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| number = ML102030120 | | number = ML102030120 | ||
| issue date = 07/15/2010 | | issue date = 07/15/2010 | ||
| title = | | title = Notification of Entergy'S Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 & 3 | ||
| author name = Bessette P | | author name = Bessette P | ||
| author affiliation = Entergy Nuclear Operations, Inc, Morgan, Lewis & Bockius, LLP | | author affiliation = Entergy Nuclear Operations, Inc, Morgan, Lewis & Bockius, LLP | ||
| addressee name = Lathrop K | | addressee name = Lathrop K, Mcdade L, Wardwell R | ||
| addressee affiliation = NRC/ASLBP | | addressee affiliation = NRC/ASLBP | ||
| docket = 05000247, 05000286 | | docket = 05000247, 05000286 | ||
| Line 16: | Line 16: | ||
=Text= | =Text= | ||
{{#Wiki_filter:Morgan, Lewis & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Tel: 202.739.3000 Fax: 202.739.3001 www.morganlewis.com | {{#Wiki_filter:Morgan, Lewis &Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Morgan Lewis COUNSELORS AT LAW Tel: 202.739.3000 Fax: 202.739.3001 www.morganlewis.com Kathryn M. Sutton DOCKETED Partner USNRC 202.739.5738 ksutton @morganiewis.com June 15, 2010 (4:45 p.m.) | ||
@morganlewis.com | PaulM. Bessette OFFICE OF SECRETARY Partner RULEMAKINGS AND 202.739.5796 ADJUDICATIONS STAFF pbessette @morganlewis.com July 15, 2010 Lawrence G. McDade, Chairman Dr. Richard E. Wardwell Dr. Kaye D. Lathrop Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Docket: Entergy Nuclear Operatgios,nc.(Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR and 50-286-LR RE: Notification of Entergy's Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3 | ||
is providing this notice to the Atomic Safety and Licensing Board ("Board") | |||
and the parties regarding Entergy's submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3 to the U.S. Nuclear Regulatory Commission | ==Dear Administrative Judges:== | ||
("NRC") on July 14, 2010. See NL-10-063, Letter from Fred Dacimo, Entergy, to NRC Document Control Desk, "Amendment 9 to License Renewal Application (LRA) -Reactor Vessel Internals Program" (July 14, 2010). A copy of NL- 10-063 is attached for your reference. | |||
Counsel is providing this notification insofar as the Reactor Vessel Internals Program may be relevant and material to admitted contention NYS-25.y,~LAT~-~ | Entergy Nuclear Operations, Inc. ("Entergy") is providing this notice to the Atomic Safety and Licensing Board ("Board") and the parties regarding Entergy's submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3 to the U.S. Nuclear Regulatory Commission ("NRC") on July 14, 2010. See NL-10-063, Letter from Fred Dacimo, Entergy, to NRC Document Control Desk, "Amendment 9 to License Renewal Application (LRA) - Reactor Vessel Internals Program" (July 14, 2010). A copy of NL- 10-063 is attached for your reference. | ||
~b2&~ | Counsel is providing this notification insofar as the Reactor Vessel Internals Program may be relevant and material to admitted contention NYS-25. | ||
Morgan Lewis | y,~LAT~-~ ~ b2&~ | ||
Office of | Morgan Lewis Lawrence G. McDade, Chairman COUNSELORS AT LAW Dr. Richard E. Wardwell Dr. Kaye D. Lathrop July 15, 2010 Page 2 | ||
'Resfuly submitted, Paul M. Bessette, Esq. | |||
.. | Counsel for Entergy Nuclear Operations, Inc. | ||
CBM/als Attachment cc: Service List | |||
gov) | |||
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and | |||
) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )) | |||
(Indian Point Nuclear Generating Units 2 and 3) ) | |||
.) July 15, 2010 CERTIFICATE OF SERVICE I hereby certify that copies of the letter entitled "Notification of Entergy's Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3," dated July 15, 2010, were served this 15th day of July, 2010 upon the persons listed below, by first class mail and e-mail as shown below. | |||
Administrative Judge Administrative Judge Lawrence G. McDade, Chair Kaye D. Lathrop Atomic Safety and Licensing Board Panel Atomic -Safety and Licensing Board Panel Mail Stop: T-3 F23 190 Cedar Lane E. | |||
U.S. Nuclear Regulatory Commission Ridgway, CO 81432 | |||
-WashingtonDC--2055 5-000O------- -(E-mail- kdt2@o)nrc-.gov ... -.. ....... | |||
(E-mail: larn @)nrc.gov) | |||
Administrative Judge Office of the Secretary* | |||
Richard E. Wardwell Attn: Rulemaking and Adjudications Staff Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Mail Stop: T-3 F23 Washington, D.C. 20555-0001 U.S. Nuclear Regulatory Commission (E-mail: hearingdocketanrc gov) | |||
Washington, DC 20555-0001 (E-mail: rew(&nrc.gov) | |||
Office of Commission Appellate Adjudication Josh Kirstein, Law Clerk U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Mail Stop: O-16G4 Mail Stop: T-3 F23 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission (E-mail: ocaamailanrc.gov) Washington, DC 20555-0001 (E-mail: Josh.Kirstein(anrc. gov) | |||
Page 2 Sherwin E. Turk, Esq. Greg Spicer, Esq. | |||
Beth N. Mizuno, Esq. Office of the Westchester County Attorney David E. Roth, Esq. 148 Martine Avenue, 6th Floor Brian G. Harris, Esq. White Plains, NY 10601 Andrea Z. Jones, Esq. (E-mail: gss I @(westchestergov.com) | |||
Office of the General Counsel Mail Stop:, 0-15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (E-mail: set(dnrc.gov) | |||
(E-mail: bniml@nrc.gov) | |||
(E-mail: david.roth(onrc.gov) | |||
(E-mail: brian.harris(nrc.gov) | |||
(E-mail: andrea.iones(dnrc.gov) | |||
Manna Jo Greene Thomas F. Wood, Esq. | |||
Environmental Director Daniel Riesel, Esq. | |||
Hudson River Sloop Clearwater, Inc. Ms. Jessica Steinberg, J.D. | |||
724 Wolcott Avenue Sive, Paget & Riesel, P.C. | |||
Beacon, NY 12508 460 Park Avenue (E-mail: mannaj o(&-clearwater.org) New York, NY 10022 (E-mail: driesel(asprlaw.com) | |||
(E-mail: isteinberg(sprlaw.com) | |||
Stephen C. Filler, Board Member John Louis Parker, Esq. | |||
-Hudson-River-Sloop--ClearwaterInc. -------Regi6nal-Attomey- - | |||
303 South Broadway, Suite 222 Office of General Counsel, Region 3 Tarrytown, NY 10591 NYS Dept. of Environmental Conservation (E-mail: sfilleranylawline.com) 21 S. Putt Comers Road New Paltz, New York 12561-1620 (E-mail: ilparker(ogw.dec.state.ny.us) | |||
Ross Gould, Member Michael J. Delaney, V.P. - Energy Hudson River Sloop Clearwater, Inc. New York City Economic Development 10 Park Avenue, #5L Corp. | |||
New York, NY 10016 110 William Street (E-mail: rgouldesq(agrmail.com) New York, NY 10038 (E-mail: mdelaney(cnycedc.com) | |||
Page 3 Phillip Musegaas, Esq.: Daniel E. O'Neill, Mayor Deborah Brancato, Esq. James Siermarco, M.S. | |||
Riverkeeper, Inc. Liaison to Indian Point 828 South Broadway Village of Buchanan Tarrytown, NY 10591 Municipal Building (E-mail: phillip(iriverkeeper.org) 236 Tate Avenue (E-mail: dbrancato(riverkeeper.org) Buchanan, NY 10511-1298 (E-mail: vob(bestweb.net) | |||
Robert D. Snook, Esq. Mylan L. Denerstein, .Esq. | |||
Assistant Attorney General Executive Deputy Attorney General, Office of the Attorney General Social Justice State of Connecticut Office of the Attorney General 55 Elm Street -of the State of New-York-P.O. Box 120 120 Broadway, 25th Floor Hartford, CT 06141-0120 New York, New York 10271 (E-mail: Robert. Snook(dpo.state.ct.us) (E-mail: Mvlan.Denerstein(ioag. state.ny. us) | |||
Andrew M. Cuomo, Esq. . Janice A. Dean Attorney General of the State of New York Office of the Attorney General John J. Sipos, Esq. of the State of New York Charlie Donaldson Esq. Assistant Attorney General Assistants Attorney General 120 Broadway, 26th Floor The Capitol New York, New York 10271 (E-Albany,:N¥-t-2-2-24-03ta..... .) -- (E-mail --Janice;Dean(aoag~stateny.,us) . | |||
(E-mail: john.sipos(ýoag.state.ny.us) | |||
Joan Leary Matthews, Esq. | |||
Senior Attorney for Special Projects Office of the General Counsel New York State Department of Environmental:Conservation 625 Broadway, 14th Floor Albany, NY 12207 (E-mail: ilmattheagw. dec. state.ny.us) | |||
Page 4 Original and 2 copies provided to the Office of the Secretary. | Page 4 Original and 2 copies provided to the Office of the Secretary. | ||
Paul M. Bessette, Esq.Counsel for Entergy Nuclear Operations, Inc.DB 1/65220145.1 | Paul M. Bessette, Esq. | ||
Counsel for Entergy Nuclear Operations, Inc. | |||
DB 1/65220145.1 | |||
Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB Entergy P.O. Box 249 Buchanan, NY 105111-0249 Tel (914) 788-2055 Fred Dacimno Vice President License Renewal NL-10-063 July 14, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 | |||
==SUBJECT:== | ==SUBJECT:== | ||
Amendment 9 to License Renewal Application (LRA) - | |||
Reactor Vessel Internals Program Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 | |||
==REFERENCES:== | ==REFERENCES:== | ||
: 1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039) | |||
: 2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings (NL-07-040) | |||
: 2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings (NL-07-040)3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References (NL-07-041) | : 3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References (NL-07-041) | ||
: 4. Entergy Letter dated October 11, 2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07-124) | : 4. Entergy Letter dated October 11, 2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07-124) | ||
: 5. Entergy Letter November 14, 2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA)Environmental Report References" (NL-07-133) | : 5. Entergy Letter November 14, 2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA) | ||
Environmental Report References" (NL-07-133) | |||
==Dear Sir or Madam:== | ==Dear Sir or Madam:== | ||
In the referenced letters, Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Energy Center operating license. This letter contains Amendment 9 to the License Renewal Application (LRA) regarding the Reactor Vessel Internals Program.If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710. | |||
NL- 10-063 Docket Nos. 50-247 & 50-286 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on Sincerely, FRD/dmt | In the referenced letters, Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Energy Center operating license. This letter contains Amendment 9 to the License Renewal Application (LRA) regarding the Reactor Vessel Internals Program. | ||
If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710. | |||
NL- 10-063 Docket Nos. 50-247 & 50-286 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on Sincerely, FRD/dmt | |||
==Attachment:== | ==Attachment:== | ||
: 1. Amendment 9 to License Renewal Application | : 1. Amendment 9 to License Renewal Application - | ||
Reactor Vessel Internals Program cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. John Boska, NRR Senior Project Manager Ms. Kimberly Green, Project Manager NRC Resident Inspector's Office Mr. Paul Eddy, New York State Department of Public Service Mr. Francis J. Murray, President and CEO, NYSERDA | |||
-added, strikethrough | ATTACHMENT 1 TO NL-10-063 Amendment 9 to License Renewal Application - | ||
-deleted)2.3.1.2 Reactor Vessel Internals The reactor vessel internals for each unit are described in the reactor coolant system description (Unit 2, Reactor Vessel Internals; Unit 3, Reactor Vessel Internals). | Reactor Vessel Internals Program ENTERGY NUCLEAR OPERATIONS, INC. | ||
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 LICENSE NOS. DPR-26 AND DPR-64 | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 1 of 90 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA) | |||
AMMENDMENT 9 The LRA is revised as described below. (underline - added, strikethrough - deleted) 2.3.1.2 Reactor Vessel Internals The reactor vessel internals for each unit are described in the reactor coolant system description (Unit 2, Reactor Vessel Internals; Unit 3, Reactor Vessel Internals). | |||
For both units, the lower core support structure, the upper core support structure, and the incore instrumentation support structure are the three major parts of the reactor internals. | For both units, the lower core support structure, the upper core support structure, and the incore instrumentation support structure are the three major parts of the reactor internals. | ||
Lower Core Support Structure The major member of the reactor vessel internals is the lower core support structure consisting of the following components included in this evaluation. | Lower Core Support Structure The major member of the reactor vessel internals is the lower core support structure consisting of the following components included in this evaluation. | ||
core baffle/former assembly: | core baffle/former assembly: bolts core baffle/former assembly: plates core barrel assembly: bolts, screws core barrel assembly: axial flexure plates (thermal shield flexures), flange, ring, shell, thermal shield, lower core barrel flange weld, upper core barrel flange weld core barrel assembly: outlet nozzles lower internals assembly: clevis insert bolt lower internals assembly: clevis insert lower internals assembly: intermediate diffuser plate lower internals assembly: fuel alignment pin lower internals assembly: lower core plate lower internals assembly: lower core support plate column sleeves lower internals assembly: lower core support column bolt lower internals assembly, lower core support column castings: column cap, lower core support lower internals assembly: radial key lower internals assembly: secondary core support (energy absorbing device) specimen guides (not subject to aging management review) specimen plugs (installed in IP2 only; not subject to aging management review) | ||
bolts core baffle/former assembly: | |||
plates core barrel assembly: | NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 90 The lower core support structure is supported at its upper flange from a ledge in the reactor vessel. Within the core barrel are a core baffle and a lower core plate, both of which are attached to the core barrel wall. The lower core support structure provides passageways for the coolant flow. The lower core plate at the bottom of the core below the baffle plates provides support and orientation for the fuel assemblies. Fuel alignment pins (two for each assembly) are also inserted into this plate. Columns are placed between the lower core plate and core support casting in order to provide stiffness and to transmit the core load to the core support casting. Adequate coolant distribution is obtained through the use of the lower core plate and a diffuser plate. | ||
bolts, screws core barrel assembly: | Upper Core Support Structure The "top hat with deep beam features" upper core support structure consists of the following components included in this evaluation. | ||
axial flexure plates (thermal shield flexures), flange, ring, shell, thermal shield, lower core barrel flange weld, upper core barrel flange weld core barrel assembly: | upper internals assembly, rod control cluster assembly (RCCA) guide tube assembly: bolts upper internals assembly, RCCA guide tube assembly: guide tube (including lower flange weld), guide plates upper internals assembly, RCCA guide tube assembly: support pin upper internals assembly: core plate alignment pin upper internals assembly: head/vessel alignment pin upper internals assembly: hold-down spring upper internals assembly: support column upper internals assembly, mixing devices: support column orifice base, support column mixer upper internals assembly: upper core plate, fuel alignment pin upper internals assembly: support assembly (including ring), upper support plate upper internals assembly: upper support column bolt The support columns establish the spacing between the upper support assembly and the upper core plate and are fastened at top and bottom to these plates and beams. | ||
outlet nozzles lower internals assembly: | The RCCA guide tube assemblies shield and guide the control rod drive shafts and control rods. | ||
clevis insert bolt lower internals assembly: | They are fastened to the upper support and are guided by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the control rod shroud tube which is attached to the upper support plate and guide tube. | ||
clevis insert lower internals assembly: | In-Core Instrumentation Support Structure The in-core instrumentation support structures consist of the following components included in this evaluation. | ||
intermediate diffuser plate lower internals assembly: | thermocouple conduit flux thimble guide tube bottom mounted instrumentation column | ||
fuel alignment pin lower internals assembly: | |||
lower core plate lower internals assembly: | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 90 An upper system (thermocouple conduit) is used to convey and support thermocouples penetrating the vessel through the head, and a lower system (flux thimble guide tube) is used to convey and support flux thimbles penetrating the vessel through the bottom. | ||
lower core support plate column sleeves lower internals assembly: | The upper system utilizes the reactor vessel head penetrations. Instrumentation port columns are slip-connected to in-line columns that are in turn fastened to the upper support plate. These port columns protrude through the head penetrations. The thermocouples are carried through these port columns and the upper support plate at positions above their readout locations. The thermocouple conduits are supported from the columns of the upper core support system. | ||
lower core support column bolt lower internals assembly, lower core support column castings: | Table 2.3.1-2-1P2 and Table 2.3.1-2-1P3 list the mechanical components subject to aging management review and component intended functions for the reactor vessel internals. | ||
column cap, lower core support lower internals assembly: | |||
radial key lower internals assembly: | |||
secondary core support (energy absorbing device)specimen guides (not subject to aging management review)specimen plugs (installed in IP2 only; not subject to aging management review) | |||
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 90 The lower core support structure is supported at its upper flange from a ledge in the reactor vessel. Within the core barrel are a core baffle and a lower core plate, both of which are attached to the core barrel wall. The lower core support structure provides passageways for the coolant flow. The lower core plate at the bottom of the core below the baffle plates provides support and orientation for the fuel assemblies. | |||
Fuel alignment pins (two for each assembly)are also inserted into this plate. Columns are placed between the lower core plate and core support casting in order to provide stiffness and to transmit the core load to the core support casting. Adequate coolant distribution is obtained through the use of the lower core plate and a diffuser plate.Upper Core Support Structure The "top hat with deep beam features" upper core support structure consists of the following components included in this evaluation. | |||
upper internals assembly, rod control cluster assembly (RCCA) guide tube assembly: | |||
bolts upper internals assembly, RCCA guide tube assembly: | |||
guide tube (including lower flange weld), guide plates upper internals assembly, RCCA guide tube assembly: | |||
support pin upper internals assembly: | |||
core plate alignment pin upper internals assembly: | |||
head/vessel alignment pin upper internals assembly: | |||
hold-down spring upper internals assembly: | |||
support column upper internals assembly, mixing devices: support column orifice base, support column mixer upper internals assembly: | |||
upper core plate, fuel alignment pin upper internals assembly: | |||
support assembly (including ring), upper support plate upper internals assembly: | |||
upper support column bolt The support columns establish the spacing between the upper support assembly and the upper core plate and are fastened at top and bottom to these plates and beams.The RCCA guide tube assemblies shield and guide the control rod drive shafts and control rods.They are fastened to the upper support and are guided by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the control rod shroud tube which is attached to the upper support plate and guide tube.In-Core Instrumentation Support Structure The in-core instrumentation support structures consist of the following components included in this evaluation. | |||
thermocouple conduit flux thimble guide tube bottom mounted instrumentation column NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 90 An upper system (thermocouple conduit) is used to convey and support thermocouples penetrating the vessel through the head, and a lower system (flux thimble guide tube) is used to convey and support flux thimbles penetrating the vessel through the bottom.The upper system utilizes the reactor vessel head penetrations. | |||
Instrumentation port columns are slip-connected to in-line columns that are in turn fastened to the upper support plate. These port columns protrude through the head penetrations. | |||
The thermocouples are carried through these port columns and the upper support plate at positions above their readout locations. | |||
The thermocouple conduits are supported from the columns of the upper core support system.Table 2.3.1-2-1P2 and Table 2.3.1-2-1P3 list the mechanical components subject to aging management review and component intended functions for the reactor vessel internals. | |||
Table 3.1.2-2-1P2 and Table 3.1.2-2-1P3 provide the results of the aging management review for the reactor vessel internals. | Table 3.1.2-2-1P2 and Table 3.1.2-2-1P3 provide the results of the aging management review for the reactor vessel internals. | ||
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 90 Table 2.3.1-4-1P2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Intended Function Lower Core Support Structure, Core baffle/former assembly Structural support.bolts Core baffle/former assembly Structural support-plates Flow distribution Shielding Core barrel assembly Structural support-bolts and screws Core barrel assembly Structural support-axial flexure plates Floherm | |||
* cast austenitic stainless steel* nickel alloy* stainless steel Environment Reactor vessel internals components are exposed to the following environments. | NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 90 Table 2.3.1-4-1P2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Intended Function Lower Core Support Structure, Core baffle/former assembly Structural support | ||
* neutron fluence* treated borated water* treated borated water > 140°F 0 treated borated water > 4820F Aging Effects Requiring Management The following aging effects associated with the reactor vessel internals require management. | .bolts Core baffle/former assembly Structural support | ||
0 change in dimensions | -plates Flow distribution Shielding Core barrel assembly Structural support | ||
* cracking* cracking -fatigue* loss of material* loss of material -wear* loss of preload* reduction of fracture toughness Aging Management Programs The following aging management programs manage the aging effects for reactor vessel internals components. | -bolts and screws Core barrel assembly Structural support | ||
0 Inservice Inspection 0 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)0 Reactor Vessel Internals Water Chemistry Control -Primary and Secondary NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 90 3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Loss of fracture toughness due to neutron irradiation embrittlement and change in dimensions (void swelling) e 4d eccur in stainless steel and nickel alloy reactor vessel internals components exposed to reactor coolant and neutron flux will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will imnlp~m~nt thp. FPRI PrP..I.iri7Pr1 W~tp~r Rp irtnr Intp~rn~lk FIiinhtinfir | -axial flexure plates dltributi Floherm | ||
To manage loss ofra;ctr,+e toughnoss in ve...l inte-rnals; | *flaRge Sil~ | ||
.components,-PEG ,il11l (1 par--ici-te in he industry pr.grams for in.'Stigating and managing aing cie n rcacto (2) cvaluato and implement esults of the.. | -thermal shield Core barrel assembly Structural support | ||
(> 700°F) as stated in the ASME Code, Section II, Part D, Table 4. No IPEC internals components operate at > 700'F. Therefore, loss of preload due to thermal stress relaxation (creep) is not an applicable aging effect for the reactor vessel internals components. | *axial flexure plates (thermal shield flexures) | ||
However, irradiation-enhanced creep (irradiation creep) or irradiation enhanced stress relaxation (ISR) is an athermal process that depends on the neutron fluence and stress: and, on void swelling if present. Neve.4heless Therefore, loss of preload of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. | Core barrel assembly Structural support Core barrel assembly Structural support | ||
to the extent that industry developed reactor vessel internals aging _managemenqt programs address thes a- ing effeGts. The I'PEC commitmert t these RVI programs iS included in IFSAR Su"pplement, Appendix A, Sections A.2. 1.1 and A.3. 1. 1.Changes in Dimensions due to Void Swelling Changes in dimensions due to void swelling ee'-ld .e... in stainless steel and nickel alloy reactor internal components exposed to reactor coolant will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. | " ringl Flow distribution | ||
T---maiage hanges in dim.en ....Of vessel ' components, 'PEG will () participate the .,dust,' Ip-rogram for investigating and ma _an .g en reacteo inRternals; (2) evaluate and implement the results Of the industry programs as NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 90 appI"cable to the reactor intornals; and (3) upon completion Of these programns, but not lcss thaR 241 mofths before enterFig the period of exterded operation;, subMit aR | " shell Shielding | ||
ThiS cOMMItmont isicue nthe UJFSAR Supplement, Appendix A, SectionRs A.2.1.41 and A.3.!.4!4 3.1.2.2.17 Cracking due to Stress Corrosion Cracking, Primary Water Stress Corrosion Cracking, and Irradiation-Assisted Stress Corrosion Cracking Cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) Gould GGeu in PWR stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. | " thermal shield Core barrel assembly Structural support | ||
To manage crackig in Vessel internals com.ponents, IPEC maintains the Water C-hemistr,' | " lower core barrel flange weld | ||
Control Primary and SeconRdary ProgramR and Will (1 ) pa~ticipate in the industry programns for investigating and m:anagfing aging effects-on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion o~f those programs, but - | " upper core barrel flange weld | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 10 of 90 Table 3.1.1 Summary of Aging Management Programs for the Reactor Coolant System Evaluated in Chapter IV of NUREG-1 801 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee Consistont with NUREG 1801. Loss of nickel alloy reactor fracture commitment to (1) commitment to -fracture toughness of stainless steel vessel internals toughness due participate in be confirmed and nickel alloy reactor vessel components exposed to to neutron industry RVI aging internals components will be managed reactor coolant and irradiation programs (2) by the Reactor Vessel Internals neutron flux embrittlement, implement ProQram. aging manago.m..t | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 90 Core barrel assembly Flow distribution | ||
-outlet nozzles Lower internals assembly Structural support | |||
*clevis insert bolt | |||
*clevis insert | |||
-fuel alignment pin | |||
-lower core support plate column sleeves | |||
-lower core support plate column bolt | |||
-radial key Lower internals assembly Flow distribution | |||
-intermediate diffuser plate Lower internals assembly Structural support | |||
-lower core plate Flow distribution | |||
-lower core support castings | |||
-column cap | |||
-lower core support | |||
-secondary core support Upper Core Support Structure-UpperInternals Assembly RCCA guide tube assembly Structural support | |||
-bolt RCCA guide tube assembly Structural support | |||
-guide tube (including lower flange welds) | |||
RCCA guide tube assembly Structural support | |||
-guide plates RCCA guide tube assembly Structural support | |||
-support pin | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 90 Core plate alignment pin Structural support Head / vessel alignment pin Structural support Hold-down spring Structural support Mixing devices Structural support | |||
-support column orifice base Flow distribution | |||
-support column mixer Support column Structural support Upper core plate, fuel alignment Structural support pin Flow distribution Upper support plate, support Structural support assembly (including ring) | |||
Upper support column bolt Structural support Inc~orInstrumentationSuppflt ~Structure Bottom mounted instrumentation Structural support column Flux thimble guide tube Structural support Thermocouple conduit Structural support | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 7 of 90 3.1.2.1.2 Reactor Vessel Internals Materials Reactor vessel internals components are constructed of the following materials. | |||
* cast austenitic stainless steel | |||
* nickel alloy | |||
* stainless steel Environment Reactor vessel internals components are exposed to the following environments. | |||
* neutron fluence | |||
* treated borated water | |||
* treated borated water > 140°F 0 treated borated water > 4820F Aging Effects Requiring Management The following aging effects associated with the reactor vessel internals require management. | |||
0 change in dimensions | |||
* cracking | |||
* cracking - fatigue | |||
* loss of material | |||
* loss of material - wear | |||
* loss of preload | |||
* reduction of fracture toughness Aging Management Programs The following aging management programs manage the aging effects for reactor vessel internals components. | |||
0 Inservice Inspection 0 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) 0 Reactor Vessel Internals Water Chemistry Control - Primary and Secondary | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 90 3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Loss of fracture toughness due to neutron irradiation embrittlement and change in dimensions (void swelling) e 4d eccur in stainless steel and nickel alloy reactor vessel internals components exposed to reactor coolant and neutron flux will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will imnlp~m~nt thp. FPRI PrP..I.iri7Pr1 W~tp~r Rp irtnr Intp~rn~lk In_*n*_nfinn *nd FIiinhtinfir Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aginq effects for reactor vessel internals. To manage loss ofra;ctr,+e toughnoss in ve...l inte-rnals; . components, | |||
-PEG ,il11l (1 par--ici-te in he industry pr.grams for in.'Stigating and managing aing cie n rcacto ffi*ternals; (2) cvaluato and implement* th* esults of the dust. | |||
.. pregFams as applicable to the.oacto. internals; and (3) upon cmR plet!on these programs, but not less than 24 moneths before entering the period of ex~te~nded operation, submit an inspection plan for ýeac~ter inernals to the NRC for review and ap~proval.-Th~is rsenoUm.tMont isicue nthe UFSAR Supplement, Appendix A, Sections A.2.!.41 ard A.3.1.41. | |||
3.1.2.2.9 Loss of Preload due to Stress Relaxation Loss of preload due to thermal stress relaxation (creep) would only be a concern in very high temperature applications (> 700°F) as stated in the ASME Code, Section II, Part D, Table 4. No IPEC internals components operate at > 700'F. Therefore, loss of preload due to thermal stress relaxation (creep) is not an applicable aging effect for the reactor vessel internals components. However, irradiation-enhanced creep (irradiation creep) or irradiation enhanced stress relaxation (ISR) is an athermal process that depends on the neutron fluence and stress: and, on void swelling if present. Neve.4heless Therefore, loss of preload of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. to the extent that industry developed reactor vessel internals aging _managemenqt programs address thes a- ing effeGts. The I'PEC commitmert t these RVI programs iS included in IFSAR Su"pplement, Appendix A, Sections A.2. 1.1 and A.3. 1. 1. | |||
3.1.2.2.15 Changes in Dimensions due to Void Swelling Changes in dimensions due to void swelling ee'-ld .e... in stainless steel and nickel alloy reactor internal components exposed to reactor coolant will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. T---maiage hanges in dim.en .... Of vessel inters*+ ' components, 'PEG will () participate the .,dust,' Ip-rogram for investigating and *an ma _an . g effe*t* en reacteo inRternals; (2) evaluate and implement the results Of the industry programs as | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 90 appI"cable to the reactor intornals; and (3) upon completion Of these programns, but not lcss thaR 241 mofths before enterFig the period of exterded operation;, subMit aR plan for reactor iRn8trals to the INIR for review and approIval. ThiS SnspectioR cOMMItmont isicue nthe UJFSAR Supplement, Appendix A, SectionRs A.2.1.41 and A.3.!.4!4 3.1.2.2.17 Cracking due to Stress Corrosion Cracking, Primary Water Stress Corrosion Cracking, and Irradiation-Assisted Stress Corrosion Cracking Cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) Gould GGeu in PWR stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. To manage crackig in Vessel internals com.ponents, IPEC maintains the Water C-hemistr,' Control Primary and SeconRdary ProgramR and Will (1) pa~ticipate in the industry programns for investigating and m:anagfing aging effects-on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion o~f those programs, but -otless than 2;* Fmonths before entering the perFod of cxtcnded ohperation, SubMit an inspect*i plan for reacRIIto Ine 1Ir a to the NRC fo review and approval: The | |||
'PEC commFitmen~t to these RVI programns is included in UFSAR Supplement, Appen~dix A, Sections A.2.1 .41 and A.3.1. .1. | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 10 of 90 Table 3.1.1 Summary of Aging Management Programs for the Reactor Coolant System Evaluated in Chapter IV of NUREG-1 801 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee Consistont with NUREG 1801. Loss of nickel alloy reactor fracture commitment to (1) commitment to -fracture toughness of stainless steel vessel internals toughness due participate in be confirmed and nickel alloy reactor vessel components exposed to to neutron industry RVI aging internals components will be managed reactor coolant and irradiation programs (2) by the Reactor Vessel Internals neutron flux embrittlement, implement ProQram. aging manago.m..t - | |||
void swelling applicable results prorams. The commitmeRnt to these (3) submit for NRC RVI progra.m is,includ ,inFSAR approval > 24 Supplement, Appendix A, Sections months before the A.2.1.41 and A.3. .14. | |||
extended period an See Section 3.1.2.2.6. | |||
RVI inspection plan based on industry recommendation. | RVI inspection plan based on industry recommendation. | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further temComponent Ag ism Management Evaluation Discussion Programs Recommended 3.1.1-27 Stainless steel and Loss of FSAR supplement No, but licensee Loss of preloead duo to stross nickel alloy reactor preload due to commitment to (1) commitment to n (Goop) is a concern for vessel internals screws, stress participate in be confirmed application. | |||
at t.mpe.atures higher bolts, tie rods, and hold- relaxation industry RVI aging than thoso of 'PEC reactor vessel and down springs programs (2) intornals .OMPen.nts. | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further temComponent Ag ism Management Evaluation Discussion Programs Recommended 3.1.1-27 Stainless steel and Loss of FSAR supplement No, but licensee Loss of preloead duo to stross nickel alloy reactor preload due to commitment to (1) commitment to relaxatio* n (Goop) is a concern for vessel internals screws, stress participate in be confirmed application. at t.mpe.atures higher bolts, tie rods, and hold- relaxation industry RVI aging than thoso of 'PEC reactor vessel and down springs programs (2) intornals.OMPen.nts. Therfore, loo implement of prFolad duo to tr*oss rolaxation applicable results 9..p) is not an applicable aging (3) submit for NRC effcct for tho roactOr 'cssel internals approval > 24 compononts. Neverthoeoss, loss of months before the preload of stainless steel and nickel extended period an alloy reactor vessel internals RVI inspection plan components will be managed by the based on industry Reactor Vessel Internals Proqram. | ||
Therfore, loo implement of prFolad duo to rolaxation applicable results 9..p) is not an applicable aging (3) submit for NRC effcct for tho roactOr 'cssel internals approval > 24 compononts. | recommendation. c.nsisont with , du*, ty devel oped rcactE)r vessel intcrnals aging anaRgqe*meRt prorams The comm tment to these RVI programs is rinluded iR UISAR SuppmonlRt, Apperndix A, RSortions A2. 1 41a;*d A. 341. | ||
Neverthoeoss, loss of months before the preload of stainless steel and nickel extended period an alloy reactor vessel internals RVI inspection plan components will be managed by the based on industry Reactor Vessel Internals Proqram.recommendation. | See Section 3.1.2.2.9. | ||
c.nsisont with , ty devel oped rcactE)r vessel intcrnals aging prorams The comm tment to these RVI programs is rinluded iR UISAR SuppmonlRt, Apperndix A, RSortions A2. 1 | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-30 Stainless steel reactor | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-30 Stainless steel reactor Cracking due Water Chemistry No, but licensee Coneictont'.wth NUREG 1801. | ||
Cracking due | vessel internals to stress and FSAR commitment Cracking _of stainless steel reactor components (e.g., corrosion supplement needs to be vessel internals components will be Upper internals cracking, commitment to (1) confirmed managed by the Water Chemistry assembly, RCCA guide irradiation- participate in Control - Primary and Secondary tube assemblies, assisted stress industry RVI aging Program and either the Reactor Vessel Baffle/former assembly, corrosion programs (2) Internals Proqram or the Inservice Lower internal cracking' implement Inspection Progqram. by-etheF RVt assembly, shroud applicable results nli, | ||
No, but licensee | -- --- ,rnm ni r-nr-~ Tka assemblies, Plenum (3) submit for NRC commitment to these other RVI cover and plenum approval > 24 programs is iAncrAlud-ed ipn. U lFS.A.Rn cylinder, Upper grid months before the Supplement, Appendix A, SectionS assembly, Control rod extended period an A. 2.1.4 aRd A."3.141. | ||
by-etheF RVt-- | guide tube (CRGT) RVI inspection plan See Section 3.1.2.2.12. | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-33 Stainless steel and Changes in FSAR supplement No, but licensee Consistent With NUREG 180!.nickel alloy reactor dimensions commitment to (1) commitment to Changes in dimensions of stainless vessel internals due to void participate in be confirmed steel and nickel alloy reactor vessel components swelling industry RVI aging internals components will be managed programs (2) by the Reactor Vessel Internals implement Proqram. RVI aging managmeont applicable results pogram.s. | assembly, Core support based on industry shield assembly, Core recommendation. | ||
The co. to these (3) submit for NRC RVI programs is included | barrel assembly, Lower grid assembly, Flow distributor assembly, Thermal shield, Instrumentation support structures) | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-33 Stainless steel and Changes in FSAR supplement No, but licensee Consistent With NUREG 180!. | |||
nickel alloy reactor dimensions commitment to (1) commitment to Changes in dimensions of stainless vessel internals due to void participate in be confirmed steel and nickel alloy reactor vessel components swelling industry RVI aging internals components will be managed programs (2) by the Reactor Vessel Internals implement Proqram. RVI aging managmeont applicable results pogram.s. The co. mmit*m*t to these (3) submit for NRC RVI programs is included inUSAR approval > 24 Supplen*.t, App.ndix A, Soctions months before the A.2.1 .41 and A.3.1 .41. | |||
extended period an See Section 3.1.2.2.15. | |||
RVI inspection plan based on industry recommendation. | RVI inspection plan based on industry recommendation. | ||
NL-1.0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-37 Stainless steel and Cracking due Water Chemistry No, but licensee Consistont with NUREG 1801.nickel alloy reactor to stress and FSAR commitment Cracking _of stainless steel and nickel vessel internals corrosion supplement needs to be alloy reactor vessel internals components (e.g., cracking, commitment to (1) confirmed components will be managed by the Upper internals primary water participate in Water Chemistry Control -Primary assembly, RCCA guide stress industry RVI aging and Secondary Program and either tube assemblies, Lower corrosion programs (2) the Reactor Vessel Internals Program internal assembly, CEA cracking, implement or the Inservice Inspection Program.shroud assemblies, irradiation-applicable results by othor RV! aging manag.m.nt Core shroud assembly, assisted stress (3) submit for NRC programs. | |||
The commitment to these Core support shield corrosion approval > 24 other RVI program. is inc.udod in assembly, Core barrel cracking months before the UF SAR SuppleMeRt, Appendix A, assembly, Lower grid extended period an Sections A.2.1.41 aRnd A.3.1.41.assembly, Flow RVI inspection plan See Section 3.1.2.2.17. | NL-1.0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-37 Stainless steel and Cracking due Water Chemistry No, but licensee Consistont with NUREG 1801. | ||
distributor assembly) based on industry recommendation. | nickel alloy reactor to stress and FSAR commitment Cracking _of stainless steel and nickel vessel internals corrosion supplement needs to be alloy reactor vessel internals components (e.g., cracking, commitment to (1) confirmed components will be managed by the Upper internals primary water participate in Water Chemistry Control - Primary assembly, RCCA guide stress industry RVI aging and Secondary Program and either tube assemblies, Lower corrosion programs (2) the Reactor Vessel Internals Program internal assembly, CEA cracking, implement or the Inservice Inspection Program. | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 I Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item- Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-63 Steel reactor vessel Loss of Inservice Inspection No The Inservice Inspection Program and flange, stainless steel material due to (IWB, IWC, and the Reactor Vessel Internals Program and nickel alloy reactor wear IWD) manages loss of-material due to wear vessel internals of the steel reactor vessel flange and exposed to reactor stainless steel and nickel alloy reactor coolant (e.g., upper and vessel internals components. | shroud assemblies, irradiation- applicable results by othor RV! aging manag.m.nt Core shroud assembly, assisted stress (3) submit for NRC programs. The commitment to these Core support shield corrosion approval > 24 other RVI program. is inc.udod in assembly, Core barrel cracking months before the UF SAR SuppleMeRt, Appendix A, assembly, Lower grid extended period an Sections A.2.1.41 aRnd A.3.1.41. | ||
assembly, Flow RVI inspection plan See Section 3.1.2.2.17. | |||
distributor assembly) based on industry recommendation. | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 I | |||
Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item- Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-63 Steel reactor vessel Loss of Inservice Inspection No The Inservice Inspection Program and flange, stainless steel material due to (IWB, IWC, and the Reactor Vessel Internals Program and nickel alloy reactor wear IWD) manages loss of-material due to wear vessel internals of the steel reactor vessel flange and exposed to reactor stainless steel and nickel alloy reactor coolant (e.g., upper and vessel internals components. | |||
lower internals assembly, CEA shroud assembly, core support barrel, upper grid assembly, core support shield assembly, lower grid assembly) | lower internals assembly, CEA shroud assembly, core support barrel, upper grid assembly, core support shield assembly, lower grid assembly) | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 16 of 90 NO)ITEs FOR TABLES 3.1.2 1 I-2 THROU | |||
G. Environment not in NUREG-1801 for this component and material.H. Aging effect not in NUREG-1 801 for this component, material and environment combination. | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 16 of 90 NO)ITEs FOR TABLES 3.1.2 1 I-2 THROU U3.1.24I 1P3 H | ||
Generic Notes A. Consistent with NUREG-1 801 item for component, material, environment, aging effect and aging management program. | |||
AMP is consistent with NUREG-1801 AMP. | |||
B. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program. | |||
AMP has exceptions to NUREG-1801 AMP. | |||
C. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP is consistent with NUREG-1 801 AMP. | |||
D. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP has exceptions to NUREG-1801 AMP. | |||
E. Consistent with NUREG-1801 material, environment, and aging effect but a different aging management program is credited. | |||
F. Material not in NUREG-1 801 for this component. | |||
G. Environment not in NUREG-1801 for this component and material. | |||
H. Aging effect not in NUREG-1 801 for this component, material and environment combination. | |||
I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable. | I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable. | ||
J. Neither the component nor the material and environment combination is evaluated in NUREG-1 801.Plant-Specific Notes 101. This component, material, environment and aging effect combination is considered in the Reactor Vessel Internals Program.As documented in MRP-227, the basis for the RVI Program, this combination warrants no additional aging management. | J. Neither the component nor the material and environment combination is evaluated in NUREG-1 801. | ||
NUREG 1801, Sectio-n XI.51 6 ctatj: "No fZrthr f ging manargemen reieW is ncessa,', if the applicant dGos!a'co)mmitm~enti the ESAR supplepmont to (1) pa~ticipate in the indIASt~' | Plant-Specific Notes 101. This component, material, environment and aging effect combination is considered in the Reactor Vessel Internals Program. | ||
programs foriG stgtn and manaaging aging offoctS on reactor inteFRnals; (2) evaluate and implcment the results of the industr,' | As documented in MRP-227, the basis for the RVI Program, this combination warrants no additional aging management. | ||
programs as applicable to the reactor internals,ý and (3) uIpon co;plotion of these program~s, bu6t not less than 24 mon)ths bePfo-ro-enrtering the period- of extenAdod operation, submit an npcinplnfreco intorn-als to the NRC f4or reView and approval." IPEC commRF:itmen-t c-an be foud.i Appendix A (JF=SAR suipplement) | NUREG 1801, Sectio-n XI.51 6 ctatj: "No fZrthr f ging manargemen reieW is ncessa,',if the applicant po*r dGos!a' co)mmitm~enti the ESAR supplepmont to (1) pa~ticipate in the indIASt~' programs foriG stgtn and manaaging aging offoctS on reactor inteFRnals; (2) evaluate and implcment the results of the industr,' programs as applicable to the reactor internals,ý and (3)uIpon co;plotion of these program~s, bu6t not less than 24 mon)ths bePfo-ro- enrtering the period-of extenAdod operation, submit an npcinplnfreco intorn-als to the NRC f4or reView and approval." IPEC commRF:itmen-t c-an be foud.i Appendix A (JF=SAR suipplement) Of the license Frnewal application. | ||
Of the license Frnewal application. | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 90 102. This item is considered a match to NUREG-1 801 even though the environments are different because the aging effect of cracking due to fatigue is independent of the environment. | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 90 102. This item is considered a match to NUREG-1 801 even though the environments are different because the aging effect of cracking due to fatigue is independent of the environment. | ||
103. These components are subject to cracking due to fatigue as identified in the generic entry in the first line of this table. | |||
104. The One-Time Inspection Program will verify effectiveness of the Water Chemistry Control - Primary and Secondary Program. | |||
105. The origqinal inconel .quide tube support pins (split pins) were replaced in both units with X-750 pins. The IP3 X-750 split pins, in service since 1987, were replaced in 2009 with stainless steel pins. The IP2 X-750 pins, installed in 1995, remain in service. Future pin replacements will be based on the pin design, industry experience, manufacturer recommendations and plant specific considerations. | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 90 Table 3.1.2-2-1P2 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table Notes Type Fntin Material Environment Requiring Management .1801 Vol. Item Type Function Management Programs 2 Item Item Reactor vessel | |||
Components that are not included in Tables 4-3, 4-6 or 4-9 are considered to be within the scope of the program, but require no specific inspections. | Components that are not included in Tables 4-3, 4-6 or 4-9 are considered to be within the scope of the program, but require no specific inspections. | ||
: 2. Preventive Actions The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions. However, primary water chemistry is maintained in accordance with EPRI guidelines by the Water Chemistry Control -Primary and Secondary Program, which minimizes the potential for stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC).Plant operations also influence aging of the vessel internals. | : 2. Preventive Actions The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions. However, primary water chemistry is maintained in accordance with EPRI guidelines by the Water Chemistry Control - Primary and Secondary Program, which minimizes the potential for stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC). | ||
The general assumptions about plant operations used in the development of the MRP-227 guidelines are applicable to the IPEC units. The units are base loaded and implemented low leakage core loading patterns within the first 30 years of operation. | Plant operations also influence aging of the vessel internals. The general assumptions about plant operations used in the development of the MRP-227 guidelines are applicable to the IPEC units. The units are base loaded and implemented low leakage core loading patterns within the first 30 years of operation. IPEC has implemented no design changes to reactor vessel internals beyond those identified in general industry guidance or recommended by Westinghouse. | ||
IPEC has implemented no design changes to reactor vessel internals beyond those identified in general industry guidance or recommended by Westinghouse. | : 3. Parameters Monitored or Inspected The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations and other aging management methods, as required. As described in MRP-227, the program contains elements that will monitor and inspect for the parameters that indicate the progress of each of these effects. The program will use NDE techniques to detect loss of material through wear, identify distortion of components, and locate cracks. | ||
: 3. Parameters Monitored or Inspected The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations and other aging management methods, as required. | Visual examinations (VT-3) will be used to detect wear. Visual examinations (VT-3) will also detect distortion or cracking through indications such as gaps or displacement along component joints and broken or damaged bolt locking systems. Direct measurements of spring height will be Used to detect distortion of the internals hold down spring. Visual examinations (EVT-1) will be used to detect crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting. | ||
As described in MRP-227, the program contains elements that will monitor and inspect for the parameters that indicate the progress of each of these effects. The program will use NDE techniques to detect loss of material through wear, identify distortion of components, and locate cracks.Visual examinations (VT-3) will be used to detect wear. Visual examinations (VT-3) will also detect distortion or cracking through indications such as gaps or displacement along component joints and broken or damaged bolt locking systems. Direct measurements of spring height will be Used to detect distortion of the internals hold down spring. Visual examinations (EVT-1) will be used to detect crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting.(MRP-227, Tables 4-3 and 4-6) | (MRP-227, Tables 4-3 and 4-6) | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 87 of 90 4. Detection of Aging Effects The RVI Program will detect cracking, loss of material reduction of fracture toughness, loss of preload and dimensional changes (distortion) of vessel internals components in accordance with MRP-227. The NDE systems (i.e., the combinations of equipment, procedure, and personnel) used to detect these aging effects will be qualified in accordance with MRP-228. The RVI Program will. conduct inspections of primary group components as follows (MRP-227, Table 4-3):* Periodic visual examinations (VT-3) will detect loss of material due to wear from control rod guide tube guide plates and thermal shield flexure plates.* Periodic visual examinations (VT-3) of the baffle former assembly plates and edge bolts will detect symptoms of distortion due to void swelling or cracking from IASCC.These symptoms include abnormal interactions with fuel assemblies, gaps or displacement along component joints, broken or damaged bolt locking systems, and failed or missing bolts.0 Direct measurements of spring height will detect distortion of the internals hold down spring due to a loss of stiffness. | |||
Measurements will be taken periodically, as needed to determine the life of the spring.* Periodic visual examinations (EVT-1) will detect crack-like surface flaws of the control rod guide tube assembly lower flange welds and the upper core barrel to flange weld.o Volumetric (UT) examinations will locate cracking of baffle former bolting. Baseline and subsequent measurements will be used to confirm the stability of the bolting pattern.Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227, Table 5-3.The relationships between primary group component inspection findings and additional inspections of expansion group components are as follows.* Indications from the EVT-1 inspections of the control rod guide tube assembly lower flange welds may result in EVT-1 inspections of the lower support column bodies and VT-3 inspections of bottom mounted instrumentation column bodies to detect cracking.* Indications from the EVT-1 inspection of the upper core barrel to flange weld may result in EVT-1 inspections of the remaining core barrel welds* Indications from the UT inspections of baffle former bolting may result in UT inspections of the lower support column bolts and the barrel former bolts for cracking. | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 87 of 90 | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 88 of 90 5. Monitoring and Trending The RVI Program uses the inspection guidelines for PWR internals in MRP-227.Inspections in accordance with these guidelines will provide timely detection of aging effects. In addition to the inspections of primary group components, expansion group components have been defined should the scope of examination and re-examination need to be expanded beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results.IPEC will share inspection results with the industry in accordance with the good practice recommendations of MRP-227. The IPEC-specific results will be incorporated into an overall industry report that will track industry progress and will aid in evaluation of potentially significant issues, identification of fleet trends, and determination of any needed revisions to MRP-227 guidelines. | : 4. Detection of Aging Effects The RVI Program will detect cracking, loss of material reduction of fracture toughness, loss of preload and dimensional changes (distortion) of vessel internals components in accordance with MRP-227. The NDE systems (i.e., the combinations of equipment, procedure, and personnel) used to detect these aging effects will be qualified in accordance with MRP-228. The RVI Program will. conduct inspections of primary group components as follows (MRP-227, Table 4-3): | ||
: 6. Acceptance Criteria The RVI Program acceptance criteria are from Section 5 of MRP-227. Table 5-3 of MRP-227 provides the acceptance criteria for inspections of the primary and expansion group components. | * Periodic visual examinations (VT-3) will detect loss of material due to wear from control rod guide tube guide plates and thermal shield flexure plates. | ||
The criteria for expanding the examinations from the primary group components to include the expansion group components are also delineated in MRP-227, Table 5-3. The examination acceptance criteria include: (i) specific, descriptive relevant conditions for the visual (VT-3) examinations; (ii) requirements for recording and dispositioning surface breaking indications that are detected and sized for length by the visual (EVT-1) examinations; and (iii) requirements for system-level assessment of bolted assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits.7. Corrective Action Conditions adverse to quality; such as failures, malfunctions, deviations, defective material and equipment, and nonconformances; are promptly identified and corrected. | * Periodic visual examinations (VT-3) of the baffle former assembly plates and edge bolts will detect symptoms of distortion due to void swelling or cracking from IASCC. | ||
In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. | These symptoms include abnormal interactions with fuel assemblies, gaps or displacement along component joints, broken or damaged bolt locking systems, and failed or missing bolts. | ||
In addition, the cause of the significant condition adverse to quality and the corrective action implemented is documented and reported to appropriate levels of management. | 0 Direct measurements of spring height will detect distortion of the internals hold down spring due to a loss of stiffness. Measurements will be taken periodically, as needed to determine the life of the spring. | ||
The Entergy (10 CFR Part 50, Appendix B) Quality Assurance Program, including relevant corrective action controls, applies to the RVI Program.Any detected condition that does not satisfy the examination acceptance criteria must be processed through the corrective action program. Example methods for analytical disposition of unacceptable conditions are discussed or referenced in Section 6 of MRP-227. These methods or other demonstrated and verified alternative methods may be used.The alternative of component repair and replacement of PWR internals is subject to the applicable requirements of the ASME Code Section XI. | * Periodic visual examinations (EVT-1) will detect crack-like surface flaws of the control rod guide tube assembly lower flange welds and the upper core barrel to flange weld. | ||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 89 of 90 8. Confirmation Process This attribute is discussed in Section B.0.3.9. Administrative Controls This attribute is discussed in Section B.0.3.10. Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, specifically with regard to cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators created a program to inspect the baffle-former bolting to determine whether similar aging degradation might be expected to occur in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections. | o Volumetric (UT) examinations will locate cracking of baffle former bolting. Baseline and subsequent measurements will be used to confirm the stability of the bolting pattern. | ||
In addition, the industry began laboratory testing projects to gather the materials data necessary to support future inspections and evaluations. | Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227, Table 5-3. | ||
Other confirmed or suspected material degradation concerns that the industry has identified for PWR components are.wear in thimble tubes, potential wear in control rod guide tube guide plates, and cracking in some high-strength bolting. The industry has addressed the last concern primarily through replacement of high-strength bolting with bolt material that is less susceptible to cracking and by improved control of pre-load.The RVI Program established in accordance with the MRP-227 guidelines is a new program.Accordingly, there is no direct programmatic history for IPEC. However, program inspections will use qualified techniques similar to those successfully used at IPEC and throughout the industry for ASME Section XI Code inspections. | The relationships between primary group component inspection findings and additional inspections of expansion group components are as follows. | ||
Internals inspections (VT-3)have been conducted at IPEC in accordance with ASME Section XI Code requirements, with no indications of component degradation. | * Indications from the EVT-1 inspections of the control rod guide tube assembly lower flange welds may result in EVT-1 inspections of the lower support column bodies and VT-3 inspections of bottom mounted instrumentation column bodies to detect cracking. | ||
IPEC has appropriately responded to industry operating experience for reactor vessel internals. | * Indications from the EVT-1 inspection of the upper core barrel to flange weld may result in EVT-1 inspections of the remaining core barrel welds | ||
For example, guide tube support pins (split pins) have been replaced in both units on the basis of industry experience. | * Indications from the UT inspections of baffle former bolting may result in UT inspections of the lower support column bolts and the barrel former bolts for cracking. | ||
As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the IPEC units. As a result, IPEC has monitored industry developments and recommendations regarding these components. | |||
Development of the MRP-227 guidelines is based upon industry operating experience, research data, and vendor evaluations. | NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 88 of 90 | ||
Reactor vessel internals aging degradation incidents in both U.S. and foreign plants were considered in the development of the MRP-227 guidelines. | : 5. Monitoring and Trending The RVI Program uses the inspection guidelines for PWR internals in MRP-227. | ||
As implemented, this program will account for applicable future operating experience during the period of extended operation. | Inspections in accordance with these guidelines will provide timely detection of aging effects. In addition to the inspections of primary group components, expansion group components have been defined should the scope of examination and re-examination need to be expanded beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results. | ||
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 90 of 90 Conclusion The RVI Program will be effective at managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls in accordance with MRP-227 and MRP-228 guidelines and current IPEC programs. | IPEC will share inspection results with the industry in accordance with the good practice recommendations of MRP-227. The IPEC-specific results will be incorporated into an overall industry report that will track industry progress and will aid in evaluation of potentially significant issues, identification of fleet trends, and determination of any needed revisions to MRP-227 guidelines. | ||
The RVI Program will provide reasonable assurance that the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.}} | : 6. Acceptance Criteria The RVI Program acceptance criteria are from Section 5 of MRP-227. Table 5-3 of MRP-227 provides the acceptance criteria for inspections of the primary and expansion group components. The criteria for expanding the examinations from the primary group components to include the expansion group components are also delineated in MRP-227, Table 5-3. The examination acceptance criteria include: (i) specific, descriptive relevant conditions for the visual (VT-3) examinations; (ii) requirements for recording and dispositioning surface breaking indications that are detected and sized for length by the visual (EVT-1) examinations; and (iii) requirements for system-level assessment of bolted assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits. | ||
: 7. Corrective Action Conditions adverse to quality; such as failures, malfunctions, deviations, defective material and equipment, and nonconformances; are promptly identified and corrected. In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. In addition, the cause of the significant condition adverse to quality and the corrective action implemented is documented and reported to appropriate levels of management. The Entergy (10 CFR Part 50, Appendix B) Quality Assurance Program, including relevant corrective action controls, applies to the RVI Program. | |||
Any detected condition that does not satisfy the examination acceptance criteria must be processed through the corrective action program. Example methods for analytical disposition of unacceptable conditions are discussed or referenced in Section 6 of MRP-227. These methods or other demonstrated and verified alternative methods may be used. | |||
The alternative of component repair and replacement of PWR internals is subject to the applicable requirements of the ASME Code Section XI. | |||
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 89 of 90 | |||
: 8. Confirmation Process This attribute is discussed in Section B.0.3. | |||
: 9. Administrative Controls This attribute is discussed in Section B.0.3. | |||
: 10. Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, specifically with regard to cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators created a program to inspect the baffle-former bolting to determine whether similar aging degradation might be expected to occur in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections. | |||
In addition, the industry began laboratory testing projects to gather the materials data necessary to support future inspections and evaluations. Other confirmed or suspected material degradation concerns that the industry has identified for PWR components are. | |||
wear in thimble tubes, potential wear in control rod guide tube guide plates, and cracking in some high-strength bolting. The industry has addressed the last concern primarily through replacement of high-strength bolting with bolt material that is less susceptible to cracking and by improved control of pre-load. | |||
The RVI Program established in accordance with the MRP-227 guidelines is a new program. | |||
Accordingly, there is no direct programmatic history for IPEC. However, program inspections will use qualified techniques similar to those successfully used at IPEC and throughout the industry for ASME Section XI Code inspections. Internals inspections (VT-3) have been conducted at IPEC in accordance with ASME Section XI Code requirements, with no indications of component degradation. IPEC has appropriately responded to industry operating experience for reactor vessel internals. For example, guide tube support pins (split pins) have been replaced in both units on the basis of industry experience. As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the IPEC units. As a result, IPEC has monitored industry developments and recommendations regarding these components. | |||
Development of the MRP-227 guidelines is based upon industry operating experience, research data, and vendor evaluations. Reactor vessel internals aging degradation incidents in both U.S. and foreign plants were considered in the development of the MRP-227 guidelines. As implemented, this program will account for applicable future operating experience during the period of extended operation. | |||
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 90 of 90 Conclusion The RVI Program will be effective at managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls in accordance with MRP-227 and MRP-228 guidelines and current IPEC programs. The RVI Program will provide reasonable assurance that the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.}} | |||
Latest revision as of 21:58, 11 March 2020
| ML102030120 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/15/2010 |
| From: | Bessette P Entergy Nuclear Operations, Morgan, Morgan, Lewis & Bockius, LLP |
| To: | Lathrop K, Lawrence Mcdade, Richard Wardwell Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| 50-247-LR, 50-286-LR, RAS E-374 | |
| Download: ML102030120 (99) | |
Text
Morgan, Lewis &Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Morgan Lewis COUNSELORS AT LAW Tel: 202.739.3000 Fax: 202.739.3001 www.morganlewis.com Kathryn M. Sutton DOCKETED Partner USNRC 202.739.5738 ksutton @morganiewis.com June 15, 2010 (4:45 p.m.)
PaulM. Bessette OFFICE OF SECRETARY Partner RULEMAKINGS AND 202.739.5796 ADJUDICATIONS STAFF pbessette @morganlewis.com July 15, 2010 Lawrence G. McDade, Chairman Dr. Richard E. Wardwell Dr. Kaye D. Lathrop Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Docket: Entergy Nuclear Operatgios,nc.(Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR and 50-286-LR RE: Notification of Entergy's Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3
Dear Administrative Judges:
Entergy Nuclear Operations, Inc. ("Entergy") is providing this notice to the Atomic Safety and Licensing Board ("Board") and the parties regarding Entergy's submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3 to the U.S. Nuclear Regulatory Commission ("NRC") on July 14, 2010. See NL-10-063, Letter from Fred Dacimo, Entergy, to NRC Document Control Desk, "Amendment 9 to License Renewal Application (LRA) - Reactor Vessel Internals Program" (July 14, 2010). A copy of NL- 10-063 is attached for your reference.
Counsel is providing this notification insofar as the Reactor Vessel Internals Program may be relevant and material to admitted contention NYS-25.
y,~LAT~-~ ~ b2&~
Morgan Lewis Lawrence G. McDade, Chairman COUNSELORS AT LAW Dr. Richard E. Wardwell Dr. Kaye D. Lathrop July 15, 2010 Page 2
'Resfuly submitted, Paul M. Bessette, Esq.
Counsel for Entergy Nuclear Operations, Inc.
CBM/als Attachment cc: Service List
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and
) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. ))
(Indian Point Nuclear Generating Units 2 and 3) )
.) July 15, 2010 CERTIFICATE OF SERVICE I hereby certify that copies of the letter entitled "Notification of Entergy's Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3," dated July 15, 2010, were served this 15th day of July, 2010 upon the persons listed below, by first class mail and e-mail as shown below.
Administrative Judge Administrative Judge Lawrence G. McDade, Chair Kaye D. Lathrop Atomic Safety and Licensing Board Panel Atomic -Safety and Licensing Board Panel Mail Stop: T-3 F23 190 Cedar Lane E.
U.S. Nuclear Regulatory Commission Ridgway, CO 81432
-WashingtonDC--2055 5-000O------- -(E-mail- kdt2@o)nrc-.gov ... -.. .......
(E-mail: larn @)nrc.gov)
Administrative Judge Office of the Secretary*
Richard E. Wardwell Attn: Rulemaking and Adjudications Staff Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Mail Stop: T-3 F23 Washington, D.C. 20555-0001 U.S. Nuclear Regulatory Commission (E-mail: hearingdocketanrc gov)
Washington, DC 20555-0001 (E-mail: rew(&nrc.gov)
Office of Commission Appellate Adjudication Josh Kirstein, Law Clerk U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Mail Stop: O-16G4 Mail Stop: T-3 F23 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission (E-mail: ocaamailanrc.gov) Washington, DC 20555-0001 (E-mail: Josh.Kirstein(anrc. gov)
Page 2 Sherwin E. Turk, Esq. Greg Spicer, Esq.
Beth N. Mizuno, Esq. Office of the Westchester County Attorney David E. Roth, Esq. 148 Martine Avenue, 6th Floor Brian G. Harris, Esq. White Plains, NY 10601 Andrea Z. Jones, Esq. (E-mail: gss I @(westchestergov.com)
Office of the General Counsel Mail Stop:, 0-15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (E-mail: set(dnrc.gov)
(E-mail: bniml@nrc.gov)
(E-mail: david.roth(onrc.gov)
(E-mail: brian.harris(nrc.gov)
(E-mail: andrea.iones(dnrc.gov)
Manna Jo Greene Thomas F. Wood, Esq.
Environmental Director Daniel Riesel, Esq.
Hudson River Sloop Clearwater, Inc. Ms. Jessica Steinberg, J.D.
724 Wolcott Avenue Sive, Paget & Riesel, P.C.
Beacon, NY 12508 460 Park Avenue (E-mail: mannaj o(&-clearwater.org) New York, NY 10022 (E-mail: driesel(asprlaw.com)
(E-mail: isteinberg(sprlaw.com)
Stephen C. Filler, Board Member John Louis Parker, Esq.
-Hudson-River-Sloop--ClearwaterInc. -------Regi6nal-Attomey- -
303 South Broadway, Suite 222 Office of General Counsel, Region 3 Tarrytown, NY 10591 NYS Dept. of Environmental Conservation (E-mail: sfilleranylawline.com) 21 S. Putt Comers Road New Paltz, New York 12561-1620 (E-mail: ilparker(ogw.dec.state.ny.us)
Ross Gould, Member Michael J. Delaney, V.P. - Energy Hudson River Sloop Clearwater, Inc. New York City Economic Development 10 Park Avenue, #5L Corp.
New York, NY 10016 110 William Street (E-mail: rgouldesq(agrmail.com) New York, NY 10038 (E-mail: mdelaney(cnycedc.com)
Page 3 Phillip Musegaas, Esq.: Daniel E. O'Neill, Mayor Deborah Brancato, Esq. James Siermarco, M.S.
Riverkeeper, Inc. Liaison to Indian Point 828 South Broadway Village of Buchanan Tarrytown, NY 10591 Municipal Building (E-mail: phillip(iriverkeeper.org) 236 Tate Avenue (E-mail: dbrancato(riverkeeper.org) Buchanan, NY 10511-1298 (E-mail: vob(bestweb.net)
Robert D. Snook, Esq. Mylan L. Denerstein, .Esq.
Assistant Attorney General Executive Deputy Attorney General, Office of the Attorney General Social Justice State of Connecticut Office of the Attorney General 55 Elm Street -of the State of New-York-P.O. Box 120 120 Broadway, 25th Floor Hartford, CT 06141-0120 New York, New York 10271 (E-mail: Robert. Snook(dpo.state.ct.us) (E-mail: Mvlan.Denerstein(ioag. state.ny. us)
Andrew M. Cuomo, Esq. . Janice A. Dean Attorney General of the State of New York Office of the Attorney General John J. Sipos, Esq. of the State of New York Charlie Donaldson Esq. Assistant Attorney General Assistants Attorney General 120 Broadway, 26th Floor The Capitol New York, New York 10271 (E-Albany,:N¥-t-2-2-24-03ta..... .) -- (E-mail --Janice;Dean(aoag~stateny.,us) .
(E-mail: john.sipos(ýoag.state.ny.us)
Joan Leary Matthews, Esq.
Senior Attorney for Special Projects Office of the General Counsel New York State Department of Environmental:Conservation 625 Broadway, 14th Floor Albany, NY 12207 (E-mail: ilmattheagw. dec. state.ny.us)
Page 4 Original and 2 copies provided to the Office of the Secretary.
Paul M. Bessette, Esq.
Counsel for Entergy Nuclear Operations, Inc.
DB 1/65220145.1
Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB Entergy P.O. Box 249 Buchanan, NY 105111-0249 Tel (914) 788-2055 Fred Dacimno Vice President License Renewal NL-10-063 July 14, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Amendment 9 to License Renewal Application (LRA) -
Reactor Vessel Internals Program Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64
REFERENCES:
- 1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)
- 2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings (NL-07-040)
- 3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References (NL-07-041)
- 4. Entergy Letter dated October 11, 2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07-124)
- 5. Entergy Letter November 14, 2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA)
Environmental Report References" (NL-07-133)
Dear Sir or Madam:
In the referenced letters, Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Energy Center operating license. This letter contains Amendment 9 to the License Renewal Application (LRA) regarding the Reactor Vessel Internals Program.
If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.
NL- 10-063 Docket Nos. 50-247 & 50-286 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on Sincerely, FRD/dmt
Attachment:
- 1. Amendment 9 to License Renewal Application -
Reactor Vessel Internals Program cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. John Boska, NRR Senior Project Manager Ms. Kimberly Green, Project Manager NRC Resident Inspector's Office Mr. Paul Eddy, New York State Department of Public Service Mr. Francis J. Murray, President and CEO, NYSERDA
ATTACHMENT 1 TO NL-10-063 Amendment 9 to License Renewal Application -
Reactor Vessel Internals Program ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 LICENSE NOS. DPR-26 AND DPR-64
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 1 of 90 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)
AMMENDMENT 9 The LRA is revised as described below. (underline - added, strikethrough - deleted) 2.3.1.2 Reactor Vessel Internals The reactor vessel internals for each unit are described in the reactor coolant system description (Unit 2, Reactor Vessel Internals; Unit 3, Reactor Vessel Internals).
For both units, the lower core support structure, the upper core support structure, and the incore instrumentation support structure are the three major parts of the reactor internals.
Lower Core Support Structure The major member of the reactor vessel internals is the lower core support structure consisting of the following components included in this evaluation.
core baffle/former assembly: bolts core baffle/former assembly: plates core barrel assembly: bolts, screws core barrel assembly: axial flexure plates (thermal shield flexures), flange, ring, shell, thermal shield, lower core barrel flange weld, upper core barrel flange weld core barrel assembly: outlet nozzles lower internals assembly: clevis insert bolt lower internals assembly: clevis insert lower internals assembly: intermediate diffuser plate lower internals assembly: fuel alignment pin lower internals assembly: lower core plate lower internals assembly: lower core support plate column sleeves lower internals assembly: lower core support column bolt lower internals assembly, lower core support column castings: column cap, lower core support lower internals assembly: radial key lower internals assembly: secondary core support (energy absorbing device) specimen guides (not subject to aging management review) specimen plugs (installed in IP2 only; not subject to aging management review)
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 90 The lower core support structure is supported at its upper flange from a ledge in the reactor vessel. Within the core barrel are a core baffle and a lower core plate, both of which are attached to the core barrel wall. The lower core support structure provides passageways for the coolant flow. The lower core plate at the bottom of the core below the baffle plates provides support and orientation for the fuel assemblies. Fuel alignment pins (two for each assembly) are also inserted into this plate. Columns are placed between the lower core plate and core support casting in order to provide stiffness and to transmit the core load to the core support casting. Adequate coolant distribution is obtained through the use of the lower core plate and a diffuser plate.
Upper Core Support Structure The "top hat with deep beam features" upper core support structure consists of the following components included in this evaluation.
upper internals assembly, rod control cluster assembly (RCCA) guide tube assembly: bolts upper internals assembly, RCCA guide tube assembly: guide tube (including lower flange weld), guide plates upper internals assembly, RCCA guide tube assembly: support pin upper internals assembly: core plate alignment pin upper internals assembly: head/vessel alignment pin upper internals assembly: hold-down spring upper internals assembly: support column upper internals assembly, mixing devices: support column orifice base, support column mixer upper internals assembly: upper core plate, fuel alignment pin upper internals assembly: support assembly (including ring), upper support plate upper internals assembly: upper support column bolt The support columns establish the spacing between the upper support assembly and the upper core plate and are fastened at top and bottom to these plates and beams.
The RCCA guide tube assemblies shield and guide the control rod drive shafts and control rods.
They are fastened to the upper support and are guided by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the control rod shroud tube which is attached to the upper support plate and guide tube.
In-Core Instrumentation Support Structure The in-core instrumentation support structures consist of the following components included in this evaluation.
thermocouple conduit flux thimble guide tube bottom mounted instrumentation column
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 90 An upper system (thermocouple conduit) is used to convey and support thermocouples penetrating the vessel through the head, and a lower system (flux thimble guide tube) is used to convey and support flux thimbles penetrating the vessel through the bottom.
The upper system utilizes the reactor vessel head penetrations. Instrumentation port columns are slip-connected to in-line columns that are in turn fastened to the upper support plate. These port columns protrude through the head penetrations. The thermocouples are carried through these port columns and the upper support plate at positions above their readout locations. The thermocouple conduits are supported from the columns of the upper core support system.
Table 2.3.1-2-1P2 and Table 2.3.1-2-1P3 list the mechanical components subject to aging management review and component intended functions for the reactor vessel internals.
Table 3.1.2-2-1P2 and Table 3.1.2-2-1P3 provide the results of the aging management review for the reactor vessel internals.
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 90 Table 2.3.1-4-1P2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Intended Function Lower Core Support Structure, Core baffle/former assembly Structural support
.bolts Core baffle/former assembly Structural support
-plates Flow distribution Shielding Core barrel assembly Structural support
-bolts and screws Core barrel assembly Structural support
-axial flexure plates dltributi Floherm
- flaRge Sil~
-thermal shield Core barrel assembly Structural support
- axial flexure plates (thermal shield flexures)
Core barrel assembly Structural support Core barrel assembly Structural support
" ringl Flow distribution
" shell Shielding
" thermal shield Core barrel assembly Structural support
" lower core barrel flange weld
" upper core barrel flange weld
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 90 Core barrel assembly Flow distribution
-outlet nozzles Lower internals assembly Structural support
- clevis insert bolt
- clevis insert
-fuel alignment pin
-lower core support plate column sleeves
-lower core support plate column bolt
-radial key Lower internals assembly Flow distribution
-intermediate diffuser plate Lower internals assembly Structural support
-lower core plate Flow distribution
-lower core support castings
-column cap
-lower core support
-secondary core support Upper Core Support Structure-UpperInternals Assembly RCCA guide tube assembly Structural support
-bolt RCCA guide tube assembly Structural support
-guide tube (including lower flange welds)
RCCA guide tube assembly Structural support
-guide plates RCCA guide tube assembly Structural support
-support pin
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 90 Core plate alignment pin Structural support Head / vessel alignment pin Structural support Hold-down spring Structural support Mixing devices Structural support
-support column orifice base Flow distribution
-support column mixer Support column Structural support Upper core plate, fuel alignment Structural support pin Flow distribution Upper support plate, support Structural support assembly (including ring)
Upper support column bolt Structural support Inc~orInstrumentationSuppflt ~Structure Bottom mounted instrumentation Structural support column Flux thimble guide tube Structural support Thermocouple conduit Structural support
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 7 of 90 3.1.2.1.2 Reactor Vessel Internals Materials Reactor vessel internals components are constructed of the following materials.
- cast austenitic stainless steel
- nickel alloy
- stainless steel Environment Reactor vessel internals components are exposed to the following environments.
- neutron fluence
- treated borated water
- treated borated water > 140°F 0 treated borated water > 4820F Aging Effects Requiring Management The following aging effects associated with the reactor vessel internals require management.
0 change in dimensions
- cracking
- cracking - fatigue
- loss of material
- loss of material - wear
- loss of preload
- reduction of fracture toughness Aging Management Programs The following aging management programs manage the aging effects for reactor vessel internals components.
0 Inservice Inspection 0 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) 0 Reactor Vessel Internals Water Chemistry Control - Primary and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 90 3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Loss of fracture toughness due to neutron irradiation embrittlement and change in dimensions (void swelling) e 4d eccur in stainless steel and nickel alloy reactor vessel internals components exposed to reactor coolant and neutron flux will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will imnlp~m~nt thp. FPRI PrP..I.iri7Pr1 W~tp~r Rp irtnr Intp~rn~lk In_*n*_nfinn *nd FIiinhtinfir Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aginq effects for reactor vessel internals. To manage loss ofra;ctr,+e toughnoss in ve...l inte-rnals; . components,
-PEG ,il11l (1 par--ici-te in he industry pr.grams for in.'Stigating and managing aing cie n rcacto ffi*ternals; (2) cvaluato and implement* th* esults of the dust.
.. pregFams as applicable to the.oacto. internals; and (3) upon cmR plet!on these programs, but not less than 24 moneths before entering the period of ex~te~nded operation, submit an inspection plan for ýeac~ter inernals to the NRC for review and ap~proval.-Th~is rsenoUm.tMont isicue nthe UFSAR Supplement, Appendix A, Sections A.2.!.41 ard A.3.1.41.
3.1.2.2.9 Loss of Preload due to Stress Relaxation Loss of preload due to thermal stress relaxation (creep) would only be a concern in very high temperature applications (> 700°F) as stated in the ASME Code,Section II, Part D, Table 4. No IPEC internals components operate at > 700'F. Therefore, loss of preload due to thermal stress relaxation (creep) is not an applicable aging effect for the reactor vessel internals components. However, irradiation-enhanced creep (irradiation creep) or irradiation enhanced stress relaxation (ISR) is an athermal process that depends on the neutron fluence and stress: and, on void swelling if present. Neve.4heless Therefore, loss of preload of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. to the extent that industry developed reactor vessel internals aging _managemenqt programs address thes a- ing effeGts. The I'PEC commitmert t these RVI programs iS included in IFSAR Su"pplement, Appendix A, Sections A.2. 1.1 and A.3. 1. 1.
3.1.2.2.15 Changes in Dimensions due to Void Swelling Changes in dimensions due to void swelling ee'-ld .e... in stainless steel and nickel alloy reactor internal components exposed to reactor coolant will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. T---maiage hanges in dim.en .... Of vessel inters*+ ' components, 'PEG will () participate the .,dust,' Ip-rogram for investigating and *an ma _an . g effe*t* en reacteo inRternals; (2) evaluate and implement the results Of the industry programs as
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 90 appI"cable to the reactor intornals; and (3) upon completion Of these programns, but not lcss thaR 241 mofths before enterFig the period of exterded operation;, subMit aR plan for reactor iRn8trals to the INIR for review and approIval. ThiS SnspectioR cOMMItmont isicue nthe UJFSAR Supplement, Appendix A, SectionRs A.2.1.41 and A.3.!.4!4 3.1.2.2.17 Cracking due to Stress Corrosion Cracking, Primary Water Stress Corrosion Cracking, and Irradiation-Assisted Stress Corrosion Cracking Cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) Gould GGeu in PWR stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. To manage crackig in Vessel internals com.ponents, IPEC maintains the Water C-hemistr,' Control Primary and SeconRdary ProgramR and Will (1) pa~ticipate in the industry programns for investigating and m:anagfing aging effects-on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion o~f those programs, but -otless than 2;* Fmonths before entering the perFod of cxtcnded ohperation, SubMit an inspect*i plan for reacRIIto Ine 1Ir a to the NRC fo review and approval: The
'PEC commFitmen~t to these RVI programns is included in UFSAR Supplement, Appen~dix A, Sections A.2.1 .41 and A.3.1. .1.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 10 of 90 Table 3.1.1 Summary of Aging Management Programs for the Reactor Coolant System Evaluated in Chapter IV of NUREG-1 801 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee Consistont with NUREG 1801. Loss of nickel alloy reactor fracture commitment to (1) commitment to -fracture toughness of stainless steel vessel internals toughness due participate in be confirmed and nickel alloy reactor vessel components exposed to to neutron industry RVI aging internals components will be managed reactor coolant and irradiation programs (2) by the Reactor Vessel Internals neutron flux embrittlement, implement ProQram. aging manago.m..t -
void swelling applicable results prorams. The commitmeRnt to these (3) submit for NRC RVI progra.m is,includ ,inFSAR approval > 24 Supplement, Appendix A, Sections months before the A.2.1.41 and A.3. .14.
extended period an See Section 3.1.2.2.6.
RVI inspection plan based on industry recommendation.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further temComponent Ag ism Management Evaluation Discussion Programs Recommended 3.1.1-27 Stainless steel and Loss of FSAR supplement No, but licensee Loss of preloead duo to stross nickel alloy reactor preload due to commitment to (1) commitment to relaxatio* n (Goop) is a concern for vessel internals screws, stress participate in be confirmed application. at t.mpe.atures higher bolts, tie rods, and hold- relaxation industry RVI aging than thoso of 'PEC reactor vessel and down springs programs (2) intornals.OMPen.nts. Therfore, loo implement of prFolad duo to tr*oss rolaxation applicable results 9..p) is not an applicable aging (3) submit for NRC effcct for tho roactOr 'cssel internals approval > 24 compononts. Neverthoeoss, loss of months before the preload of stainless steel and nickel extended period an alloy reactor vessel internals RVI inspection plan components will be managed by the based on industry Reactor Vessel Internals Proqram.
recommendation. c.nsisont with , du*, ty devel oped rcactE)r vessel intcrnals aging anaRgqe*meRt prorams The comm tment to these RVI programs is rinluded iR UISAR SuppmonlRt, Apperndix A, RSortions A2. 1 41a;*d A. 341.
See Section 3.1.2.2.9.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-30 Stainless steel reactor Cracking due Water Chemistry No, but licensee Coneictont'.wth NUREG 1801.
vessel internals to stress and FSAR commitment Cracking _of stainless steel reactor components (e.g., corrosion supplement needs to be vessel internals components will be Upper internals cracking, commitment to (1) confirmed managed by the Water Chemistry assembly, RCCA guide irradiation- participate in Control - Primary and Secondary tube assemblies, assisted stress industry RVI aging Program and either the Reactor Vessel Baffle/former assembly, corrosion programs (2) Internals Proqram or the Inservice Lower internal cracking' implement Inspection Progqram. by-etheF RVt assembly, shroud applicable results nli,
-- --- ,rnm ni r-nr-~ Tka assemblies, Plenum (3) submit for NRC commitment to these other RVI cover and plenum approval > 24 programs is iAncrAlud-ed ipn. U lFS.A.Rn cylinder, Upper grid months before the Supplement, Appendix A, SectionS assembly, Control rod extended period an A. 2.1.4 aRd A."3.141.
guide tube (CRGT) RVI inspection plan See Section 3.1.2.2.12.
assembly, Core support based on industry shield assembly, Core recommendation.
barrel assembly, Lower grid assembly, Flow distributor assembly, Thermal shield, Instrumentation support structures)
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-33 Stainless steel and Changes in FSAR supplement No, but licensee Consistent With NUREG 180!.
nickel alloy reactor dimensions commitment to (1) commitment to Changes in dimensions of stainless vessel internals due to void participate in be confirmed steel and nickel alloy reactor vessel components swelling industry RVI aging internals components will be managed programs (2) by the Reactor Vessel Internals implement Proqram. RVI aging managmeont applicable results pogram.s. The co. mmit*m*t to these (3) submit for NRC RVI programs is included inUSAR approval > 24 Supplen*.t, App.ndix A, Soctions months before the A.2.1 .41 and A.3.1 .41.
extended period an See Section 3.1.2.2.15.
RVI inspection plan based on industry recommendation.
NL-1.0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-37 Stainless steel and Cracking due Water Chemistry No, but licensee Consistont with NUREG 1801.
nickel alloy reactor to stress and FSAR commitment Cracking _of stainless steel and nickel vessel internals corrosion supplement needs to be alloy reactor vessel internals components (e.g., cracking, commitment to (1) confirmed components will be managed by the Upper internals primary water participate in Water Chemistry Control - Primary assembly, RCCA guide stress industry RVI aging and Secondary Program and either tube assemblies, Lower corrosion programs (2) the Reactor Vessel Internals Program internal assembly, CEA cracking, implement or the Inservice Inspection Program.
shroud assemblies, irradiation- applicable results by othor RV! aging manag.m.nt Core shroud assembly, assisted stress (3) submit for NRC programs. The commitment to these Core support shield corrosion approval > 24 other RVI program. is inc.udod in assembly, Core barrel cracking months before the UF SAR SuppleMeRt, Appendix A, assembly, Lower grid extended period an Sections A.2.1.41 aRnd A.3.1.41.
assembly, Flow RVI inspection plan See Section 3.1.2.2.17.
distributor assembly) based on industry recommendation.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 I
Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item- Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-63 Steel reactor vessel Loss of Inservice Inspection No The Inservice Inspection Program and flange, stainless steel material due to (IWB, IWC, and the Reactor Vessel Internals Program and nickel alloy reactor wear IWD) manages loss of-material due to wear vessel internals of the steel reactor vessel flange and exposed to reactor stainless steel and nickel alloy reactor coolant (e.g., upper and vessel internals components.
lower internals assembly, CEA shroud assembly, core support barrel, upper grid assembly, core support shield assembly, lower grid assembly)
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 16 of 90 NO)ITEs FOR TABLES 3.1.2 1 I-2 THROU U3.1.24I 1P3 H
Generic Notes A. Consistent with NUREG-1 801 item for component, material, environment, aging effect and aging management program.
AMP is consistent with NUREG-1801 AMP.
B. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program.
AMP has exceptions to NUREG-1801 AMP.
C. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP is consistent with NUREG-1 801 AMP.
D. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP has exceptions to NUREG-1801 AMP.
E. Consistent with NUREG-1801 material, environment, and aging effect but a different aging management program is credited.
F. Material not in NUREG-1 801 for this component.
G. Environment not in NUREG-1801 for this component and material.
H. Aging effect not in NUREG-1 801 for this component, material and environment combination.
I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable.
J. Neither the component nor the material and environment combination is evaluated in NUREG-1 801.
Plant-Specific Notes 101. This component, material, environment and aging effect combination is considered in the Reactor Vessel Internals Program.
As documented in MRP-227, the basis for the RVI Program, this combination warrants no additional aging management.
NUREG 1801, Sectio-n XI.51 6 ctatj: "No fZrthr f ging manargemen reieW is ncessa,',if the applicant po*r dGos!a' co)mmitm~enti the ESAR supplepmont to (1) pa~ticipate in the indIASt~' programs foriG stgtn and manaaging aging offoctS on reactor inteFRnals; (2) evaluate and implcment the results of the industr,' programs as applicable to the reactor internals,ý and (3)uIpon co;plotion of these program~s, bu6t not less than 24 mon)ths bePfo-ro- enrtering the period-of extenAdod operation, submit an npcinplnfreco intorn-als to the NRC f4or reView and approval." IPEC commRF:itmen-t c-an be foud.i Appendix A (JF=SAR suipplement) Of the license Frnewal application.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 90 102. This item is considered a match to NUREG-1 801 even though the environments are different because the aging effect of cracking due to fatigue is independent of the environment.
103. These components are subject to cracking due to fatigue as identified in the generic entry in the first line of this table.
104. The One-Time Inspection Program will verify effectiveness of the Water Chemistry Control - Primary and Secondary Program.
105. The origqinal inconel .quide tube support pins (split pins) were replaced in both units with X-750 pins. The IP3 X-750 split pins, in service since 1987, were replaced in 2009 with stainless steel pins. The IP2 X-750 pins, installed in 1995, remain in service. Future pin replacements will be based on the pin design, industry experience, manufacturer recommendations and plant specific considerations.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 90 Table 3.1.2-2-1P2 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table Notes Type Fntin Material Environment Requiring Management .1801 Vol. Item Type Function Management Programs 2 Item Item Reactor vessel Structural Stainless Treated borated Cracking - TLAA - metal IV.B2-31 3.1.1-5 A internals support steel, water fatigue fatigue (R-53) components CASS, nickel alloy Lower Core Support Structure Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1- E A, baffle/former support steel water > 140'F dimensions Internals RV4 (R-126) 33 101 assembly bolts Cracking Water Chemistry IV.B2-10 3.1.1- EA-Control - Primary (R-125) 30 4-04 and Secondary Reactor Vessel Internals -V4 GeMG4RteRt Loss of Water Chemistry IV.B2-32 .3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Typet Fnctone Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item tem Loss of Reactor Vessel IV.B2-5 3.1.1- E A, preload Internals RV4 (R-129) 27 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- E A, water> 140'F fracture Internals R-V4 (R-128) 22 101 Neutron fluence toughness eMitne~t Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1- E A-baffle/former support steel water > 140°F dimensions Internals FM4 (R-124) 33 4--
assembly Flow GOMM"e t plates distribution Shielding Cracking Water Chemistry IV.B2-2 3.1.1- E A-Control - Primary (R-123) 30 4- and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-3 3.1.1- E A, water > 140 0 F fracture Internals RV4 (R-127) 22 101 Neutron fluence toughness eemitment
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Function Material Environment Requiring Management 1801 Vol. Notes Type Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1- E G, assembly support steel water > 140'F dimensions Internals R-V4 (R-126) 33- 101 bolts-and er,A-4ix screws Cracking Water Chemistry IV.B2-10 3.1.1- EA Control - Primary (R-125) 30 4 and Secondary Reactor Vessel Internals RNI GGeR4R4tR Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-5 3.1.1- E A, preload Internals RV4 (R-129) 27 101 commit Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- E A, water > 140°F fracture Internals PMt (R-128) 22 101 Neutron fluence toughness eeAAmriteMR
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.E2-7 3.1.1- E A, assembly support steel water > 140°F dimensions Internals RV4 (R-121) 33 101 axial flexure Fsew epmitmept plates (thermal dstrlbution d Cracking Water Chemistry IV.B2-8 3.1.1- E A, shield Control - Primary (R-120) 30 04 flexures) and Secondary Reactor Vessel Internals RFM GORmitmeRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-26 3.1.1- E material - Internals (R-142) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E A, water > 140'F fracture Internals R-4 (R-122) 22 101 Neutron fluence toughness e)i4mmeRt
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 22 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- EA, assembly support steel water > 140'F dimensions Internals R-V4 (R-121) 33 101
-flange Flew A4M.m-A dw4,4butn Cracking Water Chemistry IV.B2-8 _
3.1.1- E A, Control - Primary (R-120) 30 "
and Secondary Reactor Vessel Internals R-V4 GMM4tMeRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-34 3.1.1- E material - Inspection (R-115) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E A, water > 140'F fracture Internals R-V4 (R-122) 22 101 Neutron fluence toughness eemIui4me*
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 23 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals -
Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Item Notes Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly support steel water > 140'F dimensions Internals RV4 (R-121) 33 101
" ring Flow "shell -distribution "thermal shield Shielding Cracking Water Chemistry IV.B2-8 3.1.1- EA Control - Primary (R-120) 30 101 and Secondary Reactor Vessel Internals R-V cGmmitR~nt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E A, water > 140'F fracture Internals R-V4 (R-122) 22 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 24 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Type Function Notes Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E, assembly support steel water > 140'F dimensions Internals (R-121) 33 101
" lower core barrel flange Cracking Water Chemistry IV.B2-8 3.1.1- E weld Control - Primary (R-120) 30
" upper core and Secondary barrel flange Reactor Vessel weld Internals Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E water > 140°F fracture Internals (R-122) 22 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 25 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG- Notes Component Intended Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly distribution steel water > 140'F dimensions Internals R-4 (R-121) 33 101 outlet nozzles t M.
eRM t
Cracking Water Chemistry IV.B2-8 3.1.1- E A-Control - Primary (R-120) 30 4 and Secondary Reactor Vessel Internals R-V4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page'26 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Item Notes Management Programs 2 Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- E A, assembly support water dimensions Internals RV4 (R-134) 33 101 clevis insert Ge9eFR bolt Cracking Water Chemistry IV.B2-16 3.1.1- E A, Control - Primary (R-133) 37 101 and Secondary Reactor Vessel Internals RV-I Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-14 3.1.1- E A, preload Internals RV4 (R-137) 27 101 GGMRiteRt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- E A, water fracture Internals R-V4 (R-135) 22 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 27 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E A, assembly , support water dimensions Internals RV4 (R-131) 33 101 clevis insert Cracking Water Chemistry IV.B2-20 3.1.1- E A, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-26 3.1.1- E material - Inspection (R-142) 63 wear r
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 28 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower internals Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E G, assembly distribution steel water > 140°F dimensions Internals R-V (R-131) 33 101
- intermediate GGR4R4~R4 diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1- E G, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals R-V4 roe mm i r4 PieAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
.Attachment 1 Docket Nos. 50-247 & 50-286 Page 29 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower internals Structural Stainless Treated borated Change in Rea tor Vessel IV.B2-15 3.1.1- E A, assembly support steel water > 140OF dimensions Internals P-V4 (R-134) 33 101 fuel alignment pin Cracking Water Chemistry IV.B2-16 3.1.1- E A$
Control - Primary (R-133) 37. 101 and Secondary Reactor Vessel Internals RV4 GGR;MitR~eRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- E A, water > 140°F fracture Internals R-V4 (R-135) 22 101 Neutron fluence toughness eeG itMe8t
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 30 of 90 Table 3.1.2-2-lP2: Reactor Vessel Internals CAging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower internals Structural Stainless. Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E A, assembly support steel water > 140°F dimensions Internals RV4 (R-131) 33 101 lower core Flow GGR4WRRRt plate distribution Cracking Water Chemistry IV.B2-20 3.1.1- E A-,
Control - Primary (R-130) 37 1 and Secondary RVI commitmont Inservice Inspection Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-26 3.1.1- E material - Inspection (R-142) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1- EA, water > 140°F fracture Internals R-V4 (R-132) 22 101 Neutron fluence toughness eeMM.tMeRt
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 31 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table N Component Intended Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower internals Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- E A, assembly support water > 4820 F dimensions Internals R-V4 (R-139) 33 " 101 low er core Flow GGeM ,iR, R ,
support distribution castings Cracking Water Chemistry IV.B2-24 3.1.1- E A,
- column cap Control - Primary (R-138) 30 40-0
- lower core and Secondary support Reactor Vessel column Internals -V4 bodies Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Thermal Aging and IV.B2-21 3.1.1- A water > 482'F fracture Neutron Irradiation (R-140) 80 Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)
NL-10-063 Attachment I Docket Nos. 50-247 & 50-286 Page 32 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect. Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- E A, assembly support steel water > 140'F dimensions Internals R-V4 (R-134) 33 101 lower core t t support plate column bolt Cracking' Water Chemistry IV.1B2-16 3.1.1- E A=,
Control - Primary (R-133) 37 1 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-25 3.1.1- E A, preload Internals -V4 (R-136) 27 101 GG.m44ImeRt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- E A, water > 140'F fracture Internals RV4 (R-135) 22 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 33 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Function Material Environment Requiring Management 1801 Vol. Item Notes Type Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- E A, assembly support steel water > 140'F dimensions Internals RV4 (R-139) 33 101 lower core support column plate Cracking Water Chemistry IV.B2-24 3.1.1- E A, sleeves Control- Primary (R-138) 30 101 and Secondary Reactor Vessel Internals R-V4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-22 3.1.1- E A, water > 140'F fracture Internals R-V4 (R-141) 22 101 Neutron fluence toughness EeMMit1et
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 34 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Type Function Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E A, assembly support steel water> 140°F dimensions Internals RWI (R-131)- 33 101 radial key Cracking Water Chemistry IV.B2-20 3.1.1- E A, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals RV4 Ge m*
Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-26 3.1.1- E material - Inspection (R-142) 63 wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 35 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E G, assembly support steel water > 140'F dimensions Internals R-V4 (R-131) 33 101 secondary- Flow -piireRt core support distribution____ Cracking Water Chemistry IV.B2-20 3.1.1- E G, Control- Primary (R-130) 37 101 and Secondary Reactor Vessel Internals R-V4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 36 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals UpperCre Support Structbivre- '-Upper Intemals Assembly RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-27 3.1.1- EA, tube assembly support steel water > 140'F dimensions Internals R-V4 (R-1 19) 33 101
- bolt Gr44MR P t Cracking Water Chemistry IV.B2-28 3.1.1- E A, Control- Primary (R-118) 37 101 and Secondary Reactor Vessel Internals R-V4 GeeOM Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- EG, preload Internals R-V4 (R-1 14) 27 101 GEMMtMeGRt
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 37 of 90-Table 3.1.2-2-IP2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1- E A, tube assembly support steel water > 1407F dimensions Internals R-V4 (R-117) 33 101 guide tube (including lower flange Cracking Water Chemistry IV.B2-30 3.1.1- E A, welds) Control - Primary (R-1 16) 30 4-04 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 38 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1- E tube assembly support steel water > 140'F dimensions Internals (R-1 17) 33 guide plates Cracking Water Chemistry IV.B2-30 3.1.1- E Control - Primary (R-116) 30 and Secondary Reactor Vessel Internals Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-34 3.1.1- E material - Internals (R-115) 63 wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 39 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item RCCA guide Structural Niekel-al, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1- E A, tube assembly support Stainless water dimensions Internals RV4 (R-119) 33 101.
support pin steel eOMMnffMe.t Cracking Water Chemistry IV.B2-28 3.1.1- E 105 Control - Primary (R-1 18) 37 A,-101 and Secondary Reactor Vessel Internals R-V Loss of Water Chemistry IV.B2-32 3.1.1- A material .. Control - Primary (RP-24) 83 and Secondary
NL-1 0-063 Attachment 1-Docket Nos. 50-247 & 50-286 Page 40 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Core plate Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- E A, alignment pin support steel water > 140°F dimensions Internals RV4 (R-1 13) 33 101 Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-34 3.1.1- E material - Inspection (R1 15) 63 wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 41 of 90 Table 3.1.2-2-1P2: Reactor.Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Head / vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- E G, alignment pin support steel water > 140'F dimensions Internals -V4 (R-107) 33 101 Cracking Water Chemistry IV.1B2-42 3.1.1- E G, Control - Primary (R-106) 30 101 and Secondary Reactor Vessel Internals RV4 GGemm44me Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-34 3.1.1- E material - Inspection (R1 15) 63 wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 42 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Hold-down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- E A, spring support steel water > 140'F dimensions Internals R-V4 (R-107) 33 101 GMM4,tRt Cracking Water Chemistry IV.B2-42 3.1.1- E A, Control - Primary (R-106) 30 101 and Secondary Reactor Vessel Internals R-V4 GQRm4tROR Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- E A, preload Internals RV4 (R-1 14) 27 404-GOM eR
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 43 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Mixing devices Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1- E G,
" support support water > 4827F dimensions Internals RV4 (R-1 10) 33 101 column orifice Flow base bsupport distribution Cracking Water Chemistry IV.B2-36 3.1.1- E G, column mixer Control - Primary (R-109) 30 101 and Secondary Reactor Vessel Internals R-V4 GGRentmel Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Thermal Aging and IV.B2-37 3.1.1- A water > 482°F fracture Neutron Irradiation (R-1 11) 80 Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 44 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table N Component Intended Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-35 3.1.1- E A, column support steel water > 140'F dimensions Internals R-V (R-1 10) 33 101 cQm~tRmRt Cracking Water Chemistry IV.B2-36 3.1.1- E A, Control - Primary (R-109) 30 101 and Secondary Reactor Vessel Internals R-V4 GGR44 1 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Upper core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- E A, plate, fuel support steel water > 140'F dimensions Internals PV4 (R-1 17) 33 101 alignment pin Flow distribution Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-1 12) 37 101 and Secondary Reactor Vessel Internals R-V4
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 45 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- E A, plate, support support steel water > 140°F dimensions Internals R-V4 (R-107) 33 101 assembly *
(including ring) Cracking Water Chemistry IV.B2-42 3.1.1- E A-,
Control - Primary (R-106) 30 1-04 and Secondary Inservice Inspection RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 46 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Notes Component Intended Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- E A, column bolt support steel water > 140°F dimensions Internals RV4 (R-1 13 33 101 Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals R-Vt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- E A, preload Internals R-V4 (R-1 14) 27 101 94,4,fme
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 47 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table N Component Intended Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Incore Instrumentation. Suppot Structure Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- E A, mounted support steel water > 140°F dimensions Internals RV4 (R-144) 33 101 instrumentation column Cracking Water Chemistry IV.B2-12 3.1.1- E A, Control - Primary (R-143) 30 and Secondary Reactor Vessel Internals RV4 emmnitmReRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 48 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- E A, guide tube support steel water > 140°F dimensions Internals RV4 (R-144) 33 101 GGmm4tmeR4 Cracking Water Chemistry IV.B2-12 3.1.1- E A, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals RV4 GeGR4RitiRRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Thermocouple Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- E C, conduit support steel water > 140°F dimensions Internals -V4 (R-144) 33 101 GGFRMtmeRt Cracking Water Chemistry IV.B2-12 3.1.1- E G, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals RV4 GOFR4:R44~RtGR
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 49 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 50 of 90 Table 3.1.2-2-1P3 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG-T Component Intended Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Reactor Structural Stainless Treated borated Cracking - TLAA - metal IV.B2-31 3.1.1-5 A vessel support steel, water fatigue fatigue (R-53) internals CASS, components nickel alloy Lower Core,Suppor Structu:re Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1-33 E A, baffle/former support steel water > 140°F dimensions Internals -V4 (R-126) 101 assembly GeMitmet bolts Cracking Water Chemistry IV.B2-lJ 3.1.1-30 E A-Control - Primary (R-125) 94 and Secondary Reactor Vessel Internals RV4 GeMmit4RRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 51 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Loss of Reactor Vessel IV.B2-5 3.1.1-27 E A, preload Internals R-V4 (R-129) 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 E A, water > 140'F fracture Internals R-V (R-128) 101 Neutron fluence toughness eemm~tR*eat Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1-33 E X7 baffle/former support steel water > 140'F dimensions Internals RVI (R-124) 04-assembly Flow eeGGFR.e. t
- plates distribution Shielding Cracking Water Chemistry IV.B2-2 3.1.1-30 E A, Control - Primary (R-123) 4 and Secondary Reactor Vessel Internals R-V4 GGFRReGt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-3 3.1.1-22 E A, water > 140'F fracture Internals R-V4 (R-127) 101 Neutron fluence toughness ccmmitment
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 52 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1-33 E C, assembly support steel water > 140°F dimensions Internals RV4 (R-126) 101 bolts-and eewRi4.eRt screws Cracking Water Chemistry IV.B2-10 3.1.1-30 E A-Control - Primary (R-125) 4-01 and Secondary Reactor Vessel Internals R-V4 GGFe4R4GR Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-5 3.1.1-27 E A, preload Internals RV4 (R-129) 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 E A, water> 140'F fracture Internals RV4 (R-128) 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 53 of 90 Table 3.1.2-2-1P3:. Reactor Vessel Internals Aging Effect Aging NUREG- Table N Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals R-V4 (R-121) 101 axial flexure F-lew GeeO4Mef plates distribution (thermal Shi,*,i,, Cracking Water Chemistry IV.B2-8 3.1.1-30 E A-shield Control - Primary (R-120) 04 flexures) and Secondary Reactor Vessel Internals R-V4 GGcmmt4n Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-26 3.1.1-63 E material - Internals (R-142) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E A, water> 140°F fracture Internals R-V4 (R-122) 101 Neutron fluence toughness cOmmitment
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 54 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Typet Fnctone Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals R-V4 (R-121) 101
- flange P-ew eem ittme.4 Cracking Water Chemistry IV.B2-8 3.1.1-30 E A-,
Control - Primary (R-120) 4 and Secondary Reactor Vessel Internals RV4 r-' eAt PR Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 E material - Inspection (R-1 15) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E A, water > 140'F fracture Internals R-V4 (R-122) 101 Neutron fluence toughness ccMMitme4 t
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 55 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Notes Component Intended Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals R-V4 (R-121) 101
- ring Flow eGtn
° shell distribution "thermal Shielding Cracking Water Chemistry IV.B2-8 3.1.1-30 E A shield Control - Primary (R-120) 101 and Secondary Reactor Vessel Internals R-V4 GQR4.MtM Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E A, water > 140°F fracture Internals RV4 (R-122) 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 56 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.111-33 E assembly supcport steel water > 140'F dimensions Internals (R-121) 101
" lower core barrel flange Cracking Water Chemistry IV.B2-8 3.1.1-30 E weld Control - Primary (R-120)
" upper core and Secondary barrel flange Reactor Vessel weld Internals Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E, water > 140'F fracture Internals (R-122) 101 Neutron fluence toughness
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 57 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Notes Component Intended Material Environment Requiring Management 1801 Vol. Tem Type Function Management Programs 2 Item Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly distribution steel water > 1407F dimensions Internals R-V4 (R-121) 101 outlet remmit.Me-nozzles Cracking Water Chemistry IV.B2-8 3.1.1-30 E A-,
Control - Primary (R-120) 1-0"-
and Secondary Reactor Vessel Internals RV4 GeM~Mit~Me*
Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 58 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Typoen Incten Material Environment Requiring Management 1801 Vol. te Notes Type Function Management Programs 2 Item Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 EA, internals support water dimensions Internals RN-4 (R-134) 101 assembly clevis insert bolt Cracking Water Chemistry IV.B2-16 3.1.1-37 E A, Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals R-V4 GGReMmRR Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-14 3.1.1-27 E A, preload Internals RV4 (R-137) 101 Ge4RRtme4 Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 E A, water fracture Internals RN4 (R-135) 101 Neutron fluence toughness EQFmitmeRt
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 59 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Co nnt IAging Effect Aging NUREG- Table N Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 EA, internals support water dimensions Internals RV4 (R-131) 101 assembly rum.Mtment
- clevis insert Cracking Water Chemistry IV.B2-20 3.1.1-37 E A, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R-VI Loss of Water Chemistry IV.1B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-26 3.1.1-63 E material- Inspection (R-142) wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 60 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Lower Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E G, internals distribution steel water > 140OF dimensions Internals R-V (R-131) 101 assembly
- intermediate diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1-37 E G, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R-V Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 61 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 E A, internals support steel water> 140°F dimensions Internals RV4 (R-134) 101 assembly
- fuel alignment Cracking Water Chemistry IV.B2-16 3.1.1-37 E A, pin Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals R-V4 GeR~mitmRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 E A, water> 140°F .fracture Internals R-V4 (R-135) 101 Neutron fluence toughness ee:mitMe*t
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 62 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E A, internals support steel water> 140OF dimensions Internals RV4 (R-131) 101 assembly Flow Atmmitm4 lower core distribution plate Cracking Water Chemistry IV.B2-20 3.1.1-37 E A-,
Control - Primary (R-130) 0-and Secondary RVI commitmont Inservice Inspection Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-26 3.1.1-63 E material - Inspection (R-142) wear Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1-22 E A, water> 140°F fracture Internals RV4 (R-132) 101 Neutron fluence toughness GGMMtmet
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 63 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Cmponen Fnten' Material Environment Requiring Management 1801 Vol. Notes Type Function' Management Programs 2 Item Item Lower Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 E A, internals support water > 482 0 F dimensions Internals RV4 (R-139) 101 assembly Flow eGGenMt
- lower lor core distribution Cracking Water Chemistry IV.B2-24ý 3.1.1-30 E A-castings Control - Primary (R-138) 4 - column cap and Secondary
- lower core Reactor Vessel support Internals MV4 column bodies Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Thermal Aging and IV.B2-21 3.1.1-80 A water > 4820F fracture Neutron Irradiation (R-140)
Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 64 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 E A, internals support steel water > 140'F dimensions Internals R-V4 (R-134) 101 assembly
- lower core support plate Cracking Water Chemistry IV.B2-16 3.1.1-37 E A-column bolt Control - Primary (R-133) 4 and Secondary Reactor Vessel Internals RV4 GeMmitmeRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-25 3.1.1-27 E A, preload Internals -VI4 (R-136) 101 Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 E A, water > 140'F fracture Internals RV4 (R-13'5) 101 Neutron fluence toughness eeG i t.Mfe.R
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286
.Page 65 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table i Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 E A, internals support steel water > 140'F dimensions Internals RV4 (R7139) 101 assembly c9mm tMA-t
- lower core support plate Cracking Water Chemistry IV.B2-24 3.1.1-30 E A, column Control - Primary (R-138) 101.
sleeves and Secondary Reactor Vessel Internals R-V4 eGR=*)=i4Rnl p Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-22 3.1.1-22 E A, water > 1407F fracture Internals RV4 (R-141) 101 Neutron fluence toughness GOMMitMeRt
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 66 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals C p nt eAging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Item Notes Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E A, internals support steel water > 140OF dimensions Internals R-V4 (R-131) 101 assembly ,.RMte,,
radial key Cracking Water Chemistry IV.B2-20 3.1.1-37 E A, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-26 3.1.1-63 E material - Inspection (R-142) wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 67 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Tmpoen nctone Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E G, internals support steel water > 140°F dimensions Internals R-V (R-131) 101 assembly Flow secondary distribution core support Cracking Water Chemistry IV.B2-20 3.1.1-37 E G, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals RVI Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 68 of 90 Table 3.1*.2-2-1P3: Reactor Vessel Internals I
Upper Core S'upportStructure -Upper Intemials.AssembIy
-r r I T r r I RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.1B2-27 3.1.1-33 EA, tube assembly support steel water > 140'F dimensions Internals R-4 (R-1 19) 101
- bolt AAMPA Cracking Water Chemistry IV.B2-28 3.1.1-37 E A, Control - Primary (R-118) 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 EG, preload Internals RV4 (R-1 14) 101 oemm tmet
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 69 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 E A, tube assembly- support steel water > 140'F dimensions Internals RVI (R-1 17) 101 guide tube Gemmitme*
(including lower flange Cracking Water Chemistry IV.B2-30 3.1.1-30 E A, welds) Control - Primary (R-116) 4 and Secondary Reactor Vessel Internals RV4 GGFR #;RtR Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment I Docket Nos. 50-247 & 50-286 Page 70 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 E tube assembly support steel water > 140'F dimensions Internals (R-1 17) guide plates Cracking Water Chemistry IV.B2-30 3.1.1-30 E Control - Primary (R-1 16) and Secondary Reactor Vessel Internals Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-34 3.1.1-63 E material - Internals (R-1 15) wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 71 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Agn UE-Tbe1Notes Component Intended Agn UE-Table 1 Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item RCCA guide Structural Nikekl-alley, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1-33 E A, tube assembly support Stainless water dimensions Internals PRV4 (R-119) 101 support pin steel eemitmeR Cracking Water Chemistry IV.B2-28 3.1.1-37 E, Control - Primary (R-1 18) 105 A-and Secondary 4 Reactor Vessel Internals -V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 72 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Table Notes Type Fntin Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Item Core plate Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 E A, alignment pin support steel water > 140'F dimensions Internals R-V4 (R-1 13) 101 Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals R.V GGM:ImitFR:eRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 E material - Inspection (R115) wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 73 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Type Fntin Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Item Head /vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 E,,
alignment pin support steel water > 140°F dimensions Internals R-V (R-107) 101 GGemRtG*)
Cracking Water Chemistry IV.B2-42 3.1.1-30 E G, Control - Primary (R-106) 101 and Secondary Reactor Vessel Internals R-V Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 E material - Inspection (R1 15) wear
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 74 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table N Function . Material Environment Requiring Management 1801 Vol. Notes Type Management Programs 2 Item Item Hold-down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 E A, spring support steel water> 140'F dimensions Internals R-V4 (R-107) 101 eeGmitReR4 Cracking Water Chemistry IV.B2-42 3.1.1-30 E A, Control - Primary (R-106) 101 and Secondary Reactor Vessel Internals R-V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 E A-preload Internals R-V4 (R-1 14) 4-04
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 75 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table.1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Mixing Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 E G, devices support water > 482°F dimensions Internals R-V4 (R-1 10) 101
" support 'Flow ee0M8m ra 4P-At column distribution orifice base Cracking Water Chemistry IV.B2-36 3.1.1-30 E G,
" support Control - Primary (R-109) 101 column and Secondary mixer Reactor Vessel Internals R-V GGeRMtReRt Loss of Water Chemistry IVB2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Thermal Aging and IV.B2-37 3.1;1-80 A water > 482°F fracture Neutron Irradiation (R-1 11)
Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 76 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- TableN1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 E A, column support steel water > 140°F dimensions Internals RVIM (R-110) 101 Cracking Water Chemistry IV.B2-36 3.1.1-30 E A, Control - Primary (R-109) 101 and Secondary Reactor Vessel Internals R-4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Upper core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 E A, plate, fuel support steel water> 140°F dimensions Internals RV4 (R-117) 101 alignment pin Flow E)4teRt distribution Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals RV4 GemmtRet
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 77 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Type Functioen Material Environment Requiring Management 1801 Vol. Item Management Programs 2 Item Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 EA plate, support support steel water > 140'F dimensions Internals R-V4 (R-107) 101 assembly cGmmi~tt (including Cracking Water Chemistry IV.1B2-42 3.1.1-30 E A-riQ1 Control - Primary (R-106) 404-and Secondary Inservice Inspection R-V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 78 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Copon IAging Effect Aging NUREG- Table N Type Fntin Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 E A, column bolt support steel water > 140OF dimensions Internals -V4 (R-1 13 101 Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control- Primary (R-112) 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 E A, preload Internals R-V4 (R-114) 101 GQR4R4tmeat
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 79 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Notes Type Fntin Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Item InoreSupportStructure Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1-33 E A, mounted support steel water > 140OF dimensions Internals R-V4 (R-144) 101 instrumentatio e1,Gmmef*
n column Cracking Water Chemistry IV.B2-12 3.1.1-30 E A, Control - Primary (R-143) 4 and Secondary Reactor Vessel Internals P-V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 80 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Aging Effect Aging NUREG- Notes Component Intended Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1-33 E A, guide tube support steel water > 140°F dimensions Internals R-V4 (R-144) 101 Cracking Water Chemistry IV.B2-12 3.1.1-30 EA, Control - Primary (R-143) 101 and Secondary Reactor Vessel Internals R-V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Thermocouple Structural Stainless Treated borated Change in Reactor Vessel IV.B2- 11 3.1.1-33 E G, conduit support steel water > 140'F dimensions Internals R-V4 (R-144) 101 GGcmitRmRt Cracking Water Chemistry IV.B2-12 3.1.1-30 E G, Control - Primary (R-143) 101 and Secondary Reactor Vessel Internals R-V4 G-FRnminme*t
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 81 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Function Material Environment Requiring Management 1801 Vol. tem Notes Type Management Programs 2 Item Item Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 82 of 90 A.2.1.41 Reactor Vessel Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aging effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."
MRP-227 and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI) Program.
Revisions to MRP-227 and MRP-228, including any changes resulting from the NRC review of the documents (issued as MRP-227-A and MRP-228-A) will be incorporated into the IPEC RVI Program. The RVI Program will monitor the effects of aging degradation mechanisms on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227 inspection reouirements and evaluation acceptance criteria.
The IPEC RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation requirements established in MRP-227 or MRP-228, will be dispositioned in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation. To. manage loss of fracture toughness, cracking, change in dimenscins (void swellig), and loss of preload in vessel internFals c,-RpnntS, the Site willI(1) paFic-ato in the industry programs for netgtn and managing aging effectS On re-a6ter internals; (2)
.valuate and implement the results of the i programs as applicable to the reactor
-nustry i nteFrnalS and (3) up*o*nE cpletion of these programs, but not less thaRn ,1 months befe**
entering the period of extended operation, submit an inspection plan for eco nenl to the NIRC fo-r review and approval.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 83 of 90 A.3.1.41 Reactor Vessel Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aqing effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."
MRP-227 and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI) Program.
Revisions to MRP-227 and MRP-228, including any changes resultinq from the NRC review of the documents (issued as MRP-227-A and MRP-228-A) will be incorporated into the IPEC RVI Program. The RVI Proqram will monitor the effects of aqinq deqradation mechanisms on the intended function of the internals throuqh periodic and conditional examinations. The RVI Program will detect and evaluate crackinq, loss of material, reduction of fracture toughness, loss of preload and dimensional chanqes of vessel internals components in accordance with MRP-227 inspection requirements and evaluation acceptance criteria.
The IPEC RVI Program will be implemented and maintained in accordance with the quidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Proqram Guidelines." Any deviations from mandatory, needed, or qood practice implementation requirements established in MRP-227 or MRP-228, will be dispositioned in accordance with the NEI 03-08 implementation protocol. The RVI Proqram will be implemented prior to the period of extended operation. To manage l*ss of fract*-ur. toughness, crFcking, change in dim.n.sions (void swelling), and loss o~f prcload in veSSel internals comnponents, thoesite Will (1) pa~ticipate in the industr,' progarams for investigating and mnanaging aging offocts On reactor intornals; (2) evaluate and impleMcnt the rcsults of the industr' pro)gramsE as applicable to the reactor inRternals; and (3) upon completion of those programs, but not less than :214 months; boforo entering the period of extended opr. submit an inspection plan fo- reactor n*ternals to the
-atin, for NRC Ieviw and approval.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 84 of 90 Section B.1.42 of the LRA is completely new.
B.1.42 Reactor Vessel Internals Program Program Description The.Reactor Vessel Internals Program is a new plant-specific program. Revision 1 of NUREG-1801 includes no aging management program description for PWR reactor vessel internals.
NUREG-11801,Section XI.M16, PWR Vessel Internals, instead defers to the guidance provided in Chapter IV line items as appropriate. The Chapter IV line item guidance recommends actions to:
"..(1) participate in the industry programs for investigating and managing aging effects on reactor internaIs; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval."
The industry programs for investigating and managing aging effects on reactor internals are part of the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP developed inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals. These guidelines are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The I&E guidelines include:
- summary descriptions of PWR internals and functions;
- summary of the categorization and aging management strategy development of potentially susceptible locations, based on the safety and economic consequences of aging degradation;
" direction for methods, extent, and frequency of one-time, periodic, and conditional examinations and other aging management methodologies;
- acceptance criteria for the one-time, periodic, and conditional examinations and other aging management methodologies; and
- methods for evaluation of aging effects that exceed the examination acceptance criteria.
The MRP also developed inspection procedure requirements specific to the inspection methods delineated in MRP-227, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections.. These inspection procedure requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."
MRP-227 and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI) Program.
Revisions to MRP-227 and MRP-228, including any changes resulting from the NRC review of the documents (issued as MRP-227-A and MRP-228-A), will be incorporated into the IPEC RVI Program.
NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 85 of 90 The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227 inspection recommendations and evaluation acceptance criteria.
IPEC will implement and maintain the RVI Program in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines."
Any deviations from mandatory, needed, or good practice implementation activities established in MRP-227 or MRP-228, will be managed in accordance with the NEI 03-08 implementation protocol.
Evaluation
- 1. Scope of Program MRP-227 guidelines are applicable to reactor internal structural components. The scope does not include consumable items such as fuel assemblies and reactivity control assemblies which are periodically replaced based on neutron flux exposure. The scope does not include welded attachments to the reactor vessel which are'considered part of the vessel, or nuclear instrumentation (flux thimble tubes) which forms part of the reactor
.coolant pressure boundary. Other programs manage the effects of aging on these components.
MRP-227 separates PWR internals components into four groups depending on (1) their susceptibility to and tolerance of aging effects, and (2) the existence of programs that" manage the effects of aging. These groupings include:
Primary - those internals components that are highly susceptible to the effects of at least one aging mechanism (identified in Table 4-3 of MRP-227);
- . Expansion - those internals components that are highly or moderately susceptible to the effects of at least one aging mechanism, but for which functionality assessment has shown a degree of tolerance to those effects (identified in Table 4-6 of MRP-227);
Existing Programs - those internals components that are susceptible to the effects of at least one aging mechanism and for which generic and plant-specific existing AMP elements are capable of managing those effects (identified in Table 4-9 of MRP-227); and No Additional Measures - those internals components for which the effects of aging mechanisms are below the MRP-227 screening criteria (internals components not included in Tables 4-3, 4-6 or 4-9 of MRP-227).
The categorization of internals components for Westinghouse PWRs, as presented in MRP-227, applies to IPEC Unit 2 and Unit 3 vessel internals. The component inspections identified in MRP-227, Tables 4-3 and 4-6 for primary and expansion group components, define the scope of the IPEC RVI Program inspections. Those components subject to aging management by existing programs, as delineated in MRP-227, Table 4-9, are included in V
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 86 of 90 the scope of those programs, and are not part of the RVI Program inspections.
Components that are not included in Tables 4-3, 4-6 or 4-9 are considered to be within the scope of the program, but require no specific inspections.
- 2. Preventive Actions The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions. However, primary water chemistry is maintained in accordance with EPRI guidelines by the Water Chemistry Control - Primary and Secondary Program, which minimizes the potential for stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC).
Plant operations also influence aging of the vessel internals. The general assumptions about plant operations used in the development of the MRP-227 guidelines are applicable to the IPEC units. The units are base loaded and implemented low leakage core loading patterns within the first 30 years of operation. IPEC has implemented no design changes to reactor vessel internals beyond those identified in general industry guidance or recommended by Westinghouse.
- 3. Parameters Monitored or Inspected The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations and other aging management methods, as required. As described in MRP-227, the program contains elements that will monitor and inspect for the parameters that indicate the progress of each of these effects. The program will use NDE techniques to detect loss of material through wear, identify distortion of components, and locate cracks.
Visual examinations (VT-3) will be used to detect wear. Visual examinations (VT-3) will also detect distortion or cracking through indications such as gaps or displacement along component joints and broken or damaged bolt locking systems. Direct measurements of spring height will be Used to detect distortion of the internals hold down spring. Visual examinations (EVT-1) will be used to detect crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting.
(MRP-227, Tables 4-3 and 4-6)
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- 4. Detection of Aging Effects The RVI Program will detect cracking, loss of material reduction of fracture toughness, loss of preload and dimensional changes (distortion) of vessel internals components in accordance with MRP-227. The NDE systems (i.e., the combinations of equipment, procedure, and personnel) used to detect these aging effects will be qualified in accordance with MRP-228. The RVI Program will. conduct inspections of primary group components as follows (MRP-227, Table 4-3):
- Periodic visual examinations (VT-3) will detect loss of material due to wear from control rod guide tube guide plates and thermal shield flexure plates.
- Periodic visual examinations (VT-3) of the baffle former assembly plates and edge bolts will detect symptoms of distortion due to void swelling or cracking from IASCC.
These symptoms include abnormal interactions with fuel assemblies, gaps or displacement along component joints, broken or damaged bolt locking systems, and failed or missing bolts.
0 Direct measurements of spring height will detect distortion of the internals hold down spring due to a loss of stiffness. Measurements will be taken periodically, as needed to determine the life of the spring.
- Periodic visual examinations (EVT-1) will detect crack-like surface flaws of the control rod guide tube assembly lower flange welds and the upper core barrel to flange weld.
o Volumetric (UT) examinations will locate cracking of baffle former bolting. Baseline and subsequent measurements will be used to confirm the stability of the bolting pattern.
Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227, Table 5-3.
The relationships between primary group component inspection findings and additional inspections of expansion group components are as follows.
- Indications from the EVT-1 inspections of the control rod guide tube assembly lower flange welds may result in EVT-1 inspections of the lower support column bodies and VT-3 inspections of bottom mounted instrumentation column bodies to detect cracking.
- Indications from the EVT-1 inspection of the upper core barrel to flange weld may result in EVT-1 inspections of the remaining core barrel welds
- Indications from the UT inspections of baffle former bolting may result in UT inspections of the lower support column bolts and the barrel former bolts for cracking.
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- 5. Monitoring and Trending The RVI Program uses the inspection guidelines for PWR internals in MRP-227.
Inspections in accordance with these guidelines will provide timely detection of aging effects. In addition to the inspections of primary group components, expansion group components have been defined should the scope of examination and re-examination need to be expanded beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results.
IPEC will share inspection results with the industry in accordance with the good practice recommendations of MRP-227. The IPEC-specific results will be incorporated into an overall industry report that will track industry progress and will aid in evaluation of potentially significant issues, identification of fleet trends, and determination of any needed revisions to MRP-227 guidelines.
- 6. Acceptance Criteria The RVI Program acceptance criteria are from Section 5 of MRP-227. Table 5-3 of MRP-227 provides the acceptance criteria for inspections of the primary and expansion group components. The criteria for expanding the examinations from the primary group components to include the expansion group components are also delineated in MRP-227, Table 5-3. The examination acceptance criteria include: (i) specific, descriptive relevant conditions for the visual (VT-3) examinations; (ii) requirements for recording and dispositioning surface breaking indications that are detected and sized for length by the visual (EVT-1) examinations; and (iii) requirements for system-level assessment of bolted assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits.
- 7. Corrective Action Conditions adverse to quality; such as failures, malfunctions, deviations, defective material and equipment, and nonconformances; are promptly identified and corrected. In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. In addition, the cause of the significant condition adverse to quality and the corrective action implemented is documented and reported to appropriate levels of management. The Entergy (10 CFR Part 50, Appendix B) Quality Assurance Program, including relevant corrective action controls, applies to the RVI Program.
Any detected condition that does not satisfy the examination acceptance criteria must be processed through the corrective action program. Example methods for analytical disposition of unacceptable conditions are discussed or referenced in Section 6 of MRP-227. These methods or other demonstrated and verified alternative methods may be used.
The alternative of component repair and replacement of PWR internals is subject to the applicable requirements of the ASME Code Section XI.
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- 8. Confirmation Process This attribute is discussed in Section B.0.3.
- 9. Administrative Controls This attribute is discussed in Section B.0.3.
- 10. Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, specifically with regard to cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators created a program to inspect the baffle-former bolting to determine whether similar aging degradation might be expected to occur in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections.
In addition, the industry began laboratory testing projects to gather the materials data necessary to support future inspections and evaluations. Other confirmed or suspected material degradation concerns that the industry has identified for PWR components are.
wear in thimble tubes, potential wear in control rod guide tube guide plates, and cracking in some high-strength bolting. The industry has addressed the last concern primarily through replacement of high-strength bolting with bolt material that is less susceptible to cracking and by improved control of pre-load.
The RVI Program established in accordance with the MRP-227 guidelines is a new program.
Accordingly, there is no direct programmatic history for IPEC. However, program inspections will use qualified techniques similar to those successfully used at IPEC and throughout the industry for ASME Section XI Code inspections. Internals inspections (VT-3) have been conducted at IPEC in accordance with ASME Section XI Code requirements, with no indications of component degradation. IPEC has appropriately responded to industry operating experience for reactor vessel internals. For example, guide tube support pins (split pins) have been replaced in both units on the basis of industry experience. As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the IPEC units. As a result, IPEC has monitored industry developments and recommendations regarding these components.
Development of the MRP-227 guidelines is based upon industry operating experience, research data, and vendor evaluations. Reactor vessel internals aging degradation incidents in both U.S. and foreign plants were considered in the development of the MRP-227 guidelines. As implemented, this program will account for applicable future operating experience during the period of extended operation.
NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 90 of 90 Conclusion The RVI Program will be effective at managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls in accordance with MRP-227 and MRP-228 guidelines and current IPEC programs. The RVI Program will provide reasonable assurance that the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.