|
Home
> Nuclear Reactors > Operating Reactors > Licensing > Renewal
> Guidance
Reactor License Renewal Guidance Documents
These documents can be found below (if available) or by using NRC's on-line
document retrieval system ADAMS located in the Electronic Reading Room.
Should you need assistance in locating documents of interest, we encourage you to contact the reference
staff in our Public Document Room.
On this page:
The following license renewal guidance documents have been updated:
Generic Aging Lessons Learned Report, Standard Review Plan, and Regulatory
Guide 1.188.
For additional information, see
Guidance, Updates, Schedule and Background.
Some links on this page are to documents in our Agencywide
Documents Access and Management System (ADAMS), and others are to documents
in Adobe Portable Document Format (PDF). ADAMS documents are provided in either
PDF or Tagged Image File Format (TIFF). To obtain free viewers for displaying
these formats, see our Plugins, Viewers, and
Other Tools page. If you have questions about search techniques or problems
with viewing or printing documents from ADAMS, please contact the Public
Document Room staff.
Generic Aging Lessons Learned (GALL)
The purpose of the GALL Report is to provide the technical basis for
the Standard Review Plan for License Renewal. GALL contains the staff's
generic evaluation of the existing plant programs and documents the technical
basis for determining where existing programs are adequate without modification
and where existing programs should be augmented for the extended period
of operation. The evaluation results documented in the GALL report indicate
that many of the existing programs are adequate to manage the aging effects
for particular structures or components for license renewal without change.
The GALL report also contains recommendations on specific areas for which
existing programs should be augmented for license renewal.
Generic Aging Lessons Learned (GALL) Report, NUREG-1801
- NUREG-1801, "Generic Aging Lessons Learned GALL) Report,"
Revision 1
Inspection Manual Chapters (IMCs) and Procedures (IPs)
The purpose of the IMCs and IPs is to guide the staff in conducting inspections
to ensure that licensees meet the NRC's regulatory requirements:
- IMC 2515, “Light-Water Reactor Inspection Program — Operations Phase,” provides guidance to NRC headquarters and regional office staff, as well as consultant personnel, regarding review and inspection activities to ensure the safe operation of light-water reactors, including post-approval site inspections associated with license renewal (see IP-71003, below).
- IMC 2516, "Policy and Guidance for the License
Renewal Inspection Programs," provides guidance to NRC headquarters and regional office staff, as well as consultant personnel, regarding review and inspection activities associated with
an applicant's license renewal program.
- IP-71002, "License Renewal Inspections,"
provides the procedures for inspecting and verifying the documentation, implementation, and effectiveness of the programs
and activities associated with an applicant's license renewal program. For additional information, see Frequently Asked Questions on License Renewal of Nuclear Power Reactors
(NUREG-1850).
- IP-71003, "Post-Approval Site Inspection for License Renewal," provides the procedures for inspecting and verifying the completion of license renewal commitments and license conditions that have been added as part of the renewed license, and ensuring that selected aging management programs (AMPs) are implemented in accordance with the license renewal regulations. For additional information, see Frequently Asked Questions About License Renewal Inspection Procedure (IP) 71003.
Interim Staff Guidance for License Renewal (LR-ISG)
Improved license renewal guidance documents, such as GALL Report, Rev. 1, SRP-LR Rev 1, Reg Guide 1.188, Rev. 1,
and NEI 95-10, Rev. 6, have been developed to enhance the license renewal application process. (For staff's resolution associated with LR-ISGs for 2001-2005, see the Summary of the 2001-2005 Interim Staff Guidance for License Renewal) As lessons are
learned during license renewal application (LRA) reviews, it is expected that these guidance documents may need to
be revised to capture new insights or address emerging issues. To document these lessons learned the staff has developed
an interim staff
guidance (LR-ISG) process that involves the industry and other stakeholders. LR-ISG improves the efficiency and
effectiveness of the license renewal process by providing guidance to license renewal applicants and other interested
stakeholders, until the emerging issues can be incorporated into the next revision of the improved license renewal guidance
documents. For the most recent status see Interim Staff
Guidance Associated with License Renewal Guidance.
Regulatory Guide 1.188, Revision 1
Regulatory Guide 1.188, Revision 1, provides format and content for applications and endorses NEI 95-10, Rev 6. However, applicants may elect to use other suitable methods or approaches for satisfying the Rule's requirements and completing a license renewal application.
- "Industry Guidelines for Implementing the Requirements of 10 CFR
Part 54 - The License Renewal Rule," NEI 95-10, provides potential
applicants with guidelines for identifying the systems, structures,
and components with the scope of 10 CFR Part 54 and their functions;
identifying structures and components subject to aging management review;
assuring that effect of aging are maintained, application of inspections
for license renewal; identifying and resolving time-limited aging analyses;
and identifying and evaluating exemptions containing time-limited
aging analyses.
Office Instructions (OIs)
The NRR Office Instruction (OI) program defines the processes by which NRR staff and managers develop and maintain office instructions and there by improve NRR's efficiency and consistency in performing its various activities and functions.
RNWL-100, "License Renewal Application Review Process," (ADAMS #ML053550456) describes the license renewal application (LRA) review process and the roles and responsibilities of the staff organizations involved in an LRA review.
Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues, LIC-203 (ADAMS# ML033550003), establishes procedures and provided guidance pertaining to the preparation of environmental assessments and the consideration of environmental issues for all licensing action
Regulatory Guides (RGs)
The purpose of the RG is to provide guidance to licensees and applicants
on implementing specific parts of the NRC's regulations, techniques used
by the NRC staff in evaluating specific problems or postulated accidents,
and data needed by the staff in its review of applications.
- Standard Format and Content for Applications to Renew Nuclear Power
Plant Operating License,
Regulatory Guide 1.188, endorses the industry license renewal document NEI
95-10, Revision 6.
- Preparation of Supplemental Environmental Reports for Applications
To Renew Nuclear Power Plant Operating Licenses,
Supplement 1 to Regulatory Guide 4.2, provides guidance on the format and
content of an environmental report to be submitted as part of a license renewal application.
Standard Review Plans (SRP)
The purpose of the SRP is to assure quality and uniformity of staff reviews and to present a well defined base from
which to evaluate a licensee's application.
- Standard Review Plan for License Renewal,
NUREG-1800, incorporates by reference the Generic Aging Lessons Learned Report and Regulatory Guide 1.188.
- Environmental Standard Review Plan, NUREG-1555, Operating License Renewal Supplement 1,
provides guidance in implementing the 10 CFR Part 51,
related to new site/plant applications and license renewal.
Technical Reports in the NUREG Series (NUREGs)
Document Identifier |
Title |
NUREG-1412 (Accession# 9201060078) |
"Foundation for the Adequacy of the Licensing Basis,"
describes the regulatory process that assures that any plant-specific
licensing bases will provide reasonable assurance that the operation
of nuclear power plants will not be inimical to the public health
and safety to the end of the renewal period. |
NUREG-1437 |
"Generic Environmental Impact Statement for License Renewal of Nuclear Plants," examines the possible
environmental impacts that could occur as a result of renewing licenses of individual
nuclear power plants under 10 CFR Part 54.
- Supplement
1 - Calvert Cliffs Nuclear Power Plant
- Supplement
2 - Oconee Nuclear Station
- Supplement
3 - Arkansas Nuclear One, Unit 1
-
Supplement 4 - Edwin I. Hatch Nuclear Plant
-
Supplement 5 - Turkey Point, Units 3 and 4
-
Supplement 6 - Surry, Units 1 and 2
-
Supplement 7 - North Anna, Units 1 and 2
- Supplement
8 - McGuire, Units 1 and 2
-
Supplement 9 - Catawba, Units 1 and 2
-
Supplement 10 - Peach Bottom, Units 2 and 3
- Supplement
11 - St. Lucie, Units 1 and 2
- Supplement
12 - Ft. Calhoun Station, Unit 1
- Supplement
13 - H.B. Robinson Steam , Unit 2
- Supplement
14 - Ginna Nuclear Power Plant
- Supplement
15 - V.C. Summer Nuclear Station
- Supplement
16 - Quad Cities, Units 1 and 2
- Supplement
17 - Dresden, Units 2 and 3
- Supplement
18 - Farley, Units 1 and 2
- Supplement
19 - Arkansas Nuclear One, Unit 2
- Supplement
20 - D.C. Cook Nuclear Plant
- Supplement
21 - Browns Ferry Nuclear Plant, Units 1,2,3
- Supplement 22 - Millstone Power Station, Units 2 and 3
- Supplement 23 - Point Beach Nuclear Nuclear Plant
- Supplement 24 - Nine Mile Point Nuclear Station, Units 1 and 2
- Supplement 25 - Brunswick Steam Electric Plant, Units 1 and 2
- Supplement 26 - Monticello Nuclear Generating Plant
- Supplement 27 - Palisades Nuclear Plant
- Supplement 28 - Oyster Creek Nuclear Generating Station
|
NUREG-1555 |
"Standard Review Plans for Environmental Reviews For Nuclear Power Plants," provides guidance to the staff in
implementing provisions of 10 CFR 51, "Environmental Protection Regulations for Domestic Licensing and Related
Regulatory Functions," related to new site/plant applications. |
NUREG-1555,
Supplement 1 |
"Standard Review Plans for Environmental Reviews For Nuclear Power Plants - Operating License Renewal,"
provides guidance to the staff in implementing provisions of 10 CFR 51, "Environmental Protection Regulations
for Domestic Licensing and Related Regulatory Functions," related to reactor operating license renewals. |
NUREG-1568 (Accession# 9701130209) |
"License Renewal Demonstration Program: NRC Observations
and Lessons Learned," is a summary of NRC staff's observations and
lessons learned from nuclear plant site visits performed under the
License Renewal Demonstration Program (LRDP). The LRDP was a program
established by the NEI to assess the effectiveness of NEI 95-10. |
NUREG-1611 (Accession# 9710100155) |
"Aging Management of Nuclear Power Plant Containments
for License Renewal," reconciles the technical information and agreements
resulting from NUMARC/NRC industry report reviews and the in-service
inspection requirements of Subsection IWE and IWL as promulgated in
10 CFR 5.55a for license renewal. |
NUREG-1705 |
Safety Evaluation Report Related to the License Renewal of Calvert Cliffs Nuclear Power Plant, Units 1 and 2 |
NUREG-1723 |
Safety Evaluation Report Related to the License Renewal
of Oconee Nuclear Station, Units 1, 2 and 3 |
NUREG-1739 |
"Analysis of Public Comments on the Improved License
Renewal Guidance Documents," contains the NRC staff's analysis of
the stakeholders' comments on the license renewal guidance documents,
which are Regulatory Guide-1.188, Standard Review Plan for License
Renewal, Generic Aging Lessons Learned Report, and the NEI 95-10,
Rev. 3. |
NUREG-1743 |
Safety Evaluation Report Related to the License Renewal of Arkansas Nuclear One, Unit 1 |
NUREG-1759 |
Safety Evaluation Report Related to the License Renewal of Turkey Point Nuclear Plant, Units 3 and 4 |
NUREG-1766 |
Safety Evaluation Report Related to the License Renewal of North Anna Power Station, Units 1 and 2,
and Surry Power Station Station, Units 1 and 2 |
NUREG-1769 |
Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 & 3 |
NUREG-1772 |
Safety Evaluation Report Related to the License Renewal of McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station
Station, Units 1 and 2 |
NUREG-1779 |
Safety Evaluation Report Related to the License Renewal of St. Lucie Nuclear Plant, Units 1 and 2 |
NUREG-1782 |
Safety Evaluation Report Related to the License Renewal of the Fort Calhoun Station, Unit 1 |
NUREG-1785 |
Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2 |
NUREG-1786 |
Safety Evaluation Report Related to the License Renewal of the R.E. Ginna Nuclear Power Plant |
NUREG-1796 |
Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3, and
Quad Cities Nuclear Power Station, Units 1 and 2 |
NUREG-1800 |
"Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," provides guidance to the
NRC staff reviewers in performing safety reviews of applications to renew nuclear power plant licenses in accordance
with 10 CFR Part 54. |
NUREG-1801 |
"Generic Aging Lessons Learned (GALL) Report, Summary and Tabulation of Results," contains the NRC staff's
generic evaluation of the existing plant programs and documents the technical basis for determining where existing
programs are adequate without modification and where existing programs should be augmented for the extended period
of operation.
|
NUREG-1803 |
Safety Evaluation Report Related to the License Renewal of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 |
NUREG-1825 |
Safety Evaluation Report Related to the License Renewal of the Joseph M. Farley Nuclear Plant, Units 1 and 2 |
NUREG-1828 |
Safety Evaluation Report Related to the License Renewal of the Arkansas Nuclear One, Unit 2 |
NUREG-1831 |
Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2 |
NUREG-1832 |
"Analysis of Public Comments on the Revised License Renewal Guidance Documents," contains the NRC staff's analysis
of the stakeholders' comments on the revised license renewal guidance documents 9/30/05) |
NUREG-1833 |
"Technical Bases for Revision to the License Renewal Guidance Documents" contains the NRC staff explanation of
technical changes to the SRP-LR (NUREG-1800) and the GALL Report (NUREG-1801) from Rev.0 to Rev.1 (10/28/05). |
NUREG-1838 |
Safety Evaluation Report Related to the License Renewal of the Millstone Power Station, Units 2 and 3 |
NUREG-1839 |
Safety Evaluation Report Related to the License Renewal of the Point Beach Nuclear Plant, Units 1 and 2 |
NUREG-1843 |
Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3 |
NUREG/CR-6490 (Accession # 9701130206 9701130171) |
"Nuclear Power Plant Generic Aging Lessons Learned (GALL)," contains the review results of the Nuclear Plant Aging
Research Program and related information pertaining to nuclear power plant aging effects and plant impact. |
NUREG/CR-6679 |
"Assessment of Age-Related Degradation of Structures and Passive Components for U.S. Nuclear Power Plants." |
NUREG-1900 |
Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2 |
NUREG-1856 |
Safety Evaluation Report Related to the License Renewal of the Brunswick Steam Electric Plant, Units 1 and 2 |
NUREG-1865 |
Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant |
NUREG-1871 |
Safety Evaluation Report Related to the License Renewal of Palisades Nuclear Plant |
NUREG-1875 |
Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station |
NPAR Reports
Nuclear Plant Aging Research (NPAR) Reports is a collection of literature on mechanical, structural, and
thermal-hydraulic components and systems providing a systematic review of plant aging information in order to assess
materials and component aging issues related to continued operation and license renewal of operating reactors.
Document Identifier |
Title |
BNL Tech Reports:
A-3270-11-26-84 |
Scoping Test on Containment Purge and Vent Seal Material |
BNL Tech Reports:
A-3270-12-86 |
Aging and Life Extension Assessment Program (ALEAP) Systems Level Plan |
BNL Tech Reports:
A-3270R-2-90 |
Interim Report - Aging Effects of Important Balance of Plant Systems in Nuclear Power Plants |
BNL Tech Reports:
TR-3270-6-90 |
Maintenance Team Inspection Results: Insights Related to Plant Aging |
BNL Tech Reports:
TR-3270-9-90 |
An Operational Assessment of the Babcock & Wilcox and Combustion Engineering Control Rod Drives |
BNL Tech Reports:
A-3270-6-21-91 |
Degradation Modeling: Extensions and Applications |
BNL Tech Reports:
EGG-SSRE-8972 |
Estimating Hazard Functions for Repairable Components |
BNL Tech Reports:
EGG-SSRE-9017 |
User's Guide to PHAZE, a Computer Program for Parametric Hazard Function Estimation |
BNL Tech Reports:
EGG-SSRE-9777 |
Isolation Valve Assessment (IVA) Software Version 3.10, User's Manual |
BNL Tech Reports:
EGG-SSRE-9926 |
Evaluation of EPRI Draft Report NP-7065-Review of NRC/INEL Gate Valve Test Program |
BNL Tech Reports:
EGG-SSRE-10039 |
An Evaluation of the Effects of Valve Body Erosion on Motor-Operated Valve Operability |
BNL Tech Reports:
Letter Report |
Summaries of Research Reports Submitted in Connection with the Nuclear Plant Aging Research (NPAR) Program |
NUREG-1144, V3 |
Nuclear Plant Aging Research (NPAR) Program Plan, Status and Accomplishments, Rev. 2 |
NUREG/CP-0100 |
Proceedings of the International Nuclear Power Plant Aging Symposium |
NUREG/CP-0105 Vol. 3 |
Proceedings of the Seventeenth Water Reactor Safety Information Meeting (aging session only) |
NUREG/CR-2641(ORNL/TM-8271) |
The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Collection and Methodology Report |
NUREG/CR-3154(ORNL/TM 8647) |
The In-Plant Reliability Data Base for Nuclear Power Plant Components: Interim Report - The Valve Component |
NUREG/CR-4144
(PNL-5389) |
Importance Ranking Based on Aging Consideration of Components Included in Probabilistic Risk Assessments |
NUREG/CR-4279 V1
(PNL-5479) |
Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear
Power Plants |
NUREG/CR-4302 V1 (ORNL-6193) |
Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants |
NUREG/CR-4302 V2(ORNL-6193) |
Aging and Service Wear of Check Valves Used in Engineered
Safety-Feature Systems of Nuclear Power Plants |
NUREG/CR-4380 (ORNL-6226) |
Evaluation of the Motor-Operated Valve Analysis and Test System (MOVATS) to Detect Degradation, Incorrect Adjustments,
and Other Abnormalities in Motor Operated Valves |
NUREG/CR-4597 V1 (ORNL-6282) |
Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol. 1: Operating Experience
and Failure Identification |
NUREG/CR-4597 V2 (ORNL-6282) |
Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol. 2: Aging Assessment and
Monitoring Method Evaluations |
NUREG/CR-4652
(ORNL/TM-10059) |
Concrete Component Aging and Its Significance
Relative to Extension of Nuclear Power Plants |
NUREG/CR-4692 (ORNL/NOAC-233) |
Operating Experience Review of Failures of Power Operated Relief Valves and Block Valves in Nuclear Power Plants |
NUREG/CR-4731 V1 (EGG-2469) |
Residual Life Assessment of Major Light Water Reactor Components, Vol. 1 |
NUREG/CR-4731 V2 (EGG-2469) |
Residual Life Assessment of Major Light Water Reactor Components - Overview, Vol. 2 |
NUREG/CR-4747 V1 EGG-2473) |
An Aging Failure Survey of Light Water Reactor Safety Systems and Components, V1 |
NUREG/CR-4747 V2 EGG-2473) |
An Aging Failure Survey of Light Water Reactor Safety Systems and Components, V2 |
NUREG/CR-4819 V1 (ORNL/SUB/83-28915/4) |
Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of Nuclear Power Plants |
NUREG/CR-4819 V2 (ORNL/TM-12038) |
Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of Nuclear Power Plants: Evaluation
of Monitoring Methods |
NUREG/CR-4967
(EGG-2514) |
Nuclear Plant Aging Research on High Pressure Injection Systems |
NUREG/CR-4977 V1 (EGG-2505 V1) |
SHAG Test Series: Seismis Research on an Aged United States Gate Valve and on a Piping System in the Decommissioned
Heissdampfreaktor (HDR): Summary, V1 |
NUREG/CR-4977 V2 (EGG-2505 V2) |
SHAG Test Series: Seismis Research on an Aged United States Gate Valve and on a Piping System in the
Decommissioned Heissdampfreaktor (HDR): Appendices, V2 |
NUREG/CR-5057
(PNL-6397) |
Aging Mitigation and Improved Programs for Nuclear Service Diesel Generators |
NUREG/CR-5159
(KEI-1559) |
Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems |
NUREG/CR-5248
(PNL-6701) |
Prioritization of TIRALEX-Recommended Components for Further Aging Research |
NUREG/CR-5314 V3 EGG-2562) |
Life Assessment Procedures for Major LWR Components |
NUREG/CR-5378 (EGG-2567) |
Aging Data Analysis and Risk Assessment - Development and Demonstration Study |
NUREG/CR-5379 V1 (PNL-6560) |
Nuclear Plant Service Water System Aging Degradation Assessment: Phase I, Vol. 1 |
NUREG/CR-5379 V2 (PNL-6560) |
Nuclear Plant Service Water System Aging Degradation Assessment: Phase II, Vol. 2 |
NUREG/CR-5386 (PNL-6911) |
Basis for Snubber Aging Research: Nuclear Plant Aging Research Program |
NUREG/CR-5404 V1
(ORNL-6566, V1) |
Auxiliary Feedwater System Aging Study, V1 |
NUREG/CR-5404 V2
(ORNL-6566, V2) |
Auxiliary Feedwater System Aging Study, Phase I Follow-On Study |
NUREG/CR-5406 V1
(EGG-2569, V1) |
BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualification and High Energy Flow Interruption Test,
V2: Analysis and Conclusions |
NUREG/CR-5406 V2 (EGG-2569, V2) |
BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualification and High Energy Flow
Interruption Test, V2: Data Report |
NUREG/CR-5406 V3
(EGG-2569, V3) |
BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualification and High Energy Flow Interruption
Test, V3: Review of Issues Associated with the BWR Containment Isolation Valve Closure |
NUREG/CR-5419 (BNL-NUREG-5221) |
Aging Assessment of Instrument Air Systems in Nuclear Power Plants |
NUREG/CR-5479 (ORNL/TM/11398) |
Current Applications of Vibration Monitoring and Neutron Noise Analysis: Detection and Analysis of Structural
Degradation of Reactor Vessel Internals from Operational Aging |
NUREG/CR-5490 V1 (PNL-7190) |
Regulatory Instrument Review: Management of Aging of LWR Major Safety Components |
NUREG/CR-5507 (BNL-NUREG-52222) |
Results from the Nuclear Plant Aging Research Program: Their Use in Inspection Activities |
NUREG/CR-5510 |
Evaluations of Core Melt Frequency Effects due to Component Aging Maintenance |
NUREG/CR-5515 (ETEC 88-01) |
Light Water Reactor Pressure Isolation Valve Performance Testing |
NUREG/CR-5519 V1 (ORNL-6607/V1) |
Aging of Control and Service Air Compressors and Dryers Used in Nuclear Power Plants |
NUREG/CR-5555 |
Aging Assessment of the Westinghouse PWR Control Rod Drive System |
NUREG/CR-5558 (EGG-2600) |
Generic Issue 87: Flexible Wedge Gate Valve Test Program: Phase II Results and Analysis |
NUREG/CR-5583 (KEI No. 1656) |
Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems-Wear and Impact Tests |
NUREG/CR-5587 (SAIC-92/1137) |
Approaches for Age-Dependent Probabilistic Safety Assessments With Emphasis on Prioritization and Sensitivity Studies |
NUREG/CR-5646
(EGG-2655) |
Piping System Response During High Level Simulated Seismic Tests at the Heissdampfreaktor Facility (SHAM Test Facility) |
NUREG/CR-5693
(BNL-NUREG-52283) |
Aging Assessment of Component Cooling Water Systems in Pressurized Water Reactors-Phase II |
NUREG/CR-5699 V1
(ORNL-6666/V1) |
Aging and Service Wear of Control Rod Drive Mechanisms for BWR Nuclear Plants - V1 |
NUREG/CR-5706 (ORNL-6671) |
NRC Bulletin 88-04: Potential Safety-Related Pump Loss- An Assessment of Industry Data |
NUREG/CR-5720
(EGG-2643) |
Motor-Operated Valve Research Update |
NUREG/CR-5754
(ORNL/TM-11876) |
Boiling-Water Reactor Internals Aging Degradation Study-Phase I |
NUREG/CR-5779 V1
(ORNL-6687) |
Aging of Non-Power-Cycle Heat Exchangers Used in Nuclear Power Plants |
NUREG/CR-5783
(BNL-NUREG-52299) |
Aging Assessment of the Combustion Engineering and Babcock & Wilcox Control Rod Drives |
NUREG/CR-5807 |
Improvements in Motor Operated Gate Valve Design and Prediction Models for Nuclear Power Plant Systems |
NUREG/CR-5848 |
Recordkeeping Needs to Mitigate the Impact of Aging Degradation |
NUREG/CR-5870
(PNL-8051) |
Results of LWR Snubber Aging Research |
NUREG/CR-5944
(ORNL-6734)V2 |
A Characterization of Check Valve Degradation and Failure Experience in the Nuclear Power Industry |
NUREG/CR-6001
(PNL-8020) |
Aging Assessment of BWR Standby Liquid Control Systems |
NUREG/CR-6029 |
Aging Assessment of Nuclear Air Treatment System HEPA Filters and Adsorbers - Phase 1 |
NUREG/CR-6043 V1 (PNL-8614) |
Phase I Aging Assessment of Essential HVAC Chillers Used in Nuclear Power Plants |
NUREG/CR-6048
(ORNL-TM-12371) |
Pressurized-Water Reactor Internals Aging Degradation Study --A Phase I Report |
ORNL/NRC/LTR-91/25 |
Throttled Valve Cavitation and Erosion |
PNL-5722 |
Operating Experience and Aging Assessment of ECCS Pump Room Coolers |
PNL-6287 |
Study Group Review of Nuclear Service Diesel Generator Testing and Aging Mitigation |
PNL-7516 |
Emergency Diesel Generator Technical Specifications Study Results |
PNL-7823 |
Maintenance Practices to Manage Aging: A Review of Several Technologies |
PNL-SA-20219 |
ASME Subsection ISTD Recommendations Based upon NPAR Snubber Aging Research Results |
|