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accession number: ML010890301
atomic element: Cobalt
atomic element: Iron
atomic element: Nickel
atomic element: Copper
atomic element: Sodium
atomic element: Cadmium
atomic element: Lead
atomic element: Helium
atomic element: Aluminum
atomic element: Titanium
atomic element: Gold
atomic element: Niobium
atomic element: Tin
atomic element: Protactinium
atomic element: Neptunium
CFR: 10 CFR 50
CFR: 10 CFR 50.60
CFR: 10 CFR 50.61
CFR: 10 CFR 50 Appendix G
CFR: 10 CFR 50.6
CFR: 10 CFR 50.109#c
CFR: 10 CFR 50.109#a7
contact: Jones W (301)415-7558
document: ML01089030
from: Office of Nuclear Regulatory Research
general design criterion: GDC 14
index number: 190
issue date: 2001-03-31Z
issue date#aux: 2451999.5
manufacturer: Westinghouse
NUREG: CR-1861
NUREG: CP-0029
NUREG: CR-6454
NUREG: CR-6453
NUREG: NUREG-52395
NUREG: CR-4886
NUREG: CR-6214
NUREG: NUREG-32876
NUREG: CR-5049
NUREG: CR-5338
NUREG: CR-4349
NUREG: CR-6071
NUREG: CR-4576
NUREG: CR-3886
NUREG: CR-3320
NUREG: CR-6115, PWR & BWR Pressure Vessel Fluence Calculation Benchmark Problems & Solutions, Being Made Available to Facilitate Review of Document by Interested Members of Public
NUREG: NUREG-0800, Document for February 22, 2024 ACRS Digital I&C Subcommittee Meeting
page count: 56
procedure: CH-1211
reg guide group number: 1
regulatory guide: Regulatory Guide 1.99
regulatory guide: Regulatory Guide 1.154
SSC: 95Zr-95Nb
SSC: 140Ba-140La
SSC: 137Cs-137mBa
start date: 2001-03-31Z
start date#aux: 2451999.5
state: Texas
state: Illinois
state: Colorado
state: Washington
state: Arkansas
system: Reactor Pressure Vessel
title: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence
topic: Anticipated operational occurrence
topic: Probabilistic Risk Assessment
topic: Backfit
WCAP: WCAP-11104
westinghouse report: WCAP-11104
wiki page modification date: 2025-01-17T06:56:51Z
modification date#aux: 2460692.7894792
wiki page sort key: Regulatory Guide 1.190
URL: http://pbadupws.nrc.gov/docs/ML0108/ML010890301.pdf

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BVY 11-021, NEDO-33618, Revision 0, Vermont Yankee Core Plate Bolt Stress Analysis Report, Attachment 2 to Bvy 11-021
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L-11-085, Regarding the Submittal of the Unapproved Pressure and Temperature Limits Report
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CNL-18-018, Supplement to Application to Revise Watts Bar Unit 2 Technical Specification 4.2.1, Fuel Assemblies, and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage (WBN-TS-17-028)
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RBG-47824, License Amendment Request to Correct a Non-Conservative Technical Specification Figure 3.4.11-1 Minimum Temperature Required Vs. RCS Pressure by Replacing with 54 Effective Full Power Years (EFPY) Curves
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BSEP 18-0041, Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report
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W3F1-2018-0020, Responses to Request for Additional Information Set 18 Regarding the License Renewal Application
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BSEP 18-0048, Response to Request for Additional Information Regarding Inservice Inspection Program Proposed Alternative ISI-09 in Accordance with 10 CFR 50.55a(z)(1) Regarding Reactor Pressure Vessel Circumferential Shell Weld Examinations
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RA-18-0024, Supplemental Information to Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report
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RS-17-126, Quad Cities Nuclear Power Station, Units 1 & 2, Revision 14 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
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LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 5, Reactor Coolant System and Connected Systems
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NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Table 4.3-1 (Sheet 1 of 3) Through Figure 6-2.1-15 (Sheet 14 of 65)
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WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, Chapter 18.0 - Response to NUREG-0737 Clarification of TMI Action Plan Requirements
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NRC-16-0034, Fermi, Unit 2, Revision 20 to Updated Final Safety Analysis Report, Chapter 4, Reactor
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L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 2 of 3
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L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 1 of 3
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RS-14-338, Byron/Braidwood Nuclear Stations, Updated Final Safety Analysis Report (Ufsar), Revision 15, Chapter 5 Reactor Coolant System and Connected Systems
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RS-14-054, LaSalle, Unit 1, Supplemental Information Supporting License Amendment Request to Revise Reactor Coolant System (RCS) Pressure and Temperature (Pit) Curves
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RS-13-266, LaSalle County Station, Unit 1, License Amendment Request to Revise Reactor Coolant System (RCS) Pressure and Temperature Curves
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NL-13-148, Browns Ferry, Unit 1, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature Limits (BFN TS 484)
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NG-17-0111, Duane Arnold Energy Center, Revision 24 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
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NRC-12-0037, Fermi 2, Proposed License Amendment to Relocate Pressure and Temperature Curves to a Pressure Temperature Limits Report
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M190008, GE-Hitachi Nuclear Energy Americas, LLC - Peak Cladding Temperature/10 CFR 50.46 for Advanced Boiling Water Reactor (ABWR) Design Certification Renewal Application, Supplemental Information
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ET 08-0043, Submittal of Supplemental Information on Proposed Changes to Technical Specifications
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NRC 2003-0018, Request for Exemptions to 10 CFR 50.61, Appendices G and H to 10 CFR 50, and Approval of PTS Application for Pbnp Unit 2 and Proposed Heatup and Cooldown Limit Curves for Pbnp Units 1 & 2
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L-2004-244, Proposed License Amendment Extension of the Reactor Coolant System Pressure/Temperature Curve Limits and LTOP to 35 EFPY
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RS-04-133, Request for Changes Related to Technical Specifications Section 3.4.9, Reactor Coolant System Pressure-and Temperature (Pit) Limits
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GO2-04-164, License Amendment Request to Revise the Reactor Pressure Vessel Material Surveillance Program
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PLA-5811, Proposed Amendment No. 235 to License NFP-22: Changes to Reactor Pressure Vessel Pressure-temperature (P-T) Limits PLA-5811
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RNP-RA/04-0098, Request for Technical Specifications Change to Reactor Coolant System Pressure and Temperature Limits
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NL-04-1123, Request to Revise Reg Guide 1.190 Implementation Date
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GO2-04-107, License Amendment Request to Revise Technical Specification 3.4.11, Reactor Coolant System Pressure and Temperature Limits
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NLS2004043, Supplement to License Amendment Request to Revise Technical Specification 3.4.9 Pressure Temperature (P/T) Curves, Figures 3.4.9-1, 3.4.9-2, & 3.4.9-3
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LR-N04-0136, Request for Change to Technical Specifications Pressure/Temperature Limits
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NLS2004042, Confirmation of Commitments Discussed in 03/05/04 Teleconference
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GO2-04-032, Schedule for Requesting Revision of Technical Specification P/T Curves and Adoption of the BWRVIP Integrated Surveillance Program
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NL-03-2177, Request to Revise Technical Specifications and Pressure and Temperature Limits Report
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RS-04-032, Extension of Relief for Alternative Reactor Pressure Vessel Circumferential Weld Examinations for Additional License Operating Period
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NG-04-0103, Request to Implement NDE-R047 Addressing Boiling Water Reactor Shell Weld Inspection Recommendations of the Boiling Water Reactor Vessel & Internals Project (BWRVIP) Report BWRVIP-05
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NLS2004005, License Amendment Request to Revise Technical Specification 3.4.9 Pressure Temperature (P/T) Curves Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3
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NLS2004004, Revision to Commitment Referenced to Amendment 201 Safety Evaluation
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NRC-02-0062, Proposed License Amendment for Participation in the BWRVIP ISP Plan for RPV Material Surveillance
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DCL-02-079, Units, 1 and 2, License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report
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L-20-274, Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits
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