test was conducted in March prior to power operation with the newly loaded core. Test data indicagedthatthetemperaturecoefficientturnednegative at 147 F after adding 10.14 of reactivity.
3.
A load rejection test at 30 Mwe (gross) was conducted on April 12, 1970. It was hoped that data from this test fol-loving the change from 1.7" operator to the h" operator would confirm that a considerably higher power test would be successful. However, the reactor just barely survived without a scram.on high pressure during the load rejection test. The reactor pressure spiked at 50 psi above the operating point leaving 0 margin from scram if the reactor had been at the 1335 psis (the Plant was at 1329 psig).
The frequency peaked at approximately 6h hertz and the I
pump flows increased slightly during a period of operation at 61.2 hertz prior to return to IPR control.
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15 i
VI. ' PERIODIC TESTING PERFORMED AS RECUIRED BY. THE TECHNICAL SPECIFICATIONS -
The following: tabulation shows the required frequency of testing, plus t.he testing date of the systems or functions,.which may be periodically tested per Technical Specifications:
t
.3ystem or Fu..ction' Frequency of Dates
-Undergoing Test Routine Tests 7bsted-Control Rod Drives-Continuous withdrawal and insertion - Each major. refueling and 3/13/70 of each drive over its stroke with' at least once every six normal hydraulic system pressure.
months during periods of Minimum withdrawal time shall be power operation.
.23 seconds.
Withdrawal of each drive, stopping Each major refueling and 3/13/70 at each locking position to check at least once every six latching and unlatching operations months during periods of
.and the functioning of the position power operation.
indication system.
Scram of each drive from the fully Each major re1Neling and 3/1h/70 v'.thdrawn position. Maximum scrar at least once every six time from system trip to 90% of months during periods of insertion shall not exceed 2.5 power operation.
seconds.
Insertion of each drive over its Each major refueling but 3/15 70
/
entire stroke with. reduced hy-not less than once a year.
draulic system pressure to deter-
. mine that drive friction is normal.
Control Rod Interlocks Rod withdravr.1 blocked when any Each major refueling but 3/13/70 two accumulators are at a pros-not less frequently than sure below 700 psig.
once every twelve months.
l-Rod witndrawal blocked when two Each major refueling but 3/13/70
'of three power range channels not less frequently than i.
rerd.below 5% on 0-125% scales once every twelve months.
.(or.below 2% on their 0-40%
seties) when reactor power is I
ab >ve the minimum operating i..
raage of these channels.
Rod withdrawal blocked when scram Each major refueling but
-3/13/T0 dump tank-is bypassed.
not less than once every
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twelve months.
l' i-
16 VI. PERIODIC TESTING PERFURMED AS REQUIRED BY THE TECHNICAL SPECIFICA-TIONS (Contd)
System or Function Frequency of Dates Undergoing Test Routine Tests Tbsted Control Rod Interlocks (Contd)
Rod withdrawal. blocked when mode Each major refueling but 3/13/70 selector switch is in shutdown not less frequently than
- position, once every twelve months.
Other Liquid poison system component.
Two months or less.
12/26/70 3/ 5/70 check.
5/L/70 Post-incident spray system auto-At each major refueling 3/15/70 matic'contrcl operation.
shutdown but not less frequently than once a year.
Core spray system trip circuit.
Not less frequently than 3/27/70 once every twelve months.
Dnergency condenser trip circuits.
Not less frequently than 3/27/70 once every twelve months.
Containment Containment. sphere access air Once every six months or h/ 7/70 locks and vent valves, leaka6e less.
rate.
Isolation valve operability and At least once every 3/ 8/70
. leak tests.-
tvelve months.
Isolation valve controls and Approximately quarterly.
1/ 9/70 instrumentation tests.
3/ 8/70 i '
3/16/70 6/ 5/70 Penetration inspection.
At least once every 3/15/70 twelve months.
Integrated leak test.
Once every two years.
3/25/70
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17 VI. PERIODIC TESTDIG PERFORMED AS REQUIRED BY THE TECIDIICAL SPECIFICA-TIONS (Contd)
'~
The fc110 wing instrument checks and calibrations were perfomed at least once e month:
1 1.
Reactor safety system checks not requiring plant shutdown.
2.
Air ejector off-gas monitor.
3 Stack-gas monitor calibration.
14 Etergency condenser "ent monitor.
5 Process monitor.
6.
Area munitoring system.
a%
Gerald J. Valke Nuclear Fuel Management Administrator Consumers Power Company Jackson, Michigan Date: September 2,1970 Sworn and subscribed to before me this 2nd day of September 1970 Wu (1_ uvit.'
Notary Public, Jackson County, ::ichigan
)tr commission expires January 15, 1972 V
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