ML20213E052

Forwards Input to SER Re Resolutions of Issues from 810928-30 & 1001 Meetings W/Burns & Roe in State of Ny. Final Response Will Indicate Results of Analysis on Support Base Flexability & Remainder of Work

ML20213E052
Person / Time
Site:	Columbia
Issue date:	03/03/1982
From:	Knight J Office of Nuclear Reactor Regulation
To:	Tedesco R Office of Nuclear Reactor Regulation
References
CON-WNP-0483, CON-WNP-483 NUDOCS 8203120067
Download: ML20213E052 (25)

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MAR  ? 1982 11Ef10RANDU:1 FOR: Robert L.' Tedesco, Assistant Director for Licensing, DL FROM: James P. Knight, Assistant Director for Components & Structures Engineering, DE p[ ,7 SUCJECT: SER IriPUT FOR WNP-2

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Plant !!ame: .l!! P-2 Licensing Stage: OL C' DL Branch and LP';t: LB 2, R. Auluck p DE Branch and Reviewer: fiEB, Y. Li . v Contract Laboratory and Reviewer: ETEC, J. Prevost '

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Requested CoTpletion Date: February 12, 1982 'i ,

Review Status: Safety Evaluation Report Input \

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The.!!echanical Engineering Branch has . completed its review of Sections 3.6.2, 3.7.3, and 3.9 (excluding " Seismic Qualification.of Safety-Related.flechanical Equipment" and " Pump and Valve Operability. Assurance", which is reviewed by EQB) of the Wr:P-2 FSAR through Amendment 20. Enclosed is our input to the Safety Evaluation Report. The input includes the resolutions of the open issues from the draft SER neeting held during September 28-30, October 1, 1981 at the Burns and Roe, Offices in !!cw York, The following issue is still considered to be open:

i SECTION 3.9.3.3 - COMP 0f;ENT SUPPORTS T.

The applicant has cormitted to a final response in the first quarter of 1932.

The final response .will indicate the results of the analysis on support base plate flexibility and the remainder of sampling and testing work. We will report our findings in a supplement to this SER.

The following issues are considered as confimatory items remaining in the f!EB scope of review:

l .' SER Section 3.6.2

a. The applicant is. currently conducting detailed piping system stress analyses and dynamic analyses of pipe whip and jet impingement. Fhe results of these studies.will.be supplied in the FSAR to complete Figures 3.6-11 through 3.6-36.

b. The applicant is to document analyses to detemine the finite 4

separation time for the broken pipe related to annulus pressurization.

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3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Ruoture of Piping General Design Criterion 4, " Environmental and Missile Design Bases," of 10 CFR Part 50, Appendix A, requires that structures, systems, and conponents important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant acci-dents. These structures, systems, and components shall be appropriately pro-tected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power plant.

Our review, conducted in accordance with Standard Review Plan Section 3.6.2,

" Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", of NUREG-0800, " Standard Review Plan", dated July 1981, pertains to the methodology used for protecting safety-related struc-tures, systems, and components against the effects of postul.ated pipe breaks both inside and outside containment. We also reviewed the criteria used to determine the location, size, and orientation of postulated failures and the methodology used to calculate the resultant pipe whip and jet impingement loads which might affect nearby safety-related structures, systems, and components.

Pipe whip need only be considered in those high energy piping systems having fluid reservoirs with sufficient capacity to develop a jet stream. The cri-teria for determining high and moderate energy lines is found in Branch Technical Position APCSB 3-1 of Standard Review Plan 3.6.1, " Plant Design for Protection Against Postulated Piping Failures in Fluid System Outside Containment". This criteria has been used correctly by the applicant in determining the list of all high energy systems.

The applicant has chosen to postulate preliminary pipe break locations on the basis of significant change in flexibility in high energy piping systems.

Examples of change in flexibility are pipe fittings (elbows, tees, and reducers) and circumferential connections to valves and flanges. This preliminary method of selection was chosen since it was conservative and most expedient, and did February 25, 1982 1 Y.C.LI/ Job D

not require the availability of detailed stress analyses of the piping systems.

The significance of the use of flexibility criteria for postulated pipe break locations is that it permitted the design of pipe whip restraints, and/or embed-ments for restraints, if required, at an early date in the project.

The applicant is currently conducting detailed piping system stress analyses and dynamic analyses of pipe whip and jet impingement. The results of these analyses will be used to determine final pipe break locations, final design of pipe whip restraints, and to evaluate the effect of jet impingement loads.

These studies will be in accordance with the criteria in Standard Review Plan, Section 3.6.2. The results of these studies will be supplied in the FSAR critm . to complete Figures 3.6-11 through 3.6-36. We will report the results of our review in a supplement to this SER.

Standard Review Plan 3.6.2 also sets forth certain criteria for the analysis and subsequent augmented inservice inspection requirements for high energy piping within the containment penetration break exclusion region. Breaks need 5 not be postulated in those portions of piping within the containment penetra-tion region that meet the requirements of the ASME Code,Section III, Subarticle NE-1120 and the additional requirements outlined in Branch Technical Position MEB 3-1 of Standard Review Plan 3.6.2. Augmented inservice inspection is required for those portions of piping within the break exclusion region. The applicant has committed to the additional inservice inspection that we require for these portions of piping. The criteria has been used correctly by the applicant.

In all analyses, except annulus pressurization analyses, full break with area equivalent to the pipe cross section is postulated to occur instantaneously.

In analyses related to annulus pressurization, the instantaneous approach is used in jet impingement and pipe whip restraint loads calculation. For the pressure time history, the recirculation line is postulated to break instan-taneously producing full blowdown force. Subsequently, the broken end is assumed to separate in a finite time based on momentum and energy consideration. These analyses will be documented in detail as an appendix to the FSAR in the New Loads update. We will report the results of our review in a supplement to this SER.

February 25, 1982 2 Y.C.LI/ Job D

Based on our review of FSAR Section 3.6.2 and subject to the satisfactory resolu-tion of the above identified items, our findings are as follows:

Our evaluation concludes that the pipe rupture postulation and the associated effects are adequately considered in the plant design, and, therefore, are 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 dy-namic effects of pipe whip and jet impingement provide adequate protection to the integrity and functionality of safet,y-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 sequential failures of safety-related piping, and emergency core cooling system per-formance will not be degraded by these dynamic effects.

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 provide adequate assurance that the structures, systems, and com-ponents 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 postu-YM l ated .r./?htYdwp4 of a high/ or moderate / energy piping systems inside or outside of containment.

3.9 Mechanical Systems and Components The review performed under Standard Review Plan Sections 3.9.1 through 3.9.6 of NUREG-0800, " Standard Review Plan", dated July 1981, pertains to the struc-tural integrity and functional capability of various safety-related mechanical February 25, 1982 3 Y.C.LI/ Job D

components in the plant. Our 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 standards other than the ASME Code. We review such issues as load combinations, allowable stresses, methods of analysis, summary of results, and preoperational testing. Our review must arrive at the conclusion that there is adequate assurance of a mechanical component performing its safety-related '

function under all postulated combinations of normal operating conditions, sys-tem operating transients, postulated pipe breaks, and seismic events.

3.9.1 Special Tooics for Mechanical Components The review of this section was performed following Standard Review Plan 3.9.1, "Special Topics for Mechanical Components." We have reviewed the design tran-sients and methods of analysis used for all seismic Category I components, component supports, core support structures and reactor internals designated as Class 1 and CS under the American Society of Mechanical Engineers Code,Section III, and those not covered by the Code. The~ assumptions and proce-dures used for the inclusion of transients in the fatigue evaluation of American Society of Mechanical Engineers Code Class 1 and CS have been reviewed. Our review also covered the computer programs used in the design and analysis of seismic Category I components and their supports and experi-mental and inelastic analytical techniques.

The applicant has provided a list of the design transients and the number of design cycles used for design. We have reviewed the list of transients and the number of cycles and find them to be acceptable. In the fatigue evaluation of NSSS components, the applicant has used ten peak OBE cycles. The SRP recom-mends five OBEs with a minimum of ten cycles per earthquake. The applicant has provided justification for using ten peak OBE cycles in a letter from R. Artigas, General Electric, to R. Bosnak, NRC, " Number of OBE Fatigue Cycles in the NSSS Design", dated September 17, 1981. In this generic study, G.E.

subjected a typcial BWR to three historically recorded earthquakes: El Centro, Taft and Golden Gate, and demonstrated the adequacy of ten peak OBE cycles for

. fatigue evaluation. In the letter from R. Artigas to R. Bosnak, dated December 3, 1981, it was shown that the design basis OBEs of WNP-2 are bounded February 25, 1982 4 Y.C.LI/ Job D

4 by those of the G.E. generic study. In addition, the contribution of the OBE cycles to the fatigue cumulative usuage factor has been shown to be negligible.

Based upon our review of the above information, and of the OBE as currently defined in the FSAR, we conclude that the ten peak OBE cycles used in the fatigue design of NSSS components provides a level of' safety equivalent to that provh(ed in the SRP for WNP-2 plant.

Computer programs were used in the analysis of many mechanical components. A list of the computer programs used in the static and dynamic analyses to determine the structural integrity and function capability of these components is included in the WNP-2 FSAR along with a brief description of each program. As stated in 10CFR Part 50, Appendix B, design control measures are required to verify the adequacy of the design of safety-related items. Methods of verification for all computer programs used in the design of safety-related mechanical items have been included in the FSAR. We have reviewed the list of computer codes and their methods of verification and find them acceptable.

Based upon our review of FSAR Section 3.9.1 our findings are as follows:

The staff concludes that the design trancients and resulting loads and load combinations with appropriate specified design and service limits for mechanical components is acceptable and meets the relevant requirements of General Design Criteria 1, 2, 14, 15, 10CFR Part 50, Appendix B, and 10CFR Part 100, Appendix B. This conclusion is based on the following.

1. The applicant has met the relevant requriements of General Design Criteria 14 and 16 by denonstrating 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 components and supports, and reactor 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 Criteria 2 and 10CFR Part 100, Appendix A by including seismic events in design transients which serve as design bases to withstand the effects of natural phenomena.

February 25, 1982 5 Y.C.LI/ Job D

3. The applicant has met the relevant requirements of 10CFR Part 50, Appendix B, and General Design Criteria 1 by having submitted information that demonstrates the applicability 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, Compenents, and Eouipment The staff has reviewed the criteria, testing procedures,,and dynamic analyses employed, to ensure the structural integrity and functionality of piping systems, mechanical equipment, and their supports undervibratory loadings. The principal document used in this review is SRP Section 3.9.2, " Dynamic Testing and Analysis of Systems, Components, and Equipment," Revision 2 dated July 1981. This review encompassed several areas, each of which is described below.

3.9.2.1 Pioing Preoperational Vibration and Dynamic Effects Testing As discussed in FSAR Section 3.9.2.1, Piping vibration, thermal expansion, and dynamic effects testing will be conducted during a preoperational testing pro-gram. 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 manner consistent with the design intent. During the WNP-2 plant's preopera-tional and startup testing program, the applicant will test various piping systems for abnormal steady-state or transient vibration and for restraint of thermal growth. Systems to be monitored include (1) ASME Code Class 1, 2.and 3 piping systems, (2) high energy piping systems inside seismic Category I structure, (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. The piping vibration test program will comply with the ASME Code,Section III paragraphs NB-3622.3, NC-3622.3, and ND-3622.3 February 25, 1982 6 Y.C.LI/ Job D

which require that the applicant be responsible, by observations during start-up or initial operations, for ensuring that the vibration of piping systemsis within acceptable levels. This vibration might be due to plant transients or might be associated with steady-state plant operation. This steady-state vibration, whether flow induced or caused by nearby vibrating machinery, could cause 108 or 109 cycles of stress in the pipe during its 40 year life. For this reason, the staff requires that the stress associated with steady-state vibration be limited to 50% of the alternating stress intensity, S, at 108 cycles, as defined in the ASME Code, Appendix I, Figures I-9.1 and I-9.2.

Both NSSS and BOP piping systems have two levels of criteria to evaluate vibra-tion and satisfy the above criteria. For trancient vibration, the applicant's Level 1 stress criteria as described in the FSAR, Section 3.9.2.1.5, Amendment 23, establishes the upper limits on vibration movement. If the vibration limit is exceeded, the test will be put on hold or terminated until the cause of the excessive vibration is identified and resolved. The Level 2 stress criteria demonstrates that the piping is responding in a manner ccnsistent with stress report predictions. Exceeding the Level 2 criteria will require the applicant to reevaluate the system based on test results. We have reviewed the NSSS and 80P limits of vibration and find that the applicant's procedures provides an equivalent level of safety to that provided in our acceptance criteria.

The test program should consist of a mixture of instrumented measurements and visual observation by qualified personnel. In addition, pipe whip restraint initial clearances will be checked, as will snubber response. The applicant will be recuired to provide a summary of the results of this test program upon its completion.

In a letter from R. L. Tedesco to R. Ferguson dated March 6, 1981, we conveyed our position to the applicant describing the requirement of pre-service examina-tion and pre-operational testing for snubbers to ensure snubber operability.

The applicant has responded to the issue in a letter from J. Shannon to R. Tedesco dated September 24, 1981. The preservice examination is detailed in the WNP-2 Preservice Inspection Program Plan)Section 9.3.1 and has met our requirement.

February 25, 1982 7 Y.C.LI/ Job D

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4 SM U t'S The pre-operational testing requirements for-uuMHrs will be detailed in Chapter 14 of the FSAR at a later date. This remains a confirmatory item pending completion of the detailed program. We will report the results of our review in a supplement to this SER. ,

Based upon our review of FSAR Section 3.9.2.1 and contingent upon the satisfactory resolution of the above identified item, we conclude that the applicant has met the relevant requirements of General Design Criteria 14 and 15 with respect to the design and testing of the reactor coolant pressure boundary. This assures that the probability of rapidly propagating failure and of gross rupts,re is low and 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 startup and initial operation on specified high- and moderate-energy piping, including all associated restraints and supports. The tests provide adequate assurance that the piping and piping supportthave been designed to withstand vibrational dynamic effects due to valve closures, pump trips, and other operating modes associated with the design basis flow conditions. In addition, the tests prov'dei assurance that adequate clearances and free movement of snubbers exist for unrestrained thermal movement of piping and supports during normal system heatup and cooldown operations. The planned tests will develop loads similar to those experienced during reactor operations.

3.9.2.2 Seismic Analysis The review performed under Standard Review Plan Section 3.9.2 included the appli-cant's analysis of all seismic Category I piping systems which was discussed in Section 3.7.3 of the FSAR. In addition to normal operating loads, this analysis also considers abnormal loadings such as an earthquake.

The scope of the review of the seismic system and subsystem analysis for the WNP-2 plant included the seismic analysis methods for all seismic Category I piping systems and components. It included review of procedures used for modeling and evaluating seismic Category I piping systems and components. The review included design criteria and procedures for evaluation of the interac-tion of nonseismic Category I piping with seismic Category I piping. The review also included seismic analysis procedures for reactor internals.

February 25, 1982 8 Y.C.LI/ Job D

The system and subsystem analyses are performed by the applicant on an elastic basis. Modal response spectrum multidegree of freedom and time history methods form the bases for the analyses of all major seismic Category I systems and components. When the response spectrum method is used, modal responses are combined by the square-root-of-the-sum-of-the-square rule. Closely spaced modes are combined using the double sum method as described in Regulatory Guide 1.92,

" Combining Modal Responses and Spatial Components in Seismic Response Analysis."

For both he time istory and r3sp onse J4 Aor s

Esspens a .I , % &pectrum

/ed rws*methods, the absolute sum (ABS)

% *.<*al ofgin tead of he square-root-cf-the-squares (SRSS) of three components of the earthquake motion as described in Regulatory Guide 1.92. The approach involves calculating separate responses for the two horizontal and one vertical components of motion. The largest horizontal response and the vertical response would than be combined using the absolute summation method. Comparisons of the results obtained from both the ABS and SRSS methods were made by the applicant and it was demonstrated that the ABS method used by the applicant is conservative.

In addition, the applicant has used lower values of damping than required by Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants." We find this approach conservative and provides an equivalent level of safety as that provided in the Regulatory Guide 1.92.

For the dynamic analysis of seismic Category I piping, each pipe line was idealized as a mathematical model consisting of lumped masses connected by elastic members. The stiffness matrix for the piping system was determined usingtheelfasticpropertiesofthepipe. This includes the effects of torsional, bending, shear, and axial deformations 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 spec-trum analysis was performed using the envelope response spectrum of all support points. In some cases, the multiple support excitation analyses methods have been used where separate acceleration time histories were applied to each piping system support points. We find the applicant's analysis methods to be acceptable.

February 25, 1982 9 Y.C.LI/ Job D

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In the application of the equivalent static load method on WNP-2, a conserva-tive static "g" loading was chosen for all piping systems irrespective of the I

building or building elevation. A study was conducted to demonstrate that the equivalent static analysis criteria employed on WNP-2 is conservative as com-pared to response spectrum analysis methods. Dynamic response spectra analysis was performed for several representative piping systems. Pipe stress and pipe support loads were calculated for these representative systems. Results were examined to confirm that pipe stresses are within allowables and pipe support loads are less than these calculated using the equivalent static load method.

We conclude that the equivalent static analysis criteria used in piping analy-sis on the WNP-2 project is conservative and provides an adequate basis for -

piping system design.

Based upon our review of FSAR Section 3.7.3, we conclude that the applicant '

has met the relevant requirements of General Design Criteria 2 with respect to demonstrating design a'equacy d of all Category I systems, components, equipment ,

and their supports to withstand earthquakes by meeting the regulatory posi-tions of Regulatory Guides 1.61 and 1.92 and by providing an acceptable seismic l systems analysis procedure and criteria. The scope of review of the seismic

• system analysis included the seismic analysis methods of all Category I systems, components equipment and their supports. It included review of procedures for modeling, inclusion of torsional effects, seismic analysis of Category I piping systems, seismic analysis of multiply-supported equipment and components with .

q distinct inputs, justification for the use of constant vertical static factors -

4 and determination of composite damping. The review has included design criteria and procedures for evaluation 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 and Category I buried piping outside ,

containment.

The system analyses are performed by the applicant on an elastic basis. Mocal response spectrum multidegree of freedom and time history methods form the >

bases for the analyses of all major Category I systems, components, equipment,

) and their supports. When the modal response spectrum method is used, governing

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February 25, 1982 10 Y.C.LI/ Job D

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E response parameters are combined by the square root of the sum of the squares rule. The double sum of the modal responses are used for modes with closely ra yonce spaced frequencies. The absolute sum of the largest horu.ontal c pa, = d d

the vertical response is used for both the time history and response spectrum methods. Floor spectra inputs to be used for design and test verifications of systems, components, equipment, and ttfjor supports are generated from the time history method, taking into account variation of parameters by peak widening.

A vertical seismic system dynamic analysis will be employed for all systems, and components, equipment and their supports where analyses show significant structural amplification in the vertical direction.

3.9.2.3 Preoperational Flow-Induced Vibration Testing of Reactor Internals 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.

Reactor internals for WNP-2 are substantially the same as the internals design configurations which have been tested in orototype BWR/4 plants. The caly exception is the jet pueps, which are of the BWR/5 design. The vibration measure-ment and inspection program has been conducted in the Tokai-2 plant, to verify the design of the jet pumps with respect to vibration. WNP-2 reactor internals will be tested in accordance with provisions of Regulatory Guice 1.20, Revision 2 for nonprototype, Category IV plants using Tokai-2 as the limited valid prototype.

The applicant has referenced G.E. Topical Report " Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants", NEDE-24057-P 3 0ctober 1977 which also contains information on the jet pump vibration measurement and inspection programs performed in the Tokai-2 plant. We have reviewed this report and find it to be acceptable.

Based upon our review of FSAR Section 3.9.2.4, and the Topical Report dis-cussed above, we conclude that the applicant has met the relevant requirements of General Design Criteria 1 and 4 with respect to the reactor internals being February 25, 1982 11 Y.C.LI/ Job D

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 regulatCry pcsitions of Regulatory Guide 1.20 ,

for tne conduct of preoperational vibration tests and by naving a preoperational ,

vibration program planned for the reactor internals which provides an acceptable ,

basis for veriffing the design adequacy cf these internals under test loading conditions comparable to those that will be experienced during operation. The combination af tests, predictive analysis, and post-test inspection provide adequate assarance that the reactor internals will, during their service lifetime, withstand the ficw-induced vibrations cf reactor without loss of P structural integrity. The integrity of the reactor internals in service is essential to assure the prcper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit safe reactor ,

operaticn and shutdown.

3.9.2.4 Dynamic Svstem Analysis of Reactor Internals Under Faulted Conditions The applicant has analyzed its 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 earthquake.

The results have not yet been documented in the FSAR. The results of this analysis including the effects of annulus pressurization will be included in the new loads evaluation program. Based on the results in the final documentation meeting the applicable acceptance criteria, we consider this Enproach acceptable.

We will report our finding in a supplement to this SER. I

! Based upon our review of FSAR Section 3.9.2.5 and contingent upon the satis-factory resolution of the above identified item, we conclude that the applicant hac 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 I the effects of earthquakes and the appropriate combinations of the effects of l normal and postulated accident conditions with the effects of the safe shutdown

earthquake (SSE) by having a dynamic system analysis to be performed which prcvides an acceptable basis for confirming the structural design adacuacy of i the reacter internals and unbroken piping loops to withstand the conD1ned dynamic l loads of postulated loss of ecclant accident (LOCA) and the SSE and the com-February 25, 1982 12 Y.C.LI/ Job D

_ . .. .. . . ~ _ . . . - _ - - _ -. .

bined loads of postulated main steam line rupture and SSE (for a BWR). The analysis provides adequate assurance that the combined stresses and strains in the components of th reactor coolant system and reactor internals will not exceed the allowable stress and strain limits for the materials of con-struction, and that the resulting deflections or displacements at any structural element /ofthereactorinternalswillnotdistortthereactorinternalsgeometry 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 combinations of component and system analyses are, therefore, acceptable. The assurance of structural integrity of the reactor internals under LOCA conditions for the most adverse postulated loading event provides added confidence that the design will withstand a spectrum of lesse~r pipe breaks and seismic loading events.

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 functional capability of pressure-retaining components, their supports, and core support structures which are designed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, or earlier industrial standards.

3.9.3.1 Loading Combinations, Desian Transients and Stress Limits l

The first area of review is the subject of load combinations methodology for response combinations and allowable stress limits. The applicant has evaluated all ASME Class 1, 2 and 3 components, component supports, core support structures, l

control rod drive components, and other reactor internals using the load combinations and stress limits provided in Section 3.9.3 of the FSAR. We have reviewed the load combinations and stress limits and, with one exception which '

l is discussed below, find them to be in agreement with our acceptance criteria ,

q It is the staff's position that the operating basis earthquake (OBE) and all

. safety relief valve (SRV) discharge load effects (initial actuation and sup-pression pool hydrodynamic loads) consistent with the 40 year design life of February 25, 1982 13 Y.C.LI/ Job D t

-. ._ . -_ _ __:--_.=. _ , _ - . . _ - _ _ - - - -- . . _ - - . . .:

the plant be evaluated, both separately and ccfocined, against ASME Code Service Level B r:equirements. In addition, for ASME Code Class 1 components, we require that the loads due to the SRV discharge and the OBE be comoined in the fatigue evaluation. In its ASME Class 1 fatigue evaluation, General Electric (GE) has considered these loads separately. GE maintains that its analysis has addi-tional conservatism that accounts for this load combination. In their fatigue analysis of the reactor vessel, vessel supports, and internals, GE assumes that connected piping loads are at the maximum allowable dynamic load es shSwn on their interface control documents. The OBE and the SRV events ar9 indecendently evaluated for fatigue considerations assuming that each event reaches the inter-face control document maximum allowable dynamic load. The piping designer is required to limit the combination of the OBE and SRV load on the RPV nozzle to the maximum allowable nozzle dynamic load. Thus, our requirement for treating the OBE and SRV loads in combination has been met. i The applicant has used the " square root of the sum of the squares" (SRSS) method for combining all dynamic responses. The use of the SRSS method for combining peak dynamic t~esponses due to the loss-of-coolant accident and the safe shut- ,

down earthquake has been accepted.by us in NUREG-0484 (Rev. 1). However, since the primary containment for the WNP-2 plant is a free standing steel pressure vessel and the plant is in a higher seismic zone, the staff requires that the criteria in Section 4 of the NUREG-0484, Rev.1, " Criteria for Combination of Dynamic Responses other than those of S$E and LOCA" be satisfied if the SRSS -

Mark III method is used. The applicant has committed to use data from the, generic A

program "M 'I to demonstrate ferg& free standing steel containment that an SRSS combination of dynamic responses achieves the 84% non-exceedance probability level and that this data is applicable to WNP-2. We will report the results of our review of this data in a supplement to this SER.

The staff requires that the functional capability of all pioing components in essential ASME Code Class 1, 2 and 3 piping systems designed to level C or D Service Limits be demonstrated. For both NSSS and BOP piping systems, the applicant ** has committed to demonstrate piping functional capability by using the screening criteria given in the General Electric Topical Report,

" Functional Capability Criteria for Essential Mark II Piping," (NED0-21985),

February 25, 1982 14 Y.C.LI/ Job D

dated September 1978. This topical report was approved by the staff in a letter from 2. Tedesco to. General Electric dated February 27, 1981.

The SWP; Mark II safety relief valve (SRV) and loss-of-coolant accident (LOCA) hydrodynamic icacs have been finalized as part of Unresolved Safety Issues A-8 and A-39, NUREG-0487, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," including Supplement 1 and Supplement 2 to NUREG-0487.

For the WNP-2 plant, the final accepted loads and stresses for the load combina-tions required by us are to be documented in the WNP-2 Design Assessment Report for SRV and LOCA loads. Subsequent to the final determinaton of acceptable salues for the WNP-2 plant hydrodynamic loads, the applicant will reconcile the leads used for the plant design with the final accepted loads. Provided the final load values do act exceed those values used in the plant design, we consider this approach acceptable. We will report our finding in a supplement to tris SER.

o Sisce the safety relief valve discharge piping and downcomers are ASME Class 2 and 3 domponcats, a fatigue analysis is not required in their design by Sectitr. III of the ASME Boiler and Pressure Vessel Code. A through wall leakage crack in the unsubmergad portion of these lines in the wet well resulting from fatigue caesec by f<V sctuations and small LOCA conditions would allow steam to bypass the pretst.Pe suppression pool. This could result in an unacceptable overprs.ssurization of the containment. We, therefore, required that the applicant perform a fatigue evaluation on these lines in accordance with the ASME Class 1 fatigue rules. Tne applicant has committed to perform a plant ur.ique Cia.5s 1 fatigue analysis on these lines. The results of this evaluation will be reported in the WNP-2 Design Assessment Report for SRV and LOCA loads.

Based on the applicant's resuits meeting the Class I fatigue rules, we find this approach acceptable We will report our finding in a supplement.to this 3ER.

NUREG-0619, "BWR Feedwater No221e and Control Rod Driver Return Line Nozzle Cracking " describes our technical position on the resolution of Generic Technical Activity A-10. "BWR Nozzle Cracking." The applicant has responded

, to this issue in a letter f rom G. Bouchey to A. Schwencer dated January 13, 1932. WNP-2 features tne welded sparger design. This design is specifically stated as acceptable in N'JREG-0619, Section 4.1, item (3).

Februa y 25, 1982 15 Y.C.LI/Jcb D

In addition, stainless' steel cladding was not installed in the WNP-2 reactor pressure vessel feedwater nozzle. For the control rod drive return nozzle, the applicant will cut and cap the control red drive line nozzle without rerouting the line. Since all of the above options are. allowed by NUREG-0619, we find the applicant's response acceptable from the_ mechanical design standpoint.

Additionally, we have contracted with the Energy Technology Engineering Center to perform an independent analysis of a sampip piping system in the WPPSS plant.

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This analysis will not only verify that the sample piping meets the applicable ASME Code . requirements, but will also provide a check on the applicant's ability to correctly model and analyze its piping systems. The results of the above evaluations will be presented in a future supplement to this SER.

Based on the above information in Section 3.9.3.1 of this SER and contingent upon the satisfactory resolution of the above identified items, we conclude that the applicant has met the requirements of 10 CFR Part 50, 950.55a and General Design Criteria 1, 2, and 4 with respect to the design and service load combina-tions and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 components by insuring that systems and components important to safety are desigt.ed.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 combinations of loadings as applied to ASME Code Class 1, 2, and 3 pressure retaining compo-nents in systems designed to meet seismic Category I standards are such as to provide assurance that in the ever,t of an earthquake affecting 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 loading combinations provides a conservative basis for the design of system components to withstand the most adverse combination of loading events without loss of structural integrity.

3.9.3.2 Desion and Installation of Pressure Relief Devices We have reviewed Section 3.9.3.3 of the FSAR relative to the design and installa-tion criteria applicable to the mounting of pressure relief devices used for February 25, 1982 16 Y.C.LI/ Job D

the overpressure protection of ASME Class 1, 2, and 3 safety and relief valves.

We have specifically reviewed the applicant's compliance with SRP 3.9.3.

Hydraulic transient forces in closed discharge systems including the main steam safety relief valve piping with discharge to the suppression pool have been analyzed using dynamic time history integration analysis methods. In the stress analysis of the main steam safety relief valve piping, it was assumed that all relief valves discharge concurrently, which is considered the most severe load case. Stresses were evaluated and applicable stress limits as stated in Section 3.9.3 of the FSAR were satisfied for all components of the run pipe and connecting systems. Consideration for effects due to the submerged leg of pipe in the suppression pool were included in the analysis.

Based upon our review of FSAR Section 3.9.3.3, we conclude that the applicant has met the requirements of 10 CFR Part 50, $50.55a and General Design Criteria 1, 2, and 3 overpressure relief devices by insuring that safety and relief valves and their installations are designed to standards which are commensurate wit,h their safety functions, and that they can accommodate the effects of discharge due to normal operation as well as postulated events such as loss-of-coolant accidents and tSt dynamic effects resulting from the safe shutdown earthquake.

The relevant requirements of General Design Criteria 14 and 15 are also met with respect to assuring that the reactor coolant pressure boundary design limits for normal operation, including anticipated operational occurrences, are not exceeded. The criteria used by the applicant in the design and installation of ASME Class 1, 2, and 3 safety and relief valves. provide adequate assurance that, under discharging conditions, the resulting stresses will not exceed allow-able stress and strain limits 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 instal-lation of the devices to withstand these loads without loss of structural integ-rity or impairment of the overpressure protection function.

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February 25, 1982 17 Y.C.LI/ Job D

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3.9.3.3 Component Supports The staff reviewed Section 3.9.3.4 of the FSAR relative to the criteria used by the applicant in the design of ASME Class 1, 2, and 3 component supports.

All component supports have been designed in accordance with Subsection NF of the ASME Code,Section III. It is the staff's position that for the design of component supports, the stresses produced by seismic anchor point motion of piping and the thermal expansion of piping should be categorized as primary stresses. Expansion stresses in the supports themselves may be categorized as secondary stresses. The applicant is committed to assess the design with respect to the NRC position and will identify supports not meeting our position and how they will be revised to comply. We will review the applicant's assessment and report our evaluation in a supplement of this SER.

The applicants' criteria pertaining to buckling of component supports and the design of bolts used in component supports is consistent with NRC Regulatory Guides 1.124, " Service Limits and Loading combinations for Class 1 Linear-Type Component Supports", and 1.130, " Service Limits and Loading Combinations for Class 1 plate - and - Shell Type Component Supports". We find this criteria to be acceptable.

The applicant has responded to IE Bulletin 79-02 for the WNP-2 plant to the Office of Inspection and Enforcement on June 21, 1979, August 8, 1980 and June 5,

(

1981. The applicant has committed to a final response in the first quarter of I

1982. The final response will indicate the results of the analysis on support base plate flexibility and the remainder of sampling and testing work. We have been requested by the Office of Inspection and Enforcement to review the i

applicant's response with respect to the pipe support baseplate flexibility I

and its effect on anchor bolt loads. We will report our findings in a supplement to this SER.

Based upon our review of FSAR Section 3.9.3.4 and contingent upon the satisfactory

( resolution of the above identified items, we conclude that the applicant has

- met the requirements of 10 CFR Part 50, 650.55a and General Design Criteria 1, l

i 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 l

February 25, 1982 18 Y.C.LI/ Job D l

supports by insuring 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 operation as well as postu-lated events such as loss-of-coolant accidents and the dynamic effects resulting from the safe shutdown earthquake. The combination of 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 as seismic Category I provide assurance 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 allewable 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 combination of loading events without loss of structural integrity.

Class CS component evaluation findings are covered in SER Section 3.9.5 in con-nection with reactor internals.

3.9.4 Control Rod Drive Systems Our review under Standard Review Plan Section 3.9.4 covers the design of the hydraulic control rod drive system up to its interface with the control rods.

j We reviewed the information in FSAR section 3.9.4 relative to the analyse.s and tests performed to assure ~ the structural integrity and functionality of this system during normal operation and under accident conditions. We also reviewed the life-cycle testing performed to demonstrate the reliability of the control rod drive system over its 40 year life. Based upon our review of the above information, we conclude 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 Part 50, 950.55a. This conclusion is based

.cn the following:

1. The applicant has met the requirement of GCC 1 and 10 CFR Fart 50, 650.55a, with respect to designing components important to safety to quality standards

February 25, 1982 19 Y.C.LI/ Job 0

t commensurate with the importance of the safety functions to be performed.

The design procedures and criteria used for the control rod drive system 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 earthquakes and anticipated normal operation occurrences with adequate margins to assure its reactivity control function and with extremely low probability of leakage or gross rupture of reactor coolant pressure boundary. The specified design transients, design and service loadings, combination of loads, and limiting the stresses and deformations under such loading combinations are in confor-mance 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 requirements of GDC 27 and 29 with respect to designing the control rod drive system to assure its capability of controlling reactiv-ity 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 Our review under Standard Review Plan 3.9.5 is concerned with the load combina-tions, allowable stress limits and other criteria used in the design of the WNP-2 reactor internals. Reactor internals have been designed in accordance with Subsection NG, " Core Support Structures," of the American Society of Mechan-ical Engineers Code,Section III using the loads, load combinations, and allowable stress limits as provided in Section 3.9.3 of WNP-2 FSAR. The description of the conf'iguration and general arrangement of the reactor internal structures has been reviewed and found to be complete.

Recently, cracking has been observed in BWR jet pump hold down beams. The appli-cant has responded to the issue in a letter from G. Bouchey to R. Tedesco dated December 4, 1980. The applicant has committed to reduce the preload on the February 25, 1982 20 Y.C.LI/ Job D

jet pump hold down beams from 30 kips to 25 kips in order to increase the beam operating time to a range of 19 to 40 years. Periodic inspection of the beams will be part of the applicant's inservice inspection program to be conducted at recommended intervals as determined by future testing at General Electric.

These inspections should provide adequate warning of potential beam failure.

Where excessive. cracking is identified by the inservice inspection, those beams will be replaced with improved heat treated beams purchased from GE. Tests indicate the improved beams may provide double the time to crack initiation as compared to the current beams.

Based on our review of the above procedure, we have found the resolution acceptable.

Based on our review of FSAR Section 3.9.5, we conclude that the design of reactor internals is acceptable and meets the requirements of General Design Criteria 1, 2, 4, and 10 and 10 CFR Part 50, 650.55a.

This conclusion is based on the following:

1.

The applicant has met the requirements' of GDC 1 and 10 CFR Part 50, 650.55a with respect to designing the reactor internals to quality standards commen-surate with the importance of the safety functi.ons to be performed. The design procedures and criteria used for the reactor internals are in con-formance with the requirements of Subsection NG of the ASME Code,Section III.

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 assure that capability to perform its safety functions is maintained and the specified

acceptable fuel design limits 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 provide reasonable assurance that in the event of an earth-quake or of a system transient during normal plant operation, the resulting deflections and associated stresses imposed on these structures and compo-nents would not exceed allowable stresses and deformation limits for the February 25, 1982 21 Y.C.LI/ Job 0

materials of construction. Limiting the stresses and deformations under such loading combinations provides an acceptable basis for the design of these structures and components 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 Pumos and Valves In Sections 3.9.2 and 3.9.3 of this SER, we discussed the design of safety-related pumps and valves in the WNP-2 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 accordance with the rules of Section XI of the ASME Code. The tests verify that these pumps and valves operate successfully when called upon. The periodic measurements are made of various parameters and compared to baseline measurements in order to detect 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 guidelines of SRP Section 3.9.6, and gives particular attention 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 has submitted its program for the inservice testing of pumps and valves. However, the applicant has not included all the required system drawings.

We have not completed our detail review o'f the applicant's inservice testing program. Based on the information in the applicant's program and flow diagrams in the WNP-2 FSAR, we find that it is impractical within the limitations of design, geometry, and accessibility for the applicant to meet some of the requirements of the American Society of Mechanical Engineers code. Imposition of those requirements would, result in hardships or unusual difficulties without a compensating increase in the level of quality or safety. The staff will consider these requests for relief from the pump and valve tasting requirements of 10 CFR 50, Section 50.55(g)(2) and g(4)(i) upon applicant's submission of all the required system drawings. This remains a confirmatory item pending l submission of the required system drawings. We will report the results of our l review in a supplement to this SER.

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February 25, 1982 22 Y.C.LI/ Job D

One area of concern during our review was the periodic leak testing of pressure isolation valves.

There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the sees 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 systems. The leaktight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems and thus cause an intersystem LOCA.

The applicant has provided a response to our concern regarding the periodic leak testing of the pressure isolation valves. We are currently reviewing the applicant's submittal. We will report the results of our review in a supplement to this SER.

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i February 25, 1982 23 Y.C.LI/ Job 0 l

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