ML19350D262
| ML19350D262 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/09/1981 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | Abel J COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8104150068 | |
| Download: ML19350D262 (24) | |
Text
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N(C ( f f" Mr. J. S. Abel 2-Director of Nuclear Licensing GEN 131981h -8
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Dear Mr. Abel:
Meeting Review to Discuss and Resolve Open Issues Related to
Subject:
Analyses of Mechanical Systems and Piping The Mechanical Engineering Branch and its contractor, Argonne National Laboratory have completed their review of the Byron and Braidwood FSAR through Amend We have chosen not to forward additional questions but to proceed directly to a The SER identifies and discusses open issues pertaining to draft SER input.
Issues or methods of analysis of the adequacy of piping and mechanical systems.
questions pertaining to the adequacy of seismic design as a result of the st review of Commonwealth's seismic margins assessment will be treated separatel We propose that a 3 to 5 day meeting be arranged in Chicago during the we The purpose of this extended meeting will be to resolve all open Therefore, NSSS, AE and utility representa-May 11, 1981.
issues identified in the draft SER.
tives involved should be available for discussion of technical issues during the Also, cognizant management personnel should be present at the meeting meeting.
to facilitate resolution of issues.
We request that you prepare a draft agenda and provide a list of anticipated attendees so that we can mutually agree on topics of discussion and attendees at the meeting well in advance.
Sincerely,
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i B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing cc: See next page
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Mr. J. S Abel Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690 Mr. Edward R. Crass ccs:
Mr. William Kortier Nuclear Safeguards and Licensing Division Atomic Power Distribution Westinghouse Electric Corporation Sargent & Lundy Engineers 55 East Monroe Street P. O. Box 355 Chicago, Illinois 60603 Pittsburgh, Pennsylvania 15230 Nuclear Regulatory Commission, Region III Paul M. Murphy, Esq.
Office of Inspection and Enforcement Isham, Lincoln & Beale 799 Roosevelt Road One First National Plaza Glen Ellyn, Illinois 60137 42nd Floor Chicago, Illinois 60603 Myron Cherry, Esq.
Cherry, Flynn and Kanter Mrs. Phillip B. Johnson 1 IBM Plaza, Suite 4501 1907 Stratford Lane Rockford, Illinois 61107 Chicago, Illinois 60611 Marshall E. Miller, Esq., Chairman Ms. Julianne Mahler Atomic Safety and Licensing Center for Governmental Studies Board Panel Northern Illinois University U. S. Nuclear Regulatory Commission DeKalb, Illinois 60115 Washington, D. C.
20555 C. Allen Back, Esq.
Dr. A. Dixon Callihan P. O. Box 342
_ Union Carbide Corporation Urbanan, Illinois 61820 P. O. Box Y Oak Ridge, Tennessee 37830 Thomas J. Gordon, Esq.
Waaler, Evans & Gordon Dr. Richard F. Cole 2503 S. Neil Atomic Safety and Licensing Champaign, Illinois 61820 Eoard Panel U. S. Nuclear Regulatory Commission Ms. Bridget Little Rorem Washington, D. C.
20555 Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935 Kenneth F. Levin, Esq.
J Beatty, Levin, Holland, Basofin & Sarsany 11 South LaSalle Street l
Suite 2200 Chicago, Illinois 60603 f
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BYRON /BRAIDWOOD DRAFT SER 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping Our review under Standard Review Plan Section 3.6.2 covered the implementation of pipe break and pipe crack criteria. The specific areas investigated were the location of pipe breaks and pipe cracks, and the resulting jet thrust and impingement forces which could damage nearby safety related components. The implementation of special criteria and features dealing with in-service inspection and the use of pipe whip restraints to protect against the postulated breaks were also investigated.
The following is a discussion of open issues which are based on our review thus far, and which must be resolved:
In section 3.6.2, the applicant uses WCAP-8082A as a reference to justify the selectic, of pipe breaks for loss of coolant accident analysis. A review of WCAP-8082A indicates that a set of normal and upset transients are used in the fatigue analysis which essentially determined pipe break locations. Some of these transients are different than those identified in section 3.9.1 of the FSAR. Justification should be provided for the applicability of the number and location of break locations postulated as indicated in WCAP-8082A when determined using the transients specific for the Byron /Braidwood plant reactor coolant loops.
Additionally, our position is that WCAP-8082A may only be referenced in applications specifying Westinghouse reactor coolant systems without loop isolation valves, and provided that the piping geometry, material,
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. and loads used in WCAP-8082A envelope those of the specific plants.
The applicant should, therefore, provide justification for use of the WCAP-8082A break locations with the addition of loop isolation valves and the specific loads including site related OBE loads of the Byron and Braidwood plants.
The applicant should describe the interface responsibility between Westinghouse and the BOP supplier relative to the design for component supports and other portions of the reactor coolant pressure boundary which is not part of the coolant loop as defined by WCAP-8082A, that insures that component displacement at support interfaces and at pipe break locations are within limits specified in WCAP-8082A.
On page 3.6-14 of the FSAR, the applicant states that the SATANIV Code has been used for two phase flow hydrodynamic blowdown analysis SATANIY Code is identified as WCAP-7263 which is of piping systems.
It is not precisely clear in what respect an unapproved topical report.
the SATANIV Code is different from the Multiflex Code (WC The Multiflex Code is used by the applicant for the two phase flow hydrodynamic analysis of reactor internals in Chapter 3.9 of the l
The applicant should FSAR, and is an approved topical report.
provide justification for using the SATANIV Code.
Subject to resolution of the above described open issues, our findings are as follows:
l The applicant has proposed criteria for detennining the location, type and effects of postulated pipe breaks in high energy piping
. systems and postulated pipe cracks in moderate energy piping system The applicant has used the effects resulting from these postulated pipe failures to evaluate the design of systems, components, and structures necessary to safely shut the plant down and to mitigate The applicant has the effect of these postulated piping failures.
stated that pipe whip restraints, jet impingement barriers, and other such devices will be used to mitigate the effects of these postulated piping failures.
We have reviewed these criteria and have concluded that they provid for a spectrum of postulated pipe breaks and pipe cracks which in the most likely locations for piping failures, and that the types of We find that the breaks and their effects are conservatively assumed.
methods used to design the pipe whip restraints provide adequate as that they will function properly in the event of a postulated piping We further conclude that the use of the applicant's proposed failure.
pipe failure criteria in designing the systems, components, and necessary to safely shut the plant down and to mitigate the conseque of these postulated piping failures provides reasonable assuran ability to perform their safety function following a failure in The applicant's criteria high or maderate energy piping systems.
comply with Standard Review Plan Section 3.6.2 and satisfy the applicable portions of General Design Criterion 4.
3.7.3 Seismic Subsystem Analysis Our reviews under Standard Review Plan Section 3.7.
f applicants dynamic analysis of all seismic Category I piping sys i
The dynamic analysis was performed by I
structures, and components.
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4 conversion of the real structure or component into a system of masses, springs, and dashpots which would mathematically simulate Each real displacement and motion of the structure or component.
pipeline was idealized as a system of lumped masses connected by elastic members. The stiffness matrix of the piping system was then calculated using the elastic properties of the pipe which included the effects of torsion, bending, shear and axial deforma-The tion as well as changes in stiffness due to curved members.
dynamic analysis was then performed using a mcdal analysis to deter-mine frequencies and mode shapes plus either the response spectrum method of analysis, or direct integration of the uncoupled modal equations or coupled differential equations of motion.
When piping systems were anchored or supported at different elevations or points of excitation, the response spectrum analysis was performed using the enveloped response spectra of the points of support or excitation.
The differential seismic movements of interconnected supports for seismic Category I piping systems were obtained from a time history The resulting displacements analysis of the supporting structure.
were assumed to cause moments and forces which when applied to the piping system resulted in self limiting stresses much like those A static analysis was, therefore, justified due to thermal expansion.
for use in determining these secondary stress distributions.
During plant life, five OBE's with ten stress cycles each were considered for piping subsystem fatigue analysis.
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. The applicants procedures for seismic subsystem analysis of Category I structures, piping, and components has been reviewed by us and found to be acceptable subject to resolution of the following open issues.
Though the procedures used for seismic analyses of piping systems, structures and components are acceptable, there is an outstanding issue regarding the adequacy of seismic input applied at the foundation of Category I structures.
It is the staff's position that seismic margins of all structures and mechanical and electrical equipment necessary to accomplish safe shutdown will have to be determined to demonstrate that they will be able to accommodate the seismic input acceptable under the current Standard Review Plan. The applicant is currently preparing a plan fcr assessing seismic safety margins.
Procedures or methods of analysis proposed for such reanalysis will be reviewed and the SER revised accordingly. This review will be integrated with the structural and seismological staff review.
The FSAR in section 3.7.3.4 provides a basis for se tection of l
frequencies based on reasonable judgement of the relationship between the equipment and its supporting structure. For the operating license review, however, the natural frequencies should be prpvided for major seismic Category I structures. This would allow for verification that fundamental frequencies of seismic Category I equipment is less than 1/2 or more than twice the dominant frequency of the supporting structure.
The applicant should also provide a more detailed discussion for modeling and placement of pipe supports and snubbers.
. The FSAR in subsection 3.7.3.5 acknowledges the use of static load method but does not elaborate on the extent of its use. The applicant should be asked to identify the equipment for which the fundamental natural period is not known and for which equivalent static load method of analysis is used.
The applicants position with regard to modal superposition in seismic analysis is not in accordance with Regulatory Guide 1.92. The applicants response to NRC question 110.33 part (4) is not satis-factory and further justification must be provided for use of equation 3.7-36 in lieu of equation (5) of Regulatory Guide 1.92.
With respect to buried seismic Category I piping systems and tunnels, the FSAR should outline a complete detailed procedure used for the seismic analysis. References 10,11, and 12 identified in section 3.7.3 are not sufficient to accept the procedure used.
Indentifica-tion of specific buried structures and components in the Byron plant is also required.
We will accept damping values used in the seismic analysis of reactor internals subject to the resolution of our concerns to the applicant in a meeting to be scheduled.
3.9 Mechanical Systems and Components Our review under Standard Review Plan Section 3.9.1 through 3.9.6 pertained to the structural integrity and operability of various safety-related mechanical components in the plant. This review included the general methot of analysis and modelling techniques which were used as well as the treat-
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Our review was not limited to ASME Code Components and Supports, but extended to other components such as control drive mechanisms, reactor internals and any safety related piping designed to industry standards other than the ASME Code. Our review was directed at providing assurance that all mechanical systems and components would be capable of performing their safety-related functions snder all combinations of normal operating conditions, system operating transients, and postulated accidents.
3.9.1 Special Topics for Mechanical Components In this section, we reviewed the treatment of transients as applied to fatigue analysis of ASME Code Class 1 components. Our review also covered the computer programs and any experimental or inelastic stress analysis which were used by the applicant in the design of seismic category I mechanical components.
In addition, we have contracted the Argonne National Laboratory to perform an independent analysis for the 16" loop 3 feedwater line. This analysis will provide a check on the applicants ability to correctly model and analyze his piping systems to ASME Code requirements and limits. We will report the results of this independent analysis as a supplement to this Safety Evaluation Report.
It is requested that the applicant provide a technical basis for 'Ris selection of number of masses in seismic modelling of piping and components.
In addition, the use of simplified elastic-plastic methods (NB-3228.3 and NB-3653.6) in the design of ASME Code Class 1 components and piping should be specified. A tabular sumary should be provided of the stress ranges Sm, m, n, and Ke for each locaticn for which primary plus secondary stress intensity range is either just below or just above a value of 35m.
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. In subsection 3.9.1.4.7, it is not clear why the applicant follows the rules of F-1323.1(a) for elastic system analysis and component inelastic analysis.
These rules are limited to elastic system analysis and component elastic analysis. Furthennore, the primary stress limit proposed is applicable only when inelastic system analysis and component elastic analysis is used. The specific stress limit used for primary membrane plus bending is recommended by !
ASME Section III Appendix F for core support structures only. The applicant should justify why this limit is used for general components and component supports.
When an elastic system analysis is used, Appendix F does not allow inelastic component analysis except when collapse load methods are used. When this method is used, the deformation and displacement limits which are proposed by the applicant should be specified.
The SSE seismic analysis of the reactor coolant loop and supports on page 3.9-18 of the FSAR specifies 4% critical damping. This value appears to i
be greater than that allowed by Regulatory Guide 1.61 for reactor coolant loop piping and must therefore be justified.
Subject to resolution of these open issues, our findings are as follows.
The aethods of analysis that the applicant has employed in the design of all l
seismic Category I ASME Code Class 1, 2 and 3 components, component supports, reactor internals, and other non code items are in conformance with Standard Review Plan Section 3.9.1.
Those components which form part of the reactor coolant pressure boundary have been designed and constructed so as to have an extremely low probability of leakage or failure as required by GDC 14. This is assured by design with sufficient margin such that design limits are not
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- 15. The criteria used in defining the applicable transients and the computer codes and analytical methods used in the analyses provide assurance that the calculation of stresses, strains, and displacements for the above noted mechanical systems and components conform with the current state-of-the-art and are adequate for the design of these items.
Seismic Margins and knpact of Revised Seismic Input Motion (reserved)
Measures Taken to Assure Safe Shutdown (reserved) 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment In this section, we have reviewed the criteria, test procedures and dynamic analyses which were emloyed by the applicant to insure the structural integrity and operabiltiy of piping systems, mechanical equipment, reactor internals and their supports under vibratory loadings. These same criteria procedures and analyses would be acceptable in qualifying system, components and equipment under seismic input acceptable under the current Standard Review Plan. This review covered the following specific areas:
3.9.2.1 The pre-operational startup and test program will include testing of ASME Code Class 1, 2 and 3 piping systems for vibration, thermal expansion and dynamic effects. The purpose of these tests is to confirm that these piping systems, restraints, and supports are adequately designed to withstand the flow induced dynamic loadings and that thermal motion is not restrained. The m
. applicants program is found to be satisfactory except for the following open issue. Response to Question 110.37 is not completely satisfactory with regard to sr.ubbers. Due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that snubber operability be insured through a program of pre-service examination and pre-operational testing as reauested by Question 110.63. This program is to be documented for all safety related hydraulic and mechanical snubbers and should be described in chapter 14 of the FSAR.
Subject to resolution of this item, our findings are as follows. The vibration, thermal expansion, and dynamic effects pre-operational test program which will be conducted on specified high and moderate energy piping and all associated systems, restraints, and supports, is an acceptable program. The tests provide adequate assurance that the piping and restraints of the systems have been designed and constructed to withstand the dynamic effects due to valve closure, pump trips, and other operating modes associated with design basis flow conditions.
In additon, the tests provide assurance that adequate clearance and free movement of snubbers exist for unrestrained motion of piping during normal plant heat up and cooldown. The test program, therefore, complies with Standard Review Plan Seciton 3.9.2 and satisfies tue applicable requirements of General Design Criteria 14 and 15.
3.9.2.2 The seismic qualification testing of safety related mechanical eq'uipment assures its ability to function during the safe shutdown earthquake or other dynamic events.
In those cases where no specific action of the component is required, only its structural integrity needs to be assured.
In nuny cases, however, the mechanical component must perform some act
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of physical motion in order to safely shut the plant down. These components are termed " active" components, and their operability as well as structural integrity must be demonstrated. The operability of active pumps and valves is discussed in Section 3.9.3 of this Safety Evaluation Report. Our review of the dynamic qualification of mechanical equipment is not complete and will be discussed in a supplement to this Report pending the applicants seismic safety margins assessment.
3.9.2.3 The applicant has committed to test the reactor internals in accordance with the provision of Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for Reactor Internals During Pre-operational and Intial Startup Testing", Revision 2, for non-prototype Category I plants. The purpose of this testing is to demonstrate that flow induced vibration similar to that expected during normal operation will not result in vibration of significant magnitude or structural damage. During hot functional testing, the reactor internals will be subjected to greater than 240 fuli flow hours which will 6
assure that a minimum of 10 cycles of vibration will be experienced by the main structural elements of the reactor internals. Pre and Post hot functional inspecton results, with the vessel head removed, will confirm l
that the internals are structurally adequate and sound for operation.
Acceptance standards are the same as required in the shop by original l
design drawings and specifications. The conduct of this pre-operational vibration test is in conformance with the provisions of Regulat;ory Guide 1.20 and Standard Review Plan Section 3.9.2.
This program satisfies the applicable requirements of General Design Criteria 1 and 4 by assuring the inservice integrity.of the reactor internals which is required for proper positioning of fuel assemblies and control rods.and therefore, permits the safe operation and shutdown of the reactor.
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. 3.9.2.4 The applicant has performed dynamic system analyses to confirm the strucutral adequacy of the' reactor internals and unbroken loops of the reactor coolant pressure boundary, including supports, for the combined loads due to a simultaneous loss-of-coolant accident and safe shutdown earthquake. Seismic safety margins and the inpact of the revised seismic input on assuring a safe shutdown are currently being evaluated by the applicant. Our review in this area can not be completed until this study is complete and the applicant submits the information requested in Question 110.62.
Subject to resolution of the above open issues, our findings are as follows.
The dynamic system analysis performed by the applicant provides an acceptable basis for confirming the structural design adequacy of the reactor internals and unbroken piping loops to withstand the combined dynamic loads of a postulated loss of coolant accident (LOCA) and the safe shutdown earthquake (SSE). The analysis provides adequate assurance that the combined stresses ar.d strains in the components of the reactor coolant system and reactor internals do not exceed the allowable stress and strain limits for the materials of construction, and that the resulting deflections or displacements at any structural elements of the reactor internals will not distort the reactor internals geometry to the e.itent that core cooling may be impaired.
The methods used for component analysis have been found to be compatible with those used for the systems analysis. The proposed combina,tions of component and system analyses are, therefore, acceptable. The assurance l
of structural integrity of the reactor internals under combined LOCA and SSE conditions provides added confidence that the design will withstand a spectrum of lesser pipe breaks and seismic events. Accom-olishment of the dynainic system analysis constitutes an acceptable basis for complying with Standard Reveiw Plan Section 3.9.2 and for satisfying the applicable requirements of General Design Criteria 2 and 4.
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3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures Our review under Standard Review Plan Section 3.9.3 is concerned with the structural integrity and operability of pressure retaining components, their supports, and core support structures which are designed in accordance with the rules of the American Society of Mechanical Engineers ( ASME) Boiler and Pressure Vessel Code,Section III, or earlier industry Standards. The review is divided intr, four parts, each of which is briefly described below.
3.9.3.1 Loading Combinations, Design Transients, and Stress Limits The applicant has, in general, described acceptable load combina-tions, design transients, and stress limits for all ASME Code Class 1, 2, and 3 components, component supports, and core support structures but did not discuss or specify the analytical methods used for combining these loads.
In order that we can complete our review, it is requested that the applicant provide in the FSAR a discussion of load combination methods and a sununary of the maximum loads, stresses, deformations, and usage factors for all ASME Code Class 1 components and supports. Total stress values that differ from allowable limits by less than 10% should be listed and the contribution to loading, such as seismic, deadweight, pressure, and thermal should be identified.
If the SRSS method was used for combination of OBE and SRV loads it should be justified in accordance with NUREG-0484 Rev.1.
The applicant's response to question 110.40 is not satisfactory.
It is requested that a committment be made in section 3.9.3.1 of the FSAR to address the functional capability of essential systems and equipment when service Limit B limits are exceeded for Emergency and Faulted Conditions.
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14 It is also requested that the applicant specifically address reactor internals in section 3.9.3.1 of the FSAR as requested by Question 110.16 and 110.41.
The Applicant's response to Question 110.16 is not acceptable. Prior to issuance of the Operating License, we require a comitment that degraded Steam Generator tubes will be pluged in accordance with the requirements of Regulatory Guide 1.121, " Bases for Plugging Degraded Steam Generator Tubes".
In the response to 110.61, the Applicant specifically takes exception to the Regulatory Guide requirement that a margin of 3 against tube failure during normal operation be maintained.
We require that the margin of 3 be maintained.
In Ch.16 of the FSAR "Techanical Specifications", the Applicant has comitted in Section 3/4.4.5 to plug all degraded tubes that have been reduced in wall thickness by 40 percent of the nominal tube wall thick-ness. We find tnis Technical Specification comitment acceptable and in consistent with the margin of three required by Reg. Guide 1.121. The Applicant should revise his response to 110.61 and refer to the Technical Specification tube plugging limit.
Subject to resolution of the above open issues, our findings are as follows:
The specified design and service combination of loadings as applied to ASME Code Class 1, 2, and 3 pressure retaining components in systems designated to meet Seismic Category I standards are such as to provide assurance that, in the event of an earthquake effecting the site or other service loadings due to postulated events or system operating transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the materials of construction. Limiting the stresses under such a.
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3.9.3.2 Pump and Valve Operability Assurance Program The pump and valve operability assurance program should demonstrate that all active pumps and valves, including their supports, can adequately sustain the combined service loading at stress levels which are at least equal to the specified service limits, and can perform their safety function without impairment.
Our review of the applicants program for pump and valves operability assurance has not been completed and will be discussed in a supplement to this report.
3.9.3.3 Design and Installation of Pressure Relief Devices The review objective for this section is to assure the adequacy of the design and installation of pressure relief devices such that integrity of the relieving devices and associated piping is maintained during the functioning of one or more of the relief devices.
For those safety and relief valves that discharge into closed systems, the applicants reponse to Question 110.49 is satisfactory except for the following. Time history or equivalent effects of changes of momentum should include the effects of postulated water slugs where water seals are used in the piping system.
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16-For th se safety valves that discharge into an cpIn system, the applicants response to Question 110.49 is not satisfactory. Justi-fication for use of Design Load Factors less than 2.0 must be provided by means of a dynamic analysis. For static equivalent analysis of these valves, a Design Load Factor of 2.0 must be used. The applicants position in Appendix A1.67-1 is not considered acceptable.
The applicant should specify the types of safety and relief valvet which are installed in the plant, their location and mounting arrangement, and the valve opening sequence and how this was accounted for in the analysis.
The load combinations and stress limits should aisc by specified in the FSAR.
Subject to resolution of the above open issues our findings are as follows.
The criteria used in the design and installation of ASME Code Class 1, 2 and 3 safety and relief valves provides adequate assurance that under discharging conditions, the allowable stress and strain limits will not be exceeded for the materials of construction. Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices, provides a conservative basis for the design and installa-tion of the devices to withstand these loads without loss of structural integrity or impairment of the overpressure protection function. The criteri.
used for the design and installation of ASME Code Class 1, 2, and 3 pressure relief devices constitute an acceptable basis for meeting the applicable requirements of General Design Criteria 1, 2, 4,14, and 15 and are con-sistent with those specified in Reg. Guide 1.67 and Standard Review Plan Section 3.9.3 3.9.3.4 Comoonent Supports To be acceptable, the component supports must provide adequate margins of safety under all plant operating conditions. Our review under this
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. section includes the combination cf loadings which were considered for each component support within a system, including the designation of tne appropriate service stress limits for these loadings.
The applicants response to Question 110.50 is not acceptable Justification for tne use of stress limits 50% greater than the normal allowable limits must be de=onstrated for all supports for which these limits were used. The actual calculated loads for LOCA and SSE must be used to demonstrate conservatism of the approach which was used for these calculations. One typical calculation as presented in Response to Question 110.50 does not show that this method is ecufvalent to and in compliance with Regulatory Guides 1.124 and 1.130 for all supports.
It should be made clear whether stress limits 50%
greater than allowable were used for support bolts as well.
In addition, the applicant should identify those component supports which will be at temperatures above ambient, and how these temperatures effects were accounted for in the linear elastic analysis per ASME Code Section III, Appendix XVII.
l The applicant has not reponded to questions 110.14 and 110.62 l
dealing with asymmetric loading of component supports.
l Subject to resolution of the above open issues, our findings.
are as follows.
The specified design and service loading combinations used for the design of ASME Code Class 1, 2, and 3 component supports in systems classified as seismic category I provide assurance l
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that, in the event of an earthquake or other service loadings due to postulated events or system operating transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the material of construction. Limiting the stresses under such loading combinations provides a conservative basis for the design of support components to withstand the most adverse combination of loading events without loss of structural integrity or supported component operability. The design and service loading combinations and acsociated stress and deformation limits specified for ASME Code Class 1, 2, and 3 component supports comply with Standard Review Plan Section 3.9.3 and satisfy the applicable portions of General Design Criteria 1, 2, and 4.
3.9.4 Control Rod Drive Systems The objective of the review under this section is to determine that design, fabrication, and construction of the control rod drive mechanisms provide structural adequacy and that suitable life cycle testing programs have been utilized to prove operabiltiy under service life conditions.
l For all pressure boundary items of the control drive mechanisms, the applicant has committed to follow the design rules for Class 1 components as specified in Section III of the ASME Code.
It is requested, however, that a list of loading combination and stress limits for these pressure boundary items also be specified. The applicant should also specify the design criteria and loading combinations which were used for non-pressure boundary items of the control drive mechanisms.
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. Subject to resolution of the above open issues, our findings are as follows.
The design criteria and test program conducted in verification of the mechanical operability and life cycle capabilities of the reactivity control system are in conformance with established criteria, codes, standards, and specifications which are acceptable to the Regulatory staff. The use of these criteria provide reasonable assurance that the system will function reliably when required, and form an acceptable basis for satisfying the mechanical reliability stipulations of General Design Criteria 27.
3.9.5 Reactor Pressure Vessel Internals Our review under this section is concerned with the load combinations, allowable stress, and deformation limits which were used in the design of the Byron and Braidwood reactor -internals.
It is noted that Indian Point Unit 2 was established as the prototype plant for testing the Byron and Braidwood four loop plant reactor internals.
In addition, the Trojan plant l
instrumentation program provides data to cover the effects of hardware modification such as the addition of neutron l
shielding pads to replace the thennal shield. These tests and the successful operating experience of other similar plants indicate l
that the Byron and Braidwood reactor internals will remain i
I structurally sound during the 40 year design life of the plant.
Our review also included the structural integrity of reactor internals under the combination of loads which could result from I
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. postulated events suci as the safe shutdown e;rthquake and loss of coolant accident.
As previously noted in Section 3.9.3.4 of this safety evaluation, the applicant has not responded to Question 110.62 which deals with the asymmetric loading of reactor internals due to the combined loading of a simultaneous safe shutdown earthquake and loss of coolant accident. This question must be addressed before our review can be completed.
The applicant states that the intent of the ASME Code,Section III is followed with respect to allowable stresses on page 3.9-81 of the FSAR. This statement should be clarified and the exact manner of compliance with the intent of the ASME Code should be specified. For example, if an elastic system analysis followed by elastic analysis of the reactor core support structure is conducted, the applicant should specify if the design rules of paragraph NG-3000 of the ASME Code,Section III, Class 1 are followed.
The reference by the applicant to the use of unirradiated material properties should be clarified to show how the use of these proper-ties is conservative for fatigue evluation of reactor internals.
Subject to resolution of the above open issues our findings are as follows.
The specified transients, design and service loadings, and combination of loadings as applied to the design of the reactor internals
l 21-provide reasonable assurance that in the event of an earthquake or of a system transient during normal plant operation, the resulting deflections and associated stresses imposed on these reactor internals would not exceed allowable stresses and deformation limits for the materials of construction. Limiting the stresses and deformations under such loading combinations provides an acceptable basis for the design of these reactor internals to withstand the most adverse loading events which have been postulated to occur during service lifetime without loss of structural integrity or impairment of function. The design procedures and criteria used by the applicant in the design of the reactor internals comply with Standard Review Plan Section 3.9.5 and constitute an acceptable basis for satisfying the applicable requirements of General Design Criteria 1, 2, 4, and 10.
3.9.6 Inservice Testing of Pumps and Valves In Sections 3.9.2 and 3.9.3 of this Safety Evaluation Report, we discussed the design and seismic qualification of safety related pumps and valves. The design of these pumps and valves is their safety function (open, close, start, etc.) at any time, and for any postulated design basis operating condtion during the life of the plant.
In order to provide added assurance of the reliability of these components, the applicant is required to periodically test all-safety related pumps and valves. These tests are performed in general accordance with the rules of Section XI of the ASME Code and verify that these pumps and valves operate successfully when cr.lled upon.
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. In addition, periodic measurements are made of various parameters for comparison to base line measurements in order to detect possible long term degradation of these components. Our review under Standard Review Plan Section 3.9.6 covers the applicants program for preservice and inservice testing of pumps and valves and specifically addresses those areas of the test program for which the applicant requests relief from the requirements of Section XI of the ASME Code.
The applicant has not yet submitted his prograu for preservice and inservice testing of pumps and valves as requested by Question 110.64, therefore, we have not yet completed our review of this section.
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