ML20126A123
| ML20126A123 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/09/1992 |
| From: | Cofie N, Giannuzzi A STRUCTURAL INTEGRITY ASSOCIATES, INC. |
| To: | |
| Shared Package | |
| ML20126A121 | List: |
| References | |
| SIR-92-082, SIR-92-082-R01, SIR-92-82, SIR-92-82-R1, NUDOCS 9212170244 | |
| Download: ML20126A123 (20) | |
Text
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't to JPN-92-067 E
Report No.: SIR 92-082 Revision No.: 1 REFERENCE ONIX,l Project No.: NYPA 360 Decernber 1992 Root Cause Failure Analysis of the Equalization Line on Reactor Water Oeanup Valve 12RWC 46 Prepared for:
New York Power Authority Prepared by:
StructuralIntegrity Associates,Inc.
San Jose, CA 12/9/S4 Prepared by:
)
Date:
N. G. Cofie p to ed y: -
e.IM Date: -[
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A. [/$iEnuzzi 0
9 92121702449212h333-4 pDR ADOCK 0500 PDR.
Table of Contents Section hgg 1.0 B ACKG RO UbrD..............................................
I 2.0 EVALUATION OF POTENTIAL FAILURE MECHANISMS...........
2 2
2.1 The rmal Fatigue.........................................
3 2.2 Mechanical (Vibration) Stress...............................
2.3 Intergranular Stress Corrosion Cracking (IGSCC).................
3 4
3.0 REP AIR....................................................
4 4.0 CON CLU SION...............................................
SIR.92 082, Rev.1 i
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List of Figures figuIn Eagt 1
Failure Location on Equalization Line.............................. 5 1
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SIR-92-082, Rev.1 ii
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1.0 BACKGROUND
During the current refueling and maintenance outage at the James A. FitzPatrick Nuclear Power Station (JAF), visual examination of a 6" valve in the reactor water cleanup (RWCU) line, valve 12RWC-46, revealed evidence of leakage from the 1/2" diameter Type 316L stainless steel equalization line at the socket weld to the valve bonnet. Figure 1 illustrates the 6" carbon steel valve, the equalization line with its 1/2" valve and the leak location. An attempt was made to repair the valve by welding. The repair attempt revealed a larger leak on the same weld adjacent to the initialleak. The presence of the second leak prompted the technical and management personnel at JAF to reconsider the repair by welding and to consider an alternate repair approach which would allow for operation of the valve for one additional cycle, ne objectives of this report are to examine the potential failure mechanisms, identify the most likely mechanism or rne.anisms, and demonstrate the manner in which the repair which was selected for dis location remedies the potential failure mechanisms.
The potential failure mechanisms for this valve include thermal fatigue, mechanical fatigue (vibration) and intergranular stress corrosion cracking (IGSCC). Since the failure location has not been removed at this time (the repair involves sleeving part of the equalization line thereby covering the leaking location), no metallurgical failure analysis was possible at this time. The failure analysis will be performed when the valve is permanently repaired at the next refueling outage The following paragraphs describe the potential failure mechanisms in greater detail, identify the most probable failure cause and describe the manner in widch the repair addresses the potential failure mechanisms.
SIR-92-082, Rev.1 i
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l 2.0 EVALUATION OF POTENTIAL FAILURE MECHANISMS 2.1 Therinal Fatigue Thermal fatigue is a likely mechanism for the cracking that has been observed. Since the valve body is carbon steel and the equalization line is stainless steel, relative thermal expansion stresses develop as the component is heated and cooled. Since the thermal expansion coefficient of stainless steelis greater than that of carbon steel, the pipe wants to grow more than the stainless steel. There are several relevant stressors:
a)
Isothermal heatup of the valve and equalization line produces a bending moment in the equalization line, especially at the lccations where it attach d to the valve body.
The locations of the cracks at the toe of the 'lllet weld are where the maximum stress cycling would occur due to cycling between operating and shutdown conditions.
b)
Isothermal heatup of the valve and equalization line produces a local throughwall bending stress distributica at the point where the equalization line attaches to the valve body. This stress would be additive to that of a) and 'vould be uniform around the circumDrence of the pipe.
c)
Transient heatup of the valve and equalization line would produce gradients through the valve and pipe wall which would tend to produce some additinnal stresses at the location where the failure was observed.
d)
Tran: dent heatup of the valve with a thermal lag of the fluid in the equalization line could also produce significant gross bending stresses as in a) above. However, these may be less than the steady stiesses described ia a).
SIR-92-.082, Rev.1 2
e)
It is expected that the fluid in the equalization line is relatively stagnant. Since the equalization line and the attached valve are not insulated, the line could run considerably cooler than the valve body, especially for conditions when there is no flow in the RWCU line. The relative temperature difference could produce bending stresses with an opposite sign as that described in a).
2.2 Mechanical (Vibration) Stress Vibration of the equalization line could result from a resonance with the recirculation pump vane passing frequency. The pump vane passing frequency determined to be 24 hz. The equalization line lowest natural frequency has been determined to be 42 hz. Thus, it does not appear likely that vibration is the principal cause of the failure of this line.
2.3 Intergranular Stress Corrosion Cracking (IGSCC)
IGSCC is a phenomenon which has occurred often in sensitized austenitic stainless steel piping and components in operating BWRs. Due to the oxidizing nature of the BWR environment (in the absence of hydrogen water chemistry), this phenomenon cannot be niled out in the equalization line. The line is fabricated from Type 316L stainless steel, a material wnich is considered resistant to IGSCC by the technical community and by the NRC, thus the probability ofIGSCC as a failure mechanism is somewhat reduced. However, the socket weld by the very nature of its fabrication, produces a crevice which can cause IGSCC even in "L" grade stainless steels. The NRC document NUREG 0313 Rev. 2 discusses the possibility of IGSCC in crevices of conforming IGSCC resistant stainless steels, like Type 316L stainless steel.
SIR 92-082, Rev.1 3
C
DEC 03 '92 16:00 STRUCTURA INTEGRITY g y pgg e,
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'0 REPAIR The repair consists of providing an outer sleeve made of Type 316L stainless welded to the valve and the existing %" diarneter pipe elbow. He repair is designed t address all of the thermal stress conditions described above producing a design whic Class I the requirements of Section III of the ASME Boiler and Pressure Vessel Code fo fh components. In addition the stiffness of the new sleeved line will be greate 1/2" inch thereby increasing the natural frequency of the line thus making fatigue mechanism less likely. Finally, use of sensitization and IGSCC r stainless steel as the sleeving material provides a pressure boundary which conf NUREG-0313 and is resistant to IGSCC. This is not to say that IGSCC is im i
crevices are created as a result of the sleeving. However, given the sensitization res of the material combined with the recognition that this repair is intended for o cycle only suggests that IGSCC or crevice corrosion will not be a conce
4.0 CONCLUSION
In conclusion, the most probable cause of the failure of the equalization line in 12RWC 46 is thermal fatigue, originating from one or more of five possible sources
~
Vibrational fatigue is not likely and the repair will make it less likely. IGSCC or c corrosion is a possible failure mechanism for the joint. The use of Type 316L The failure analysis to be reduces the probability of IGSCC but does not eliminate it.
performed following the next operating cycle will determine definitively failure.
4 SIR 92-082, Rev.1
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SIR-92 082, Rev.1
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New York Power Authority James A. FitzPatrick Nuclear Power Plant ATTACHMENT 3 to JPN-92-067 Modification Information including:
Minor Modification Package'for.
Plant Modification M1-92-383 Nuclear Safety Evaluation JAF-SE-92-220 ASME Section XI Repair Program for Repair.of the RWCU Valve 12RWC-46 1/2 " Equalization Line
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CONTROL FORM tv Authority MINOR MOD TITLE Encaosulation of 12RWC-46 MINOR MOD (MM) NO. Mi-92-383 Ecualization line MM REY.
O NSE NO. JAF-SE-92-220 QA CLASS (circle highest category):
(IP3)
Category I "ategory M Non-Category I (JAF)
Category i Category H Category II/III ON-SITE DEPARTMENT / PROGRAM REVIEWS: (fjIGNATURE/DATE) Check box if required:
( ) FIRE / SAFETY N/A NOPS d
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( ) If,C N/A
( ) SIMut.ATOR N/A
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'NYPA FORM MCM-5, ATTACHMENT 4.1 (FEBRUARY 1991) 0
JAP NUCLEAR POWER PLANT MINOR MODIFICATION PACKAGE (MMP)
MM NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 1.0 SYSTEM (8) :
012 EQt Yes No
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COMPONENTS:
12RWC-46 12RWC-46-1 SUB SYSTEM N/A 2.0 PURPOSE AND SCOPE:
This minor modification is required to repair a 1/2" diameter schedule 80 stainless steel bonnet equalization line on a 6" diameter carbon steel reactor water cleanup (RWCU) manual isolation valve 12RWC-46 located inside the drywell.
This 1/2" pipe has experienced a through-wall circumferential crack at the toe of the weld attaching the line to the bonnet portion of the 6" RWCU Isolation Valve.
This modification repair will be removed during the next refuel outage.
The leak in the equalization line was discovered during post work testing of 12RWC-49 and 50.
3.0 DESCRIPTION
OF CHANGE:
The 1/2" diameter equalization line (ASTM A312 TP316L) will be encapsulated with a 1-1/2" diameter schedule 80 stainless steel ( ASTM A312 TP316L) pipe between the 6" carbon steel valve body and the 90' stainless steel (ASTM A-182, TP316L) elbow.
This same encapsulation detail will be applied to both ends of the 1/2" equalization pipe.
The 1-1/2" encapsulation pipes measure approximately 2-1/2" and 4" long.
A window will be cut longitudinally in the 1\\"
pipe for installation.
The installation and welding requirements will include full penetration longitudinal window seam welds, full penetration circumferential attachment weld to the 6" valve and.circumferential fillet weld to the outside surface of'the 1/2" stainlessisteel 90*
elbow.
. Seal welding over the circumferential crack will be applied to the \\" pipe sufficiently to stop leakage prior to installing the encapsulation repair.
This encapsulation pipe will be fitted with threaded plugs to permit drainage of.any potential leakage from the crack in~the pipe and to.
provide a purge path-for the welding operation.
A-threaded plug will be inserted in'the 1 "_ pipe and seal welded.- To reduce the effects of thermal expansion, 12RWC-46-1 will be changed by procedure to be normally closed and the equalization line will be insulated.
NYPA FORM MCM-5, ATTACEMENT 4.2 Page _E_ of _11_
.c.
e JAF NUCLEAR POWER PLANT MINOR MODIFICATION PACKAGE (MMP)
MM NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 4.0 DESIGN INPUT
SUMMARY
DCM-13, Attachment 4.1 (attached) 5.0 TECENICAL EVALUATION:
5.1 Technical Evaluation Summary This plant modification to the reactor water cleanup manual isolation valve 12RWC-46-bonnet pressure equalization line is Q2. Category I.
This line provides the ability to vent the bonnet cavity between the wedges on this 6" double disc gate valve, manufactured by Anchor Darling Valve Company.
The 1/2" line contains an isolation valve 12RWC-46-1.
During normal plant operations.both the 6" valve and the 1/2" valve were open.
Although the 1/2" bypass valve will be administrative 1y closed by this modification, the normal function of the 6" isolation valve will not be changed by this modification and therefore operation of reactor water cleanup system will not be impaired.
This modification serves to. restore the pressure boundary of the equalization line.
The thermal expansion coefficients are different between carbon steel and stainless and the heatup rates are different between the \\" equalization line and the 6" 12RWC-46 valve-because of wall-thickness differences.
As a result, thermal stress may have contributed to the \\" line failure.
To minimize the thermal effects, this modification will require that the \\" equalization valve 12RWC-46-1 be close:d and controlled by operating procedures..This valve is used during maintenance activities on the 6" 12RWC-46 valve to provide isolation from the reactor vessel. inventory.
The 12RWC-46-1 valve is opened.to vent the bonnet area of the 12RWC-46 valve prior to opening it to preclude-bonnet pressure locking.
The modification also stiffens the piping to reduce vibration fatigue potential.
The new materials used'are low carbon' stainless steel, which is resistant to potential integranular stress corrosion cracking (IGSCC).
A root cause evaluation has been: conducted to determine the probable mechanism that cause the \\" pipe to' fail (Ref. 12.10).
The evaluation investigated potential mechanisms such as thermal fatigue, vibration fatigue and intergranular stress corrosion.
It appears, at this time, that thermal fatigue originating from the dissimilar metals for the \\" line and 6" valve,-was a NYPA FORM MCM-5, ATTACEMENT 4.2 Page 3 of_11_
' h.
I JAP
_ NUCLEAR POWER PLANT,
,7
- MINOR MODIFICATION PACKAGE- (MMP);
MM NO. - M1-92-383 MM REV.-
0 NSE NO. JAF-SE-92-220' potential mechanism.
A metallurgical analysis will b'e-performed when the \\" pipe is removed-during.the next refueling outage.
The 6" valve 12RWC-46 including theE1/2" equalization piping was manufactured by Anchor Darling in accordance.
With New York Power Authority specification MDA 0493-A and installed under plant modification F1 053.
The valve was designed and manufactured'in.
accordance with the requirements of-ANSI B16.34 (1981).
All materials and workmanship will meat the applicable design and installation requirements.
The 6" reactor water cleanup line has an operating temperature of 532'F and design pressure of 1147.psig.- Per analysis by Structural Integrity-Associates (Ref. 12.9)-the 1--
1/2" diameter schedule-80 stainless steel encapsulation sleeve of ASTM A312-TP316L pipe with welds as shown on-drawing 6.37-231 is acceptable for'the design' pressure and temperature, material chemistry, vibration and thermal loading.
The-encapsulation pipes will be? installed on both ends of the 1/2" line:since the configurations are-similar-and any fatigue potential would be reduced in this piping which connects.to the: valve l body.
Since thermal stress cycling still exists the repair with encapsulation pipe will be-limited 1to one plant operating cycle.
The design, installation, and-testing of-the encapsulation pipe is performed-in accordance with ASME-Section XI (1980-winter.1981 Addenda) and: ASME Section -
III, subsection ND (1986 Edition).
' Further-discussion ;
of'these codes isiprevided~1n-the1ASME Section XI' Repair Program document,fAttachmentL13.1 to this
~
modification.
f The unl'ikely' potential for water intrusion from the K
crack being. locked in the cavity and pressurized due to-thermal expansion of the' water was evaluated and'itywas-i-
determined air in the cavity would' allow sufficient pressure relief.
NYPA-FORM MCM-5,. ATTACHMENT 4.2_
Page 4 of J1_
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JAF NUCLEAR POWER PLANT l
MINOR MODIFICATION PACKAGE - (MMP)
MM NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 5.2 Design Input Summary Evaluatiens This modification will not impair the basic I.1 function of the equalization piping.
12RWC-46-1 valve will be normally closed and required to be open only when operating the 6" manual isolation valve 12RWC-46.
The 12RWC-46 valve including the I.3 equalization piping was replaced under modification F1-84-053 under the requirements of ASME Section XI (1980 Edition through Winter 1981 Addenda).
The valve was manufactured in accordance with ANSI B16.34 (1981).
This modification will be performed in accordance with the attached ASME Section XI Repair Plan document.
The NYPA valve specification MDA-85-0493-A I.7 1 provides the design requirements for Rev.
the 12RWC-46 valve including the equalization line.
These requirements were considered in the Repair Design.
I.10 This modification is not in conflict with modification F1-84-053.
This modification is QA Category I as I.11 identified on the Master Equipment List.
II.1,4,5 1-1/2" diameter schedule 80 ASTM A312 TP316L stainless steel pipe has been selected for the sleeve based on the design pressure, temperature, material strength and compatibility, and ease of welding.
V.2,3,4,5 Structural Integrity Associates have analyzed deadweight, thermal, vibration and seismic stresses for the Encapsulation Repair and determined ths.t stress levels are within the ASME Section III Subs. NB (1986) design code allowables (Ref. 12.9).
The Reactor Water Cleanup valve 12RWC-46 equalization piping is Seismic Class I.
Page 5 of _lL_
NYPA FORM MCM-5, ATTACEMENT 4.2
m m. ed4rA aM.A hMAA fWER PULW1
'(ypENGINEERINO CHANGE NOTICE MOD # Nb ib b ECNOObSH-OF DOCUM MM ENGR [#
R E V..O_.
ii MINOR MODIFICAT:
8
' VERIFMR APPR %W MM NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 VI.4 The annular space between the 1-1/2" diameter encapsulation pipe and the 1/2" diameter equalization line on both the reactor side and bonnet side of 12RWC-46-1 will be hydrostatically tested at 1.1 times the operating pressure as per ASME section XI article IWB-5200.
(1980 edition through Winter 1981 addenda).
VI.10 The Authorized Nuclear In-Service Inspector (ANII) will be notified of the repair and post work testing prior to beginning the repair.
An ASME Section XI Repair Program is issued with this modification for the repair activity.
The ANII will perform an overall review of the documentation at the conclusion of the modification.
VII.2 JAF ALARA has been notified.
6.0 DESIGN VERIFICATION:
DCM-4, Attachments 4.1 and 4.2 (attached) 7.0 EFFECT ON FINAL BAFETY ANALYSES REPORT The proposed minor modifications Require a change to the FSAR (x) Does
( ) Does Not Sections Affected Sect 16.5.5.1.5 NYPA FORM MCM-5, ATTACHMENT 4.2 Page 6 of _1L m
n'
.x JAP
_ NUCLEAR. POWER.PLANTT MINOR MODIFICATION-PACKAGE (MMP).
MM:NO.
M1-92-383 MM REV.
O NSE NO.-JAF-SE-92-220-8.0 NUCLEAR SAFETY EVALUATION The proposed minor Modification:
1.
() Does Increase the probability of occurrence (x) Does-Not or consequences of an accident or malfunction of equipment important to safety previously' evaluated in~the safety analysis report,-because this repair to the 1/2" diameter equalization line will be designed, installed, inspected and tested to codes that are equal to.or more conservative than the original code l
for this valve and piping assembly.
The: design considers potential
~
causes for the existing leak to preclude recurrence.
The-modification meets the requirements of the applicable section of the James A. FitzPatrick FSAR.
1 2.
( ) Does Create the possibility of an accident or
'l (x)-Does Not' malfunction of a.different type than any evaluated previously in the safety-analysis report,.because
-this modification will restore the; piping pressure boundary.
Then u
repair does not introduce a new failure mode to those which have-
.been,previously evaluated in
. Chapter 14-of~the James A.
i L
FitzPatrick FSAR.-
y 3.
( ) Does Reduce the margin of safety as-defined-(x) Does.Not in the-basis for any Technical u
Specification, because l
this modification will restore the-l piping pressure boundary.
[
Integrity of the installation will be verified-by examination and
-l l-testing.in_-accordance~with t
- applicable codes.
4.
(.F Does Involve an'unreviewed safety; question, (x) Does Not
' based on 1, 2 or 3 above.
NYPA FORM MCM-5,'ATTACKMENT 4.2.
- Page 7 of _J1_
JAF NUCLEARL POWER PLANT MINOR MODIFICATION PACKAGE (MMP)
MM NO.
M1-92-383 4
MM REV.
O NSE NO. JAF-SE-92-220 f
5.
( ) Does Involve a change in the Technical!
-(x) Does Not Specifications incorporated in the-license.
6.
( ) Does Require pre-implementation review by the (x) Does-Not NRC because It does not involve an unreviewed safety-question.
7.
( ) Does Degrade the Security Plan, Quality (x) Does Not Assurance Program:or the Fire Protection System, because the repair of 12RWC-46 equalization line.ls unrelated to Plant Security or Fire Protection._
The repair-will be performed in accordance with the Plant Quality Assurance Program.
B.
( ) Does Affect the environmental iopact of the (x).Does Not plant or involve an unreviewed environmental question.
REMMARY AND CONCLUSION OF NUCLEAR SAFETY EVALUATION This modification to repair the reactor _ water cleanup isolationy valve 12RWC-46 equalization line is_QA; Category-I.
The piping forms part of the reactor coolant pressure-boundary. _The repair
-of'l/2"-diameter schedule 80Lstainless: steel line by-3; encapsulating it with-a_ larger _ diameter-pipe-restores pressure
,~
boundary and reinforces the piping.
.The design, installation, examination and testing meets the requirements of.ASME;Section XI as described in the repair plan document. _Non Destructive L-Examination-was performed on all the.\\" equalization' piping fittings and welds.to confirm _the integrity of these: items.
No further. unacceptable defects exist in these items.
This repair will be removed at.the next-Aefuel Outage.
This modification does not create the possibility of an accident or malfunction previously evaluated _or offa different type than l'
any evaluated.in-the: Final Safety Analysis Report and is therefore acceptable.
l-NYPA FORM MCM-5,: ATTACHMENT 4.2 Page 8 :of._ll_
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e JAF NUCLEAR POWER PLANT MINOR MODIFICATION PACKAGE (MdP) -
MM NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 WR NO.
9.0 LIST OF AFFECTED DOCUMENTS Document Location Installation Identi fication (if not attached)
Document Yes No Design 6.37-231 Rev. Al-0 x
Documents 6.37-232 Rev. A3-0 x
FM-24A Rev. 378-0 X
Tech tiONE Manuals Plant OP-28 Procedures Other NONE MEL Yes x
-No NYPA FORM h0M-5, ATTACHMENT 4.2 Page 9 of _11_
MENGINEERIN CHANG 5" NOTI E~
MOD # M N b~ b ECNS9I.SH-b F
4
-O DOCUM# j C C EV O_
MINOR MODIFICA' ENGR M ".VERIF M APPR W MM-NO.
M1-92-383 MM REV.
O NSE NO. JAF-SE-92-220 10.0 SPECIAL ENGINEERING INSTALLATION REQUIREMENTS Repair equalization piping in accordance with the following requirements:
1.
Moisture from leakage through the existing 1/2" pipe shall be removed or sealed to ensure optimum conditions for welding.
2.
Preheat shall be strictly followed and h>4t input i
during the welding process shall be min 4 Sized and controlled so as to minimize residual stresses as a result of the welding operation.
3.
Weld sequencing shall be implemented to control weld shrinkage and potential distortions.
4.
Assure that means are available to stop or control unanticipated leakage during welding via catch containment and back seating of 12RWC-46 valve.
5.
After post work testing and NDE, replace insulation on 12RWC-46, 1/2" equalization piping and 12RWC-46-1 valve in accordance with IS-M-05.
11.0 BPECIAL MODIFICATION TEST REQUIREMENTS 11.1 Hydrotest bonnet side 1\\" encapsulation pipe at 1135 psig (1.1 x operating pressure) test pressure, 11.2 Hydrotest reactor side 1\\" encapsulation pipe at 1135 psig (1.1 x operating pressure) test pressure.
^
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NYPA FCR4 MCM-5, ATTACEMEET 4.2 Page 10 of _ll.
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4 JAF NUCLEAR POWER PLANT MINOR MODIFICATION PACKAGE (MMP)
MM NO.
liL 92-383 MM REV.
O NSE NO. JAF-SE-92-220 12.0 REFERENCE CALCULATIONS, REPORTS, ANALYBES, ETC.
12.1 MMP F1-84-053, Revision 0 12.2 WR # 106441 10/20/92 12.3 JAF Welding Manual 12.4 MDA-85-0493-A R/1 New York Power Authority Gate Valve Specification 12/1/86 12.5 FM-24A Rev. 37 12.6 FM-20A Rev. 34 12.7 FM-26A Rev. 29 12.8 FM-47A Rev. 29 12.9 Structural Integrity Associates Evaluation SIR-92-081 Rev. 1 "ASME III, NB Stress Report" 12.10 Structural Integrity Associates Evaluation SIR-92-082 Rev. 1 " Root Cause Failure Analysis" 12.11 Anchor Darling Calculation Order No. E6879-2 Seismic Analysis Report 12.12 JTS-92-1062 Assessment Of The Requirement For The Bonnet Equalizing Valve on 12RWC-46 13.0 ATTACHMENTS 13.1 ASME Section XI Repair Program for 12RWC-46 Bypass 13.2 DCM-13, Attachment 4.1 13.3 DCM-4, Attachment 4.1 and 4.2 NYPA FORM MCM-5, ATTACHMENT 4.2 Page _l L of _l L
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