ML20031C434
| ML20031C434 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 08/23/1966 |
| From: | Robert Carlson US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20031C422 | List: |
| References | |
| 50-029-66-03, 50-29-66-3, NUDOCS 8110070141 | |
| Download: ML20031C434 (15) | |
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U. S. ATOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE August 23, 1966 CO REPORT NO. 29/66-3
Title:
YANKEE ATOMIC ELECTRIC COMPANY LICENSE NOS.
DPR-3 and SNM-906 Dates of Visits:
July 27 - 29, 1966 and August 4, 1966 By R. T. Carlson, Reactor Inspector 90 lis
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SUMMARY
r Mr. H. A. Autio, Shift Supervisor, was promoted to Assistant
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Chief Engineer effective July 1, 1966.
This and related
?T moves completes the personnel changes associated with the staffing of Connecticut Yankee.
The primary to secondary system leakage has increased from 6 to 16 gallons per day.
Sodium tracer tests are planned to determine if steam generators other than Unit No. 3, a known leaker, are involved.
I A problem with electrical storms, one of which caused the
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loss of one af the outside transmission lines and resulted l
in a reactor scram, is discussed in the report.
The results of the licensee 's test program on the pres-t surizer indicate possible stratification of the gases i
within the gaseous phase.
The licensee has proposed equipment modifications to permit a more effective in-
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vestigation of this condition.
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Y-A problem with control rod follower latching joints is discussed in the report.
The equipment design precludes f
(continued)
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5 Summary (continued) any safety implications; however, operational considerations have prompted the licensee to plan to replace all the con-trol rods and followers with new units employing welded joints.
Results differing from those previously experienced and reported were noted in the magnitude and rate of reactivity gain following the addition of ammonia for Core V stretchout operation.
3; 4D I The -status of the defects in the Core V fuel and the problem lek
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with the thermal barrier in the No. 2 main coolant pump L
remain essentially unchanged.
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. :. y The end of Core V operation is currently scheduled for early f;
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october 1966.
4 No items of noncompliance or of safety significance were noted or reported to the inspector.
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DETAILS I.
Scope of Visits Mr.
R. T. Carlson, Reactor Ins pector, Region I, Division 4
of Compliance, made an unannounced visit to the Yankee Atomic s
Electric Company (Yankee) reactor facility in Rowe, Massachusetts, on July 27 - 29, 1966.
Mr. Carlson made an announced visit to jf '
the facility on August 4, 1966, in conjunction with a visit i.j.
by Dr. R. L. Doan, Director, Division of Reactor Licensing.
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The July 27 - 29, 1966, visit included a tour of the facility Y.
and a review of the following subject areas (continued)
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- v. Scope of Visits (continued)
A.
Changes in licensee organization.
B.
Plant operation since the last visit, including significant problem areas.
C.
Control rods - the problem with the latching joint.
The principal persons contacted included:
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Yankee I
M ff j Mr. Lawrence E. Minnick, Vice President Mr. Wendell P. Johnson, Plant Superintendent g,w Mr. John B. Randazza, Chief Engineer g.pj#t b - ^
Mr. Herbert A. Autio, Assistant Chief Engineer
'; b ' 2 Mr. David A. Hanson, Jr., Technical Assistant to (f 'l.
Superintendent
}N Mr. G. Carl Andognini, Jr.,
Plant Reactor Engineer Mr. Louis H. Heider, Chemical Engineer
(( aj Mr. John Burnsee, Health Physicist
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Mr. David P.
Pike, Plant Health Physicist Mr. Edwin R. Taylor, Shift Supervisor yI Mr. Karl E. Jurentkuff, Jr., Control Room Operator Mr. Donald B. Vassar, Control Room Operator B.
Westinghouse Electric Corporation N
Mr. Benjamin James, Field Engineer, Irradiated
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~ :'i Fuel Shipments f.
II.
Results of Visits
.l A.
Organization g!
5 Mr. Herbert A. Autio, Shift Supervisor, was promoted i
to Assistant Chief Engineer effective July 1, 1966.
Mr.
Richard A. Herzog, Jr., Control Room Operator, was promoted (centinued)
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l l Resdits of Visits (continued) to Shift Supervisor vice Mr. Autio effective that date.
These moves complete the personnel changes relating to the staffing of the Connecticut Yankee facility, discussed in CO REPORT NO. 29/66-2, paragraph II.A.
According to Mr. Johnson, the Yankee facility is currently staffed with a total complement of 66 persons.
B.
Plant Operations 7g 1.
Previously Reported Problem Areas
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, Fdp4 The inspector reviewed the status of previously ff$f
,[jy reported problem areas.
Significant developments noted are discussed in the following paragraphs a.
Primary to Secondary System Leakage i
The existence of a primary to secondary i
system leak of 6 gallons per day, determined
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by the licensee to be in steam generator No.
3, had been reported previously and is dis-cussed most necently in CO REPORT NO. 29/66-2, paragraph II.B.5.
The results of a secondary o' '
system sample taken by the licensee just prior to the August 4, 1966, visit indicated that
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[ {4 the primary to secondary system leakage had increased from 6 to 12 gallons per day.
In-formation received from the licensee on August c2 15, 1966, indicates that leakage rates as high as 16 gallons per day have been measured.
l' The licensee has been unable to determine by
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normal means if the other three steam generators U
are involved.
The licensee plans to inject small amounts of sodium into the prima ry sys-tem to be used as a tracer in an attempt to
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. Results of Visits (continued) 1
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identify any new sources of leakage.
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i cording to Mr. :linnick, this method was suc-cessfully employed previously in locating a 10%
leak in a moisture separator.
j The licensee is continuing to dilute the
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secondary coolant as necessary to prevent 185 tritium leak-through from becoming a problem 7{p' of any significance.
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b.
Fuel Defects
- ri:9 The results of the licensee's analyses of
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recent main coolant samples indicate an ap-L
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parent stabilization in the concentration of
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I-131, and in the I-131/I-133 atomic ratio,
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at values slightly higher than previously 1
reported.
The facility records indicate that temporary increases in the values of
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these parameters and in the concentration t
of I-133 were noted following deboration of the main coolant system on June 22, 1966.
The licensee attributes the temporary in-l creases to changes in power distribution l
resulting from the insertion of control 7
l rods in compensation for the removal of l
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The results of representative samples ob-p tained by the inspector from the facility records, together with previously reported sample results, are shown below:
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- Discussed in CO REPORT NO. 29/66-2, paragraph II.B.l.
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\\_ Results of Visits (continued) 5/20/66 6/17/66 6/24/66 7/22/66 I-131, uc/ml 4.0 x 10-4 3.8 x 10-4 9.8 x 10 4.8 x 10~4
-4 I-133, uc/ml 1.9 x 10-3 1.6 x 10-3 2.9 x 10-3 1.6 x 10-3 I-131/I-133 (Atomic) 2.02 2.22 3.10 2.84 The inspector's review of the facility records and discussions with die licensee indicate that the fuel defects still have not contributed significantly to the e!!
radioactive wastes and effluents, nor have
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they resulted in any significant safety or operational problems.
c.
Main Coolant Circulating Pump No. 2-
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Bearing Temperature Problem
- a The facility records indicate that there have been no significant changes in the status of this problem.
The licensee has limited the f
maximum temperature of the pump bearings to s
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164 F by varying, as necessary, the flow rate j[ 1 3.
i of the cooling water.
Assuming no change f'$
for the worse, the licensee plans to continue I.
operation under these conditions until the Core V - VI outage, during which time any l }.h necessary inspections and repairs on the pump will be made.
T 2.
Electrical Storyp - Reactor Scram jt '
T The licensee experienced a series of severe electrical storms in the vicinity of the U
(continued)
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l O.i 3 review of the circumstances re-lating to the occurrence indicates that all systems functioned properly during the scram and that no problems of safety significance resulted.
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Pressurizer Test Program
- The licensee has repeated the no-capillary-vent test on the pressurizer.
The inspector noted that the test results, reported by the licensee in Operation Report No. 66, are in good agreement g.
with previous results with respect to the com-4
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position of the gaseous phase (normal) and to the volume of noncondensable gases (about one-fifth k[.h the amount expected).
However, the dissolved oxygen content in the pressurizer, previously 1p reported to be 2 - 3 ppm, could not be verified idh valve was subsequently repaired during the brief
' 'M;f due to a malfunction of a sampling valve.
The
'K shutdown on July 10, 1966, which was caused by the l{F scram previously discussed.
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The licensee has concluded, based on the results h.
of the no-capillary-vent tests, that stratification E
of the gases is being experienced in the gaseous phase in the pressurizer.
According to Mr. Helder, it is probable that the volume of noncondensable gases per unit of condensate is greater at the surface of the liquid phase than elsewhere.
This
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could, in his view, account for the noted differences.
Mr. Heider postulates further that the percentage of oxygen in the gaseous phase may be greater at the surface of the water than elsewhere.
This j,
could account for the relatively high dissolved I'
oxygen content in the liquid phase.
The licensee I
has requested authorization to install a sampling b
tap nearer to the surface of the liquid phase **
f to permit a more effective investigation of these
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conditions.
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- Discussed in CO REPORT NO. 29/66-2, paragraph II. B,6.
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- See proposed change to the Technical Specifications dated F.:
August 1, 1966.
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l Results of Visits (continued)
In discussing the licensee's inspection program for the pressurizer during the next refueling outage, Mr. Minnick told the inspector that ultrasonic tests of the base metal are being considered as part of the investigation of the pressurizer cladding defects.
This subject will be reviewed further during future inspection visits.
NY 4.
Core Reactivity Depletion bl$
'M TB The facility records indicate that the licensee deborated the main coolant system on June 22, 1966.
jy The end of Core V full power operation with unadjusted
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pH occurred on July 8, 1966.
The addition of am-i~*
monia for stretch-out operation was postponed until i
July 14, 1966, because of the scram on July 10,
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1966.
The inspector 's review of the facility records re-
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lating to the ammonia addition indicated that where previous experience showed that a gain of 0.7%
delta k/k should be expected, only 0.5% delta k/k f
was actually experienced.
Of the 0.5% delta k/k,
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j 0.25% delta k/k was realized initially at the rate previously experienced; however, the re-
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maining 0.25% delta k/k was realized as an ex-l l's
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tended flat spot in the core excess reactivity plot, lasting for about 2 weeks at essentially full power ope ration.
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The licensee was unable, at the time of the visit, to provide a complete explanation for the unusual f
nature of this most recent ammenia induced reactivity
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gain.
Mr. Johnson told the inspector that they j
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. Results of Visits - (continued) believed it to be related to the fact that relatively 1cw concentrations of crud have been experienced in the main coolant during Core V*.
The inspector's review of the licensee's chemistry records indicated that the main cool-and crud concentrations averaged about 0.4 ppm during the early stages of Core IV.
The crud pc concentrations have averaged 0.15 ppm during a P
similar period in Core V.
- [v Discussions with Messrs. Minnick and Johnson f?i indicate that they have no immediate concern i>
regarding the unusual nature of the reactivity b;
respense relating to the ammonia addition, a I
position shared by the inspector.
This is based I~
principally on the following:
The variances from previous experiences were in the conservative y
direction; the licensee was able to follow, and I-did follow, the reactivity change closely; the
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difference between the predicted and the actual reactivity gain, 0.2% delta k/k, was relatively small.
The licensee, in the interest of gaining further i
insight into the overall pH-induced reactivity
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phenomenon, asked Westinghouse to take additional b
measurements during the recent period of reactivity gain.
The results of the measurements were not available to the inspector at the time of the I
visits.
Any significant developments in this
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area will be reviewed by the inspector during future visits.
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(ccntinued)
- The subject of pH ver sus rea ctivity, including the role
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played by crpd, discussed it CO FEPORT NO. 29/65-2, paragraph
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! l Results of Visits (continued)
C.
Observations in Control Room As part of the facility tours, the inspector observed the indications on the control room instrumentation on several occasions during the visits.
All indications observed were noted to be within the requirements of the license.
Typical readings of pertinent parameters obtained by the inspector at 11:15 am (EDT) on July 29, 1966, are shown below:
O Parameters Reading
- S ;
Load, Mwe 175.0
'!U ew Condenser backpressure, inches Hg 3.18 Power level (from licensee curves using load and backpressure), Mwt 599.5 Power level (intermediate and pcuer range i
instrumentation - six channels), % full power 99 - 100 Primary system pressure - psig 2000 I
Main coolant average temperature (T avg.),
F 526.6 l
t Maximum outlet temperature (from in-core T/C's - fuel position C-8),
F 583 Condenser cooling water inlet temperature, oF 53 O
Condenser cooling water outlet temperature, F
73 Control red positions, inches withdrawn Group A 88-7/8 i;
Group B 89-5/8 Group C 89-5/8 Group D 89-5/8 t
(centinued)
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. Results of Visits (continued)
The portions of the control room panels relating to the safety injection system were visually examined.
All the system controls, including those for pumps and valves, were noted to be indicating operability and to be positioned in i
accordance with the requirements of the Technical Specifica-tions, paragraph B.1.
- Section 212.
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D.
Control Rods - Problem with Latching Joints C;
A problem with the latching joints between two con-trol rods and their followers, including the results of the Qt}
licensee's preliminary investigation of the problem, is dis-
!W cussed in CO REPORT NO. 29/65-4, paragraph II.B.4.
As was reported, both control rods and their followers were re-
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placed.
This work was done during the core IV - V outage.
The licensee subsequently completed a more detailed g
examination of the defective joints.
As a result of the j
findings, the licensee plans to request authorization to replace all 24 control rods and their followers with new units employing welded joints.
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The inspector reviewed the results of the licensee's l
l detailed examination of the defective latching joints.
In-cluded was an examination of pertinent wax impressions. The impressions showed that one side of the V-notch in one of the four vanes in each of the control rods was badly worn.
The wear was noted to be in the area where the follower spring fits into the V-notch *.
The facility records in-dicate that the springs were found to be rounded at the g_
top with slight curling of the metal.
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In discussing the significance of these findings,
[J Mr. Johnson reiterated his previously reported positicn, f
with which the inspector concurs, that the condition does I
(centinued)
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- s b Results of Vis its (continued) not represent a safety problem.
This is based en the fact that the control rods and followers must be rotated with relation to one another to effect disassembly, and the design precludes rotation within the assembled reactor vessel *.
According to Mr. Johnson, the licensee is going to a welded joint to avoid the possibility of inadvertent dis-assembly during refueling operations, and the resultant need for extensive recovery and repair operations.
The new rods gyy will be silver-indium-cadmium with inconel cladding.
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aM one problem associated with the welded joints is that f},
the assembled units will be too long to be moved through the gggQg fuel transfer chute.
Mr. Johnson stated that the welds b[ ".
would be cut prior to any transfer operations.
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Li E.
Miscellaneous
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1.
Irradiated Fuel Shipments - Cask Damage (SNM-906)
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The inspector's review of facility records re-
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lating to irradiated fuel shipments, showed
.y that the ten assembly shipping cask was damaged at Nuclear Fuel Services, Inc. (NFS),
i following the unloading of Shipment No. 8.
Discussions with Mr. Benjamin James, the Westinghouse field representative for the fuel I
shipments, who was at Yankee during the July 3,
27 - 29, 1966 visit, indicated that the damage was relatively minor and consisted of bent cask
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head guides and a jammed telescoping section of j
the primary coolant piping.
According to Mr. James, l
who said that he visited NFS in regard to the cc-c.
cu rrenc e, the damage resulted from having the
'i, head improperly oriented during its reinstallat ion.
(continued)
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The cask was subsequently repaired at Yankee.
Mr. James stated that there have been no problems with the cask since the occurrence.
Information received by CO:I from the licensee subsequent to the visits indicates that Shipment No.12 departed the site on August 17, 1966.
2.
Licensee Opera tion Report A1 53 The licensee's Monthly Operation Reports to the jfr Division of Reactor Licensing are routinely re-L7M viewed by the inspector.
Included in the review
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have been the reports covering the first six y '.
months of calendar year 1966.
In the inspector 's judgment, based on observations made at the site and documented in the inspection reports for this period, the Operation Reports provide an adequate j
summary of the significant aspects of plant opera-
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tions for the pariod covered, and are in compliance j
with the requirements of the license, paragraph 3.D. (2).
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3.
Core V - VI outage The licensee plans to conclude Core V operation on
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or about October 3,1966.
Discussions with Messrs.
f Johnson and Andognini indicated that Core VI rill g
'i be composed of the following-l 36 l:
t 4.94% enriched assemblies, new h
4.94% enriched assemblies, recycled 34 g,.
4.1% enriched assemblies, recycled 4
1 2.9% enriched zirconium clad h
2 f-assemblies, recycled i:
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The licensee plans to submit the necessary proposed change to the Technical Specifications relating to the Core VI loading during the latter part of August 1966.
Mr. Johnson estimates that the outage will require four weeks.
F.
Exit Inte rview s,;
x, te An exit interview was held with Mr. Johnson on July M;
29, 1966.
The principal subjects discussed were as folless:
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The primary to secondary system leakage.
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The recent ammonia addition a nd the nature of 2,v the resultant reactivity change.
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3.
The control red latching j oint problerr.,
f Significant ccmments by Mr. Johnson are included in the body of the report.
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