ML20210C307
| ML20210C307 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/19/1986 |
| From: | Pandey S, Stilwell T CALSPAN CORP. |
| To: | NRC |
| Shared Package | |
| ML20210C313 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130, TASK-03-07.B, TASK-3-7.B, TASK-RR TER-C5506-435, NUDOCS 8603240177 | |
| Download: ML20210C307 (44) | |
Text
a TECHNICAL EVALUATION REPORT NRC DOCKET NO. 50-219 FRC PROJECT C5506 NRC TAC NO. -- FRC ASSIGNMENT 18 NRC CONTRACT NO. N RC-03-81-130 FRCTASK 435 FINAL SUPPLEMENTARY REPORT
' REVIEW OF LICENSEE RESPONSES TO SEP TOPIC III-7.B, DESIGN CODES, DESIGN CRITERIA, AND IDADING COMBINATIONS j GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION TER-C5506-435 Prepared for I Nuclear Regulatory Commission FRC Group Leader: T. Stilwell Washington, D.C. 20555 NRC Lead Engineer: P. Y. Chen i
March 19, 1986 Revised April 4, 1986 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third par'y's use, or the results of such use, of any information, appa-retus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. ,
Prepared by: Reviewed by: Approved by:
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( ALw W !ain h y p Principal Author f Department Director
, Date: E N 'M Date: S,b b. Date: 3/M/M j FRANKLIN RESEARCH CENTER b ~' Olvi$10N OF ARVlN/ CAL 5 PAN 30tn & sect STeests.pienAcouposa.en esics
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i CONTENTS
- Section Title Page
' f 1 INTRODUCTION . . . . . . . . . . . . . 1 t
i 2 DESIGN CODE CHANGES DESIGNATED SCALE A . . . . . . . 2 2.1 Shear Connectors for Composite Beams . . . . . . 2 2.2 Composite Beams or Girders with Formed Steel Deck . . . 3 2.3 Flange Stress in Hybrid Girders . . . . . . . 3 2.4 Stresses in Unstiffened Compression Elements . . . . 4 2.5. Maximum Load in Riveted or Bolted Tensile Members . . . 4 i 2.6 Shear Load in Coped Beams . . . . . . . . . 6 2.7 Column Web Stiffeners at Frame Joints . . . . . . 7 2.8 Lateral. Support Spacing in Frames . . . . . . . 7 l 1
2.9 Brackets and Corbels . . . . . . . . . . 8
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2.10 Special Provision for Walls . . . . . . . . 8 2.11 Elements Loaded in Shear with No Diagonal Tension . . . 9 2.12 Elements Subject to Temperature Variations. . . . . 9 2.13 Columns with Spliced Reinforcing . . . . . . . 9 1
2.14 Embedments. . . . . . . . . . . . . 9 i
! 2.15 Ductile Response to Impulse Loads . . . . . . . 10 ,
l 2.16 Tangential Shear (Containments) . . . . . . . 10 g
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2.17 Areas of Containment Shell Subject to Peripheral Shear. . 11 2.18 Areas of Containment Shell Subject to Torsion . . . . 11 2.19 Thermal Loads . . . . . . . . . . . . 11 2.20 Areas of Containment Shell Subject to Biaxial Tension . . 12
! 2.21 Brackets and Corbels (on the Containment Shell) .. . . 12 i
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f TER-C5506-435 CONTEN*.'S (Cont. )
Section Title Page
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' 3 REVIEW METHOD AND TABULAR PRESENTATIONS. . . . . . . 13 4 TABULAR
SUMMARY
OF FINDINGS OF LICENSEE COMPLIANCE STATUS CONCERNING IMPLEMENTATION OF SEP TOPIC III-7.B IMPACT OF DESIGN CODE CHANGES . . . . . . . . . 15 5 LOADS AND LOAD COMBINATIONS . . . . . . . . . 31 6
SUMMARY
OF REVIEW FINDINGS . . . . . . . . . . 33 7 CONCLUSIONS AND RECOMMENDATIONS. . . . . . . . . 38 8 REFERENCES . . . . . . . . . . . . . . 39 i
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FOREWORD .
, This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC~ operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by l the NRC.
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- 1. INTRODUCTION z
Current design criteria for nuclear power plant structures contain requirements that were not in effect when older plants were designed and
.- licensed. Consequently, one aspect (designated Tcpic III-7.B) of the implementation of NRC's Systematic Evaluation Program requires licensees to review changes that have occurred in structural design criteria since their plant was built and also to review the loads and load combinations used for design of plant structures by comparing them with the loads and load combinations now specified for current construction. The licensee's objective is to assess the impact that these changes may have on margins of safety of Seismic Category I structures as they were originally perceived and as they would be perceived under current criteria. Upon completion of this work, licensees report their findings to the NRC.
To assist in this review, the NRC provided licensees with plant-specific J Technical Evaluation Reports (TERs) concerning these issues (e.g., Reference l!. The TERs listed design code changes and, on a building-by-building basis, the load and loading combination changes to be addressed in the licensee review. The items listed were ones judged to have the greatest potential to j degrade the originally perceived margins of safety.
In May 1983, under contract NRC-03-81-130, the NRC retained the Franklin Research Center (FRC) to assist in its review of licensee findings. This report describes the review for the Oyster Creek Nuclear Generating Station and sumarizes GPU Nuclear Corporation's compliance status with respect to the i
implementation of SEP Topic III-7.B.
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1 TER-C5506-435 1
- 2. DESIGN CODE CHANGES DESIGNATED SCALE A I
Current structural design codes contain provisions that differ from, or did not appear in, the codes to which older plants were designed and con-j structed. Changes that were judged to have the potential to significantly affect perceived margins of safety have been designated as Scale A.
4 I For reference, changes in ACI-63 and AISC-63 designated Scale A are ,
briefly discussed in this section of the report. Although all such changes were considered in the Topic III-7.B assessment of all plants constructed to these codes, not all appear among the issues the Licensee was requested to address. On a plant-specific basis, some were eliminated as not applicable to the type of construction employed. When this was done, the eliminated code changes were listed (together with the reason they were considered inapplica-ble) in Appendix A of Reference 1, and the Licensee was requested to confirm l
the validity of Appendix A.
2.1 SHEAR CONNECTORS FOR COMPOSITE BEAMS Four major modifications to the 1963 AISC Code [2] related to the type, distribution, and spacing of shear connectors for composite beams occur in the
- l 1980' Code (3). These modifications are:
- a. Permission to use lightweight structural concrete (concrete made with C330 aggregates) in composite designs
- b. Allowance of design for composite action in the negative moment region of continuous beams and provision of design guidance for including the longitudinal reinforcing steel in the negative moment resisting section
- c. Design requirements for the minimum number of shear connectors in regions of concentrated load
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- d. Maximum and minimum spacing requirements in terms of stud diameters.
The first two modifications will not affect old designs because they were not l allowed by the previous code. The new provisions concerning the number of studs in the region.near concentrated loads and the new limits concerning spacing of studs may adversely affect the margin of safety in older designs
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TER-C5506-435 when checked against the new code provisions. These new requirements are of
' special concern in the case of composite beams subject to large concentrated
' loads, such as those associated with extreme environmental or critical
. accident conditions.
2.2 COMPOSITE BEAMS OR GIRDERS WITH FORMED STEEL DECK The 1980 AISC Code [3] contains a new section covering stay-in-place formed steel deck when used in a composite design. These provisions for I
formed steel decking, depending on the rib geometry and the direction of the ribs relative to the beam, may affect the load capacity of the shear studs and
provide for reduction factors, to be applied to the shear stud allowable capacity, which account for the structural irregularity introduced into the composite slab.
Composite beams with formed steel decks that were designed to the previous code could have less conservative margins of safety when compared to present requirements, especially in cases where extreme loadings are to be !
considered. ,
- l 2.3 FI.ANGE STRESS IN HYBRID GIRDERS r
The AISC Code section covering reduction of bending stress in the compression flange was modified in the 1980 Code.
- The original flange stress ceduction formula in the old code was needed
- to account for stress transfer which may occur in ordinary beam webs if the
, compression region should deflect laterally, thereby changing the bending capacity of the cross section. In hybrid girders, the amount of the loss of bending resistance resulting from this phenomenon will vary depending on the relative properties of the web and flange steel. A reduced bending stress formula reflecting this interaction was introduced. In order to keep the-
, formulation relatively simple, the reduced banding strees was made applicable to both flanges of the hybrid member.
1 Where the ratio of web yield stress to flange yield stress is less than ;
O.45 and the ratio of the web area to flange area is low, beams or girders l l
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fabricated from plate where the flange and web steels are different could have l l lower margins of safety under the new code than were thought to exist under
'I older code requirements.
, 2.4 STRESSES IN UNSTIFFENED COMPRESSION ELEMENTS New requirements provide stress roouction factors for unstiffened elements subject to compression with one edge free parallel to the compressive stress.
Previous code provisions allowed the designer to neglect a portion of the area of such elements. The new code requirements provide equations for var-ious elements based on the critical buckling stress for plates. The new j analytical approach is more conservative for the stems of tees and less conservative for all other cases.
Where structural tees are used as main members and the tee stem is in f
1 l compression, the margin of safety for older designs checked under the new code could be significantly less than was thought under prior code requirements.
Since buckling is a non-ductile type failure, these new requirements are of special concern in the case of tee shapes subjected to the extreme environ-mental or critical accident conditions.
I l 2.5 MAXIMUM LOAD IN RIVETED OR BOLTED TENSILE MEMBERS l 1
The 1980 AISC Code [3] introduces codes changes which affect the maximum load permitted in tensile members.
Two interacting code changes are involved in establishing this limit, and the mutual effects of both must be considered in assessing the impact of the
, new code upon the perception of margins of safety in tension members. The two i
provisions involved concern: .
- 1. the tensile area permitted to be used in establishing load carrying capacities
- 2. the allowable stresses to be used in conjunction with these areas.
Both effects are taken into account in ranking this change. The potential magnitude of the mutual effects of the two changes is discussed below.
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The 1980 AISC Specification definition of " Effective Net Area" introduces a reduction coefficient which is to be applied to the traditional d;finition of net area. This essentially changes the design capacity of a tension member
. when compared to older versions of these specifications. First consider only
,I the effect of the critical area used for the design of a tension member as defined in the new code compared to the critical area used for the design of the same member as defined in the old code. Clearly, if all other factors are equal, the new code is more conservative. However, all other factors are not the same. The changes in allowable tensile stress definition (on the gross area and on the effective net area) which were introduced simultaneously with
! the new definition of effective net area modify the above conclusion. In addition, the traditional upper limit en the critical net area of 85% of the gross area (a requirement of the old code) is no longer a requirement of the new code. Both of these changes interact with the new effective net area requirement.
A valid assessment of the effect of these changes is best accomplished by a comparison of the allowable load each code permits in tension members. If If one considers the allowable load on the effective net area, the value based on the new code is a function of three variables: the new reduction coefficient, I
j the net area,* and the ultimate tensile strength of the steel. The allowable load based on the old code is a function of only two variables: the net area and the yield strength of the steel. First, form the load ratio of the allowable load defined by the new code criteria to the allowable load defined
- by the old code criteria. Next, consider the ranges of all of the parameters mentioned above, this ratio will have defined upper and lower limits which are a function of the ratio of the net areas, the new code net area reduction factor, and the ratio of the steel ultimate strength to the yield strength.
l For all the steels allowed under the new code, this load ratio ranges from 1.5 to 0.69. For all the steels allowed under the old code, this load ratio ranges from 1.6 to 0.88. It is apparent that, for those steels with
- In making this comparison, one must be careful to note that the net area is not always the same under the old and new codes.
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I' load ratios less than 1.0, .the new code is less conservative than the old.
The margin of safety of some older designs therefore could be significantly lower when checked against the new code requirements.
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,! 2.6 SHEAR LOAD IN COPED BEAMS r
The 1980 AISC Code [3] introduces additional control over the shear load permitted at beam end connecticas where the top flange has been coped.
Web shear control in older codes did not distinguish between coped and uncoped beams or between shear allowed at connections and over the free span
. (except for requiring reinforcement of thin webs at connections). The shear load allowed was given by:
allowable shear load = 0.4 (yield strength) (gross web section).
The 1980 Code retains this limit, but introduces an additional
- requirement to protect against a failure mode associated with coped beams.
t For coped beams (and similar situations), a portion of the web may sever, failing along the' perimeter of the connection holes. In particular, coped i beam web connections where the fastener holes lie close to the butt end of the beam may be prone to such failures.
l This web " tear out" failure is actually a combination of shear failure through the line of fasteners together with tensile failure across the I
i shortest path to the beam end. The failure surface turns a corner with shear failure along a line trending upward through the holes, combined with tensile failure across a more-or-less horizontal line running out to the beam end.
i The newly introduced shear limit is given as a function of the minimum f net failure surface and the steel ultimate strength. Thus, the new requirements mr / or may not control a coped beam's allowable capacity in shear. Whether or not it does depends on both the connection geometry and the type of steel used.
When this requirement is controlling, coped beams designed by previous rules may be found, if checked against the new criteria, to have significantly smaller margins of safety than previously thought.
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2.7 COLUMN WEB STIFFENERS AT FRAME JOINTS The more recent editions of the AISC code mandate which columns must be stiffened at locations where beams of girders are rigidly attached to the l column flange and also establish requirements for the geometry of such web
- stiffeners. These requirements are introduced to preclude local crippling at
, a such frame joints.
No such guidance was provided by AISC-63 [2]. Older codes (such as AISC-63) left such matters to the designer's discretion. Consequently, there is no assurance that all such columns are adequately stiffened for current
, accident and faulted loadings.
2.8 LATERAL SUPPORT SPACING IN FRAMES (PLASTIC DESIGN METHOD)
The 1980 AISC Code contains changed spacing requirements for lateral supports in portions of members in frames where failure mechanisms are expected to form at ultimate load.
f Members of such frames must not only be capable of developing a plastic
! hinge, but must also be stable enough to sustain moments larger than those computed on an elastic-perfect-plastic theory (because real steels work-harden at strains expected to occur at hinge locations). Previous lateral bracing requirements were developed for a limited range of steels. Research on I high-strength steels has shown that, for certain ranges of slenderness ratio of the compression flange of such frame members, older specification bracing
- requirements were not sufficiently conservative.
I The new specification requirements'make the slenderness ratio limits a function of the steel yield strength and the member curvature (as expressed by the ratio of the lesser bending moment at the ends of the unbraced segment to the plastic moment). ,
The new specifications are more conservative for (1) any segment bent in
, double curvature regardless of its steel specification and (2) very I
i high-strength steel members. The adequacy of frame members bent in single curvature and constructed of steels whose yield strength exceeds 36 ksi should l
' I be examined on a case-by-case basis.
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1 TER-C5506-435 The new requirements may reduce the margins of safety thought to exist in:
j 1. structures designed under the plastic requirements of older codes
- 2. elastically designed structures sized to carry a smaller maximum load than is now required by current accident and faulted load combinations. In this case, plastic logic may have to be invoked to justify the adequacy of exisiting structures. Nonconformance with current bracing requirements may substantially restrict the l capability of frame members to carry code-acceptable overloads.
2.9 BRACKETS AND CORBELS ACI 349-76 [4), Section 11.13 contains design requirements for short brackets and corbels which are considered primary load-carrying members; no comparable requirements are provided in ACI 318-63 (5]. l i
The requirements apply to brackets and corbels having a shear span-to-depth ratio of unity or less. They provide minimum and maximum limits on
, tension and shear reinforcement, limits on ultimate shear stress in concrete,
'i and constraints on member geometry and location of reinforcement. ,
l Brackets and corbels designed under earlier codes may or may not satisfy the newly imposed limits. If they do not, they may be prone to non-ductile failure (which occurs suddenly and without warning) and may exhibit smaller i margins of safety than those currently required.
2.10 SPECIAL PROVISIONS FOR WALLS 1
2.10.1 Shear Walls 1 1
ACI 349-76, Sections 11.15.1 through 11.15.6 specify requirements for i 1
l reinforcing and permissible shear stresses for in-plane shear loads on walls.
The ACI 318-63 Code had no specific requirements for in-plane shear on shear walls. .
2.10.2 Punching Shear ACI 349-76, Section 11.15.7 specifies permissible punching shear stresses for walls. ACI 318-63 had no specific provisions for walls for these stresses. Punching loads are caused by relatively concentrated lateral loads
I TER-C5506-435 on the walls. These loads may be from pipe supports, equipment supports, duct j supports, conduit supports, or any other component producing a lateral load on a wall.
2.11 ELDENTS LOADED IN SHEAR WITH NO DIAGONAL TENSION (SHEAR FRICTION)
The provisions for shear friction given in ACI 349-76 did not exist in ACI 318-63. These provisions specify reinforcing and stress requirements for situations where it is inappropriate to consider shear as a measure of diagonal tension.
2.12 ELEMENTS SUBJECT TO TEMPERATURE. VARIATIONS The ACI 349-76 [4), Appendix A requirements for consideration of temperature variations in concrete were not contained in ACI 318-63. These new provisions require that the effects of temperature gradients and the difference between mean temperature and base temperature during normal operation or accident conditions be considered. The new provisions also i
require that thermal stresses be evaluated considering the stiffness and rigidity of members and the degree of restraint of the structure.
! 2.13 COLIJciNS WITH SPLICED REINFORCING i
The ACI 349-76, Section 7.10.3 requirements for columns with spliced reinforcing did not exist in the ACI 318-63 Code. The ACI 349-76 Code raquires that splices in each face of a column, where the design load stress in the longitudinal bars varies from fy in compression to 1/2 fy in tension,
'be developed to provide at least twice the calculated tension in that face of l the column (splices in combination with unspliced bars can provide this if applicable). This code change requires that a minimum of 1/4 of the yield capacity of the bars in each face of the column be developed by.both spliced and unspliced bars in that face of the column.
2.14 DGEDMENTS Appendix B of ACI 349-80 provides rules for the design of. steel embedments in concrete; the design of embedments is not specifically addressed 1
in ACI 318-63. >
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1 Current requirements of Appendix B are based upon ultimate strength l
design using factored loads. The anchorage design is controlled by the i
ultimate strength of the embedment steel. Ductile failure (i.e., steel yields before concrete fails) is postulated.
- Under the provisions of ACI 318-63, the design of embedments was left to 1 the discretion of the designer. Working stress design methods vere widely used.
I Consequently, it is likely that original embedment designs do not fully I
conform to current criteria. Reviev of such designs to determine the implications with respect to margins of safety is therefore judged a desirable precaution. ;
2.15 DUCTILE RESPONSE TO IMPULSE LOADS Appendix C to ACI 349-76 [41 contains design rules for structures which l
! may be subjected to impulse or impact loads; no such provisions occur in ACI 318-63 [5].
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The rules of Appendix C are intended to foster ductile response (i.e.,
steel yields prior to concrete failure) of nuclear structures if and when they
- experience impulse or impact loads. For structures built to codes not containing such provisions, there is no assurance that sufficient design effort was directed toward proportioning members to provide energy absorbtion capability. Consequently, such structures might be prone to non-ductile, sudden failure should they ever experience postulated accident loadings such as jet impingement, pipe whip, compartment depressurization, or tornado missiles, i
2.16 TANGENTIAL SHEAR (CONTAINMENTS)
Paragraph CC-3421.5, Tangential Shear, of Section III, Division 2 of the ASME Boiler and Pressure Vessel Code [6] addresses the capacity of reinforced concrete containments to carry horizontr.1 shear load. It provides code-acceptable levels of horizontal shear stress that the designer may credit to the concrete. No specific guidance in this matter exists in ACI 318-63.
The provisions associate the allowable concrete stress in horizontal shear with the concrete properties, the manner in which lateral loads are i
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Sufficient diagonal reinforcement (or its demonstrated equivalent) is to be supplied to carry, without excessive strain, shear in excess of that
- permitted in the concrete. A major consideration here is the preservation of the structural integrity of the liner.
In containments constructed to older codes, such matters were left to the discretion of the designer, who may or may not have provided the horizontal shear capacity at controlled strains that the code currently requires.
2.17 AREAS OF CONTAINMENT SHELL SUBJECT TO PERIPHERAL SHEAR Concrete containment design is currently governed by the ASME Boiler and Pressure Vessel Code,Section III, Division 2, 1980 (6). The provisions for peripheral (punching) shear appear in code Section CC-3421.6. These provisions are similar to the ACI 318-63 Code [5] provisions for slabs and footings, except that the allowable punching shear stress in CC-3421.6 includes the effect of shell membrane stresses. For membrane tension, the f
allowable concrete punching shear stress in the ASME Code is less than that allowed by ACI 3'8-63. 1
, 2.18 AREAS OF CONTAINMENT SHELL SUBJECT TO TORSION i
Concrete containment design is currently governed by the ASME Boiler and Pressure Vessel Code,Section III, Division 2, 1980. Section CC-3421.7 of the i code contains provisions for the allowable torsional shear stress in the concrete. Such provisions were not contained in the ACI 318-63 Code. The present allowable torsional shear stress includes the effects of the membrane 3
stresses in the containment shell and is based on a' criterion that limits the i principal membrane tension stress in the concrete. -
2.19 THERMAL LOADS ACI 349-76 Appendix A and ASME B&PV Code,Section III, Div. 2, CC-3440
- contains requirements for consideration of temperature variations in concrete
'that are not contained in ACI 318-63.
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The new provisions require consideration of the effects of thermal J
! gradients and of the effects depending on the mean temperature distribution and the base temperature distribution during normal operation or accident conditions. The new provisions also require that thermal stresses be eval-uated considering the stiffness and rigidity of members and the degree of !
restraint of the structure. i An assessment is to be made of the analytical methods used to determine thermal stresses as compared to current code-acceptable practices, e.g., those l 1
discussed in ACI 349.1R-80 and the commentary to ACI 349R-80. )
i If the methods used for design produce stress results which are signifi- !
l cantly different from those current procedures generate, perceived margins of' l safety could be affected.
I 2.20 AREAS OF CONTAINMENT SHELL SUBJECT TO BIAXIAL TENSION l
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Increased tensile development lengths are required by Section CC-3532.1.2 of Reference 6 for reinforcing steel bars terminated in areas of reinforced concrete containment structures which may experience biaxial tension. For biaxial tension loading, bar development lengths, including both straight
, embedment lengths and equivalent straight length for standard hooks, are l'
required to be increased by 25% over the standard development lengths required n for uniaxial loading. Nominal temperature reinforcement is excluded from these special provisions. ACI 318-63 had no requirements related to this increase in development length.
I 2.21 BRACKETS AND CORBELS (ON THE. CONTAINMENT SHELL)
I The ACI 318-63 Code did not specify requirements for brackets and corbels. Provisions for these components are included in the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Section CC-3421.8. These provisions apply to brackets and corbels having a shear-span-to-depth ratio of .
unity or less. The provisions specify minimum and maximum limits for tension and shear reinforcing, limits on shear stresses, and constraints on the member !
1 geometry and placement of reinforcing within the member.
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- 3. REVIEW METHOD AND TABULAR PRESENTATIONS 3.1 REVIEW DOCUMENTS The information relating to SEP Topic III-7.B which was supplied to the NRC by GPU Nuclear Corporation and made available for this review is contained in the following document and its numerous attachments:
- 1. Peter R. Fiedler, GPU Nuclear Corporation Letter (with Attachments I through XI) to Dennis M. Crutchfield, USNRC
Subject:
Oyrter Creek Nuclear Generating Station - SEP Topic No.
III-7B, Design Codes, Design Criteria, and Load Combinations June 4, 1984 .
3.2 REVIEW PRESENTATION Before undertaking licensee report reviews, FRC prepared tabular forms to be used as a working tool during the review process and also to document the review work and its findings.
f These tables are intended to:
- 1. establish a systematic and comprehensive review procedure
- 2. standardize, as much as possible, the review process for all licensees
- 3. present a relatively corapact overview of each licensee's SEP Topic III-7.B compliance status.
The form sheets summarize key information reported in licensee responses. Certain items (such as descriptions of Scale A code changes, conclusions, and comments) frequently are not adaptable to abbreviated summary. For such items, the form sheets refer the reader either to sections of this TER where the matter is developed more fully or to an extended note list compiled on separate sheets. The note list, although detached from the main table in order to allow a fuller discussion, should be regarded as an integral part of it.
i The form sheet consists of four major columnar sections which:
- 1. identify each Scale A item
- 2. state the action that the licensee took or the logic that the licensee presented to resolve the item
TER-C5506-435 I
- 3. provide an assessment of engineering conclusions that may be 1 reasonably drawn from the evidence provided
- 4. summarize the licensee's compliance status with respect to the item.
Items listed on the tables are design code changes designated Scale A. ,
This list is drawn directly from TER-C5257-320, the previous Technical Evaluation Report on this topic [1].
Form sheets summarizing the review findings concerning the licensee's compliance status with respect to the implementation of SEP Topic III-7.B ;
aspects related to design code changes follow in Section 4. A discussion of the review findings concerning the licensee's compliance status with respect to load and load combination changes is presented in Section 5.
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- 4. TABULAR
SUMMARY
OF REVILM FINDINGS OF LICENSEE COMPLIANCE STATUS CONCERNING IMPLEMENTATION OF SEP TOPIC III-7.B IMPACT OF DESIGN CODE CHANGES Form sheets summarizing the review findings concerning technical aspects
' with respect to the implementation of SEP Topic III-7.B as related to design code changes follow.
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StP Topic Ill-7.8 OtilGs CODES. OtilGI CRift984 AND LOAD COMBINATIDW5 SINWRY OF LICENSEE COMPLIANCE STATUS surer i Or ,
!! ACT OF DCSIGN CODC CHANGES MT ove =
CODt CNANGt CIff0 AS SCAtt A tlCt45ft*5 ACTION TO RESOLVI (VALUAil04 0F tittR5ft'5 ACTION LICtn5tt 5fATUS In it R.5257 -320 P0ftnTIAL CONCtRN 15 $UHICIENT .
Rfffatutte C N S OtsCRIPTIon 0F RtFittutt 15 METHOD (VIDinCE CONClut10ns STATUS WITH FURTMER AND PARAGRAPM ACfl0N CM Cnw VALID An0 REP 0eitD TO A40 COBOEINTS RESPECT TO ma7 I N RWRfD (See Indicated note) APP 90PRIAft? g, (See note)
CURetui M5tm IR R (See note)
A ISC-l_940 At3C-8965 I...I.,- I. .! C.epree.t . t. Att. . 5 ..< . ~.r. . .e r e h,, ,o . <l..no.e - .. .
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4 j NOTES:
i 5' In the following notes, the Licensee's conclusion is presented first,
- followed by the reviewer's comments, if any, in brackets.
)
}' C-1. New width / thickness criterion does not have an adverse effect on i any of the reviewed elements of the reactor and turbine buildings. '
No other Category 1 structures are affected.
i
! C-2. Revised effective area calculational methods for tension members do
! not have an adverse effect on any of the reviewed elements of the :
l reactor and turbine buildings. No other Category 1 structures are 1 affected.
t I New shear criterion for beam ends with top flange coped does not
- C-3.
j have an adverse effect on any of the reviewed elements of the i reactor and turbine buildings.
j C-4. New shear criterion for provision of column web stiffness in moment l'
connections does not have an adverse effect on any of the reviewed I elements of the reactor and turbine buildings.
1}
l C-5. New criterion for spacing of lateral bracing does not affect the
- Oyster Creek plant since it refers to plastic design which was not used in the original design.
! C-6. One corbel which supports a 10-ton monorail in the reactor building was identified for analysis. All requirements of ACI 349-76 are
!] }i met with the exception of providing horisontal steel parallel to
}
the main flexural steel. The provision of the horisontal steel is j t intended to insure the distribution of shear friction steel across i the portion of concrete above the neutral axis. Since calculated j bending and shear stresses in the corbel are very low and the area i j of steel provided is well in excess of that required, this l exception is not deemed to be significant.
C-7. The reactor building shear walls meet referenced ACI 349 requirements for load combinations including CBE and SSE.
\
l C-8. The diesel generator building roof sections were reviewed and found i . to conform with all requirements of ACI 349 for shear as a measure i
of diagonal tension.
- C-9. As part of the original design the reactor building biological J
l shield wall was analysed for toeperature differential of 15'r to 55'r.
2 i
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IER-C5506-435 The spent fuel pool analysis was performed by Burns and Roe in support of Final Design Safety Analysis Report (FDSAR) Amendment No. 78 dated March 18, 1976. GPUN has also performed a spent fuel pool analysis as reported in TDR 231 dated May 29, 1981.
Analysis of the reactor building exterior walls performed by Burns and Roe was submitted to the NRC. This analysis considered a temperature differential of 70'F and the wall sections were found to be adequate.
C-11. Insert plates in the biological shield wall for support of the horizontal truss at elevation 82'-7" were reviewed and found to meet all the requirements of current ACI codes.
Insert plates in the biological shield wall for support of radial platform framing were reviewed and found to be adequate for the applied loads. However, in some cases, the anchors do not have sufficient development length to ensure ductile tensile failure as required by the current ACI Code. This has no impact on margin of safety and does not represent a safety concern.
Insert plates in the biological shield wall for attachment of snubbers were reviewed for compliance with current ACI codes with findings as follows.
j The existing construction does not provide sufficient development
! length to ensure ductile tensile failure as required by the current ACI code. This has no impact on margin of safety and need not be considered significant provided adequate development exists for j calculated tension.
The original design procedure divided shear transfer between i
bearing on the edge of the anchor plate and shear in the anchor bolts. Anchor bolt shear was considered in accordance with ACI 512. The current ACI code prohibits consideration of multiple mechanisms and, in place of the ACI 512 provisions, requires shear friction calculations.
Anchorage calculation in accordance with the current code results I in higher anchor tensile forces than the original design. While I the existing anchors have adequate tensile capacity to resist these forces within allowable stress limits, i'n some cases they do not
' have sufficient embedment to develop load transfer to the concrete within prescribed code stress levels for bond. The overstress is a maximum of 40 percent and occurs for 13 of 35 inserts.
'I In assessing the impact of the calculated overstress, there are specific construction and code conditions which are of significance. First, as indicated in the ACI Code Conenentary, the code formula for development length is based on the tendency of l
l i
?
TER-C5506-435 1
t highly stressed bars to split thin sections of restraining concrete. It is further noted that a single bar embedded in a mass of concrete does not require as great a development length. The
' Ph ysical construction under review is that of single bars embedded in mass concrete which makes the code requirements conservative.
ACI Monograph No. 5 indicates that if splitting and cracking do not occur (e.g., single bars in mass concrete), bond stresses well in excess of those present in the attachments are permissible. This would indicate that the attachments are adequate as they currently exist and would not fail under the applied loading.
Also of considerable importance are the details of the construction which differ from those assumed in the ACI formulation. The bolt is encased in a manufactured wire cage of 1 3/4 inch diameter which may be considered to reinforce the concrete within the bond area and increase the effective diameter of the bolt. No credit for the extra reinforcing was taken in the analysis procedure since the code does not address this condition. We do, however, have manufacturer's unpublished data which indicate that bolt assemblies utilizing this construction have developed bond values in excess of those required for the Oyster Creek bolts.
I I In summary, although the bolts do exhibit calculated bond stress I beyond code allowable values, there are sufficient differences j
between ACI Code construction assumptions and actual conditions that the existing construction is deemed adequate for the applied loads.
C-12. In the TER-CS257-320, some items have been judged to potentially I impact margins of safety regarding the containment vessel as a I result of comparing the original containment design code (ASME BPV Section VIII, 1962) to ASME BPV Section III Division 1, 1980.
Results of our evaluation for these items are summarised below:
l 1. Non-Code Materials The original construction materials extracted from Chicago Bridge & Iron (CB&I) Drawing 9-097 and Form U-1 and U-2 are the following ASTM materials.
- 1) plate: A212-61T Gr "B" F3X to A300 21 forgings: A 350 Gr LFI
- 3) pipe: A 333 Gr "O" ,
- 4) bolts: A 320-L7 Current ASME Code acceptable material designations for the above materials are:
- 1) SA-515 and SA-516 See Comment (i) below
- 2) SA-350 Gr LTI Same as the original material l
- 3) SA-33 Gr 1 to 8 See Comment (11) below
- 4) SA-320 L7 Same as the original material I
l 27 l .
f TER-C5506-435 j Comment (i)
ASTM A212 Gr "B" has the same chemical composition as ASME
) SA-515 Gr 70. ASTM A300 has the identical manganese content to SA-516 Gr 70.
l l Allowable strength values are identical for all the compared materials:
I Tensile strength 70-83 Ksi i Yield point 38 Ksi Therefore, we conclude that the original materials are acceptable.
Comment (ii) i ASTM A333-67 and more recent ASTM and ASME Codes show grades 1 4
to 8, but not "0". In its report, CB&I states that SA-333 Grade 0 as listed in ASME Code,1962 Ed. is similar to Sa-333 l
Gr 1 as listed in ASME Code,1980 Ed. except the following:
i
! ; a. Grade 0 allowed keyhole slot for the Charpy Impact Testing i I with values of 15 ft.-Ibs 4-50*F, while Grade 1 requires V-Notch for same test with values of 13 ft.-lbs 4-50*F.
1
! b. Oyster Creek SA-333 Grade O piping was supplied in compliance with the impact requirements of Burns & Roe Spec. 5-2299-4, which required a V-Wotch test with Impact Testing values of 13 ft.-lbs 0 0*F (not 4-50*F).
! Although the original certified Material Test Reports are not !
! ! retrievable, the Oyster Creek Impact Test requirements of Burns
& Roe Spec S-2299-4 included the same requirements given by the ASME SA-333 Gr 1 material except temperature parameter. This i temperature deviation should not present problems for indoor components which will not see temperatures below O'F.
! C-14. Oyster Creek FDSAR, Vol IV, " primary Containment Design Report" l provides design details which are indicative of design by i analysis. Following are some of the examples:
- i
- i. Part III, para. 2.4.5, Drywell Design .
]
ii. Part III, para. 2.4.20 Absorption Chamber Design
! Design by analysis meets Code requirement.
i j I
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I TER-C5506-435 i ,
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C-18. Oyster Creek Technical Specification requires the drywell to be j
locked and inerted (N2 ) before operation. Before each startup of
- the reactor operation, the drywell personnel-lock is checked to assure that its holding elements are fully engaged in their l'
intended operating position before pressure can be built up in the
! vessel (Plant Operating Procedure 201.1 -- Approach to Critical).
4 The personnel air lock is provided with two gasketed doors in series opening inwards against the drywell pressure.
C-19. No pressure indicating device is visible from the operating area. ,
- l. However, the intent of the Code is met since the door interlock l mechanism could only be activated when the pressure on both sides of the door is equalised.
C-20. Volume IV, page III-15-1 of FDSAR,l" Criteria for Design" provides l the following information:
! i. Containment vessel stresses at penetrations are calculated in accordance with Bulletin 107 of the Weld Research Council,
) dated August, 1975.
l i 11. Containment vessel and penetrations designed per ASME B&PV Code and Code Case 1272N, except allowable stresses are 1.5 SM l (28,875 psi) for accident design condition, and 0.9 Sy (29,700 i t psi) for pipe rupture (based on accident loads plus jet
- l forces). The nozzle calculations for area replacement and i loads on the 7'-10" vent in drywell were performed by CB&I.
)- The nozzle calculations for loads due to connecting piping were performed by Burns & Roe and reviewed by CB&I. i l Additional information on piping / nozzle design calculations is !
available in Oyster Creek Mark I Containment Long-Term Program i Document (MPR-712).
C-21. CB&I has reviewed the design covered by nossle calculations for i i compliance with area replacement reinforcement in accordance with the ASME Code,Section III Subsection NE, 1980 Ed. Also, CB&I has I reviewed the design of the balance of the drywell nossles for '
! compliance with the area replacement requirement of the ASME Code, '
) 1900 Edition. l l
1 .
l C-22. Paragraph NE-3365.1(f) of 1980 ASME Code requires expansion joints j over 6" in diameter to be provided with internal sleeves when the
< fluid velocity exceeds 25 fps for gases or 10 fps for liquid.
j Volume IV, page III-15-1, para. E of the Oyster Creek FDSAR states j that penetrations between the drywell and process lines are
! provided with guard pipes.
1 j - _ . _ _ _ _ - - - _ _ _ _ _ _ .--_ _ .____ _ _ , _ . _ . . _ _ . _ _ . -_-- , _ _ _. --
- g. -
l TER-C5506-435 C-23. Paragraph NE-3365.2 of the 1980 ASME Code establishes design
,. requirements for bellows expansion joints. " Analysis of Torus / Vent
.! Line Bellows Expansion Joints (MPR-727)" provides detailed l' calculations for Oyster Creek bellow design based on Subsection NE (1977 Edition). The analysis shows that the bellows design complies with the new requirements.
e i
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30
Y TER-C5506-435 l
- 5. LOADS AND LOAD COMBINATIONS l
t In reviewing the discrepancies between the load combinations currently
- required for the des'ign of Seismic Category I structures and those actually used for such structures at the Oyster Creek plant--and the effects such discrepancies may have on perceived margins of safety--GPU Nuclear supplemented review of individual structures by adopting a unique approach made possible by circumstances specific to the Oyster Creek plant. These i circumstances are:
3 1. At the Oyster Creek plant, the preponderate number of systems and components which are designated Seismic Category I are housed within structures (or portions of structures) made of concrete.
- 2. In the preliminary review of load combinations contained in I
TER-CS257-320 (1), the structure-specific load combinations selected for Licensee review (and consequently designated either scale A or i
scale Ax) were for every structure either load combination 10 or load I
! combination 13 or both, l
i Because of these circumstances, GPU Nuclear was able to approach the review by considering each of the two load combinations (i.e., combination 10 and combination 13) individually and the plant-wide effects of each by f analysis of sampic structural components (beams, columns, walls, slabs, etc.)
selected from those composing the various Seismic Category I structures. The examples initially selected were those expected to be the most critical. Once a small set of critical examples has been analyzed, the status of other similar, but less critical components could often be established by comparison.
l, Two assumptions are made to simplify these analyses i 1. The principle of linear superposition is employed.
- 2. A " reserve capacity" for structural members is invoked. Loads which constitute the load combination are individually invest'igated. For l each load, the maximum stress on the section which is critical (for l
that particular load) is computed and compared to the ultimate load-
- carrying capacity of that section. The resulting percentage is taken as the fraction of the structural element's load-carrying capability used up by that load.
i
P r TER-C5506-435 FRC offers the following comments concerning these assumptions:
- 1. The two assumptions used in conjunction provide a tractable approach to assessing the adequacy of structural elements under somewhat complex loading combinations.
. 2. Linear interac* ion between individual loads, although this general principle is subject to exceptions (particularly in regimens of non-linear material behavior), does not appear to be unreasonable for analyses made within the context of Topic III-7.B review.
- 3. The concept of structural element " reserve capacity" is conservative because loads entering a given load combination do not necessarily produce maximum stress on the same cross section. Furthermore, even when they do, maximum stresses from different individual loads may r not occur simultaneously or at the same point.
- 4. The approach is essentially a screening process (rather than a true stress computation). Overall, it appears to be conservatively biased and adequate to provide a reasonable basis for making valid judgments concerning the adequacy of structural elements which satisfy the ;
criteria.
l The sampling investigations undertaken by the Licensee individually considered all structures identified in Reference 1 and all loads (including the extreme environmental snow load) and load combinations that TER-CS257-320 suggested be used as test cases for demonstration of compliance with the intent of SEP objectives. However, the status of masonry walls is not here considered, but is addressed in an independent investigation.
The Licensee concludes that, in general, Oyster Creek Seismic Category I structures are capable of withstanding currently postulated loading combinations. For the exceptions found, the Licensee concludes that these do not expose public safety to sources of risk precluded by current design requirements.
e t
d t
l TER-C5506-435
- 6.
SUMMARY
OF REVIEW FINDINGS 4 i
l This section discusses items not fully addressed in the Licensee's
- submittal. There ar five such items, and they may be considered to fall into
- j. two broad categories:
I jl 1. Items for which the responses, although partial, may be considered
- ' acceptable. There are two of these - the first two of the individual discussions below. A basis for acceptance of the Licensee's conclusion is provided.
i,
- 2. Items where the Licensee's response itself indicates that i investigation is incomplete. A course of action for completion is suggested.
6.1 D(BEDMENTS Appendix B of ACI 349-80 provides rules for the design of steel i embedments in concrete; the design of embedments is not specifically addressed in ACI 318-63 but instead is left to the discretion of the designer, jc Consequently, it is likely that original embedment designs do not fully comply 4
with current criteria. The Licensee was requested to review the implications with respect to margins of safety at the Oyster Creek plant.
4
- The Licensee's response was as follows:
i r " Insert plates in the biological shield wall for support of the i horizontal truss at elevation 82'-7" were reviewed and found to meet all the requirements of current ACI codes.
- Insert plates in the biological shield wall for support of radial 4
platform framing were reviewed and found to be adequate for the applied i loads. However, in some cases, the anchors do not have sufficient i development length to ensure ductile tensile failure as required by the current ACI code. This has no impact on margin of safety and does not represent a safety concern.
i l Insert plates in the biological shield for attachment of snubbers were l' reviewed for compliance with current ACI codes with findings as follows.
)! '
The existing construction does not provide sufficient development length to ensure ductile tensile failure as required by the current ACI code.
j This has no impact on margin of safety and need not be considered a
- significant provided adequate development exists for calculated tension.
) The original design procedure divided shear transfer between bearing on j
the edge of the anchor plate and shear in the anchor bolts. Anchor bolt L . . . - . ----.----..-.---..-.------.-.--_--.-_-J. . - .
e s TER-C5506-435 shear was considered in accordance with ACI 512. The current ACI code l prohibits consideration of multiple mechanisms and, in place of the ACI 512 provisions, requires shear friction calculations.
Anchorage calculation in accordance with the current code results in higher anchor tensile forces than the original design. While the existing anchors have adequate tensile capacity to resist these forces within allowable stress limits, in some cases they do not have sufficient
- embedment to develop load transfer to the concrete within presc' ribed code j stress levels for bond. The overstress is a maximum of 40 percent and occurs for 13 of 35 inserts.
In assessing the impact of the calculated overstress, there are specific construction and code conditions which are of significance. First, as indicated in the ACI Code Conunentary, the code formula for development j length is based on the te'ndency of highly stressed bars to split thin sections of restraining concrete; It is further noted that a single bar l'
embedded in a mass of concrete does not require as great a development length. The physical construction under review is that of single bars a embedded in mass concrete which makes the code requirements conservative. ACI Monograph No. 5 indicates that if splitting and
, cracking do not occur (e.g., single bars in mass concrete), bond stresses well in excess of those present in the attachments are permissible. This
' would indicate that the attachments are adequate as they currently exist
- and would not fail under the applied loading.
l' '
Also of considerable importance are the details of the construction which l
j differ from those assumed in the ACI formulation. The bolt is encased in 1 a manufactured wire cage of 1 3/4 inch diameter which may be considered to reinforce the concrete within the bond area and increase the effective diameter of the bolt. No credit for the extra reinforcing was taken in the analysis procedure since the code does not address this condition.
! We do, however, have manufacturer's unpublished data which indicate that
! bolt assemblies utilizing this construction have developed bond values in excess of those required for the Oyster Creek bolts.
i i In summary, although the bolts do exhibit calculated bond stress beyond ,
I code allowable values, there are sufficient differences between ACI code construction assumptions and actual conditions that the existing construction is deemed adequate for the applied loads."
1 ,
FRC recommends this issue be considered resolved based on the following '
considerations:
e
- 1. The Licensee's response reflects an understanding of the issue.
j
- 2. On a sample basis, the Licensee has undertaken an investigation of i i the issue and reported findings both favorable and unfavorable. I i
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h - .
TER-C5506-435 f
- 3. For circumstances where criteria of the current code are not met, the Licensee has provided rational grounds for the belief that these components can be credited with additional strength.
i
- 4. The Licensee has documented an engineering judgment of structural
, adequacy.
i .
6.2 DESIGN BY FORMULA The design computations for the Oyster Creek containment vessel were
! ' carried out by hand. Use was made of design formulas provided by ASE Section j VIII, 1962 - the then current code. Identical formulas are retained in the now current code under NE-3300, but their application is restricted by NE-3131 which reads:
"NE-3131 General Requirements (a) The containment vessel design shall be such that rules of NE-3200 are l satisfied. However, in the absence of substantiall mechanical or j thermal loads other than pressure, the rules of NE-3300 may be used in lieu of the rules of NE-3200 for those configurations which are i explicitly treated in NE-3300. ;
1For the purpose of this provision, substantial loads are defined as I
those which cumulatively result in stresses which exceed 10% of the j [
primary stresses induced by the Design Pressure, such stresses being defined as maximum principal stresses."
f.
In contrast to NE-3300 (design by formula),'NE-3200 (design by analysis) introduces the new requirements brought into the ASME Code at the time that i
! Section III was written. These requirements were intended to provide guidance
, 1 l
for the assessment of the structural integrity of nuclear components when l modern methods of structural analysis (i.e., computerised techniques) are used.
i It is very likely that treatment of the contal'nment vessel under the
- I rules of NE-3200 with modern analytical tools will produce different margins j of safety than those resulting from the use of formula methods (which are not I' endorsed by the current code for components having substantial mechancial and 1
thermal loadings). The Licensee was requested to examine the impact such treatment might have upon the perception of the margins of safety in the Oyster Creek containment vessel.
i
TER-C5506-435 1
The Licensee's response paraphrased below does not appear to address evaluation of comparative margins of safety.
" Oyster Creek FDSAR, Vol. IV, " Primary Containment Design Report" i
provides design details which are indicative of design by analysis.
Following are some of the examples:
- i. Part III, para. 2.4.5, Drywell Design I li. Part III, para. 2.4.20 Absorption Chamber Design .
Design by a-alysis meets Code requirements."
! The referenced calculations include consideration of mechanical and thermal loadings, but used in conjunction with a formula approach. Although the response does not address potential differences that may exist in the perception of margins of safety, it does indicate that consideration is given to all loading effects. As such, the response could qualify under the criterion that the plant design meets the intent of modern codes.
1' 6.3 CONCRETE SUBJECT TO HIGH TEMPERATURE AND/OR THERMAL TRANSIENTS I
Appendix A of ACI-349-76 contains design requirements for concrete subject j i to high temperatures and thermal transients; no corresponding requirements
- were present in the provisions to which Oyster Creek structures were designed.
I. ! The Licensee's response indicates that the impact of these new provisions I
- was investigated. The Licensee found all structures other than the drywell i
adequate to meet current criteria. The Licensee stated that recent estimates I of drywell thermal conditions will necessitate reconsideration of this structure, f An investigation of the consequences of the latest estimates of drywell 1
temperature should be carried out.
6.4 REINFORCEMENT OF OPENINGS
{ NE 3334.1 and NE 3334.2 of Section III of the ASME Code provide rules for emplacement of reinforcement around openings: the original design was in accordance with UG-40 of Section VIII of the ASME Code.
! l l s j l .__ -. - - _. . - . - - - - - . .. . - . -. _- ---
a t
TER-C5506-435 The Licensee has provided a thorough study of the reinforcement existing at openings. The results of this investigation show that the disposition of reinforcement is in accordance with current criteria for some openings but not
, for others.
l For openings shown not to be in compliance with current criteria (or for I a representative sample of such openings), the Licensee should supply a conse-1 quences analysis or other justification of their acceptability.
6.5 REINFORCEMENT OF OPENINGS SUBJIET TO CYCLIC LOADINGS NE-3331(b) of Section III of the ASME Code imposes design requirements for openings subject to cyclic loading; UG-36 of Section VIII of the ASME Code does not impose this requirement.
No cyclic analysis is required for openings meeting the exclusion rules of Subsection NE. The Licensee's response expresses an opinion that all openings may satisfy the exclusion rules.
No documentation of this opinion is offered, nor is it reported that a review was made. Proper documentation should be provided.
l 9
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TER-C5506-435
- 7. CONCLUSIONS AND RECOMMENDATIONS s
GPU Nuclear Corporation has submitted a responsive and substantive report
' addressing Topic III-7.B issues as they relate to the Oyster Creek Nuclear r
~
Power Station. Of concern are the effects that changes in design codes and postulated load and load combinations, introduced since plant construction, may have on perceived margins of safety in Seismic Category I structures at the Oyster Creek plant.
GPU findings indicate that, although some Seismic Category I structures are not in total compliance with all design requirements imposed by current codes, the exceptions are not of a character that significantly degrades overall structural integrity below that provided by construction to current standards.
Section 6 individually addresses issues which are not considered fully f resolved in the Licensee's response and suggests a basis for resolution for each.
With respect to current loads and load combinations, GPU submitted evidence based on screening analyses of critical Seismic Category I
- components. Elements in the reactor building (steel structure and concrete structure), the turbine building (steel frame and concrete structure), the diesel generator building, spent fuel pool, and intake structure were examined. Excluded from these investigations were masonry walls (examined in study). Based on these results, GPU concludes that the examined a separate elements either have adequate margins of safety under current loads and load combinations or, if not, do not expose public safety to risks which current criteria are designed to preclude.
E l
l 1
l j
+ ** a TER-C5506-435 :
l l 8. REFERENCES j l
- 1. Franklin Research Center, Technical Evaluation Report Design Codes, Design Criteria, and Loading Combinations (SEP Topic III.7.B), Jersey Central Power and Light Company, Oyster Creek Nuclear Generating Station Plant Unit 1, TER-C5257-320
- 2. " Specification for Design, Fabrication, and Erection of Structural Steel l for Buildings," Sixth Edition American Institute of Steel Construction, Inc. ;
New York, NY l 1963 ,
1 l'
- 3. " Specification for Design, Fabrication, and Erection of Structural Steel for Buildings," Eighth Edition American Institute of Steel Construction, Inc.
)
New York, NY 1 1980 j
- 4. " Code Requirements for Nuclear Safety Related Concrete Structures" (ACI 349-76) l American Concrete Institute, Detroit, MI j
i 5. " Building Code Requirements for Reinforced Concrete" (ACI 318-63)
American Concrete Institute, Detroit, MI
- 6. ASME Boiler and Pressure Vessel Code,Section III, Division 2 l
" Code for Concrete Reactor Vessels and Containments" New York, NY 1980 l
- 7. Standard Review Plan NRC, Rev. 1, July 1981 !
NUREG-0800 (formerly NUREG-75/087)
- 8. Peter R. Fiedler, GPU Nuclear Corporation Letter with Attachments I through XI to Dennis M. Crutchfield, USNRC
Subject:
Oyster Creek Nuclear Generating Station - SEP Topic No.
' III-7.B, Design Codes, Design Criteria, and Load Combinations June 4, 1984 4
l i -