ML20087A786
| ML20087A786 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/09/1970 |
| From: | Caphton D, Robert Carlson US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20086U000 | List:
|
| References | |
| FOIA-95-36 50-219-70-02, 50-219-70-2, NUDOCS 9508070171 | |
| Download: ML20087A786 (13) | |
Text
,_
(.;.
' 'k a
', ' ? -
U. S. ATOMIC ENERGY COMMISSION P2GION I DIVISION OF COMPLIANCE Report of Inspection CO Report No. 219/70-2 Licensee:
JERSEY CENTRAL POWER & LIGHT COMPANY l
OYSTER CREEK 1 License No. DPR-16 Category C l
Date of Inspection:
February 5, 1970 Date of Previous is ti
- p nyary 6-8, 1970 W 9b N/hhh Inspected by:
D. L. Capht'on, Reactor Inspector D' ate
!N Reviewed by:
R. T. Carlson, Senior Reactor Inspector Date Proprietary Information:
None SCOPE Type of Facility:
Boiling Water Reactor Power Level:
1600 Mwt Location:
Lacey Township, Ocean Country, N.J.
Type of inspection:
Special, Unannounced, Inspection Accompanying Personnel:
None Scope of Inspection:
To determine details of MSIV (main steam isolation valve) test and review results.
To determine the nature of any repair work completed or being per-formed on the MSIVs.
9500070171 950227 PDR FOIA DEKOK95-36 PDR
3,,
t a
- Safety Items - MSIV leakage for all OC-1 valves (4 total) was determined to be above technical specification limits.
This fact negates the primary containment for the reactor facility.
It appears that the facility had been operated with this condition.*
(Paragraph K)
!I Noncompliance ' Items - Failure to report promptly, per paragraph 3.C. (1) of license, that the MSIV leakage rates were in substantial variance with the technical specifications and therefore may have prevented this containment system from performing its intended function as described in the technical specifications paragraph 4.5.
Status of Previously Reported Items - To be reported in the next routine inspection.
Other Significant Items - None Management Interview - Neither Mr. McCluskey nor Mr. Hetrick were at the facility at the conclusion of this visit; therefore, the inspector discussed with Mr. Ross, the ranking staff member at the facility, specific parts of the license ** and technical specifica-tions*" pertaining to reporting the fact that the MSIV leakage appeared to be in excess of the technical specification limit.
Mr. Ross acknowledged to the inspector that the individual valve leakage was in excess of the technical specification limit ** ** and that a report to the AEC appeared to be required.
He also stated that he had just talked with Mr. McCluskey by telephone and had received instructions to start work on the report.
The inspector observed Mr. Ross to note the pertinent technical specification and license reference's discussed by the inspector.
The inspector stated that the matter appeared to be an abnormal occurrence.*****
The inspector also stated that review by the Plant Operations Review Committee appeared to be required.******
- Reference to letter dated December 24, 1969, George H. Ritter j
to Peter A. Morris regarding MSIV.
- License Paragraph 3.C. (1).
- Technical Specifications Paragraph 6.3.
- Technical Specifications Paragraph 4.5F.d.
- Technical Specification Paragraph 1.15 defines " abnormal occurrence".
- Technical Specifications Paragraph 6.3A.
C..
J e
(,
y 4 The inspector requested Mr. Ross to notify the inspector in time to witness the final test of the MSIV's.
Mr. Ross stated that he had been instructed to do so by Mr. McCluskey, and would provide notice.
s Mr. Ross reflected to the inspector some confidence that the
[
repairs and changes being effected on the MSIV's would solve the leakage problem.
Mr. Ross stated that it was his opinion that the current MSIV leakage problem could be correlated with the large number of tests made on the valves during the startup test program.
He further stated that with fewer tests scheduled from now on, he did not believe that Obe MSIV leakage problems would continue.
Mr. Ross stated that no plans were made for additional leak rate tests of the valves when the reactor was started up again other than meeting the requirements of the technical specifications.*
The day after the inspection, February 6, 1970, the inspector telephoned the facility for the purpose of insuring that Mr.
McCluskey had fully understood the matters discussed with Mr. Ross.
Mr. McCluskey was stated to be out, however, he returned the inspector's call some three hours later.
The inspector stated to Mr. McCluskey that it appeared that the MSIV leakage appeared to be reportable under item 3.C. (1) of the license.
The inspector further stated that it was the inspector's opinion that the primary containm nt had been demonstrated to have been negated as of the last May, February 1,1970.
Mr. McCluskey argued that the MSIV test run was inconclusive and preltminary.
The inspector re-stated his position.
Mr. McCluskey stated that a report would be submitted.
The inspector told Mr. McCluskey that people at AEC Headquarters were very concerned regarding the MSIV matter.
DETAILS A.
Personnel Contacted:
Personnel contacted during the visit included the following:
T. McCluskey, Station Superintendent D. Ross, Technical Supervisor J.
Sullivan, Associate Engineer F.
Kossate, Maintenance Foreman
- See Paragraph 4.5D.
77 a
(
. B.
Administration and Organization i
{
l.
General Mr. Ross Etated that JC personnel were performing all testing, inspection and maintenance work on the MSIV's.
j Mr. Ross stated that Mr. Boyd Brooks, GE valve expert from i
San Jose, had spent several days at OC-1 providing assistance with the valve problem.
Mr. Ross stated Mr. Brooks arrived on February 2, 1970, and departed from OC-1 on or about February 4, 1970.
Mr. Ross stated that the MSIV manufacturers' representative, Mr. John Festa of Atwood Morrill, was also at the site to provide consultation concerning the valves from January 31, through February 4, 1970.
2.
Plant Operations Review Committee (PORC)
The inspector asked Mr. Ross if PORC had met
- to review the abnormal occurrence ** of the MSIV leakage.
He stated "No."
He further stated that they (OC-1 staff) considered that the whole matter was still under investigation.
C.
Operations The reactor was determined to have been shutdown and was still shutdown for an extended maintenance outage scheduled to take from one to two weeks.,
Prior to the shutdown, the reactor had been operating at full power.
A check of the operations log determined that the JC system had authorized release of the reactor for the maintenance outage at 2033 hours0.0235 days <br />0.565 hours <br />0.00336 weeks <br />7.735565e-4 months <br /> on January 31, 1970.
The power reduction was initiated at 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br /> on January 31, 1970.
The generator was taken off line at 0054 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> and the reactor was manually scrammed at 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on February 1, 1970.
K.
Containment 1.
MSIV Inspection Upon arrival at the site the inspector proceded to inspect
- Required by Technical Specifications 6.3A.
- Defined in Technical Specification Paragraph 1.15.
~
C
)
. the work in progress on the MSIV's.*
Mr. Ross accompanied the inspector.
The outer valve (NSO4B) in the south main steam line was observed to be disassembled and the work of lapping the valve's seat was in progress.
The maintenance man i
performing the work stated that he was using a 250 grit size lapping compound.
The maintenance foreman, Mr. Fred Kossate, also at the work stie, stated that they varied the grit size depending upon the amount of metal to be removed.
Mr. Kossate stated that no measurements were made as to the amount of metal to be removed, he stated that this was determined by visual observation of the lapping compound distributed around the seating surface.
Estimates given as to the amount of time to complete the lapping work on this valve's seat ranged from 6 to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
Mr. Kossate stated that periodically they would remove the lapping tool to the shop to check and refinish the tool as necessary.
The inspector visually examined the valve's seating surface during a time the tool was removed.
The seating surface had been wiped clean.
No pits or other irregularities were visually observed.
The inside of the valve appeared clean.
The seating surface of the plug on valve NSO4B appeared to be good.
No pits, groves or other imperfections that would be expected to contribute to leakage was visually detectable by the inspector.
Mr. Kossate stated that no work had been done on the plug.
The seat' ng surfaces of the pilot valve's plug and seat i
were also examined.
No pits, groves or other significant imperfections were observed.
The inspector observed [a whitish discoloration on the NSO4B valve body studsi located at.the bonnet flange.
- See Attachment I for layout of valves.
/.
~u'
<r
% bavarme,
.a-.
4m 6
- mamer4 MA.
-J.
s.
/ l l
This discoloration indicated to the inspector the l.
possibility that steam leakage may have occurred at the bonnet flange.
Mr. Ross stated that while '
conducting the leak rate test, air leakage had been
+
detected at the bonnet flanges of both-of the outer i
MSIV valves.
Inspection of the bonnet and body -
gasket surfaces determined some pitting and other minor imperfections in the surfaces, however, none appeared to the inspector to be of significant magnitude as to cause a leakage problem.
Mr. Kossate stated that a flexitallic gasket was used at the j
Mr. Kossate stated that they had found the south inner.
valve (NSO3B) seat to have been distorted.
He stated that the lapping work just completed had corrected the problem on that valve.
1 i
[
f
=
s=
2.
Test Equipment a.
Air Supply to Reactor i
The inspector, accompanied by Mr. Ross, inspected the t
setup on elevation - 19'6" used to supply air to pressure the reactor for testing the inner MSIVs.
The supply air was obtained from a valve with a hose connection off oi an ~1-1/2 inch building air supply line.* The hose (es3/4 g
inch 1.D.) ran to a portable flow measuring and regulating apparatus equipped with a e-3/4 inch bypass valve.
Mr. Ross stated that the bypass valve was opened for all their testing and the regulator and flow meter had been valved out.
Another hose ran from the portable flow measuring and regulating setup to an ~3/4 inch valve that was tagged as No. V-31-3.
Mr. Ross stated that this valve was a part of the reactor head cooling system.
The inspector subse-quently verified this by reference to GE drawing 886D403.
Mr. Ross stated that the air supply to the reactor entered the reactor through the reactor head cooling system.
Mr. Ross stated that the piping run to the reactor vessel contained several hundred feet of piping.
He stated that the actual air flow capability into the reactor was un-known.
b.
Leakage Measuring Apparatus The inspector observed that each of the two main steam lines in the tunnion room had identical flow measuring apparatuses."
Mr. Ross reviewed the purpose ' of the individual equipment with the inspector.
The equipment consisted of piping, valves, hose and fittings for connecting to the main steam lines.
The connection location was at the outer MSIV in the trunnicn room and the connection point was between the inner and outer MSIVs.
The flow measuring equipment con-sisted, in series, of a polydhylene 5 gallon bottle, arranged to serve as a bubbler, and a Rockwell flow meter.
A pressure gauge was also connected on the MSIV side of the bubbler. (See Attachment II).
Mr. Ross stated that the j
flow equipment, the bubbler and the Rockwell flow meter were used to determine leakage through the inner valves.
- This supply has a maximum pressure of 115 psig and is supplied by two air compressors.
- See Attachment II for leakage measuring apparatus.
j
W.
. J:
+ s, ~
]
~ L The pressure gauge was stated to be used to obtain data on pressure decay wh,en leak testing the outer valve.
The j
inspector noted that the two pressure gauges had'cali-bration tags ' dated November 1969 on one and January 1970 on the other.
Mr. Ross stated that both of the Rockwell flow meters had been calibrated on January 30, 1970.
3 l
A manometer was connected to the turbine side of the outer MSIV to determine that there was no back pressure in the line when the tests were being conducted.
Mr. Ross stated tha t the reactor pressure during testing was determined'from permanently installed equipment and i
was monitored in the reactor control room.
l 4
- 3. MSIV Test (All information and test data wac supplied by
,Mr. Ross).
l a.
Initial Test of both inner MSIV's on February 1, 1970 The primary system temperature was 1400 F.
for this teEt; the reactor had just been shut down.
All MSIV's were closed.
Air pressure was supplied to the reactor through the reactor vessel head cooling piping (reference to paragraph K. 2.a. ).1 Mr. Ross stated that the. maximum.
air pressure that could be and was obtained in the reactor was 2.5 psig* with the available air supply.
A pressure buildup of 1.0 psig occurred during the test between the two isolation valves in both main steam lines.
Both steam line flow meters indicated 18 cfh leakage ** through both of the inner valves.
Mr. Ross stated that due to the flow restriction in the bubblers, the maximum flow that could be obtained through the flow measuring apparatus was 18 cfh.
The inspector stated that the leakage was probably greater than the 18 cfh.
Mr. Ross stated "probably".
Mr. Ross stated that the valves had been closed in the required manner for the test and had not been cycled.
- Technical Specifications specify 20 psig as the test pressure.
- Technical Specification limit is 11.5 cfh leakage through any one MSIV at a test pressure of 20 psig.
G.
s h
~
. Mr. Ross stated that since the valves were leaking so much such that the reactor could not be pressurized,*
L the prepared test program plan was of no.use.
Mr. Ross stated that the intended plan had been reviewed and approved by PORC.
1 l
Mr. Ross stated that another improvised test was tried j
on both lines as follows:
1 A measured air flow input was piped into the space 1
between the inner and outer MSIVs with the reactor i
pressure at zero.
l North line (A valves)
Air flow in was 55 cfh, the pressure built up to 4.6 psig maximum.
When the supply was cut off, Mr. Ross stated that pressure held for j
several minutes before decaying to zero.
1 l
South line (B valves)
Air flow in was 240 cfh and no pressure build up took place;'when the supply was cut off the pressure went immediately to zero.
(No'e - location of pressure gauge is in line used to
]
t supply air, see Attachment II).
i b.
Test on North line (A valves).** February 2, 1970 The steam line between the two north valves was pressured to slightly above 20 psig air pressure and the air supply was valved off.
There was no water leg on the upstream side of the inboard valve for this test.
Pressure readings were taken every five minutes for 30 minutes and the pressure decay rate was determined from this curve for
?"
ysig.
(i.e.,
tangent at 20 psig was used to calculate rate)
The leakage rate was calculated to be 20.8 cfh.
February 3, 1970 Same test as above except data was taken over a 55 minute time period and the pressure decay ranged from 21.5 to 16.8.
The 3 eakage was calculated to be 21.8 cfh at 20 psig.
- Mr. Ross stated that there was no other known leaks from the reactor I
other than the MSIVs.
- Mr.
Ross stated that testing of A valves continued while work was begun on the inboard B valve NSO3B.
The A valves were cycled several times.
}..
Ll 0
J.
r The hangers.for the valve operators were removed
- from both A valves and the same test was rerun, except the:
test time was for 30 minutes.
The pressure decay was from 20.8 to 17.7 psig,the leakage rate at 20 psig was calculated j
to be 17. 5 cfh.
- c. Test Following B inner valve Repair February 4, 1970 1
I Upon completion of the repair of NSO3B valve, kith all' MSIVs closed, the reactor was pressured to 20 psig air i
pressure via the reactor head cooling piping connection.
Mr. Ross stated that the MSIVs had been closed normally i
and that there was no water in'the main steam lines, valve hangers were disconnected.
Mr. Ross stated that the flow meters (see Attachment II) did not indicate any flow through either inner MSIVs during this test. He stated over a 15 minute time period i
with the reactor at 20 psig, there were no bubbles de-tected at the south, B line. bubbler.
He stated that one bubble was detected at the north, A line bubbler.
Mr. Ross stated that this test verified the capability of both in-board valves NSO3A and NSO3B to meet the 11.5 cfh leakage requirement.
d.
Test of outer valves February 4, 1970 f
f Both main steam lines were flooded with water for this test of the outer valves.
The purpose of the water is to insure that no leakage could pass the inner valves while testing the outer valves.
4 The pipe between the two valves of both lines were drained and then pressurized with air.
North, A Line - The test period was one hour and the pressure decay was from 24.1 to 16.8 psig.
The calculated leakage rate at 20 psig was 23 cfh at valve NSO4A.
Mr. Ross stated that some air could be heard leaking out of the valve bonnet flange into the tnnvion room (which is outside of j
-I
- Mr. Ross stated tha t the hanger on valve NSO3B was very tight and was believed adding considerable stress to the valve.
i
)
1 I
_ _' 3., n
, _ _ 2 ~~- AL -
j.
. c;.
}.
H South, B Line - They were unable, per Mr.1Ross, to get any pressure rise with-theLavailable input of 240 cfh.
Mr. Ross stated that the leakage through valve NSO4B was therefore at lease.240 cfh.
Retest of North, A Line - Mr. Ross stated that a recheck was made to insure that water'had completely filled the.
{
A line and was against the inner valve. HThe leakage rate thru the outer valve NSO4A for this test was determined to be 17.6 cfh at 20 psig air pressure.
2nd Retest of North, A Line - Mr. Ross stated that tape was wrapped around the bonnet flange of valve NSO4A in order to stop air leakage out of this flange.
This test was re-run at 25.8 psig air pressure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 50 minutes.
The leak rate was determined to be 14.8 cfh through valve NSO4A.
- 4. Outer MSIV Support Both outer MSIVs, NSO4A and NSO4B, have a heavy channel steel support welded to the bottom of each valve.
Mr. Ross stated that it was decided to remove, via grinding, a weld between the valve and one end of the channel.- This was done so as to l
eliminate any possibility of 'dtstorting the valve from possible differences in temperature expansion between the channel and valve.
Mr. Ross stated-that when the weld on valve NSO4B was removed, relative motion occurred between the valve and channel thus further reinforcing this theory.
The'aimiliar channel weld on valve NSO4A had not been removed while the inspector was at the site, however it was JC's intention to remove this weld also.
The two inner valves do not have this support.
I I
I:-
, - y; ;
- 3
-e
/
CO REPORT NO.
219/70-2 MWlk Tyunnion bn1 b wy i
(inter vais) IG, 5 wd (ouL u\\ts) i l
l 24 s4k
/
b n) %.twc A
R SDBE R$o4b l
im3
- %9eh I
go Tub l24."Nos%
/
(
A
'Mawsbhw
/\\
M503A 11504 A i
1 l
kAN
\\llEW OF l'/S W L L
Ah6ern I
e.
', :2.
- /.*[*'
llo' **
- Y CENTRAL POWER & LIGHT Cr')
JE CO REPORT NO. 219/70-2
'Dvywell Twnion'l2 con r
.j-
+
q T
Nthi
~~
/
\\
M To lufbly
- --> \\
r n
nuyn IMW odet OS W vgeg.
3 essue xwn p
u,m if v
T
.A m
f a
/
Exbud
/-
Aw l
NXt$h
&bbb As %
ki b%
ydit hw\\j i
LEAKAGE MEAsV12lNC, APPARATUS (Tyrai 5m ben mm &. hnd ATEACWEhT IL
.e
3944 March 5, 1970 J. P. O'Reilly, Chief, Reactor Inspection & Enforcement Br.,
Division of Cornpliance, Headquarters INQUIRY MEMORAND1/M JERSEY CENTRAL POWER & LIGHT COMPANY, (OYSTER CREEK 1),.219/70-A PROPOS3D CHANGES IN STATION STAFF On March 3,1970, during a routine telecon with T. McCluskey, Station Superintendent, Region I learned of the planned transfer of both I. Finfrock, Maintenance Supervisor, and D. Hetrick, Operations Supervisor, from the plant to new assignments in the GPU offices in Parsippany, New Jersey.
Finfrock's move would be in the immediate future whereas Hetrick's would be in a couple months.
E. Riggle, Instrument and Electrical Foreman, would be promoted to Maintenance Supervisor vice Finfrock and J. Carroll, Shift Foreman, would be promoted to Operations Super-visor vice Hetrick.
Our concerns in this matter were communicated to McCluskey in a telecon held for that purpose on March 4, 1970.
We plan to review this matter further during our next inspection scheduled for the week of March 16, 1970.
You will be kept informed of future developments.
R. T. Carlson, Senior Reactor Inspector cc:
E. G. Case, DRS R. S. Boyd, DRL (2)
S. Levine, DRL D. J. Skovholt, DRL (2),
L. Kornblith, Jr., CO Regional Directors, CO REG File 4365e3togg
/
7 omct,
__COM,PLIANCE k,,
CARL maz su A
v-H IL 1
394'4 j
March 16, 1970 1
J. P. O'Reilly, Chief, Reactor Respection & Enforcement Br.,
Division of Compliance, Headquarters 1
INQUIRY )GEMORANDUM JERSEY CEbrfRAL POWER & LIGHT COMPANY (OYSTER CREEK 1), 219/70-B RESIGNATION OF NUCLEAR ENGINEER AND TARSINESS IN SUBMITTAL OF REQUIRED REPORTS.
During a routine telecon with T. McCluskey, Station Superintendent, on March 16, 1970, Region I learned of the resignation of C. Agan, Assistant Technical Engineer, effective March 20, 1970.
Mr. Agan is the last of the three GE-trained nuclear engineers originally assigned to the Oyster Creek staff.
Rer Mr. McCluskey, two replacement assistant technical engineers are currently undergoing BWR training at GE-San Jose and will report to the site on May 1, 1970.
I In response to inquiry, Mr. McCluskey stated that the license required reports on excessive main steam isolation valve leakage i
(10-day report due February 11, 1970) and reactor operations (first semi-annual report due January,1970) were still in preparation.
Similar responses were received on previous inquiries by Region I regarding these and other reports required by the
- license, j
The subjects of dilution of operating staff experience level and tardiness in submittal to DRL of required reports continue to be of concern to us, and rightfully belong on the agenda for the meeting with licensee management scheduled for March 25, 1970.
These matters will be pursued 82rther during the next inspection scheduled for March 18-20, 1970.
R. T. Carlson, Senior Reactor Inspector cc:
E. G. Case, DRS R. S. Boyd, DRL (2) j' S. Levine, DRL 49 7 m e-s -> - --
b k:s ff}
U s L/ A > >00 n_
y nnuwni e - nm.
orr4e.KO.rnblit.h.,,,.7.L, CO Region V CA
- mat sun m a s c..y i l e-
J
- .4.
u 3944 March 27, 1970 J. F. O'Reilly, Chief, Reactor Inspection & Enforcement Br.,
Division of Camp 11ancs, Headquarters INQUIRY M MORANDtM JERSEY GNERAL 70WER AMD LIGHT C(MFANT (0YSTER CRERK 1), 119/70-C OPENING W REACTOR HIGH PRESSIEtB INSTRIDENT LIM - LOSS OF SAFETY STSTEM RRDtBIDANCY The Station Superintendent, Tom McCluskey, telephoned the Region I office at 4:30 p.m. on March 25, 1970, and stated that at 11:35 a.m., a tubing connection to the reactor high pressure scram switch RE03C became disconnected while an instroent technician was working at the swithh. He stated the occurrence caused a loss of instrumentation redundancy relative to the limiting conditions i
for operation (Reference Technical Specifications Table 3.1.1.)
He stated that scram protection was not lost. He stated that a written report (in-house) was being prepared and that the Plant Operations Review Comunittee would convene either on March 25 or 26,1970, to review the matter.
Other pertinent facts stated were:
-1 s.
The reactor was not shutdown and continued operation at full power equilibrium conditions.
]
b.
The systems dispatcher was notified regarding intent to shutdown, however repairs were completed before the actual shutdown was begun.
c.
Redundancy was stated to have been lost on reactor protection instrumentation as follows:
High pressure scram switch Righ pressure emergency condenser switch Reactor vessel 600 pois permissive to bypass low condenser vacusan switch.
Tripple low level d.
Core spray valve permissive was stated to have been activated.
e.
The inefAent occurred while an instrument technician was performing surveillance testing.
f.
The excess flow check valve was stated to have closed immediately, and
{
1eakage from the reactor to secondary contaissent was stated to have been l
n.ths..
ur, met.a,mem*a che eka -
flow -Lek valve later omcc, c olLE k.LA.E C.3 A oar-a m
_0_ _. [
CAPAON:maz CARLhS o w w puu~
w.
.un,
n 3/27/70
/V mmn
i n reopened to reestablish the instrumentation to the reactor. The exact length of time the valve was closed was not given, however repairs were stated to have been completed in 15 minutes and everything back to normal within approx-imately 50 minutes.
(Note - When the excess flow check valve was closed, all of the instruneatation on the header with the reactor high pressure switch was effectively inoperable. There is no visible indication of the position of the excess flow check valve.)
l Upon inquiry, Mr. McCluskey stated that JC does not consider this occurrence reportable (written) to the AEC. Based on the information provided to date.
l Region 1 concurs in this respect. The assigned inspector plans to investigate l
the entire matter during the next regularly scheduled inspection.
l R. T. Carlson Senior Reactor Inspector cc E. G. Case, DRS R. S. Royd, DRL (2) i S. Levine, DRL D. J. Skovholt, DRL (2)
L. Kornblith,Jr., CO REG File
(
-