ML19221B049
| ML19221B049 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-03.9.5, NUREG-75-87, NUREG-75-87-3.9.5, SRP-03.09.05, SRP-3.09.05, NUDOCS 7907120312 | |
| Download: ML19221B049 (4) | |
Text
NUREG 75/087 fj n,g%
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U.S. NUCLEAR REGULATORY COMMISSIGN
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STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 3.9.5 REACTOR PRESSURE VESSEL INTERNALS REVIEW RESPONSIBILITIES Pr-imary - Mechanical Engineering Branch (MEB)
Secondary - None I.
AREAS Of REVIEW for the purpose of this SRP section, the term " reactor internals" refers to all struc-tural and mechanical elements inside the reactor pressure vessel with the exception of the following:
fuel assemblies, control rod drive mechanisms (defined and reviewed in SRP Section 3.9.4), and reactivity control elements out to the CRDM coupling interf ace.
The CRDM quide tubes are considered to be part of the reactor internals.
In-core instrumentation (in-core iaitrumontation support structures are considered part of the reactor internals).
The Materials Engineering Branch reviews the reactar internals mater ials in SRP 5ection 4.5.2.
The Core Performance Branch reviews the fuel assembly design in SRP Section 4.2.
The staff reviews the following specific areas to assure conformance with the equire-ments of General Design Criteria 1, 2, 4 and lu:
1.
The physical or design arrangements of all reactor internal structures, components, assemblies and systems. This should include the manner of positioning and securing such items within the reactor pressure vessel, the manner of providing for axial and lateral retention and support of the internals assemblies and components, and the manner of accontmodating dimes ional changes due to thermal and other ef fects.
2.
The plant and system operating conditions and design basis events which yovide the basis for the design of the reactor internals.
3.
The specific design and servic. loading combinations and the appropriate i.athod of combination of these loads. The design and service stress limits associated with these various loading combinations are reviewed, including maximum allowable
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stress, deflection limits, and cycling and fatigue limits, as specified in Table I of Branch Technica' Position MEB 3-2 of SRP Section 3.9.3.
The extent of compli-ance with subsection NG of the ASME Code is reviewed. Details of dynamic analysis, input forcing functions, and response loadings are discussed in SRP Section 3.9.2.
II.
ACCEPTANCE CRITERIA Compliance with the following criteria ccnstitutes an acceptable basis for satisfying the applicable port! ns of General Design Criteria 1, 2, 4 and 10.
A discussion of the acceptance criteria for the design and service loading combinations and stre.s limits applicable to reactor interaals is presented in SRP Section 3.9.3 (Ref. 7).
The design and construction of the core support structures should conform to the requirements of subsection NG, " Core Support Structures," of Section III of the ASME Code (Ref. 5).
The design and service loadings, stress limits, and analyses that provide the basis for the design of reactor internals other than the core support structures should meet the guidelines of NG-3000 and be constructed su as not to adversely affect the integrity of the core support structures (NG-ll22).
Deformation limits for reactor internals shculd be established by the applicant and presented in his Safety Analysis Report. The basis for these limits shculd be included.
The stresses associated with these displacements should not exceed the specified design limits. The requirements for dynamic analysis of these components are dist.assed in SRP Section 3.9.2.
l III. REVIEW PROCEDURES The reviewer will select and emphasize material from the procedures described below as may be appropriate for a particular case.
The configuration and general arrangement of all mechanical and structural internal elements covered by this SRP section are reviewed and compared to those of previously l
licensed similar plants. Any significant changes in design are noted and the applicant is asked to verify that these char.ges do not affect the flow-indutad vibration test results required by SRP Section 3 9.2.
l With respect to the design and analysis of these comp >nents, a statement by the v pli-cant that they are designed in accordance with subsection NG, " Core Support Structures,"
of the ASME Code,Section III, and meet the criteria for design and service loads and limits in SRP Section 3.9.3, is acceptable. In lieu of such a commitment, the reviewer must determine that the design and analysis of these components are consistent with the requirements discussed in II, above. This is accomplished by requiring that the appli-cant describe the design procedures and criteria used in the design of these components.
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3.9.5-2 146 318
This includes a list of the design and service stress limits used for all of the applicable loading conditions.
The deformation limits specified for these components are reviewed to verify that tne appl' cant has stated that these deflections will not interfere with the functioning of related components, e.g.,
control rods and standby cooling systems, and that the stresses associated with these disnlacements are less than the specified limits for the core support structures.
At the operating license stage, the calculated str?sses and deformations are reviewed to determine that they do not exceed the specified limits.
l Any deviations that have not been adequately justified are identified and findings t0 that effect are trar3mitted to the applicant with a request for conformance with the requirements discussed in subsection II, or for additione.1 technical justification.
l IV.
EVALUATION FINDINGS The reviewer verifies that suf ficient information has been provided in accc" dance wit' this SRP section, and that his evaluation supports conclusions of the follow:nq type, to be included in the staf f's %fety Evaluation Report:
"The design procedures and criteria that the applicant has used for the reactor internals are in conformance with established technical procedures, positions, standards, and criteria which are acceptable to the staff.
"The specified transients, design and service loadings, and combination of loadings l
as applied to the design of the reactor internals structures and components provide reasonable assurance that in the event of an earthquake or of a system transient l
during normal plant operation, the resulting deections and associated stresses imposed on these structures and components would not exceed allowable stresses and deformation limits for the materials of construction. Limiting the stresses and deformations under such loading combinations provides an acceptable basis for the design of these structures and csmponents to wishstand the most adverse loading events which have been postulated to occur M ring service lifetime without loss of structural integrity or impairment of function. In addition, the design procedures and criteria used by the applicant in the design of the reactor internals consti-tutes an acceptable basis for satisfying the coplicable requirements of General Design Criteria 1, 2, 4 and ;0."
V.
REFERENCES 1.
10 rFR Part 50, Appendix A, General Design Criterion 1, " Quality Standards and l
Records."
2.
10 CFR Part 50, Appendix A, General Design Criterion 2 " Design Basis for Protec-l tion Against Natural Phenomena."
146 319 3.9.5-3 Rev. 1
3.
10 CFR Part 50, Appendix A, Gen (ral Design Criterion 4, " Environmental and Missi!e l
Design Cases."
4.
10 CFR Part 50, Appendix A, General Design Criterion 10, " Reactor Design."
l S.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, " Nuclear Power Plant Components," American Society of Mechanical Engineers.
6.
SRP Section 3.9.2, " Dynamic Testing and Analysis of Systems, Components, and l
Equipment."
7.
SRP Section 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures."
O Rev. 1 3.9.5-4 7'O
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