ML20086P032
| ML20086P032 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Nine Mile Point |
| Issue date: | 04/30/1982 |
| From: | Polk P Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20083L677 | List: |
| References | |
| FOIA-84-14 NUDOCS 8402240234 | |
| Download: ML20086P032 (42) | |
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NUCLEAR REGULATORY COMMISSION d. ';.
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April 30,1982 Docket No. 50-220 LICENSEE: Niagara 2 hawk Power Corporation
- FACILITY:. Nine Mile. Point Nuclear Station, Unit 1
SUBJECT:
SUBJECT OF APRIL 22, 1982 MEETING ON REPLACEMEllT OF SAFE-ENDS AT THE NINE MILE POINT NUCLEAR STATION, UNIT 1 A meeting was held on April 22, 1982 with Niagara Mohawk Power Corporation to discuss:
(1) The present status of the Nine Mile Plant. and (-2) The licensee's intentions regarding decontamination of the recirculation.
system, ALARA considerations, and removal and replacement of safe.. ends.
The meeting agenda and attendees are provided in Enclosures (1) and (2),
respectively.
Regarding the present status of the Nine Mile Plant, Niagara Mohawk indi-cated that the reactor has been defueled ar d that all fuel and control rods have been placed in the spent fuel pool.
In addition, all five recirculation loops have been decontaminated between loop isolation valves.
With respect to the repair effort 10 replacement safe-ends are on-site, some elbows are a'vailable, spool pieces have been ordered, and some of the custom shielding is available.
In essence, the licensee indicated the implementation of a 1979 contingency plan and estimated a one-year outage.
A mid-May commencement of drywell work is presently planned.
Niagara Mohawk presented a review of the Spring 1981-UT examinations on the affected safe-ends. No indications were found at that time, and a
-review of the records indicates that the 1981 tests were performed and evaluated correctly.
A crack growth ' analysis.was also presented by Niagara Mohawk. Seven areas were _ investigated as potential causes of failure. These areas are delineated in Enclosure (3) together with the disposition of each. The licensee was 'not able to single out one mechanism which could cause through wall cracks in less than a year, i.e., the only conclusion which could be
. drawn was that there were many contributing factors which could have caused rapid crack growth.
During the investigation of the seven potential causes of failure, the i
licensee decided to perfonn UT examinations on recirculation pump to piping welds. Niagara Mohawk indicated that these welds were tested due to un-certainty rel_ative to potential cracks in the~~entir'e" recirculation system.
All five pump discharge welds were inspected. Two of the five resulted in UT indications. The licensee indicated that dye penetrant and radiographic
-tests would be conducted in order to verify pump weld UT findings.
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I With respect to the accuracy of prior UT examinations and the identification of the failure mechanism, the staff advised the licensee that a conclusion must be reached in order to justify resuming plant operation. Although sensitized safe-end problems were known and replacement considered to be an acceptable solution, the pump weld indications necessitate a better understanding of the crack growth process.
Regrding decontamination, the licensee presented various applications of the London Nuclear, Inc., Candecon process. A decision had not-been reached regarding the extent of decontamination, i.e., whether to include the lower portion of the reactor vessel during recirculation system decontamination.
The licensee was advised by the staff that the details of the decontamination decision were required prior to beginning work. Further, we advised the
~ licensee that all possible measures should be taken to reduce radiation doses received by workers.
With respect to shielding, Niagara Mohawk proposed the use of custom sheilding in order to reduce direct doses originating from the reactor vessel.
(Other shielding alternatives were being investigated, but details were not available.)
The proposed design is shown in Enclosure. (4). Based on this design and on a recirculation system decentamination factor of 2, the licensee presented a worst case man-rem estimate for safe-end replacement of 3750 man-rem, as sumarized in Enclosure (5). The licensee was advised by the staff that a more accurate estimate would be required, as well as the details of how as-low-as-reasonably achievable (ALARA) criteria were to be satisfied.
In conclusion the licensee was advised by the staff that at least two sub-sequent meetings would be required. One meeting would be held in Bethesda to review the decontamination and ALARA areas. The other meeting would be held on-site in order to review welding procedures.
In consonance with the
'10 CFR 50.54(f) letter forwarded on April 21, 1982 details of the Nine Mile work should be submitted before significant and irreversible programs are undertaken.
9 Philip J. Polk, Project Manager Operating Reactors Branch #2 Division of Licensing
Enclosures:
'1. Meeting Agenda
- 2. Attendees
- 3. Overview Summary and 7 Areas of Investigation
- 4. Proposed Shielding Design
- 5. Worst Case Man-Rem Estimate cc w/ enclosures: All NRC Meeting Participants cc w/o enclosures: Standard Distribution List l
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Mr. Donald P. Dise
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Leonard M. Trosten Esq.
LeBoeuf. Lamb., Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Suite 1100 Washington, D. C.
20036 State University College at Oswego Penfield Library - Documents Oswego,-New York 13126 Resident Inspector c/o U.S. NRC P.O. Box 126 f
Lycoming, New York 13093,
Carl D. Hobelman, Esq.
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Suite 1100 Washington, D.C.
20036 Ronald C. Haynes Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 3
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SUMMARY
';0F, PRO,BLEM;.;G.J..Gresock
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A.
Two.throu..gh wall' cracks are in the.s,afe-end 1.
1S pump outlet 2..* 11 pump inlet
.Ind$ca'iodon#13inletsafe-end
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danse - Belie'ved to be' IGSCC.'
1.
Confirming by analysis at Battelle Columbus and GE.
Preliminary results indicate IGSCC.
2.
Evaluation of other potential causes have been eval-uated and no cther can be identified.
III.
PLANS FOR REPLACEMENT:
G.J. Gresock
()
A.
Plans began in 1978 (contingency) 1.
. Replacement specifications
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- 4.
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of all ten safe-ends B.
Material 1.
Safe-ends on site 2.
Some elbows available 3.
Spool pieces ordered 4.
Portions of custom shield material available.
Remain on order.
C.
Summary of sequence of outlet safe-end.
D.
Summary of inlet safe-end E.
Actual drywell work scheduled to begin in mid May F.
Potential to decon loops in two phases 1.
Prefer to discuss the details of decon later by London Nuclear
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Exams Performed - 1981 N'$
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Equipment
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Re-examination 2.
Technique Evaluation
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CRACK GROW'1E ANALYSIS:
.W. Schmidt - MPR
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. Summary of EPRI data
'.Br.!..:Cokr.islati.dn 'o..'f'<EPRI,'slata'. with NMPf.i' dat'a... '.
C.
Results VI.
DECONTAMINATION:
R. Hemings - London Nuclear A.
Description of process - generic B.
Description of process to be used in NMP#1 recire loops C.
Summary of experimental data which indicates no adverse effect from an IGSCC standpoint
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VII.
'ALARA A.
Decontamination:
T.W. Roman 1.
Analysis ' performed 'by NMPC of benefits of decen.
2.
Results B.
Custom shielding design:
P. Kasik - MPR 1.
Outlet safe-end (pump suction) a)
Shield curtain b)
Shield plug c)
No::le shielding (between no::le and bio shield) i 2.
Inlet safe-end (pump discharge) a)
No::le shielding (between no::le and bio shield) b)
No::le shield plug 3.
- Dose assumptions basis a
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.u.' ALARA (Continusd)
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Continual ALARA review:
R. Gallagher - NNI lN
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Description of shielding method a
3.
Personnel assigned
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PROCEDURES FOR REPLACEST.NT A.
Controlled work instructions:
R. Gallagher - NNI 1.
Outlet safe-end (pump suction) 2.
In1'et safe-end. (pump suction) -<
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1.
B.
Cutting procedure; R. Gallagher - NNI
- ,. J.,
C.J., Welding' piocedures;:.
T.. '311'1 man
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Qualification of welders 2.
Qualification of welding procedures a)
Stainless b)
Inconel c)
Heat sink weld (riser to elbow)
IX. --
OPE 4 DISCUSSION - ALL PARTIES v.i-
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LIST. F INVOLVED ORGANIZATIONS AND SPECIFIC INDIVIDUALS
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NIAGARA MOHAWK:
' T.J. Perkins C.V. Mangan T.W. Roman S.W. Wilczek, Jr.
G.J. Gresock
.J.J.: Clarke.',;
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NEWPORT. NEWS; R. Gallagher.
T. Gillman MPR:
P. Kasik*
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W. Schmidt NES:
G. Schmidt.
P. Berry T. Cook F. Martsen LONDON NUCLEAR:
B. Hemmings J. LeSurf 4
M' GE:
J. Gordon
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s-g ATTENDANCE LIST OF 4/22/82 Niagara Mohawk NRC Chuck Mangan Phil Polk Tom Roman Dom Vassallo Greg Gresock Gus Lainas Tom Perkins T. A. Ippolito Pete Barry, NES K. R. Wichman Karl F. Schmidt, NES Paul O'Connor William R. Schmidt, MPR Conrad McCracken P. M. Kasik, MPR P. R. Matthews Ron Gallagher, NNI James Wing Tom Gillman, NNI Horace Shaw J. Eric LeSurf, London Nuclear Frank C. Skopec R. L. Bob Hemmings, London Nuclear Joseph Halapatz Fred. Marsten, NES Warren S. Hazelton Ted Koch, NES Bernard Turovlin Jim Morris, GE Steve Hudson (Res. Insp.)
Robin L. Jones, EPRI Bill Koo G. D. Clarke, NNI M. L. Boyle M. R. Eshelman, NNI John G. Roberts, New
' G. M. Gordon, GE York State Public Service Comm.
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The purpose of my presentation is to (1) outline the problem and the I
cause, and (2) describe the general methods of replacement of safe ends.
4 My presentation is intended to be a broad overview, not a detailed tech-nical discussion of the cause or replacement procedure. The detailed tech-nical asp'ects of the cause or replacement methcdology will be covered in subsequent presentation.
y On March 19, 1982, Nine Mile Point Unit #1 was shut down due to in-creasing drywell. lea} cage. The cause of the excessive leakage was thought to be recirculation pump seal failure. Two pump seals were replaced, and the Reactor Building closed loop-cooling system was repaired. The plant was in the process of hydro testing on March 23, 1982, when leaks in No. 15 recirculation outlet and No. 11 recirculation inlet safe-ends were discovered.
l
' Figure 1 shows the location of the through-wall crack and the LTT indica-
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tions on No.11 pump discharge safe-end. Figure 2 shows the location of the through-wall crack on No. 15 pump suction safe-end (no complete UT exam-ination has been peifoiiiie~'d7However, the leak area was UT'd and a reflector i
identified).
Subsequently, on March 31, 1982, No. 13 pump discharge safe-end was ultrasonically examined. The results of that examination are shown in Figure 3.
There is one indication from approximately 12 o' clock to 3 o' clock, a
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which is a 100% DAC. There are two smaller indications between 7 o' clock n
and 9 o' clock. Since Niagara Mohawk has decided to replace all ten (10) f safe-ends, no further ultrasonic examinations of the safe-ends in place is
~
warranted.
(3 examinations and 3 indications)
The preliminary evaluation of samples at Bette11e Columbus indicate that the cracks are intergranular stress corrosion cracks (IGSCC). Prior to that determination, Niagara Mohawk investigated other potential causes of 4.
Table 1 lists these potential causes and the disposition of each.
failures.
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.Due to !!iagara Mohawk's uncertainty relative to potential cracks in s
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theentirerecirculation. system,itlwasdecided..toinspectthewelds between the piping and pump body discharges on all five (5) recircula-tion loops (pump suctions.were uninspectable),. Two, (2) of the five (5).
l inspections revealed a reflection when testing from the pipe s'ide.
No exam-
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ination could be niade from the pump body side, so that a geometric determination 1
i; could not be made.
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,/..., p.. 7.',. pump h'as been decont'aminated, di's~asse'mbled'and a dye-cenetran'.
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esam'was pe' formed. Inierm1tte'nt hairIine 'indicitions' were obs'erved irt the b'
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~ regionfrom7o'clockto9o' clock,one3/8inchindicationatklo' clock j
and a one inch indication at 12 o' clock at the longitudinal weld.
In an i[
attempt to verify this a radio graph is to be performed.
The locations of the indications on 13 pump are such that these may be explained by the weld root geometry. However, the pump is being decontam-Ij.
inated, and disassembled so that a dye-penetrant exam can be performed. This should be accomplished by the end of this week.
4 j.
. 2 c:.In t1978,, Niagara-Mohawk realized-the. potential of IGSCC and the furnace..
-4_
l sensitized safe-ends. 'Iherefore, we initiated a contingency project for safe-end replacement. This contingency project consisted of developing a replacement specification, drafts of necessary procedures and the procure-ment of certain items. A portion of this software must now be updated to reflect more efficient techniques, such as automatic welding. These re-
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visions are in the process at this time.
0 Briefly, this is the procedure for removing and installing recirculation h
3 obtlet safe-end. Figure 4 shows the installation of the shield curtain in 6
the annulus, and the external nozzle shielding. Figure 5 shows the location
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of the riser to elbow and safe-end cuts.
Figure 6 depicts the removal of i
j the elbow and the installation of the internal shielding.
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Figure 7 represents the removal of the internal shielding and the instal-lation of a. new e.lbow/s.afe-end ass,embly The proposed method of removing and replacing the recirculation inlet safe-ends will now be di.scussed Inlet. Figure 3 depicts the installation of external shielding.. Figure 9 gives the 1ccations of cut 1 and 2.
Figure ~10 shows the removal of the elbow and the installation of the shield plug.
Figure 11 I.
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represents removal of the spool piece and transit from the drywell.
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'shows $ie 'n'ew.s, pool pi' ce bein.g'~.instailed. '.'F.igurY 13 is~ the removal of' the. '.. ;...
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shield 'p' lug 'an'd fin'allf, Figure14"is 'th' new e-lbow in' stalled.
e The schedule for accomplishing the work previously described has not been finali:ed. Niagara Mohawk has estimated that the outage could last as long as one (1) year. Actual drywell work will not likely begin until mid-May.
1 Niagara Mohawk is investigating the possibility of decontaminating the l
recirculation loops prior to the actual work.
If performed, the decon would be performed in two phases. Phase 1 (see Figure 15) would be from the pump
. :, discharge. valve to. just.below,.tha.nos:le - (there is, a potential of putting a
.s,,,
8 seal plug in the annulus to decontamination through the no::le, which is still being reviewed). Phase II (see Figure 16) would be completed after all j
the pump suction safe-ends are replaced and the water level is lowered.
Niagara Mohawk has evaluated the ALARA and IGSCC concerns relative to decon. Tentatively, these evaluations have yielded a favorable benefit j
for deconning. However, the final determination has not been made.
Mr. Hemmings 1
will speak in more detail later concerning the process itself.
1
(.
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With this broad brush overview, I will now turn the podium over to f
Mr. Pete Barry of NES.
Mr. Barry will discuss the ISI procedures used in 1981 I
and the potential reasons for not detecting indications at that time.
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Potential Cause of Failure Disposition 1.
Startup Anomalities No appearance significant anomalities' 2.
Thermal Lockup of Vertical and Reading from the snubber LVDTs indicate i-Horizontal Snubbers proper snubber movement after startup I
in 1981 and during cooldown in March.
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'.,,.,..: ~ ;/.. ' 1.*of -1982., In.. addition,. UT of pump - dise -
r ch'arge ' elbows.
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3.
Chemical Cleanup During Last Review of London Nuclear Process reveals Outage that no adverse IGSCC effects would be expected. In addition, the area in the vicinity of the safe-ends was not cleaned.
i 4.
Proper Setting and Functioning Field review indicates no anomalities
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of the System Spring Hangers f.
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Reactor Coolant Chemistry No apparent intrusions which would cause i
a failure.
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Piping for Oyster Creek Nine Mile Point Unit #1 would be more susceptible to cracking.
7.
Parametric Study of Effects of An evaluation of the worst case vibra-J I-Pump Frequency, Pump Forcing tional frequency has been perfor:ned.
i Function Amplitude and Piping The resultant stresses attained from
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System Stiffness on Safe-end the evaluation do not approach the yield a
Nozzle Stresses stress of the material.
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A. GENERAL' N OTES :
- 1. AXIAL INDICATIONS ARE THOSE ORIENTED PERPENDICULAR TO THE WELD.
2.CIRCUMFERENTIAL INDICATIONS ARE THOSE ORIENTED r
PARALLEL TO THE WELD.
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- 3. ALL INDICATIONS DETECTED APPEARED TO ORIGINATE AT h
THE I D SURFACE.
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I B. EXAMINATIONS PERFORMED FROM SAFE END
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SIDE ( REFER TO. CLOCK):
dLEAK CRACKS WERE READILY DETECTABLE AXIAL INDICATIONS.
dCLUSTERS OF AXIAL INDICATIONS WERE DETECTED IN AREAS lt OF THE SAFE END OTHER THAN AT LEAK LOCATIONS. THESE 1
RANGED IN LENGTH FROM OVER I" TO BARELY DETECTABLE.
C. EXAMINATIONS FROM PIPE SIDE:
j dONE AXlAL INDICATION WAS ALSO DETECTED FROM THE PIPE M
SIDE SUGGESTING THAT A CRACK HAS POSSIBLY PROPAGATED h
INTO THE WELD FROM THE SAFE END.
ATHREE CIRCUMFERENTIAL INDICATIONS WERE DETECTED AT l'.
THE WELD ROOT EXTENDING FROM LOW AMPLITUDE ROOT GEOMETRIC.
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INDICATION ORIENTATION L
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duTLET Figure 4 Water Level.
Install Shi.e.ld.- Curtains pj, Nozzle. Shielding s.
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If decontamination is accomplished, it will be prior to this j
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Note:
Cut 1 approximate 1y' 2 inches below the elbow to b
riser weld. Cut 2 is to be on the safe end side i
of the inconel weld.
It i
i ll il p.
{
)
1 OUTLET Figure 6 Remove Elbow & Install Shielding
.y s
s 6-
+.
s N
g 1
i
.s g
y.
N s
s V
g essel A nnulus g.
N N
... =
0 Core gShroud; 4 g
' 5 N
C
?,
r Recirc' Outlet.
- .M. s l
Nozz..le
? *i
. 5 s
Reactor S.. ld ni e.
[. Wall.
.g.
4..
~
s y
.1
.Lcwer Corb Support g
PTake N
s
/
l s
yy
[
\\
t e ube s
Recire. Inlet Elbow Shielded N
=
i and Rigged Out-A C ozzle N
s of Drywell s
N 4
\\
3
, e-,
bi
.4 L
x
.i
-.a-
[
j s
xss i
(r l
7 i
N I
= =...
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I
/
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. _ = _
<
)
8
'f,3 e :.,.:
r l
..e
.o Cv s'">
J a
g 1
4
}
Note: After the elbow is removed, the riser and nozzle will j-be weld prepared..The ID and OD of the nozzle must be 4
etched to determine the inconnel to nozzle interface.
h Niagara Mohawk has specified that 3/16" of inconnel j
must remain so code requirements are not violated.
?
en s
w -
.e.--m<..-..
==e
.me-m.m.=um
=
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"-u'u==*
4-*
e.
'0UTLET i
1 Figure 7 Remove Inner Shielding
- f..
, Re.. place Elbow &. Safe End i
x s
s
,r p
s s
s s
s N
s s
s
,-Vessel A nnulus y
~~ \\
g a..,.
\\.
,g *
>j, Core' Shroud
, ')
Reactor Shield
.r e
).
'.4 7. s f
Wall Weld s
N Recire. Outlet
'A
. Lower Core Support j
Nozzle -
Dfa te N
s l
1
, Control Rod l
Water Level.
3 g' N N
\\
Guide Tube l
weld s
s l
Water Level *W
,, ' ; }
j s
Recirc. Inlet t
g Nozzle s
.. ;.. g.., -
~~;.)...y.;
\\ s 1
\\
?
I
=.
\\ '\\
l II f
l l
n.
)
c n
s 4*
.n
,. - r..:
I.
e 1.
f y., *.y
..e r
g l
cw b
sC) 1 i>
1.
L l
Note: Prior to installation, a template is made to the final 1
site of the safe end. The safe end is then welded to i
the elbow prior to installation. The elbow to riser Il weld is planned to be a heat sink weld.
... - - -..... -... -....,.. -,, _...... _. ~ -...
......... -. - -.,. - -. =,. - -
]
l INLET Figure 8 Install Shielding
.s.
i 4-
\\
s N
s N
s N
s N
s yVessel A nnulus
.g
- g
\\
s d.
+.. +
g Core' Shroud
- }
N
.(f g
<*l'i
. g
,Wat'er In Gaide Tubes \\
Reactor Shield s
r Wall Recire. Outiet M
Lower Core Support Nozzle %
~
i Pfa te N
Te:nporary u de ube
\\s Shielding i
! I N
- ~! )
i.
s Recire. Inlet
't tN s
}l h
r Nozzle l
Water Level c ~;
\\
i N
l t
Water Level NN j
I
_[-
l
'I*
y
\\
~
~ ~ -
,e s
[
i 1
r *,.,, ?.
a.
t 1
c-
-o 1
s 1
Note:
Constant flow through CRD, thus keeping the guide i
tubes filled. Constant drain through one CR0 stub tube.
I i
1
y
~...... _.......
...._ I- ~ ~
.i INLET Fig'ure 9 Cut Elbow x.
. c.
,, e
~-
s s
s c
s e,
..g.
s s
s N
N N
i s
N g Vessel A nnulus 4
. /.
s s
N s
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Reactor Shield -
/ a'ter In.Gu ide Tu'bes \\
t 1' Y
~W s
\\
f Wall
.I Rectre. Outlet Lowerg supp rt Nozzle %
e s
s
{h ud ue s
n 11 s
- e. ':
,s s
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A, g
r Nozzle r
i N.
"ut 2 1
s Water Level a
N ri- ~ PM.. T-i Qut 1 s
i J
l
.s x.s I
Water Level
'l i
l
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i
= = -
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)
i s
4-
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I.
e y *,' '.?.
..., J.
i s
b c-
- C) 1 i
'i l
1 1
i Note: Riser to elbow cut is to be approximately 2" below 1
the weld.
l
. _ _..._.I
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l INLET
,\\ -
Figure 10 Remove Elbow & Install Shield Plug i
F'.-
~
w
. e :.'
s
/.
n.-
g,
I:
g
/
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s i
N n...t, s.
s s
s
~
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y Vessel A nnulus
\\
5 s
~
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I
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.sY Reactor Shield
,.--Water In Gu ide Tubes s
T f
'Jaf )
g
'Reci'rc. Qu'let g
Ldwir Sub: ort s
t Nozzle m
~
s N
u e ube N
Nf Shield Plug I
1 5
I
-s Installation Tool f
16 Lead
/
ji Segments l
1 J
iWater Level Drive iT l
V
- Ro6-I C"3 Water Level l
t -
y NA
..J.,
f l
?
l l
T-l l
y.<,
-a i,,,
}
8*
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Cw
-s:'">
i 3
Elbow Shie ded-and Removed From Drywell
,w---
- m.,,,. -
,-,--e g
4
-,p
-g
,,,,w,7
1 4
I e
k.-
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INLET Figure.11 Remove Spool Piece
)..=::i..-n.?'.;.;;h... ]%.,W. 3 - ;
.1.
\\
,;... a_.. >.
,s.-
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s s
g s
g.
t 1.
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s s
[
i gVes.:e1 A nnulus
/
l..
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+
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- {
i-
.s -
r-,-
Reactor Shield
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Ir-Water in Gu ide Tubesm
[
Wall Recire. Outlet LcaerCg,r Support p
Nozzle m t
I s
Centrol R::d N
N.
Guide Tube s
Cut
[a
' J',
Cutting Machine r.
/.
Added Shielding A
s a
s h
/
l
//
Water Level
. ~. "
L.:v.-
v,-
7
,e c
I Water Level l
x.s I
'_. l. _
l b
[
)
(;
e J'
=I.-
r u-
., ' t.
r
)
- ~.e
!. {
- e jj Fi C-so f4 l
l 3
3 f.
Shield & Rig Spool Piece Out of Drywell ui:
J Note:
Inlet nozzle is prepared the same as the outlet nozzle to assure code compliance.
L I'
l1 L
v.
j i.
s INLET.
{
Figure 12 Install Spool Piece
~ '
- 24*..;.. \\', 7C.
- W
^ '. '
^T.* ! v..*: Y 'l7s
.a~
m s. :..
\\
s s
s j.
s s
}-
s s
s s
s s
j Vessel A nnulus i
(:
s s
. Core'Shrrud %
N N
at
.e Reactor Shield s
pWater. In Guide.. Tubes g
.4 s
s
-r Wall N
N Lower h, Support Recke. Outlet
\\
Nozzle --,
g s
^l jj Centrol Red 5
N g\\//
Guide Tube
\\s l[
' ].*
[
'N Weld I
m ll l
///
Water Level I
- .) ~ -
e
,A--
o m,.
y f.
44 c
-...-.2.
c.
s.
x B
l w
3 I
Water Level l
\\-
)
I l
I I
= =.
-t 4
.f 4*
a
{
g.'- r,.
y 1
,-l. '.?.
.=
a
- .e cw so I
k t
g j
j Note: A template will be made to the final size of the 3
safe end. The safe end will then be welded to the
}
spool piece prior to installation on vessel,
,m
,c.-
.1 1...
IhtET
.j Figure 13 Remove Shield Plug d..'..}. i=:. e :...' %,,.:,.. @
?*s,:.
y
.. 7 :
.y - t,7yi" x
s N
s i,
i.
j.
i s
N
.i s
s s
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gVessel A nnulus s
/
s N
~
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.i f.a.Y Reactor Shield s-P. Water.In Gu ide Tubes N f ' k'all
\\
Lower Core Supcort
.j Recire. Outiet p1a te Nozzie -
s t
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l Temporaq
.t Guide Tube s s s
4 Shielding O
\\
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~
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t a
Water Level
- s. u - - =. ~...,...
..s..
s
,- l
.e l
-Water Level xx t
t T-I L-t' i
i-1 T-t-
.=
.< e.,
~2 J'
l I A
c-L i
1
= _ - - _ _ -
l
}
j Note: Plug installation tool and plug will then be
.i removed from the general area.
l l.
I, I
<l:-
['
INLET.
Figure 14 Install New Elbow o
- . -v
.y.,; u ';.. % i t:
S. ;. w. 4...a '.
- c
... o
- t e.
- .p n...
..* w --
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s
. s v
.s s
p s
s s
s i
1 s
s 7-- Vessel A nnulus
/
e s
s s.
Co're 'Shrou'd
,(
/ ater In'Gu ide' Tubes' w Reactor. Shield
,e..
[
Wal'1
I 5
W Lowergr!esupcort l
Recire. Outlet i
Nozzle %
s l
s Temporary
'C ntrol Rod q\\
. Guice Tube Shielding s
g
.[
'.].'., h
\\
v.
s s
ee: ire. Inlet
' ?
f****
b Weld
.a f
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- s v.
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,a j;-.
.'s M
(
l
.i P
i j
- Water Level s
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= =.
e
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,.[.
i
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Note: The elbow will be fitted prior'to removal of the shield I
plug. The riser to elbow weld will be a heat sink weld.
{
9 b
C FIGURE 15
..g.......
=
PHASEI t
=
.=,
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CUSTOM SHIELDING DESIGN e
1
' BASIS FOR DOSE ESTIMATES o-o OUTLET N0ZZLE. SHIELDING o
INLET N0ZZLE SHIELDING 1
e
- i 9
l 4
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o GUNDREMMINGEN ' KRB (SAFE-END REPAIR) o DUANE ARNOLD (SAFE-END REPAIR) o NINE MILE POINT UNIT 1 (RADIATION SURVEYS) 4 h
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ANTICIPATED RADI ATICil.LE_VELS, NORP.AL SHUTD741 CC!iDITIO*;S - VESSEL FILLED WITH WATER, FUEL IfiSTALLED, PLA'IT Ifl COLD Cl!UID0;ci, ;0 TEf?0RARY 59tELDI::G 1:XSTALLED
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i LOCAL LEVELS I
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tilTICIPATED RADIATIC:1 LEVELS ENTIRE VESSEL DRA!;ED, FUEL A:D 'iEUTR0:1 SOURCES RE!10VED, C0iTROL R0C: I;SERTED, ;0 TEI'PORARY SHIELDING 1:; STALLED
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e SPECIAL SHIELDING DESIGNS o
SHIELD CURTAIN o
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At t.UTLET N0ZZLE FRIOR TO CUTTING:
'NOULUS ONLY DRAINED, FUEL rid NEUTRGil 50.URCES RE"0'/ED, CONTROL RODS I:i!ERTED, TETTORARY ~lllELDING IN PLACE
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Potential Cause of Failure Disposition 1.
Startup Anomalities No appearance significant anomalities 2.
Thermal Lockup of Vertical and Reading from the snubber LVDTs indicate i
Horizontal Snubbers proper snubber movement after startup I.
in 1981 and during cooldcwn in Mar.ch
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Chemical Cleanup During Last Review of London Nuclear Process reveals Outage that no adverse IGSCC effects would be expected.
In addition, the area in the vicinity of the safe-ends was not cleaned.
4.
Proper Setting and Functioning Field review indicates no anomalities of the System Spring liangers t
i 1
5.
Reactor Coolant Chemistry No apparent intrusions which would cause a failure.
?
h. Configuration of. Supports and.--.
No differences which would indicate that ".
Piping for Oyster Creek Nine Mile Point Unit #1 would be acre susceptible to cracking.
I 7.
Parametric Study of Effects of An evaluation of the worst case vibra-1 Pump Frequency, Pump Forcing tional frequency has been performed.
{
Function Amplitude and Piping The resultant stresses attained from g
System Stiffness on Safe-end the evaluation do not approach the yield Nozzle Stresses stress of the material.
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i Os SIMfARY OF NAN-REM TOTALS 170R RECIRCULATION
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Case 1-Case 2 Recircu-No Decon-
.lation Loop; Man E1!ENT tamination Decontamination Savings. Hours flours 1.
Recirculation Safe Ends Suction Side (Outlet From Reactor) j A.
Preliminary Drywell Work 500 400 100 715 3000 l
B.
Removal of Existing Safe. Ends j
And Elbows 350 200 150 550 1100 I
C.
Installation of New Safe Ends' -
And Elbows 450 300 150 1100 2150 D.
Post Drywell Work 100 90 10 80 300 1
SUBTOTAL:
1400 990 410 2445 6550
~
2.
Recirculation Safe Ends Discharge Side (Inlet to Re, actor)
A.
Preliminary Drywell Work 185 170 15 150 500 r
l B.
Removal of Existing Safe Ends 8
and Elbows 1050 800 250 850 1700 f
C.
Installation of New Safe Ends and Elbows 1950 1700 250 1750 3800 D.
Post Welding Drywell Work 100 90 10 400 1300 SUBTOTAL:
3285 2760 525 3150 7300 I
RECIRCULATION LOOP SAFE END DEPLACEMENT TOTAL:
4685 3750 935 5595 13,850 3.
y.-
NOTES 1.
. Average decontamination factor (DF) assumed was 2.
(Actual DF could be as high as 10 or more).
2.
Figures indicate a significant man-rem savings if decontamination work is performed. Approximate cost per man-rem is $535,/ man-rem.
Larger savings could be reali:ed if any rework is performed.
3.
Man-rem estimates are based on conservative man-hour estimates and crew sizes, but there is no contingency added for any rework such as rewelding or multiple radiographs.
4.
Man-hour estimates assume respirators wcrn on certain jobs, normal craft productivity and a learning curve for initial work.
5.
Figures are "first-cut" estimates and will be refined as mock-up work is observed and radiation surveys are performed when the vessel and recire loops are drained.
6.
The man-rem savings could be reduced to 900 due to the potential exposure which could be received during the decontamination process.
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ANCITIPATED FADIATI0il LEVELS TEP.POPARY $111ELDit:G 1*l PLACE PRIOR TO PAXI:G CUTS AT 13LET tiOZZLES. VESSEL FULLY DPAltlED,.. FUEL A1D NEUTR0!! SOURCES REPOVED. C0!!TRO'l RODS !!1SERTED
1B Y NIAGARA RUMOHAWK NtAGARA MOHAWK POWER CORPORATION /300 ERIE BOULEVARD WEST SYRACUSE. N Y 13202/ TELEPHONE (315) 4741511
~
i August 6, 1982 l'
k t
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j-) 7,m '; 4 v, Mr. Darrell G. Eisenhut, Director I \\ i L C ; 9 '- F 'G v~
Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission liUG101987 Washington, D.C.
20555 Re: Nine Mile Point Unit 1
- -$((f,.g Docket No. 50-220 r.os rio.-
OPR-63
Dear Mr. Eisenhut:
On_May 11, 1982, Niagara Mohawk provided information on the replacement of the recirculation safe ends at Nine Mile Point Unit 1.
That letter was in response to your letter of April 21, 1982.
The purpose of this letter is to confirm orevious verbal communications with members of your staff that we olan
~
to reolace all the 28-inch recirculation system pioing. The basis for Niagara Mohawk's decision to replace the recirculation Dioing is provided herein.
Chronology ano Conclusions On March 23, 1982, through wall cracks were detecteo in two of the ten recirculation safe ends.
On March 26, 1982, ultrasonic examinations were performed on these two safe ends and one other. The results of those examinations confirmed crack indications.
Based uoon that information, Niagara Mohawk decioed to replace all ten safe enos.
In mid-Aoril, 1982, two boat samples were obtained from one of the safe ends in the vicinity of the through wall cracks. One each of these samoles was sent to General Electric and Battelle Laboratorics for evaluat'an.
The results of those evaluations in mid-May confirmeo the presence of intergranular stress corrosion cracking.
Prior to receiving the results of the evaluations from Battelle ano p W.":/' $ *. bMleMil O;triq%mi'? & phenomena were evaluate.d.by Niagara Mohawk as General Elec iother.
I fian f
qualitative basis' t,o determine high stress areas of' 'Se recirculation pioing'-
system under various coerating scenarios (i.e. locked Dumo snubbers, etc.).
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' ' met' hods,ditMs' detshiirted"th'at 't'wo of 't'he f'iveNelus 'hao coce ~rea'drtable- ~ -
' ~ -
indications. When increased ultrasonic transducer gain was used, the remaining three welos exhibited indications.
S* pl8
A'ugust 6, 1982 Page 2 The two pump discharge casting to riser elbow welos with normal ultrasonic indications were later examined by dye penetrant methods on the inside diameter. The results of those examinations indicateo the presence of cracks.
1 In early May,' 1982, a replication process (i.e., obtaining a negative of the crack with a piece of #11340 celluose acetate taoe) was implemented. The i
results of General Electric and Niagara Mohawk's review indicated the presence of intergranular stress corrosion cracking. A boat sample of that same region (taken on June 13, 1982) was evaluated by Sylvester Associates and confirmed the presence of intergranular stress corrosion cracking.
Based upon the confirmation of cracking,at the safe ends and the oumo discharge casting to riser elbow welds, it was decided to ultrasonically inspect all of the remaining welds, where the radiation fields permitted.
The results of those examinations indicated cracking in a large number of welds.
In most cases, these indications could only be obtained when using the increased gain ultrasonic technique procedure.
Attached Figures 1 through 5 outline the location of all welds in the recirculation loops.
Table I summarizes the examinations and results of the examinations.
Based on the results of our~ examinations and investigations, it was cecideo to reolace the 28-inch recirculation piping.
Preliminarily, it acoears that it will be advantageous to replace branch oiping also.
- However, all of the technical issues have not yet been resolved. Therefore, no final oecision has been made to replace this branch piDing.
Reolacement Program All reolacement material will be 316 NG or equivalent, with a carbon content of less than 0.02 percent.
This material is of the grade which does not require augmented in service inspection per NUREG 0313, Revision 1.
The replacement safe ends were manufactured in 1979 to the 1977 ASME Boiler and Pressure Vessel Code Section III (adoenda through summer 1977).
The remainder of the pioing was ordered and is being manufactured to the 1980 code through winter 1980 Adaenda.
The actual replacement will be accomolished in accordance with IWA4000 and IW84000 of ASME Boiler and Pressure Vessel Code,Section XI, 1977 addition (addenda through summer 1978). All weloing will be in accordance with Section IX,1978).
The fitJuo reouirements will:be.in accordance l
%W. ';l: y.hwith jaddehd5ftrdssh$1 hidr 07s) Tode~.Ttir@dsTi6d di,hih,gIy jh ANSI p,3,h.},, ",", - (, *3, i Since the configuration of the system will be.the,same as.the or,igi.na.l... hk khd b,.- 8! k 1 n J.'s ~.. Safety Ana'lys.is.Recort;. -- ^ " '~" r " ' ' " ~ ~ ' * ' " ' ' ' ' ' ' ' ' ' ' ' ' " ' ' ' " ' " i
August 6, 1982 Page 3 i i Nondestructive examination requirements for field welds which are applicable to this replacement are as follows: Radiographic and Dye Penetrant Section III of ASME Boiler and Pressure Vessel i 1977 Sumer 1978 addenda i Ultrasonic Section XI 1977 Sumer 1978 addenda The replacement methodology is in the process of being develooed. I Preliminarly, the method of reDl8 Cement is to dismantle all of the oiping in all five loops. The loop furthest from the equipment hatch would be rebuilt, from the pumo uo. The uppermost elbows would be used for the, final closure and fit up, on both inlet and outlet sides"of the loop. Rebuilding of the loops would continue until the final loop (closest to.the equipment hatch) is completed. A revised occupational dose estimate for the expanded scope of work will be included with the next quarterly repair program status as required by oaragraphs 2.D.(6) c ano d of our operating license. Very truly yours, NIAGAF,A M0 HAWK POWER CORPORATION T. E. Lemoges l Vice President Nuclear Generation GJG/kmb s & O a,..w. .<.=..<:.e. . w a.r. >.i.e.. + ~. m a ~.< ~. m. ~ e p e p. .. ** f =*% 6% - / 6 ~ ~
1 t TABLE 1 l i. Ultrasonic Dye Penetrant ( Examination Examination Weld No. (UT) (PT) Results(l) l No. 11 Recirc. Loop P32-FW-1-W Not Inspected P32-SW-1-W Not Inspected UT Indications ( P32-SW-2-W X (Increased Gain) P32-FW t.'-W X UT Indications (Increased Gain) UT Indications P32-FW-3-W X (Increased Gain) UT Indications P32-SW-3-W X -(Increased Gain) UT Indications P32-FW-4-W X (Increased Gain) UT Indications P32-FW-26-W X (Increased Gain) P32-SW-17-W Not Inspected P 32-FW-2 5-W Not Inspected P32-SW-16-W Not Inspected P32-SW-15-W Not Inspected P32-FW-23-W Not Inspected P32-FW-22-W Not Inspected No. 12'Recirc. Loop P32-FW-5-W Not Inspected P32-SW-4-W Not Inspected UT Indications P32-FW-6-W X (Increased Gain) UT Indications P32-FW-7-W X (Increased Gain) UT Indications P32-SW-5-W X (Increased Gain) UT Indications P32-FW-8-W X (Increased Gain) l l P32-FW-31-W X UT Indications (Increased Gain) l l P32-SW-20-W Not Inspected P32-FW,30-W Not Inspected i k-@5 'NNp3tykyg'-v $M b.L.. M,vo W.'Mdt Jn'sphned. * <. ' -,."F.c/M.e 4.<q. :./ t..%: N ti(4i<,yf'..- P32 ~SW-19-W ,Not Inspected P32-SW-18-W Not Inspected .'p.pf.c. 'h5?.h.[,hhk'.hy; @gq{.4p.ig $Y YC<N h k!*$'$Wiffch?$if,hk ij<. :f, } . Plu s L e ak s :.< :.......a.x.. . 7.,....., 2..:...... ..s... (1) Where increased gain is indicated, the normal code ultrasonic examination showed at least one recordable defect indication. However, using an ~ increased gain, intermittent indications were observed along the circumference of the inside diameter. E.
TABLE 1 (Continued) Ultrasonic Dye Penetrant Exaniination Examination Weld No. (UT) (PT) Results(l) No. 13 Recirc. Loop P32-FW-9-W Not Inspected P32-SW-6-W Not Inspected UT Indications P32-SW-7-W X (Increased Gain) UT Indications P32-FW-10-W X (Increased Gain) P32-FW-ll-W X UT Indications (Increased Gain) UT Indications P32-SW-8-W X (Increased Gain) UT Indications PI2-FW-12-W' X (Increased Gain) P32-FW-36-W X X UT Indications (Increased Gain) PT Indications P32-FW-35-W Not Inspected ~ P32-FW-34-W Not Inspected P32-SW-22-W Not Inspected P32-5W-21-W Not Inspected P32-FW-33-W Not Inspected Code UT Indications P32-FW-32-W X i No. 14 Recirc. Looo .P32-FW-13-W Not Inspected P32-SW-9-W Not Inspected UT Indications P32-5W-10-W X (Increased Gain) UT Indications P32-FW-14-W X (Increased Gain) P32-FW-15-W X UT Indications (Increased Gain) UT Indications P32-FW-11-W X (Increased Gain) UT Indications P32-FW-16-W X {si,;p;r. M/.R. i ;sw.4.v ecy.,/V.,;h :e.y pr.y,z:g ey;s:t;..,;. g:. ;.g..:,b.,.y;,.{,l.ncr.e4.5eff.1Gg.j tt);.,. . t ~ r 4 UT Indications P32-FW-41-W X-(Increased Gain) I ?..>R32..5W9.6t? '.... 8 :.~ - Rot. Insp.ected - bEEbWPI2fN-404W[4%f' s ?iM'YGJf?f 71%d.M1W.hiFdNd.?.@td.2,.$ @;Jh W$$.V.N.tyW.Oik)ig-{., i
- . c, -
Not 4n'spected ;. -
- s. -.,.....,.......
u.- '.,=.>'P32-FW-39-W ' e s P 32-SW-2 5-W Not Inspected P32-SW-24-W Not Inspected P32-FW-38-W Not Inspected P32-FW-37-W Not Inspected l
TABLE 1 (Continued) Ultrasonic Dye Penetrant Examination Examination Weld No. (UT) (PT) Re sults(l) No. 15 Recire. Loop P32-FW-17-W Not Inspected P32-SW-12-W X X UT Indications (Increased Gain) PT Verification After Removal UT Indications P32-SW-13-W X (Increased Gain) UT Indications P32-FW-18-W X (Increased Gain) UT Indications P32-FW-19-W X (Increased Gain) - UT Indications P32-FW-20-W X (Increased Gain) UT Indications P32-SW-14-W X (Increased Gain) P32-FW-21-W X UT Indications (Increased Gain) P32-FW-46-W X X Code UT Indications PT Indications P32-SW-30-W Not Inspected P32-FW-45-W Not Inspected P32-FW-44-W Not Inspected P 32.-SW-2 9-W Not Inspected P32-SW-28-W Not Inspected P32-FW-4 3-W Not Inspected P32-SW-27-W Not Inspected P 32-FW-4 2-W Not Inspected Through Wall Crack (Leakage) kv Nt::.<y k.G':%<.1. @y.rm.,.c A'.:;. > '.g :l..<:;.',.p :.'.:e.i;a..l.. > g. 4 < y.~.:, p.:l. 5Ly:,. m:.;. g.s. s +. * *... - ,***g 'I'. f. . # h* k*. M
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o FIGURE I RECIRCULATION LOOP NO. ll MATCH LINES s x P32-FW-I -W l 1 - r. f l P32-SW - l -W s I \\ { l SAFE END l @32-sW-2 -W/ P32-FW-22-W L I P32-FW W [ / P 32 - SW -W y I l REACTOR RECIRC SKEE END i 'PU MP HO. I'l C P3 2 - SW W l P32.- FW-2 -W C / g P32-FW-3 -W V ~ l g N i, a P32 - FW W [ kb l P32. - FW -2 5 -W \\ 'I P32 - SW W s.;.~... v.L..., s..v 2 3a.a.., g. } 'I p.s.,>e..s.. Ggi.,.. >,, _ 9_....., _..... _.... j [ REACTOR VESSEL ~ i, i i y N n OWSs ik5.+1nb.bQ W *" 4.s.>.:.b.tedi.Sv.mkhi+gf.n.nkwyup: n .,..i,. A \\ P32-FW-26-W ~ P 32 - FW W
FIGURE 2 RECIRCULATION LOOP NO.12 MATCH LINES . \\ P31-FW-5 -W l l / 32-SW-4 -W x ( i j SAFE END 'N P S E-FW W P32-FW W P32.- SW W y R'EACTOR. REClRC 5AFE END . PUMP HO.12 c esz-sW _19 _W 53 2-- FW W [ ] g fi k3 2-FW W ~ l A J P32 -FW W / 9~ I ' / P32 - FW-30 -W I ( / P 32 - SW W l r i. hyf.::..+.,..... ::.., ;.,. y:.
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{ REACTOR YESSEL il ~ i l-y pp&yc:v.s... V..gre.+.wmpum.vi.sy;.i+wsmysa3:., na'wsw+ w <w \\ P32 - FW-31 -W I N 1 P31 - F W-8 -W =
t riounc 3 REClRCULATION LOOP NO.13 s MATCH ' LINES s x Q, P32-FW-9-W, e X z 5 2-SW-6 -W \\ l SAFE END .\\ [ P32-FW W '32-SW-7-W l g PS E-FW W / y P 32-SW W l l REACTOR RECIRC SKEE END . PUMP HO.13 c P32-SW w 32-FW-lO-W c / ss i l f 32-FW-i l -W l A P32-FW-34-W h l P31-FW-3 5-W I \\ 'IP32-S'il-2 3 -W c 2- '. n...;...}. / I. h.&r~;s. a:.n y
- j. 4.
. g.,., ;;e:,.;. v.:. o., 4, < c.y. .'. >y..i..:. <3.g, g. i )A f b~b4.Ihd,jb. hhb 9 sis".d57.)$; N,rdk i.W.p.4i.;.dp;.#.QI#;qi[Nj;.?S3. '$c.#f$ f .. ;,\\ c i. \\ JP3 2.-- FW -3 6.-W. '.... t i P32-FW-12 -W i
FIGURE 4 RECIRCULATION LOOP NO.14 MATCH LINES s x P32 - FW-13 -W I I 'N \\ f e i B2-sW-9 -w
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g [ l SAFE END 7, \\ i {l I P32-FW W l P31-FW -3 8 -W P32-SW -2 4-W y l l REACTOR. REClRC SAFE END FUMP HO.14 C P32-SW-25-W $32-FW W l \\ N32-FW-15-W 'l ~~ A 4 P 32'- FW W / C I-l P32-FW W I ) 7 /P32-SV.' - 2 6-W i i .,h', ,:... h.;j., ,,.ky .{... :. ,,,7..g..f,... 3,. <h. bs.;:~: A. : .~; e. ( REACTOR VESSEL-i i t y ...W w ,.; no s.n i y.:.ypi.w w w g..:y g,.,g y nV4w#om91.1w & g u n.s g.. w ...,.1, .. pg ........g., .m P31-F W W
FIGURE.5 L ~ RECIRCULATION LOOP NO.15 MATCH LINES / N 3 P 32F FW-17-W l 1 \\ '3 2 -- SW W ~ h l SAFE END f 32-SW-13-W l i P32-FW-42-W j { i 32-FW-18 -W/ P32-SW-27 -W 7 ~ y P32 -FW-43 -W e l [ ~ SAFE END REACTOR REClRC P32-SW-2 B-W ~FUMP NO.15 f31-FW-19 -W / \\ h c. P32-SW - 2 9 -W ~~ )32-FW-20-W c l / A l l P32 -FW-4 4 -W l b I l i / P32-FW-4 5-w 2 \\ /P52 -SW-30 -W . h lI - n [T
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asioq% UNITED STATES NUCLEAR REGULATORY COfnMISSION j{ WASHINGTON, D. C. 20555 f SEP 2 ISE, Docket No. 50-220 Mr. Donald P. Dise Vice President - Engineering c/o Miss Catherine R. Seibert Niagara Mohawk Power Curporation 300 Erie Boulevard - West Syracuse, New York 13202
Dear Mr. Dise:
SUBJECT:
RECIRCUI.ATION SYSTEM PIPING Re: Nine Mile Point Nuclear Station, Unit No.1 By letter dated August 6,1982 you provided information regarding plans for replacing all of the the 28-inch recirculation system piping, such work to be done in conjunction with your safe-end replacement program authorized by Amendment No. 49 to Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Staiton, Unit No.1. This letter addresses the acceptability of your program related specifically to those aspects associated with recirculation system piping removal and the attendant mitigation of worker radiation doses. Section 2.D(6)c. and d. of Anendment No. 49 to your license established various license conditions related to safe-end replacement. These conditions required: (1) collective occupational dose estimates weekly, (2) notification if estimates exceed 2906 person-rem by more than 10% and (3) progress reports at 90-day intervals. By letter dated August 6,1982, as supplemented by letters of August 16 and 26,1982, you provided a revised dose estimate associated with safe-end replacement, as well as new dose estimates for the removal of recircu-lation system piping. The dose estimate associated with removal of the recirculation piping was estimated to be 533 person-rem. You also provided information which revised the dose estimate associated with safe-end replacement downward from 2906 person-rem to 2036 person-rem. We have reviewed your revised dose estimates as related to the 2906 person-rem specified-in Section 2.D(6)c. of Amendment No. 49 to your license. We find that the combined total person-rem exposure estimate associated with the removal of all of the recircu-4 'lation system piping, as well as the safe-end replacement, is bounded by the license condition. \\ J v o r I s -3 L-
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Based on our review, we conclude that the expected total person-rem exposure estimate associated with recirculation piping removal is as-low-as-reasonably-achievable (ALARA) and is bounded by the cited conditions specified in license Amendment No. 49. Therefore, by this letter the recirculation piping removal is approved and you are authorized to proceed with the removal activities. As discussed with members of your staff, we require that you provide additional information for our review before we can authorize replacement of the recirculation piping. Sincerely, ~ Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing cc: See next page i l 1 i
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