ML20030A507
| ML20030A507 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/24/1965 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090794 | |
| Download: ML20030A507 (17) | |
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A' Docket No. 50-155
'Filo COP 3hff1I Report of Operation of Big Rock Point Nuclear Plant License No. DPR-6 May 1, 1965 Through October 31, 1965 This report, submitted in accordance with Paragraph 3.r.(3) of Operating License No. DPR-6 (effective May 1,1964), covers the tiird six-month period of operation of the Big Rock Point Nuclear Plant (Plant) under this license.
I.
SUMMARY
OP OPERATIOg A.
Thermal Shield Modifications The Plant resumed operation on September 4, 1965, following the extended outage to modify the thermal shield and install and test a new thermal shield seal.
The early part of this reporting period was spent fabricating and installing additional instrumentation to be used during further cold flow tests in an effort to pinpoint the exact excitation mechanism causing tne thermal shield vibration.
Installation of the instrumentation and the 12 stainless Jteel seal stabilizer weights (weights) was completed on June h, 1965 Puel loading, to a hh-bundle core, was then started in preparation for further flow tests.
The first flow test (weights installed - 44 fuel bundles) was conducted on June 15 at core pressure drops up to 11.8 psi.
The thermal shield was completely stable with no signs of self-excited vibra-tion.
This test appeared to indicate that the weights were an adequate "fix."
However, further inspection of the seal ring prior to and during the instrument installation had shown the ring to be deformed and the decision was made to start fabrication of a new top-mounted thermal shield seal while continuing the testing to verify that the lower seal ring was indeed the excitation mechanism.
Proceeding on this basis, the weights were removed and an
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inflatable rubber seal was installed at the top of the thermal shield.
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This seal, when inflated with demineralized water, completely sealed the r
annulus area between the thermal shield and the reactor vessel vall. When deflated, the seal occupied slightly less than half of the annulus area.
' Cold flow tests were conducted on June 2h with this seal both inflated and deflated in hopes that the thermal shield vibration could be " turned off" and'then " turned on."
The tests showed the thermal shield to be stable under b'oth conditions (inflated or deflated) of the inflatable seal. How-ever, measurements of flow and core pressure drop did demonstrate that the leakage flow past the lover seal ring was very hig% These measurements.
. tended to confirm the seal excitation theory and the desirability of dis-l abling the lower seal ring completely and installing a new seal at the top.
j A further flow test with the inflatable seal was performed l
on June 28 after si: fuel bundles had been removed from the core to allow a
j' flow testing up to 12 psi core pressure drop. Again, the thermal chield was stable with the inflatable seal either inflated or deflated.
In view of these results, it was decided to remove the inflatable seal and rerun the flow test. During this test, which was performed on July 1, the thermal shield " broke into" sustained self-excited vibration, demonstrating that even the slight dampening effect of the deflated top seal was sufficient 1
to stabilize the thermal shield.
Now that the thermal shield vibration had been reproduced, a final test was performed with the inflatable seal reinstalled.
This test took place on July h and demonstrated conclusively that the top seal, 4
in either the inflated or deflated condition, would stabilize the thermal shield. Therefore, with the excitation mechanism now clearly defined, preparations for installation of the new top-mounted seal were started with removal of fuel and core internals from the reactor vessel.
The first two figures appended to this report show the new top seal design. The seal is composed of five 72 segments, each containing a seal bar which is spring-loaded against the reactor vessel vall. All parts are either 304 SS or Inconel. The seal segments and threaded parts of the spring housing are nitrided 304 SS.
Leakage paths
.between segments and through the housings are designed so as to give j
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proper cooling flow in the annulus. The last figure shows one segment
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of the sea' as installed.
4 3
The installation of the top seal was completed on August 8, 1965 In conjunction with this installation, the lower seal was disabled to preclude the possibility of any instability resulting from operation of the two seals in series.
This disabling was accomplished by vedging the bottom seal open (about 3/8") with a self-locking vedge which was fur-ther secured by velding to the thermal slield.
Following the above work, the core was again loaded and flow tests with the nev (top) seal installed vere performed on August 20.
The thermal shield was completely stable during these tests and the top seal functioned satisfactorily.
B.
Plant Start-Up With all equipment instulled and checked out, preparations for resuming power operation commeneec. All special instrumentation was removed or disconnected and the reactc.r core was loaded to 84 fuel bundles.
Following the low-power physics tests (see Section V-B, Special Tests),
three high-temperature displacement fauges were reinstalled at the top of the thermal shield to provide monitoring during the approach to power and the reactor vessel head was installed.
All preparations for power operation were completed on September 3 On September 4, the Plant was once more connected to Consumers Power Company's electric system.
The special instruments on the thermal shield were monitored continuously during the start-up and approach to power, with the last instrument remaining operable for approximately two weeks after full power had been reached. The thermal shield was completely stable during all modes of operation throughout this period, confirming our belief that the thermal shield vibration problem had been resolved.
C.
J'over Operation The Plant reached nominal full power on September 10 and maintained full load until Captember 17, when a relaying malfunction on the 138 kv transmission system resulted in the complete loss of load on the Plant. This resulted in a turbine trip and reactor scram.
Recovery from the scram was normal and power operation j
resumed on September 18. The Plant then held nominal full power until September 30 when the unit was taken off the line for about 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> to i
repair a steam leak in a turbine stage drain line. Following this short
4 outage, full power was maintained until October 30 when the Plant again was taken off the line on a scheduled outage to repair minor steam leaks under the turbine.
Indications of possible leaking fuel were evident almost immediately when power operation resumed in September.
By the middle of the month, the off-gas activity had risen to about 1500,ue/second and there was little doubt that we were seeing a progressive failure of one or more of the stainless steel clad fuel rods rather than a pinhole leak in one of the new fuel bundles that had just been inserted in the core.
From mid-September through mid-October, the off-gas activity rose quite consistently on an exponential curve to about 15,000 pc/second, and then started showing signs of leveling off. At the time of shutdown on October 30, the off-gas activity level was about 18,000 pc/second. con-tinued observations since the end of this reporting period have definitely shown the off-gas activity to have leveled off at about 18,000 to 19,000
,pc/second, with this level still holding as of December 21, 1965 During preparations for start-up and return to power on October 31, difficulty was encountered with one of the peripheral control rod drives. At first, the drive would not withdraw from full in and, after movement of the drive was achieved, it was noted that the collet assembly would not relatch except when the drive was scrammed.
Since the primary system had been "bott?ed up" since the shutdown on October 30, cool down of the system was ict?diately initieted preparatory to replacing the de-fective drive.
At the same time, all o~mer control rod drives were re-While all the drives jogged pl ;perly prior to being scram tested, i
tested.
three additional drives showed evidence of sticking or sluggish collet assemblies after being scrammed. The decision was made at this time to inspect, clean and modify all control rod drives that had rot been modi-fied previously. This amounted to 22 drives (also see section on Mainte-nance Activities). It should be noted that none of the 10 drives modified previously were among those with collet difficulties.
Inspection of the drives showed the collet malfunctions, in every case, to be due to small metal particles lodged in the close-clearance area between the guide sleeve and the collet tube.
In all other aspects, the
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drives were in excellent' condition. Inspection of the drive screens showed small Nnounts of resin and only occasionally a small metal chip from the i
.recent machining operations, indicating that a thorough cleaning job had l
been done following the thermal shield modification work.
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In retrospect, it is apparent that the 4 to 6 days of addi-j tional downtime should have been taken, prior to the September start-up, to clean, inspect and modify all 32 control rod drives. We believe that the small metal particles drifted down into the drives during those periods of time. When the control rod drive pumps were not on and the temporary cover f
plate'over the drives was off. The clearance between the index tube and the l
upper screen is sufficient to allow a few of these small particles to bypass l
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the screen and drop directly into the collet assembly.
i D.
Statistics The reactor was brought critical 21 times and was critical for 1300.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during this reporting period. Total heat produced by the reactor was 280,073 Mvh(t). The turbine generator was on the line for 1278 7
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hours with a gross electrical generation of 90,263 Mvh. Net Plant electrical generation for this period'was 83,493 Mvh.
E.
R&D Program No special R&D Program tests were performed during this re-porting period with the exception of the R&D fuel verification tests (see section on Special Tests).
Twenty-two R&D fuel bundles were in the core for the September start-up. Irradiation on these-bundles continued through the end of this re-porting' period. The 22 bundles included 2 of the 8 Phase I bundles, 15 of the Phase II bundles (described in Section 8 of the Technical Specifications) and'5 additional Phace II bundles that had not been in the core previously.
Of the 5 new Phase II bundles, 2 are 11-mil svaged Incoloy-800, identical to the previous 6 Phase II ll-mil Incoloy bundles except that they are svaged-over-pellets rather than svaged-over-powder, and the remaining 3 bundles are
_l 30-mil Zirealoy-2 clad and identical to the 3 previous Zr-2 clad Phase II i
bundles with the exception that they have vibratory compacted UO p uder rather 2
than UO p 11 ts.
These 5 new Phase II bundles complete the developnental 2
fue) program as originally planned. At this time, there are no plans for 6'
an further R&D fuel to be fabricated.for insertion into the Big Rock Point i
reactor under the original R&D Program.
6 II.
ROUTINE RELEASES, DISCHARGES AND SHIPMENTS OF RADIOACTIVE MATERIALS A.
The gaseous radioactivity released from the stack during power operation averaged 5000 ac/second of activation and fission gases. Based J
upon 1167 effective full power hours of operation during this reporting period, this results in a total release of approximately 2.1 x 10 curies of activation and fission gases to the environs. Gaseous release during periods of shutdown was negligible.
B.
During this reporting period, the liquid radioactivity releases to Lake Michigan, by way of the circulating water discharge canal, numbered 88 batches, with a total activity of 0.83 curie. Four batches were released on a partially identi?ied basis wherein at least 90% of the activity was determined to be a combination of Co and Zn All other batchec were released under unidentified limits.
C.
There were 7 off-site chipments of radioactive material during this reporting period:
Shipment Type of Number Date Transfer License Numbers Radioactive Material 1
8-9-65 DPR-6 to 0017-60 (Consumers Two Antimony Source Pins -
Power to General Electric -
300 Curies Vallecitos) 2 8-10-65 DPR-6 to 0017-60 (Consumers Three Contaminated Diving Power to General Electric -
Suits - 0.001 me - Mixed Vallecitos)
Fission Products 3
3 10- 1-65 DPR-6 to 20-685-2 (Consumers 794 Ft - Dry. Lov< avel Power to Allied Crossroads Waste - 219.h me Jnemical Nuclear Corporation)
NOS h
10- 7-65 DPR-6 to 0017-60 (Consumers Three Liters of Radicactive Power to General Electric -
Liquid 10)1c - Chemical Vallecitos)
NOS 3
5 10-21-65 DPR-6 to 20-685-2 (Consumers 48 Ft Radioactive Spent Power to Allied Crossroads Demineralizer Resin -
Nuclear Corporation) 81 5 Curies 3
6 10-22-65 DPR-6 to 20-685-2 (Consumers 190 Ft Radioactive Spent Power to Allied Crossroads Demineralizer Resin -
Nuclear Corporation) llh Curies 3
7 10-27-65 DPR-6 to 20-685-2 (Consumers 200 Ft Radioactive Spent Power to Allied Crossroads Demineralizer Resin -
Nuclear Corporation) 60 Curies I
7 III. RADIOACTIVITY LEVELS IN PRINCIPAL FLUID SYSTD.tS (FOR 6 M0'ITHS)
A.
Primary Coolant Minimum Average
, Maximum Reactor Water Filtrate *
-1 1
pc/cc 7 3 x 10 2 9 x 10 2.0 x 10
- Reactor Water Crud *
-2 0
pe/ce/ Turbidity h.h x 10 5.8 x 10 1.2 x 10 Iodine Activity
-1 0
pc/cc
Background
1 x 10 1 3 x 10 B.
Reactor Cooling Water System Reactor Cooling Water (*}
-2
-2 pc/cc
Background
1 5 x 10 h.3 x 10 C.
Spent Fuel Pool Spent Fuel Pool
-3
-2
-2 pc/cc h.3 x 10 1,9 x 10 2 9 x 10 Iodine Activity (b)
-6
-5 pc/cc
Background
2 x 10 8 x 10 The principal radionuclides in the reactor cooling water system were and Cr These resulted from the activation of the potassium-chromate inhibitor. The principal radioactivity in the spent fuel pool was activated corrosion products going into solution from the fuel and core components (stured in spent fuel pool during thermal shield modification work).
IV.
PRINCIPAL MAINTENANCE PERF0hCD A.
The primary maintenance effort during the reporting period was associated with the extensive flow testing of the thermal shield and with the installation of tI_? new top seal on the thermal shield. While the installation work was performed mainly by the reactor manufacturer, Big Rock Point personnel provided assistance in all areas.
" Based on APHA turbidity units and 500 ml of filtered samples.
" 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 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after campling.
(b) Based on efficiency of Iodine-131, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after sampling.
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i B.
The control rod drives have been inspected, cleaned and modified by the installation of a modified collet sleeve and an additional screen (see our informational letter to Dr. Doan, dated December 11, 196h).
Ten of the drives were modified in August 1965, just prior to returning the Plant to power, while the remaining 22 drives were modi-fled in late October.
C.
Two new in-core monitor strings were installed, prior to the September 1965 start-up, due to mechanical damage to cables and connet' ors.
The remaining original in-core strings will be replaced during the next re-fueling outage as they are losing sensitivity due to burn-up of the sensing material (U235),
D.
The first shipments of solid radioactive vaste materials from the Plant site occurred during this reportit.g period (794 cubic feet of dry low-level baled vaste and 438 cubic feet of spent demineralizer resins were transferred to Allied Crossroads Nuclear Corporation for disposal).
V.
,3ANGES, TESTS AND EXPERIMENTS PERFORMED PURSUANT TO 10 CF3 50 59(a)
This section describes the changes made to the lucility within the six-month reporting period without prior Commission approval pursuant to Section 50 59(a) of Title lo, Code of Federal Regulations, to the extent that such changes constitute changes in the facility an described in the Final Hazards Summary Report (FHSR).
It also includes tests and experiments carried out at the Plant without prior Commission approval pursuant to Section 50 59(a).
Each change, test or experiment described was authorized only after a finding by Consumers Power Company that it did not involve a change in the Technical Specifications incorporated in Operating License DPR-6 (ef-fective May 1, 196h) or an unreviewed safety question.
A.
Facility Changes 1.
The mcdifications to the thermal shield, as discussed in the
" Summary of Operations" of this report and the previous report, have now been completed.
Changes to the thermal shield included (a) installation of six
" stilts" te replace the original support brackets; (b) installation of twelve
" stabilizer" weights in the annulus between the thermal shield and the reactor vessel to provide additional dampening to the thermal shield; (c) installation of a new segmented bypass-leakage seal at the top of the thermal shield; and
(
(d) installation of six locking vedges to open and disable the lower seal ring.
9 8
2.
In a continuing effort to eliminate all possible avenues for inadvertent entry of demineralizer resins into the primary system, the follow-ing strainers were installed:
a.
A "Y" strainer in the demineralized waterline to the clean-up demineralizer.
b.
A "Y" strainer in the inlet line to the clean-up de-mineralizer.
c.
A strainer in the demineralized water supply line to the sphere.
d.
A strainer in the " treated vaste" line to the sphere.
3 The high-radiation door to the nuclear steam supply system (NSSS) pipevay was moved back to the opening between Room h26 and Room 424 due to the increased radiation levels being experienced in the area of the clean-up pump.
4.
The main steam bypass valve electronic control system has been completely replaced by a new solid-state system.
The new system has been designed to provide high reliability and to minimize the possibility of spur-ioas opening of the bypass valve.
5 Modification kits, each composed of a new flange strainer and a modified collet sleeve, have been installed in each of the control rod 1
drives as a step to reduce the sensitivity of these drives to foreign material (see letter to Dr. Doan, dated December 11, 1964, describing this modification).
B.
Tests and/or Experiments 1.
Control Rod Blackne.;c Test -
The 32 new control rods (see Change No. 4 to the Techni-cal Specifications dated February 11,1965) were checked for blackness using a 5-curie Pu-Be neutron source and B-lO proportional counter.
In order to check each control rod, a stand was built so that the neutron source, the control rod and the proportional counter could be positioned in the same exact geometry for each test. The neutron source "as left in its storag?
container for these tests in order to utilize the 4 inches of paraffin around the source for thermalizing the neutrons.
Instrument readings of Type 1 control rods showed an at-tenuation of approximately a factor of 10, while readings of Type 2 control 4
rods showed an attenuation of approximately a factor of 3 By comparison,
10 t
apieceofstainlesssteel,1/8" thick (anapproximationofthethicknese of SS in a control rod blade), caused an ettenuation of only 1.1 in instru-ment reading.
2.
Control Rod k' orth Comparison Test -
A complete set or new control rods were installed in the core during this period. The new Type 2 control rods differed from the original Type 2 ccatrol rods with respect to the location and number of B C poison rods.
g A comparison test was conducted in order to verify that the worth of these new Type 2 control rods was not significantly different from the worth of the original Type 2 control rods. An original Type 2 control rod was installed in the E-2 position during core reconstitution (August 1965).
The control rod withdrawal pattern was set so that this control rod would be controlling at critical and period data were then taken over a range of control rod movement (the practical limit was four notenes).
This original Type 2 control rod was then replaced with a new Type 2 control rod and the period measurements re-peated. Data show?d that the new TyIe 2 control rod had slightly less worth than the original Type 2 control rod. Maximum difference in core reactivity, throughthefour-notchrange,was05x10'b1k/t.
3 Fuel Verification Test -
Fuel verification tests were conducted on the 5 new Phase II developmental fuel bundles and established that there were no gross reactivity variations from calculated values.
Each fuel bundle (in turn) was placed in the core test position and core reactivity compared with that obtained when using a fuel bundle of known k in the same core position.
g h.
Temperature Coefficient Test -
The objective of thic test (conducted each time a signi-ficant core change takes place) is to determine the amount of reactivity added. to the core as the system heats from ambient temperature to that point where the temperature coefficient turnu negative. Results of this test showed that the temperature coefficient became negative at 89 F after adding approxi-mately 3 cents of reactivity.
5 Flux Tilting Test -
Indications of leaking fuel had been apparent for several veeks prior to the October 30, 1965 shutdown. This planned outage afforded lI
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the opportunity to experiment with flux tilting in order to determine the effectiveness of ~this method in locating-leaking fuel bundles.
Power was reduced to approximately 70% of rated, an even
. control rod pattern was established, and data. vere recorded from power-indicating instrumentation as well as off-gas activity instrumentation.
A selected control rod was then fully inserted and data again recorded
'after steady-state conditions had.been reached. This control rod was then Jwithdrawn to its original position in the pattern-and anoth7r control rod a
was fully inserted. -Testing continued until data had been obtained for each of the 32 control rod positions.
The resolution of this experiment was insufficient to-2
-be able to determine the exact location of a leaking fuel bundle but it was sufficient enough to indicate the area of the core which vas mest suspect. The data tended to confirm the suspicions that one of the two 10-mil stainless steel clad fuel bundles presently in the core is probably the leaker.
i
- 6. - Cold Recirculation Flow Test -
j A number of recirculation flow tests were run during the reporting. period in order to verify the excitation mechanism involved in'the thermal shield vibration and to evaluate the performance of the modifications.
These tests involved only the recirculation pumps. The reactor was not 4
taken critical and' no control rod movement was involved.
Careful surveillance
'of the special instrumentation was maintained at all times and any periods of vibration were held to the absolute minimum time consistent with obtaining meaningful data.
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VI.1 PERIODIC TESTING PERFOPMED AS REQUIRED.BY THE TECHNICAL SPECIFICATIONS The 'following table.shows the required frequency of testing, plus the_. testing dates of the systems or functions which must be tested periodically as required by the Technical SIecifications:
~
System or Function Frequency of Dates Undergoing-Test Routine Tests Tested Control Rod Drives 6-3-65
~
Each major refueling)and-
. 8-1445
-Continuous withdrawal and insertion at leau quarterly (l of each drive over its stroke with normal hydraulic system pressure.
- during periods of power 8-25-65.
Minimum withdrawal time shall be 23 -
cperation.
7-seconds.
6-' 3 Each major refueling)and Withdrawal of each drive, stopping at least quarterly (l 8-14-65
. at each locking position to check latching and unlatching operations.
during periods of power
-and the functioning ~of the position operation.
- indication system.
6-2-65 Each major refueling)and Scram of each drive from the fully at least quarterly (l 8-14-65 withdrawn position. Maximum scram time from system trip' to 90 percent; during periods or power of insertion shall not exceed 2.5 operation.
seconds.
l Insertion of each drive over its Each major refueling-6-2-65 i
entire stroke with reduced hydraulic but not less frequently 8 6 5 i
. system pressure to detcrmine that _
than once a year.
j drive' friction is normal.
I Control Rod Interlocks I
Rod withdrawal blocked when any Each major refueling but 6-2 65 t"o accumulators are at a pressure not less frequently than 8-14-65
~oelow 700 psig.
once a year.(2)
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l' Change No. 7 dated 7-9-65 changes the frequency of testing from
"... quarterly" to "... once every 6 months."
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Change No. 7 changes the frequency of testing from "... once a year" l
to "... once every 12 months."
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System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Interlocks (Contd)
Rod withdrawal blocked when two of Each major refueling 6-3-65 three power range channels read but not less frequently 8-15-65 below 5% on 0 - 125% scales (or than once a year.(2) below 2% on their 0 - 40% scales) when reactor power is above the minimum operating range of these channels.
Rod withdrawal blocked when scram Each major refueling 6-2 65 dump tank is bypassed.
but not less frequently 8-1k-65 than once a year.(2)
Rod withdrawal blocked when mode Each major refueling 6-2 65 selector switch is in shutdown but not less frequently 8-14-65 position.
than once a year.(2)
(2) Change No. 7 changes the frequency of testing from "... once a year"
(
to "... once every 12 months."
1h System or Function Frequency of Dates Undergoing Test Routine Tests Tested Othe r Reactor safety system scram circuits At each major refueling 5-15 65 requiring plant shutdown to check.
shutdown but not less 9-2 65 year.(3) y than once a frequentl Liquid poison system component check. TVo months or less.
4-9-65 6-10 65 9 65 10-10 65 Test firing of explosive valves.
One year or less.
5-21 65 Post-incident spray system At each major refueling 5-28 65 automatic control oleration, shutdown but not less frequently than once a year.
Core spray system trip circuit.
At each major re.' eling 8-13-65 shutdown but not less year.(4) y than once a frequentl Emergency condenser trip circuits.
At each major refueling 6-21 65 shutdown but not less 9-2 65 year.(4) y than once a frequentl Containment Containment sphere access air locks Once every six months 8 6 5 and vent valves, leakage rate, or less.
8-20 65 Isolation valve olerability and At least once every 8-18 65 leak tests.
twelve months.
Isolation valve controls and Approximatelf quarterly.
6 65 instrumentation tests.
9-2 65 Penetration insIection.
At least once every 4-21-65 twelve months.
( 3) Change No. 7 changes the frequency of the test from "... once a year" to "... once every 12 months."
(4) Change No. 7 deleted the entire paragraph under "Frequen:y of Routine Tests" and substituted the following:
"Not less frequently than once every 12 months."
1 15 l-The following instrument checks and calibrations were per-
' formed at least once a month: Reactor safety system checks on circuits,
not ' requiring plant shutdown to check; air ejector off-gas monitor cali-bration; stack-gas monitor calibration; emergency condenser vent monitor.
calibration; process monitor calibration; and the area monitoring system calibration.
I By Robert L. Haueter (Signed)
Robert L. Haueter Assistant Electric Production Superintendent - Nuclear Consumers Power Company Jackson, Michigan Date: December 24, 1965 I
[
Sworn and subscribed to before me this 24th day of December 1965 (SEAL)
Grace R. Warner (Signed)
Notary Public, Jackson County, Michigan My commission expires February 16, 1968 O
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