ML19221B091
| ML19221B091 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-05.3.3, NUREG-75-87, NUREG-75-87-5.3.3, SRP-05.03.03, SRP-5.03.03, NUDOCS 7907120407 | |
| Download: ML19221B091 (6) | |
Text
NU R EG-75/087 n REcg fa o',e U.S. NUCLEAR REGULATORY COMMISSION kc' }
STANDARD REV EW PLAN OFFICE OF NUCLEAR REACTOR REGULATION LCTION 5.3.3 REACTOR VESSEL INTEGRITY REVIEW RESPONSIBILITIES Primary - Materials Engineering Branch (MTEB)
Secondary - None I.
MEAS OF REVIEW The portions of the applicant's safety analysis report (SAR) listed below are reviewed. These portions are all related to the integrity of the reactor vessel. Although most of these areas are reviewed separately in accordance with other review plans, the integrity of the reactor vessel is of such impo-tante that a special sumary review of all factors relating to the integrity of the reactor vessel is warranted. The information in cach area is reviewed to ensure that the information is complete, and that no inconsistencies in information or requirements exist that would reduce the certainty of vessel integrity, l.
Design The basic design of the re tctor vessel reviewed.
2.
M3terials of Construction The materials of construction are each taken into consideration.
3.
Fabrication Methods The processes used to fabricate the reactor vessel, including forming, welding, cladding, and machining, are reviewed.
4.
Inspection Requirements The inspection test methods and requirements are reviewed.
5.
Shipment and Ins tall a tig Protective measures taken during shipment of the reactor vessel and its installation at the site are reviewed.
6.
Operating Conditions All the operating conditions as they relate to the integrity of the reactor vessel are reviewed.
USNRC STANDARD REVIEW PLAN Stendere rev*ew piene are prepared for the guidance et the Otfice of Nucteer Reactor Reguietson staff responsible f or the review of applications to construct and openete nucteer power plante These documents are modo aveilable to the public as part of the Commission e po66cy to 6nform the nucieer industry and the ponered public of reganotory procedures and policies Standard rev+ew p6ene are not oubetetutes f or requietory guides or the Commasoson a reguistsono and comptience weth them se not requered The etendeed review paen sectione are hoved to Revision 2 of the S+endard f ormet and Content of Setevy Aretyees Reporte for muc6eer Power Ptento Not oss sections of the Stenderd Foeme+ have a corresponding review plan Pubreeked stendard review pte e will be rev*eed periodically. se oppropriete to accommodete commente end to reffect new enformation end empen c Commente end suggesteene for 6mprovement wdl be concedered and should be sent to the U S Nucleer Regulatory Commeseson. Office of Nucleer Reactor Regotetson. Weekengton D C 2Nd5 7 90712 0 4v7
7.
Inservice Surveillance Plans and provisions for inservice surveillance of the reactor vessel are reviewed.
II.
ACCEPTANCE CRITERI A The basic acceptance criteria for each review area are covered by other standard review plans, so they will be discussed here only in general terms. References are made to the review plans that include detailed criteria. Interrelationships among review areas, and criteria for consistency, compatability, and technical coherence among resiew areas, are emphasized in the following discussion.
1.
Desijn The basic acceptance criteria for the design of the vessel are detailed in the standard review plans for SAR Sections 3 and 5.2.
These cover the requirements of the General Design Criteria and Section III of the ASME Boiler and Pressure Vessel Code (hereaf ter
" the Code"). The design of the reactor vessel nust be compatible with the preparties of the naterials used, and must permit construction by the use of standard and well proven fabrication rtthods. The design details should not include new or novel concepts unless they are substantiated by a comprehensive justification showing that no aspects of the design will compromise the overall integrity of the vessel in any ra nn e r.
The design details must be adequate to permit all required inspections and to provide required access to all areas requiring inservice inspection in conformance with Section XI of the Code, as detailed in Standard Review Plan 5.2.4, " Reactor Coolant Pressure Boundary Inservice Inspection and Testing."
If the neutron radiation exposure of the reactor vessel becomes high enough that the predicted value of the adjusted reference temperature of the naterial exceeds 200*F, the design must be adequat; to permit in-place annealing of the vessel to restore ductility and toughness, in accordance with Appendix G, 10 CFR Part 50, and as detailed in the review plan for SAR Section 5.3.1, " Reactor Vessel Materials.
2.
Materials of Construction The basic acceptance criteria for the materials used in the construction of the reactor vessel are the reeirenents of Section III of the Code, as augmented by Appendix G, 10 CFR Part 50, "Fractse Toughness Requirements." These criteria are detailed in Standard Review Dlans 5.2.s, " Reactor Coolant Pressure Boundary Materials," and 5.3.1,
" Reactor Vessel Materials."
The materials must be compatible with the design requirenents. Acceptability is based on standard practice and engineeririg judgement, with consideration being given to such factors as material fom, size-related variations in properties, and nonisotropic characteristics.
Although many materials are acceptable for reactor vessels according to Section III of the Code, the special considerations relating to fracture toughness and radiation 5.3.3-2
effects effectively limit the basic nuterials that are currently acceptable for cost parts of reactor vessels to St 533 Gr B Cl 1, SA 508 C12,and SA 503 C1 3.
Accept-ability criteria for other grades will have to be developed before they can be used.
The relationships among material compositions, expected neutron fluence, and require-ments for the material surveillance program must be compatible. The reviewer uses publishec data to ensure that the predicted shif t in toughness properties (RT NDT upper shelf eNrgy) is conservative, based on actual material composition and predicted fl uence. Acceptability of the caterial surveillance progran, as spccified in Appendix H,10 CFR Pert 50, depends on these relationships.
3.
Fat;ricati_on Athods Acceptability criteria for the bi. sic fabrication processes and their qualification and control requirements are given in Sections III and IX of the Code, cnd detailed in Standard Review Plan 5.3.1," Reactor Vessel Materials."
Although a particuiar fabrication process (such as multiple wire-high heat input welding) ruy be generally acceptable, it may not be suitable for reactor vessel fabrication for sone nucerials without further justification or qualification. The reviewer uses " state-of-the-art" criteria and past practice to evaluate the acceptability of raterials-process combinations.
Because fabrication methods, materials, and the ef fectiveness of non-aestructive evaluation metbds are interrelated, the reviewer mJst rely on state-of-the-art knowledge and past practice to determine whether the proposed combinations are compatible and acceptable.
4.
Inspection Requirements The basic requirenents for perfonaing nondestructive inspections and the quality assurance criteria for the reactor vessel are contained in Sections III and V of the Code. These are detailed in Standard Review Plan 5.3.1, " Reactor Vessel Ma teri al s. "
Acceptance cri t 2 for compatibility with raterials and fabrication areas are discussed in orevious sections.
Very important relationships are thase among in-process and final shop inspections, and the inservice inspection requirenents of Section XI of the Code. The reviewer must determine that the nethods of inspection, the sensit vity levels, and flaw evaluation criteria are compatible with Section XI, and that the results of the presr vice baseline inspection can be correlated with the results of later inservice inspections.
5.
Shipnent and Installation The basic acceptance criteria for procedures and cara used in shipping, storage, and installation of the vessel are given in Regulatory Guides 1.37, " Quality Assurance iU i25 5.3.3-3
Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Reactor Plants," 1.33, " Quality Assurance Requirements for Packaging, Shioping, Receiving, Storage, and Handling of Items for Water-Cooled T4uclear Power Plets,' and 1.39, " Housekeeping Requirements for Water-Cooled T4uclear Power Plants."
The purpose of this area of review 1s to verify that the as-built characteristics of the reactor vessel are not degraded by improper handling. Acceptability in these areas is assured for current designs and materials by comoliance with the basic acceptance criteria. If nonstandard materials or designs are used, the reviewer nust determine that these criteria will be adequate, based on current technology.
If the basic criteria are nct followed, either intentionally or through error, the reviewer must evaluate, on a case basis, whether the integrity of the reactor vessel is compromised, using current technology, past practice, and experience as applicable.
6.
Operating Conditions Acceptance criteria for operating limits for the reactor vessel are given in Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements," and are detailed in Standard Review Plan 5.3.2, " Pressure-Temperature Limits." In addition, Regulatory Guide 1.33,
" Quality Assurance Program Requirements (Operation)" provides acceptable criteria for other phases of operational procedures.
Abnormal operational occurrences must not result in loss of reactor vessel integrity.
The most severe postulated transient is the thermal shock to the vessel caused by emergency ccre cooling system operation af ter a loss-of-coolant accident. The criterion for acceptable behavior is that the vessel must remain leaktight enough to suppcrt adequate core cooling. The generally accepted principles and procedures of linear elastic f racture mechanics provide the basis for acceptance of analyses that support conforrance with this criterion.
7.
Inservice Surveillance The acceptance criteria for adequacy of the reactor vessel raterials surveillance program are based on the requirem nts of Appendix H to 10 CFR Part 50, " Reactor Vessel Material Surveillance Program Requirements," and detailed in Standard Review Plan 5.3.1, " Reactor Vessel Materials."
The SAR also provides inforration regarding the inservice inspections to be performed on the reactor vessel. The acceptance criteria for accessibility and inspection plan details are those of Section XI of the Code, and are detailed in Standard Review Plan S.2.4, " Reactor Coolant Pressure Boundary Inservice Inspection and Testing."
III. _ REVIEW PROCEDURES The reviewer will select and emphasize material from the procedures described below, as may be appropriate for a particular case. The reviewer initially determines that the basic criteria are net in each review area covered by this plan. Although he will not normally
\\
5.3.3-4
be responsible for the basic reviews of all of these areas, he will consuit wiw um responsible for basic review of the other areas to determine that all areas are individually acceptable.
He then reviews each area again, considering the information presented in other areas that interrelate with it, as discussed in 11 above.
Because the reviewer is familiar with the specific procedures used by the reactor vendor, he can readily pick out any differences from past practice. He will evaluate these in detail, consulting with other MTEB members as appropriate.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information is provided to satisfy the requirements of this review plan, and that the conpleteness and technical adequacy of his evaluation will support conclusions of the following type, to be included in the staf f's safety evaluation report:
"We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude there are no special considerations that make it necessary to consider potential reactor vessel tailure for this plant. The bases for our conclusion are that the design, materials, fabrication, inspection, and quality assurance require-nents for the plant will conform to applicable AEC Regulations and Regulatory Guides, and tG the rules of the ASME Boiler and Pressure Vessel Code,Section III. The stringent fracture toughness requirements of the Regulations and ASME Code Section III will be met, including requirenents for surveillance of vessel material properties throughout service life. Also, operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G, " Protection Against Non-Ductile failure," of ASME Code Section III, and Appendix G,10 CFR Part 50.
"The integrity of the reactor vessel is assured because the vessel (1) will be designed and fabricated to the hip standards of quality required by the ASME Gailer and Pressure Vessel Code and any pertinent Code Cases; (2) will be made from materials of controlled and demonstrated high quality; (3) will be subjected to extensive preservice inspection and testing to provide assurance that the vessel will not fail because of naterial or fabrication deficiencies; (4) will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions will not be exceeded during nornal reactor operation, and that the vessel will not fail under the conditions of any of the postulated accidents; (5) will be subjected to periodic inspection to demonstrate that the high initial quality of the reactor vessel has not deteriorated significantly under service conditions; and (6) nay be annealed to restore the material toughness properties if this becones necessary."
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\\h I
5.3.3-5
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants."
9 2.
10 CFR Pact 50, Appendix G, " Fracture Toughness Requirements."
3.
10 CFR Part 50, Appendix H. " Reactor Vessel Material Surveillance Program Requi remen ts. "
4.
ASME Boiler and Pressure Vessel Code,Section III, especially Appendix G, " Protection Against Nonductile Failure," Arrerican Society of Mechanical Engineers.
5.
ASME Boiler and Pressure Vessel Code, Sections II, V, IX, and XI, American Society of Mechanical Engineers.
6.
Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)."
7.
Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Corponents of Water-Cooled Nuclear Power Plants."
8.
Regulatory Guide 1.33, " Quality Assurance Requirer ents in Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Plants."
9.
Regulatory Guide 1.39, " Housekeeping P.equirements for Water-Cooled Nuclear Power Plants."
10.
Standard Review Plan 5.2.3, "PCPB Materials."
11.
Standard Review Plan 5.2.4, "RCPB Inservice Inspection and Testing."
- 12. Standard Review Plan 5.3.1, " Reactor Vessel Materials."
13.
Standard Review Plan 5.3.2, " Pressure-Temperature Limits."
147 128 5.3.3-6