ML20030A509
| ML20030A509 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/25/1965 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090801 | |
| Download: ML20030A509 (14) | |
Text
..
\\p 9
q, Q '
DOCKETED f 'H I O g
USAEC
.'S Qf(
f}l 52G1955n {
g vf'4,'?
c-4: ace:r 4
CONSUMERS POWER COMPAITf e
Dociet'No.-50-155 V'
O h&
- ';:.1 e
?'"'=%
c
/o 3
\\
9
',Ello Cong(.h[/
~C ;id 4 %
//
f Report of Operation of Big Rock Point Nuclear Plant
'b License No. DPR-6' November 1, 1964 Through April 30,'1965 This report, submitted in accordance with Paragraph 3.D.(3) of Operating License No. DPR-6 (effe tive May 1,196h), covers the second six-month period of operation of the Big Rock Point Nuclear Plant (Plant) under this license.
I.
SUMMARY
OF OPERATIONS The Plant was shut down for the entire reporting period while analyses, testing and repair work were conducted on the thermal shield hold-down assemblies. Early in th's period, the decision was made to remove the thermal shield support brackets and to support the thermal si teld by a low-stressed flexible " stilt" at each of the six suppor' pads.
These stilts were rigidly bolted to the r.upport pads and also bolted and pinned to the thermal shield. Figure I shows a stilt prior to installation. Figure II shows several of the stilts after installation.
(The stilt, in Figure I, is one of three that 'is instrumented to provide further information on thermal shield behavior during subsequent. testing.)
Installation of the stilts was completed and cleanup of the reactor-versel (vessel) started by mid-February. A small am;unt of machining chips had worked through a gap between the vessel wall and the nylon net chip catcher which necessitated vacuuming of the bottom of the vessel. On completion of this work, (in early March) reinstallation of vessel internals was started.
A series of control rod blackness checks on the new control rods was'also apleted during February. These checks, utilizing the l
plutonium-beryllium neutron source, verified the presence of boron in I
each control' rod.
/
22L m
2 Reloading the reactor to a kh-bundle co.e was started on March 17 in preparation for the cold recirculation alow tests with the new stilts and with thorough instrumentation of reactor internals. All preliminary work was completed by March 27, at which time the cold recirculation flow tests were started.
These tests showed that the tiermal shield would " break into" a lateral vibration at a core pressure drop of about 115 psi and that this vibration was self-excited through some sort of hydraulic feed-back. Movement of the top of the thermal r.hield was about 32 to 36 mils peak-to-peak maximum amplitude, with a frequency of about h cycles per second.
From these tert results, it can be surmised that the previous thermal shield hold-down stud failures were due to a self-excited vibration rather than the fretting wear mechanism previously felt to be the most probable cause. However, the differences between the present thermal shield support and the original support are such that this can only be a supposition.
Analysis of-test data has pointed toward lifting of the thermal shield seal ring as being the most likely cause for the self-excited vibration. Uneven lifting of the seal ring at the higher corc pressure drops could result in unbalanced pressures in the annulus between the thermal shield and the vessel wall.
Changes in the leakage path resulting from lateral mov(ment of the thermal shield also could provide the positive pressure feedback in the annulus.
As a result of the analysis of the tests and postulated causes for the excitation, further tests are planned. Additional in-strumentation will be installed on the tht rmal shield. Instrumentation-will be in=L 11;l on several fuel channels to eliminate fuel and channel movement as a possible excitation cause.
A series of 12 stainless steel weights has been fabricated.
They will be installed early in the next reporting period as part of the next planned tests. These weights (see Figure III) will provide a pre-load of about 13 psi on the outer edge of the seal ring which should prevent _ its lif t' ng.
M
4 4
3 Following the above flow tests, the core internals and fuel again were removed from the vessel In preparation for installation of.the seal weights.and new instrumentation. At the end of thic re-porting period (4-30-65), the. vessel was empty.
Considerable cleaning of the primary system was necessary during this time due to the inadvertent back flushing of the clean-up demineralizer which flushed several cubic feet of resin into the primary system during the previous cold flow test.
The back flushing of the resin resulted from a check valve sticking open, coup eu with unusual valving sequences on the clean-up loop due to the flow test sequences. A strainer on the clean-up loop demineralizer inlet line has been, Ided to prevent any recurrence.
All control rod drives were removed, cleaned inspected and reinstalled during this clean-up operation. Other than the resin and a few metal chips, the control rod drives were found to be in excellent condicion.
II.
ROUTINE RELEASES, DISCHARGES A?D SHIPMENTS OF RADIOACTIVE MATERIAL A.
The gaseous radioactivity released from the stack was negligible since the Plant was shut down throughout the period.
B.
During this reporting period, the liquid radioactivity releases to Lake Michigan, by way of the circulating water discharge canal, nuc.bered 90 batches, with a total activity of 1.1 curies. Twelve batches were released on a partially identified basis wherein at least 90% of the 65 activity was determined to be a combination of Co and Zn All other batches were released under unidentified material limits.
C.
There were six shipments of radioactive material during the reporting period as follows :
Shipment No.
Date Transfer License No.
Radioactive Material 4
11/ 6/64 DPR-6 to General Electric 1 3 me - Unidentified (Vallecitos 0017-60)
(Lead Liner for 6E-Val-601 Cask) 5 11/ 6/6h DPR-6 to General Electric 150 Curies - Activated (Vallecitos 0017-60)
Stainless Steel - Misc Parts 6
12/1/64 DPR-6 to General Electric 20 pc - Liquid Radwaste (Vallecitos 0017-60)
Sampler - Activated Corrosion Products 7
12/3/6L DPR-6 to General Electric 200,pc - Activated Cor-(Vallecitos 0017-60) rosion Products on Profilometer and Parts
h L
Shipment' i
No. _.
Date Transfer License-No.
Radioactive Material 4
-8!
-1/14/65 DPR-6 to U.S.-Naval 50 Curies-- Activated Research Laboratory-Steel and Flux Wires.
I-
_g-1393-2(A-66g Reactor Vessel Surveil-lance Specimens #119 and 0122 9-h/ 8/65' DPR-6 to General Electric 600;ic - Activated Cor-(Vallecitos 0017-60) rosion Products on Pmfilometer and Parts
- III. RADIOACTIVITY LEVFLS IN PRINCIPAL FLUID SYSTEMS ~
A.
Primary Coolant The activity levels in the primary-coolant during-this 4
period of shutdown were :
Min avg Max
-0
-5
-2 Reactor' Water Filtrate ** pc/cc 1.1 x 10 36x10 5.8 x 10
-3
- Reactor Water Crud ** pc/cc/Turb -
k.h x 10 1 5 x 10-2.6 x.lo
-8 Iodine Activity *** pc/cc Bkg-5 9 x lo'9 7 5 x 10 B.
Reactor Cooling Water System j
The principal radionuclides in the reactor cooling water 24 51 system were Na and Cr, which resulted from activation of the sodium chromate inhibitor.
(A portion of this water flows through the biological 4
shield enoling jacket.} This activity is mainly residual activity since the system was drained and flushed the first week of this reporting period and refilled with potassium chromate inhibitor.
j i.
Min Avg Max
-6
-5 Reac_ tor Cooling Water ** pc/cc 2.6 x 10 7 3 x lo-2.2 x 10 C.
Spent Fuel Pool Radioactivity in the spent fuel pool is principally activated corrosion products going into solution from fuel and core components. During this six-month period, there has been a substantial amount of movement of fuel and -core components into and out of the spent fuel pool.
Min Avg Max
-3
-2
-2 Fuel Pool Activity ** pc/cc 2 5 x 10 1,7 x 19 6 5 x 10 Iodine ' Activity *** pc/cc BP4 1 x 10-7 5 x 10-i
- (Based on APHA-turbidity units and 500 ml of filtered sample. )
{
- (A counter efficiency. based on a gamma energy of 0.662 Mev and one gamma-1-
photon per disintegration decay scheme is assumed to convert count rate to miccoeuries. '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 sampling.)
~
31
- (Basad on efficiency of I
, a hours after sampling.)
5 IV.
PRINCIPAL MAINTENANCE PERF0PMED A.
The primary maintenance item during the reporting period was the installation and testing of the thermal shield hold-down system. While this work was mainly performed by the reactor manufacturer, Big Rock Pbint personnel provided assistance in all areas.
Installation of core internals and fuel, as well as cleanup of the system,was performed by Big Rock Point personnel.
B.
The second largest maintenance item (discussed previously in Summary of Operations) was the cleanup of the resin from the system fol-loving the flow tests. All vessel internals and the steam drum were thoroughly vacuumed. The dcwncomers and pump discharge piping were flushed.
All control rod drives were removed, inspected and cleaned, and the con-trol rods and fuel were inspected and flushed clean as necessary.
As a final item, screens will be placed in the recirculstion suction valves and the system flushed using the recirculation pumps.
Selected control rod drives will be inspected following further flow testing.
C.
Other miscellaneous maintenance items of interest are :
1.
The turbine vacuum relief diaphragms were replaced after an air-leakage test disclosed several leaks around studs.
2.
The gas charging valves, on the control rod drive ac-cumulators, were replaced with ball type valves and snap-on hose connectors to simplify recharging and reduce valve packing leaks.
3 During a routine test run, the emergency diesel generator failed due to a cooling water failure.
It sustained slight damage. The engine was removed, shipped off site for rebuilding and upon return to site was reinstalled.
4.
Two reactor-vessel surveillance coupons (No.119 and No.122) were removed from their holders and shipped to the U.S. Naval Research Laboratory for analysis. One coupon had been irradiated in a position between the core and the thermal shield with a calculated fast flux about 10 times that seen by the vessel vall. The other coupon had been irradiated between the thermal shield and the vessel vall with a calculated fast flux about 1.2 times that seen by the vessel vall.
5 The "x" and "y" phases of No.1 house service water pump shorted during operation. The motor was removed and shipped off site for rewinding.
4 6
s 6.
All Magnetrol water-level switches used the reactor protection system were tested at a minimum of 310 F fe er magnetic operation. It had been reported that a group of svite'.cs had been manu-factured with a material that lost its magnetic properties with increased temperatures. All Magnetrol water-level switches in service at Big Rock Point tested satisfactorily.
- V.
CHANGES, TESTS, AND EXPERIMENTS PEEPORMED PURSIWU TO 10 CFR 50 59(a)
This section briefly describes the changes made to the facility, within the six-month reporting period, without prior Commission approval pursuant to Section 50 59(a) to the extent that such changes
. constitute changes in the facility as described in the Final Hazards Summary Report.
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, or an unreviewed safety question.
A.
Facility Changes 1.
The overflow line from the reactor vessel refueling extens: on tank has been modified so that it can be drained to the clean sump if desired as well as the other routes previously provided.
2.
The packing leak-off line from the turbine main steam stop valve has been rerouted to the turbine room sump rather than to the main condenser.
It is suspected that this has been a major source of the air in-leakage to the condenser.
3 Pressure switches and relays have been added to the starting circuits of the condensate pumps to provide automatic starting of the stand-by pump upon low reactor feed pump suction pressure.
h.
Modifications and improvements to the main steam bypass valve system have continued. The pressure compensated servo-flow control valve (which had previously given trouble) was replaced with a needle valve.
The drain capacity from the isolate solenoid valves was also increased to reduce the pressure spikes during "ir alate" operation.
The ele tronic controllers have been modified to improve 4
their reliability during the interim period until the new control system
7 (now on order) can be checked out and installed. The present controllers have been modified by installation of a permissive circuit so that both controllers must be calling for an open signal before the valve will open.
Alarm circuitry is provided to alert the operator to both normal and abnormal operation of the system.
5 The d-c motor on the spare gland seal exhauster has been replaced with an a-c motor to allow for normal service use of this exhauster 1or long periods of time. The d-c motor placed too much load on the d-c supply system for more than a few hourt operation.
D-c operation of.this exhauster is not required since the exhauster is not needed following loss of a-c power to the Plant.
6.
A new enclosure, utilizing a locked and alarmed folding-type metal fence, has bet ir. stalled around the high-radiation area of the turbine in place of the old locked-chain installation. The area enclosed by the fence also has been increased somewhat to compense:o for the slightly higher radiation levels which resulted from 240 Mwt operation.
7 The chrome-moly turbine stage drain manifold was re-placed with a stainless steel manifold to combat the erosion problem.
Experience on a similar turbine had indicated the desirability of making this change at a convenient time.
8.
Four nitrided 304 ss index tubes and two nitrided 304 Ss inner piston tubes have been installed in four different control rod drives to get experience on the performance of this material.
(See letter to Dr. Doan, dated December 11,196h, regarding our analysis of this material.)
9 Modifications to the condensate demineralizer tanks were made to enable the bottom six inches of resin to be sluiced from the tanks, thus improving performance. Also the regeneration cycle of the regeneration facility has been modified to provide an air scrub and separate back washes for the anion and cation resins to improve the cleaning of the resin beads during regeneration.
10.
The control rod oscillator equipment, used for R&D testing, has been removed and returned to General Electric as this type
{
of testing has been completed.
8 B.
Test and/or Experiments 1.
A 72-hour "no-purification" test was condue ed on the water _ from both the reactor and the spent fuel pool in order to try and compare the rate of dissolved activity build-up (reactor water vs spent fuel pool water) during shutdown conditions. Results of the test were comewhat inconclusive, but they did show that the Zn goes into solution in the spent fuel pool water from reactor internals and the fuel thus 3
giving a steady rise of 1 x 10 cpm /ml/ day in gross filtrate activity and a steady rise of 0.1 pmho/cm/ day in conductivity.
2.
The cold recirculation flow test, to check the new thermal shield support s at m,is described under " Summary of Operations."
This testing involved
, the recirculation pumps. The reactor was not taken critical and no control rod movement was performed. Careful sur-veillance of special instrumentation was maintained at all times. The periods of vibration were held to the absolute minimum while still ob-taining meaningful data.
VI.
PERIODIC TESTING PERF0FIED 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 Specifications :
System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives Continuous withdrawal and insertion Each major refueling 3-15-65 of each drive over its stroke with and at least quarterly 6-3-65 normal hydraulic system pressure.
during periode of power Minimum withdrawal time shall be 23 operation.
ceconds.
Withdrawal of each drive, stopping Each major refueling 3-15-65 at each locking position to check and at least quarterly 6-3-65 latching and unlatching operations during periods of power and the functioning of the position operation.
indication system.
Scram of each drive from the fully Each major refueling 3-1h-65 withdrawn position. Maximum scram and at least quarterly 6-2-65 time from system trip to 90,ercent during periods of power
(
of insertion shall not exceed 2 5 operation.
seconds.
4 A-9
- l
. System' or Function Frequency of Dates-Undergoing-Test
-Routine Tests-Tested
< Control Rod Drives (Contd)
Insertion of ~each drive 'overf its Each major refueling.
3-15-65 entire scope with reduced hydraulic but'not less frequently 6-2-65
+
1 system pressure to. determine that than once a year.
drive friction is normal.
- Control Rod Interlocks Rod withdrawalIblocked'when any
_ Each major refueling 3-15-65
-two accumulators are at a pressure but not less frequently 6-2-65
~
below 700-psig.
than once a. year.
Rod withdrawal blocked when two of Each major refueling 3-15-65
.three power range channels read but not less frequently 6-3-65 i
below 5%.on 0 '125% scales (or than once a year.
. below 2% on their. 0 -- h0%_ scales) j.
when reactor power is above the 4
minimum operating range of these
]
channels.
Rod withdrawal blocked when s cram Each major refueling 3-15-65 3
. dump tank is bypassed.
but not less freq.uently 6-2-65 L
tnan once a year.
Rod withdrawal blocked when mode-Each major refueling 6-2-65 selector switch is in shutdown but not less frequently position.
than once a year.
Other 1
Reactor safety system scram At each major refuel-3-16-65 i
circuits requiring plant shut-ing shutdown but not down to check.
less frequently than i
once a year.
t Liquid poison system component Two months or less.
2-12-65 check.
h-12-65 Test firing or explosive valves.
One year or less.
5-21-65 Post-incident' spray system At each major refuel-10- 1-64 i
automatic control operation.
ing shutdown but not 5-28-65 less frequently than once a year.
j
. Core spray system trip circuits.
At each major refuel-3-15-65
' - (-
ing shutdown but not i
less frequently than once a year.
4 2------
..m
,,... - -<~
g.
,_.r
-p
10 System or Function Frequency of Dates Undergoing Test Routine Tests Tested Other-(Contd)
Emergency condenser trip circuits.
At each major refuel-(Not ing shutdown but not.
~ tested less frequently than during' once a year.
this period.)
Cor.tainment.
Containment. sphere access air-Six months or less.
2-9-65 locks and vent valves, leakage rate.
Isolation valve operability and At least every twelve 3-25-65 leak tests.
months.
'I nlation valve controls and Approximately quarterly.
3-15-65 instrumentation-tests.
(Except main steam and main steam bypass-valves which
(-
have been closed thrcughout the period.)
Penetration inspection.
At least every twelve 4-21-65 months.
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 calibration; stack gas monitor calibration; emergency condenser vent monitor calibration; process monitor ~ calibration; and the-area monitoring system calibrations.
VII. C_0NTAINMENT LEAK RATE TEST The containment leak rate test, as required by Section 3 7 of the Technical Specifications, was performed during the period of April 13, 1964 through April 18, 1964. Results of this test were inadvertently left out of.the' previous semiannual report.
The testing was conducted in accordance with the procedures
.specified in Section 3 7 The leakage test was performed at a pressure of 10 psig ar.d held for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Tne Technical Specifications also specified that a controlled leakage test be conducted. This test was
^
11 3
- performed with a. leak _ off of 1040 ft / day- (or o.066% leakage per day) with the leakage measured by an integrating gas flow meter.
The reference system method was employed (using the same reference system that had been installed by Chicago Bridge and Iron Company-and Bechtel. Corporation for the initial leak rate tests).
During the 10 psi _ hold test, there were vide and rapid temperature fluctuations and no periods of stable conditions.w( h experienced. Temperature measure-ments taken inside and outside of the reference chambers showed that the reference system.as installed was unable to cope with the rapid changes.
The test results, therefore, were not as precise as would be desired.
The leakage rate was calculated using three different methods:
(a) Readings were averaged for two periods of time during which the manometer readings were decreasing at a maximum rate; (b) end points were established to include all meaningful data and the maximum and minimum rea(.ing during this period used; and, (c) a least squares fit of all mer.ningful data was obtained and the leakage rate determined from the slope of the line.
Results from the three calculations yielded leakage rates of: (a) 0.103% per' day; (b) 0.099% per day; and, (c) 0.080% per day. The allowable leakage rate at 10 psi, calculated in accordance with 3 7(g) of the Technical Specifications, is 0.125% per day. We have concluded that the leakage rate was within the specified limits.
The controlled leakage test was held for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, during which there was one period of 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> where conditions were very stable allowing establishment of the slope of the leakage line.
Results of calculations using these data yielded a leakage rate of 0.169% per day, which is in reasonable agreement.
By Robert L. Haueter (Signed)
Robert L. Haueter Assistant Electric Production Superintendent - Nuclear Consumers Power Company Jackson, Michigan Date: June 25, 1965 Sworn and subscribed to before me this 25th day of June 1965 (SEAL)
Grace Warner (Signed)
Notary Public, Jackson County, Michigan My commission expires February 16, 1968
p-.
o o
s
\\.
s s
f N
- f.,# P o
,y m r.L21a" p
g, L;:'
)
,/
j!
)
Y, t
I
,r
/*
"t e
i
'e
^
l f,
7 e.
y e'
o t
9 L
[
b
~
t.' d ;
f-9
- 7. n
{
ss i'.~-f__;F%
T O
D N
E j
t:
s
,,j
't 4
4
- t.. >.,
f
- 7, 'l
'[
,)
ij t
.'*. x j'
?
~
f T
6
)
b
.9'
..a a. a...
+
p
., f tA
\\
A r
1 1
P00R ORIGINAL s
.f I l..
s.
)* h l.I I l 5
[
(;
=
. =. -
m_ >..
.e.4
- - - ~ - ~ - -
~
~ - ~
M Q
1 O
(~I 11 MI'ORARY ACCELL ROMETERS ON BOTTOM OF THERN.AL SHIELD I
(
n-
{
~ -
,,,m
. _. ~..
^*.
t.
~ M
\\
.]
N EWLY IN D '.'.E D i
[.
>'-[
/
,j k
~ '
THERMAL SHIELD
[.
- j M,f-j,. /,.
.c.,
SUPPORT STILTS
-5/,ar, 1
L t.,
q., ',W...g a(
i ru
/: 3 TEM PO RA RY i-
,A f ;t'-
'p
~.
1.h,.w ;l '. '4.#
(s
- 1. j A
ACCELEROMETERS
,s
?
U
+
ON INLET FLOW g
w' b
s.
s m
i,
.f oirFUSER
{,
jg j '.j g,
.x
\\
, :s h
b~
'j 4, M.\\;;, g V$.,
jy' 5(#- m.n
. A, s
_..,,_, T vf. 4 %-
s
.4 (4 w' '.
.]
?
,, ny.,e 7
[4 y,
3
%k m
- /
e d.,
pp n
J l
CORE SUPPORT
\\
- ,i-s
~
g, f
PLATE 2
p
.~/-
y
\\
8 m
g.
s CD D'
i O
c::3 P
- c3 S
I m
s m
h
- 23 6
e
. :E ^:.:.,:,.
m l
p (VM
-l
-<7,,
l Z
i.. - _.. _ _ '
- f..
...,_2.
.._..m. -
._~... S. A-i i
PF
]
VIEW 0F COMPLETED STILT INSTALLATION t
f / l's is I Its sM lit I lit i INI e I'l A l l ( D R M f iliu st it all it s t r t-l t il W All ig
[
1-
O
~'~
3
~'
e p
c-N. e.
., N',
,]
t,"
.I%
/
s-J.
r.
.i, t,,
-1
'y
- )
4 *
'i,
. L, t
- f,
,t
/,. >.
')
l i.
"4 e e1
,e e....
s
[ [,
jl'\\ 1
} _,,
c'-'.
't
's
.p
.i s
s ge,b. 3 -
t l
e q.2 n
. 'f; I.
4 l. },, I-e i
.. t t
..i su i..
,f g7 ', c
't
, g (e
{
i l i
)
ir if l. l
- 1 J
.I e
4 s
h d,
j, Ig a
j i
t
' 8,,. J f
Ji g j
j i:
(
f q
.t s,
l e
t t
- ) )
t
, ~ *,
I
}
i 3 -
)
.,t
(;, s
[
. -.. :i
- 1 4
0 4
7 Lg '
I 3.
., y t
.', 0 -
'gf
' gl s
s
,4
. ' -, 's.
6
- l' p',pi n k
.P.
.' s a
i f
4
+
.4 i
1, 4
s t
t
,p e
a
>i
'n' t
- i s
1-d *i 1,
{
4 e-u I 3
b 4
't t
s~
L' 2
' i
(
y,
?
- If i
U
',_d.
g i' j
i f' : I
! ' l'l g s i
$l I
i.' e. j
,i 4
f
{ I, $-, $,
I,l.
j i
4, y
', Il I
{.
6 ' {l ;
a 1,
1;1 t,
af. P'
.,[
i
' i t
r 9;
g;
.i
. 1 0 f
sg
- j t
{ i,p ) y -
- i s
t.
4 4 s
t
\\
2 1
1
!'tI l ;
l[3 i
4 g.
n
- gja, r,c
-wer.=
i i
r 4
iil ti x
(
4
[8 i
1. )
/
l 6 1, ni j
i O',.
, t 4
,et 4
4
.a a
[j'. - {.
j' - { ; ',
I;
.i
- g
,.t g
46 j
4 1
I I,
g.
i
.o H Y..
I 8
p".,,."'
t
., l, i v 3
,i...,
' ' ' - ~.v/ 18 7
,s i I L L
- w L.4
-c '.J bei J he s=s
. m.e a' b
-/.i,
l =.,= * ** * ** "'
'd w "
)
,'t...,.
~I eh.n.l y'
m s-s-1
.-'t.
- r. g G<.
/v. s
+-
g
.i f s
r -.
6 3
(
- 1
. H,I ', 4 1
l'..
g ' tJ M g j' - { ' '~~
' !}
_~_. #
~e fg..{q f ',p' \\
g f*
.w ?
Y ;..,., 'i s' < h; ' ',. g,r.
p.-.
.).
gj y
.(.
n as....-,,
- g. V - 3 3 '.'~
- kQ} -
,. b.;
- Q s-~ i '. h ' i U < b, _
y
.. Q...h,..,' W[
-- ~
-~.m y,g,~ m f.*
I L.,1
,{
ei.
~ a:,
f, n,"
g' y,;, (%('i/.Jll/
,.:.n f }
I, y *.. k.. e.. g f,
c c-TLJ-
)-
l
,,.,,-- @b( %%[pi g
-4;-
p.
j
. [-
}-
}
h h e,
~.
_7-
)3,,. ~
. {/,
<8-r->
4 L., t.
.; 8
,/ 4 ; -A 't
p
., e s
1, - *4 t y t j< s
. 'i t
t
_ "4
,,...I,,,._.--,,
r;
.*7*r g.w.
.. g/..
- 1*,s f
- l. ' [ b M~ y.'1 : * ' 'j
,d
,,7 ) l' j
g
~
Y N K'pg ?l - ;
a rm
- m. j.
. t
) ~ G..,C,,, ' j
-d t
, 2...
.-C
-Lt gI
%'l"?.
, l,.."3, o
p. - Z :,g. w-as.
"ld ' """'"'j i ty' w
,g.
Y^. ~..
a, g,uk n P,
f p
t
,.,; o t + < di
- m.
Q;3 g
g pg
.- s.. c g
i i
f,,,,*
c,.s,
i
-t,..~.. ~.,
1 s
er-k.,w
., ~
g'
-,a.,a...
y.
i w
-. _., w.
%., d. h
,"*[
", +
4 I
P00R '0 RENAL
~,
e
~
TilERMAL SillELD SICAL RING STA111LIZER (One of Twelve Assemblies - Shown Attacheil to Instalbtion.it.c' Pleure 1II g,.
w -.
,