ML20031C826
| ML20031C826 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 09/23/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8110090008 | |
| Download: ML20031C826 (26) | |
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-tRicket i40. 50-329/330 LB #4 r/f SEP 2 31981 DEisenhut EAdensam MDuncan SHanauer
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'Y bec: Local PDR RVollmer Docket Nos.: 50/329/330 NRC PDR h,.
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RHartfield, MPA Nr. J. W. Cox OCT1 1981* d TIC OELD Vice President ACRS (16)
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Consumers Power Company S
ACappucci 1945 West Parnall Road fi,,
RHernan Jackson, Nichigan 49201
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FCherny
Dear itr. Cook:
HBramer Sub. ject:
Transmittal of Preliminary SER Dratt Sections 3.6.2, 3.7.3 and 3.9, Hidland Plant. 'Jnits 1 and 2 Enclosed for your review and coinent are the following preliwiriary dratt sections of tne hAC staff's Safety Evaluation Report for af dland Plant, Units 1 and 2:
3.6.2 Determination of Break Locations and Dynamic Ettects Associateu with tne Postulated F.upture of Piping 3.7.3 Seismic Subsystem Analysis 3.9 Hechanical hystens and Components Tnese sections reflect only tnose portions of Hialand FShx Chapter 3 wnich are the review responsibility of the Staff's nechanicai cngineerin3 Branch.
Your attention is directed in particular to numerous open items contained within these draf t sections. A principal objective of this transmittal is to provf ue for timely Joentification and resolution of any adaittonal analysis, missin, information, clarifications or other work necessary to resolve outstanding issues. Also, resolution of these open items, in several cases, need to be integrated with resolution of several unresolved 50.55(e) reports not yet acknowledged in these draf t Skk sections. Please contact tne Staff's Project Manager regarding the need for any meetings and telephone conferences to this end.
four comments, including scheoules for coupletion of any further analyses or other work associated witn resolution of open iteas, are requestea witnin two weeks of receipt of this letter.
Sincerely, i
q 110090006 10923; DRADocK0g00032f Drfginalsigned by:
L Pm Rober'. L. Tedescu, Assistant utrect'or
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.e DISTRIBUTION Docket File R. Hernan LB#4 r/f DBoyd, IE-RIII DEisenhut FCherny EAdensam HBrammer MDuncan bec: L/PDR Docket Nos.: 50/329/330 SHanauer NSIC/ TIC /TEPA RTedescu ACRS (16)
RVollmer Mr. J. W. Cook TMurley Vice President RMattson Consumers Power Company RHartfield, MPA 1945 West Parnall Road OELD Jackson, Micnigan 49201 OIE (3)
ACappucci
Dear Mr. Cook:
Subject:
Transmittal of Preliminary 5ER Draf t Sections 3.6.2, 3.7.3 and 3.9, hidland Plant, Units 1 and 2 Enclosed for your review and comment are the foilcwing preliminary draft sections of the NRC staff's Safety Evaluation Report f or Fiidland Plant, Units 1 and 2:
3.6.2 Determination of Break Locations anc Dynamic Effects Associated with the Postulated Rupture of Piping 3.7.3 Seismic Subsystem Analysis 3.9 Hechanical Systems and Components These sections reflect only those portions of Midland FSAR Chapter 3 which are the review responsibility of the Staff's Mechanical Engineering Branch.
Your attention is directed in particular to numerous open items contained within these draft sections. A principle objective of this transmittal is to provide for timely identification and resolution of any additional analysis, missing information, clarifications or other work necerssary to resolve outstanding issues. Also, resolution of tnese open items, in several cases, need to be integrated with resolution of several unresolved 50.55(e) reports not yet acknowledged in these draft SEk sections. Please contact the Staff's Project Manager regarding the need for any meetings and telephone conferences to this end.
Your comments, including schedules for completion of any further analyses or other work associated with resolution of open items, are requested within two weeks of receipt of this letter.
Sincerely,
~
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Robert L. Tedesco, Ass:stant Director for Licensing Division of Licensing
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l MIDLdWD Mr. J. W. Cook Vice President Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201, cc: Michael I. Miller, Esq. Mr. Don van Farrowe, Chief Ronald G. Zamarin, Esq. Division of Radiological Health Alan S. Farnell, Esq. Department of Public Health Isham, Lincoln & Beale P.O. Box 33035 Suite 4200 Lansing, Michigan 48909 1 First National Plaza Chicago, Illinois 60603 William J. Scanlon, Esq. 20!,4 Pauline Boulevard James E. Brunner, Esq. Anri Arbor, Michigan 48103 Consumers Power Conpany 212 West Michigan Avenue U.S. Nuclear Regulatory Commission Jackson, Michigan 49201 Resident Inspectors Office RoJte 7 Myron M. Cherry, Esq. Midland, Michigan 48640 1 IBM Plaza Chicago, Illinois 60611 Mt.. Farbara Stamiris 5795 N. River ~ Ms. Mary Sinclair Freeland, Michigan 48623 5711 Summerset Drive Midland, Michigan 48640 Mr. Paul A. Perry, Secretary Consumers Power Company Stewart H. Freeman 212 W. Michigan Avenue Assistant Attorney General Jackson, Michigan 49201 State of Michigan Environmental Protection Division Mr. Walt Apley 720 Law Building c/o Mr. Max Clausen Lansing, Michigan 48913 Battelle Pscific North West Labs (PNWL) Battelle Blvd. Mr. Wendell Marshall SIGMA IV Building Route 10 Richland, Washington 99352 Midland, Michigan 48640 t Mr. Steve Ganier 2120 Carter Avenue St. Paul, Minnesota 55108 1 0 6 i I
s. /' UNITED STATES g g NUCLEAR REGULATORY COMMISSION wasmworow. o. c. 2oses \\, / / l SEP 4 1981 / \\ / MEMORANDUM FOP: E. Adensam, Chief / ~ \\ Licensing Branch #4, OL /' FROM: R. Bosnak, Chief [ Mechanical Engineering Branch, DE
SUBJECT:
DRAFT INPUT TO MIDLAND, UNITS 1 & 9 SER The MEB and its c tractor, Energy Technology Eng neering Center, have completed the review of the 'dland FSAF through Amendment /34. We have chosen not to develop a round of adqstions but to proceeo directly to a draft SER input. We request that you forwa d this draft to the appl,1 cant. The applicant should then prepare an agenda for a eting in which we can discuss and resolve the open issues in our review. W anticipate this meeting being held over a 3-5 day period at a mutually agree ble. site about 50/ days after the applicant has received this drsft. After this meeting and any necessary fo; low-up, we will i update the SER input into a rm sufficient / y clean for publication. You should emphasize to the appli ntthatwe/txpectthisextendedmeetingtoresolve almost all of these open issues Therefo e, it should bring the NSSS, AE, and utility people necessary to both discuss technical details and make binding comitments. We strongly reccme the, eting be held at the Bechtel offices in Ann Arbor, Michigan. The draft SER contains those sections oi 3.6, 3.7, and 3.9 applicable to MEB's scope of responsibility. As stated in the April 3, 1981 let er fr J. Knight to R. Tedesco, the MES has tentatively reserved the week f Nove, er 15, 1981 for the Midland SER meeting. 9.-/ %4 R. J. B nak, Chie Mechanic Engineering Branch Division o Engineering
Enclosure:
As stated cc w/ encl: R. Vollmer, DE . Branmar, OE R. Tedesco, OL F. Cherny, DE N/ A. Cappucci-R. Hernan, DL L. Auge, ETEC (3) J. Knight, DE cc w/o enc 1: W. Pike, MPA
- 0. Eisenhuti, OL
Contact:
A.,Cappucci, DE:MEB, x29476 i
MIDLAND 1 AND 2 DRAFT SER 3.6 Protection Against Dynamic Effects Associated with the Postulated Ruoture of Piping The review performed under this section pertains to the applicant's program for protecting safety-related components and structures against the effects of postulated pipe breaks both inside and outside containment. The effect that breaks or cracks in high and moderate energy fluid systems would have on adjacent safety-related components or structures are required to be analyzed with respect to jet impingement, pipe whip, and environmental effats. Sev-eral means are riormally used to assure the protection of these safety-related items. They include physical separation, enclosure within suitably designed structures, the use of pipe whip restraints, and the use of equipment shields. 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Ruoture of Piping Our review under Standard Review Plan Section 3.6.2 was concerned with the locations chosen by the applicant for postulating piping failures. We also reviewed for the size and orientation of these postulated failures including the methodology used by the applicant in calculating the resultant pipe whip and jet impingement loads which might affect nearby safety-related components. In the Midland plant, Babcock & Wilcox and Bechtel have not followed ident1.al criteria in postulating pipe breaks inside and outside containment, but, as discussed below, both programs are in general acceptable. In one instance the applicant has chosen postulated pipe break locations not entirely in accordance with Standard Review Plan Sections 3.6.1 and 3.6.2. Specifically, the areas of concern are high energy piping syster, running along the roof of the auxiliary building which are not required to be seismic Category I. The applicant has optionally chosen to design this piping to withstand the fe 3-1
shutdown earthquake. However, this piping has not received the requisite que '+y assurance documentation to be designated leismic Category I. Never-th' .,, the applicant has chosen postulated pipe break locations for this piping 'using rules applicable to seismic Category I piping (terminal ends and intermediate points of highest stress). For piping which is not seismic Cate-gory I, the Standard Review Plan would essentially require consideration of postulated piping failures everywhere along the pipina length. However, our review has determined that the design of the subject piping is very conser - vative with the peak calculated stress only 56% of the allowable value. We find that these piping systems will behave essentially as a seismic Category I system and that the pipe break criteria used by the applicant are appropriate. Therefore, this deviation from the Standard Review Plan is acceptable. The applicant tas proposed a program of augmented inservice inspe.ction of high energy piping systems in the containment penetration region in which no pipe breaks are postulated. We would note that the Standard Review Plan Sections 3.6.1 and 3.6.2 do not require the postulation of pipe breaks in the containm'nt penettg+on region if certain criteria are met, including low e stress and an augmented inservice inspection program. Standard Review Plan Sections 3.6.1 and 6.6 call for such an augmented inservice inspection pro-gram to con;ist of 100% volumetric ultrasonic examination of each pipe weld. The applicant's program specifies a combination of surface and volumetric examination for piping with greater than 1/2 inch wall thickness and a surface exmnination alone for piping with less than 1/2 inch wall thickness. We would note that the applicant's procedures are consistent with the Section-XI of the - ASME Code. We find that the applicant's augmeqted icservice inspection pro-gram, although deviating from the Standard Review Plan, provides an equivalent level of safety and is acceptable. In order to account for strain rate effects, the applicant has assumed a 20% increase in the minimum yield strength of the materials used in the con-struction of the pipe whip restraints supplied by Bechtel. Standard Review Plan Section 3.6.2 allows only a 10% increase. The applicant has referenced the approved Bechtel Topical Report BC-TOP-9A, " Design of Structures for Missile Impact," which provides justification for a 20% increase when designing 3-2
4, structural steel for loads due to impinging missiles. However, Topical Report BC-TOP-9A is not applicable to the design of pipe whip restraints. For-the design of pipe whip restraints, the ap>1icant has referenced the approved Topical Repont BN-TOP-2, " Design for, Pipe,9reak Effects." In Topical Report BN-TOP-2, a commitment is made to assdne no' more than a 10% increase in material yield strength for the design of pip;e whib rostraints. We believe thatinthissituationBN-TOP-2is-thecontrollingl document. Therefore, we require the applicant to provide assurance that tha 10% limit has not been exceeded. We will report the resolution of this 2 issue in a supplement to this Safety Evaluation Report. The applicant has stated that a break is not pos-tulated at any branch connection unless it is one of tae two highest stressed intermediate points in the piping run. Standard Review Plan 3.6.2 states that break locations should be postulated at the terminal ends of the branch run except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant effect on the main run behavior. In many cases the branch connection is the terminal end for the branch run which in turn should have the required intermediate breaks. Therefore the applicant will be required to postulate breaks at the branch connec+. ion unless it meets the criteria stated above. The applicant also states th.. Sce a high energy piping system has been analyzed and break locations have been identified and evaluated, the postulated intermediate break locations will' not be altered upon subsequent stress reanalysis, provided there have been no major changes in the routing of the high energy piping in the vicinity of the original intermediate breaks. However, if the reanalysis results in stresses or usage factors in excess of the criterion of 2.45, or 0.1 respectively, additional breaks will be considered in these areas. We find this acceptable provided breaks at the new highest stress locations are not significantly apart from the orig. locations and do not result in consequences to other safety related systems. The results of the LOCA and jet impingement analysis of the major reactor coolant loop components has not been included in tne FSAR (Sect. 3.6.3.3) as committed to by the applicant. The stress intensity values provided for ASME Section III Code Class 1 piping postulated break locations shown in the tables of FSAR Section 3.6 were calculated by the rules for ASME Section III Code 3-3
Class 2 and 3 piping per the applicant. These stress intensities have.not been updated to reflect the rules for Class 1 piping as committed to by the ) applicant. We have reviewed addicional information from the applicant concerning the methods used to calculate limited breck sizes and finite break opening times for postulated pipe breaks in the reactor coolant system. Babcock & Wilcox has performed'a three-dimensional, inelastic, time-history analysis of the response of the E dland reactor coolant system to postulated pipe breaks at various locations within the system. This analysis assumed an instantaneous pipe severance, then predicted the movement of the broken pipe while taking into consideration deformation of the reactor vessel and other components, deformation of the broken pipe itself, and deformation of the pipe whip restraints. The ANSYS computer code was used for performing the structural analysis while the CRAFT computer code developed the hydraulic forcing function. Our review found '.his method of analysis to be acceptable. Subject to resolu-tion of the above, our overall findings with respect to postulated pipe failures are as follows. The applicant has proposed criteria for determining the location, type and effects of postulated pipe breaks in high energ-piping systems and postulated pipe cracks in moderate energy piping systems. The applicant has used these postulated effects to evaluate the design of systems, components, and struc-tures necessary to safely shut the plant down and to mitigate the effects of these postulated piping failures. He has indicated that pipe whip restraints,- jet impingement barriers, and other such devices will be used to mitigate the effects of these postulated piping failures. We have reviewed these criteria and have concluded that they provide for a spectrum of postulated pipe breaks and pipe cracks which includes the most likely locations for piping failures, and that the types of breaks and their effects are conservatively assumed. We find that the methods used to design the pipe whip restraints provide adequate assurance that they will function properly in the event of a postulated piping failure. We further conclude that the use of the applicant's proposed pipe failure criteria in designing 3-4
-. 7 I the systems, components, and structures necessary to safely shut the plant down and ty mitigate the consequencks of these postulated piping. failures pro-vides reasonable assurance of their ability to perform their safety function following a failure in high or moderate energy piping systems. The applicant's criteria comply with Standard Review Plan Section 3.6.2 except as noted above and satisfy the applicable portions of General Design Criterion 4. 3.7.3 Seismic Subsystem Analysis The review performed under Standard Review Plan Section 3.7.3 includes the applicant's seismic analysis methods, modeling techniques and criteria for the seismic Category I piping systems, the reactor coolant system and the reactor internals. The seismic analysis method used for-Category I subsystems other than NSSS subsystems and piping is stated to be that of BC-TOP-4-A. This is a staff approved Topical and is considered an acceptable method. Additional informouco is required concerning the dynamic analysis methods of the major reactor coolant loop components. The computer program used and the means of verification of same should be given. It is stated that torsional, shearing, bending and axial deformations are included in the flexibility cal-culations. It is not clear, however, if this includes the deformations of the structures to which the components are mounted. There is no justification to show that a sufficient namber of modes have been considered to assure partici-pation of all significant modes. Information is required as to how significant effects such as piping interactions, externally applied structural restraints, hydrodynamic loads, etc., c"e considered. A three dimensional seismic analysis is performed on the reactor coolant loop using the response spectra and the normal mode approach. The seismic evalua-tion methods and procedures described by the applicant for the reactor coolant loop are acceptable except as noted above. 3-5 w + 9
r--- - The applicant's methods or justification for determining the number of earthquake, cycles meets or exceeds the minimum required per Section 3.7.2 of the SRP. The applicant has described the mathematical models and methods used to analyze the Category I NSSS subsystems, subsystems other than NSSS, and Cate-gory I piping. The methods used for the NSSS subsystems are dynamic analysis, simplified dynamic analysis, and equivalent static analysis. The systems other ~ than NSSS and exposed piping are analyzed using the techniques of the staff approved Topical BC-TOP-4-A. The non-NSSS Category I piping is analyzed using a dynamic analysis, equivalent dynamic analysis or static analysis. The apoli-cant has indicated the method used for each system and provided technical justification for use of the equivalent static load and the simplified dynamic analysis methods. The applied seismic loads are obtained from the appropriate j response spectrum. The three components of earthquake motion, multiple supports, and the torsional effects of eccentric masses have been satisfactorily accounted for. We have reviewed the applicant's procedure and conclude that the seismic evaluation methods and procedures described for the NSSS and non-NSSS subsystems and equipment are acceptable based on the resolution of the following: 1. There is no statement as to how the applicant determines that all the significant modes are considered in the dynamic analysis methods or the equivalent dynamic analysis for the piping systems. 2. The applicant has listed two computer programs used in the seismic analysis (ME 632 and ME 101). A brief description is given as to what ME 101 does but there is no verification information given. It is not clear as to how ME 632 was verified after the significant program changes mentioned were incorporated. 3. The primary system component analysis in the B&W scope of supply and' the Midland plant piping design use the square root of the sum of the squares method for combining seismic modal responses. The effects of closely spaced modes upon the analysis of the NSSS was not addressed. Our position is that when closely spaced modes exist (frequencies within l 3-6 j
1 ~ 10% of each other), an acceptable method to obtain the system response is to combine the responses of the closely spaced modes by the absolute sum method and combine this sum with the other andal responses by SRSS. Other approaches,, such as Regulatory Guide 1.92 which give an equivalent degree of conservatism and which are adequately justified are also acceptable. Therefore we will require that the effects of closely spaced modes be addressed for the NSSS. To demonstrate the adequacy of the Mid-land plant, design representative systems were chosen and reanalyzed using the absolute sum method of Regulatory Guide 1.92. The reana7ysis of these systems per Regulatory Guide 1.92 showed only one support exceeded design allowables. However, the comparative analysis showed the seismic piping stresses increased up to 6 maximum of 13% and the seismic nozzle and anchor loads increased uo to a maximum of 45%. The seismic load increase on the supports was such that a detailed reanalysis of 25% of those supports considered was deemed necessary by the applicant. Our position is that the applicant should show that all Category I piping, supports, anchors, etc. are adequate if the seismic modes are combined per Regulatory Guide 1.92. 4 The seismic analysis review of the buried piping cannot be completed until the effect of the excessive soil settlement has been determined, the properties of the soil through which the buried piping runs determined, and an acceptable stress or deflection criteria developed for the piping. 5. The applicant has satisfactorily described how the icteraction of seismic Category I piping with non-Category I piping is accounted for when the two classes are connected. A description is also required as to how Category I piping is protected from adjacent non-Category I piping. The applicant has described the analysis of the reactor internals and the methods used in FSAR Section 3.9.2. The comments made regarding that section are aise applicable to this section. i 3-7
3.9 Mechanical Systems and Comoonents The review performed under Standard Review Plan Section 3.9.1 thru 3.9.6 pertains to the structural integrity and operability of variou's safety-related mechanical components in the plant. Our review is not limited to ASME Code components, but is extended to other components sucr as control rod drive mechanisms, certain reactor internals, pumps and valves which perform a safety related function, and piping designed to industry standards other than the ASME Code. We review such things as load combinations, allowable stresses, methods of analysis, summary results, seismic qualification, pre-operational testing, and inservice testing of pumps and valves. Our review must arrive at the conclusion that there is adequate assurance of a mechanical component performing its safety-related function under all postulated combina-tions of normal operating conditions, system operating transients, postulated pipe breaks, and seismic events. 3.9.1 Scecial Taoics For Mechanical Comoonents The review performed under Standard Review Plan Section 3.9.1 pertains to the design transients, computer programs, experimental stress analysis and elastic-plastic analysis methods that were used in the analysis of seismic Category I ASME Code and non-Code items. The applicant has provided a complete list of transients to be used in the design and fatigue analysis of all Code Class 1 and CS components and of component supports and reactor internals within the reactor coolant pressure boundary. - The number of events postutated-for each transient has been included and is acceptable. The methods of analy.is that the applicant has employed in the design of all seisric Category I ASME Cole Class 1, 2, and 3 components, component supports, reactor internals, and otter non-Code items are in conformance with Standard Review Plan Sectic, 3.9.'. and satisfy the applicable portions of General Design Criteria 2, 4, 14, and 15. The criteria used in defining the applicable transients and the computer codes and analytical methods used in the analyses provide assurance that the 3-8
calculations of stresses, strains, and displacements for the above noted items confctm with the current state-of-the-art and are adequate for the design of these items. In addition, the Energy Technology Engineering Center has performed a partial independent piping analysis of the Midland Decay Heat Removal and Core Flood-ing System. The purpose of this analysis was to verify that the piping system meets the applicable ASME Code requirements and to provide a cneck on the applicant's ability to correctly model and analyze their piping systems. The analysis performed was considered to be preliminary because the "as built" drawings were not available and additional information requested concerning snubber and support stiffnesses, response spectra, loading conditions other than design conditions, and clarification were not received by our consultants, ETEC. Therefore, based on the available information the design, normal and seismic conditions were analyzed. The results of this analysis verify that the referenced piping system meets the applicable ASME Code requirements for the conditions checked and provides a partial check of the applicant's ability to correctly model and analyze its piping s,;tems. 3.9.2 Dynamic Testing and Analysis of Systems, Comoonents and Eouioment The review performed under Standard Review Plan Section"3.9.2 pertains to the criteria, testing procedures, and dynamic analyses employed by the applicant to assure the structural integrity and operaoility of piping systems, mechan-ical equipment, reactor internals and their supports under vibratory loadings. This review is divided into four parts, each of which is discussed briefly below. During the Midland Plant's preoperational and startup test program, the applicant will test various piping systems for abnormal steady-state or transient vibration and for restraint of thermal growth. The systems to be tested include all safety-related ASME Code Class 1, 2, and 3 systems, all other high energy piping systems, and all seismic Category I moderate energy piping systems. This test program will consist of a mixture of instrumented 3-9
i..L. __ L, l measurements and visual observation by qualified personnel. Our findings are as follows-: } The vibration, therma'l expansion, and dynamic effects test program which will .be conducted during startup and initial operation on specified high and mod-erate energy piping, and all associated systems, restraints and supports is an acceptable program. The tests provide adecuate assurance that the piping and piping restraints of the system have been designed to withstand vibra-1 tional dynamic effects due to valve closures, pump trips, and other operating modes associated with the.esign basis flow conditions. In addition, the tests provide 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 operation. This test program complies with Standard Review Plan Section 3.9.2 and constitutes an acceptable basis for fulfilling, in part, the requirements of General Design Criteria 14 and 15. 9 I f 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 tnose expected during } operation will not cause unanticipated flow-incuced vibrations of significant I magnitude or structural damage. The applicant has described the preoperational tests that Gill Ye'perto m6d' ~ - ~ on the Midland reactor internals to verify their long term structural integrity l under normal operating loads. The applicant has referenced the approved topical report BAW-10039,. Revision 0 and Supplement 1, " Prototype Vibration j Measurement Results for B&W's 177-Fuel Assembly, Two-Loop Plant," which l describes the tests run on the prototype reactor internals at Oconee 1. The Oconee Unit I reactor internals have been accepted as the valid prototype for i the 177-fuel assembly units and Davis Besse Unit 1 is the limited valid proto- } type for the surveillance specimen holder tube design which includes Midland Units'1 and 2. The applicant has stated that the tests and inspections at 3-10 r
Midland will be in compliance with Regulatory Guide 1.20 rega" ding non prototype Category I reactor internals. The operational vibration program planned for the reactor internals provides an' acceptable basis for verifying the Usign adequacy of these internals under test loading conditions comparable to those that will be experienced during operation. The combination of tests, predictive analysis, and post-test inspec-tion provide adequate assurance that the reactor internals will, during their ~ service lifetime, withstand the flow-induced vibrations of reactor operation without loss of structural integrity. The conduct of the preoperational vibration tests is in conformance with the provisions of Regulatory Guide 1.20 and Standard Review Plan Section 3.9.2, and satisfies the applicable requirements of General Design Criteria 1 and 4. 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 earth-quake. The original analysis of the reactor internals for this particular loading combination was reported in a topical report referenced by the applicant, BAW-10008, Part 1, " Reactor Internals Stress and Oeflection Analysis Due to LOCA and Maximum Hypothetical Earthquake." This report was reviewed and approved by the staff in September 1972. This report did not include the loading effect due to asymmetric cavity pressurization due to a postulated pipe break within a subcompartment such as the reactor vessel cavity. There-fore, this loading effect was not-considered in the applicant's original analysis. ~ Subsequently, the staff has approved a Babcock & Wilcox Topical Report BAW-10132 which describes newly leveloped analytical tecnniques for calculating LOCA related loads. The staff's position is that reanalysis of the Midland reactor internals and core support structures is required. This analysis should include the load-ing conditions of BAW-10008 with the addition of reactor vessel motion caused by asymmetric cavity pressure differentials. The thermal-hydraulic analysis shall be in accordance with the staff approved version of BAW-10132. Also, that the effects of the cavity pressurization loads on the reactor vessel 3-11 (
~ supports be based on the final approved redesign of these supports as discussed in Section 3.9.3. The applicant is now ' reanalyzing the reactor internals and unbroken loops of thq' reactor coolant pressure boundary, including the supports, for the com-bined loads due to a loss-of-coolant accident and safe shutdown earthquake, and the effects of asymmetric cavity pressurization. The applicant has committed to the use of BAW-10132 in the reanalysis but there is no commitment concerning using the loading conditions and allowables of BAW-10008. The results of the reanalysis have not been incorporated into the FSAR as yet (Talles 3.9-8, -9, -10, -11). We note that the applicant is not perfo ming a plant-specific analysis of the Midland 1 and 2 reactor internals for combined SSE and LOCA loads, including asymmetric cavity pressurization. Rather, it is using the results of the analysis of the Davis Besse 2 and 3 reactor internals as a baseline, then using scaling factors to produce calculated stresses for the Midland internals. Due to the fact that the physical configuration nf the Midland and Davis Besse internals are not hntical, we asked for further information. During a meet-ing on April 20, 1979, with the applicants, Babcock & Wilcox discussed-tha details of this analysis and provioed justification of the conservatism of its approach. We find t.1at the applicant's proposed method of analyzing the Midland reactor internils is acceptable except as noted above. 3.9.3 ASME Code Clas: 1, 2 and 3 Components, Component Supports, and Core Support Structures Our review under Standard Review Plan Section 3.9.3 is concerned with the structural integrity and operability of pressure-retaining components, their supports, and core support structures which are cesigned in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, or earlier industry standards. 3-12 ~
I The applicant describes loading combinations, design transients and stress limits. Wp have reviewed the loading combinations and allowable. limits used 1 by the applicant for all components and find they provide a level of safety equivalent to that provided in Regulatory Guide 1.48 with the' exceptions noted. Table 3.9-1 which lists the loading combinations and stress limits for safety-related mechanical systems and components is not complete for all items. Table 2.9-3b " Loading Combinations and Stress Limits for Seismic Class 1 Components (NSSS Scope)" should also include the applicable transients of Table 3.9-2. In Tables 3.9-9, -10, -11 and -15, the calculated stress intensity exceeds the allowable stress show1 in some instance. It is noted that an elastic plastic evaluation was performed for justification in these instances. The allowables used in t.hese etaluations should be shown. The applicant has not as yet completed the assessment of the effects of building and ground settlement on piping and other mechanical components, and what remedial actions, if any, ire required. Concerns yet to be resolved are: (1) The apparent excessive stresses on the buried piping due to soil settlement. (2) Anchor and component loads due to the deformation of the undarground piping. (3) The prediction and monitoring of future settlements of the underground piping due to the questionable soil properties. (4) The effects of the rupture of non-safety related piping on safety-related piping, compon'nts and structures. e (5) Seismic Analysis of the deformed buried piping considering questionable soil properties and deformation of the piping. (6) The effects of overburden loads such as soil dead weight, heavy equipment, etc. on the buried piping taking into consideration that the soil properties and pipe bedding qualities are uncertain. 3-13 -~
We have reviewed the criteria used by the applicant in the de i piping, ASME Class 1, 2 and 3 components, and core support structures s gn of ANSI 831.7 noted that.this subject wa's extensively discussed and substantially resol It is during pre-docketing meetings with the applicant on compliance Guide 1.48, " Design Limits and Loading Combinations for Seismic C with Regulatoi, Fluid System Components." ry I The applicant was requested to provide assurance that under en; erg nc faulted plant conditions ASME Class 2 and 3 piping in essential s e y and i.e., those systems required for safe plant shutdown and continued shut
- ystems, heat removal, will maintain their functional capability own refers to the ability of the piping system to deliver rated flow when the culated piping stresses predict some amount of plastic deformation cal-The results of recently completed studies performed at Oak Ridge Nati e.
Laboratory and Battelle-Columbus Laboratory under NRC contract indicat onal piping components stressed to levels permitted by the ASME Code Lev e that 0 stress limits will retain pressure boundary integrity but could pos ibl n experience large plastic deformations. s y Based upon the results of these studies which have been documented in report ORNL/Sub-2913/8 , " Evaluation of the Plastic Characteristics of Piping Products in Relation to ASME C Criteria," we have developed criteria to assure piping functional cap bili ode The applicant has -met-these urheria oy designing all essential ASME Cl a ty. and 3 piping to maintain their primary stresses below their yield st engt ass 2 under emergency and faulted plant conditions. The applicant was asked to commit to comply with Regulatory Guide 1 121 " Bases for Plugging Degraded PWR Steam Generator Tubes " Compliance with this guide assures that any degraded tubes discovered during the inservice inspections will be removed from service by plugging bef periodic structural integrity is compromised. ore their with the intent of Regulatory Guide 1.121.The applicant has committed to comp Our review under Standard Review Plan Section 3.9.3 evalua thich the applicant has used to assure operability of active p a valves. umps and Active pumps and valves are those which must perform a mechanical { 3-14
motion in order to shut down the plant or mitigate the consequences of an accident.. For instance, under accident conditions, certain valves' must open or close, and certain pumps are required to start. On the other' hand, passive pumps and valves are only required to maintain their position'during an accident. We have reviewed the applicant's program for assuring the operability of " active" pumps and valves. The applicant has supplied information concerning the analyses and testing performed to assure the cperability of these pumps and valves under their specified loading conditions. For the Midland plant, the 18 inch containment purge valves are classified as active. The information submitted to date concerning the design of these valves is not sufficient for us to complete our review. We will report our findings in a supplement to this Safety Evaluation Report. We have reviewed the criteria used by the applicant in designing its ASME Class 1, 2 and 3 safety and relief valves, their attached piping, and their supports. We h ve specifically reviewed the applicant's compliance with Reg-ulatory Guide 1.67, " Installation of Overpressure Protection Devices." Excludino the seismic Qualification of certain safety and relief valves, we feel that the overall design is sufficiently acceptable to report ;ur find-ings. These findings are contingent upon successful resolution of seismic qualification with the Equipment Qualification Branch and are as follows. The criteria used 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 allowable stress and strain limits for the materials of construction. Limiting the stresses under the loading combinations associated Lith the actuation of these pressure relief devices provides a conservative basis for the design ano installation of the devices to withstand these loads without loss of structural integrity or impair-ment of the overpressure protection function. The criteria used for the design and installation of ASME Class 1, 2 and 3 overpressure relief devices con-stitute an acceptable basis for meeting the applicable requirements of General 3-15 ^^ ^
Design Criteria 1, 2, 4,14 and 15 and are consistent with those specified in Regulatory Guide 1.67 and Standard Review Plan Section 3.9.3. We have aviewed the criteria used by the applicant in the design of ASME i Class 1, 2 and 3 component supports. All supports for ASME Code components 1 supplied by Bechtel and some supplied by Babcock & Wilcox have been designed in accordance within Subsection NF of the ASME Code, Section III. The remain-ing supports for ASME Code components supplied by Babcock & Wilcox were designed to the same criteria as the supported component because Subsection NF .had not yet been published at the time they were being designed. This is an acceptable approach. The failure of three reactor vessel hold down studs due to decreased resistance to stress corrosion (as indicated in Teledyne Reports TR-3887-1 and addendum) because of excessive surface hardness has caused Consumers Power to propose modifications to the reactor vessel supporting scheme. These modifications consist of detensioning the remaining studs to reduce the service stresses and modifying the existing shield plug support brackets to provide additional upper lateral support. Based on our review and subsequent meetings and con-versations with the applicant, we find this design concept to be an accept,able approacn to satisfy the requirements of the Standard P' view Plan 3.9.3. riowever, final approval of this redesign will be basec. .he final detailed analysis and report. We have recently issued two Regulatory Guides related to the design of component supports. These are 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." Consistent with other recent operating license reviews, we have determined the applicant's compliance with two major issues addressed ~ by the guides. These subjects are buckling of component supports and the design of bolts used in component supports. With respect to buckling, the applicant complies with the faulted condition buckling criteria of the two guides. With respect to bolt design, the applicant has supplied information concerning the design of not only the bolts, but also the baseplates into 3-16
3. which the bolts are inserted and which the bolts connect to the underlying concrete or steel structures. We have reviewed this information and conclude that the applicant has adequately complied with Regulatory Guides 1.124 and 1.130. As noted previously in this section, the applicant is currently assessing the effect of building settlement upon piping and other mechanical equipment. It is possible that some component supports might be affected. We will review the applicant's assessment upon its completion and report our findings in a supplement to this Safety Evaluation Report. We have reviewed the applicant's criteria for designing, installing and testing its hydraulic snubbers. The only hydraulic snubbers in the Midland plant are attached to the reactor coolant pumps. These snubbers have been designed in accordance with Subsection NF of the ASME Code, Section III. They will be periodically inspected in accordance with the Midland plant Technical Specifications. We find that these actions by the applicant adequately assure the operability of its hydraulic snubbers. 3.9.4 Control Rod Drive Systems Our review under Standard Review Plan Section 3.9.4 covered the design of the control rod drive system up to its interface with the control rods. We reviewed the analyses and tests performed to assure the structural integrity and oper-ability of this system during normal-eperation and under accident conditions. We also reviewed the life cycle testing performed to demonstrate the relia-bility of the control rod drive system over its 40 year life. The applicant has referenced Topical Report BAW-10029, Revision 3, " Control Rod Drive Mech-anism Test Program," which describes the tests and analyses performed far the prototype of the control rod drive system installed in the Midland plant. We approved this topical report on February 25, 1976. During our review we verified that the seismic loads assumed'in the stress analyses of the topical report were large enough to envelop the Mid1:..c site specific seismic loads. We find BAW-10029 to be an acceptable reference for the Midland plant. Our findings are as follows. 3-17
r The design criteria and the testing program conducted in verification of the mechanical operability of life cycle capabilities of the control rod drive system are in conformance with Standard Review Plan Section 3.9.4. The use of these criteria provide reasonable assurance that the system will function reliably when required, and form an acceptable basis for satisfying the mechanical reliability stipulations of General Design Criterion 27. 3.9.5 Reactor Pressure Vessel Internals Our review under Standard Review Plan Section 3.9.5 is concerned with the load combinations, allowable stress limits, and other criteria used in the design of the Midland reactor internals. We have reviewed the physical configuration of the Midland reactor internals and would note that they are essentially identical to the reactor internals at several operating plants which use the Babcock & Wilcox 177-fuel assembly twc-loop nuclear steam supply system. As stated in Section 3.9.2 of this Safety Evaluation Report, the prototype of the Midland reactor internals were tested for steady state and transient vibration at Oconee Unit 1. These tests and analyses plus the successful operating experience at Oconee and other similar plants indicate that the Midland reactor internals will remain structurally sound during their design life of 40 years of normal operation. Additionally, the applicant has ref-erenced a Babcock & Wilcox Topical Report, BAW-10051, Revision 1, and Supplement 1, " Design of Reactor Internals and Incore Instrument Nozzles for Flow Induced Vibrations," which we reviewed and found acceptable-in Aprt1- ~~ 1979. Our review is also concerned with the structural integrity of the Midland reactor internals under the combination of loads that would be experienced during postulated events such as the safe shutdown earthquake and loss-of-coolant accident. As noted in Section 3.9.2 of this Safety Evaluation Report, the applicant is ( currently reai.clyring the reactor internals for the combined loads due to a i simultaneous safe shutdown earthquake and loss-of-coolant accident, including 3-18
~ any asymmetric cavity pressurization effects. This reanalysis for this one particular load combination is the only unresolved issue with respect to the design of the Midland reactor internals. Therefore, subject to resolution of this issue, our findings are as follows. The specified transients, design and service loadings, and combination of loadings as applied to the design of the Midland reactor internals provide reasonable assurance that in the event of an earthquake or of a system transient during normal plant operation, the resulting deflections and associated stresses imposed on these reactor internals would not exceed allowable stresses and deformation limits for the materials of construction. Limiting the stresses and deformations under such loading combinations provides an acceptable basis for the design of these reactor internals to withstand the most adverse loading events which have been postulated to occur during service lifetime without loss of structural integrity or impairment of function. The design procedures and criteria used by the applicant in the design of the Midland reactor internals comply with Standard Review Plan Section 3.9.5 and constitute an acceptable basis for satisfying the applicable requirements of General Design Criteria 1, 2, 4 and 10. 3.9.6 Inservice Testing of Pumps and Valves The applicant has submitted a description of its preliminary inservice testing program for pumps and valves. The program includes both baseline-preservice testing and periodic inservice testing. 4t provides for-both-functional testing of components in the operating state and for visual inspection for leaks and other signs of degradation. The date of the applicant's construction permit (December 12, 1972) places this plant under 10 CFR 50.55a(g)(2) which requires design and access to comply with the 1971 Edition of Section XI of the ASME Boiler and Pressure Vessel Code through the Winter 1973 Addenda. Since inservice testing require-ments for pumps and valves were not included in the Code.until the Summer 1973 Addenda of the 1971 Edition, the applicant has chosen to optionally-J f 3-19
meet the requirements of the 1977 Edition through the Winter 1977 Addenda to the extent ~ practical and has requested relief from certain Code requirements. The. surveillance requirement for which the program is applicable is: Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). We have not completed our detailed review of the applicant's submittal. However, based on our oreliminary review, we find that it is impractical within the limitations of design, geometry, and accessibility for the appli-cant to meet certain of the American Society of Mechanical Engineers C0de requirements. Imposition of those requirements would, in our view, result in hardships or unusual difficulties without a compensating increase in the level of quality or safety. Therefore, based on the resolution of the follow-ing concerns, and pursuant to 10 CFR 50.55a(g)(2) and (g)(6)(1), the relief tnat the applicant has requested from the pump and valve testing requirements of the American Society of Mechanical Engineers Code is granted for that portion or the initial 120-month period during which we complete our review. Since the applicant's request for relief has been granted and the applicant will be required to comply with Sectitm-XI ef the-American-Scciety of ~ Mechanical ~ Engineers Boiler and Pressure Vessel Code endorsed by 10 CFR 50.55a 12 months prior to issuance of an operating license and/or the Technical Specifications, we find the Midland inservice testing program for pumps and valves acceptable for the period described above. Certain areas of concern during our preliminary review are described below. (1) The applicant has proposed to full stroke exercise the pressure isolation check valves in the safety injection system during each cold shutdown longer than 72 hours. If the applicant is proposing to full stroke 3-20
exercise these check valves by passing rated flow through them at cold shutdown, we will require a description of the planned test procedure to assure protection of the reactor vessel and reactor internals due to safety injaction' hydrodynamic loads. (2) There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS pressure). There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure. In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the 11terface between the high pressure RCS and the low pressure systems. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA. Periodic leak testing of pressur e isola-tion valves shall be performed after all disturbances to the valve are completed. The pressure isolation valves to be tested will be listed in the technical specifications. The applicant has agreed to categorize their pressure isolation valves for the pressurizerspraymakeupandpurification,andthedecayheatremovalsyst$ ems as Category A or AC. These categorizations meet our requirements and we find them acceptable. Pressure isolation valves are required to be Category A or AC and to meet the appropriate valve leak rate test requirements of IWV-3420 of Section XI of. the ASME Code-except as-disettesed-below--4he aMowstrit- - - --- -- leakage rate shall not exceed 1.0 gallon per minute (GPM) for each valve as stated in the technical specifications. The applicant has committed to test all pressure isolation valves to the 1.0 GPM leak rate criteria. Limiting Conditions for Operation (LCO) will be added to the technical specification which will require corrective action, i.e., shutdown or system isolation when the leakage limits are not met. Also, surveillance requirements, 3-21
l .} ~. i l l which will state'the acceptable leak rate testing frequency, will.be provided l in the technical specifications. l-We conclude that Consumer's Power commitments to periodic leak testing of pressure iso 16 tion valves between the reactor coolant system and low pressure systems will provide reasonable assurance that the design pressure of the low l pressure systems will not be exceeded, and thus reduce the probability of an occurrence of an inter-system LOCA. Criterion 55 of the General Design (' Criteria of Appendix A of 10 CFR 50 partially considers this matter. ( I i l l l ,e a e, _q. m .,es 6 . an e e l l l 3-22}}