NRC Generic Letter 1987-02
A -X.'$Jt Rt44UNITED STATESNUCLEAR REGULATORY COMMISSION* WASHINGTON. D. C, 20555FEB 13 t Vi.ITO: -All Holders of Operating Licenses Not ReviewedLicensing Criteria on Seismic Qualification ofto CurrentEquipmentGENTLEMEN:SUBJECT:VERIFICATION OF SEISMIC ADEQUACY OF MECHANICAL AND ELECTRI-CALEQUIPMENT IN OPERATING REACTORS, UNRESOLVED SAFETYISSUE((USI) A-46 (Generic Letter 87-02)As a result of the'technical' resolution of 'USI A-46, "Seismic Qualificationof Equipment in Qperating Plants," the NRC has concluded that the seismic ade-quacy of certain equipment in operatinginuclear power"'plants must be reviewedagainst seismic criteria not in use when these plants were licensed. The tech-nical basis for this conclusion is set forth in References 1 and 2.Direct application of current seismic criteria to older plants could requireextensive, and probably impracticable, modification of these facilities. Analternative resolution of this problem is set out in the enclosure to thisletter. This approach makes use of earthquake experience data supplemented bytest results to verify the seismic capability of equipment below specified 'earthquake motion bounds. In the staff's judgment, this approach is the mostreasonable and cost-effective means of ensuring that the purpose of GeneralDesign Criterion 2 (10 CFR Part 50 Appendix A) is met for these plants.Because affected plants are being asked to carry out this evaluation againstcriteria not used to establish the design basis of the facility, this resolutionis a backfit under 10 CFR 50.109. The backfit analysis and findings may befound in the Regulatory Analysis (Reference 2) at pp. 31.Seismic verification may be accomplished generically, as described in theenclosure. Utilities participating in a generic program should so state intheir reply to this letter, identifying the utility group and the scheduleestablished for completion of the effort. The implementation schedule will benegotiated with utility groups or individual utilities in accordance with theNRC policy on integrated schedules for plant modifications. See Generic Letter83-20, May 9, 1983. Utilities not participating in a generic review may beallowed some additional time to complete the review.' i.. ICR~ I~cct5C0 ooow3IDA'-i51--
We therefore request*that you provide within 60 days of receipt of this letter aschedule for implementation of the seismic verification program at your facility.Sincerely,arold R. Denton, Direct eOffice of Nuclear Reactor RegulationEnclosure:Seismic Adequacy VerificationProcedureReferences:(1) NUREG-1030, "Seismic Qualification of Equipment in OperatingNuclear Power Plants (USI A-46)," February 1987(2) NUREG-1211, "Regulatory Analysis for Resolutionof Unresolved Safety Issue A-46, Seismic Qualification ofEquipment in Operating Plants," February 1987C B X Y90_* This request is covered by Office of Management and Budget ClearanceNumber 3150-0011 which expires September 30, 1989. Comments on burdenand duplication may be directed to the Office of Management and Budget,Room 3208, New Executive Office Building, Washington, DC 20503. -2
-IENCLOSURESEISMIC ADEQUACY VERIFICATION PROCEDUREThe proposed procedure for verifying seismic adequacy of equipment is addressedin the following paragraphs. Each licensee will be required to perform theverification steps and submit a report to the NRC including an affidavit thatthe verification has been completed and all equipment within the scope definedbelow has been found to be acceptable. A generic resolution will be acceptedin lieu of a plant-specific verification review subject to the guidance presentedherein.1. Scope of Seismic Adequacy ReviewEach licensee will determine the systems, subsystems, components, instrumenta-tion, and controls required during and following a design-basis seismic eventusing the following assumptions:(1) The' seismic event does not cause a loss-of-coolant accident (LOCA), asteam-line-break accident (SLBA), or a high-energy-line-break (HELB), anda LOCA, SLBA, or HELB does not occur simultaneously with or during a seis-mic event. However, the effects of transients that may result from groundshaking should be considered.(2) 'Offsite power may be lost during or following a seismic event.(3) The plant must be capable of being brought to a safe shutdown conditionfollowing a design-basis seismic event.The equipment to be included is generally limited to active mechanical and elec-trical components and cable trays. Piping, tanks, and heat exchangers are notincluded except-that those tanks and heat exchangers that are required toachieve and maintain safe shutdown must be reviewed for adequate anchorage.Seismic sys.tem interaction is included in the scope of review to the extentthat equipment within the scope must be protected from seismically induced-physical interaction with all structures, piping, or equipment located nearby.Lessons learned from studies of nuclear and nonnuclear facilities under earth-,.quake loading indicate that the effect of failure of certain items--such assuspended ceilings and lighting fixtures--could influence the operability ofequipment within the scope of reviews. Instrument air lines and electrical and..'instrumentation cabling must be verified to have sufficient flexibility fromthe connection to equipment so that relative movement of anchor points willnot jeopardize their integrity. Air lines and electrical and instrument cabl-ing are not included in the scope of review except for that portion from theequipment item to the first anchor point. The failure of masonry walls thatcould-affect the operability of nearby safety-related equipment is of concern.However, this concern has been addressed by IE Bulletin 80-11, which requiresthat all such masonry walls be identified and re-evaluated to confirm theirdesign adequacy under postulated loads and load combinations. This concernis, therefore, not considered as part of A-46 implementation. The requiredseismic interaction reviews will be based on, and consistent with, observationsmade in the seismic experience data base augmented by expert judgement of1Enclosure SQUG/SSRAP. The review procedures will be reviewed by the NRC staff and SSRAPprior to plant specific implementation.For some pressurized water reactor plants, the seismic adequacy of auxiliaryfeedwater (AFW) systems has been verified by licensee'actions taken in responseto Generic Letter 81-14, dated February 10, 1981. Review of the AFW systemsmay be deleted from consideration under USI A-46 if staff acceptance has beendocumented in a Safety Evaluation Report or if the licensee has committed tomeet the requirements of the generic letter.For the purpose of seismic adequacy verification, the following guidance isgiven. Each licensee must identify equipment necessary to bring the plant toa hot shutdown condition and maintain it there for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The72-hour period is sufficient for inspection of equipment and minor repairs, ifnecessary, following a safe-shutdown earthquake (SSE) or to provide additionalsource(s) of water for decay heat removal, if needed, to extend the time at hotshutdown.Equipment required includes that necessary to maintain the supporting functionsrequired for safe shutdown. For all equipment within the defined scope, theverification must closely follow the procedure outlined in paragraph 2 below.Each licensee must show practical means of staying at hot shutdown for a minimumof 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If maintaining safe shutdown is dependent on a single (not redun-dant) component whose failure, either due to seismic loads or random failure,would preclude decay heat removal by the identified means, the licensee mustshow that at least one practical alternative for achieving and maintaining safeshutdown exists that is not dependent on that component.Each licensee must develop an equipment list. This list will include all equip-ment within the required scope.The equipment to be considered depends on the functions required to be performed.'Typical plant functions would include:(1) Bring the plant to a hot shutdown condition and establish heat removal.(2) Maintain support systems necessary to establish and maintain hot shutdown.(3) Maintain control room functions and instrumentation and controls necessaryto monitor hot shutdown.(4) Provide alternating current (ac) and/or direct current (dc) emergency poweras needed on a plant-specific basis to meet the above three functions.2. General Verification Procedure for Plant-Specific ReviewThe licensee will be required to conduct a plant walk-through and visual inspec-tion of all identified equipment items necessary to perform the functionsrelated to plant shutdown. The inspection team must consist of as a minimum,(1) an experienced structural engineer familiar with seismic anchoragerequirements2Enclosure
(2) an. experienced mechanical engineer familiar with plant mechanical equipment(3) an experienced electrical.engineer familiar with plant electrical equipmentFurthermore, an operations supervisor or a licensed Senior Reactor Operatormust be available for consultation before and during the walk-through process.Not all members of the inspection team are-required to participate in all partsof the walk-through; however, appropriate technical, expertise must be includedfor each review area, and a person with proper structural background.mustalways be present to inspect the anchorage for all equipment.As an alternative, licensees may use consultants instead of their staff for(1), (2),-and (3) above.Before.-the walk-through inspection,-the licensee will be required to verifythat the appropriate data base spectra envelope the site free-field spectra atthe ground surface defined for the plant. There are a number of nuclear plantswhose free-field'SSE spectra are defined at the foundation level. For theseplants, an estimate of the free field spectra at the ground surface must bemade-for comparison with the data base bounding spectra..-The licensee mustidentify all equipment on the plant's equipment list that is located at anelevation higher than 40 feet above grade level.* For equipment above 40 feet,one-and-one-half times the appropriate data base bounding spectrum (defined inparagraph.6 below) must envelope the floor response spectra.for the equipment.For those cases where floor response spectra are needed, NUREG/CR-3266,"Seismic and Dynamic Qualification of Safety-Related Equipment in OperatingNuclear Power Plants: The Development of a Method to Generate Generic FloorResponse.Spectra," may bemused as one alternative to develop the necessaryfloor response spectra on a-case-specific basis. The appropriate boundingspectra for equipment belonging to the original eight types in the-data baseare defined in paragraph 6 below. For equipment types that are not included inthe eight types in the data base but that exist in the data base plants, andfor equipment unique to nuclear plants, the appropriate bounding spectra.aredefined in paragraph 7 below.The walk-through inspection must cover anchorage review and identification ofpotential "deficiencies" and ."outliers." "Deficiency" in this context meansequipment, components, and their anchorages/supports that are identified asobviously inadequate by the A-46 criteria during.plant-specific walk-throughreviews-and confirmed as inadequate by further engineering studies. "Outlier"in this context means equipment items that-are subject to the caveats and ex-clusions defined in this generic letter, or that are otherwise not covered bythe experience data. The treatment of deficiencies is further described inparagraphs 4 and S below. The walk-through inspection must cover the following:(1) For equipment within scope, verify equipment anchorage (including requiredcable trays, tanks, and heat exchangers) using the guidance provided inparagraph 3 below, and identify potential deficiencies. Utilities partici-pating in a generic implementation-may use the walk-through procedures,.;.being developed by SQUG/EPRI when these are approved by SSRAP and NRC.*IlGrade level" is the top of the ground surrounding the building.3Enclosure
\-d I(2) For equipment belonging to the initial eight types in the data base, ident-ify data base exclusions and caveats (outliers) from the guidance providedin paragraph 6 below.(3) For equipment types that exist in the data base plants but that are notincluded in the eight types in the data base, the guidelines provided inparagraph 7 below and the guidelines being developed by SQUG (to be approvedby SSRAP and NRC prior to implementation) must be used for identificationand review of "outliers" and "caveats" during the walk-through inspectionfor this equipment.The licensee must specify all equipment items that are required to functionduring the period of strong shaking. The licensee must demonstrate the oper-ability of these items by means other than comparison with the experience database; otherwise, the licensee must determine that any change of state will' notcompromise plant safety. The period of strong shaking is defined to'be thefirst 30 seconds of the seismic event and should be considered in conjunctionwith the loss of offsite power. On the basis of the seismic experience datagathered to date, the only concern remaining on equipment functional capabilityis the concern regarding relays. Contactors and switches are considered asrelays in this context. In addition, mercury switches are known to malfunctionduring testing and should be replaced by other types of qualified switches.Unless the test data being collected by the Electric Power Research Institute(EPRI) and the NRC Office of Research (RES) reveal otherwise, certain types ofrelays are the only equipment whose functional capability will need to beverified.The essential plant functions that are required to achieve and maintain hotshutdown during and after an SSE must be identified. The associated systemsand electrical circuits required to provide these functions-must then beidentified. Next, these functions must be evaluated and the essential relaysmust be identified. Essential relays are relays that must remain functionalwithout chatter during an SSE.These essential relays must be qualified by test, in a manner consistent withcurrent licensing requirements (Section 3.10 of the Standard Review Plan(NUREG-0800), NRC Regulatory Guide 1.100/IEEE Standard 344-1975), verified bycomparison with the test data base beihg developed by EPRI/RES, or replaced byrelays qualified to current licensing requirements. As an alternative, theredesign of circuitry, strengthening of relay supports/cabinets to reduce seis-mic demand, or relocation of relays to reduce demand can be used.The licensee must identify all relays that could potentially change state duringan SSE as a result of contact chatter and preclude use of equipment neededafter the SSE to place the plant in safe shutdown. The redesign of circuitry,strengthening of relay supports/cabinets to reduce demand, or relocation ofrelays to reduce demand can also be used. As an alternative, the licensee mayshow that chattering or change of state of the relays does not affect systemperformance or preclude subsequent equipment or system functions. In addition,credit can be taken for timely operator action to reset the relays in case changeof state occurs during an SSE, provided detailed relay resetting procedures aredeveloped and there is sufficient time for resetting the relays.4Enclosure For components included in the data base by type but outside the limits of ex-perience data or test data , or of a type not included in the data base, as ageneral guideline the seismic verification can be deferred until additional testdata is-developed, endorsed by SSRAP, and approved by the NRC staff, providedthat the seismic verification is completed no later than about 36 months fromthe date of issuance of the USI A-46 final resolution. Actual schedule dateswill be based on negotiations with the generic group or with individual utili-ties. The proper integration of the proposed work scope into each plant'sschedule for plant modifications will be considered.If a utility replaces components for any reason, each replacement (assembly,subassembly, device) must be verified for seismic adequacy either by using A-46criteria and methods or, as an option, qualifying by current licensing criteria.This provision also applies to future modification or'replacements. "Component"in this context means equipment and assemblies (including anchorages and sup-ports)--such as pumps and motor control centers--and subassemblies and devices--such as motors and relays that are part of assemblies.3. Verification of AnchorageTo verify acceptable seismic performance, adequate engineered anchorage must beprovided. There are numerous examples of equipment sliding or overturning underseismic loading because anchorage was absent or inadequate. Inadequate anchoragecan include short, loose, weak, or poorly installed bolts or expansion anchors;inadequate torque on bolts; and improper welding or bending of sheet metal framesat anchors. Torque on bolts can normally be ensured by a preventive maintenanceand inspection program.In general, checking of equipment anchorages requires the estimation of equipmentweight-and its approximate center of gravity. Also, one must either estimatethe fundamental frequency of the equipment to obtain the spectral accelerationat this frequency or else use the highest spectral acceleration for all fre-quencies. When horizontal floor spectra exist, these spectra may be used toobtain the equipment spectral acceleration. Alternatively, for equipment mountedless than about 40 feet above grade, one-and-a-half times the free-field hori-zontal design ground spectrum may be used to conservatively estimate the equip-ment spectral acceleration. For equipment mounted more than about 40 feet abovegrade, floor spectra must be used. This restriction may be modified if addi-tional data become available to justify raising the 40-foot-limit.Equipment anchorage must not only be strong enough to resist the anticipatedforces-but must also be stiff enough to prevent excessive movement of the equip-ment and potential resonant response with the supporting structure. The reviewof anchorages should include consideration of both strength and stiffness ofthe anchorage and its component parts.Additional discussions on seismic motion bounds and equipment supports andanchorage for each of the original eight classes of equipment in the experiencedata base are included in paragraph 6 below. This guidance supplements thegeneral guidance above.,During the walk-through inspection, anchors and supports of equipment withinthe scope'of review will be carefully inspected. The detailed guidance devel-oped is the preferred method for review of anchorages. The detailed guidancehas been developed jointly by SQUG and EPRI. It was approved by SSRAP and is5Enclosure being reviewed by the NRC staff. It will be approved by the NRC staff beforeimplementation. If the adequacy of supports and anchors cannot be determinedby inspection, an engineering review of the anchorage or support will be con-ducted. This engineering review will include a review of design calculationsor the performance of new calculations and/or verification of fundamentalfrequency of equipment to ensure adequate restraint and stiffness. Physicalmodifications may be necessary if engineering review determined the anchorageor support to be inadequate.4. Generic ResolutionThe NRC will endorse and encourage a generic resolution of USI A-46 providedthe guidelines presented below are followed:(1) All member utilities of the SQUG would be eligible to participate.(2) The generic group must be responsible (a) for developing procedures toidentify relays to be evaluated, (b) for defining functionality require-ments, and (c) for developing evaluation procedures for relays. This pro-cedure must be reviewed and endorsed by SSRAP and the NRC staff.(3) The generic group must submit to the NRC a generic schedule for the de-velopment of implementation procedures and for workshops/training seminarsfor participating utilities. A pilot walk-through must be conducted on afew selected plants to test the procedure. Afterwards, the generic groupmust hold workshops/training seminars for participating utilities to ensureuniformity in approach. Each individual utility must submit an implementa-tion schedule to the NRC within 60 days of receipt of the A-46 genericletter. Individual utilities must then perform the plant-specific imple-mentation reviews.(4) Each utility must submit to the NRC an inspection report that must include:certification of completion of the review, identification of deficienciesand outliers, justification for continued operation (JCO) for identifieddeficiencies if these deficiencies are not corrected within 30 days, mod-ifications and replacements of equipment/anchorages (and supports) made asa result of the reviews, and proposed schedule for future modificationsand replacements.The objective of the requirement to submit a JCO is to provide assurancethat the plant can continue to be operated without endangering the healthand safety of the public during the time required to correct the identifieddeficiency.The JCO may consider arguments such as imposition of administrative controlsor limiting conditions for operation (LCOs) or consideration of the impor-tance of the safety function involved and/or identification of alternatemeans to perform that function.(5) Consultants to the generic group must perform audits of plant-specificreviews. All plants must be audited. The NRC staff will participate inplant audits on a selective basis. The generic group must submit a reportof audits performed and results of these audits to the NRC. This reportcovers all participating utilities, and must also include the results ofany reviews and/or audits performed by the SSRAP.6Enclosure
(6) The SSRAP and .the NRC staff must perform a limited review of the genericgroup audit process to evaluate effectiveness.(7) Final approval of the implementation will be made by the NRC in the formof a plant-specific Safety Evaluation Report for each affected plant afterNRC receives a final report from the utility..involved certifying completionof implementation, reviews and equipment/anchorage modifications andreplacements.'(8) The generic group must provide for the'0ontinuation of the SSRAP as anindependent review body. The SSRAP must-.be consulted during the. develop-ment of the'.generif program and walk-through procedure, and must audit theimplementation.(9) NRC,,staff.members must be invited to participate in all,meetings.betweenthe generic group and the SSRAP.5. Provisions for Resolution for Individual UtilitiesThegeneric resolution described in-paragraph 4 above, Generic Resolution, isthe method preferred by the NRC for the resolqtion of A-46. This paragraphoffers provisions for resolution of A-46 for individual utilities not partici-pating in the.generic .group.Each utility must develop a detailed review procedure that must be submitted tothe NRC staff.for'review. This procedure.must'reflect the guidance given inparagraph.2 above. The data'and procedures.developed by the SQUG will not, ingeneral, be available to'non-participating utilities. ' All information that hasbeen made publicly available by SQUG or the staff can be 'used.Each utility must perform plant-specific verification reviews according to guid-ance in paragraphs.2 and must also maintain an auditable record of implementationof USX A-'46.Within 60 days of receiptof the A-46 generic. letter', each utility must submitto the NRC.a schedule for implementation of'the A-46 requirements. 'Utilitieswho may not'have access to SQUG imolementation procedures or data base may havedifficulty in establishing implementation schedules within 60'days. For theseutilities the NRC will negotiate time extensions on a case by case basis. Theutility must submit an inspection report to the NRC after the plant-specificwalk-through inspection. It should consist of the following:(1) Certification of completion of the walk-through inspection and a descriptionof the procedures used.(2) A list of the equipment included in the review scope. Equipment requiredto function during the strong shaking period should be identified.(3) Identified deficienci.es. .(4) Identified outliers.(5) Modifications and replacements of'equipment/anchorages (and supports) madeas a result of the inspection.7Enclosure
(6) The proposed schedule for future modifications and replacements.(7) A JCO for identified deficiencies if these deficiencies are not correctedwithin 30 days.Following the completion of implementation reviews and all necessary modifica-tions and replacements of equipment/anchorages, the utility must'submit a finalreport to the NRC. A description of the procedures used for the implementationreviews and the modifications and replacements must be included.The NRC will review the inspection procedure, inspection report, and the finalreport and will audit all plant-specific reviews before granting final NRCapproval. The final N~rapproval will be in the form of plant-specific SERS.6. Guidance on Use of Seismic Experience Data for the Eight Equipment Typesin the Experience Data Saw(1) Seismic Motion BoundsTo compare the potential performance of equipment at'a given nuclear power plantwith the actual performance of similar equipment in the data base plants inrecorded earthquakes, SSRAP has developed seismic motion bounding spectra tofacilitate comparison. The purpose of these bounding spectra is to compare thepotential seismic exposure of equipment in a nuclear power plant with the esti-mated ground motion that similar equipment actually resisted'in earthquakesdescribed in the data base. For convenience, the bounding spectra are expressedin terms of ground response at the nuclear 'site'rather than floor response orequipment response. These bounding spectra represent approximately two-thirdsof the free-field ground motion to which the data base equipment was actuallyexposed.Three different seismic motion'bounds (types A, B, and C) are used. Differentbounding spectra were developed, not to infer different ruggedness of equipment,but to represent the actual exposure of significant numbers of each class ofequipment within the data base to ground motion. These bounds are defined interms of the 5% damped horizontal ground response spectra shown in Figure A-1.The seismic motion bounds may be used for the equipment class as defined below.Equipment Class BoundMotor control centersLow-voltage (480-V) switchgear Type BMetal-clad (2.4 to 4-kV) switchgearUnit substation transformers*Guidance in this paragraph is based on the SSRAP report dated January 1985.The SQUG is in the process of expanding the data base to include more recentearthquake experience and 20 classes of equipment which cover all the equip-ment needed for plant hot shutdown. The SSRAP report also is being revisedaccordingly. The final guidance in the SSRAP report may differ from thatmentioned here. The revised SSRAP report should be followed for implementa-tion guidance.8Enclosure Equipment Class BoundMotor-operated valveswith large eccentric-operator- Type Clengths-to-pipe-diameterratiosMotor-operated valves (exclusive ofthose with large eccentric-operator- -lengths-to-pipe-diameter ratios)Air-operated valves Type AHorizontal pumps and their motorsVertical pumps and their motorsThese spectrum bounds are intended for comparison with the 5% damped designhorizontal ground response spectrum at a given nuclear power plant. In otherwords, if the horizontal ground response spectrum for the nuclear plant siteis less than a bounding spectrum at the approximate frequency of vibration ofthe equipm'ent and at all greater frequencies (also referred to as the frequencyranrge'of interest), then the equipment class associated with that spectrum isco64idered to be included within the scope of this method. Alternately, onemay compare 1.5 times these spectra with a given 5% damped horizontal floorspectrum' in the nuclear plant.The comparison of these seismic bounds with the design horizontal ground responsespectrum is judged to be acceptable for equipment mounted less than about 40 feet*above grade (the top of the ground surrounding the building) and for moderatelystiff structures. For equipment mounted more than about 40 feet above grade,comparisons of 1.5 times these spectra with the horizontal floor spectrum isnecessary. In all cases such a comparison with floor spectra is also acceptable.The vertical component will not be any more significant relative to the horizon-tal components for nuclear plants than it was for the data base plants. There-fore, it was decided that seismic bounds could be defined purely in terms ofhorizontal motion levels.The criteria are met so long as the 5% damped horizontal design spectrum liesbelow the appropriate bounding spectrum at frequencies greater than or equalto the fundamental frequency range of the equipment. This estimate can be madejudgmentally by experienced engineers without the need for analysis or testing.The recommendation that the seismic bounding spectrum can be compared with thehorizontal design ground response spectrum for equipment mounted less than about40 feet above grade is based upon various judgments concerning how structuresrespond in earthquakes. However, this 40-foot above grade criterion must beapplied with some judgment because some structures may respond in a differentmanner.(2) Motor Control CentersMotor control centers contain motor starters (contactors) and disconnectswitches. They also provide over-current relays to protect the system from*In most cases where numerical values are given in this section they should beconsidered as either "approximate" or "about," and a tolerance about thestated value is implied.9Enclosure v-overheating. In addition, some units will contain small transformers and dis-tribution panels for lighting and 120 V utility service.Motor control centers of the 600-V class (actual voltage is 480-V) are con-sidered. The general configuration of the cabinets must be similar to thosespecified in the Standards of the National Electrical Manufacturers Association(NEMA). This requirement is imposed to preclude unusual designs not covered inthe data base. Cabinets that are configured similarly to NEMA standards willperform well if they are properly anchored. Cabinet dimensions and materialgauges need not exactly match NEMA standards.On the basis of a review of the data base and anticipated variations in. condi-tions, it appears that the motor control centers are sufficiently rugged tosurvive a seismic event and remain operational thereafter provided the following.conditions exist in the nuclear facility:(a) The spectrum for the nuclear facility is less than the type B bounding.spectrum described in Figure A.1 for frequencies above theestimated funda-mental frequency of the cabinet, and'the motor control center is located'less than 40 feet above exterior grade and has'stiff anchorage, as discussedbelow. If the motor control center is located higher-than-40 feet aboveexterior grade or does not have stiff anchorage, the floor spectrum shallbe compared to 1.5 times the type B bounding spectrum. In all cases acomparison with floor spectra is also acceptable.(b) The cabinets have stiff engineered anchorage. Both the strength and stiff-ness of the anchorage and its component parts must be considered. Stiffnesscan be evaluated by engineering judgment based'on the cabinet.constructionand the location and type of anchorage, giving special attention to thepotential flexibility between the tiedown anchorage and the walls of thecabinet. One concern is with the potential flexibility associated withbending of a sheet metal flange between the anchor and the cabinet wall.Stiffly anchored cabinets will have a fundamental frequency greater thanabout 8 Hz under significant shaking.The intent of this recommendation is to prevent excessive movement of thecabinet and to ensure that under earthquake excitations the natural fre-quency of the installed: cabinet will not be in resonance with both thefrequency content of the earthquake and the fundamental frequency of thestructure, thereby allowing comparison of the ground response spectrawith the type B bounding spectrum.(c) Cabinets with sufficiently strong anchorage that do not have the stiffanchorage as recommended above are still considered in the data base;however, the floor response spectrum must be compared to 1.5 times thetype B bounding spectrum.(d) Cutouts in the cabinet sheathing are less than about 6 inches wide and12 inches high including side sheathing between multi-bay cabinets.(e) All internal subassemblies are securely attached to the motor controlcabinets that contain them.10Enclosure
1.2-.I- O 6% DAMPINGCO.86, YPE A0.26;_ -.020 2 0.2 1 2 4 280FREQUENCY. (Hz)Figure A.1' Seismic motion bounding spectra, horizontal ground motion11Enclosure (f) Adjacent sections of multi-bay cabinet assemblies are bolted together.(g) Equipment and their enclosures mounted externally to motor control centercabinets and supported by them have a total weight of less than 100 pounds.Functional capability (that is, inadvertent change of state or failure to changestate on command of relays during an earthquake) is not considered here. Func-tional capability must be established by other means. The structural integrityof relays contained in the motor control centers and their ability to functionproperly after earthquakes, as defined in Figure A.1, has been demonstrated.(3) Low-Voltage SwitchgearLow-voltage switchgear consists of low voltage (600 V or less) distributionbusses, circuit breakers, fuses, and disconnect switches.Low-voltage switchgear of the 600-V class (actual voltage is 480-V) is con-sidered. The general configuration of cabinets must be similar to tho~sespec1-fied in Standard C37.20 of the American National Standard Institute (ANSI).' Thisrequirement is imposed to preclude unusual designs not covered in the data base.Cabinets that are configured similarly to those defined in the ANSI standardswill perform well if they are properly anchored., Cabinet dimensioniarid materialgauge need not exactly match the ANSI standard. .All the conclusions, limitations, and bounding -spectra for motor control centersare applicable to low-voltage switchgear. -(4) Metal-Clad Switchgear -Metal-clad switchgear consists primarily oftciitcuit breakers and associatedrelays (such as over-current relays or ground fault protection relays), inter-locks, and other devices to protect the equipinentWthat it services.Metal-clad switchgear of 2.4 kV and,4.16 kV is considered. The general config-uration of cabinets must be similar to those specified in ANSI C3t.20j' Thisrequirement is imposed to preclude unusual designs not covered in the data base.Cabinets that are configured similarly to. those specified in the ANSI standardswill perform well if they are properly anchored. Cabinet dimensions and materialgauges need not exactly match ANSI standards.:All the conclusions, limitations, and bounding spectra for motor control centersare applicable to metal-clad switchgear, except that the cutouts in the cabinetsheathing shall be less than about 12 inches by 12 inches.(5) Motor-Operated ValvesMotor-operated valves consist of an electric motor and gear box cantileveredfrom the valve body by a yoke and interconnected by a drive shaft. The motorand gear box serve as an actuator to operate the valve.On the basis of a review of the data base and anticipated variations in condi-tions, it appears that motor-operated valves are sufficiently rugged to survive12Enclosure a seismic event and remain operational thereafter provided the following condi-tions exist in the nuclear facility:(a) The spectra for the nuclear facilityspectrum described in Figure A.1 formental frequency of the piping-valveare less than the appropriate boundingfrequencies above the estimated funda-system.(b) The valve is located less than 40 feet above exterior grade. If the valveis located higher than 40 feet above exterior grade, the floor spectrashall be compared with 1.5 times the appropriate bounding spectrum.(c) The valve body and yoke construction is not of cast iron.(d) The valve is mounted on a pipe at least 2 inches in diameter.-(e) The actuator is supported by the pipe and not independently braced to orsupported by the structure unless the pipe is also braced immediatelyadjacent to the valve to a common structure.The following limitations on operator weight and eccentric length relative topipe diameter are derived from the data base for motor-operated valves thatwas provided by SQUG.*(a) A type A bounding spectrum shall be used for the following cases: (seeFigure A.2):Valves mounted on 12-inch diameter or largerdistance from the pipe centerline to the topthe approximate actuator weight is less thanpipes with a 60-inch or smallerof the motor actuator, and400 pounds.Valves mounted on 24-inch diameter or larger pipes with a 100-inch orsmaller distance from the pipe centerline to the top of the motor actuator,and the approximate actuator weight is less than 300 pounds.(b) A type C bounding spectrum shall be used for the following cases: (seeFigure A.3):Valves mounted on a pipe diameter of at least 2 inches but less than6 inches, with a 30-inch or smaller distance from the pipe centerline tothe top of the motor actuator, and the approximate actuator weight is lessthan 100 pounds.Valves mounted on a pipe diameter of at least 6 inches but less than8 inches, with a 40-inch or smaller distance from the pipe centerline tothe top of the motor actuator, and the approximate actuator weight is lessthan 300 pounds.*The data base contains relatively few heavy operators and small pipe diameterssubjected to severe ground shaking. These limitations could be less restrictiveif more motor-operated valves had been located and documented in the areas ofhigher shaking. Additional data, either from other earthquake experience orseismic qualification tests, could expand the scope of these recommendations.13Enclosure aIwe%0*8100RCI-NCwC.CL00.0I-F0012 24PIPE DIAMETER (inches)I*APPROXIMATE MAXIMUM OPERATOR WEIGHTFigure A.2 Motor-operated valves for which type A spectrum is to be used14Enclosure J ,. -1 Q OUTSIDE EXPERIENCE DATAA Q WITHIN EXPERIENCE DATA¶00 t E. .TOC 4004300 -t440PIPE DIAMETER (inches)*j I ...*arPROXNAtTE, MAMMUM OPERATOR WEIGHTF A M vae f which .t..4*, ~ *~ .4.$: ."v *.Figure A.3 Motor-operated valves for which type C spectrum is to be used15Enclosure Valves mounted on a pipe diameter.of at least 8 inches but less than10 inches, with a 50-inch or smaller distance from the pipe centerline tothe top of the motor actuator, and the approximate actuator weight is lessthan 400 pounds.Valves mounted on a pipe diameter of at least 10 inches with a 70-inch orsmaller distance from the centerline of the pipe to the top of the motoractuator, and the approximate actuator weight is less than 640 pounds; orthe weight is more than 300 pounds for cases where the distance from thecenterline of the pipe to the top of the motor actuator is not greaterthan 100 inches.For motor-operated valves not complying with the above limitations, the seismicruggedness for ground motion not exceeding the type A bounding spectrum may bedemonstrated by static tests. In these tests, a static .forte equal to threetimes the approximate operator weight shall be applied npon-concurrently in eachof the three orthogonal principal axes of the yoke. Such:tests should includea demonstration of operability following the application of the static load.The limitations other than those related to the operator weight and distancefrom the top of the operator to the centerline of the pipe, given above shallremain in effect.(6) Unit Substation TransformersUnit substation transformers convert the distribution voltage to low voltage.In this discussion, unit substation transformers that convert 2.4-kV or 4.16-kYdistribution voltages to 480 V are considered.On the basis of a review of the data base and anticipated variations, it appearsthat unit substation transformers are'sufficiently rugged to survive a seismicevent and remain operational thereafter provided the following.conditions existin the nuclear facility:(a) The spectrum for the nuclear facility is less than the type B boundingspectrum described in Figure A.1 for frequencies'abpve the estimated funda-mental frequency of this equipment, and the unit substation tiransformer islocated less than 40 feet above exterior grade. If the unit substationtransformer is located higher than 40 feet above exterior grade, the floorspectrum shall be compared with 1.5 times the bounding spectrum. In allcases a comparison with floor spectra is also acceptable.(b) Both unit substation transformer enclbsures and the transformer itself musthave engineered anchorage.The functional capability of properly anchored unit substation transformers during andafter earthquakes, as defined above, has been demonstrated. ^(7) Air-Operated ValvesAir-operated valves consist of a valve (controlled by a solenoid valve) operatedby a rod actuated by air pressure against a diaphragm attached to the rod. Theactuator is supported by the valve body through a cantilevered yoke.16Enclosure On the basis of a review of the data base and anticipated variations in condi-tions, it appears that air-operated valves are sufficiently rugged to survive aseismic event and remain operational thereafter provided the following conditionsexist in the nuclear facility:(a) The ground motion spectra for the nuclear facility are less than the type Abounding spectrum for frequencies above the estimated fundamental frequencyof the piping-valve system.(b) The valve body is not of cast iron.(c) The valve is mounted on a pipe of 1-inch diameter or greater.(d) If the valve is mounted on a pipe less than 4 inches in diameter, the dis-tance from the centerline of the pipe to the top of the operator shall notexceed 45 inches. If the valve is mounted on a pipe 4 inches in diameteror larger, the distance from the centerline of the pipe to the top of theoperator shall not exceed 60 inches (see Figure A.4).(e) The actuator and yoke are supported by the pipe, and neither is indepen-dently braced to the structure or supported by the structure unless the pipeis also braced immediately adjacent to the valve to a common structure.The air supply line is not included in this assessment.For air-operated valves not complying with the above limitations, the seismicruggedness for ground motion not exceeding the type A bounding spectrum may bedemonstrated by static tests. In these tests, a static force equal to threetimes the approximate operator weight shall be applied non-concurrently in eachof the three orthogonal principal axes of the yoke. Such tests should includedemonstration of operability following the application of the static load. Thelimitations other than'those related to the distance of the top of the operatorto the centerline of the pipe given above shall remain in effect.(8) Horizontal and' Vertical PumpsHorizontal pumps in their entirety and vertical pumps above their flange arerelatively stiff and very rugged devices as a result of their inherent designand operating requirements. Motors for these pumps are also included. Subjectto the limitations set forth below, all pumps meet the criteria for the type Abounding spectrum.For horizontal pumps, the driver (electric motor, turbine, etc.) and pump mustbe rigidly connected through their bases to prevent damaging relative motion.Of concern are intermediate flexible bases, which must be evaluated separately.Thrust restraint of the shaft must also be ensured in both axial directions.The data base covers pumps up to 2500 hp; however, the conclusions appear to beequally valid for horizontal pumps of greater horsepower.For vertical Rumps, the data base has many entries up to 700 hp and several upto 600 h p. However, vertical pumps, above the flange, of any size at nuclearplants. appear to be sufficiently ruggeU to meet the type A bounding spectrum.17Enclosure
0%-I0IC.LU0IL09-'ILILiIp1 Io OUTSIDE EXPERIENCE DATAo WITHIN EXPERIENCE DATAvL~~60I00 **.............*S.:;:1:1I .1 .j,F. r. .g.-..:.'X 'f" , .OXIM"-aM ---MX I..:.F.:!.451-0..O*\~<* sY~' :M:x::&:....... .... x N9y .. .xvx m .0%. * ..x:.:/ .* %. .....,~s6N.\. 00. ' .. O ...V. < ... ......y.......*: vr >.. ........ .....0 $~'.4~t. *~.*'0%9-i0-... .... ..-1:.::-:-.-:.,-,;,......,...-...-ff1I4VPIPE DIAMETER (Inches)Figure A.4 Air-operated valves for which type A spectrum is to be usedisEnclosure
- .*The variety of vertical pump configurations and shaft lengths, below the flange,and the relatively small number of data base points in several categories pre-clude the use of the data base to screen all vertical pumps. Vertical turbinepumps (deep well submerged pumps with cantilevered casings up to 20 feet inlength and with bottom bearing support of the shaft to the casing) are wellenough represented to meet the bounding criteria below the flange as well.Either individual analysis or use of another method should be considered as ameans of evaluating other vertical pumps below the flange. The chief concerns* would be damage to bearings as a result of excessive loads, damage to the im-peller as a result of excessive displacement, and damage as a result of inter-floor displacement on multi-floor supported pumps.7. Guidance on Review of Equipment that Exists in the Experience Data BasePlants but that Is Not Included in the Eight Types in the Data BaseOn the basis of the above experience, reviews conducted by the staff in theSEP Program and licensing activities (SQRT audits), and the observation of thebehavior of equipment beyond the original eight classes found in the data baseplants, the staff concludes that the seismic adequacy of equipment other thanthe eight types can be achieved by (1) anchorage verification; (2) a carefulreview of caveats, outliers, and exclusions observed; and (3) documentation bySQUG of the basis for seismic adequacy of each equipment type.The SQUG is in the process of broadening the data base to include more recentearthquake experience (notably the 1985 earthquakes of Chile and Mexico). Theequipment covered by the experience data base will be expanded from the originaleight to twenty which will encompass all equipment needed for plant hot shutdown.The SSRAP report is also being revised accordingly. The guidance in the finalrevised SSRAP report may differ from that mentioned in the January 1985 SSRAPreport. The revised SSRAP report should be followed for implementation guidance.For individual utilities not participating in the generic group, the detailedprocedures used to review the seismic adequacy of all equipment should be sub-mitted to the NRC for review. Items such as equipment caveats and exclusions,bounding spectra to be used, and the like should be included in the submittal.19Enclosure
"., -T-_) ILIST OF RECENTLY ISSUED GENERIC LETTERSGenericLetter No.Date ofSubject IssuanceIssued ToGL 87-01GL 86-17GL 86-16GL 86-15GL B6-14GL 86-13PUBLIC AVAILABILITY OF THE NRCOPERATOR LICENSING EXAMINATIONQUESTION BANKAVAILABILITY OF NUREG-U169,"TECHNICAL FINDINGS RELATED TOGENERIC ISSUE C-BBWR MSIC LEAKAGE AND LEAKAGECONTROL SYSTEMWESTINGHOUSE ECCS EVALUATIONMODELSINFORMATION RELATING TOCOMPLIANCE WITH 10 CFR 50.49,"EQ OF ELECTRICAL EQUIPMENTIMPORTANT TO SAFETY"OPERATOR LICENSINGEXAMINATIONSPOTENTIAL INCONSISTENCYBETWEEN PLANT SAFETY ANALYSESAND TECHNICAL SPECIFICATIONS01/08/8710/17/8610/22/8609/22/860/20/8607/23/86ALL POWERREACTORLICENSEES ANDAPPLICANTS FORAN OPERATINGLICENSEALL LICENSEESOF BOILINGWATER REACTORSALLPRESSURIZEDWATER REACTORAPPLICANTS ANDLICENSEESALL LICENSEESAND HOLDERS OFAN APPLICATIONFOR ANOPERATINGLICENSEALL POWERREACTORLICENSEES ANDAPPLICANTSALL POWERREACTORLICENSEES WITHCE AND B&WPRESSURIZEDWATER REACTORSGL 86-12 CRITERIA FOR UNIQUE
PURPOSE
EXEMPTION FROM CONVERSION FROMTHE USE OF HEU FUEL07/03/e6ALL NON-POWERREACTORLICENSEESAUTHORIZED TOUSE HEU FUELGL 86-11GL 86-10DISTRIBUTION OF PRODUCTSIRRADIATED IN RESEARCHREACTORSIMPLEMENTATION OF FIREPROTECTION REQUIREMENTS06/25/8604/28/B6ALL NON-POWERREACTORLICENSEESALL POWERREACTORLICENSEES ANDAPPLICANTSGL 86-09 TECHNICAL RESOLUTION OFGENERIC ISSUE NO. B-59(N-I) LOOP OPERATION IN BWRSAND PWRS03/31/86ALL BWR ANDPWR LICENSEESAND APPLICANTS