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OCT 03'534 MEMORANDUM FOR:
Thomas M. Novak, Assistant Director for Licensing, DL FROM:
James P. Knight, Assistant Director for Components & Structures Engineering, DE
SUBJECT:
DRAFT SER INPUT FOR V0GTLE 1 & 2 Plant Name:
Vogtle Electric Generating Plant - Units 1 & 2 Docket No.:
50-424/50-425 Licensing Stage:
OL DL Branch & PM:
LB #4, M. Miller DE Branch '& Reviewer:
MEB, D. Terao Review Status:
DRAFT The Mechanical Engi.neering Branch and its contractor laboratory (Pacific North-west ~ Laboratory) have completed the revidw of the Vogtle FSAR.
As described in the memorandum from R. Bosnak to B. J. Youngblood dated December 14, 1933, the Q(O attached draft SER includes a list of all our open issues and staff positions i
which we have identified in our FSAR review. '
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We request that you forward the list of.. questions, open items, and staff post-tions to the applicant.
The applicant should then prepare an agenda for a meeting in which we can discuss and resolve the open items from our review. We anticipate this meeting being held over a three day period at a mutually agree-able site about 45-60 days after the applicant has received our open items.
-After this meeting and any necessary follow-up, we will update the SER input into a form sufficiently clean for final publication.
It should be emphasized to the applicant that we expect this extended meeting to resolve almost all of these open items.
therefore, it should bring the NSSS, AE, and utility repre-sentatives necessary to both discuss technical issues and make binding commit-ments.
We strongly recommend the meeting be held at the Bechtel offices in
.Los Angeles, CA during the week of January,7, 1985.
The draft SER contains those sections of 3.6 and 3.9 aplicableto MEB's scope of review.
Sections 3.2 and 5.2 will be reviewed separately and SER input will be provided under a separate cover memo.
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James}P.Knighty, Assistant Director
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for Components & Structures Engineering 4
Div s on of Engineering Branch cc: See.P a4 34l t6 V I %
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i Thomas M. Novak 2
Att'achments:
- 1) draft SER
- 2) SER Questions cc:
R. Vollmer, DE R. Bosnak,DE E. Adensam, DL M. Miller, DL H. Brammer, DE R. Kirkwood, DE D. Terao, DE J. Alzheimar, PNL (5) 4 i
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- .,1 3.6.2.
Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping General Design Criterion 4, " Environmental and Missile Design Bases", of 10 CFR'Part 50, Appendix A,. requires that structures, systems, and components important to safety shall be designed to be compatible with and to accommodate the effects of the environmental conditions as a result of normal operations, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be adequately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from eve.its and conditions outside the nuclear power plant.
The staff's review, conducted in accordance with Standard Review Plan (NUREG-0800), Section 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", pertains to the methodology used for protecting safety-related structures, systems, and com-ponents against the effects of postulated pipe breaks both inside and outside containment. The staff has used the review procedures identified in SRP 3.6.2 to evaluate the effect that breaks in high energy fluid systems would have on adjacent safety-related structures, systems, or components with respect to jet impingement and pipe whip..The staff also reviewed the location, size, and orientation of postulated failures and the methodology used to calculate the resultant pipe whip and jet impingement loads that might affect nearby safety-related structures, systems, or components.
The details of the staff's review follow.
Pipe whip need only be considered in those high-energy piping systems having fluid reservoirs with sufficient capacity to develop a jet stream.
The criteria for determining high-and moderate-energy lines is found in Branch Technical Position ASB 3-1 of Standard Review Plan 3.6.1, " Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment".
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VCGILE DSER SEC 3.6.2 IMPUT
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This criteria has been used correctly by the applicant. A list of all high
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energy systems is included in the FSAR.
For high energy piping within the containment penetration area where breaks are not postulated, SRP Secticn 3.6.2 sets forth certain criteria for the analysis and subsequent augmented inservice inspection requirements. Breaks need not be postulated in those portions of piping within the containment penetration region that meet the requirements of the ASME Code,Section III, Subarticle NE-1120 and the additional requirements outlined in Branch Tech 11 cal Position MEB 3-1 of SRP Section 3.6.2.
Augmented inservice inspection is required for those portions of piping within the break exclusion region.
For ASME Section III Class 1 high e'nergy fluid syste.n piping not in the contain-ment penetration area, SRp Section 3.6.2 states that breaks are to be postulated at every location where the fatigue cumulative usage factor, as determined by the ASME Code, is greater than 0.1.
Additionally, breaks arc also to oe postu-lated at those ASME Class 1 piping locnions wh:re the primary or secondary p
stress intensity range (including che ::ero load set) as calculated by equation (10) and either equation (IT) er (13) in Paragraph NB-3653 of ASME Section III exceeds 2.4 Sm for normal and upset conditions including the OBE.
The applicant has provided drawings of break locations showing types of breaks, structural barriers, restraint locations and constrained directions for each restraint for the primary coolant loop and all breaks inside containment.
The following are considered to be open items.
The applicant has not yet completed the final pipe whip and jet impingement evaluation for all high energy piping systems.
The staff's review cannot be completed until this information is available for review.
Clarification is required on some aspects of the applicant's pipe break criteria and pipe whip restraints.
Based on the staff's review of FSAR Section 3.6.2 and resolution of the above open items, the staff's findings are as follows.
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V0GTLE OSER SEC 3.6.2 INPUT c-
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In its evaluation, the staff concludes that the pipe rupture postulation and the ascociated effects are adequately considered in t'e plant design, and, there-fofe, ari acceptable and meet the requirements of General Design Criterion 4.
This conclusion is based on the following.
(1) The proposed pipe rupture locations have been adequately assumed and the
, design of piping restraints and measures to deal with the subsequent dynamic effects of pipe whip and jet impingement provide adequate protec-tion to the structural integrity of safety-related structures, systems and components.
(2) The provision for protection against dynamic effects associated with pipe ruptures of the reactor coolant pressure boundary inside containment and the resulting discharging fluid provide adequate assurance that design basis loss-of-coolant accidents will not be aggravated by the sequential failures of safety-related piping, and emergency core cooling system performance will not be degraded by these dynamic effects.
Q (3) The proposed piping and restraint arrangement and applicable design con-siderations for high-and moderate-energy fluid systems inside and outside of containment, including the reactor coolant pressure boundary, will pro-vide adequate assurance that the structures, sy' stems, and components important to safety that are in close proximity to the postulated pipe rupture will be protected. The design will be of a nature to mitigate the consequences of pipe ruptures so that the reactor can be safely shut down and maintained in a safe shutdown condition in the event of a postulated rupture of a high or moderate energy piping system inside or outside of containment.
3.9 MECHANICAL SYSTEMS AND COMPONENTS The review performed under SRP Sections 3.9.1 through 3.9.6 of NUREG-0800 per-tains to the structural integrity and functional capability of various safety-relat~ed mechanical components in the plant.
The staff's review is not limited to ASME Code components and supports, but is extended to other components such as control rod drive mechanisms, certain reactor internals, and any safety-related piping designed to industry stadards other than the ASME Cede.
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VCGTLE DSER SEC 3.6.2 INPUT
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The-staff reviews such issues as load combinations, allowable stresses, methods
-of. analysis, summary of results, and preoperational testing. The staff's refliew must arrive.at the conclusion that there is adequate assurance of a mechanical component performing it's safety-related function under all postu-lated combinations of normal operating conditions, system operating transients, postulated pipe breaks, an'd seismic events.
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3.9.1 Special Topics for Mechanical Components The review of this section was performed following SRP Section 3.9.1, "Special Topics for Mechanical Components". All areas of review and review procedures identified in SRP Section 3.9.1'were followed.
The staff has reviewed the design transients and methods of analysis used for all seismic Category I com-ponents, component supports, core support structures and reactor internals designated as Class 1 and CS under the ASME Code,Section III, and those not covered by the Code. The assumptions and procedures used for the inclusion of transients in the fatigue evaluation of ASME Code Class 1 and CS components have been reviewed. The staff's review also covered the computer programs used in the design and analysis'of seismic Category I components and their supports and experime'ntal and inelastic analytical techniques.
The applicant has provided a list of the design transients and the number of cycles for each design transient used for design.
Five OBEs of ten cycles each and one SSE of ten cycles have been included. This is in conformance with the requirements of SRP 3.9.1.
Analysis of mechanical components by the use of computer programs was performed by the applicant. A list showing all computer programs used by the applicant for static and dynamic analyses to determine the structural integrity and func-tional, integrity of seismic Category I Code and non-Code items, and the analyses to determine stresses along with a description of the program is included in the FSAR.
Design control measures to verify the adequacy of the design of safety-
.related components is required by 10 CFR Part 50, Appendix B.
Based upon the staff's review of FSAR Section 3.9.1 our findings are as follows.
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VCGTLE DSER SEC 3.6.2 INPUT
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The staff concludes that the design transients and resulting loads and load combinations with appropriate specified design and service limits for mechanical codponents and supports are acceptable and meets the relevant requirements of General Design Critert'a 1, 2, 14, 15; and 10 CFR Part 50, Appendix B; and 10 CFR Appendix A.
This is based on the following.
(1). The applicant has met the relevant requirements of General Design Cri-teria 14 and 15 by demonstrating that the design transients and resulting loads and load combinations with appropriate specified design and service limits which the applicant has used for designing Code Class 1 and CS com-ponents and supports, and reacto: internals provide a complete basis for design of the reactor coolant pressure boundary for all conditions and events expected over the service lifetime of the plant.
(2) The applicant has met the relevant requirements of General Design Cri-teria 2 and 10 CFR Part 100, Appendix A by including seismic events in design transients which serve as design bases to withstand the effects of natural phenomena.
(3) The applicant has met the relevant requirements of 10 CFR Part 50, Appendix 8, and General Design Criteria 1 by having submitted information that demonstrates the appitcability and validity of the design methods and computer programs used for the design and analysis of seismic Category I Code Class 1, 2, 3 and CS structures, and non-Code structures within the present state-of-the-art limits and by having design control measures which are acceptable to assure the quality of the computer programs.
3.9.2.
Dynamic Testing and Analysis of Systems, Components and Equipment The staff has reviewed the methodology, testing procedures, and dynamic analyses employed by the applicant to ensure the structural integrity and func-tionality of piping systems, mechanical equipment, and their supports under vibratory loadings.
The principal document used in this review is SRP (NUREG-0800) Section 3.9.2, " Dynamic Testing and Analysis of Systems, Compo-nents, and Equipment". All areas of review and review procedures identified in SRP Section 3.9.2 were followed. The staff's review included (1) the piping 10/22/34 5
VCGTLE DSER SEC 3.o.2 input
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vibration, thermal expansion, and dynamic effect testing, (2) the seismic system i
analysis methods, (3) the dynamic responses of structural components within the reictor caused by steady-state and operational flow transient conditions for non prototype reactors, (4) flow-induced vibration testing of reactor internals to be conducted during the preoperational and start-up test program, and (5) the dynamic analysts methods used to confirm the structural design adequacy and functional capability of the reactor internals and piping attached to the reac-tor vessel when subjected to loads from a loss-of-coolantaccident (LOCA) in combination with an SSE.
3.9.2.1 Piping Preoperational Vibration and Dynamic Effects Testing Piping vibration, thermal expansion, and dynamic effects testing will be con-ducted during a preoperational testing program. The purpose of these tests is to assure that the piping vibrations are within acceptable limits and that the piping system can expand thermally in a manner consistent with the design intent. During the Vogtle plant's preoperational and start-up testing program, p
the appiteant will test various piping systems for abnormal, steady-state or transient vibration and for restraint of thermal growth.
Systems to be mont-tored will include 1) ASME Code Class 1, 2 and 3 piping systems, 2) high energy piping systems inside seismic Category I structures, 3) high energy portions of systems whose failure could reduce the functioning of seismic Category I plant features to an unacceptable safety level, and 4) seismic Category I portions of moderate energy piping systems located outside containment.
Steady-state vibration, whether flow-induced or caused by nearby vibrating machinery, could cause 10' or 10' cycles of stress in the pipe during it's 40 year life.
For this reason, the staff requires that the stresses associated with steady-state vibration be minimi ed and limited to acceptable levels. The test program will consist of a r.txture of instrumented measurements and visual observations by qualified personnel.
The following are considered to be open items.
+ -A description is needed for the methods to be used to relate measured vibration values to stress levels.
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VCGTLE DSER SEC 3.6.2 INPUT 9
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Assurance is needed that all essential safety-related instrument lines
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will be included in the vibration monitoring program during preopera-tio'nal or start-up testing.
Based upon the staff's review of FSAR Section 3.9.2.1 and resolution of the open items, the staff concludes that the applicant has met the relevant require-ments of General Design Criteria 14 and 15 with respect to the design and testing of the reactor coolant pressure boundary. This provides reasonable assurance that rapidly propagating failure and gross rupture will not occur as a result of vibratory loadings.
In addition, the testing assures that design conditions are not exceeded during normal operation including anticipated operational occurrences by having an acceptable vibration, thermal expansion, and dynamic effects test program which'will be conducted during start-up and initial operation of specified high and moderate energy piping, including all associated restraints and supports. The tests provide adequate assurance that the piping and piping supports have been designed to withstand vibrational dynamic effects due to valve closures, pump trips, and other operating modes
{.N associated with the design basis flow conditions.
In addition, the tests pro-vide assurance that adequate clearances and free movement of snubbers exist for unrestrained thermal movement of piping and supports during normal system heatup and cooldown nperations.
The planned test will develop loads similar to those experienced during reactor operations.
3.9.2.2 Seismic Subsystem Analysis The staff's review performed according to Standard Review Plan Section 3.9.2 included Section 3.7.3 of the applicant's FSAR, " Seismic Subsystem Analysis".
Areas reviewed were seismic analyses methods, determination of the number of earthquake cycles, basis for selection of frequencies, the combination of modal responses and spatial components of an earthquake, criteria used for damping, torsional effects of eccentric masses, interaction of other piping with Category I piping, and Category I buried piping systems.
The scope of the review of the Vogtle's seismic system and subsystem analysis includes the seismic analysis methods for all seismic Category I piping systems j
and components.
The staff has reviewed the manner in which the dynamic system i
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VCGTLE DSER SEC 3.6.2 INPUT
analysis is performed, the method of selection of significant modes, whether the number of masses or degrees of freedom is adequate, and how consideration is'giveri to maximum relative displacements.
The review included design method-ologies and procedures used for the evaluation of the interaction of non-seismic Category I piping with seismic Category I piping, ahd the seismic methods which consider the effect of settlement and movement at suppert points, ynetration, and anchors for seismic Category I buried piping systems.
In addition, the staff reviewed seismic analysis procedures for reactor internals. The system and subsystem analyses are performed by the applicant on an elastic basis.
Modal response spectrum, multi-degree of freedom and time history methods form the basis for the analyses of all major seismic Category I systems and compo-nents. When the response spectrum method is used, modal responses are combined by the square-root-sum-of-the-squares (SRSS) rule.
-For the dynamic analysis of seismic Category I piping, each piping system was idealized as a mathematical model consisting of lumped masses connected by elastic members. The stiffness matrix for the piping system was determined using the elastic properties of the pipe. This includes the effects of torsional, bending, shear, and axial deformatio'ns as well as change in stiff-ness due to curved members. Next, the mode shapes and the undamped natural frequencies were obtained. The dynamic response of the system was calculated by using the response spectrum method of analysis.
For a piping system which was supported at points with different dynamic excitations, the response analysis was performed using an enveloped response spectrum.
The following are considered to be open items.
The staff has requested further information on the design of seismic interface anchors.
Clarification is required on the use of damping values and equivalent static factors other than those discussed in the SRP.
- *The staff has requested detailed information regarding the piping analysis procedures used for the main steam and feedwater piping outside containment.
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V0 GILE DSER SEC 3.6.2 INPUT
e Based upon the staff's review of FSAR Section 3.7.3, and contingent upon resolution of the open items, the staff concludes that the applicant has met th[relehantrequirementsofGeneralDesignCriteria2withrespecttodemon-strating the design adequacy of all Category I piping systems, components, and their supports to withstand earthquakes by meeting the regulatory positions of Regulatory Guides 1.61 and 1.92 and by providing acceptable seismic analysis procedures and criteria. The scope of review of the seismic system analysis included the seismic analysis methods of all Category I piping systems, components, and their supports.
It included review of procedures for model-ing, and inclusion of torsional effects, seismic analysis of multiply-supported equipment and components with distinct inputs, and determination of composite damping. The review has included design criteria and procedures for evalua-tion of the interaction of non-Category I piping with Category I piping. The review has also included criteria and seismic analysis procedures for reactor internals.
3.9.2.3 Preoperational Flow-Induced Vibration Testing of Reactor Internals
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Flow-induced vibration testing of reactor internals will be conducted during the preoperational and startup test program.
The purpose of this test is to demonstrate that flow-induced vibrations similar to those expected during operation will not cause unanticipated flow-induced vibrations of significant magnitude, or structural damage.
The Indian point No. 2 reactor has been established as the prototype for the Westinghouse four-loop plant internals verification program.
The only signifi-cant differences between Vogtle's internals and the Indian Point No. 2's intarnals are the replacement of the annular thermal shield with neutron shield panels and the substitution of 17x17 fuel assemblies for 15x15 assemblies, and the change to the UHI-style inverted top hat support structure configuration.
The change to the neutron shield panels and 17x17 fuel assemblies has been tested at the Trojan plant. The change to the UHI-style inverted top hat suppdrt structure configuration has been tested at the Sequoyah Unit 1 plant.
The Four Loop Internals Assurance Program conducted on Indian Point No. 2 supplemented by the Trojan and Sequoyah Unit 1 data jointly satisfy Regulatory Guide 1.20.
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VCGTLE DSER SEC 3.6.2 INPUT
The appitcant has committed to test the reactor internals in accordance with the provisions of Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program'for Reactor Internals During Preoperational and Start-up Testing",
Revision 2, for non prototype Category I plants.
The applicant will conduct a visual inspection prior to hot functional testing and after hot functional testing the applicant has committed to inspecting all major load-bearing sur-faces, torsional, lateral, and vertical restraints, locking and bolting devices whose failure could adversely affect the structural integrity of the internals, and all other locations examined on the prototype design. The inside of the vessel will be inspected with all the internals removed both prior. to and sub-sequent to hot functional testing to verify that no loose parts or foreign material are present.
The applicant will subject the internals to an operating time of sufficient duration to assure that a minimum of 108 cycles of vibration will be experi-enced by the critical components. At completion of the flow test, the vessel head will be removed and the internals will ha 1.n,pected for evidence of wear
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and loose parts. The inspection will over all components which were examined on.the prototype design.
Important welds, bearing surfaces, and alignment and locking devices in the internals will be inspected with t% aid of 5x or 10x magnifying glass.
The staff finds the inspection program to be sufficient and the hot functional test to be of adequate length.
Based upon the staff's review of FSAR Section 3.9.2.4, findings are as follows.
The staff concludes that the applicant has met the relevant requirements of General Design Criteria 1 and 4 with respect to the reactor internals being designed and tested to quality standards commensurate with the importance of the safety functions being performed and being appropriately protected against dynamic effects by meeting the regulatory positions of Regulatory Guide 1.20 for the conduct of preoperational vibration tests and by having a preopera-tional vibration program planned for the reactor internals which provides an acceptable basis for verifying the design adequacy of these internals under test loading conditions comparable to those that will be experienced during operation.
The combination of tests, predictive analysis, and post-test 10/22/84 10 VC3TLE DSER SEC 3.6.2 INPUT
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inspection provide adequate assurance that the reactor internals will, during their service lifetime, withstand the flow-induced vibrations of the reactor without' loss of structural integrity.
The integrity of the reactor internals in service is essential to assure the proper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit safe reactor operation and shutdown.
3.9.2.4 Dynamic System Analysis of Reactor Internals Under Faulted Conditions The applicant has analyzed it's reactor internals and unbroken loops of the reactor coolant pressure boundary, including the supports, for the combined loads due to a simultaneous loss-of-coolant-accident and safe shutdown earth-quake.
The applicant has described the methodology used in developing the dynamic loads resulting from an asymmetric load from a postulated pipe break at the RPV nozzle safe-end in FSAR Section 3.9 N.2.5.
Based on the staff's review of FSAR Section 3.9.N.2.5 and the load combinations
,- S and stress limits as presented in tables contained in FSAR Section 3.9.3, the staff concludes that the applicant has met the relevant requirements of General Design Criteria 2 and 4 with respect to the design of systems and components important to safety to withstand the effects of earthquakes and the appropriate combinations of the effects of normal and postulated accident conditions with the effects of the safe shutdown earthquake (SSE) by performing a dynamic system analysis which provides an acceptable basis for confirming the struc-tural 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 SSE.
The analysis provides adequate assurance that the combined stresses and strains in the components of the reactor coolant system and reactor internals will not exceed the allowable design stress and strain Ifmits for the materials of construction, and that the resulting deflections or dis-placements at any structural element of the reactor internals will not distort the reactor internals geometry to the extent that core cooling may be impaired.
The methods used for component analysis have been found to be compatible with those used for the system analysis.
The proposed combination of component and system analyses are, therefore, acceptable.
The assurance of structural 10/22/84 11 VCGTLE DSER SEC 3.6.2 INPUT
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o integrity of the reactor internals under LOCA conditions for the most adverse postulated loading event provides added confidence that the design will with-st'and a* spectrum of lesser pipe breaks and seismic loading events.
3.9.3 ASME Code Class 1. 2, and 3 Components. Comp ~onent Supports and Core Support Structures The staff's review under Standard Review Plan Section 3.9.3 is concerned with the structural integrity and functional capability of pressure-retaining com-
'ponents, their supports, and core support structures which are designed in accordance with the ASME Boiler and pressure Vessel Coie,Section III, or earlier industrial standards. All areas of review and ieview procedures identified in SRp Section 3.9.3 were followed.
The staff has reviewed loading
- combinations and their respective stress limits, the design and installation of pressure relief devices, and the design and structural integrity of ASME Code
. Class 1, 2, and 3 components and component supports. Details of the staff's review are included in the following sections.
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3.9.3.1 Loading Combinations, Design Transients and Stress Limits The first area of review is the methodology used for load combinations and allowable stress limits in FSAR Section 3.9.3.
The following is considered an open item.
The applicant's methodology used for load combinations does not appear to conform to the acceptance criteria in SRP 3.9.3.
Specifically, the applicant does not appear to have included the LOCA loads in evaluation of the faulted condition limits for ASME Class 2 and 3 components and their supports where such loads are appropriate.
The applicant is to provide the basis for assuring that ASME Code Class 1, 2, and 3 piping can perform its intended function for service levels C
- and 0 loadings.
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Based upon the staff's review of FSAR Sections 3.9.B.3.1 and 3.9.N.3.1 and contingent upon the satisfactory resolution of the open items, the staff's fi'ndingi will be as follows.
The applicant has met the requirements of 10 CFR 50.55a and General Design Criteria 1, 2, and 4 with respect to the design and service load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 components by ensuring that systems and components important to safety are designed to quality standards commensurate with their importance to safety
'and that these systems can accommodate the effects of normal operation as well as postulated events such as loss-of-coolant accidents and the dynamic effects resulting from earthquakes. The specified design and service combina-tions of loading as applied to ASME Code Class 1, 2, and 3 pressure retaining components in systems designed to meet seismic Category I standards are such as to provide assurance that in the event of an earthquake affecting the site for other service loading caused by postulated events or system operating tran-sients, the resulting combined stresses imposed on system components will not (7 y exceed allowable stress and strain Itmits for the materials of ccnstruction.
Limiting the stresses under such loading combinations provides a conservative basis for the design of system components to withstand the most adverse com-bination of loading events without loss of structural integrity.
3.9.3.2 Design and Installation of Pressure Reitef Device The staff has reviewed Section 3.9.3.3 of the applicant's FSAR with respect to the design and installation, and testing criteria applicable to the mounting of pressure relief devices used for the overpressure protection of ASME Class 1, 2, and 3 components.
This review, conducted in accordance with SRP Section 3.9.3 (NUREG-0800),includesevaluationoftheapplicableloadingcombinationsand stress criteria. 'The design review extends to consideration of the means pro-vidad to accommodate the rapidly applied reaction force when a safety valve or relief valve opens, and the transient fluid-induced loads applied to the piping downstream of a safety or relief valve in a closed discharge piping system.
The staff also reviewed the applicant's relief and safety valve test results as required in Item II.O.1 of NUREG-0737.
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kn accordance with Item 1I.0.1 of NUREG-0737, pressurized water reactor and boiling water reactor licensees and applicants are required to conduct testing to' qualify the r,eactor coolant system relief and safety valves, block valves, and associated piping and supports under expected operating conditions for design-basis transients and accidents.
The Electric Power Research Institute (EPRI) was contracted by the PWR Owners Group to develop and carry out a generic test program and to provide the generic test data to be used by the PWR utilities to satisfy the NUREG-0737, Item II.D.1, requirements.
Testing of valves in the EPRI program was completed by December 21, 1981.
By letter dated April 1,1982, from D. P. Hoffman, Chairman of the PWR Safety and Relief Valve Test Program Subcommittee, the EPRI/PWR Owners Group transmitted the following reports to NRC:
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Valves Seleetion/ Justification Report 2.
Valve Inlet Fluid Condition for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants (note:
two other NSSS vendor reports were also received) 3.
Test Condition Justification Report 4.
Safety and Relief Valve Test Report 5.
Application of RELAPS/ MOD 1 for Calculation of Safety ad Relief Valve Discharge Piping Hydrodynamic Loads.
Additionally, by letter dated June 1, 1982, from R. C. Youngdahl to H. Denton, reports documenting block valve testing performed by EPRI were transmitted to NRC. These generic reports are currently being reviewed by the staff. On the basis of a preliminary review of the EPRI generic reports, the staff has con-cluded that they contain data that can be used by the appiteant to prepare an 10/22/84 14 V0GTLE DSER SEC 3.6.2 INPUT
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Item II.D.1 plant-specific response for the valves and associated piping for-Vogtle.
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The staff re' quires that these plant-specific submittals be made before fuel load in accordance with the schedule of NUREG-0737 and the September 29, 1981,
~ larification letter on this matter. Once the staff has received this c
information, it will report its findings in a supplement to this SER.
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C' The staff requires cdditional information on the design of safetysand relief valves.
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The applicant'has met the requirements of 10 CFR 50.55a and General Design Criteria 1, 5; "and 3 with respect to the criter'f a used for design and instal-lation of ASME Code Class 1, 2, and 3 cva@pressura, relief devices by ensuring that safety and rellief valves and their installations are designed to standards which are commensurate with their safety functions, and that they can accommo-date-the effects of discharge caused by norinal operation as well as postulated s
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events such as loss-of-coolant accidents and the dynamic effects resulting from T _'
the safe shutdown earthquake. The relevant requirements of General Design Criteria 14 aitd.15 are also met with respect'to. assuring that the reactor coolant pressure boundary design.lfmits for normal operation including antici-N pated operational occurrences are not exceeded. The criteria used by the applicant in the design'and installation of ASME Cla'ss 1, 2, and 3 safety and relief valves provide adequate assurances that. under discharging conditions, the resulting stresses will not exceed allowable stress and' strain limits for the materials of construction.
Limiting the stresses under the loading combi-nations associated with the actuation of these pressure reitef devices provides a conservative basis for the design and installation of the device to withstand these loads wlthout loss of structural integrity or impairment of the over-pressure protection function
'3.9.3.3 Component Supports T
- The staff's review of Section 3.9.3.4 of the applicant's FSAR relates to the methodologyusedbytheapplicantin'thedesfgnofA!kEClass1,2,and3 s.
component supports.
The review includes asses'sment of design and structural r
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integrity of the supports.
The review addresses three types of supports:
plate and'shell, linear, and component standard types. Additional information to' ensuie a ccmp,lete basis and consistent approach for the design and construc-tion of component supports is required. The specific concern has been trans-
~
mitted to the applicant.
Based upon the staff's review of FSAR Section 3.9.3.4 and contingent upon the resolution of the open items, the staff findings will be as follows.
The applicant has met the requirements of 10 CFR 50.55a and General Design Criteria 1, 2, and 4 with respect to the design and service load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 component supports by ensuring that component supports important to safety are designed to quality standards commensurate with their importance to safety, and that these supports can accommodate the effects of normal opera-tion as well as postulated events such as_ loss-of-coolant accidents and the dynamic effects resulting from the safe shutdown earthquake. The combination of r ~cs loadings (including system operating transients) considered for each component support within a system, including the designation of the appropriate service stress-limit for each loading combination, has met the positions and criteria of Regulatory Guides 1.124 and 1.130 and are in accordance with NUREG-0484, Revision 1.
The specified design and service loading combinations used for the design of ASME Code' Class 1, 2, and 3 component supports in systems classified p
as seismic Category I provide assurance that in the event of an earthquake or other service loadings caused by postulated events or system operating tran-j sients, 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 loading combinations provides a conservative basis for the design of support components to withstand the most adverse com-t bination of loading events without loss of structural integrity.
Class CS component evaluation findingt are covered in SER Section 3.9.5 in connection with reactor. internals.
1
(
"y-I 10/22/S4 15 VCGTLE DSER SEC 3.6.2 INPUT i
V f 7
-3.9.4 Control Rod Drive Systems
. The staff's review under Standard Review Plan Section 3.9.4 covers the design of the control rod drive system up to it's interface with the control rods.
The rods and drive mechanism shall be capable of reliably controlling reactivity changes either under conditions of anticipated normal plant operational occur-rences, or under postulated accident conditions. The staff reviewed the.
information in FSAR Section 3.9.4 relative to the analyses and tests performed to assuie the structural integrity and functionality of this system during normal operation and under accident conditions. The staff also re. viewed the life-cycle testing performed to demonstrate the reliabilit'y of the control rod drive system over it's 40-year life.
A detailed review of the design of the control rod drive system with respect to its capability of controlling reactivity and cooling the reactor core with appropriate margin in conjunction with either the emergency core, cooling system or the reactor protection system was not performed because of the system Q
similarity with other Westinghouse plants which were found to be acceptable.
~
The staff is not aware of any significant design changes in the control rod drive system for the Vogtle plant.
Based on the staff's review of the above information, it was concluded that the design of the control rod drive system is acceptable and meets the requirements of General Design Criteria 1, 2, 14~, 26, 27, and 29, and 10 CFR 50.55a. This conclusion is based on the following.
(1) The applicant has met the requirements of GDC 1 and 10 CFR 50.55a, with respe,ct to designing components important to safety to quality standards commensurate with the importance of the safety functions to be performed.
The design procedures and criteria used for control rod drive systems are in conformance with the requirements of appropriate ANSI and ASME codes.
(2) The applicant has met the requirements of GDC 2, 14, and 26 with respect to designing the control rod drive system to withstand effects of earth-quakes and anticipated normal operation occurrences with adequate margins to assure its structural integrity and functional capability and with l
i 10/22/84 17 VCGTLE DSER SEC 3.6.2 INPUT I
i
.c extremely low probability of leakage or gross rupture of reactor coolant pressure boundary.
The specified design transients, design and service loa' dings, combination of loads, and limiting the stresses and deformations under such loading combinations are in conformance with.the requirements of-appropriate ANSI and ASME codes and acceptable regulatory positions specified in SRP Section 3.9.3.
(3)' The applicant has met the requirements of GDC 27 and 29 with respect to designing the control rod drive system to assure its capability of con-trolling reactivity and cooling the reactor core with appropriate margin, in conjunction with either the emergency core cooling system or the reactor protection system.
The operability assurance program is acceptable with respect to meeting system design requirements in observed performance as to wear, functioning times, latching, and overcoming a stuck rod.
3.9.5 Reactor pressure Vessel Internals 4.
,Q The staff's review under Standard Review Plan 3.9.5 is concerned with the load combinations, allowable stress limits and other criteria used in the design of the Vogtle reactor internals.
The staff has limited their review of SRP Sec-tion 3.9.N.5 to include the design and analysis of the reactor internals and the deformation limits specified for those components.
A detailed review of the configuration and general arrangement of the mechanical and structural internal elements was not performed because of the similarity with other Westinghouse plants which were found acceptable. The staff is not aware of any significant design changes in the reactor internais for the Vogtle plant.
Based on the staff's review of FSAR Section 3.9.5, the staff concludes that the l
design of reactor internals is acceptable and meets the requirements of General Design Criteria 1, 2, 4, and.10 and 10 CFR 50.55a. This conclusion is based on the following.
(1) The applicant has met the requirements of GDC 1 and 10 CFR 50.55a with respect to designing the reactor internals to quality standards commen-surate with the importance of the safety functions to be performed. The i
10/22/S4 IS VCGILE DSER SEC 3.6.2 INPUT
design procedures and criteria used for the reactor internals are in con-formance with the requirements of Subsection NG of the ASME Code, Sedtion III.
9 (2) The applicant has met the requirements of GDC 2, 4, and 10 with respect to designing components important to safety to withstand the effects of earthquake and the effects of normal operation, maintenance, testing, and postulated loss-of-coolant accidents with sufficient margin to ensure that capability to perform its safety functions is maintained and the specified
~
acceptance fuel design ifmits are not exceeded.
The specified design transients, design and service loadings, and combination of loadings as applied to the design of the reactor internals structures and components provided 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 structures and components would not exceed allowable stresses and deformations under such loading combinations.
~
'Q This provides an acceptable basis for the design of these structures and com-
/
ponents 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.
3.9.6 Inservice Testing of pumps and Valves The review under Standard Review Plan 3.9.6 is concerned with the inservice testing of certain safety-related pumps and valves typically designated as ASME Class 1, 2, or 3.
Other pumps and valves not categorized as Code Class 1, 2, or 3 may be included if they are considered.to be safety-related by the staff.
In Sections 3.9.2 and 3.9.3 of the Safety Evaluation Report, the staff discussed the design of safety-related pumps and valves in the Vogtle plant. The load combinations and stress limits used in the design of pumps and valves assure that the component pressure boundary integrity is maintained.
In addition, the applicant will periodically test and perform periodic measurements of all its safety related pumps and valves. These tests and measurements are performed in
'g' accordance with the rules of Section XI of the ASME Code. The tests verify 10/22/84 19 VCGTLE OSER SEC 3.6.2 INPUT
that these pumps and valves operate successfully when called upon. The periodic
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measurements are made of various parameters and compared to baseline measurements in'ordei to dete_ct long-term degradation of the pump or valve performance. The staff reviews the applicant's program for preservice and inservice testing of pumps and valves using the guidance of SRP Section 3.9.6, and gives particular attention to the completeness of the program and to those areas of the test program for which the applicant requests relief from the requirements of Section XI of the ASME Code. The applicant must provide a commitment that the inservice testing of ASME Class 1, 2, and 3 components will be,,in accordance with the rules of 10 CFR Section 50.55a, paragraph (g).
The applicant has n,ot yet submitted its program for the preservice and inservice testing of pumps and valves; therefore, the staff has not yet completed their review.
The staff will report the resolution of these issues in a supplement to this safety Evaluation Report.
There are several safety systems connected to the reactor coolant pressure Q
boundary that have design pressure below the rated reactor coolant system (RCS)
_./
. pressure. There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure system. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems.
Pressure isolation valves are required to be Category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME l
Code, except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action; i.e., shutdown or system isolation when the final approved leakage limits are not met. Also, surveil-(
.lanc6 requirements, which will state the acceptable leak rate testing frequency, l p..-,
'shall be provided in the technical specifications.
10/22/84 20 V0GTLE DSER SEC 3.6.2 INPUT I[.
a.
f^
Periodic _ leak testing of each pressure isolation valve is required tn be pc~
formed at least once per each refueling outage, after valve maintenance prior to'retuin to service, and for systems rated as les: than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.
Leak testing should also b'e performed after all disturbances to the valves are complete such as prior to reaching power opera-tion following a refueling outage, and maintenance.
The staff's position on leak rate limiting conditions for operation is that leak rates must be equal to or less than 1 gallon per minute (GPM) for each valve to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degrada-tion over a finite period of time.
Significant increases over this limiting value would be an indication of valve degradation from one test to another.
The Class Jeto Class 2 boundary will be considered the isolation point which must be protected by redundant isolation va1ves.
In cases where pressure p
isolation is provided by two valves, both will be independently leak tested.
When three or more valves p'rovide isolation, only'two of the valves need to be mj leak tested.
The applicant has provided~a list of Vogtle pressure isolation valves to be included in the leak rate testing program. Howevar, the applicant has not committed to the staff's position on acceptable leak rates. This is an open item and will be addressed in a supplement to this SER.
1 10/22/34 21 VCGTLE DSER SEC 3.6.2 INPUT
. '1; V0GTLE SER QUESTIONS SECTION 3.6.2 210.25 Provide a clarification of the criteria for the elimination of circum-ferential and longitudinal breaks as provided in sections 3.6.2.1.2.2.A.1 and 2 which appear to have axial and circumferential sEresses transposed.
210.26 Provide further explanation of paragraphs 3.6.2.3.2.2.B and 3.6.2.3.2.3.8 regarding permissible propagation of postulated breaks in the RCL and branch lines.
Specifically discuss how these criteria are satisified.
210.27 Section 3.6.2.3.4.1.B.5 refers to a dynamic increase factor.
Provide a discussion on how this factor is de,termined.
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J 210.28 SRP 3.6.2 states that rise times for fet thrust not exceeding one millisecond should be used unless justified. Address how break propa-gation and rise times were considered in postulated pipe breaks.
210.29 Address where, if any, limited area circumferential and longitudinal breaks have been assumed and the basis for assuring that those limited area breaks can not propagate to a full area break.
210.30 Provide the results of the evaluation of pipe whip and jet impingement effects on safety related systems, components, and structures as indicated in Table 3.6.2-2.
210.31 Breaks in non-nuclear high energy piping which are not seismically qualified should be postulated at those locations which produce the greatest effect on an essential component or structure.
Provide assur-ance that the above position has been satisified.
10/22/84 i
VCGTLE DSER QUESTIONS
rr-210.32 Provide assurance that 100% volumetric inservice examination of all pipe welds in the break exclusion regions will be conducted during each l inspecti,on interval as defined in IWA-2400, ASME Code,Section XI. The break exclusion regions should be clearly identified in the applicable sheets of Figure 3.6.1-1.
SECTION 3.7.3 210.33 Sections 3.7.B.1.3 reference possible use of damping values higher than those listed in Tables 3.7.B.1-1 and 3.7.N.1-1 if justified.
If higher damping values were used, provide justification.
210.34 Section 3.7.B.3.5 references possible use of an equivalent static factor of less than 1.5.
If such a factor is used, provide justifi-cation.
210.35 Provide a discussion of the design t:onsiderations used to assure that p
the dynamic effects associated with steam-hammer in the main steam l%
lines and water-hammer in the feedwater lines have been minimized.
210.36 The staff requests information regarding the analysis methods and assumptions used for the piping stress analyses of the main steam and feedwater piping. Provide the drawing details of the 5-way restraints and forged sections used in the main steam and feedwater systems.
-Include a discussion of the method used for the modelling of these 5 way restraints in the piping stress analyses, the extent to which those non-seismic portions of the main steam and feedwater piping are included in the seismic analytical model, and the design considerations given to assure adequate flexibility of the main steam and feedwater piping outside containment for the thermal expansion loadings.
210.37 In Table 3.7.N.1-1 a damping value of 4 percent for large diameter piping (12 inch NPS and greater) is indicated for the faulted condition and is not consistent with the staff position.
Regulatory Guide 1.61 recommends a damping value of 2 percent for 12 inch piping and 3 percent 10/22/84 2
VCGTLE OSER QUESTICNS L
f for piping greater than 12 inch diameter for the safe shutdown earth-
-quake.
Furthermore, the staff evaluation of the Westinghouse Topical
' Report WCAP-7921 AR as provided in a letter from D. B. Vassallo (NRC) to R. Salvatori (W) dated May 16, 1974 specifically stated that the higher damping value was approved for Westinghouse primary coo.lant loop system components and large piping with a configuration similar to that used in the test. Provide a list of piping systems for which the higher damping value was used and additional justification for your position.
SECTION 3.9.1 210.38 Section 3.9.B.1.1.
please provide a clarification of what is meant by design transients being " compatible" with those described in paragraph 3.9.N.1.1.
SECTION 3.9.2 A
210.39 Provide the procedures used for the design of piping anchors which reparate seismically designed piping and non-seismic Category I piping.
Include a discussion of the loads and load combinations used and how the local pipe wall stresses are considered.
210.40 Provide a summary of how piping vibration amplitudes measured during preoperational tests will be related to a stress level?
In addition, clarify what is meant by the " endurance limit as defined in the ASME Code,Section III."
Itisthesttg,positionthatallessentialsafety-relatedinstrumen-210.41 f
L tation liner should be included in the vibration monitoring program during pre-operational or start-up testing.
We require either a visual or instrumental inspection (as appropriate) be conducted to identify any excessive vibration that will result in fatigue failure. Provide a list of all safety-related small bore piping and instrumentation lines that will be included in the initial test vibration monitoring program.
10/22/84 3
V0GTLE D5ER QUESTIONS
e
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f' SECTION 3.9.3 21'0.'42 1 Provide,the basis for assuring that ASME Code Class 1, 2, and 3 piping systems in the NSSS scope are capable of performing their safety func-tion under all plant conditions. Describe 'the methodology used to assure the functional capability of essential piping systems when service levels C or D are specified.
210.43 Provide a discussion of the loading combinations used for the faulted condition in_ Table 3.9.B.3-1.
Specifically address whether LOCA loads (where appropriate) have been considered in the design of ASME Class 2 and 3 components and supports.
SECTION 3.9.3 210.44 Provide a discussion of the design sonsiderations used for safety and relief valve loads and piping react-ions.
Include in your discussion Q
(1) the basis for assuring that the valve end loads are acceptable, (2) the support arrangement for the affected piping, and (3) the methodology used to calculate the hydraulic transient forces in the piping due to valve blowdown.
210.45 The staff review of FSAR Section 3.9.8.3.4 and 3.9.N.3.4 finds that there is insufficient information regarding the design of component supports.
Per SRP'Section 3.9.3, our review includes an assessment of design and structural integrity of the supports:
(1) plate and shell, (2) linear, and (3) component standard types.
For each of the above
-three types of supports, provide the following information (as applicable) for our review:
(a) Describe (for typical support details) which part of the support is designed and constructed as component supports using Subsec-tion NF of the ASME Code and which part is designed and constructed as supplementary steel using the AISC Code, s
f
+'
10/22/24 4
V0GTLE DSER QUESTIONS
_ y.;
v (b) Provide the complete basis used for the design and construction of both the component support and the supplementary steel up to the butiding structure.
Include the applicable codes and standards used in the design, procurement, installation, examination, and inspection.
(c) Provide the loads, load combinations, and stress limits used for the component support and the supplementary steel up to the building structure.
(d) Provide the deformation Itmits used for the component supports.
(e) Describe the buckling criteria used for the design of component supports.
210146 Valve discs are considered part of the pressure boundary and as such should have allowable stress limits.
Provide these limits for our
. review.
w/
210.47 Due to a long history of problems dealing with inoperable and incor-rectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, installa-tion of snubbers should conform to the following procedures:
Pre-Service Examinaltra A pre-service examination should be made on all snubbers listed in Tables 3.7-4a and 3.7-4b of Standard Technical Specification 3/4.7.9.
This examination should be made after snubber installation but not more than six months prior to initial system preoperational testing, and should as a minimum verify the following:
(1) There are no visible signs of damage of impaired operability as a result of storage, handling, or installation.
,e, 10/22/84 5
VCGTLE OSER QUESTIONS
e (2) The snubber location, orientation, position setting, and con-figuration (attachments, extensions, etc.) are according to design dra, wings and specifications..
(3) Snubbers are not seized, frozen or jammed.
(4) Adequate swing clearance is provided to allow snubber movement.
(5). If applicable, fluid is to the recommended level and is not leaking from the snubber system.
(6) Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, and cotter pins are installed correctly.
If the period between the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situa-gT-)(
tions, re-examination of items 1, 4, and 5 shall be ' performed. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
Pre-Operational Testing Ouring pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250*F should be verified as follows:
(a) During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
(b) For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
10/22/84 5
V0GTLE OSER QUESTIONS O
m
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E (c) Verify the snubber swing clearance at specified heatup and cool-down intervals. Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
.The above described operability program for snubbers should be included and documented by the pre-service inspection and preoperational test programs.
.The pre-service inspection must be a prerequisite for the.preopera-tional testing of snubber thermal motion. This test program should be specified in Chapter 14 of the FSAR.
Confirm that Vogtle meets the above.
SECTION 3.9.6 210.48 There are several safety systems connected to the reactor coolant pres-sure boundary that have design pressure below the rated reactor coolant p' f '
system (RCS) pressure.
There are also some systems which are rated.at full reactor pressure on the discharge side of pumps but have pump l
suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems.
The leak tight integrity of each of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems.
. Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code expect as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action; i.e.,
~
shutdown or system isolation when the final approved leakage limits are not met. Also, surveillance requirements which will state the accept-7s able leak rate testing frequency shall be provided in the technical specificaticns.
10/22/84 7
VCGILE OSER QUESTIONS
e o F
Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve main-
' 'tenance, prior to return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is giv'en. The testing interval should average to be approximately one year.
Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute (GPM) for each valve to ensure the integrity of the valve, demonstrate the ade-quacy of the redundant pressure isolation function and give an indica-tion of valve degradation over a finite, period of time. Significant increases over this limiting value would be an indicattor, of valve degradation from one test to another.
(w The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves p. ovide isola-tion, only two of the valves need to be leak tested.
Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping and Instrument Ofagrams which describe your reactor coolant system pressure isolation valves. Also discuss in detail how your leak testing program will conform to the above staff position.
210.49 Provide a schedule for the completion of your program for inservice testing of pumps and valves. The program should contain any relief requests from ASME Section XI requirements together with the justifica-tion for requesting relief.
10/22/84 8
VCGTLE OSER QUESTIONS
.