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Forwards Safety Evaluation of SEP Topics III-6 Re Seismic Design Considerations & III-11 Re Component Integrity. Facility Has Adequate Seismic Capacity to Resist Postulated SSE
	ML20054L744
Person / Time
Site:	Millstone
Issue date:	06/30/1982
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Counsil W; NORTHEAST NUCLEAR ENERGY CO.
References
	TASK-03-06, TASK-03-11, TASK-3-11, TASK-3-6, TASK-RR LSO5-82-06-126, LSO5-82-6-126, NUDOCS 8207080427
	Download: ML20054L744 (61)
	v • d • e

{{#Wiki_filter:f June 30, 1982 Docket tio. 50-245 LS05-82 126 fir. W. G. Counsil, Vice President fluclear Engineering and Operations flortheast fluclear Energy Company Post Office Box 270 Hartford, Connecticut 06101

Dear fir. Counsil:

SUBJECT:

SEP SAFETY TOPICS III-6, SEISMIC DESIGfi C0flSIDERATIONS AND III-ll, C0f!P0fiENT IfiTEGRITY - !!ILLST0flE NUCLEAR POWER STATI0fi UNIT 1 We have completed our seismic review of Millstone Nuclear Power Station Unit 1. Enclosed is a copy of our combined safety evaluation report of the two subject topics. As discussed in this report, four equipment items (1-the structural integrity of heat exchangers, 2-the structural integrity of motor operated valves attached to small piping, 3-the structural integrity of transformer, and 4-the structural integrity of control room ~ electrical panels) remain open due to lack of design infomation. In addition, the staff is unable to conclude that the turbine building pile foundations are capable of withstanding the postulated SSE loads because design information of pile foundations was not available to the staff for review. A supplement to this report will be issued after the review of your responses to these five open items is completed. 5 6CO This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as p ss6 built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if flRC criteria go.' relating to this topic are modified before the integrated assessment is completed. (,. 5 c Ef T. C 4 6-} 8207080427 820630 DR ADOCK 05000245 PDR OF F6CE ) susw4us ) omo OFFICIAL RECORD COPY usam mi-us. co enc ronu m om useu ma

k .. 9a4 !!r. W.G. Counsil Your response is requested within 30 days of receipt of this letter. If no response is received within that time, we will assume that you have no colmlents or corrections. Sincerely, hriginni cicned by! Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing Enclosure : As stated cc w/ enclosure: See next page l l lih 8l .... T. E PB../..B...C ..D..L..:. 0. RB [5../..P. M D..L... B.. C... ...D.. iA D. A.... .V...L...:.. f..e. omen > .D..L..:.S E P.B....,.#c...D..L..:.S E PB../..S. L. 4.,.. DL..: T...(..'. [.,... a. DPe rsi nko':'sh RHermann WRusselll JShea DCr fi eld Gla s sumue > 6/..r..l./. 82 6/77/82 6 /... /.82 6 /.p.../. 82 6L.t /. 8. 2 6..././..c.../.82 .. g. /...>...., f)... oun> OFFICIAL RECORD COPY uso m usi-sw u nne ronu ais o>soi nncu uaa

t Mr. W. G. Counsil CC William H. Cuddy, Esquire State of Connecticut Day, Berry & Howard Office of Policy & Management Counselors at Law ATTN: Under Secretary Energy One Constitution Plaza Division Hartford, Connecticut 06103 80 Washington Street Hartford, Connecticut 06115 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission Region I Office 631 Park Avenue King of Prussia, Pennsylvania. 19406 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant P. O. Box 128 .Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC P. O. Box Drawer KK Niantic, Connecticut 06357 ~ ' Ffrst-Selectman of the Town of Waterford Hall of Records ~ 200 Boston Post Road Waterford, Connecticut 06385 John F. Opeka Systems Superintendent Northeast Utilities Service Company I - P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency ' Region I Office ' ATTN:. Regional Radiation Representative ~ JFK Federal Building .f: Boston, Massachusetts 02203 } h g l 9 a g Q

f S SYSTEMATIC EVALUATION PROGRAM TOPICS III-6 & III-ll MILLSTONE NUCLEAR POWER STATION 1 TOPICS: III-6, SEISMIC DESIGN CONSIDERATIONS III-ll, COMPONENT INTEGRITY I. INTRODUCTION The nuclear power plant facilities under review in the SEP received con-struction permits between 1956 and 1967. Seismic design procedures evolved significantly during and after this period. The Standard Review Plan (SRP) first issued in 1975, along with the Regulations 10 CFR Part 50, Appendix A and 10 CFR Part 100. Appendix A constitute current licensing criteria for seismic design reviews. As a result, the original seismic design of the SEP facilities vary in degree from the Uniform Building Code up through and approaching current standards. Recognizing this evolution, the staff found that it is necessary to make a reassessment of the seismic safety of these plants. Under SEP seismic reevaluation, these eleven plants were categorized into two groups based upon the original seismic design and the availability of seismic design documentation. Different approaches were used to review the plant facilities in each group. The approaches were: Group I: Detailed NRC review of existing seismic design documents with limited reevaluation of the existing facility to i confirm judgments on the adequacy of the original design with respect to current requirements. Licensee were required to reanalyze their facilities and Group II: to upgrade, if necessary, the seismic capacity of their facility. The staff will review the licensee's reanalysis methods, scope and results. Limited independent NRC analysis will be performed to confirm the adequacy of the licensee's method and results.

I s s , Based upon the staff's assessment of the originai seismic design; the Millstone 1 Plant was placed in Group I for review. The Millstone 1 Plant, a Mark I boiling water reactor (BWR), is located on the Atlantic coast, about 5 miles southwest of New London, Connecticut. General Electric Company (GE), the primary contractor for the plant, I engaged EBASCO Services, Incorporated, as their architect-engineer. Most seismic analysis of the plant were performed by John A. Blume and Associates, Engineers. The plant received its construction permit on May 19, 1966 and provisional operating license on October 26, 1970. The licensee filed its application for a full-term operating license on September 1,1972. The Millstone 1 plant was originally designed for an earthquake (equivalent to the operating basis carthquake or OBE) with a horizontal peak ground acceleration (HPGA) 0.079 and reviewed for an earthquake (equivalent to the safe shutdown earthquake, or SSE) with a PGA of 0.179 A smoothed design response spectrum recomended by John Blume and Associates and the North 69 West component of the 1952 Taft earthquake record normalized to the specified PGAs were used as seismic input for the analyses and design. The vertical component of ground motion was assumed to be two-thirds of the horizontal components. For the dynamic analyses of Seismic Class I Structures, the buildings (or structures) were modell'ed as lumped mass-spring systems with fixed base to simulate the rock founded foundations. Different analysis methods were used to obtain the dynamic responses of structures: (1) reactor building, drywell, ventilation stack, radwaste/ building / control room and condensate storage tank were analyzed by time history approach,

9 , (2) gas turbine building and suppression chamber were analyzed by response spectrum method, and (3) turbine building and intake structure were analyzed by equivalent static approach. 'Two methods were used for the analysis of safety related piping systems and equipment: (1) the response spectrum analysis approach with smoothed response spectrum recommended by John Blume and Associates as input, and (2) the equivalent static method using peak structural responses as input. Chapter 4 of the NRC NUREG/CR-2024 report, " Seismic Review of the Millstone 1 Nuclear Power Plant as Part of the Systematic Evaluation Program" (Reference 1), summaries the details of the original analysis and design. The SEP seismic review of Millstone 1 facility addressed only the Safe Shutdown Earthquake, since it represents the most severe event that must be considered in the plant design. The scope of the review included three major areas: the integrity of the reactor coolant pressure boundary; the integrity of fluid and electrical distribution systems related to safe shutdown; and the integrity of mechanical and electrical equipment and engineered safety features systems (including containment). A detailed review of the facilities was not conducted by the staff; rather our evalua-l tions relied upon sampling representative structures, systems and components. Confirmatory analyses using a conservative seismic input were performed for the sampled structures, systems and components. The results of these analyses served as the principal input for our evaluation of the seismic capacity of the facility. t L

6 9 , II. REVIEW CRITERIA Since the SEP plants were not designed to current codes, standards, and NRC requirements, it was necessary to perform "more realistic" or "best estimate" assessments of the seismic capacity of the facility and to consider the conservatisms associated with original analysis methods and design criteria. A set of review criteria and guidelines was developed for the SEP plants. These review criteria and guidelines are described in the following documents: 1. NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," by N. M. Newmark and W. J. Hall, May 1978. 2. "SEP Guidelines for Soil-Structure Interaction Review," by SEP Senior Seismic Review Team, December 8,1980. For the cases that are not covered by the criteria stated above, the follow-ing SRPs and Regulatory Guides were used for the review: 1. Standard Review Plan, Sections 2.5, 3. 7, 3.8, 3. 9, and 3.10, 2. Regulatory Guides 1. 26, 1. 29, 1. 60, 1. 61, 1. 92, 1.100, and 1.122. III. RELATED TOPICS AND INTERFACES The related SEP topics to the review of seismic design considerations and component integrity are 11-4, II-4. A. II-4.B. II-4.C. These topics relate to specification of seismic hazard at the site, i.e., site specific ground response spectrum for the Millstone 1 site. The seismic input selected for the confirmatory analysis of Millstone 1 facility, namely the Regulatory Guide 1.60 spectrum scaled to 0.2g peak ground acceleration, envelopes the i

. Millstone I site specific ground response spectrum as shown in Figure 1. The results for these four safety topic evaluation confirm the con-servatism of using a 0.2g Regulatory Guide 1.60 spectra for reevaluation of structures, systems and components. IV. EVALUATION A. General Approach The seismic reevaluation of Millstone 1 Nuclear Power Plant was / initiated by conducting a detailed review of the plant seismic documentation. The results of this review are summarized in the draft report, " Seismic Review of Millstone 1 Nuclear Power Plant - Phase I Then, the staff and our consultants conducted a site-visit. Report " The purposes of this site-visit were: (1) to observe the as-built plant specific feature relative to the seismic design of the facility, (2) to obtain seismic design information which was not available to the staff in the docket, (3) to discuss, with the licensee, seismic design information that the staff and our consultans had reviewed, and (4) based on the results of this field inspection, experience and judg-ment, to identify sample structures, systems and components for which The the confirmatory analyses (or audit analyses) would be performed. results of these analyses, then, served as the basis for safety assess-ment of the plant facility. When a structure was evaluated, it was judged adequately desigred if the results from the structural analysis met one of the following three criteria:

j S - 1. The loads generated from confinnatory analysis were less than original loads; The seismic stresses from confirmatory analysis were low compared 2. to the yield stress of steel or the compressive strength of concrete; and 3. The seismic stresses from confirmatory analysis exceeded the steel yield stress or the concrete compressive strength, but estimated reserved capacity (or ductility) of the structure was such that inelastic deformation without failure would be expected. If the above criteria were not satisfied, a more comprehensive reanalysis was required to demonstrate its design adequacy. [n For piping reevaluation, the results from the audit analysis,of each of the sampled piping systems were compared with ASME Code requirements N This for Class 2 piping systems at appropriate service conditions. ] comparison provided the basis for reevaluating the structural adequacy Because limited documentation exists regarding of piping systems. the original specifications applicable to procurement of equipment, as well as for the qualification of the equipment, the seismic review ~ Two levels of equipment was based on expert experience and judgment. of qualification were performed, structural integrity and operability. l The results of this reevaluation of equipment served as the basis for modifications or reanalysis to be undertaken by the licensee. 2 B. Confirmatory Analysis In order to provide independent analytical results for the reevaluation, a relatively complete seismic confinnatory analysis, which started with a definition of seismic input ground motion and ended with responses of the safety related structures and selected systems and components, 1 The analysis during the postulated earthquake event, was performed. procedures and results are briefly discussed on the following sections. I /

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" ' ~, ' ..I ' l. Seismic Input When seismic review of Millstone 1 plant started in mid 1979, ~ -the 4 site specific ground response spectra were not available. '~~~ s - O"',In order to, perform the review on a sampling basis that could

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be applied wi1!h confidence, a more conservative ground motion, namely Regulatory Guide 1.60Jorizontal ground response spectrum 2' (R. G.1.60 sppctrum) scaled-to,. 0.2g was used fas, the horizontal s component of postulated grou6d motion for analysis.' The input a motion in the vertical directio.n was taken as/2/3 of the value ~. ~. _, L' s inhorizontaldirectioiPacio}3'IYentire[f[equencyrange. th w..e Recently, the.' site specifit 3pectra developmant program was completed, and the spectrum generated for the Millstone 1 site s s wAs issued to the licensee on June 17,1981 (Reference 2) for ,w e '~ a., C The basis for the develop-C'anyfutureworkthatmaybe, required. fy ' ~b. ment of site specific spectra was documented in NRC NUREG/CR-A 1582, report, rhismic Hazard' Analysis",(Reference 3). This site \\ 4_x 3;[ _, [j s'phific,spectium is appropriate-for essessing the actual safety + y ~. margins 'present for any structures, systems and components that ~ ~ havebeenidentifiedas, nap. items. In Figure 1, a comparison i is made for thg.; ground response spectra that were used for the N. ,a original plant design and for SEP seismic reevaluation (Regulatory e 1 Guide 1.60 spectrum and the site specific spectra). ,-.s; e s ,~ e, f ..y e =* y r r.'- e ,x ~ ,i V ~,_ ~ f ~ - f'3 I u

, 2. Acceptance Criteria and Scope The specific SEP reevaluation criteria are documented in NUREG/CR-0098 and SEP Guidel,ines for Soil-Structures Interaction Review. These documents provide guidance for: a) Selection of the earthquake hazard b) Design seismic loadings c) Soil-structure interaction d) Damping and energy absorption e) Methods of dynamic analysis f) Review analysis and design procedures g) Special topics such as under ground piping, tanks, and vaults, equipment qualification, etc. These criteria are felt to more accurately represent the actual stress level in structures, systems and components during a postulated earthquake event and consider, to certain extent, nonlinear behavior of the systems. The SEP seismic reevaluation of Millstone 1 facility was a limited review centering on: (1) Assessment of the general integrity of the reactor coolant 1 pressure boundary. Evaluation of the capability of essential structures, systems (2) and components required to shutdown the reactor safely and to maintain it in a safe shutdown condition (including 'the capability for removal of residual heat) during and after a postulated seismic event.

9-A total of five (5) structures, four (4) piping systems and seventeen (17) equipment components (mechanical and electrical) were fully evaluated and several others samples were evaluated on a limited basis in this work. They are: (1) Structures - reactor building, drywell, radwaste/cor. trol building ventilation stack, and turbine building. (2) Piping systems - feedwater, shutdown cooling, condensate transfer, and diesel oil lines. (3) Equipment - 9 mechanical equipment items and 8 electrical equipment items. Auxiliary structures - condensate storage tank, underground (4) piping, buried tank, suppression chamber, and gas turbine building. 3. Analysis of Structures Analytical procedures and methods conforming with the current state of the art were used. These procedures and methods considered the three dimensional dynamic response effects of 8 buildings, interaction between buildings, equipment masses, structural damping in accordance with calculated stress levels, and so forth. (A) Analysis of Reactor Building /Drywell Complex The reactor building /drywell complex (including founda-tion) was modelled as two cantilever lumped mass-spring closely coupled systems with fixed base to simulate the rock foundation. Because of the high degree of asymmetry of this building, a fully three dimensional model was developed for simulating the dynamic behavior of the building during the postulated SSE. The input ground

. motion, R. G. 1.60 spectrum scaled to 0.29, was defined at free field ground surface and was input directly to the fixed base of the model. The response spectrum analysis approach conformed with the NRC Standard Review Plan (SRP) requirements in that a combination of modal and directional responses, etc. was used to generate the structural responses. The details of analyses and final results are summarized in Reference 1. The time history analysis approach together with an artificial time history record (acceleration) normalized to the same PGA, namely 0.29, was used for generator in-structure (or floor) response spectra. After the peaks were broadened + 15% of corresponding frequency in accordance with Regulatory Guide 1.122, the smoothed response spectra were used as input motions for the evaluation of piping systems and equipment. Appendix A of Reference 1 contains a summary of all the generated in-structure response spectra. The results of these evaluations showed that reactor building and drywell are capable to withstand the postulated seismic l event. (B) Analysis of Radwaste/ Control Building The original analysis model, with the modifications by adding actual eccentricity to each floor mass to simulate the three dimensional response of the building, was used. The same acceptance criteria and analytical approach used i

for reactor building /drywell complex were used to evaluate this building. The details of modelling techniques, analysis procedures and analysis results (dynamic forces used for structural evaluation and in-structure response spectra used for equipment and piping evaluation) are found in Chapter 5 and Appendix A of reference 1. The results of evaluation showed that the radwaste/ control building is capable to withstand the postulated seismic event. (C) Analysis of Turbine Building Since dynamic analyses were not originally performed by the licensee for this building, a new three dimensional lumped, mass-spring model with fixed base to acccount for the partially rock founded and partially pile i supported foundations was developed for the confirmatory analysis. The response spectrum analysis method was used to evaluate the dynamic responses (dynamic shear, moments, and axial forces) to seismic excitations of the building and the time history analysis method was applied for generating in-structure (or floor) response spectra. The details of modelling techniques, analysis criteria, analyses procedures, and results are found in Reference 1. The results of this evaluation showed that the turbine building is capable of withstanding the postulated seismic event.

9 . However, since no information was provided by the licensee to demonstrate the design adequacy of piles, especially the pile cab, for resisting the lateral forces due to the postulated SSE as discussed in the SEP Topic II-4.F evaluation report (Reference 5), the staff is unable to conclude that the turbine building will withstand the postulated SSE loads until the concern about pile design adequacy is resolved under SEP Topic II-4.F review. 4. Analysis of Auxiliary Structures (A) Ventilation Stack The 386 ft. ventilation stack was modelled as a two dimensional, fixed base, lumped ness-spring system

and was analyzed by response spectrum analysis approach using R. G.1.60 spectrum as input motion. The details of modelling techniques, analysis procedures, and results are summarized in Reference 1 The results of this evalua-tion showed that the stack is capable to withstand the postulated SSE loads.

The original analysis model was checked and used for the

Note:

confirmatory analysis. t 6

(B) Condensate Storage Tank A number of potential failure modes of the tank had been evaluated under this review. Among these were: buckling of the sidewall due to the overturning moment; yielding; fracture or pull-out of the anchor bolts; sliding at the base with subsequent rupture of connections; and failure due to high tensile loop stresses as a result of the hydrodynamic pressure occurring simultaneously with hydrostatic pressure. An equivalent static analysis with 0.29 R. G. 1.60 spectrum as input ground motion was performed for this tank. The detailed analyses and evalua-tion are described in Reference 1. The results of this evaluation showed that the above critical elements of the tank are adequately desianed to withstand the postulated SSE loads. (C) Buried Tank and Underground Piping Since no information (soil data, original analysis results, etc. ) was available, some conservative assumptions had been made for the confirmatory analysis. The equivalent i static analysis method, including the wave passage effects, l was used for the evaluation. The detailed analysis and evaluation were described in Reference 1. The results of the evaluation showed that the underground piping and buried tank are capable to withstand the postulated SSE loads.

. (D) Analysis of Suppression Chamber (Ring Header Torus, and Support System) Based on the criteria discussed in NRC NUREG/CR-0098, an audit analy' sis, with R. G. spectrum anchored to 0.29 as input ground motion, was performed for this system. The detailed analysis and evaluation were described in reference 1. The results of this evaluation showed that the suppression Chamber system is capable to withstand the postulated SSE loads. However, current licensing criteria would require that the design of this system be based on an evaluation which includes both suppression pool dynamic loads due to LOCA and seismic loads. This review is a part of Unresolved Safety Issue (USI) A-7 which considers the safety of suppression pool structures under the combined load of SSE and LOCA as required by the FSAR. Since the suppression pool is supported on the base mat (about 25 ft. below ground surface) of rock founded reactor building and the peak ground acceleration specified in FSAR is only 10% lower than that for Millstone 9 vs 0.19 ), it is the 1 site specific spectrum (0.17 9 staff's judgement that the resolution about this issue under USI A-7 review will satisfy SEP concern and this item is considered closed. i l i

i

(E) Gas Turbine Building The gas turbine building is a reinforced concrete structure founded on top of piles which were driven to competent The building, modelled as a single-degree-of-freedom rock.

system with fixed base, was originally analyzed by equivalent static method using the response spectrum generated from 1952 Taft N69 W earthquake record normalized to 0.079 i (5% damping) as input motion. No effects of the flexibility Since of the overburden and pile system were considered. no information of supporting media (soil data and pile system) was available for review, the evaluation was performed based on a " load comparison" approach by assuming no frequency shift or amplification resulting from overburden flexibility. (See Reference 1). The preliminary results of this evaluation recommended that the effects of foundation flexibility should be considered before a conclusion is reached for the adequacy of the building. After the incorporation of information provided by the licensee, Reference 6, reevaluation results demonstrate that this building is capable to withstand the postulated SSE loads ( Attachment 1).

-le-5. Analysis of Piping Systeas As discussed in the subsection B.2 above, four piping systems were sampled and analyzed to verify the adequacy of the original design. The piping systems selected were portions of the feedwater, shutdown cooling, condensate transfer, and diesel oil lines. The selections were based on: (1) the expert's judgment and observations during the walkdown of the facility, (2) review of the original analyses and design, and (3) a desire to provide a range of piping sizes. Audit analyses which incorporated current ASME Code and Regulatory Guide Crituia and used the floor response spectra as input motion were performed for each portion of piping system selected. The results from these analyses were compared to ASME Code requirements for Class 2 piping systems at the appropriate service conditions. This comparison provided the bases for assessing the structural adequacy of the piping under the postulated seismic loading condition. Assumptions made for the analysis, methodology employed and detailed preliminary results are found in the INEL report (Reference 4). The preliminary results of confirmatory analysis showed that some locations of piping systems were found to be overstressed and some relatively large deflections were identified under the postulated seismic loading. After the incorporation of additional

, information which reflected the modifications made in response to IE Bulletin 79-14 (Reference 7 ), revised audit analysis was performed for three of the sampled piping systems. The condensate transfer line, which was not originally classified as safety related piping, was not reanalyzed by the staff. However, it is being upgraded by the licensee. The results of the revised audit analyses demonstrate that the sampled piping systems with the modifications due to IE Bulletins 79-02 and 79-14 (the installation of required modifications is con-tinuing and is scheduled for the completion by early 1983) are capable of withstanding the postulated SSE loads. 6. Analyses of Selected Mechanical and Electric Equipment The evaluation of equipment was done on sampling basis. Safety related components required for safe shutdown, the primary pressure boundary, and engineering safeguard features were categorized as active or passive and as rigid or flexible according to the criteria in R. G. 1.45 and SRP 3.9.3. A representative sample (or samples) from each ground was selected and evaluated to determine the seismic design margin or adequacy of each group. In this way, groups of similar components were evaluated without the need for detailed reevalua-tions of all individual components. The sampled mechanical i

, and electrical equipment items and the basis for this sampling are described in Table 1 below: TABLE 1: Mechanical and electrical components selected by the review team for seismic evaluation and the basis for selection. Item No. Description Reason for selection Mechanical Components 1. Emergency service water pump This item has a long, vertical unsupported intake section which may be limiting for lateral loads which are seismically induced. In addition, material may be cast iron. 2. Isolation (emergency) condenser This item is a horizontally mounted component supported by three saddles that do not appear to be seismically restrained except at the center support. Concern was expressed about the saddle's ability to carry required seismic loads, particularly in the longitudinal direction. 3. Shutdown heat exchanger Horizontally mounted leading to concern regarding capability to carry the load in longitudinal direction. l 4. LPCI/ Containment Spray Heat This item is vertically mounted Exchanger and may not be adequately restrained. 5. Recirculation pep support This item is a vertical component supported by spring hangers whose functionality is critical in insuring reactor coolant system integrity. 6. Emergency diesel oil Anchor-bolt system for in-structure storage day tank flat-bottom tanks that are l flexible may be overstressed if tank and fluid contents were assmed rigid in the original analysi s. -2

l _19 7. Motor-operated valves A general concern with respect to air and electric motor,-operated valves, particularly for lines 4 in or less in diameter, is that the relatively large eccentric mass of the motor, when not externally supported, will cause excessive stresses in the attached piping. In addition, overstress and excess deformation of valve yoke and sten may also occur. 8. CRD hydraulic control system Item is particularly critical to including tubing and supports insure reactor coolant systen integrity. 9. Reactor vessel, supports, Sane as Iten 8. and internals Electrical Components 10. Battery racks The bracing required to develop lateral load capacity may not be sufficient to carry the seismic load. 11. Motor control centers Typical seismically qualified electrical equipment. Functional design adequacy may not have been demonstrated. In addition, anchorage to support structure may not be adequate and might permit sliding er overturning during a seismic event. 12. Transf omers Same as Item 11. 13. Switchgear panels Same as Item 11. 14. Control room electrical The control panels appear panel s adequately anchored at the base. However, there appear to be many components cantilevered off the front panel, and the lack of front panel stiffness may permit significant seismic response of the panel, resulting in high acceleration of the attached components. 15. Diesel generator remote Same as Iten 11. control board 16. Battery room distribution Same as Iten 11. panels 17. Electri . ble raceways The cable tray support system does not appear to have positive lateral-restraint load-carrying capacity.

O . The licensee was asked to provide seismic qualifications data for each sampled component including design drawings, specifica-tions, and design calculations. After a detailed evaluation of each component was completed, conclusions were drawn as to the overall seismic capacity of the safety related equipment at the Millstone 1 facility. The description of selected components, analytical procedures and evaluations are found in Reference 1. A total of 12 open items (structural integrity and/or operability) out of 17 sampled equipment items were identified in Reference 1. Most of these items remain open due to lack of design information. After the review and incorporation of additional information provided by the licensee (References 8 thru 12 and attachment 2), the results of evaluation are summarized below: (a) Seven (7) mechanical equipment items and one (1) electrical equipment item were found to be adequately designed. (b) The structural integrity of LPCI/ containment spray beat exchanger remains open due to lack of design information. (c) The structural integrity of motor operated valves attached of small piping (4" D and smaller) remain open due to lack of design information. l (d) The structural integrity of motor control centers, switch-gear panels, diesel generator remote control boards and battery room distribution panels is considered to be capable of withstanding postulated SSE loads. (e) The structural integrity of transformers and control room electrical panels remains open due to lack of design information about the modified anchorage systems. (f) The structural integrity of internal mounted electrical devices and operability of all safety related electrical equipment were not evaluated.

(g) Qualification of electrical cable trays is being evaluated by testing through SEP Owners Group program. This program is scheduled for completion by December of 1982. V. CONCLUSION Based on the review of the original design analyses, the results of con-firmctary analyses performed by the staff and its consultants, and the licensee's responses to the SEP seismic related safety issues, the following conclusions can be drawn: Structures - All safety related structures, structural elements and auxiliary structures of the Millstone 1 facility are adequately designed to resist the postulated SSE loads considering that the design adequacy of pile foundations under turbine building will be demonstrated through SEP Topic II-4.F review. Piping Systems - According to the results of revised SEP piping audit analysis (Attachment 3), all sampled piping systems with the modification due to IE Bulletins 79-02 and 79-14 are found to be capable of withstanding the postulated SSE loads. The modifications to all safety related piping are scheduled for completion by early 1983. Therefore, the staff concludes that all safety related piping systems are capable of withstanding the postulated SSE loads pending on the completion of all The necessary modifications required by these Bulletins. staff also concluded that the plant can continue to operate between now and early 1983 based on the operability review under these Bulletins. Electrical Equipment - As a result of SEP seismic review, three (3) activities have been or are being completed by the licensee: upgrading of anchorage and support of all safety related a)electrical equipment required by NRC letters dated January 1, 28, 1980 (References 13 and 14) has been completed, and July and found to be adequately designed (Attachment 3), (b) a program has been initiated for assessing the similarity of electrical equipment to facilitate seismic qualification (functionability of the equipment and structural integrity of internal components) of all safety related electrical equipment, namely the SEP Owners Group program, and (c) a program for seismic qualification of electrical cable trays based upon testing by the SEP Owners has been implemented. These latter two programs are intended to confirm the adequacy of existing designs and equipment. l

~ . Recently, MRC has initiated a generic program to develop criteria for the seismic qualifications of equipment in operating plants; Unresolved Safety Issue (USI) A-46. This program is scheduled for the completion in Marci. Under this program, an explicit set of guidelines (or criteria) 1983. that could be used to judge the adequacy of the seismic qualifications (both functional capability and structural integrity) of safety related mechanical and electrical equipment at all operating plants will be developed. Considering tht: All safety related electrical equipment has been properly anchored 1. by assuming that all block walls to which the safety related equip-ment items are supported are capable of withstanding the postulated seismic loads and the open items about the structural integrity of transformers and control room panels will be closed in the near future), Past experience and testing results (from both nuclear and nonnuclear 2. facilities) indicate in general that electrical equipment will con-tinue to operate under dynamic loading conditions with only limited transient behavior, if the equipment is adequately anchored; and The SEP Owners Group programs from which a set of general analytical 3. methodologies is being developed for the seismic qualifications of cable trays and for assessing similarity of other safety related electrical equipment to facilitate qualification for operability, it is our judgement that for the interim period until a technical resolution of USI A-46 is reached regarding methods for assessing seismic qualifica-tion of equipment in operating plants, the safety related electrical equipment at Millstone 1 Plant will function during and after an earthquake up to and including the postulated SSE. If additional requirements are imposed, as a result of USI A-46, regarding functional capability of safety related electrical equipment, the Millstone 1 facility will be required to address these new requirements along with other operating reactors.

l . Furthermore, since the ground response spectrum (0>.2g R. G.1.60 spectrum) used for Millstone 1 seismic reevaluation envelopes the Millstone 1 site specific ground response spectrum, additional safety margins in the structures, systems and components do exist for resisting SSE seismic loadings. Thus, the staff concludes that Millstone 1 plant has an adequate t seismic capacity to resist a postulated SSE, and therefore, there is l ressonable assurance that the operation of the facility will not be inimical to health and safety of the public.

REFERENCES 1. NRC NUREG/CR-2024 Report, " Seismic Review of the Millstone 1 Nuclear Power Plant," July 1981. 2. NRC letter, " Site Specific Ground Resppnse Spectra for SEP Plants Located in the Eastern United States," June 17, 1981. 3. NRC NUREG/CR-1582 Report, " Seismic Hazard Analysis," Vols. 2-4, October 1981. 4. EGG-EA-5391 Report, " Summary of the Millstone Unit 1 Piping Calculations Performed for the Systematic Evaluation Program " May 1981. 5. Letter from J. Shea, NRR, to W.G. Counsil, NNECo, dated June 15, 1982. 6. J. A. Blume and Associates, " Millstone Point Nuclear Station, Earthquake Analysis : Gas Turbine Building," March 1968. 7. EBASCO Isometric Dwgs. 11-1 sheets 1 & 2, VIII-l sheets 1 & 2, and IX-1. sheets 3 & 4. 8. Letter from W.G. Counsil, NEUCo, to D.M. Crutchfield, NRC, dated July 13,1981. l 9. Letter from W.G. Counsil, NEUCo, to D.M. Crutchfield, NRC, dated August 14, 1981. 10. Letter from W.G. Counsil, NEUCo, to D.M. Crutchfield, NRC, dated September 10, 1981. 11. Meeting Summary of December 9,1981 meeting, dated January 29, 1982. 12. Letter from W.G. Counsil, NEUCo, to D.M. Crutchfield, NRC, dated April 28,1982. 13. Letter from D.G. Eisenhut, NRC, to W.G. Counsil, NEUCo, dated January 1,1980. 14. Letter from D.M. Crutchfield, NRC, to W.G. Counsil, NEUCo, dated July 28, 1980. l l

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h/c men f 5.6 GAS TURBINE BUILDING The gas turbine building was originally analyzed as a single-degree-of-freedom system using a fixed-base cantilever beam model (Refer-ence 12). No effects of the flexibility of the overburden and pile system on which the structure is founded were considered. The analysis 0 was conducted using the response spectrum from the 1952 Taft north 69 west earthquake record normalized to 0.079 with 5% damping. The analysis was performed for the ' earthquake in two directions but only the critical (E-W) direction results were reported (Reference 12). The period was calculated to be 0.089 seconds and the response at the roof was computed to be 0.121. It was recommended in Reference 12 that the structure be 9 checked for safe shutdown for twice the loads developed for the 0.07g case in conjunction with a 0.10g vertical component. It is not clear that the structure was checked for the 0.179 SSE or whether 0.14g was used. 5.6.1 Pile Foundation Characteristics In order to account for the pile foundation amplification char-acteristics, a confirmatory analysis was conducted for the pile structure system. The number, lengths, and section properties of the piles are available, but the orientation of the HP-section piles relative to the building axes was not provided. The foundation material has been described as quarry trailings over bedrock. However, no information was available on the dynamic characteristics of the overburden. Consequently, a fairly wide range of soil properties was assumed. Information for l crushed rock backfill to typical gravelly soils from a number of sources was utilized. Upper and lower bound soil properties were estimated based on values for fill materials considered representative, and intemediate " worst case" values resulting in maximum structure responses were included in the anal'ysis. The range of the soil shear modulus used in the analysis was assumed at approximately 560 ksf to approximately 6350 ksf. Maximum dynamic response of the structure / foundation system was found to occur for a shear modulus of approximately 4000 ksf for the transverse direction and I 2800 ksf for the longitudinal direction. Obviously, the maximum response in the two directions cannot occur simultaneously for any one given soil condition. i 5-1

Piles are HP 10x42 pile sections driven to refusal as defined as five blows per 1/4 inch. Fifty-three piles were originally specified for the 81'-6" by 38'-2" foundation, with somewhat closer spacing under the gas turbine pedestal. However, a total of 65 piles were actually installed due to no refusal for some of the piles. The average pile length was approximately 16 feet. However, the driven length varied from approximately 6'-7" to 38'. No overall trends in pile length with plan location throughout the foundation were apparent, so the pile foundation system stiffness and damping parameters were computed using the average pile length. For the upper bound shear modulus, the strong axis of the pile was assumed oriented in the direction of maximum earthquake input, while for the lower bound, the weak axis was oriented in the direction of maximum earthquake input. The stiffness and damping properties of the individual piles were computed using methods outlined by Novak (Ref 13 and 14). End bearing piles with pin-ended caps were assumed. Novak's stiffness and damping coefficients are derived for circular piles. Becaus'e the equivalent pile radius is determined by the enclosed area, the pile modulus of elasticity was adjusted depending on the item of interest such as axial, flexural, or shear stiffness. The dimensionless frequencies computed for the individual piles for all soil conditions exceeds the dimensionless frequency recommended by Novak for considering frequency dependent effects. Therefore, frequency independent stiffness and damping parameters were used throughout the confirmatory analysis. Pile group effects were considered using methods developed by Poulos (Ref 15 and 16). A summary of the combined pile foundation stiffnesses computed for the various soil conditions is shown in Table 5-1. 5.6.2 Structure Model A three-dimensional lumped mass model was developed for the confimation analysis. Masses for the model included the weight of the roof and floor slabs including the equipment pedestals and the tributary masses of the walls. In addition, a 75 psf equipment weight for the entire plan area was assumed. This total weight was lumped at the equipment pedestals on the ratio of their plan areas. The original JAB analysis (Ref 12) included a 60 psf live load on the roof which was not included in the current analysis. Due to the presence of a number of relatively 5-2 l

large cutouts in the walls, the lateral stiffnesses were computed as fixed-free piers, including both shear and bending flexibilities. These stiffnesses were converted to equivalent shear areas. The modulus of concrete was computed using the 3000 psi design strength with no provision for aging. As a result of the eccentricities introduced by the pile arrangement, non-uniform mass distribution, and cutouts in some walls, the mass center and center of rigidity of the system are not coincident and some torsional response results. 5.6.3 Structure Response The response of the combined structure-pile foundation model described above was determined by response spectrum analysis methods using R.G.1.60 spectra anchored to a 0.29 peak ground acceleration. Composit.e modal damping was based on 5% of critical for the concrete structure and 10% of critical for foundation. Geometrical damping for the piles was calculated using the same methods as the stiffness (Ref 13 and 14), and an additional 3 to 4% (depending on the mode) was added to account for the soil damping. As a check on the structure frequency characteristics, a fixed base frequency in the transverse direction of 12.5 Hz was calculated compared to the 11.2 Hz from the original JAB analysis. The combination of amplification through the overburden and use of the 0.2g R.G.1.60 response spectra results in substantially higher seismic loads than were developed in the original design analysis. Figure 5-6 shows a plot of the shear and moment diagrams for the worst case trans-verse direction. A load r.atio of approximately 0.57 is indicated assuming the structure as originally checked for the 0.17g SSE. No structure loads were available for the longitudinal direction from the original analysis. The north wall for transverse response is the most highly stressed wall of the structure. This wall has a factor of safety of nearly 4 for shear and 1.09 for moment using ACI 318-77 as a basis. The worst case stress condition for the piles occurs for the minimum soil shear modulus. For this condition, pile stresses are expected to exceed yield for the 0.2g earthquake. The ductility ratio for this case is approximately 1.3. This is considered acceptable in accordance with recommendations set forth in 5-3

Ref 1. The gas turbine building is therefore considered to be capable of withstanding the 0.29 SEE with only possibly minor yielding occurring in the piles for the worst assumed overburden soil characteristics. 5.6.4 In-Structure Response Spectra Seismic input motion for piping and equipment is typically defined by means of in-structure (or floor) response spectra for items with relatively small mass. For the gas turbine building, all equipment items of importance are located on the floor slab (or on rigid pedestals approximately 2' high which are integrally connected to the floor slab). The mass of the equipment was included as rigid masses connected to the structure in the building model as described in section 5.6.2. As part of the SEP evaluation, in-structure response spectra were generated using direct ground spectra to floor spectra methods (Ref 17) based on random vibration theory. In-structure spectra were generated for the gas turbine building floor slab for both vertical and the two principal horizontal directions. Spectra were generated for 2, 3, and 7% of critical equipment da ping in accordance with the recommenda-tions in Reference 1. Spectra were generated for each of the individual soil cases, and smoothed and broadened (15%) envelopes encompassing the complete soil range were developed. Plots of the new in-structure response spectra for the gas turbine building are shown in Figures 5-7 through 5-9. No information concerning spectra used in the original qualification of gas turbine building equipment was available. However, it is assumed that the ground response spectra used for design of the structure were also used for equipment qualification. Consequently, due to the use of R.G.1.60 spec-tra anchored to a 0.2g peak acceleration level together with the amplifi-cation of the pile foundation system, the in-structure response spectra developed for the SEP evaluation are expected to exceed the original design spectra throughout the frequency range. i 5-4

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~ .. J ~ ,~'- O- [ REFERENCFS (Continued) Y,- ~ j' tr V

10. " Millstone Point Nuclear Statiop, Earthquake Analysis:

Pressure Suppress hn Chamber", J. A. Blume and Associates. 11. " Plant Unique' Analysis Report for Torus Support System and Attached Piping for Millstone Point Nuclear Power Station", Teledyne Materials Research, July 26, 1976.

12. " Millstone Point Nuclear Station, Earthquake Analysis: Gas Turbine Building", J. A. Blume and Associates, March,1968.

lb.

Novak, M., " Dynamic Stiffness and Damping of Piles" University of Western Toronto, January 1974.

14. Novak, M. and F. Aboul-Ella, " Stiffness and Damping of Piles in Layered Media" Earthquake Engineering & Soil Dynamics, ASCE Geo-technical Division, Vol. II, Pasadena, CA, June 1978.

15. Poulos. H.

G., " Group Factors zfor Pile-Deflection Estimation", J. Geotechnical Division ASCE, December 1979. 16. Poulos, H. G., and E. H. Davis, Pile Foundatior Analysis and Design, John Wiley and Sons,1980.

17. Singh, M.

P., " Generation of Seismic Floor Spectra", J. Engineering Mechanics Division, ASCE, October,1975. R-1,

4 .g A& Led 2 Conclusions regirdinq equisnent selected for review for scismic desi n 11equacy of *1111 stone finit 1. (Jona 193?) 1 Iten Gescription Conclusion and Reco7,endation 1. Energency service water A Licensee analysis of the oumo was pump reviewed. Resulting stresses in the anchor bolts were found to be well belos vield and puno casing, impeller and shaf t and column oice ~ stresses were low. The ES'.4? is considered to be structurally adequate to perform its intended functions during an SSE. 2. Emergency condenser 0.K. 3. Shutdown heat exchanger

0. K.

i 4. Energency coolina water NMEC0 stated that 'tillestone Unit heat exchanger 1 does not have an EC'.44X. Details of a revised support system for the Containment Spray Cooling Heat Exchancer were instead submitted for review. Although design cal-culation results were not sub-mitted, the revised CSCHX suoport systen apoears to be adequate to resist SSE loading. 5. Recircul ati-oumo 0.K. for structural intecritv; supoor t however, no evaluation has been performed to determine seismic loads on the Dumo snub 5er suoports or to evaluate its functionti integrity since sufficient data is unavailable. 6. Emerqency diesel oil 0.K. j storage d3y tan'< P 1

s Item Description Conclusion and Recommendation 7. Motor oparated valves The Licensee stated in Reference I that there are a small number of - 4" and smaller motor ooerated valves. They h. ave reviewed piping stress levels at valve locations for one 2" and two 4" valves and found pipe stresses, without consideration of valve coerator eccentricity, to be less than 10% of the code allowable. For a fourth valve, the ooerator eccen-tricty has been included in a oice stress analysis. Based upon the examples subnitted, valve operator eccentricity does not appear to be a oipe stress oroblem. The licen-see has stated that operability of the valves will be addressed within a separate program. 8. CR0 hydraulic control Results of an analysis of CR0 system in-luding insert and withdraw lines 5etween tubing and supports the accumulators and the shield wall indicate that the lines will remain functional durinq an SSE, Reference 1. ~ 9. Reactor vessel, supports, 0.K. and intern 31s 10. Ba'.tery racks Positive restraint of the batteries has been demonstrated via a Licensee analysis of the battery / rack system, References 1 and 2. 11. Motor control centers Blockwall structural inteority has been demonstrated via the Licensee response to-IE Bulletin 80-11. local attachments to block walls are detailed in Reference 1. Anchorage of electrical cabinets is considered ade19 ate. Function-ability is to be addressed in an SEP owners group program. 2

l I Iten Description Conclusion and Reconnendation 12. Transformers A sunmary of the results of an investigation of anchorage and supoort of safety rel3ted equionent was submitted by the Licensee, Reference 3. Typical details of support for a 490 volt transformer were provided via Reference 1. Transf ormer supoorts appear adequate to resist an SSE. 13. Switchgear panels Blockwall structural integrity has been demonstrated vi s the Licensee response to IE Bulletin 80-11. Local attachments to block walls are detailed in Reference 1. Anchorage of electrical cabinets is considered adequate. Function-ability is to be addressed in an SEP owners qroup oroqraq. 14. Control room electric 11 Control room electrical panels are panels welded to embedded ol ates. Licen-see calculations have demonstrated that no modifications are re-quired. Functionality is to be addressed in an SEP Owners Group program. 15. Diesel generator remote Blockwill structural integrity has control boards been demonstrated vi a the Licensee resoonse to IE Bulletin 80-11. Local attachments to block walls are detailed in Reference 1. Anchorage of electrical cabinets is considered adequate. Function-ability is to be addressed in an SEP owners group orogram. 3

Iten De,:ription Conclusion and Reconnendition 16. Battery room distri5stion Blockwall structural inteqrity has pinels been demonstrated via the Licensee response to IE 9:11etin 80-11. Local attachments to block walls are detailed in Reference 1. Anchoraqe of electrical cabinets .is considered idequate. Function-ability is to be addressed in an SEP o.iners qroup orogran. 17. Elec tricil cable receniys The Licensee is carticiostinq in an SEP Owners Gro so effort for qualif ying cable trays and cor.duit raceway systems, Reference 1. 4

REFE RE'4CE S 1. Northeast Nuclear Enerqy Comp 3ny ' letter, '.4. G. Council to D. 'f. Crutchfield, USNRC, April 23,193?. i 2. Northeast fluclear Energy Company letter,

.4. G. Council to D. 'i.

Crutchfield, USNRC, September 10, 1931. 3. Northeast fluclear Energy Co,7any letter, 'l. G. Council to D. 'i. Crutchfield, USNRC, April 13,193?. \\ t

. Attachment d May 20, 1982 g ggj Saff-194-82

SUMMARY

OF MILLSTONE UNIT 1 FEEDWATER AND SHUTDOWN COOLING DIPING REANALYSIS D. K. Morton Applied Mechanics Branch EG&G Idaho, Inc. May 12, 1982 s

~ Audit calculations to verify the original analyses and new calculations using current ASME Code and Regulatory Guide Standards were previously performed on the Millstone Unit 1 feedwater and shutdown cooling piping (Reference 1). Information subsequently received provided as-built information of the piping systems per the NRC's I&E Bulletin 79-14. This updated information typically reflected pipe and valve dimensional changes and pipe support variations. After incorporating this updated information into the NUPIPE-II finite element structural models, a " current criteria" analysis was then repeated using the methods and procedures described in Reference 1. The results of the analysis using the updated configurations are described below. This supplement should be used in conjunction with the Reference 1 report. FEEDWATER ANALYSIS A complete plot of the feedwater piping model is shown in Figure 1. Figures 2 and 3 displaf isolated portions of various sections of the complete piping model. It should be remembered that the dimensions shown on these plots are approximate and are intended only to give an indication of scale. Two feedwater models were analyzed. One model reflected the entire system as shown in Figure 1. The second model reflected the removal of the reactor water clean-up system (8"-CUW-28). The removal of this piping is currently scheduled for late 1982. f The analytical assumptions made for the feedwater piping listed in l Reference 1 were again utilized for this analysis. One exception is that a design pressure of 1250 psi was used rather than the 2300 psi Es reported in Reference 1. The same response spectra listed in Reference 1 was utilized for this analysis. l l l 1

l MILLSTONE FEEDWATER. IN CONTAINNENT. AS i NUPIPE NATNEMATICAL NODEL GW I.4. ..teseNo.. Reactor C - "*'' L'e*'t'a Vessel %

e-

.e.i.. / O- ..v..a. N -.nsio.ver.at t 12"-FDW-12b " "."......, I ~ .u m m -.....<...... s,.. m

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eit,

.~ 14" thermal sleeve .. - m - **' \\ 18*-FDW-Sa Anchor Figure 1. Millstone Unit I feedwater piping -- complete model. PAGE 2

d MILLSTSNE FEEDWATER. IN CONTAIMMfMT. AS ^ NUPIPS NATHOMAftCAL MODEL IV l.4. 0 m N Penetration ' ~ " " " ' ' " " " X-15 o O - .a....t.T L. cats.m e. o-12.3 ft. u............, _g....... .u....... o ~ _ g. ,u n.u ..c... meme. TaL.. y". 8"-CUW-2B y 12.6 ft. X I W 1 l i Vll-31 ..TATI.. A...? T*A.I. Me .-2 ,La.. TILT 33 4 . ~., -

~

V11-32 e'e 12"-FDW-12b ,/ / Figure 2. Millstone Unit I reactor water clean-up system. PAGE 8

MILLSTONE FEEDWATER. IN CSNTAINMENT. AS i NUPIPE MATHEMATICAL MODEL (V 1.4.

- LEGEND==

.ao. t.c.,a.. O -.. sop.s.,t.c.,i.. R.V. g e-R.V. t. sca -====. 4Q. .468D SUPPe., j j e _ g -..... .e..,. Ull-22.4 ft. [ q - .6....t. l l 23.0 ft. l'- o Y 1I' '... y 1-FW-11a x a ier JP* l-FW-10a '" / .-,et...

rit, a

"'" ses - 1-FW-9A lg.. 30.3 ft. Figure 3. Millstone Unit I feedwater piping without 8"-CUW-28. PAGE 4

Table 1 lists the first ten natural frequencies for both models and those listed in Reference 1. A comparison between the frequencies previously determined and the updated configuration with 8"-CUW-28 still installed shows variations which were expected, considering the model changes made. Of course, the frequencies of the model without 8"-CUW-28 increased in magnitude. This was also expected since the 8"-CUW-28 was core flexible than the 12, 14, and 18 inch piping. A comparison of ASME Code, Class 2, Equation 10, primary stress results is displayed in Table 2. This table compares results at identical points between the various models. All models are within the allowable limits; however stresses on the 8"-CUW-28 are higher than previously predicted in the Reference 1 report. Typically, stresses on the 12 and 18 inch piping are lower for both updated configuration models. These changes in stress levels were not unexpected, especially in light of the piping support changes made (both deletions and additions), r Table 3 presents a comparison of support loads. This table lists seismic loads for all models. Since the support configuration has been altered (one deletion and four additions) from the Reference 1 analysis, a comparison of loads provides little useful information. It is assumed that all supports have been adequately designed. This assumption is partially supported by knowing that the design load for the support at Node 45 was 13.9 kip. This is greater than the weight plus SSE imposed load. No support analysis was performed due to lack of complete information. Table 4 indicates the anchor loads for all models. Again, since the support configuration has been altered, a comparison of anchor loads provides little useful information. However, due to the increased number of supports, most of the anchor loads are smaller than those reported in Reference 1. No anchor analysis was performed due to lack of information. 5

TABLE 1. COMPARISON OF FIRST TEN NATURAL FREQUENCIES FOR MILLSTONE UNIT 1 FEEDWATER PIPING Ref. 1 Current Feedwater Feedwater System Frequencies System Frequencies Frequencies Without _ Hz) (Hz) 8 "-CUW-28 (Hz) ( Mode 1 1.76 2.32 2.98 2 3.73 3.41 4.93 3 4.04 3.91 5.15 4 4.74 4.83 6.91 .l 5 5.38 5.38 8.08 6 5.84 5.59 8.64 7 5.94 6.99 10.3 8 6.68 7.37 13.2 9 7.72 r 8.23 15.7 10 9.31 9.18 17.3 l l 6

TABLE 2. COMPARISON OF ASME, CLASS 2, EQUATION 10, STRESS VALUES - " CURRENT" SSE STRESS (KSI) Node Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Allowable Coment 5 13.1 12.2 9.7 9.1 13.1 11.9 38.8 R. V. Anchor 285 13.9 13.0 4.7 4.6 4.9 4.8 38.8 Anchor Near Penetration X-9A 350 9.16 8.50 10.4 9.5 11.8 10.8 38.8 R. V. Anchor 465 30.6 27.3 34.2 32.1 35.0 Penetration X-15 Anchor 360 29.6 26.3 30.4 27.6 35.0 Elbow 355 26.3 23.3 28.6 26.0 5.4 5.4 35.0 End of Tee Leg 110 28.5 25.2 26.6 24.2 11.9 10.9 38.8 Tee Inte section u NOTES: For the previous configuration, the case numbers correspond to the following situations: a. Case 1 = 2% damping Case 2 = 3% damping b. For the updated configuration, the case numbers correspond to the following situations: Case 3 =.2% damping Case 5 = 2% damping with 8"-CUW-28 removed Case 4 = 3% damping Case 6 = 3% damping with 8"-CUW-28 removed c. Allowable stress is 2.4 S - h d. Equation 10 stresses include the effects of pressure, weight, and SSE. e. The reported stress at Node 465 for Case 3 already reflects the use of the alternate pressure term.

r TABLE 3. SUPPORT LOAD

SUMMARY

-- FEEDWATER MODEL -- SEISMIC LOADS ONLY LOADS (KIP OR IN.-KIP)

Node Direction Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 11.9 11.1 11.9 10.5 45 X 145 INC 14.2 12.5 13.7 11.7 11.0 9.4 11.3 10.0 19.0 16.5 174 INC 195 X 13.1 11.5 16.2 13.9 20.5 17.4 245 X 15.4 12.7 21.1 18.0 Y 5.9 5.1 7.3 6.3 258 X 8.2 7.0 11.8 10.2 Y 0.2 0.2 0.3 0.3 ZZ NOTES: a. Directions are global axis orientations. INC -- Inclined to global axes and ZZ -- rotational fixity about global Z axis. b. Case situations defined in Table 2.

TABLE 4. ANCHOR LOAD

SUMMARY

-- FEEDWATER MODEL -- SEISMIC LOADS ONLY ANCHOR LOAD (KIP OR IN.-KIP)

Node Direction Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 5 X 6.1 5.4 3.9 3.7 3.6 3.3 Y 7.1 6.2 7.6 7.0 6.8 5.9 Z 5.4 4.6 5.5 4.7 9.3 8.0 518. 451. 885. 766. XX 565. 489. YY 351. 324. 153. 140. 151. 134. ZZ 706. 650. 270. 249. 253. 220. 285 X 6.3 5.5 1.9 1.9 1.9 1.9 Y 9.2 7.9 1.3 1.3 1.4 1.4 Z 16.9 15.5 15.2 13.9 20.3 18.0 XX 1650. 1420. 94.3 83.8 127. 112. YY 1140. 973. 95.6 89.3 108. 98.4 ZZ 532. 450. 119. 107. 172.

150, 350 X

6.3 5.3 5.0 4.4 8.5 7.4 Y 18.2 15.6 9.7 8.6 15.0 13.0 Z 5.9 5.0 4.0 3.3 4.0 3.4 XX 404. 350. 409. 346. 504. 432. YY 106. 93. 91.2 78.3 111. 94.6 ZZ 350. 305. 310.9 257. 369. 313.4 465 X 7.5 6.4 3.2 2.8 Y 3.0 2.6 3.7 3.3 Z 8.6 7.4 3.1 2.8 XX 255. 225. YY 575. 514. 322. 285. ZZ NOTES: a. Directions listed are in the global coordinate system. "X" implies a force in the global X direction. "XX" implies a moment about the X axis. b. Case situations defined in Table 2.

In conclusion, the results discussed above can be used to conclude that the Millstone Unit 1 feedwater piping stresses will be within ASME Code allowable limits during an SSE event. SHUTDOWN COOLING ANALYSIS A complete plot of the shutdown cooling piping model is shown in Figure 4. Figures 5 and 6 display isolated portions of various sections of the complete piping model. It should be remembered that the dimensions shown on these plots are approximate and are intended only to give an indication of scale. Two shutdown cooling models were analyzed. One model was utilized and significant overstresses were found. Therefore, a second analysis was made incorporating an additional support which is planned to be installed according to the licensee. Again, the analytical assumptions used in Reference 1 were incorp-orated into this analysis. Also, the same response spectra listed in Reference I was utilized for this analysis. Table 5 lists the first ten natural frequencies for both models and those listed in Reference 1. A comparison between the frequencies previously determined and the updated configuration without the additional support shows significant variations. These decreases in frequency magnitudes were expected because of the support variations. Of course, the frequencies, with the additional support modeled in, increased over the frequencies associated with the piping without the support. A comparison of ASME Code, Class 2, Equation 10, primary stress l results is displayed in Table 6. This table compares results at identical points between the various models. The results show that the stresses increased significantly in the updated model without the additional l support. This is due to the change in supports and applied stress intensif-f ication factors. An additional support was incorporated into the model and a reanalysis was performed because ASME Code allowables were exceeded. The additional support was sufficient to permit the predicted stresses to fall below Code allowables. 10

WILLSTONE SHUTDSWN CSSLING IN CSMTAINMEN NUPIPE NATH84ATICAL NSDEL EV t.4. 't .. L E G E se D.. i - m..e t.ca,s.. t O - .a..p.s.1

t. cats.e wa-

.P.s.e .a.s.. o- . __4 - g - ..c .u... _ g. .u..a ..c... ..t.. O Penetration X-14 y ..;~ i X f ~ ..."< 28"-recire. 100p V14-1 suction .r .c E-I PLa.E flLT MA m Penetration / "* X-12 Figure 4. Millstone Unit I reactor shutdown cooling piping -- complete model. PAGE 2

o MILLSTONE SHUTOGWN CSSLING IN CSNTAINNEN NUPIPB NATHEN.,8 CAL M6 DEL 4F t.4. 16.4 ft. ..te euo.. .c..... o....... .,t.c..... .P.l G 98. 45. >O - ..v..t. +-.4 ..nse s.re.., V14-1 28" recirculation _p.-..... .t.... 8"-CUW-1 loop suction 4 -g - n u i.u ..c... ..t.. ~ Y o H 14"-SHC-1 ~ ~ _m 17.3 ft. ....,......,v.ui. .-t PL... ,tt, ma Penetration X-12 Figure 5. Millstone Unit I reactor shutdown cooling inside drywell. PAGE 8

MILLSTONE SHUTDONN CSSLING IN CONTAINNEN NUPIPE NATHLMATICAL NSDEL tv l.4. li I l

. LEG 9NDe*

e s. i - moos tocaties O - masePoset tocaties M-aresus mausse O-suwooes M SIGID $5PP9ef -Q - Ancuee e, 3 - EL&OTS JStut -( - PLSE B BLE AuCMSW ~ m. 22.5 ft. ... n D Penetration X-14 X w V11-2 E-3 PLANS TILT Me 8.7 ft \\ 8"-CUW-1 i. Vll-1 ) /.14"-SHC-1 Figure 6. Millstone Unit I reactor clean-up water system. PAGE E d

\\ TABLE 5. COMPARISON OF FIRST TEN NATURAL FREQUENCIES FOR MILLSTONE UNIT 1 SHUTDOWN COOLING PIPING Ref. 1 Without With Mode Frequencies Additional Support Additional Support 1 5.03 2.53 3.97 2 7.13 3.95 4.61 3 7.81 4.76 5.53 4 8.69 5.04 5.81 5 10.4 5.59 7.17 6 12.4 5.97 7.48 7 13.3 7.26 8.86 8 13.5 8.89 10.81 9 14.1 11.60 13.34 10 15.3 12.56 14.95 t NOTE: Frequency units are Hz. t f f I l ? r I l 14

TABi.E 6. COMPARISON OF ASME, CLASS 2, EQUATION 10, STRESS VALUES - " CURRENT" SSE STRESS (KSI) Node Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Allowable Comment 5 21.2 18.9 45.1 39.5 32.8 28.6 41.4 28 in. Pump Suction Line Anchor 110 7.92 7.76 19.8 17.7 23.8 21.0 38.9 Penetration X-12 Anchor 285 10.3 9.87 45.4 39.7 26.7, 23.4 41.4 Penetration X-14 Anchor 15 49.2 42.5 99.0 85.6 38.2 34.0 41.4 14x8 Weldolet Junction 115 57.4 49.4 68.3 59.3 28.1 25.3 41.4 Branch End of Weldolet 120 36.6 32.0 32.7 29.1 18.6 17.1 41.4 Straight Pipe 140 26.6 23.7 35.7 31.5 13.7 12.7 41.4 8 In. Elbow 150 26.3 23.4 34.9 31.0 15.2 14.1 41.4 8 In. Elbow 280 8.1 7.7 36.2 31.8 22.0 19.4 41.4 Straight Pipe NOTES: For the previous configuration, the case numbers correspond to the following situations: a. Case 1 = 2% damping Case 2 = 3% damping b. For the updated configuration, the case numbers correspond to the following situations: Case 3 = 2% damping Case 5 = 2% damping with additional support Case 4 - 3% damping Case 6 = 3% damping with additional support c. Allowable stress is 2.4 S - h d. Equation 10 stresses include the effects of pressure, weight, and SSE.

Table 7 presents a comparison of support loads. This table lists seismic loads for all models. Since the support configuration has been altered (two deletions and one addition) from the Reference 1 analysis, a comparison of loads provides little useful information. It is assumed that all supports have been adequately designed. No support analysis was performed due to lack of detailed information. Table 8 predicts the anchor loads for all models. Again, since the support configuration has been altered, a comparison of anchor loads provides little useful information. However, due to a decrease in the number of supports, most of the anchor loads are larger than those pre-sented in Reference 1. No anchor analysis was performed due to lack of information. In conclusion, the results discussed above can be used to conclude that the Millstone Unit 1 shutdown cooling piping stresses will be within ASME Code allowable limits during an SSE event provided the additional support (CVW-R004) is installed. REFERENCE 1. M. E. Nitzel, " Summary of the Millstone Unit 1 Piping Calculations Performed for the Systematic Evaluation Program," EGG-EA-5391, May 1981. 16 t

s TABLE 7. SUPPORT LOAD SUP94ARY -- SHUTDOWN COOLING MODEL -- SEISMIC LOADS ONLY LOADS (KIP) Node Direction Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 80 Y 0.93 0.83 LAT 6.45 5.47 255 Y 2.99 2.58 LAT 5.38 4.61 3.99 3.35 165 X 4.46 3.69 Z -w NOTES: a. Directions are global axis orientations. INC -- Inclined to global axes, b. Case situations defined in Table 6.

rn.

a "*

TABLE 8. ANCHOR LOAD

SUMMARY

-- SHUTDOWN COOLING MODEL -- SEISMIC LOADS ONLY LOAD (KIP OR IN.-KIP)

Node Direction Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 5 X 8.55 7.13 19.0 16.2 11.71 9.80 Y 5.72 4.77 10.1 8.38 7.86 6.50 Z 13.4 11.33 21.6 18.3 17.28 14.4 XX 1200. 1000.

1752, 1485.

1256. 1046. YY 386. 322. 867. 730.

712, 590.

ZZ 680. 566. 1367. 1165. 710. 600. 110 X 2.79 2.32 7.13 5.93 8.30 6.88 Y 0.89 0.77 4.19 3.54 5.11 4.29 Z 4.87 4.09 4.60 3.90 6.15 5.15 XX 41.7 35.7 735. 617. 971. 813. YY 73.4 62.6 1114. 931. 1459. 1213. ZZ 58.8 48.7 179. 151. 213. 179. 285 X 4.33 3.80 3.13 2.66 2.50 2.08 Y 1.40 1.30 7.80 6.72 2.39 2.09 Z 3.94 3.46 3.65 3.01 3.30 2.72 XX 74.8 66.5 516. 445. 138. 123. YY 42.3 37.7 551. 461. 472. 390. ZZ 74.5 68.2 684. 591. 206. 181. NOTES: a. Directions listed are in the global coordinate system. "X" implies a force in the global X direction. "XX" implies a moment about the X axis. b. Case situations defined in Table 6.}}

ML20054L744

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