ML15160A221
| ML15160A221 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/09/2015 |
| From: | State of NY, Office of the Attorney General |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 27910, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR | |
| Download: ML15160A221 (8) | |
Text
NYS000521 Submitted: June 9, 2015 NUREG-0800 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN BRANCH TECHNICAL POSITION 5-3 FRACTURE TOUGHNESS REQUIREMENTS REVIEW RESPONSIBILITIES Primary - Organization responsible for the review of component integrity issues related to reactor vessels Secondary - None A. BACKGROUND NRC requirements regarding fracture toughness, pressure-temperature limits, material surveillance, and pressurized thermal shock (PTS) (PWR only) are contained in Appendices A, G, and H to 10 CFR Part 50 and in 10 CFR 50.61; these requirements also refer to relevant sections of the ASME Code. The purpose of this branch technical position is to summarize these requirements and provide guidance, as necessary.
Since many of these requirements were not in force when some plants were designed and built, this position also provides guidance for applying these requirements to older plants. Also included is a description of acceptable procedures for making the conservative estimates and assumptions for older plants that may be used to show compliance with the new requirements.
Revision 2 - March 2007 USNRC STANDARD REVIEW PLAN This Standard Review Plan, NUREG-0800, has been prepared to establish criteria that the U.S. Nuclear Regulatory Commission staff responsible for the review of applications to construct and operate nuclear power plants intends to use in evaluating whether an applicant/licensee meets the NRC's regulations. The Standard Review Plan is not a substitute for the NRC's regulations, and compliance with it is not required. However, an applicant is required to identify differences between the design features, analytical techniques, and procedural measures proposed for its facility and the SRP acceptance criteria and evaluate how the proposed alternatives to the SRP acceptance criteria provide an acceptable method of complying with the NRC regulations.
The standard review plan sections are numbered in accordance with corresponding sections in Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)." Not all sections of Regulatory Guide 1.70 have a corresponding review plan section. The SRP sections applicable to a combined license application for a new light-water reactor (LWR) are based on Regulatory Guide 1.206, "Combined License Applications for Nuclear Power Plants (LWR Edition)."
These documents are made available to the public as part of the NRC's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Individual sections of NUREG-0800 will be revised periodically, as appropriate, to accommodate comments and to reflect new information and experience. Comments may be submitted electronically by email to NRR_SRP@nrc.gov.
Requests for single copies of SRP sections (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to DISTRIBUTION@nrc.gov. Electronic copies of this section are available through the NRC's public Web site at http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/, or in the NRC's Agencywide Documents Access and Management System (ADAMS), at http://www.nrc.gov/reading-rm/adams.html, under Accession # ML070850035.
B. BRANCH TECHNICAL POSITION
- 1. Preservice Fracture Toughness Test Requirements.
The fracture toughness of all ferritic materials used for pressure-retaining components of the reactor coolant pressure boundary shall be evaluated in accordance with the requirements of Appendix G, 10 CFR Part 50, as augmented by the criteria of Section III of the ASME Code. The fracture toughness test requirements for plants with construction permits prior to August 15, 1973 may not comply with the new Codes and Regulations in all respects. The fracture toughness of the materials for these plants should be assessed by using the available test data to estimate the fracture toughness in the same terms as the new requirements. This should be done because the operating limitations imposed on old plants should provide the same safety margins as are required for new plants.
1.1 Determination of RTNDT for Vessel Materials Temperature limitations are determined in relation to a characteristic temperature of the material, RTNDT, that is established from the results of fracture toughness tests. Both drop weight nil-ductility transition temperature (NDTT) tests and Charpy V-notch tests should be run to determine the RTNDT. The NDTT temperature, as determined by drop weight tests (ASTM E-208-1969) is the RTNDT if, at 33 EC (60 EF) above the NDTT, at least 68 J (50 ft-lbs) of energy and 0.89 mm (35 mils) lateral expansion (LE) are obtained in Charpy V-notch tests on specimens oriented in the weak direction (transverse to the direction of maximum working).
In most cases, the fracture toughness testing performed on vessel material for older plants did not include all tests necessary to determine the RTNDT in this manner. Acceptable estimation methods for the most common cases, based on correlations of data from a large number of heats of vessel material, are provided below for guidance in determining RTNDT when measured values are not available.
(1) If dropweight tests were not performed, but full Charpy V-notch curves were obtained, the NDTT for SA-533 Grade B, Class 1 plate and weld material may be assumed to be the temperature at which 41 J (30 ft-lbs) was obtained in Charpy V-notch tests, or -18 EC (0 EF), whichever was higher.
(2) If dropweight tests were not performed on SA-508, Class II forgings, the NDTT may be estimated as the lowest of the following temperatures:
(a) 33 EC (60 EF).
(b) The temperatures of the Charpy V-notch upper shelf.
(c) The temperature at which 136 J (100 ft-lbs) was obtained on Charpy V-notch tests if the upper-shelf energy values were above 136 J (100 ft-lbs).
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(3) If transversely-oriented Charpy V-notch specimens were not tested, the temperature at which 68 J (50 ft-lbs) and 0.89 mm (35 mils) LE would have been obtained on transverse specimens may be estimated by one of the following criteria:
(a) Test results from longitudinally-oriented specimens reduced to 65% of their value to provide conservative estimates of values expected from transversely oriented specimens.
(b) Temperatures at which 68 J (50 ft-lbs) and 0.89 mm (35 mils) LE were obtained on longitudinally-oriented specimens increased 11 EC (20 EF) to provide a conservative estimate of the temperature that would have been necessary to obtain the same values on transversely-oriented specimens.
(4) If limited Charpy V-notch tests were performed at a single temperature to confirm that at least 41 J (30 ft-lbs) was obtained, that temperature may be used as an estimate of the RTNDT provided that at least 61 J (45 ft-lbs) was obtained if the specimens were longitudinally oriented. If the minimum value obtained was less than 61 J (45 ft-lbs), the RTNDT may be estimated as 11 EC (20 EF) above the test temperature.
1.2 Estimation of Charpy V-Notch Upper Shelf Energies For the beltline region of reactor vessels, the upper shelf toughness must account for the effects of neutron radiation. Reactor vessel beltline materials must have Charpy upper shelf energy, in the transverse direction for base material and along the weld for weld material according to the ASME Code, of no less than 102 J (75 ft-lbs) initially and must maintain Charpy upper shelf energy throughout the life of the vessel of no less than 68 J (50 ft-lbs).
If Charpy upper shelf energy values were not obtained, conservative estimates should be made using results of tests on specimens from the first surveillance capsule removed.
If tests were only made on longitudinal specimens, the values should be reduced to 65% of the longitudinal values to estimate the transverse properties.
The predicted end-of-life Charpy upper shelf energy and adjusted reference temperature for the reactor vessel materials must meet the requirements of 10 CFR 50, Appendix G, paragraph IV.B. Reactor vessel materials that do not meet the specified end-of-life acceptance criteria are reviewed in accordance with paragraphs V.C and V.D of 10 CFR 50, Appendix G. NUREG-0744 provides an acceptable methodology for performance of fracture analysis for demonstrating adequate margins of safety for continued operation in accordance with 10 CFR Part 50, Appendix G, paragraph V.C.3.
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1.3 Reporting Requirements Fracture toughness information identified by the Code and by Appendix G, 10 CFR Part 50, should be reported in the FSAR to provide a basis for evaluating the adequacy of the operating limitations given in the Technical Specifications. In the case of older plants, the data may be estimated, using the procedures listed above, or other methods that can be shown to be conservative.
- 2. Operating Limitations for Fracture Toughness 2.1 Pressure-Temperature Operating Limitations As required by Appendix G, 10 CFR Part 50, the following operating limitations shall be determined and included in the Technical Specifications. The basis for determination shall be reported, and is the responsibility of the applicant, but in no case shall the limitations provide less safety margin than those determined in accordance with Appendix G, 10 CFR Part 50, and Appendix G to Section III of the Code.
(1) Minimum temperatures for performing any hydrostatic test involving pressurization of the reactor vessel after installation in the system.
(2) Minimum temperatures for all leak and hydrostatic tests performed after the plant is in service.
(3) Maximum pressure-minimum temperature curves for operation, including startup, upset, and cooldown conditions.
(4) Maximum pressure-minimum temperature curves for core operation.
2.2 Recommended Bases for Operating Limitations 2.2.1 Leak and Hydrostatic Tests (1) It is recommended that no tests at pressures higher than design pressure be conducted with fuel in the vessel.
(2) For system and component hydrostatic tests performed prior to loading fuel in the reactor vessel, it is recommended that hydrostatic tests be performed at a temperature not lower than RTNDT plus 60 EF.
(3) For system and component hydrostatic tests performed subsequent to loading fuel in the reactor vessel, the minimum test temperature should be determined as discussed in Section III of SRP 5.3.2.
2.2.2 Heatup and Cooldown Limit Curves Heatup and cooldown pressure-temperature limit curves may be determined using single pr/t stress calculations, using the method given in Appendix G of the Code. The effect of thermal gradients may be conservatively approximated by BTP 5-3-4 Revision 2 - March 2007
the procedures in Appendix G of the Code or from the report, Tabulation of thermally-Induced Stress Intensity Factors (KIT) and Crack Tip Temperatures for Generating P-T Curves per ASME Section XI-Appendix G, ORNL/NRC/LTR-03/03.
Calculations need only be performed for the beltline region, if the RTNDT of the beltline is demonstrated to be adequately higher than the RTNDT for all higher stressed regions.
Alternatively, more rigorous analytical procedures may be used, provided that the intent of the Code is met, and adequate technical justification for all assumptions and bases is provided.
2.2.3 Core Operation Limits To provide added margins during actual core operation, Appendix G, 10 CFR Part 50 requires a minimum temperature during core operation, and a 22 EC (40 EF) margin in temperature over the pressure-temperature limits as determined for heatup and cooldown in 2.2.2 above. The minimum temperature, regardless of pressure, is the temperature calculated for the inservice hydrostatic test according to 2.2.1 above.
2.2.4 Upset Conditions The pressure-temperature limits described in 2.2.2 and 2.2.3 above are applicable to upset conditions. Normal operating procedures should permit variations from intended operation, including all upset conditions, without exceeding the limit curves.
2.2.5 Emergency and Faulted Conditions It is recognized that the severity of a transient resulting from an emergency or faulted condition is not directly related to operating conditions, and resulting temperature-stress relationships in the reactor coolant boundary components are primarily system dependent, and therefore not under direct control of the operator.
For these reasons, operating limits for emergency and faulted conditions are not a requirement of the Technical Specifications.
The SAR should present descriptions of the continued integrity of all vital components of the RCPB during postulated faulted conditions. It is recommended that such descriptions be made in as realistic a manner as possible, avoiding grossly over conservative assumptions and procedures.
2.3 Reporting Requirements The Technical Specifications should include the operating and test limits discussed above, and the basis for their determination. The Technical Specifications should also include information on the intended operating BTP 5-3-5 Revision 2 - March 2007
procedures, and justify that adequate margins between the expected conditions and the limit conditions will be provided to protect against unexpected or upset conditions.
- 3. Inservice Surveillance of Fracture Toughness The reactor vessel may be exposed to significant neutron radiation during the service life. This will affect both the tensile and toughness properties. A material surveillance program in conformance with Appendix H, 10 CFR Part 50, must be carried out.
3.1 Surveillance Program Requirements The minimum requirements for the surveillance program are covered by Appendix H, 10 CFR Part 50. The selection of material to be included in the surveillance program should be in accordance with ASTM E-185-82, unless the intent of the program is better realized by using more rigorous criteria. For example, the approach of estimating the actual RTNDT and upper shelf toughness of each plate, forging, or weld in the beltline as a function of service life, and choosing as the surveillance materials those that are expected to be most limiting, may be preferable in some cases. This would include consideration of the initial RTNDT, the upper shelf toughness, the expected radiation sensitivity of the material (based on copper and nickel content, for example) and the neutron fluence expected at its location in the vessel.
3.2 SAR Criteria With respect to the adequacy of the surveillance program, information requested for beltline materials includes the following:
(1) Tensile properties.
(2) Dropweight test and Charpy V test results used to determine RTNDT.
(3) Charpy V test results to determine the upper shelf toughness.
(4) Composition, specifically the copper and nickel content.
(5) Estimated maximum fluence for each beltline material.
(6) List of materials included in the surveillance program, with basis used for their selection.
3.3 Surveillance Test Procedures Surveillance capsules must be removed and tested at intervals in accordance with Appendix H, 10 CFR Part 50. The proposed removal and test schedule should be included in the Technical Specifications.
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3.4 Reporting Criteria All information used to evaluate results of the tests on surveillance materials, evaluation methods, and results of the evaluation should be submitted with the evaluation report. This should include:
(1) Original properties and compositions of the materials.
(2) Fluence calculations, including original predictions, for both surveillance specimens and vessel wall.
(3) Test results on surveillance specimens.
(4) Basis for evaluation of changes in RTNDT and upper shelf toughness.
(5) Updated prediction of vessel properties.
3.5 Technical Specification Changes Changes in the operating and test limits recommended as a result of evaluating the properties of the surveillance material, together with the basis for these changes, shall be submitted to the Office of Nuclear Reactor Regulation for approval.
- 4. Pressurized Thermal Shock (PWR only) 4.1 Pressurized Thermal Shock Requirements As required by 10 CFR 50.61, the following is a summary of requirements for the PWR reactor vessels:
(1) RTPTS values must be projected using end-of-life fluence for each weld, plate or forging in the reactor vessel beltline region. The projected EOL RTPTS values must be approved by the NRC.
(2) PTS screening criteria is 132 EC (270 EF) for plates, forgings, and axial weld materials, and 149 EC (300 EF) for circumferential weld materials.
(3) If reactor vessel is projected to exceed the PTS screening criteria, 10 CFR 50.61(b)(3) requires the applicant to implement a flux reduction program that is reasonably practicable to avoid exceeding the PTS screening criteria.
(4) If the flux reduction program does not prevent the reactor vessel from exceeding the PTS screening criterion at the end of life, the applicant choose between the two options in 10 CFR 50.61 to meet PTS requirements: (a) submit a safety analysis pursuant to 10 CFR 50.61(b)(4) to determine what, if, any, modifications to equipment, systems, and plant operation to prevent failure of the reactor vessel from a postulated PTS event, (b) perform a thermal-annealing BTP 5-3-7 Revision 2 - March 2007
treatment of the reactor vessel pursuant 10 CFR 50.61(b)(7) to recover fracture toughness. 10 CFR 50.61 requires details of the approach selected to be submitted for NRC approval at least 3 years before the reactor vessel is projected to exceed the PTS screening criteria.
C. REFERENCES
- 1. 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants.
- 2. 10 CFR Part 50, Appendix G, Fracture Toughness Requirements.
- 3. 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
- 4. 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
- 5. NUREG-0744, Pressure Vessel Material Fracture Toughness.
- 6. ASME Boiler and Pressure Vessel Code,Section III, including Appendix G, Protection Against Nonductile Failure, American Society of Mechanical Engineers.
- 7. ASTM E-185-82, Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels.
- 8. ORNL/NRC/LTR-03/03, Tabulation of Thermally-Induced Stress Intensity Factors (KIT) and Crack Tip Temperatures for Generating P-T Curves per ASME Section XI -
Appendix G.
PAPERWORK REDUCTION ACT STATEMENT The information collections contained in the Standard Review Plan are covered by the requirements of 10 CFR Part 50 and 10 CFR Part 52, and were approved by the Office of Management and Budget, approval number 3150-0011 and 3150-0151.
PUBLIC PROTECTION NOTIFICATION The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.
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