ML19221B032
| ML19221B032 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-03.8.3, NUREG-75-87, NUREG-75-87-3.8.3, SRP-03.08.03, SRP-3.08.03, NUDOCS 7907120270 | |
| Download: ML19221B032 (27) | |
Text
NUREG-75/037
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STANDARD REVIEW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 3.8.3 CONCRETE AND STEEL INTERNAL STRUCTURES OF STEEL OR CONCRETE CONTAINMENTS REVIEW RESPONSIBILITIES
(
Primary - Structural Engineering Branch (SEB)
Secondary - Mechanical Engineering Branch (MEB)
Conta:irent Systems Branch (CSB)
I.
AREAS OF REVIEW The following creas relating to the containment internal structures are reviewed:
1.
Description of the Internal Structures The descriptive information includirig plans and sections of the various internal structures is reviewed to establish that sufficient information is provided to define the primary structural aspects and elements relied upon to perform the safety-related functions of these structures. The internal structures have several safety-related functions for which their structural integrity is important. By pro-viding support during normal operation and seismic disturbances, they shoul: prevent the occurrence of a loss of coolant accident (LOCA). If such an accident does occur, however, they should act to mitigate its consequences by protecting the containnent and other engineered safety features from the effects induced by the accident such as jet forces and whipping pipes.
The najor containnent internal structures that are reviewed, together with the primary structural function of each structure, and the extent of descriptive information required for each structure, are indicateo below. For equipment supports that are not covered by this plan, reference is nade to Standard Review Plan 3.9.3.
For PWR Dry Containment Internal Structures a.
Reactor Supports The PWR vessel should be supported and restrained to resist non"al operating loads, seismic loads, and loads induced by postulated pipe rupture including the loss of coolant accident. The support and restraint system should limit the movement of the vessel to within allowable limits under the applicable combinations of loadiras.
USNRC STANDARD REVIEW PLAN st.ndard review piene e,e pre,ered for the guidance of ihe o++ ice of==ciese Reecto, Regenerion even respeasibie for the review of.oni catione is construct end opersee nucsese pc -ce plante These docuciente are made evoitable to the public ee part of the Comm.esion a pohcy to entosm t'.e nueteer 6advetry and the generes pubhc of regusetory procedusee sad pohsees Stendard review p6ene are not sub.tetutes for reguietorv guedes or the Commission a regulation. end e
comphence wvth them as not requ6 red The standard review plan sectione are hoved to Rev'eson 2 r,f the Standeed Format and Con'ont of Safety Anesysis Reporte foe Nwcseet Nwee Ptente Not oil sections of the Stenderd Format have e careesponding rowen pasa Pubhehed etsaderd review p6ase weit be revised periodicen, se sopropnete to accoenmodete cominente and to refioct new enf ormat6on end empenerne Comniente sad musgest'ons for improvement will be considered and should be sent to the U 5 Nucleet RegWetery Commission. Office of Nucteet Reector Regveetten. Weenington D C 20685 146 220 7907120270
The support system should nevertheless minimize resistance to the thermal movements expected during operation.
With these functional requirenents in mind, the general arrangement and principal features of the reactor vessel linear supports are reviewed with emphasis on methods of transferring loads from the vessel to the support and eventually to the structure and its foundations. Shell-type supports and component standard supports are reviewed by the Mechanical Engineering Branch (MEB). The definition of linear, shell type, and standard supports is in accordance with Subsection NF of the ASME Boiler and Pressure Vessel Code,Section III, Division 1 (Ref.1).
Where uplift supports are utilized, the method of anchoring such supports in the concrete is also reviewed.
b.
Steam Generator Supports Steam generators should be supported and restrained to resist normal operating loads, seismic loads, and loads induced by pipe ruptuie. The support system should prevent the rupture of the primary coolant pipes due to a postulated rupture in steam or feedwater pipes and vice versa. The system should nevertheless minimize resistance to the thermal novements expected during operation.
With these functional requirements in mind, the general arrangement and principal features of the steam generator linear supports are reviewed with emphasis on rethods of transferring loads from the vessel to the support and eventually to the structure and its foundations. Shell-type supports, standard supports, and mechanical restraints such as hydraulic snubbers are reviewed by the Mechanicai Engineering Branch (MEB).
c.
Reactor Coolant Pu,p Supports Reactor coolant pumps should be supported and restrained to prevent excessive deflections during normal operating, seismic, and pipe rupture conditions. Under LOCA loads, the pump should not become a missile and should not generate missiles that might damage other safety-related components. The pump support system should also minimize resistance to thermal novements expected during operation.
With these f ctional requirements in mind, the ge.1eral arrangement and principal features of
? ump linear supports are reviewed with emphasis on methods of transferring loads from the pump to the support and eventually to the structure and its fourdations. Shell-type supports, standard supports, and rechanical rastraints such as hydraulic snubbers are reviewed by the Mechanical Engineering Branch (MEB).
d.
Primary Shield Wall and Reactor Cavity The primary shield wall forms the reactor cavity and usually supports and restrains the reactor vessel. It is usually a thick wall that surrounds the reactor vessel and may be anchored through the liner plate to the containnent base slab.
The general arrangement and principal features of the wall and cavity are reviewed including the main reinforcement and anchorage system.
)kb 3.8.3-2
e.
Secondary Shield Walls The secondary shield walls surround the primary loops, forming the steam generator compartments, and protecting the containment from the effects of pipe rupture acci-dents inside the compartment. They may also support intennediate floors and the operating floor. The general arrangement and principal features of these walls are reviewed with emphasis on the method of structural framing and exoected behavior under compartment pressure loads.and jet forces, particularly those associated with the LOCA.
f.
Other Interior Structures The other major interior structures of PWR dry containments that are reviewed in a similar manner are the pressurizer U:; ear supports, refueling pool walls, the operating floor, other inunediate floors, and the polar crane supporting elements.
For PWR Ice-cond7nser Containment Internal Structures In PWR plants where the ice-condenser containment system is utilized, in addition to the applicable structures reviewed in dry PWR containments, the following elements are also reviewed:
a.
The Divider Barrier In the PWR ice-condenser containment system, which utilizes the pressure-suppression concept, the divider barrier surrounds the reactor coolant system.
The upper portion of the divider barrier is nearly sur.munded by the ice-condenser which is bounded by the containment shell on the outside and by the divider barrier wall on the inside. Several venting doors connect the space inside the divider barrier to the ice-condenser.
In the event of a LOCA, the divider barrier will contain the steam released from the reactor coolant system and, temporarily acting as a pressure-retaining envelope, will channel the steam through the venting doors and into the ice-condenser. The ice will condense the steam and the energy released to the containment will thus be minimized.
Following such a LOCA and before blowdown is completed, the divider barrier will be subjected to a differential pressure and possibly jet forces, and any structural failure in its boundary may result in steam bypassing the ice-condenser and flowing directly into the containment, possibly generating a containment pressure higher than that for which it has been designed.
With this functional requirement in mind, the general arrangement and principal features of the divider barrier are reviewed with emphasis on structural framing and expected behavior when subjected to the design loads.
b.
Ice-Condenser A major feature of the ice-condenser containment is the ice-condenser which contains the baskets of ice forming the heat sink essential for pressure suppression. The structurally significant components of the ice-condenser that are reviewed are the vent doors, ice baskets, brackets, couplings and lattice framings, lower and upper supports, and insulating and cooling panels.
46 230 3.8.3-3
The general arrangment and principal features of these major components are reviewed with emphasis on the structural framing, supports, and expected behavior when sub-jecter to design loads.
For BWR Coc :nment Internal Structures Since it is expected that future BWR applications will utilize the Mark III contain-ment concept, this Standard Review Plan is oriented towards and based on this type of containment. For other types of BWR containments, modifications to this plan are made on a case-by-case basis.
Among the major Mark III contali ment internal structures that are reviewed, together with the primary structural function of each structure, and the extent of descr iptive information required for each structure, are the following:
a.
Drywell In the BWR Mark III containment systen, which utilizes the pressure-suppression concept, the drywell surrounds the reactor coolant system. The lower portion of the drywell is surrounded by the suppression pool which is bounded by the containment shell on the outside and by a weir wall located just inside the drywell wall. Several vent holes connect the drywell to the suppression pool.
In the event of a loss-of-coolant accident, the drywell will contain the steam released from the reactor coolant system and, tenporarily acting as a pressure-retaining envelope, will channel the steam through the vent holes and into the suppression pool. The pool water will condense the steam and the energy released to the containment will thus be minimized.
Following such a LOCA and before blow wn is completed, the drywell will be sub-jected to a differential pressure and possibly jet forces, and any structural failure in its boundary would result in steam bypassing the suppression pool and flowing directly into the containment, possibly generating a containment pressure higher than that for which it has been designed.
With this functional requirement in mind, the general arrangement and principal features of the drywell are reviewed with emphasis on structural framing and expected behavior under loads. Since the drywell geometrically resembles, to a certain degree, a containment, the descriptive infomation reviewed is similar to that reviewed for containments as delineated in Section I.1 of Standard Review Plan 3.8.1.
The major components of the drywell that are so reviewed, other than the main body of the drywell, include the bottom vent region, the roof and drywell head, and major penetrations.
b.
Weir Wall The weir wall foms the inner boundary of the suppression pool and is located inside the drywell. It completely surrounds the lower portion of the reactor coolant system. The general arrangement and principal features of the weir wall are reviewed with emphasis on structural framing and behavior under loads.
O
)ffl 3.8.3 4
c. Refueling Pool and Operating Floor The refueling pool walls are located on top of the drywell. The outer walls form a rect.ngular peol that is usually subdivided by two interior crosswalls. The base slab )f the pool is corron to the drywell roof slab. The pool may be filled continuously with water for shielding purposes during operati3n. The general arrangement and principal features of the refueling pool are revised with emphasis on structural framing and behavior under loads. The operating floor is intended to provide laydown space for refueling operations and is usually a combination of reinforced concrete and structural steel framing. The containment walls and the refueling pool walls may support the floor. The general arrangement and principal features of the operating floor are reviewed. d. Reactor and Recirculation Pumo Supports The support systems of the BWR vessel and recirculation pumps have the same functions as the support systems for PWR vessels and pumps are similarly reviewed. e. Reactor Pedestal The reactor pedestal is usually a cylindrical structure located below and supporting the reactor vessel, which is anchored to the top of the pedestal. The general arrangenent and principal features of the reacto; pedesta, are reviewed wi th erThasis on structural framing, main reinforcerent and the manne r in which the pedestal is anchored to the containnent base slab. f. Reactor Shield Wall This is usually a cylindrical wall surrosing the reactor vessel for radiation shielding purposes. It is supported on the reactor pedestal. The wall may be linea on both surfaces with steel plates which also may act as the main structural components of the wall. The wall may also be utilized as an anchor for pipe restraints. The general arrangement and principal features of the wall are reviewed with particular emphasis on structure framing and behavior under loads. g. Other Interior Structures The other major interior structures constructed of rtinforced concrete or struc-tural steel or combinations thereof that are also reviewed in a similar manner are the floors located inside the drywell and in the annulus between the drywell and the containment, and the polar crane supporting elements. The general arrangerwnt and principal features of these structures are reviewed. 2. Applicable Codes, Standards, and Specificatinns The ir. formation pertaining tr design codes, standards, specifications, regulations, general design criteric and regulatory guides, and other industry standards that are applied in the design, fabrication, construction, testing, and surveillance of the containnent internal structures, is reviewed. The specific edition, date, or addenda identified for each document are also reviewed. )kb 3.8.3-5
J. Loads and Loading Combinations Information pertaining to the applicable design loads and various load combinations thereof is reviewed. The loads nemally applicable to containment internal structures include the following: a. Those loads encountered during nomal plant startup, operation, and shutdown, including dead loads, live loads, themal loads due to operating temperature, and hydrostatic loads such as in refueling and pressure suppression pools. b. Those loads to be sustained during severe environmental conditions, including those induced by the operating basis earthquake (OBE) specified for the plant site, c. Those loads to be sustained during extreme environmental conditions, including those induced by the safe shutdown earthquake (SSE) specified for the plant site. d. Those loads to be sustained during abnomal plant conditions. The most critical abnomal plant condition during which most of the containment internal structJres have to perform their primary function is the design basis LOCA. Ruptures of other high-energy pipes should also be considered. Time-dependent and dynamic loads induced by such accidents include elevated temperatures and dif ferential pressu es across compartments, jet impinoament, impact forces associated with the postulated ruptures of piping, and loads e.r;.icable to some structures such as pool swell loads in tr. BWR Mark III containment and drag forces in the PWR ice-condenser containment. The various combinations of the above loads that are nomally postulate; and reviewed include the following: nomal operating loads; nomal operating loads with severe environmental loads; normal operating loads with extreme environmental loads; nomal operating loads with abnomal loads; normal operating loads with severe en'.ironmental and abnomal loads; and normal operating with extreme environnental and abnormal loads. In addition, the following infomation is reviewed: a. The extent to which the applicant's criteria comply with the " Building Code Require-ments for Reinforced Concrete," /.CI 318-71 (Ref. 2) for concrete, and with the AISC " Specification for Design," Fabrication and Erection of Structural Steel for Buildiags" (Ref. 3) for steel, as applicable. b. For concrete portions of the divider barrier of the PWR ice-condenser containment and for concrete portions of the drywell of the Mark III BWR containment, the extent to which tne applicant's loading criteria comply with Article CC-3000 of the proposed " Standard Code for Concrete Reactor Vessels and Containments," ACI-ASME (ACI-359) (Ref 4). For steel pressure-resisting portions of these two structures, the ey'ent to which the applicant's loading criteria comply with Article NE-3000 of Subser : ion NE cf the ASME Code, Section III, Div.1, (Ref. 5) as augmented by Regu-latory Guide 1.57 (Ref. 9). c. For steel linear supports of the reactor coolant system, the extent to which the applicant's criteria comply with Subsection NF of tiie ASME Code, Section III, Division 1 (Ref.1). )46 2J7' 3.8.3-6
4. Design and Analysis Procedures The design and analysis procedures utilized for the containment internal structures are reviewed with emphasis on the extent of compliance with the applicable codes as indicated in Section I.3, including those applicable to the following areas: For PWR Dry Containment Internal Structures a. Reactor Coalant System Supports The support system for the reactor vessel, steam generators, and reactor coolant pumps, as described in Section I of this plan, shc designed to resist various combinations of loadings, including noma! c -ating loads, seismic loads, and loss of coolant and other pipe rupture accident loads. Analytical procedures for determining nomal operating loads and accident loads are reviewed by the Mechanical Engineering Branch (MEB). Analytical procedures for detemining seismic loads are as described in Standard Review Plan 3.7.3. Af ter the procedur es for detemining individual loads and combinations thereof are so reviewed, the design and analysis methods utilize. for the linear supports are reviewed including the type of analysis (elastic or piastic), the methods of loa'i transfer, and the assumptions on boundary condi;ns. Specifically, the extent of compliance with design and analysis procedures :elineated in Subsection NF of the ASME Section III Code, Division 1 (Ref.1), n revi ewed, b. Primary Shield Wall and Reactor Cavity The primary shield wall sh uld withstand all the applicable loads including those transmitted throtgh the reactor supports. It is subjected to most of the loads described in Section I.3 of this plan and should be designed and analyzed for all the applicable load combinations. During noma' plant operation, a themal gradient across the wall is generated by the attenuatior heat of gama and neutron radiation originating from the reactor core. Insulation and cooling systems may be provided to reduce the severity of this gradient by limiting the rise in temperature to an acceptable level. Procedures Tor determining seismic loads on the primary shield wall are reviewed in accordance with Standard Review Plan 3.7.2. Loss of coolant accident loads that are applicable to the primary shield wall include a differential pressure created across the reactor cavity by a pipe break in the vicinity of the reactor nozzles. Such a transient pressure may act on the entire cavity or on portions thereof. Procedures for detemining such pressures are reviewed by the Containment Systems Branch (CSB). Other less of coolant accident loads that apply are those transmitted to the wall through the reactor supports including pipe rupture reaction forces which may induce simultaneous shear forces, torsional momants, and bending moments at the base of the wall. Further, the elevated temperature within and around the primary shield 3.8.3-7 \\46 2M
created by the accident may produce transient thermal gradients across the thick wall. Design and analysis procedures for such accident effects are accordingly reviewed. c. Secondary Shield Walls The secondary shield walls surrounding the primary loops and supporting the operating floor should be designed for loads similar to those applicable to the primary shield wall including loads of fluid jets from a postulated break of a primary pipe which can impinge on these walls. The analytical techniques utilized for these walls are reviewed including their structural framing and behavior under loads. Where elasto-plastic behavior is assumed and the ductility of the walls is relied upon to absorb the energy associated with Jet loads, the procedures and assumptions are reviewed with particular emphasis on such areas as modeling techniques, boundary conditions, force-time functions, and assumed ductility. For the time-dependent differential pressure, however, elastic behavior is required and the methods of detemining an equivalent static load are accordingly reviewed. d. _0ther Interior Structures Most of the other interior structures that are also reviewed are combinations of slabs, walls, beams and columns, classified as Category I structures and subject to most of the loads aad combinations described in Section I.3 of this plan. Analytical techniques for these structures are reviewed on the same basis as for the structures described above. For PWR ice-Condenser Containment Internal Structures a. Divider Barrier Since the divider barrier has to maintain a certain degree of leak-ti,ghtness during a LOCA and is thus a critical structure with respect to the proper functioning of the containment, it is treated on the same basis as the containment. The loads that usually govern the design of the divider barrier are those induced by the LOCA, including the time-dependent differential pressure across the barrier and any concurrent concentrated jet impingenent loads. As the divider barrier is typically a combination of walls and slabs framed together, the design and analysis procedures are of the conventional type. They are accordingly reviewed with emphasis on the assuned boundary conditions and behavior under loads. Since the differential pressure ed jet impingement loadings are dynamic impulsive loads that vary with time, the techniques utilized to detemine their equivalent static loads are reviewed, b. Ice-Condenser The design of the ice-condenser and its various components may be based on a combination of analysis and testing. The analytical and testing procedures that are reviewed include those for the ice baskets and brackets (couplings); the lattice frames and columns including attachments; the supporting struc*ures com-prising the lower supports; the wall panels and cooling duct and supports of various auxiliary components. 146 235 3.8.3-3
The ice-condenser and its components should be analyzed or tested for various loads and combinations thereof including dead and live loads, themal loads induced by differ antial thermal expansion within the various elements, seismic loads and loads induced by the loss-of-coolant accident. Accident loads include pressure differential drag loads and loads induced by the change of momentum of the flowing steam. Elastic analysis is vmally utilized for the ice-condenser and its components. However, plastic analysis may also be used as an alternate. Accordingly, the load factors that are applied to each of the applicable loads and the basis and justification of these load factors are reviewed. Where experimental verification of the design using simulated load conditions is used, the procedures used to account for similitude relationships which exist between the actual component and the test model are reviewed to assure that the results obtained from the test are a conservative representation of the load carrying capability of the actual component under the postulated loading. For BWR Containment Internal Structures a. Drywell The drywell, which has to maintain a certain degree of leak-tightness during a LOCA, is critical with respect to the proper functioning of the containment. Accordingly, and since it geometrically resembles a contaiment, the design and analysis procedures utilized for the drywell are reviewed on a basis similar to those of containments as described in Section I.4 of Standard Review Plans 3.8.1 and 3.8.2 for concrete and steel portions, respectively. b. Weir Wall One of the major loads to which the weir wall may be subjected is a jet impingement load induced by a pipe rupture in a nearby recirculation loop under such a con-centrated load, the weir wall should not defom to an extent that might impair or degrade the pressure-suppression perforrance. Accordingly, the procedures utilized to analyze the wall for such dynanic time-dependent loads are reviewed with particular emphasis on modeling techniques, assumptions un boundary con-ditions, and behavior under loads. c. Refueling Pool and Operating Floor In the BWR Mark III containments reviewed recently, the refueling pool is continuously filled with water to provide biological shielding above the reactor. The operat ing floor, which may be Supported on the walls of the refueling pool on one nue and on the containment shell on the other side, is a combination of reinforced concrete and structural steel. The design and analysis procedures for the refueling pool and the operating floor are of the comentional type and are accordingly reviewed, with particular emphasis on the structural framing and behavior under loads. In cases where the floor beams are supported vertically on the containment shell, they should be laterally isolated to minimize interaction between the cortainment and its interior. 3.8.3-9
d. Reactor and Recirculation Pump Supports The design and analysis procedures utilized for the reactor and recirculation pump supports are reviewed in a similar manner to that for PWR reactor and pump supports, as already described in this plan, e. Reactor Pedestal The reactor pedestal supports the reactor and has to withstand the loads trans-mitted through the reactor supports. It is thus subjected to most of the loads described in Section I.3 of this plan and is designed and aralyzed for all the applicable load combinations. Because of the similarity in geometry and function of the BWR reactor pedestal to the PWR primary snield wall, the design and analysis procedures are similar and are reviewed accordingly as has already been discussed in this plan, f. Reactor Shield Wall This cylindrical wall, which surrounds the reactor and provides biological shield-ing, is also subjected to most of the loads described in Section I.3 of this plan. In most cases, the wall is utilized to anchor pipe restraints placed around tr 3 reactor coolant system piping. Moreo/er, a pipe rupture in the vicinity of the reactor nozzles may pressurize the space within the wall. The wall is usually lined on both faces with steel plates which may constitute the major structural elements relied upon to resist the design loads. The analytical and design techniques utilized to detemine the effect of the design Icads on the wall are eviewed with particular enphasis on the assumed boundary conditions and the behavior of the wall under loads. g. Other Interior Structures There are several platfoms within the BWR Mark III containment some of which are inside the drywell and the others outside in the annulus becween the drywell and the containment. Platfoms inside the drywell are usually of structural steel and their main structural function is to provide foundations for the pipe restraints inside the drywell. Platforms outside the drywell are usually combinations of steel and concrete and have to be designed to resist the various applicable loads particularly the effects of pool swell during a loss-of-coolant accident. The analytical procedures for detemining pool swell loads are reviewed by the Contain-ment Systems Branch (CSB). Design and analysis procedures for these platfoms are reviewed with particular emphasis on the framing and structural behavior under loads. 5. Structural Acceotance Criteria The design limits imposed on the various parameters that serve to quantify the struc-tural behavior of the various interior structures of the containment are reviewed, specifically with respect to stresses, strains, defamations, and factors of safety against structural failure, with emphasis cn the extent of compliance with the applicable codes as indicated in Section I.3 of this plan. 146 237 3.8.3-10
6. Materials, Quality Control, and Special Construction Techniques Information provided on the materials that are used in the construction of the contain-ment internal structures is reviewed. Among the major materials of construction that are reviewed are the concrete ingredients, reinforcing bars and splices, and structural steel and various supports and anchors. The quality contral program that is proposed for the fabrication and construction of the containment 'nterior structures is reviewed including nondestructive examination of the materials to determine physical properties, placement of concrete, and erection tolerances. Special, new, or unique construction technl ques, if proposed, are reviewed on a case-by-case basis to determine their effects on the structural integrity of the completed interior structure. In addition, the following information should be provided: The extent to which the naterials and quality control programs comply with the a. " Building Code Requirements for Reinforced Concrete," ACI 318-71 (Ref. 2), for concrete, and with the AISC " Specifications for Design, Fabrication and Erection of Structural Steel for Buildings,"(Ref. 3), for steel, as applicable. b. For steel linear supports of the reactor coolant system, the 2xtent to which the material and quality control programs comply with Subsection NF of the ASME Code, Section III, Division 1 (Ref.1). For quality control in general, the extent to which the apolicant complies with c. ANSI N45.2.5 (Ref. 7), d. If welding of reinforcing bars is proposed, the extent to which the applicant com-plies with the applicable sections of the proposed " Stand Concrete Reactor Vessels and Containmc7ts " ACI-ASME (ACI-359) (F 4), should be described and any exceptions taken should be justified. 7. Testing and Inservice Surveillance Programs If applicable, any post-construction testing and in-service surveillance programs are reviewed on a case-by-case basis. The structural test for the drywell of the BWR Mark III containment is reviewed in a similar manner to that of the containment. II. ACCEPTANCE CRITERIA The acceptance criteria for the areas of review are as follows: 1. Descri". ion of the Internal Structures The uescriptive infomation in the SAR is considered acceptable if it meets the minimum requirements set forth in Section 3.8.3.1 of the " Standard Fomat and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2 (Ref. 8). 146 2M 3.8.3-11
Deficient areas of descriptive information are identified by the reviewer and a request for additional informaticn is initiated at the application acceptance review. New or unique design features that are not specifically covered in the " Standard Format" may require a more detailed review. The reviewer determines if additional information is required to accomplish a meaningful review of the structural aspects of such new or unique features. 2. Applicable Codes, Standard, and Specifications The design, materials, fabrication, erection, inspection, testing, and in-service sur-veillance, if any, of interior structures of containments are covered by codes, standards, an'. guides that are either applicable in their entirety or in portions there-of. The following codes, Standards, specifications, and guides are acceptable. Code, Standard,or Speci fica ti on Title ACI 318-71 Building Code Requirements for Reinforced Concrete ACI/ASME (ACI-359) Proposed Standard Code for Concrete Reactor Vessels and Containments, ASME Boiler and Pressure Vessel Code, Section III, Division 2 ASME Boiler and Pressure Vessel Code, Section III, Subsections NE and NF AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings ANSI M5.2.5 Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants Regulatory Guides 1.10 Mecha.kal (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures 1.15 Testing of Reinforcing Bars for Category I Concrete Structures 1.55 Concrete Plac ant in Category I Structures 3. Loads and Load Combinations With the exception of the dividar-barrier and ice-condenser elements of the ice-condenser PWR containment, the drywell of the BWR Mars III contair. ment, and the steel linear supports of the reactor coolant system, the loads and load combinatior; for all other containment interior structures described in Section I.1 of this plan, are acceptable if found in accordance with the following: Loads, Definitions,and Nomenclature All the major loads to be encountered or to be postulated are listed below. All the loads listed, however, are not necessarily applicable to all the interior structures. Loads and the applicable load combinations for which each structure has to be designe6 will depend on the concitions to which that particular structure could be subjected. 3.8.3-12 146 239
Normal loads, which are those loads to be encountered during normal plant operation and shutdown, include: D --- Dead loads or their related internal moments and forces, including any pemanent equipment loads and hydrostatic loads. For equipment supoorts, it also includes static and dynamic head and fluid flcw effects. L --- Live loads or their related internal moments and forces, including any movable equipment loads and other loads which vary with intensity and occurrence. For equipment supports, it also includes loads due to vibration and any support movement effects. T --- Themal effects and loads during nomal operating or shutdown conditions, g based on the most critical transient or steady state condition. R --- Pipe reactions during nomal operating or shutdown conditions, based on g the most critical transient or steady state condition. Severe environmental loads include: E --- Loads generatea by the operating basis earthquake. Extreme environmental loads include: E' --- Loads generated by the safe shutdown earthquake. Abnormal loads, which are those loads generated by a postulated high-energy pipe break accident, include: P --- Pressure equivalent static load within or across a compartment generated a by the postulated break, and incluaing an appropriate dynamic load factor to account for the dynamic nature of the ioad. 9 T --- Themal loads : s e thermal conditions generated by the postulated break a and including T. g R --- Pipe reactions under themal conditions generated by the postulated break a and including R.g Y --- Equivalent static load on the structure generated by the reaction on the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factcr to account for the dynamic nature of the load. Y --- Jet impingement equivalent static load on a structure generated b, the j postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Y --- Missile impact equivalent static load on a structure generated by or during the postulated break, as f rom pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load. In determining an appropriate equivalent static load for Y ' i, ard Y, elasto-plastic r j behavior may be assumed with appropriate ductility ratios, provided excessive deflec-tions will not result in loss of function of any safety-related system. Load Combinations for Concrete Structures For concrete interior structures, the load combinations are acceptable if found in accordance with the following: }kh 3.3.3-13
a. For service load conditions, either the workinc, stress design (WSD) method or the strenght design method may be m ed. (i) If the WSD method is used, the following load combinations should be considered: (1) D+L (2) D+L+E If thermal stresses due to T and R are present, the following combinations g g should be also considered: (la) D + L + T + Rg (2a) D + L + T +R +E g g Both caces of L having its full value or being completely absent should be checked. (ii) If the strength design method is used, the following load combinations should be considered: (1) 1.4D + 1.7L (2) 1.4D + 1.7L + 1.9E If thermal stresses due to T and R are present, the following combinations g g should also be considered: (lb) (0.75) (1.4D + 1.7L + 1.7T + 1.7R ) g g + 1.7R ) (2b) (0.75) (1.4D + 1.7L + 1.9E + 1.7Tg g b. For factored load conditions, which represent extreme environmental, abnormal, abnormal / severe environmental, and abnormal / extreme environmental conditions, the strength design method should be used and the following load combinations should be considered: (3) D+L+T +R
- E' g
g (4) D+L+T + R, + 2. 5 P a a j + Y ) + 1.25E (5) D+L+T +R + 1.25 P + 1.0 (Y + Y m a a a p (6) D + L + T, + R + 1.0 P + 1.0 (Y +Y j + Y ) + 1.0 E' m a a p In combinations (4), (5), and (6), the maximum values of P,, T, R, Y, Y, and a a j r Y, including an appropriate dynamic load factor, should be used unless a m time-history analysis is performed to justify otherwise. Combinations (5) and (6) and the correspondirg structural acceptance criteria of Sec+ in 11.5 of this plan should first be satisfied without Y, Y, and Y,. When considering these r j loads, local section strength capacities may be exceeded under these concentrated loads, provided there will be no loss of function of any safety-related system. Both cases of L having its full value or being completely ament should be chGked. Load Ccmbinations for Steel Structures For steel interior structures, the load combinations are acceptable if found in accordance with the following: a. For service load conditiors, either the elastic working stress design methods for Part 1 of AICC, or the plastic design methods of Part 2 of AISC, may be used. O \\46 wL 3.9.3-14
(1) If the elastic working stress design methods are used: (1) D+L (2) D+L+E If thermal str' esses due to T and R are present, the following cc.nbirations o g should also be considered: (la) D+L+T +R g g (2a) D+L+T +R +E g g (ii) If the plastic design methods are used: (1) 1.70 + 1.7L (2) 1.7D + 1.7L + 1.7E If themal stresses due to T and R are preser.t, the following combinat1ons g g should also be considered: (lb) 1.3 (D + L + T +R) g g (2b) 1.3 (D + L + E + T +R) g g b. For factored load conditions, the following load combinations should be considered: (i) If the elastic working stress design methods are used: (3) D+L+T +R + E' g g (4) D+L+T + +P a a a (5) D+L+T +R +P + 1.0 (Y3+Y +Y)+E a a a m (6) D+L+T +R +P + 1.0 (Yj+Y +Y)+E' a a a (ii) If the plastic design methods are used: (3) D+L+T +k + E' g g 9 (4) 0+L+T +R + 1.5 P a g a (5) D+L+T +R + 1.25 P + 1.0 (Yj+Y + Y ) + l.25 E a a 3 r m (6) D+L+T + R, + 1. 0 P 1.0 (Y) + Y + Y ) + 1.0 E' + a a r g In the above combinations, themal loads can be neglected when it can be shown that they are secondary and self-limiting in nature. In combinations (4), (5), and (6), the maximun values of P, T, R, Y, Y, nd Y ' 3 a 3 3 r m including an appropriate dynamic load factor, should be used unless a time-history analysis is performed to justify otherwise. Combinations (5) and (6) and the corresponding structural acceptance criteria of Section II.5 of this plan should first be satisfied without Y, Y, and Y. When considering these loads, however, local section strengths or stresses may be exceeded under these concentrated loads, provided there will Fa no loss of function of any 2.ic'c related system. For the divider barrier, ice-condenser elements, tne Mark III containment drywell, and for the steel linear supports of the reactor coolant system, the loading criteria are acceptable if found in accordance with the following: a. Divider barrier As the structural integrity of the divider barrier and, to a certain extent, its leak-tight integrity as well, are important to the proper funct;cainq of the ice-condenser containment system, it is treated for design purr ts similar to the containment itself. 3.8.3-15
Accordingly, for concrete pressure-resisting portions of the divider barrier, the loads and load combinations of Article CC-3000 of /CI-359 (Ref. 4) will apply, with the following exceptions. For Table CC-3200-1 Ii) Jet impingement loads, Y, and impact loads of missiles associated with the j loss-of-coolant accident, Y, should be included. (ii) The 6th combination, representing abnomal conditions, need not include Y in combination with 1.5 P ' r a (iii) In the 7th, 8th, and 9th combinations, representing abnormal / severe environmental and abnormal / extreme environmental load conditions, the "and/or" between R and Y should be deleted and, in addition to R nd Y ' a r a r the combinations should include Y and Y
- j m
(iv) It should be indicated that the maximum values of P, T, R, Y, Y, and Y ' a a a r 3 m including an appropriate dynamic load factor, should be applied simultaneously, unless a time-history analysis is perfomed to justify otherwise. Steel portions of the divider barrier which resist the d' sign differential pressure and are not backed by concrete, such as penetrations, hatches, locks and guard r,ipes, should be designeo in accordance with the appropriate sections of Subsection NE of the ASME Code, Section III, Division 1, (Ref. 5) together with the applicable loads, load combinations, and acceptance criteria of Regulatory Guide 1.57, (Ref. 9). Specifically, the load combinations of Section II.3 of Standard Review Plan 3.8.2 apply. b. Ice-condenser Elements In the ice-condenser containment system the structural integrity of the ice baskets, ice bed framing, and their supports, is important to the functional integrity of the containment system. The major loads that are applicable to the ice-condenser elements are: D, L E, E', and P. For this structure, P is the a a LOCA pressure load induced by drag and change in momentum of the flcwing air and steam. Load combinations for the ice-condenser elements are acceptable if found in accordance with the following: (i) For service load conditions, if elastic working stress design methods are used: (1) D+L (2) D+L+E (ii) For service load conditions, if plastic design methods are used: (1) 1.7 D + 1.7 L (2)
- 1. 7 D + 1. 7 L F 1.7 E (iii) For service load conditions, if an experimental test verification of the design is used:
(1) 1.9 0 + 1.9 L (2) 1.9 D + 1.9 L + 1.9 E If thermal stresses are significant and have to be considered, an acceptable pro-cedure for accounting for SJch themal loads is contained in item (a) of Subarticle NF-3231.1 of Subsection Nt of the ASME Code, Section III, Division 1 (Ref.1). liv) For factored load conditions, if elastic working stress design methods are used: 3.8.3-16 146 243
(3) D + L + E' (4) D+L+P a 9 (5) D + L + P, + E' (v) For factored load conditions, if plastic design methods are used: (3) 1.3 0 + 1.3 L + 1.3 E' (4) 1.3 D + 1.3 L + 1.3 Pa (5)
- 1. 2 0 + 1. 2 L + 1. 2 P + 1. 2 E '
a (vi) For factored load conditions, if an experimental test verification of the design is used: (3) 1.4 D + 1.4 L + 1.4 E ' (4) 1.4 D + 1.4 L + 1.4 Pa (5) 1.3 0 + l 3 L + 1.3 P + 1.3 E' 3 c. BWRMarkIIIContainmentDryneR As the structural integrity of the drywell and, to a certain extent, its leak-tight integrity as well, are critically important to the proper functioning of the Mark III pressure-suppression system, the drywell is treated, for design and testing purposes only, similar to the containment itself. Accordingly, for concrete pressure-resisting porticas of the drywell, the loads and loading combinations of Article CC-3000 of ACI-359 (Ref. 4) will apply, with the exceptions listed for concrete portions of the PWR ice-condenser divider barrier. For steel components of the drywell that resist pressure and are not backed by concrete, such as the drywell head, the appropriate sections of Subsection NE of the ASME Code, Section III, Division 1, (Ref. 5) should be used together with the applicable loads, load cortbinations, and acceptance criteria of Regulatory Guide 1.57 (Ref. 9). Specifically, the load conbinations of Section II.3 of Standard Review Plan 3.8.2 apply. For the lower vent portion of the drywell: (i) If the main reinforcement of the drywell is carried down between the vent holes and the reinforced concrete section is relied upon for structural purposes, the criteria that apply to concrete portions of the drywell as described above will apply. (ii) If the main reinforcement of the drywell is terminated above the vent holes and two steel plates lining both faces of the drywell are alone utilized for structural purposes, the criteria that apply to steel portions of the drywell as described above will apply. (iii) If other structural systems are used in the vent region, the loads and load combinations are reviewed and judged on a case-by-case basis. d. Reactor Coolant System Supports Steel linear supports for the reactor vessel, steam generators, reactor coolant pumps, and recirculation pumps, as describeri in Section I of this plan, are governed by Subsection NF of the ASME Code, Section III, Division 1 This Code does not explicitly delineate load combinations for the design of these supports. Accordingly, the following combinations should be sa.isfied as a minimum: 1.8.3-17 146 24,-D
Load Combinations (1) If the elastic method r? analysis of paragraph NF-3231.1 of Subsection NF of the ASME Code. Section III, Division 1, is used, the following combinations should be satisfied as a minimum: (i) D+L+E (ii) D + L + E ' + P, + Y +Yj+Y r In addition, the conditions of item (a) of paragraph NF-3231.1 shall be satisfied. (2) If the limit method of analysis of paragraph NF-3231.2 of Subsection NF is ased, the following combinations should be sathfied as a minimum: (i) 1.7 (D + L + E) (ii) 1.0 (D + L + E' + P +Y +Y +Y) r j 4. Design and Analysis Procedures The design and analysis procedures utilized for the interior structures of the containment are acceptable if found in accordance with the following: For PWR Dry Containnent Internal Structures a. Reactor Coolant System Supports The linear support systems for the reactor vessel, steam generators, and reactor ccolant pumps, as described in Section I of this plan, should be anaiyzed for and designed to resist various combinations of loadings as indicated in Section II.3 of this plan. Design and analysis prccedures for such supports are acceptable if in accordance with Subsection NF of the ASME Section III Code, Division 1, (Ref.1), particularly with Appendix XIII. b. Primary Shield Wall and Reactor Cavity The design and analysis procedures utilized for the shield wall are acceptable if in accordance with the ACI 318-71 Code (Ref. 2). This code is mostly based on the strength design method. However, the use of Section P.10 of the Code, which is based on the working stress design method where actual elastic / linear stresses in the concrete and reinforcement are detemined and compared with their cur-responding allowables, is considered acceptable. Analyses for loss-of-coolant accident loads applicable to the primary shield wall, such as for the cavity differential pressure combined with pipe rupture reaction forces, are acceptable if these loads are treated as dynamic time-dependent loads whereby either a detailed time-history analysis is performed, or a static analysis utilizing the peak of the forcing function amplified by an appropriately chosen dynamic load factor is utilized. Elastic behavior of the wall should be maintained under the differential pressure. However, for the concentrated accident loads such as Y and Y, elasto-plastic behavior may be assumed as long as the deflections are limited to maintain functional requirenents. Simplified methods for deter-nining effective dynamic load factors for elastic behavior are acceptable if in accordance with recognized dynamic analysis methods. 3.8.3-18
c. Secondary Shield Walls Design and analysis procedures utilized for the secedary shield walls are acceptable if in accordance with conventional beam / slab design and analysis procedures described in the ACI 318-71 Code. Similar to the primery shield wall, the secondary shield walls are also subject to dynamic loss-of-coolant accident loads and the same methods desc-ibed in paragraph b. above are, therefore, applicable and acceptable, d. Other Interior Structures fiost of the other interior structures that are reviewed are combinations of reinforced concrete and steel slabs, walls, beams, and columns, which are clas-sified as Category I structures subject to the loads and load combinations described in Section II.3 of this plan. Analytical techniques for these struc-tures are acceptable if found in accordance with those described in the ACI 318-71 Code for concrete and with those in the AISC specifications for steel. For PWR Ice-condenser Containment Internal Structures a. Divider Barrier The most important loads that usually govern the design of the dividar barrier are those induced by the loss-of-coolant accident, including the differential pressure across the barrier and any concentrated jet impingemer.t loads. As thc divider barrier is a combination of walls and slabs framed together, the design and analysis procedures are acceptable if in 3cccrdance with those contained in Section 8.10 of th? ACI 318-71 Code for the concrete portions of the divider barrier. These methods are based on the elastic / linear working stress design method where actual stresses are determined. For steel portions of the divider barrier that resist pressure but are not backed by structural concrete, the design and analysis procedures are acceptable if fcand in accordance with the applicable provisions of Subsection NE of the ASME code, Section III, Division 1. b. Ice-condenser Elements The design and analysis procedures for the ice-condenser and its various components are acceptable if in accordance with either the elastic / linear design method of Part 1 of the AISC Specifications or with the plastic design method of Part 2 of the same Specifications. For components where experimental testing is utilized to verify the design, the testing procedures are acceptable if in accordance with recognized prototype or modal testing procedures whe e the effect of scaling and similitude are taken into consideration. For BWR Containment Internal Structures a. Drywell The design and analysis procedures utilized for concrete portions of the drywril are acceptable if in accordance wi th Section II.4 of Standard Review Plan 3.8.1. For steel portions of the drydell that resist pressure but are not backed by struc-tural concrete, the design and analysis procedures are acceptable if found in kb 3.8.3-1)
accordance with the applicable provisions of Subsection NE of the ASME Code, Section III, Division 1. b. Weir Wall One of the major leads to which the weir wall may be subjected is a jet impingement load induced by a pile rupture in a nearby recirculation loop. The deflection of the wall under such a load must be limited so as not to impair the pressure-suppression perfomance. The procedures utilized to analyze the wall for such a dynamic time-dependent load are acceptable if a detailed time-history dynamic analysis is performed or if an equivalent static analysis S performed utilizing the peak of the jet ioad amplified by an appropriately chosen dynamic load factor. c. Refueling Pool and Operating Floor The refueling pool and the operating floor, which may be supported on the walls of the refueling pool on one side and on the containment shell on the other side, are a combinatior of reinforced concrete and structural steel. The design and analysis procedures are acceptable if found in accordance with conventional methods described in th; ACI 318-71 Code for concrete and in the AISC Specifications for structural steel. d. Reactor Supports The linear support system for the reartnr vessel, described in Section I of this plan, should be designed to resist various combinations of loadings as indicated in Section II.3 jf this plan. Among the major loads that should be considered are normal operating loads, seismic loads, and loss-of-coolant accident loads. Desigr, a.m analysis procedures are acceptable if in accordance with those delineated in Suhsection NF of the ASME Section 1[I Code, Division 1, particularly alth ApNndix XVII. e. RactorPedestal lhe reactor pedestal, which supports the reactor and has to withstand the loads transmitted through the reactor supports, should be sabjected to most of the loads described in Section II.3 and should be designed for all the applicable load combinations. The design and analysis proceoures are acceptable if found tn he similar to those referenced for the primary shield wall of PWR containments in para. graph (b) under PWR dry containments, f. Reactor Shield Wall This cylindrical wall, which surrounds the reactor and provides biological shielding, should be subjected to most of the loads described in Section II.f of this plan. In most cases, the wall is utilized to anchor most of the pipe restraints placed arounc the reactor coolant system piping. A pipe rupture in the vicinity of the reactor nozzies may pressurize the space within the wall. The wall may be lined on both f ates with steel plates which may constitute the major etructural elements relied upon to resist the design loads. 9 fg 3.8.3-20
Similar to the reactor pedestal, the biological shield wall is also subjected to dynamic loss-of-coolant accident loads and the same methods are, therefore, applicable and acceptable. g. Miscellaneous Platfoms Platforms inside the drywell are usually of structural steel and their main structural function is to pro.ide foundations for the pipe restraints inside the dryweli. Platforms outside the drywell are usually combinations of steel and concrete. The analytical and design procedures for these platforms are acceptable if in accordance with the ACI 318-71 Code for reinforced concrete, and with the AISC Specifications for structural steel. Of particular interest are the dynamic loads induced on these floors by pool swell during a LOCA. Computer programs used in the design and analysis of containnent interior structure should be described and validated by any of the procedures described in Section II.4.e of Standard Review Plan 3.8.1. 5. Structural Acceptance Criteria With the exception of the divider barrier and ice-condenser elements of the ice-condenser PWR containment, the drywell of the BWR Mark III containment, and the steel linear supports of the reactor coolant system, the structural acceptance criteria for all other interior structures of the containment described in Section I.1 of this plan are acceptable if found in accordance with the following: For each of the loading combinations delineated in the beginning of Section II.3 of this plan, the following defines the allowable limits which constitute the structural acceptance c.'teria: In Combinations for Concrete Internal Structures Limit (a)(i) 1, 2 S(I) (a)(i) la, 2a 1.35 (a)(ii) 1, 2 U(2) (a)(ii) lb. /b U (b) 3, 4, 5, 6 U In Combinations for Steel Internal Structures Limit (a)(i) 1, 2 S (a)(i) la, 2a 1.5 S (a)(ii) 1, 2 Y( ) (a)(ii) lb, 2b Y (b)(i) 3, 4, 5(4) 1.6 S (b)(i) 6(4) 1.7 S (b)(ii) 3, 4, 5, 6 .9 Y Notes (1) S --- FJr concrete structures, S is the required section streng+.h based on the working stress design method and the allowable stresses defined in Section 8.10 of ACI 318-71. 1,R 1-?l 146 249
For structural steel, 5 is the required section strength based on the clastic design methods and the allowable stresses defined in Part 1 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969. The 33% increase in allowable stresses for concrete and steel due to seismic loadings is not pemitted. (2) U --- For concrete structures, U is the section strength required to resist design loads based on the strength design methods described in ACI 318-71. (3) Y --- For structural steel, Y is the section strength required to resist design loads and baseo on plastic design methods described in Part 2 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969. (4) --- For these two combinations, in computing the required section strength, S, the plastic section modulus of steel shapes may be used. For the divider barrier, ice-condenser elements, the drywell, and the linear steel supports of the raactor coolant system, the structural acceptance criterla are acceptable if found in accordance with the following: a. Divider barrig (i) For concrete portions of the divider barrier, the specified limits for stresses and strains are acceptable if found in accordance with Subsection CC-3400 of the ACI-359 Code, but with the following exceptions: CC-3421.1 - The footnote on page 196 should be revised to indicate that the 33-1/3: increase in allowable stresses is permitted only for temperature loads and not for seismic loads. CC-3422.1 - Item (c) should be deleted. CC-3422.2_ - The footnote on page 197 should be deleted. (ii) For steel portions of the divider barrie which resist the design differential pressure and are not backed by concrete, the design should be similar to that of steel containments. Accordingly, the load combinations and stress limits of Section 11.3 of Standard Review Plan 3.8.2 app', b. Ice-condenser Elements for load combinations delineated in Section II.3 of this plan for the ice-condense:' elements, the stress limits are acceptable if found in accordance with the following: 3.8.3-22 146 250
For Combinations: Limit (i) (1),(2) S(I) (ii) (1), (2).. Y(2) C( } (iii) (1). (iv) (3), (4). 1.35 (iv) (5). 1.65 (v) (3),(4),(5). Y (vi) (3),(4),(5). C Notes (1) S --- As defined in " Notes" under first tables in II.5 above. (2) Y --- As defined in " Notes" under first tables in 11.5 above. (3) C --- Where experimental testing is used for verification of the design, C shall be the ultimate load carrying capacity of the member. Size effects and any simil-itude relationship which may exist between the actual component and the test model shall be accounted for in the evaluation of C. c. BWR Mark III Containment Drywell (i) For concrete portions of the drywell, the acceptance criteria of paragraph (a)(i) as described for the divider barrier apply. (ii) For steel portions of the drywell that resist pressure and are not backed by structural concrete, the acceptance criteric of paragraph (a)(ii) as described for the divider barrier apply. (iii) For the lower vent portion of the drywell: - If the main reinforcement of the drywell is carried down between the vent holes and the reinforced concrete section is relied upon fcr structural purposes, the structural acceptance criteria is the same as for (i) above. - If the nain reinforcement of the drywell is terminated above the vent holes and two steel plates lining both faces of the wall are utilized for structural purposes, the acceptance criteria for (ii) above will apply. - If other structural systems are used in the vent region, the acceptance criteria are reviewed on a case-by-case basis, _eactoc Coolant Systen Supports d. R The structural acceptance criteria for the steel linear supports of the reactor coolant system are acceptable if found in accordance with the following: For load combinations delineated in paragraph (d) of Section II.3 of this plan for the reactor coolant system linear supports, th ? following acceptance criteria will apply: (1) If the 'lastic analysis method is used: Combination Allowable Limits (i) Limits of XVII-2000 of Appendix XVII of the ASME Code, Section III. (ii) Limits of F-1370 of Appendix f he ASME Code, Section III. 3.8.3-23
(2) If the limit method of analysis is used: Combination Allowable Limits Limits of XVII-400 of Appendix XVII (i) of the ASME Code, Section III. (ii) c'me en above. 6. Materials, Quality Control, and Special Construction Techniques The specified materials of construction and quality control programs are acceptable if in accordance with the applicable code or standard as indicated in Section I.6 of this plan. Epecial construction techniques, if any, are treated on a case-by-case basis. 7. Testing and In-service Surveillance Requirements Each BWR Mark III cor.tainment drywell should be subjected to a structural proof test. Such a test is acceptable if in accordance with the following: The drywell should be subjected to an acceptance test that increases the drywell internal a. pressure in three or more approximately equal pressure increments from atmospheric pressure to at least the design pressure. The drywell should be depressurized in the same number of increrents. Measurements should be recorded at atmospheric pressure and at each pressure level of the pressurization and depressurization cycles. At each level, the pressure should be held constant for at least one hour before the deflections and strairy are recorded. b. So that the overall deflection pattern can be determined in prototype drywells, radial deflections should be measured at least at three pokts along each of at least t! ree meridians equally spaced around the drywell, including locations with varying stiffness characteristics. Radial deflections should be measured at the lower vent region, at about mid-height and at near the top of the cylindrical uall. The measurement points may be relocated depending on the distribution of stresses and deformations snticipated in each particular design. In prototype drywells only, strain measurements sufficier;t to permit an evaluation of c. strain distribution should be recorded at least at two opposing meridians at the following locations on the wall: (1) at the bottom of the wall, and (2) at mid-height of the wall. These strain measurements should be made at least at three positions within the wall section; one at the center and one each near the inner and outer surfaces. d. In nonprototype drywells, deflection and strain measurements need not be made i; strain levels have been correlated with deflection measurements during the acceptance test of a prototype drywell if measured strains and deflections are within the precefined toler-antes of their predicted response. e. Any reliable system of displacement meters, optical devices, strain gauges, or other suitable apparatus may be used for the measurements. f. If the test pressure drops due to unexpected condition 5 to or below the next lower pressure level, the entire test sequence should be repeated. Significant deviations from the previous test should be recorded and evaluated. g. If any significant modifications or repairs are made to the drywell following and because of 'he initial test, the test should be repeated. gh 2 3.8.3-24
h. A description of the proposed acceptance test and instrumentation require-ments should be included in the preliminary safety analysis report. i. The folicwing information should be submitted prior to the performance of the test-(i) The numerical values of the prefitted responses of the structure which will be neasured. (ii) The tolerances to be pemitted on the predicted responses. (iii) The bases on which the predicted responses and the tolerances thereon were es ta bl i s hed, i. The following infomation should be included in the final test report: (i) A description of the actual test and instrumentation. (ii) A comparison of the test measurenents with the allowable limits (predicted response plus tolerance) for deflections and strains. (iii) An evaluation of the accuracy of the measurements. (iv) An evaluation of any deviations (i.e., test results that exceed the allowable limits), the disposition of the deviations, and the need for corrective measures. (v) A discussion of the calculated safety margin provided by the structure as deduced from the test results. For steel linear supports of the reactor coolant sjstem, testing and in-service surveillarce requirements are acceptable if in accordance with Subsection NF of the ASME Section III Code, Division I. III. REVIEW PROCEDURES 9 The reviewer sc;ects und emphasizes raterial from the review procedures described below, as may be appropriate for a particular case. l. Description of the Internal Structures Af ter each structurc and its functional characteristics are identified, information on similar structures of previously licensed applications is obtair,ed for reference. Such infomation, which is available in safety analysis reports and amendments of licensed plants enables identification of differences for the case under review which require additional scrutiny. New or unique features that have not been used in the past are Of particular interest. The information furnished in the SAR is reviewed for sufficiency in accordance with the " Standard Fomat..." Revision 2. A decision is then made with regard to the sufficiency of the descriptive infomation provided in the SAR. Any additional required information is requested from the applicant at an early stage of the review process. 2. Applicable Codes, Standards, and Specifications The list of codes, standards, guides, and specifications is checked against the list in Section 11.2 of this plan. The reviewer assures himself that the applicable edition and stated effective addenda are utilized. 3. Loads and loading Combinations The reviewer verifies that the loads and load conbinations are as conservative as those specified in Section II.3 of this plan. Any deviations from the acceptance critieria for loads and load corbinations that have not been adequately justified are 146 253 3.8.3-25
identified as unacceptable and transmitted to the applicant for further consideration. 4. Design and Analysis procedures The reviewer familiari es himself with the design and analysis procedures that are generally utilized for the type of structure being reviewed. Since the assurptions made on the expected behavior of the structure and its various elements under loads may be significant, the reviewer determines that they are conservative. The behavior of the structure under various loads arid the manner in which these loads are treated in con-junction with other coexistent loads, are reviewed to establish compliance with pro-cedures delineated in Section II.4 of this plan. 5. Structural Acceptance Criteria The limits on allowable stresses and strains in the concrete, reinforcement, structural steel, etc., are compared with those specified in Section 11.5 of this plan. Where the applicant proposes to exceed sone of these limits for some of the load combinations and at some localized points on the structure, the justification provided to show that the functional integrity of the structure will not be affected is evaluated. If such justification is not acceptable, a request for the required additional justificatica and bases is made. 6. Materials, Quality Control, and Special Construction Techniques The infomation provided on materials, quality control prograns, and special construction techniques, if any, is reviewed and compared with that specified in Section II.6 of this plan. If a new material not used in prior license applications is utilized, the applicant is requested to provide sufficient test and user data to establish the acceptability of such a material. Similarly, any nes quality control programs or construction techniques are reviewed and evaluated to assure that there will be no degradation of '.tructural quality that might affect the structural integrity of the structure. 7. Testing and In-service Surveillance Requirements Procedures for the structural test of the BWR Mark III containment drywell are reviewed and compared with the procedures described in Section II.7 of this plan. Any other proposed testing andin-service surveillance programs are reviewed on a case-by-case basis. IV. EVALUATION FINDINGS The reviewer verifies that sufficient infomation has been provided in accordance with the requirements of this review plan, and concludes that his evaluation is sufficiently complete and adequate to support the following type of conclusive statement to be incli ' " the s fety evaluation report: staff's a "The criteria used in the design, analysis, and construction of the containment internal structures to account for anticipated loadings and postulated conditions that may tse imposed upon the structures during their service lifetime are in confomance with established crite.-ia, and with codes, standards, and specifications acceptable to the Regulatory staff. 3.8.3-26 146 254
"The use of these criteria as defined by applicable codes, standards, and specificatioas; the loads and loading combinations; the design and analysis procedures; the structural acceptance criteria; the naterials, quality control orograms, and special construction techniques; and the testing and in-service surveillance requirement.s provide reasonable assurance that, in the event of earthquakes and various postulatec accidents occurring within the containment, the interior structures will withstand the specified design conditions without impaiment of structural integrity or the perfomance of required safety functions. Conformance with these criteria constitutes an acceptable basis for satisfying in part the requirements of General Design Criteria 2 and 4." V. REFERENCES _ 1. ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF, " Require-nents for Co,nponent Supports," American Society of Mechanical Engineers. 2. ACI 318-1971, " Building Code Requirements for Reinforced Concrete," A erican Concrete Institute (1971). 3. AISC, " Specification for Design, Fabrication, and Erection of Structural Steel for Buildings," Ameri:an Institute of Steel Construction (1969), 4. ASME Boiler and Pressure Vessel Code, Section III, Division 2 (ACI-359), " Proposed Standard Code for Concrete Reactor Vessels and Containments," issued for interim trial use and comment, April 1973, American Society of Mechar.ical Engineers. 5. ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsection NE, " Class MC Components," American Society of Mechanical Engineers. 6. Regulatory Guide 1.55, " Concrete Placement in Category I S:ructures." 7. ANSI N45.2.5, " Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel 'uring the Construction Phase of Nuclear Power Plants," Draf t 3, Revision 1, January 1974, American National Standards Institute. 8. Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2 (in preparation). 9. Regulatory Guide 1.57, " Design Limits and Loadirg Combinations for Metal Primary Reactor Containnent Systen Components." 10. 10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena. 11. 10 CFR Part 50, Appendix A, General Dnign Criterion 4, " Environmental and Missile Desig,1 Bases." 146 255 3.8.3-27}}