NUREG-1557, Technical Info from Industry Repts Addressing License Renewal, Presented at 970817-22 14th Intl Conference on Structural Mechanics in Reactor Technology in Lyon,France

Technical Info from Industry Repts Addressing License Renewal, Presented at 970817-22 14th Intl Conference on Structural Mechanics in Reactor Technology in Lyon,France

ML20217H371
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Issue date:	08/17/1997
From:	Christopher Regan NRC
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RTR-NUREG-1557 NUDOCS 9804290354
Download: ML20217H371 (10)

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14th International Conference on Structural Mechanics i

in Reactor Technology (SMIRT-14)

Lyon, France, August 17-22,1997 TechnicalInformation From Industry Reports Addressing License Renewal Christopher M. Regan U.S. Nuclear Regulatory Commission, U.S.A.

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ABSTRACT The USNRC teviewed and documented, in NUREG-1557, technical information and agreements on aging issues that resulted from the review of nine industry reports submitted by NUMARC in the early 1990's. The detrimental effects of aging and acceptable aging management programs are delineated in the text with the exception of 15 issues which have been high-lighted for continued analysis. Both the USNRC and industry have been using this report to support preparation and review of license renewal applications.

1.

INTRODUCTION In order to establish the United States Nuclear Regulatory Commission (USNRC) understanding of technical issues related to renewal of operating licenses for nuclear power plants, under Title 10, Part 54, of the U. S. Code of Federal Regulations (10 CFR Part 54) the USNRC reviewed and documented the detrimental effects of aging and the aging programs to manage these effects for certain systems, structures and components.

..e USNRC effort resulted in the publication of NUREG-1557, " Summary of Techn' cal Inforniation and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," in late 1996. This information on the detrimental effects of aging and the pertinent management programs were orighially documented in the license renewal mdustry reports (irs) submitted to the USNRC for review by the Nuclear Utilities and Resources Mangement Council (NUMARC) beginning in the late 1980's. The purpose of this paper is to st.mmarize the information contained in NUREG-1557 and the notable findings and conclusio*is reached in the aforementioned document.

2 REPORT DEVELOPMENT In the late 1980's, NUMARC, now the Nuclear Energy Institute (NEI), submitted for USNRC review ten irs addressing aging issues associated with specific structures and components of nuclear power plants [1-10], and one IR addressing the screening methodology for performing an integrated plant assessment (IPA)[11], under Title 10, Part 54, of the j

l} i 9804290354 970817 T'"

PDR NUREO 1557 C PDR

United States Code of Federal Regulations (10 CFR Part 54). The ten irs for specific structures and components are:

1. Pressurized Water Reactor (PWR) Reactor vessel [1]

2. Boiling Water Reactor (BWR) Reactor Vessel [2]

3. PWR Containment [3]

4. BWR Containment [4]

5. PWR Reactor Coolant Pressure Boundary [5]

6. BWR Reactor Coolant Pressure Boundary [6]

7. PWR Reactor Vessel Internals [7]

8. BWR Reacto-Vessel Internals [8]

9. Class I Structurc.s [9]

10. Low-Voltage, In-Containment, Environmentally Qualified Cable [10]

The original intent of the irs for specific structures and components was to serve as a referenceable surrogate for carrying out the integrated plant assessment (IPA) requirements l

of the USNRC license renewal rule,10 CFR Part 54, as published in 1990. The IPA irformation to be submitted by a prospective applicant was to describe aging affects applicable to the systems, structures, and components within the scope of license renewal for their plant and to describe and justify the programs for managing those effects for a period of extended operation beyond the original 40 year operating license. The detrimental effects of aging on certain systems, structures, and components within nuclear power plants and the 4

suggested programs for managing these detrimental effects of aging had been described in the irs for use by individual utilities seeking to submit an application to the USNRC for a renewed operating license.

In 1992 the USNRC staff and industry resources were redirected and review of the irs was I

terminated. USNRC and industry efforts were concentrated on revising the license renewal rule to focus on the effects of aging rather than an indeterminable number of aging mechanisms. Nonetheless, it was determined after finalizatica of the revised license renewal rule in 1995, that the effort already expended on review of the irs should be utilized. The technical information and agreements reached at the point of review cessation, therefore, were to be incorporated into the draft 'USNRC standard review plan for license renewal i

(SRP-LR).

After a hiatus in IR review activities, NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," was completed.

This report summarized the technical information and NUMARC/USNRC agreements reached from nine of the ten irs, the cable IR [10] was excluded. The cable IR addresses the issue of environmental qualification (EQ) 7 of electric equipment, which at the time of NUREG-1557 development, had been superseded by the USNRC EQ action plan to address aging of cables. The technical information and agreements documented in NUREG-1557 represent the status of the USNRC staff's review when the USNRC and industry resources were redirected to address license renewal rule implementation issues.

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3.

TECHNICAL INFORMATION AND AGREEMENT FORMAT NUREG-1557 encompasses nine of the original ten NUMARC irs, the relevant USNRC staff review documentation for each IR, and the NUMARC responses / positions taken with respect to the USNRC review. All the relevant documentation has been compiled and is listed in Appendix A of NUREG-1557. The technical information and agreements from the USNRC review have been compiled into tables, see examples Exhibits 1 and 2, and are presented in Appendix B of NUREG-1557. Each table consists of seven columns. The columns list among other things the aging-related degradation mechanisms (ARDMs) addressed in the irs and their effects on structures and components. The effects of ARDMs were based primarily on information in the irs. In summary the following "ARDMs" and their corresponding " effects" heve been considered to affect structures and components in the l

Reactor Pressure Vessel (RPV), Reactor Vessel Internals, and the Primary Coolant Pressure Boundary (PCPB).

Anine Mechanism Anine Effects j

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1. Corrosion, Microbiologically Loss of material +"+

induced corrosion, Boric Acid +

corrosion

2. Creep Change in dimension

3. Erosion / Corrosion (E/C)

Wall thinning

4. Fatigue Cumulative fatigue damage l

5. Stress Corrosion Cracking (SCC)"

Crack initiation and growth (to include:IGSCC, TGSCC, & IASCC) l

6. Neutron Irradiation Embrittlement

• Loss of fracture toughness l

-7.

Stress Relaxation Loss of preload l

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8. Wear" Attrition i

9. Thermal Embrittlement ("+x"*)

Loss of fracture toughness

• Boric acid wastage of external surfaces j

• *1GSCC-Intergranular SCC; TGSCC.Transgranular SCC; IASCC-Irradiation Assisted SCC.

• Includes Cast Austenitic Stainless Steel (CASS) (BWR Primary Coolant Pressure Boundary)

(

• * *

• Also listed as an Effect: " corrosion product buildup" (PWR Vessel Internals and BWR Pressure Vessel)

'Also listed as Irradiation Embrittlement (PWR Vessel Internals)

~Also listed as Fretting and Wear (PWR and BWR Pressure Vessel and BWR Vessel Internals) and Mechanical Wear (BWR and PWR Primary Coolant Pressure Boundary)

~*Also listed as Thermal Aging (PWR Pressure Vessel) 721-4

The following "ARDMs" and their " effects" have been considered to affect structures and components in the Reactor Containment and Class I Structures:

Anine Mechanhm Anine Effects Concrete Structures -

1. Freeze Thaw Scaling, cracking, and spalling

2. Leaching of Calcium Hydroxide Increase of porosity & permeability

3. Aggressive Chemical Attack' Increase of porosity & permeability, cracking, and spalling

4. Reaction with Aggregates Expansion and cracking

5. Elevated Temperature Loss of strength and modulus

6. Irradiation of Concrete Loss of strength and modulus

7. Creep Deformation

8. Shrinkage Cracking

9. Corrosion Loss of material j

10. Abrasion and Cavitation Loss of material

11. Restraint, Shrinkage, Creep, &

Cracking of masonry walls

)

Aggressive Environment

12. Concrete Interaction with Aluminum Loss of strength

13. Cathodic Protection Current Cathodic protection effect on bond strength Structural Steel & Stainless Steel Liner -

1. Corrosion, Local corrosion, Loss of material Atmospheric corrosion

2. Elevated Temperature Loss of strength and modulus

3. Irradiation,

IAss of fracture toughness

4. Stress Corrosion Cracking Crack initiation and growth j

Reinforcing Steel (Rebar) -

l. Corrosion of Embedded Steel Cracking, spalling, loss of bond,

& Loss of material

2. Elevated Temperature Loss of strength and modulus

3. Irradiation Loss of strength and modulus Miscellaneous -

1. Fatigue Cumulative fatigue damage

2. Settlement" Cracking, distortion, increase in component stress level

3. Mechanical Wear Lockup""

4. Strain Aging (of Carbon Steel)+

Loss of fracture toughness

5. Loss of Prestress"*

Reduction of design margin

6. Corrosion of Steel Piles Loss of material

7. Corrosion of Tendons Loss of material

' Includes Stainless Steel-Bellows (PWR Containments)

"Also listed as " Aggressive Chemicals" (PWR Containments)

'"Also listed as " Differential Settlement" (BWR Containments)

• • 'Also listed as 'Prestress Losses" (BWR Contaimnents)

~~~Also listed as ' Attrition" (PWR Containments) 721-5

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Me M Drief sumrnary q/technkal fr/anatkat art:i NUMARC/NRC agreernerats frtpri TWR corttatrunent structures (rt:fustry reppt Z Renated NRC Degr=l=H=t Aging Comment NUMARC/NRC Dania for Mechanism Efterte Components Matertalaa Numbert>

nt or N b.. ant or Pmpoeal Agreessve increase of Concrete Contmartmen -

Concrete G - 12.

For concrete containment Degradation caused by aggres-l Chemical poroasty &

Reinforced / Prestressed S S.

structures that meet the be-sive chemical attack is non-Attack permembthty

• Concrete Donic S-38 to sus requirements. aggressive signtAcant for concrete contain-l-

cracking. &

• Concrete Containment Wall S 41 chemkal attack is non-ment structures not espaard to j

l spalung Above Grade significant ARDM.

aggreserve erwtronment (pH I

<5.5L or to chlortde or sulfate solutions beyond defined hmits 1 500 ppm chloride.5 and 1500 ppni sulfateteor i en-poord to ground water that ex-ceeds the pH. chk>rlde sulfate hmits the exposure is for inter-mittent perkids only Agresolve increase of Concrete Containments Concrete G - 10.

Arresalble concrete surfaces in cases where containment Chem 6 cal porostry &

Retnforced/ Prestressed G 13.

are periodically examined in concrete is exposed to aggres-Attack permembthey.

= Concrete Containment Wall G 15.

accordance with the pmce-save groundwater (pH e5 5.

cracking. &

Helow Grade S 25.

dures of Type A Integrated chloride >500 ppm, & sulfale

?

spalhng

• Concrete fleermat S 36.

leak rate test. or en accordance s l500 ppml. periodic inspec-Free-Standing Steel Containment S 37.

with ASME Sect XI. Subsart.

tion of accessible concrete sur-with Flat Bottom & en Ice S 41.

IWL a fores as part of Type A Inte-Condeneer S 65.

grened leak rate test performed

• Concrete Basemat S 66.

Management for the effect; of under Appendix J.10CFR50.7 5 ti9.

aggresetve chemscal attack of or in accordance with ASME S 72.

coricrete surfaces that are not Sect. XI. Subecci fWL.8 exam.

S 75 rerkidically examined due to category L A. & guidelb.cs of inaccessiblitty requires further ACI20189 plant operthe evaluation Further evaluation for management of inaccesaible areas is to be justifled on a plant speranc bams EXHIBIT 1. Sample Table (B3) from NUREG-1557:PWR Containment Structures'IR I

Tame D8. Drief sumanary of techrikal trVarmatirm and NUhtARC/NRC agreemeritsfnan UMN sm4 interrtals industry report Agtng-Related NRC Degradataan Ag:ing Comment NUMARC/NRC Bases for Mechanism Effects Components Materials Numtere' Agreement or Proposal Agrement or Proposef IGSCC Crack Access Hole Cover Alloy (500 G 2.

GESIL 462SI 8 recommends Recommendations of GESIL initiatkin &

G S.

volumetric inspections. tm-462S18 & aslety analysts are gmwth o.9.

piementauon is plant specthc.

current & ettertsve insperunn G-il.

& recommended repair is to programs for detection & eval G 22.

attach reinforcement hardware usuon repair of access hole G 24.

covers Core Shroud Head Dohs SS G 27 GESIL 4332 recommends IR Recommenda tons t'GESIL Allov bnD G 2R.

examinatkm durtng outages.

4332 & rep 6 cement with crevare G 29.

Implementatkan is plant free design are current & effec-G 33.

specific, a replacement is with uve Inspection programs for de-G 34.

crevere free design tection & evaluation replace-S3 ment of core shmud head tmits Control Blade SS S4 GESIL 157 routine replace.

Routine replacement & opera.

S6 nent 3 operat6onal parameter tional pararneter monnortng are 5 17.

monnorirt inspertkin.

rurrent & cflective programs for 4

detertkm & evaluatinn-S 22.

evaluation. & replacement S 25.

reptarement of contrnt blades Contml Rod Detve (CHDI SN S 26 to AsME Sett XISrequires vul ASME Sect XI. Subscrt IWDJ Haussne 5 29 umetric esam of welds & VI 2 czam. categories it O & D P is S 31.

of pressure retaming cu-rent & cfictuve program for S 32.

Imundary & system leakage &

detection & evaluatkm repair S 3M hyttm staur teus replacement of CHD housing Core Spray Sparger SS S 49 NHC liulietln HD 136 rerum NRC ftulletin MO 136 & safety 5 54.

mends venual inspertkm during analysts are effective insgertkm S 55 refucieng outages, analytical programs for detection. evalum

)

evalmenon & rep 4 sr tum repair of core aps,araparger_

Intermediate kar*c Momtor/

SS GLSIL 4(F). key IIrrt um Retnmmendations of GLSIL 1

Sourre Range Moristor mrnds visual insperterm.

404.7 leakage monstonna. & re i

.i (IRM/SRMI Dry Tubes leakage monnonna re-plarement with crevere free de placement is with trevire free sign. are effernve annpection i

design. & resistant maternal programs for deterthm & evalua tion replacement nldry tubre EXHIBIT 2. Sample Table (B8) from NUREG-1557:BWR Vessel Internals IR i

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In addition, the tables list the specific structures and/or components, and their construction materials. For each IR, a complete list of what are defined as structures and components is included in the end of the table (see Exhibit 3).

In general, the irs present only representative examples and do not provide a comprehensive list of the type, grade, and specification of materials used for various reactor structures and components but give a good baseline from which comparative reviews by a prospective license renewal applicant can begin. For most irs, only material categories such as stainless steel (SS), cast austenitic stainless steel (CASS), Ni Alloy, or carbon steel (CS), are described. A detailed list of material type and grade is, however, provided in the PWR and BWR reactor vessel irs.

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l UST OF BWH COMIAINMDrf COMPONENTS MAM I STEEL CONTAINMENT MAAK $ CONCRETE CONTMMMENT MARK III STEEL CONTMNMENTS Drywellintertor Surface Drywell uner intertor Surfm e Contatrunent ShellInternor Surface Drywell Extertor Surfa c Drywell uner Exterkw Surface Contamment Shell Exterkw Surface j-Drywell Head Torus Laner Interior Surtan Suppr. Chamter Shell interkw Surface l

Embedded Shell Region Torus Liner interior Surim e at Wateriene Suppr. Chamter Snell Exterkw Surfare Drywell Support Skirt Torus Larr Exterior Surfaa r flasemme uncr Saml Perlu t Hrgkm lure Aswhors uner Arrtywa Torus Inscrior Surfm c litywell Cmwrrte C<merete Ikasemat Torus interior Surfai e at Walcrime Torus Curu rric Concrete Fillin Annulus Torus Esecrka Surf.n e Drywell Corw rcle Heinfurring Sacri Emledded Shell Hegion l

Torum Hang Gwder Torum Cum retr Hemhn mg hiert MARK III CONCRETE CONTMMMENTS Vent Lancs VentLums Contamment Liner bterkw Surfe e Vens u elirilows Vent 1 inw licllows Containmeni Laner Emiertor Surfaw I

Vene llender Vent licaders Suppr Chamler Lar.cr or Claddma Interior Surlare f

Duwnromers and br.u mg lbwm muern ami lir.u ing Suppr. Chamter uner Emerrior Surim e I

Veni System Supports Vent Synarm Suppurte Contreie Containment Wall Ahrwe Grade l

Torus Seismic Hestramte Drywell licad Corwrcle Cunlainment Wall itelow Gnade Torum Support Columns / Saddles MARK II CONCRETE CONTMNMENTS Concrete Dunw t

ECCS Sucelon Header Drywell uner Inscriur Surfare Itasemat Liner Orean Plant with Uncoated CS Surfares Drywell1.iner Emienor Suriarc Concrete Elasemat i

Unrusted Submerged CS Surias en Suppr Chamter L6ner interka Surface unct Anchorm l

MARE II STEEL CONTMMME,rTS buppe. Chamler Liner intertor Surtare at Walerkne Containment Wall Helnforritig Steel l

Drywee interior Surface Suppr Chamter uner t'atertur Surim e Dome Heinfuscing Steel Drywell Extertor Surfac e Liner Arnhors llancmat Heinforcing Steel Drywell Head Liner Region Shielded by Diapteragm Fkor COMMON COMPOfrnNTS Suppe Chamber Exterkw Surfare Containment Concrete Penetration $1ceves l

Suppr. Chamter Interkw Surface Concrete Containnwnt Hemforcing Secci Dessuntlar Metal Welds Suppr. Chamtwr interior Surfac e at Waterhne Drywell Head Penetratkm llellows Regen Shielded by Diaphragm Floor Downcomer Pales and Itracing Personnel Airlutk Embedded Shell Hegion Concrete liasemat Equipment Hatches Sand Pucket Reghm Banemat uner CHD Hatch Support Skirt Danemat Heinforrtng Sicci Duwiummer Pipes and Praring Prestressmg Tendons and Durtn Oman Plant with Unconted CS Surfaces Uncoated Submerged CS Surfaces EXHIBIT 3. Sample Listing of IR Components: BWR Containment Components The USNRC staff review of each original IR resulted in a set of comments / issues for each IR which are referenced in NUREG-1557 in order to recreate the lineage of the aging issue agreement.

NUMARC/USNRC agreements or proposals on whether a ARDM or ARDM/ component combination is potentially significant, and if it is potentially significant, is given and a brief description of the program that can adequatcly manage the effects of 1

aging is presented in a similar fashion. The technical basis for these agreements or proposals, including assumptions and references, are also described in briefin NUREG-1557.

A few examples of the information delineated in the report as aging management programs and their bases are given below.

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1.

For a specific ARDM or ARDM/ component combination, if the effects of aging are not potentially significant, "non-significant" is listed in the agreements column.

The technical basis, assumptions, and references for the agreement are presented in column seven. For example, the effect of creep is non-significant for BWR primary coolant piping and fittings fabricated from carbon steel (CS) or stainless steel (SS) because the reactor operating temperatures are significantly lower than the temperatures at which creep is a concern to CS and for SS components. Also, if the effects of aging are not potentially significant when certain bounding conditions are met, then "for components that meet the basis requirements, this ARDM is non-significant" is listed in the agreements column. For example, the effects of freeze thaw is non-significant for Class I concrete structures that meet the following criteria: located in geographic regions of negligible weathering conditions (weathering index <100 day-inch / year): and if located in severe

~

l weathering conditions (weathering index 100-500 day-inch / year) the concrete mix design meets the air content and water-to-cement ration requirements of American l

Concrete Institute (ACI) 318-63 or ACI-349-85.

l 2.

If a specific ARDM/ component combination is potentially significant and the effects of aging are adequately addressed by current management programs, then l

a brief description of the program is provided in column six. For example, the l

program delineated in NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," July 31, i

1977, and implemented through USNRC Generic Letter 88-01, "USNRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," January 25,1988, is a current and adequate program to manage the effects of intergranular stress corrosion cracking (IGSCC) of SS piping and fittings of BWR primary coolant l

pressure boundary.

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3.

If a specific ARDM/ component combination is potentially significant and the l

current programs are not adequate for managing the effects of aging, then column six simply states " current practices to be enhanced, select plant-specific aging management." For these cases, the NUMARC recommended aging management l

options are described in Chapter 6 of the irs. However, chapter six was not the l

focus of the USNRC review of the irs.

4.

An ARDM or ARDM/ component combination is listed as " unresolved issue" if no agreement was reached between NUMARC and the USNRC staff. For these cases, both the NUMARC and USNRC proposals are briefly described in the agreements column. An example of an unresolved issue is the effects of thermal aging embrittlement on PWR primary coolant system components fabricated from cast austenitic stainless steel (CASS). The NUMARC proposal considers a ferrite content screening criterion and American Society of Mechanical Engineers

- (ASME) Code Section XI, Subsection IWB, inspection to be an adequate program for managing the effects of thermal embrittlement.

The USNRC proposal, however, considers that ferrite content criterion is inadequate for screening and VT-3 visual examination is not intended or reliable for detecting tight cracks.

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SUMMARY

AND OBSERVATIONS Nine of the ten irs submitted by NUMARC addressing the detrimental effects of aging associated with specific structures and components of nuclear power plants were reviewed by the USNRC. The technical information and NUMARC/USNRC agreements for each IR have been compiled into tables (Appendix B to NUREG-1557). The information presented in each of the tables includes specific structures and components and their materials of construction; ARDMs and their effects on structures and components; relevant comments of the USNRC staff; and the NUMARC/USNRC agreements or proposals and their technical bases, including assumptions and references.

Considerable effort was expended by industry representatives and the USNRC staff to come to agreement on the aging issues pertinent to several major structures and components in nuclear power plants. These aging issues and associated structures and components are of prime importance for a prospective applicant aiming to satisfy the requirements of the revised license renewal rule,10 CFR Part 54, published in 1995. NUREG-1557 documents the USNRC understanding of the information delineated in the irs and the agreed upon and disagreed upon technical positions from the previous USNRC review effort. Resulting from the industry /USNRC exchanges it was determined that fifteen open technical issues existed.

These open technical issues include the following:

)

(1) Fatigue in Metal Components (2) Environmental Qualification of Cables (3) Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components (4) Irradiation-Assisted Stress Corrosion Cracking of Reactor Internals Components (5) Stress Relaxation of PWR Internals Components (6) Primary Water Stress Corrosion Cracking of High-Nickel Alloys'

'(7) Stress Corrosion Cracking in PWR Metal Components (8) Neutron Irradiation Embrittlement (Definition of Reactor Vessel Beltline)

(9) Freeze-Thaw Damage in Concrete Structures (Significance of Effects)

(10) Alkali-Aggregate Reactions in Concrete Structures (Significance of Effects)

(11) Differential Settlement of PWR Containments and Class I Structures (Significance of Effects)

_(12) Reinforcement Corrosion in PWR Concrete Containments (Significance of Effects)

(13) One Time Inspections of Concrete and Steel Structures (14) Ultrasonic Inspection of Pressure Vessels and Components (15) Visual Inspection of Components and Structures using certain American Society of Mechanical Engineers Code Acceptance Criteria Resolution of these technical issues have been slated to be the focus of USNRC license renewal reviews. Resolution of these technical issues, for a prospective license renewal applicant, is paramount to successfully meeting the requirements of the revised license renewal rule,10 CFR Part 54. Several of these issues are also cur ently under review by the USNRC for applicability to operating reactors for their existing 40 year operating license.

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In the future, as U.S. utilities begin preparation of an application for license renewal, NUREG-1557 will have established a starting point to begin an aging management analysis.

Documented in NUREG-1557 are the aging management programs deemed acceptable by the USNRC and the technical issues where the USNRC and the industry hold diverse positions.

It is resolution of the open technical issues elucidated by this report that is necessary to provide a stable and predicable regulatory environment for a license renewal applicant. By referencing this publicly available report, a prospective license renewal applicant can become familiar with the level of effort necessary to initiate aging studies of other systems, structures, and components that may also be subject to the requirements of the license renewal rule. NUREG-1557 can be used as a time saving and a cost cutting mechanism by a license renewal applicant. In addition to the benefits for industry utilities, the USNRC is currently considering incorporation of the appropriate technical information and agreements as the basis for a revised draft USNRC SRP-LR. This new SRP-LR will establish the l

methods and acceptance criteria that meet 1he requirements of 10 CFR Part 54 that the l

USNRC staff will utilize to perform a review of future license renewal applications.

5.

CONCLUSIONS Industry submittal and subsequent USNRC review of nine of the ten irs resulted in the establishment of a clear understanding of industry and USNRC positions related to the detrimental effects of aging and the programs necesse7 or managing them. NUREG-1557 f

concisely summarizes these agreements and disagreements in a format usable to a prospective license renewal applicant. The information delineated may afford a prospective applicant an introduction to the significant aging issues needing resolution. Although several aging issues remain unresolved the report contains valuable information that may be referenced by a l

prospective applicant saving them both time and resources. It should also be noted that the l

USNRC will consider utilizing this report by incorporating resolved aging issues into a draft SRP-LR that will meet the requirement of the revised license renewal rule,10 CFR Part 54.

l 6.

REFERENCES 1.

Nuclear Management and Resources Council. September 1992. Pressure Water Reactor Vessel License Renewal Industry Report, May 1990. Revision 1. Report Number 90-04.

2.

Nuclear Management and Resources Council. September 1992. Boiling Water Reactor Vessel License Renewal Industry Report, October 1989. Revision 1. Report Number 90-02.

3.

Nuclear Management and Resources Council. September 1991. Pressurized Water Reactor Containment Structures License Renewal Industry Report, August 1989. Revision

1. Report Number 90-01.

4.

Nuclear %..iage~" ed Resources Council. December 1991. Boiling Water Reactor Containments License Renewal lhdc. y Report, July 1990. Revision 1. Report Number 90-10.

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5.

Nuclear Management and Resources Council. May 1992. PWR Reactor Coolant System License Renewal Industry Report, October 1990. Revision 1. Report Number 90-07.

i 6.

Nuclear Management and Resources Council. April 1992. BWR Primary Coolant Pressure Boundary License Renewal Industry Report, September 1990. Revision 1. Report Number 90-09.

7.

Nuclear Management and Resources Council. December 1992. Pressurized Water Reactor Vessel Internals License Renewal Industry Report, September 1990. Revision 1.

Report Number 90-05.

i i

8.

Nuclear Management and Resources Council. June 1992. Boiling Water Reactor Vessel Internals License Renewal Industry Report, February 1990. Revision 1. Report Number 90-03.

9.

Nuclear Management and Resources Council. December 1991. Class I Structures License Renewal Industry Report, June 1990. Revision 1. Report Number 90-06.

10. Nuclear Management and Resources Council. March 1993. Low-Voltage, In-Containment, Environmentally-Qualified Cable License Renewal Industry Report, July 1990.

Revision 1. Report Number 90-08.

11. Nuclear Management and Resources Council. October 6,1989. Methodology to i

Evaluate Plant Equipment for License Renewal.

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