NRC Generic Letter 1987-02
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X.'$Jt Rt44 UNITED STATES
NUCLEAR REGULATORY COMMISSION
- WASHINGTON. D. C, 20555 FEB 13 tV
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TO: - All Holders of Operating Licenses Not Reviewed to Current Licensing Criteria on Seismic Qualification of Equipment GENTLEMEN:
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 Qualification of Equipment in Qperating Plants," the NRC has concluded that the seismic ade- quacy of certain equipment in operatinginuclear power"'plants must be reviewed against 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 require extensive, and probably impracticable, modification of these facilities. An alternative resolution of this problem is set out in the enclosure to this letter. This approach makes use of earthquake experience data supplemented by test results to verify the seismic capability of equipment below specified '
earthquake motion bounds. In the staff's judgment, this approach is the most reasonable and cost-effective means of ensuring that the purpose of General Design Criterion 2 (10 CFR Part 50 Appendix A) is met for these plants.
Because affected plants are being asked to carry out this evaluation against criteria not used to establish the design basis of the facility, this resolution is a backfit under 10 CFR 50.109. The backfit analysis and findings may be found in the Regulatory Analysis (Reference 2) at pp. 31.
Seismic verification may be accomplished generically, as described in the enclosure. Utilities participating in a generic program should so state in their reply to this letter, identifying the utility group and the schedule established for completion of the effort. The implementation schedule will be negotiated with utility groups or individual utilities in accordance with the NRC policy on integrated schedules for plant modifications. See Generic Letter 83-20, May 9, 1983. Utilities not participating in a generic review may be allowed some additional time to complete the review.
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We therefore request*that you provide within 60 days of receipt of this letter a schedule for implementation of the seismic verification program at your facility.
Sincerely, arold R. Denton, Direct e Office of Nuclear Reactor Regulation Enclosure:
Seismic Adequacy Verification Procedure References: (1) NUREG-1030, "Seismic Qualification of Equipment in Operating Nuclear Power Plants (USI A-46)," February 1987
(2) NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety Issue A-46, Seismic Qualification of Equipment in Operating Plants," February 1987 C B X Y90_
- This request is covered by Office of Management and Budget Clearance Number 3150-0011 which expires September 30, 1989. Comments on burden and duplication may be directed to the Office of Management and Budget, Room 3208, New Executive Office Building, Washington, DC 20503. -
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ENCLOSURE
SEISMIC ADEQUACY VERIFICATION PROCEDURE
The proposed procedure for verifying seismic adequacy of equipment is addressed in the following paragraphs. Each licensee will be required to perform the verification steps and submit a report to the NRC including an affidavit that the verification has been completed and all equipment within the scope defined below has been found to be acceptable. A generic resolution will be accepted in lieu of a plant-specific verification review subject to the guidance presented herein.
1. Scope of Seismic Adequacy Review Each licensee will determine the systems, subsystems, components, instrumenta- tion, and controls required during and following a design-basis seismic event using the following assumptions:
(1) The' seismic event does not cause a loss-of-coolant accident (LOCA), a steam-line-break accident (SLBA), or a high-energy-line-break (HELB), and a LOCA, SLBA, or HELB does not occur simultaneously with or during a seis- mic event. However, the effects of transients that may result from ground shaking 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 condition following 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 not included except-that those tanks and heat exchangers that are required to achieve and maintain safe shutdown must be reviewed for adequate anchorage.
Seismic sys.tem interaction is included in the scope of review to the extent that 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 as suspended ceilings and lighting fixtures--could influence the operability of equipment within the scope of reviews. Instrument air lines and electrical and
'instrumentation cabling must be verified to have sufficient flexibility from
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the connection to equipment so that relative movement of anchor points will not jeopardize their integrity. Air lines and electrical and instrument cabl- ing are not included in the scope of review except for that portion from the equipment item to the first anchor point. The failure of masonry walls that could-affect the operability of nearby safety-related equipment is of concern.
However, this concern has been addressed by IE Bulletin 80-11, which requires that all such masonry walls be identified and re-evaluated to confirm their design adequacy under postulated loads and load combinations. This concern is,therefore, not considered as part of A-46 implementation. The required seismic interaction reviews will be based on, and consistent with, observations made in the seismic experience data base augmented by expert judgement of
1 Enclosure
SQUG/SSRAP. The review procedures will be reviewed by the NRC staff and SSRAP
prior to plant specific implementation.
For some pressurized water reactor plants, the seismic adequacy of auxiliary feedwater (AFW) systems has been verified by licensee'actions taken in response to Generic Letter 81-14, dated February 10, 1981. Review of the AFW systems may be deleted from consideration under USI A-46 if staff acceptance has been documented in a Safety Evaluation Report or if the licensee has committed to meet the requirements of the generic letter.
For the purpose of seismic adequacy verification, the following guidance is given. Each licensee must identify equipment necessary to bring the plant to a 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 />. The
72-hour period is sufficient for inspection of equipment and minor repairs, if necessary, following a safe-shutdown earthquake (SSE) or to provide additional source(s) of water for decay heat removal, if needed, to extend the time at hot shutdown.
Equipment required includes that necessary to maintain the supporting functions required for safe shutdown. For all equipment within the defined scope, the verification must closely follow the procedure outlined in paragraph 2 below.
Each licensee must show practical means of staying at hot shutdown for a minimum of 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 must show that at least one practical alternative for achieving and maintaining safe shutdown 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 necessary to monitor hot shutdown.
(4) Provide alternating current (ac) and/or direct current (dc) emergency power as needed on a plant-specific basis to meet the above three functions.
2. General Verification Procedure for Plant-Specific Review The licensee will be required to conduct a plant walk-through and visual inspec- tion of all identified equipment items necessary to perform the functions related to plant shutdown. The inspection team must consist of as a minimum,
(1) an experienced structural engineer familiar with seismic anchorage requirements
2 Enclosure
(2) an. experienced mechanical engineer familiar with plant mechanical equipment
(3) an experienced electrical.engineer familiar with plant electrical equipment Furthermore, an operations supervisor or a licensed Senior Reactor Operator must be available for consultation before and during the walk-through process.
Not all members of the inspection team are-required to participate in all parts of the walk-through; however, appropriate technical, expertise must be included for each review area, and a person with proper structural background.must always 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 verify that the appropriate data base spectra envelope the site free-field spectra at the ground surface defined for the plant. There are a number of nuclear plants whose free-field'SSE spectra are defined at the foundation level. For these plants, an estimate of the free field spectra at the ground surface must be made-for comparison with the data base bounding spectra..-The licensee must identify all equipment on the plant's equipment list that is located at an elevation higher than 40 feet above grade level.* For equipment above 40 feet, one-and-one-half times the appropriate data base bounding spectrum (defined in paragraph.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 Operating Nuclear Power Plants: The Development of a Method to Generate Generic Floor Response.Spectra," may bemused as one alternative to develop the necessary floor response spectra on a-case-specific basis. The appropriate bounding spectra for equipment belonging to the original eight types in the-data base are defined in paragraph 6 below. For equipment types that are not included in the eight types in the data base but that exist in the data base plants, and for equipment unique to nuclear plants, the appropriate bounding spectra.are defined in paragraph 7 below.
The walk-through inspection must cover anchorage review and identification of potential "deficiencies" and ."outliers." "Deficiency" in this context means equipment, components, and their anchorages/supports that are identified as obviously inadequate by the A-46 criteria during.plant-specific walk-through reviews-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 by the experience data. The treatment of deficiencies is further described in paragraphs 4 and S below. The walk-through inspection must cover the following:
(1) For equipment within scope, verify equipment anchorage (including required cable trays, tanks, and heat exchangers) using the guidance provided in paragraph 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.
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(2) For equipment belonging to the initial eight types in the data base, ident- ify data base exclusions and caveats (outliers) from the guidance provided in paragraph 6 below.
(3) For equipment types that exist in the data base plants but that are not included in the eight types in the data base, the guidelines provided in paragraph 7 below and the guidelines being developed by SQUG (to be approved by SSRAP and NRC prior to implementation) must be used for identification and review of "outliers" and "caveats" during the walk-through inspection for this equipment.
The licensee must specify all equipment items that are required to function during the period of strong shaking. The licensee must demonstrate the oper- ability of these items by means other than comparison with the experience data base; otherwise, the licensee must determine that any change of state will' not compromise plant safety. The period of strong shaking is defined to'be the first 30 seconds of the seismic event and should be considered in conjunction with the loss of offsite power. On the basis of the seismic experience data gathered to date, the only concern remaining on equipment functional capability is the concern regarding relays. Contactors and switches are considered as relays in this context. In addition, mercury switches are known to malfunction during 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 of relays are the only equipment whose functional capability will need to be verified.
The essential plant functions that are required to achieve and maintain hot shutdown during and after an SSE must be identified. The associated systems and electrical circuits required to provide these functions-must then be identified. Next, these functions must be evaluated and the essential relays must be identified. Essential relays are relays that must remain functional without chatter during an SSE.
These essential relays must be qualified by test, in a manner consistent with current licensing requirements (Section 3.10 of the Standard Review Plan (NUREG-0800), NRC Regulatory Guide 1.100/IEEE Standard 344-1975), verified by comparison with the test data base beihg developed by EPRI/RES, or replaced by relays qualified to current licensing requirements. As an alternative, the redesign 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 during an SSE as a result of contact chatter and preclude use of equipment needed after the SSE to place the plant in safe shutdown. The redesign of circuitry, strengthening of relay supports/cabinets to reduce demand, or relocation of relays to reduce demand can also be used. As an alternative, the licensee may show that chattering or change of state of the relays does not affect system performance or preclude subsequent equipment or system functions. In addition, credit can be taken for timely operator action to reset the relays in case change of state occurs during an SSE, provided detailed relay resetting procedures are developed and there is sufficient time for resetting the relays.
4 Enclosure
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 a general guideline the seismic verification can be deferred until additional test data is-developed, endorsed by SSRAP, and approved by the NRC staff, provided that the seismic verification is completed no later than about 36 months from the date of issuance of the USI A-46 final resolution. Actual schedule dates will be based on negotiations with the generic group or with individual utili- ties. The proper integration of the proposed work scope into each plant's schedule 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-46 criteria 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 Anchorage To verify acceptable seismic performance, adequate engineered anchorage must be provided. There are numerous examples of equipment sliding or overturning under seismic loading because anchorage was absent or inadequate. Inadequate anchorage can include short, loose, weak, or poorly installed bolts or expansion anchors;
inadequate torque on bolts; and improper welding or bending of sheet metal frames at anchors. Torque on bolts can normally be ensured by a preventive maintenance and inspection program.
In general, checking of equipment anchorages requires the estimation of equipment weight-and its approximate center of gravity. Also, one must either estimate the fundamental frequency of the equipment to obtain the spectral acceleration at this frequency or else use the highest spectral acceleration for all fre- quencies. When horizontal floor spectra exist, these spectra may be used to obtain the equipment spectral acceleration. Alternatively, for equipment mounted less 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 above grade, 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 anticipated forces-but must also be stiff enough to prevent excessive movement of the equip- ment and potential resonant response with the supporting structure. The review of anchorages should include consideration of both strength and stiffness of the anchorage and its component parts.
Additional discussions on seismic motion bounds and equipment supports and anchorage for each of the original eight classes of equipment in the experience data base are included in paragraph 6 below. This guidance supplements the general guidance above.,
During the walk-through inspection, anchors and supports of equipment within the scope'of review will be carefully inspected. The detailed guidance devel- oped is the preferred method for review of anchorages. The detailed guidance has been developed jointly by SQUG and EPRI. It was approved by SSRAP and is
5 Enclosure
being reviewed by the NRC staff. It will be approved by the NRC staff before implementation. If the adequacy of supports and anchors cannot be determined by inspection, an engineering review of the anchorage or support will be con- ducted. This engineering review will include a review of design calculations or the performance of new calculations and/or verification of fundamental frequency of equipment to ensure adequate restraint and stiffness. Physical modifications may be necessary if engineering review determined the anchorage or support to be inadequate.
4. Generic Resolution The NRC will endorse and encourage a generic resolution of USI A-46 provided the 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 to identify 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 seminars for participating utilities. A pilot walk-through must be conducted on a few selected plants to test the procedure. Afterwards, the generic group must hold workshops/training seminars for participating utilities to ensure uniformity in approach. Each individual utility must submit an implementa- tion schedule to the NRC within 60 days of receipt of the A-46 generic letter. 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 deficiencies and outliers, justification for continued operation (JCO) for identified deficiencies if these deficiencies are not corrected within 30 days, mod- ifications and replacements of equipment/anchorages (and supports) made as a result of the reviews, and proposed schedule for future modifications and replacements.
The objective of the requirement to submit a JCO is to provide assurance that the plant can continue to be operated without endangering the health and safety of the public during the time required to correct the identified deficiency.
The JCO may consider arguments such as imposition of administrative controls or limiting conditions for operation (LCOs) or consideration of the impor- tance of the safety function involved and/or identification of alternate means to perform that function.
(5) Consultants to the generic group must perform audits of plant-specific reviews. All plants must be audited. The NRC staff will participate in plant audits on a selective basis. The generic group must submit a report of audits performed and results of these audits to the NRC. This report covers all participating utilities, and must also include the results of any reviews and/or audits performed by the SSRAP.
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(6) The SSRAP and .the NRC staff must perform a limited review of the generic group audit process to evaluate effectiveness.
(7) Final approval of the implementation will be made by the NRC in the form of a plant-specific Safety Evaluation Report for each affected plant after NRC receives a final report from the utility..involved certifying completion of implementation, reviews and equipment/anchorage modifications and replacements.'
(8) The generic group must provide for the'0ontinuation of the SSRAP as an independent review body. The SSRAP must-.be consulted during the. develop- ment of the'.generif program and walk-through procedure, and must audit the implementation.
(9) NRC,,staff.members must be invited to participate in all,meetings.between the generic group and the SSRAP.
5. Provisions for Resolution for Individual Utilities Thegeneric resolution described in-paragraph 4 above, Generic Resolution, is the method preferred by the NRC for the resolqtion of A-46. This paragraph offers 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 to the NRC staff.for'review. This procedure.must'reflect the guidance given in paragraph.2 above. The data'and procedures.developed by the SQUG will not, in general, be available to'non-participating utilities. 'All information that has been made publicly available by SQUG or the staff can be 'used.
Each utility must perform plant-specific verification reviews according to guid- ance inparagraphs.2 and must also maintain an auditable record of implementation of USX A-'46.
Within 60 days of receiptof the A-46 generic. letter', each utility must submit to the NRC.a schedule for implementation of'the A-46 requirements. 'Utilities who may not'have access to SQUG imolementation procedures or data base may have difficulty in establishing implementation schedules within 60'days. For these utilities the NRC will negotiate time extensions on a case by case basis. The utility must submit an inspection report to the NRC after the plant-specific walk-through inspection. It should consist of the following:
(1) Certification of completion of the walk-through inspection and a description of the procedures used.
(2) A list of the equipment included in the review scope. Equipment required to 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) made as a result of the inspection.
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(6) The proposed schedule for future modifications and replacements.
(7) A JCO for identified deficiencies if these deficiencies are not corrected within 30 days.
Following the completion of implementation reviews and all necessary modifica- tions and replacements of equipment/anchorages, the utility must'submit a final report to the NRC. A description of the procedures used for the implementation reviews and the modifications and replacements must be included.
The NRC will review the inspection procedure, inspection report, and the final report and will audit all plant-specific reviews before granting final NRC
approval. 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 Types in the Experience Data Saw
(1) Seismic Motion Bounds To compare the potential performance of equipment at'a given nuclear power plant with the actual performance of similar equipment in the data base plants in recorded earthquakes, SSRAP has developed seismic motion bounding spectra to facilitate comparison. The purpose of these bounding spectra is to compare the potential seismic exposure of equipment in a nuclear power plant with the esti- mated ground motion that similar equipment actually resisted'in earthquakes described in the data base. For convenience, the bounding spectra are expressed in terms of ground response at the nuclear 'site'rather than floor response or equipment response. These bounding spectra represent approximately two-thirds of the free-field ground motion to which the data base equipment was actually exposed.
Three different seismic motion'bounds (types A, B, and C) are used. Different bounding spectra were developed, not to infer different ruggedness of equipment, but to represent the actual exposure of significant numbers of each class of equipment within the data base to ground motion. These bounds are defined in terms 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 Bound Motor control centers Low-voltage (480-V) switchgear Type B
Metal-clad (2.4 to 4-kV) switchgear Unit 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 recent earthquake experience and 20 classes of equipment which cover all the equip- ment needed for plant hot shutdown. The SSRAP report also is being revised accordingly. The final guidance in the SSRAP report may differ from that mentioned here. The revised SSRAP report should be followed for implementa- tion guidance.
8 Enclosure
Equipment Class Bound Motor-operated valves with large eccentric-operator- Type C
lengths-to-pipe-diameter ratios Motor-operated valves (exclusive of those with large eccentric-operator- -
lengths-to-pipe-diameter ratios)
Air-operated valves Type A
Horizontal pumps and their motors Vertical pumps and their motors These spectrum bounds are intended for comparison with the 5% damped design horizontal ground response spectrum at a given nuclear power plant. In other words, if the horizontal ground response spectrum for the nuclear plant site is less than a bounding spectrum at the approximate frequency of vibration of the equipm'ent and at all greater frequencies (also referred to as the frequency ranrge'of interest), then the equipment class associated with that spectrum is co64idered to be included within the scope of this method. Alternately, one may compare 1.5 times these spectra with a given 5% damped horizontal floor spectrum' in the nuclear plant.
The comparison of these seismic bounds with the design horizontal ground response spectrum 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 moderately stiff structures. For equipment mounted more than about 40 feet above grade, comparisons of 1.5 times these spectra with the horizontal floor spectrum is necessary. 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 of horizontal motion levels.
The criteria are met so long as the 5% damped horizontal design spectrum lies below the appropriate bounding spectrum at frequencies greater than or equal to the fundamental frequency range of the equipment. This estimate can be made judgmentally by experienced engineers without the need for analysis or testing.
The recommendation that the seismic bounding spectrum can be compared with the horizontal design ground response spectrum for equipment mounted less than about
40 feet above grade is based upon various judgments concerning how structures respond in earthquakes. However, this 40-foot above grade criterion must be applied with some judgment because some structures may respond in a different manner.
(2) Motor Control Centers Motor control centers contain motor starters (contactors) and disconnect switches. They also provide over-current relays to protect the system from
- In most cases where numerical values are given in this section they should be considered as either "approximate" or "about," and a tolerance about the stated value is implied.
9 Enclosure
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 those specified in the Standards of the National Electrical Manufacturers Association (NEMA). This requirement is imposed to preclude unusual designs not covered in the data base. Cabinets that are configured similarly to NEMA standards will perform well if they are properly anchored. Cabinet dimensions and material gauges 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 to survive 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 discussed below. If the motor control center is located higher-than-40 feet above exterior grade or does not have stiff anchorage, the floor spectrum shall be compared to 1.5 times the type B bounding spectrum. In all cases a comparison 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. Stiffness can be evaluated by engineering judgment based'on the cabinet.construction and the location and type of anchorage, giving special attention to the potential flexibility between the tiedown anchorage and the walls of the cabinet. One concern is with the potential flexibility associated with bending of a sheet metal flange between the anchor and the cabinet wall.
Stiffly anchored cabinets will have a fundamental frequency greater than about 8 Hz under significant shaking.
The intent of this recommendation is to prevent excessive movement of the cabinet and to ensure that under earthquake excitations the natural fre- quency of the installed: cabinet will not be in resonance with both the frequency content of the earthquake and the fundamental frequency of the structure, thereby allowing comparison of the ground response spectra with the type B bounding spectrum.
(c) Cabinets with sufficiently strong anchorage that do not have the stiff anchorage as recommended above are still considered in the data base;
however, the floor response spectrum must be compared to 1.5 times the type B bounding spectrum.
(d) Cutouts in the cabinet sheathing are less than about 6 inches wide and
12 inches high including side sheathing between multi-bay cabinets.
(e) All internal subassemblies are securely attached to the motor control cabinets that contain them.
10 Enclosure
1.2 I-O
-. 6% DAMPING
CO.8
6, YPE A
0.26;_ - .02
02 1 2 4 280
0.2 FREQUENCY. (Hz)
Figure A.1' Seismic motion bounding spectra, horizontal ground motion
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(f) Adjacent sections of multi-bay cabinet assemblies are bolted together.
(g) Equipment and their enclosures mounted externally to motor control center cabinets and supported by them have a total weight of less than 100 pounds.
Functional capability (that is, inadvertent change of state or failure to change state on command of relays during an earthquake) is not considered here. Func- tional capability must be established by other means. The structural integrity of relays contained in the motor control centers and their ability to function properly after earthquakes, as defined in Figure A.1, has been demonstrated.
(3) Low-Voltage Switchgear Low-voltage switchgear consists of low voltage (600 V or less) distribution busses, 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).' This requirement is imposed to preclude unusual designs not covered in the data base.
Cabinets that are configured similarly to those defined in the ANSI standards will perform well if they are properly anchored., Cabinet dimensioniarid material gauge need not exactly match the ANSI standard. .
All the conclusions, limitations, and bounding -spectra for motor control centers are applicable to low-voltage switchgear. -
(4) Metal-Clad Switchgear -
Metal-clad switchgear consists primarily oftciitcuit breakers and associated relays (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' This requirement is imposed to preclude unusual designs not covered in the data base.
Cabinets that are configured similarly to. those specified in the ANSI standards will perform well if they are properly anchored. Cabinet dimensions and material gauges need not exactly match ANSI standards.:
All the conclusions, limitations, and bounding spectra for motor control centers are applicable to metal-clad switchgear, except that the cutouts in the cabinet sheathing shall be less than about 12 inches by 12 inches.
(5) Motor-Operated Valves Motor-operated valves consist of an electric motor and gear box cantilevered from the valve body by a yoke and interconnected by a drive shaft. The motor and 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 survive
12 Enclosure
a seismic event and remain operational thereafter provided the following condi- tions exist in the nuclear facility:
(a) The spectra for the nuclear facility are less than the appropriate bounding spectrum described in Figure A.1 for frequencies above the estimated funda- mental frequency of the piping-valve system.
(b) The valve is located less than 40 feet above exterior grade. If the valve is located higher than 40 feet above exterior grade, the floor spectra shall 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 or supported by the structure unless the pipe is also braced immediately adjacent to the valve to a common structure.
The following limitations on operator weight and eccentric length relative to pipe diameter are derived from the data base for motor-operated valves that was provided by SQUG.*
(a) A type A bounding spectrum shall be used for the following cases: (see Figure A.2):
Valves mounted on 12-inch diameter or larger pipes with a 60-inch or smaller distance from the pipe centerline to the top of the motor actuator, and the approximate actuator weight is less than 400 pounds.
Valves mounted on 24-inch diameter or larger pipes with a 100-inch or smaller 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: (see Figure A.3):
Valves mounted on a pipe diameter of at least 2 inches but less than
6 inches, with a 30-inch or smaller distance from the pipe centerline to the top of the motor actuator, and the approximate actuator weight is less than 100 pounds.
Valves mounted on a pipe diameter of at least 6 inches but less than
8 inches, with a 40-inch or smaller distance from the pipe centerline to the top of the motor actuator, and the approximate actuator weight is less than 300 pounds.
- The data base contains relatively few heavy operators and small pipe diameters subjected to severe ground shaking. These limitations could be less restrictive if more motor-operated valves had been located and documented in the areas of higher shaking. Additional data, either from other earthquake experience or seismic qualification tests, could expand the scope of these recommendations.
13 Enclosure
a I
%0*
8100
we C
w C.
CL
R
0
0.
0
I-
I-
F NC
0
0
12 24 PIPE DIAMETER (inches) I
- APPROXIMATE MAXIMUM OPERATOR WEIGHT
Figure A.2 Motor-operated valves for which type A spectrum is to be used
14 Enclosure
J ,. -
1 Q OUTSIDE EXPERIENCE DATA
A QWITHIN EXPERIENCE DATA
¶00 t E. .
TO
C 4004
40
300 -t4
- j I . . .
. .4
~. *, *~
.4.$:
PIPE "vDIAMETER
- .
(inches)
- arPROXNAtTE,MAMMUM OPERATOR WEIGHT
F A M vae f which .t Figure A.3 Motor-operated valves for which type C spectrum is to be used
15 Enclosure
Valves mounted on a pipe diameter.of at least 8 inches but less than
10 inches, with a 50-inch or smaller distance from the pipe centerline to the top of the motor actuator, and the approximate actuator weight is less than 400 pounds.
Valves mounted on a pipe diameter of at least 10 inches with a 70-inch or smaller distance from the centerline of the pipe to the top of the motor actuator, and the approximate actuator weight is less than 640 pounds; or the weight is more than 300 pounds for cases where the distance from the centerline of the pipe to the top of the motor actuator is not greater than 100 inches.
For motor-operated valves not complying with the above limitations, the seismic ruggedness for ground motion not exceeding the type A bounding spectrum may be demonstrated by static tests. In these tests, a static .forte equal to three times the approximate operator weight shall be applied npon-concurrently in each of the three orthogonal principal axes of the yoke. Such:tests should include a demonstration of operability following the application of the static load.
The limitations other than those related to the operator weight and distance from the top of the operator to the centerline of the pipe, given above shall remain in effect.
(6) Unit Substation Transformers Unit substation transformers convert the distribution voltage to low voltage.
In this discussion, unit substation transformers that convert 2.4-kV or 4.16-kY
distribution voltages to 480 V are considered.
On the basis of a review of the data base and anticipated variations, it appears that unit substation transformers are'sufficiently rugged to survive 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'abpve the estimated funda- mental frequency of this equipment, and the unit substation tiransformer is located less than 40 feet above exterior grade. If the unit substation transformer is located higher than 40 feet above exterior grade, the floor spectrum shall be compared with 1.5 times the bounding spectrum. In all cases a comparison with floor spectra is also acceptable.
(b) Both unit substation transformer enclbsures and the transformer itself must have engineered anchorage.
The functional capability of properly anchored unit substation transformers during and after earthquakes, as defined above, has been demonstrated. ^
(7) Air-Operated Valves Air-operated valves consist of a valve (controlled by a solenoid valve) operated by a rod actuated by air pressure against a diaphragm attached to the rod. The actuator is supported by the valve body through a cantilevered yoke.
16 Enclosure
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 a seismic event and remain operational thereafter provided the following conditions exist in the nuclear facility:
(a) The ground motion spectra for the nuclear facility are less than the type A
bounding spectrum for frequencies above the estimated fundamental frequency of 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 not exceed 45 inches. If the valve is mounted on a pipe 4 inches in diameter or larger, the distance from the centerline of the pipe to the top of the operator 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 pipe is 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 seismic ruggedness for ground motion not exceeding the type A bounding spectrum may be demonstrated by static tests. In these tests, a static force equal to three times the approximate operator weight shall be applied non-concurrently in each of the three orthogonal principal axes of the yoke. Such tests should include demonstration of operability following the application of the static load. The limitations other than'those related to the distance of the top of the operator to the centerline of the pipe given above shall remain in effect.
(8) Horizontal and' Vertical Pumps Horizontal pumps in their entirety and vertical pumps above their flange are relatively stiff and very rugged devices as a result of their inherent design and operating requirements. Motors for these pumps are also included. Subject to the limitations set forth below, all pumps meet the criteria for the type A
bounding spectrum.
For horizontal pumps, the driver (electric motor, turbine, etc.) and pump must be 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 be equally valid for horizontal pumps of greater horsepower.
For vertical Rumps, the data base has many entries up to 700 hp and several up to 600 hp. However, vertical pumps, above the flange, of any size at nuclear plants. appear to be sufficiently ruggeU to meet the type A bounding spectrum.
17 Enclosure
p1 I vL~~
o OUTSIDE EXPERIENCE DATA
0% o WITHIN EXPERIENCE DATA
-I
0
IC.
60
I
LU 00 **
.............
- S.:;:1:1 -a I .1 . j,
0 'N'I'MMO
F.r. .g .-..:.
IL
0 M --- 'X OXIM" 'f" , . ::e::
X I..:.
M F.:!.
9-
451-
0..O
- \~<*
sY~' :M:x::&:.......
.... x N9y .. .
'IL
.
i xvx m x:.:/ 0%. * .. .* .....
%. ,~s IL 6N.\. 00. ..' ... O V. < ... ...... y.......
- vr>.. ........ . ....
I 0 $~'.4~t.
- ~.*'0%
9-i
0
-...
- .... .
...-1:.::-:-.-:.,-,;,......,.... -ff
1I 4 V
PIPE DIAMETER (Inches)
Figure A.4 Air-operated valves for which type A spectrum is to be used is Enclosure
- . *
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 turbine pumps (deep well submerged pumps with cantilevered casings up to 20 feet in length and with bottom bearing support of the shaft to the casing) are well enough represented to meet the bounding criteria below the flange as well.
Either individual analysis or use of another method should be considered as a means 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 Base Plants but that Is Not Included in the Eight Types in the Data Base On the basis of the above experience, reviews conducted by the staff in the SEP Program and licensing activities (SQRT audits), and the observation of the behavior of equipment beyond the original eight classes found in the data base plants, the staff concludes that the seismic adequacy of equipment other than the eight types can be achieved by (1) anchorage verification; (2) a careful review of caveats, outliers, and exclusions observed; and (3) documentation by SQUG of the basis for seismic adequacy of each equipment type.
The SQUG is in the process of broadening the data base to include more recent earthquake experience (notably the 1985 earthquakes of Chile and Mexico). The equipment covered by the experience data base will be expanded from the original eight to twenty which will encompass all equipment needed for plant hot shutdown.
The SSRAP report is also being revised accordingly. The guidance in the final revised SSRAP report may differ from that mentioned in the January 1985 SSRAP
report. The revised SSRAP report should be followed for implementation guidance.
For individual utilities not participating in the generic group, the detailed procedures 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.
19 Enclosure
"., - T-_
)I
LIST OF RECENTLY ISSUED GENERIC LETTERS
Generic Date of Letter No. Subject Issuance Issued To GL 87-01 PUBLIC AVAILABILITY OF THE NRC 01/08/87 ALL POWER
OPERATOR LICENSING EXAMINATION REACTOR
QUESTION BANK LICENSEES AND
APPLICANTS FOR
AN OPERATING
LICENSE
GL 86-17 AVAILABILITY OF NUREG-U169, 10/17/86 ALL LICENSEES
"TECHNICAL FINDINGS RELATED TO OF BOILING
GENERIC ISSUE C-B WATER REACTORS
BWR MSIC LEAKAGE AND LEAKAGE
CONTROL SYSTEM
GL 86-16 WESTINGHOUSE ECCS EVALUATION 10/22/86 ALL
MODELS PRESSURIZED
WATER REACTOR
APPLICANTS AND
LICENSEES
GL 86-15 INFORMATION RELATING TO 09/22/86 ALL LICENSEES
COMPLIANCE WITH 10 CFR 50.49, AND HOLDERS OF
"EQ OF ELECTRICAL EQUIPMENT AN APPLICATION
IMPORTANT TO SAFETY" FOR AN
OPERATING
LICENSE
GL B6-14 OPERATOR LICENSING 0/20/86 ALL POWER
EXAMINATIONS REACTOR
LICENSEES AND
APPLICANTS
GL 86-13 POTENTIAL INCONSISTENCY 07/23/86 ALL POWER
BETWEEN PLANT SAFETY ANALYSES REACTOR
AND TECHNICAL SPECIFICATIONS LICENSEES WITH
PRESSURIZED
WATER REACTORS
GL 86-12 CRITERIA FOR UNIQUE
PURPOSE
07/03/e6 ALL NON-POWER
EXEMPTION FROM CONVERSION FROM REACTOR
THE USE OF HEU FUEL LICENSEES
AUTHORIZED TO
USE HEU FUEL
GL 86-11 DISTRIBUTION OF PRODUCTS 06/25/86 ALL NON-POWER
IRRADIATED IN RESEARCH REACTOR
REACTORS LICENSEES
GL 86-10 IMPLEMENTATION OF FIRE 04/28/B6 ALL POWER
PROTECTION REQUIREMENTS REACTOR
LICENSEES AND
APPLICANTS
GL 86-09 TECHNICAL RESOLUTION OF 03/31/86 ALL BWR AND
GENERIC ISSUE NO. B-59 PWR LICENSEES
(N-I) LOOP OPERATION IN BWRS AND APPLICANTS
AND PWRS