ML20206T733
| ML20206T733 | |
| Person / Time | |
|---|---|
| Issue date: | 05/14/1999 |
| From: | Matthews D NRC (Affiliation Not Assigned) |
| To: | Pietrangelo A NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| PROJECT-689 NUDOCS 9905240244 | |
| Download: ML20206T733 (18) | |
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May 14, j999 Mr. Anthony Pi:tr:ng:lo, Director I
i Licensing l
Nuclear Energy Institute Suite 400 17761 Street, NW Washington, DC 20006-3708
SUBJECT:
STAFF POSITION ON DESIGN BASIS INFORMATION AS DEFINED IN 10 CFR 50.2 i
Dear Mr. Pietrangelo:
j This letter forwards the final draft staff position on what constitutes design basis information as defined in 10 CFR 50.2. The staff and the industry have had many meetings recently on this i
subject, and I agree with the consensus developed during these meetings that it is important to reach a clear, common understanding. We received your letter on this subject dated May 7, 1999. The enclosed staff position was developed after we considered the Nuclear Energy Institute (NEI) proposed guidance in your letter.
The enclosed position paper gives some background information, the staff's objectives in clarifying the definition of design bases, and a discussion of the role of design bases in the regulatory process. Attachments to the paper provide guidance for determining desig 1 basis information and examples. The guidance is an updated version of the draft critce provided to NEl in a letter from the staff dated January 4,1999. The examples were developed to illustrate the level of detail of design bases and to serve as a focus for a future meeting.
As discussed in our letter of April 14,1999, the staff anticipates that NEl will submit a revised guidance document by June 1,1999. The staff intends to continue to work with NEl to attempt to reach consensus on and to endorse an industry guidance document on this subject, however, the staff is prepared to develop guidance unilaterally. As we discussed, another meeting to discuss how the examples provided in this letter match up with your proposed guidance would be useful before you submit your revised guidance document. Please feel free to call me or Stewart Magruder of my staff with any questions.
Sincerely, Scott Newberry for/
9905240244 990514 David B. Matthews, Director REVOPERGNgG Division of Regulatory Improvement Programs PDR Office of Nuclear Reactor Regulation i
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May 14,1999 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 i Street, NW Washington, DC _20006-3708
SUBJECT:
STAFF POSITION ON DESIGN BASIS INFORMATION AS DEFINED IN 10 CFR 50.2
Dear Mr. Pietrangelo:
l This letter forwards the final draft staff position on what constitutes design basis information as l
defined in 10 CFR 50.2. The staff and the industry have had many meetings recently on this l
subject, and I agree with the consensus developed during these meetings that it is important to i
reach a clear, common understanding. We received your letter on this subject dated May 7, l
1999. The enclosed staff position was developed after we considered the Nuclear Energy Institute (NEI) proposed guidance in your letter.
'l The enclosed position paper gives some background information, the staff's objectives in clarifying the definition of design bases, and a discussion of the role of design bases in the regulatory process. Attachments to the paper provide guidance for determining design basis information and examples. The guidance is an updated version of the draft criteria provided to NEl in a letter from the staff dated January 4,1999. The examples were developed to illustrate J
the level of detail of design bases and to serve as a focus for a future meeting.
As discussed in our letter of April 14,1999, the staff anticipates that NEl will submit a revised guidance document by June 1,1999. The staff intends to continue to work with NEl to attempt to reach consensus on and to endorse an industry guidance document on this subject, however, the staff is prepared to develop guidance unilaterally. As we discussed, another meeting to discuss how the examples provided in this letter match up with your proposed guidance would be useful before you submit your revised guidance document. Please feel free to ca!! me or Stewart Magruder of my staff with any questions.
Sincerely, David B. Matthews, Directo Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 689
Enclosure:
As stated l
cc w/ encl: See next page 1
1
Design Basis Position Paper Introducti2D This paper clarifies the staff's position on what constitutes design basis information as defined in 10 CFR 50.2. The staff has developed guidelines for determining which information contained in an updated final safety analysis report (UFSAR) is considered to be design basis information. These guidelines are included in Attachment 1 to this paper. The staff has also developed examples of design basis information for two plant systems and one component.
These examples are included in Attachment 2.
1 Backaround l
Although the staff and the nuclear industry have always agreed that it is important to understand what constitutes the design bases of a plant, there have been disagreements about the definition in 10 CFR 50.2. In the mid-1980s, the staff conducted many system-specific engineering inspections and devdoped inspection findings that demonstrated that some licensees had not adequately naintained their design basis information as required by NRC regulation. In response to tne problems identified during the NRC inspections and those identified by licensees, mc st reactor licensees initiated design bases reconstitution programs.
These programs sought to identify missing design documentation and to selectively regenerate missing documentation.
In October 1990, the Nuclear Management and Resources Council (NUMARC) published its
" Design Bases Program Guidelines," NUMARC 90-12. The staff concluded that these guidelines provided a useful standard framework for implementing design reconstitution programs. The guidelines briefly discussed the definition of design basis information but did not focus on it.
In February 1991, the staff published NUREG-1397,"An Assessment of Design Control Practices and Design Reconstitution Programs in the Nuclear Power Industry." This report gave the results of a survey reflecting the scope and performance of several utility design i
change control programs and design document reconstitution programs. This report included a definitions section that stated that design bases include only the design constraints that are included in current licensing bases and form the bases for the staff's safety judgments.
In August 1992, the Commission published a policy statement on " Availability and Adequacy of Design Bases Information at Nuclear Power Plants." In the policy statement, the Commission j
concluded that--
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[M]aintaining current and accessible design documentation is important to ensure that (1) the plant physical and functional characteristics are maintained j
and are consistent with the design bases as required by NRC regulation, (2) l systems, structures, and components can perform their intended functions, and (3) the pl ant is operated in a manner consistent with the design bases.
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in the policy statement, the Commission also said that all power reactor licensees should assess the accessibility and adequacy of their design bases documentation and decide whether a design reconstitution program is necessary. With regard to the NUMARC guidance, the l
Commission stated that--
1 Enclosure
m Yhe guidance outlines a framework to organize and collate nuclear power plant design bases information. This information provides the rationale for the design bases consistent with the definition of design bases contained in 10 CFR 50.2.
In response to the findings relating to the regulatory burden of team inspections identified in the 1991 Regulatory impact Survey and voluntary implementation of the NUMARC guidance by licensees, the staff reduced its effort on specific, resource-intensive, design-related team inspections and followed the issue of accurate and accessible design documentation at plants principally as an element of inspection and followup of operations-related activities.
In 1996, the staff's findings during inspections and reviews began to identify broad programmatic weaknesses that resulted in design and configuration deficiencies at some plants, which could affect the operability of required equipment, raise unreviewed safety questions, or irdcate discrepancies between the plant's UFSAR and the as-built or as-modified plant or plant operating procedures. As a result of these findings, the staff issued a letter in accordance with 10 CFR 50.54(f) to all licensees requesting information to provide the NRC added confidence and assurance that the plants were operated and maintained within the design bases and any deviations were reconciled in a timely manner.
SECY-97-160, dated July 24,1997, informed the Commission of tir followup activities resulting from the staff's review of licensee responses to the $ 50.54(f) request. In this paper, the staff stated that-Based on the review of licensee responses to the 50.54(f) letter, the staff concluded that while licensees had established programs and processes to maintain their facility's design bases, there was a need to implement plant-specific followup activities. This determination was based upon the staff having identNied: (1) instances in which licensees failed to reconcile regulatory performance with their assertions that their programs and processes were effective in maintaining their design bases, or (2) that there was a need to gain a
- better understanding or to validate a particular aspect of a licensee's programs and processes.
SECY-97-160 referred to the above-mentioned followup activities as Phase 4 and stated that they were to be a combination o' erchitect-engineer design team inspections led by the Office of Nuclear Reactor Regulation and region-led inspections, such as safety system functional inspections and safety system engineering inspections.
In addition to the 950.54(f) letters and the inspection activities, the staff conducted lessons-learned reviews of Millstone and Maine Yankee. One of the conclusions of these reviews was that the definition of design bases should be clarified. In SECY-97-205, dated September 10,1997, the staff provided the Commission with several options for an integrated approach to solving the problems identified during the lessons-learned reviews. In the staff requirements memorandum on SECY-97-205, dated March 24,1998, the Commission directed the staff to continue to develop guidance regarding design basis issues, such as specifying the type of information to be considered as design basis information. This effort was subsequently included in the staff's response to the Chairman's tasking memorandum of August 7,1998.
In October 1997, NEl submitted an update to NUMARC 90-12, NEl 97-04, also titled " Design Bases Program Guidelines," which gave additional examples of design basis information and directly addressed the reportability of conditions outside the design basis of the plant. The staff commented on NEl 97-04 in August 1998 and has met with NEl several times since then.
Objective The staff's objective is to provide a clear definition of what constitutes design basis information.
A clear, common understanding is required so that regulations that use the term " design bases" are implemented correctly and are well understood. A clear definition is also required to make clear the functions that structures, systems, and components (SSCs) are designed to perform, and under what conditions they are required to perform them, when resolving or ieviewing the resolution of degraded and nonconforming conditions.
Discussion The term " design bases" is defined in 10 CFR 50.2 because it is used in several regulations in Part 50. Specifically, it is currently found in $$50.34(a) and (b),50.72, and 50.73. The definition is provided below for reference.
Design Bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted " state of the art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.
The importance of understanding what constitutes design basis information with regard to its use in 650.34 is that the design bases are required to be included in the final safety analysis report (FSAR) and, through 650.71(e), the updated FSAR. The staff has worked closely with the industry and is endorsing, through a Regulatory Guide, NEl 98-03, " Guidelines for Updating Final Safety Analysis Reports." For example, this document describes a method for complying with the requirements of @50.71(e) and states that for new regulatory requirements--
UFSARs must be updated to reflect changes to the facility resulting from new or amended requirements, e.g., Appendix R, the Station Blackout rule (10 CFR 50.63), the Anticipated Transient Without Scram (ATWS) rule (10 CFR 50.62), or plant-specific orders. As a result of such new requirements, the following information must be incorporated in the UFSAR, as applicable:
new or modified design bases e
summary of new or modified safety analyses e
appropriate UFSAR description as defined in Section 3.7 of this guideline e
Thus, without a clear understanding of the term "new or modified design bases," the industry and the staff will have difficulty reaching agreement on what information should be included in UFSAR updates.
With regard to the reporting requirements of G 50.72 and 50.73, licensees are currently required to notify the NRC following the occurrence of any event or condition that results in the nuclear power plant's being in a condition that is outside of the design basis of the plant. The a
e issue of when a plant is outside its design basis and what reports the NRC needs has been the subject of much discussion between the industry and the staff recently. In SECY-99-119, dated April 19,1999, the staff forwarded a proposed rule to the Commission that would modify the
- reporting requirements of $650.72 and 50.73. The proposed rule would, among other things, delete the requirement for reporting a condition that is outside the design basis of the plant and replace it with a requirement to report significant design issues that affect the ability of a structure or system to perform its specified safety function. If the proposed rule is approved by the Commission, questions regarding when to report design issues should be minimized and discussions about what it means to be outside the design basis of the plant will have less urgency.
Although no reference to design bases is found in other regulations now, proposed revisions to
.$50.59 would add new criteria that would use the term. Specifically, the staff is proposing to
- add a criterion to require prior NRC approval if a change, test, or experiment would result in a
- departure from a method of evaluation used in' establishing the design bases or in the safety 1
analyses. The staff is also proposing a criterion to require prior NRC approval if a change, test, or experiment would result in a design basis limit for a fission product barrier's being exceeded i
or altered. Since 650.59 evaluations are performed so frequently by diverse members of a licensee's staff, the importance of understanding what constitutes design bases will clearly be important in this application.
Aside from its use in interpreting regulations, understanding design bases is also important in the evaluation of degraded and nonconforming conditions. Licensees routinely encounter situations in the plant in which the performance of an SSC has degraded from its original J
intended performance. NRC inspectors use the NRC Inspection Manual Part 9900:" Technical Guidance - Resolution of Degraded and Nonconforming Conditions" to evaluate the licensee's actions. A clear, common understanding of the functions and controlling parameters for plant
- SSCs is important to allow licensoes to tako appropriate actions and to allow NRC inspectors to review the actions in the proper context.
Recent joint efforts among staff and industry to develop a new reactor oversight process have highlighted the importance of inspections in evaluating licensees' engineering programs. When the task forces working on the new reactor oversight process developed a set of performance indicators they determined that design issues are difficult to assess and that added emphasis on inspection will t'e required. it should be noted that these design engineering inspections will be focused on risk-significant systems. The definition of design bases is important to the NRC staff's evaluation of a licensee's compliance with the quality assurance requirements in Appendix B to 10 CFR 50, such as Criterion til for design control. Therefore, a clear understanding of what constitutes design bases is important to both the inspectors and the licensees.
Guidance.
The staff has developed guidance that is intended to clarify the 950.2 definition of design bases. This guidance is an updated version of the guidance transmitted in a letter to NEl dated January 4,1999, and is included as Attachment 1 to this paper. The guidance is meant to provide an explanation of how the staff interprets some of the key phrases in the definition.
Although intended to be a stand-alone document, several points deserve further discussion here.
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The first point is that the scope of the functions to be considered for design bases is not limited to safety functions or some other subset of the functions described in the UFSAR. Previously, the staff had attempted to distinguish these functions as those required by NRC regulations, license condition, or order. After additxmal research and discussion, however, the staff has i
determined that the $50.2 definition does not leave room for such interpretation and that all i
functions described in the UFSAR, that are relied on to satisfy NRC regulatory requirements, must be considered.
The second point is that the term " design basis functio.f means not just the action or response of an SSC but also the conditions or environment under which it is performed. Examples are l
provided to illustrate this point in the guidance.
The third point is that the staff expects that design bases will describe the functions that SSCs -
perform along with a short description of how they will be performed. The staff does not expect i
the design bases to include detailed descriptions of SSCs or analyses of their performance it should also be noted that there is not a direct association between design basis functions and l
design basis values. That is, there can be functions without an associated value.
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Finally, the staff does not expect licensees to add information to their UFSARs to comply with this clarification of design basis information. Assuming that each licensee has properly updated L
its UFSAR in accordance with the requirements of 650.71(e), no additional information should 1
be required.
To aid in understanding the attached guidance and to ensure that the guidance can be l
implemented, the staff developed several examples of design basis information. These examples are provided as Attachment 2 and were developed by reviewing the improved l
standard technical specification bases, several plant UFSARs and associated safety evaluation reports, and the Standard Review Plan (NUREG-0800).
Attachments: As stated i
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y Guidance for Determining Design Basis Information as Defined in 10 CFR 50.2 10 CFR 50.2 Definition: Design Bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.
These values may be (1) restraints derived from generally accepted " state of the art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.
General Guidance:
- 1. The 10 CFR 50.2 design basis functions of a plant's structures, systems, and components (SSCs), are the functions described in the UFSAR relied upon to satisfy NRC regulatory requirements.
- 2. The 10 CFR 50.2 design basis values are the bounding reference values set forth in the UFSAR and are derived from the design process, in determining these values for a particular SSC, a licensee must consider all conditions affecting the SSC's capability to perform their intended functions.
Soecific Guidance:
- 1. The term " design basis function" means the action or response of an SSC, given the conditions that may affect its performance when it must perform its functions, and presuming that its support systems also perform their functions. Paragraph 5. provides a list of requirements and conditions that should be considered when determining whether or not a function is a design basis function. Some examples of SSCs and their design basis functions are included here. Additional examples are provided in Attachment 2.
- a. The function of an isolation valve is not merely to close. Its function is to close under specified differential pressure, seismic and normal structural loading, and environmental conditions; assuming the valve meets specified material and mechanical requirements and assuming the support systems also perform their functions.
- b. The function of a pump is not merely to provide a specified quantity of water per unit time, its function is to do so against a specified head, if necessary, given a minimum NPSH and environmental conditions; assuming the pump meets specified material and mechanical requirements and assuming the support systems also perform their functions.
- c. A function of the electrical power system is to provide electrical power from the transmission network to the onsite electric distribution system by two physically independent circuits designed and located so as to minimize, to the extent practical, the likelihood of their si.multaneous failure under operating and postulated accident and environmental conditions.
- 2. The term " functions described in the UFSAR" means the functions of an SSC as described in the UFSAR relied upon to satisfy regulatory requirements. The regulatory requirements include those in NRC regulations, orders, and licenses. The level of detail of the discussion of the functions will vary from plant to plant depending, generally, on when the plant was licensed.
The design basis functions are specified in the UFSAR and were found acceptable by the staff for initiallicensing.
The design bases of a facility are a subset of the licensing basis and are required pursuar,t to 10 CFR 50.34(a)(3)(ii) and (b) to be included in the FSAR. The design bases of the plant may change over the life of the plant, as documented in the licensee's analyses performed pursuant to 10 CFR 50.59 and staff SERs on amendments issued, and each licensee is required to update the UFSAR in accordance with the requirements of 10 CFR 50.71(e). The staff does not intend to impose a new minimum level of detail for design basis information.
- 3. Structure and component functions may be implicitly subsumed within UFSAR descriptions of system functions. (Some functions, such as a valve body's pressure retaining function, may not be explicitly identified in the UFSAR.)
- 4. Functional requirements are applicable to all conditions of plant operation during which an SSC must operate. These conditions include, in addition to accidents and anticipated operational occurrences: plant startup, normal operation, shutdown, special or infrequent operation, and system abnormal or emergency operation. However, an SSC may not be required to perform particular design functions in one (or more) of these conditions, as set forth in the safety analysis.
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- 5. The design bases should provide sufficient detail to permit a conclusion on how requirements are met. Sufficient detail may require the inclusion of the following information:
- a. A definition of those events, transients, and accidents for which the SSC must function as designed,
- b. A definition of those events, transients, and accidents for which the SSC must be designed to withstand.
- c. Commitments to codes, standards, and Regulatory Guides, including the applicable issue and/or addenda.
- d. Design conditions such as pressure, temperature, fluid chemistry and voltage.
- e. Loads such as seismic, wind, thermal, and dynamic.
- f. Environmental conditions anticipated during normal operation and during transient and accident events such as pressure, ternperature, fatigue, humidity, corrosiveness, nuclear radiation, electromagnetic radiation, water table fluctuations, and duration of exposure.
- g. Interface requirements including definition of the functional and physicalinterfaces involving SSCs (e.g., electrical / mechanical system interface, safety /non-safety system interface)
- h. Material requirements including such items as compatibility, electrical insulation properties, protective coating, and corrosion resistance.
- l. Mechanical requirements such as vibration, stress,. shock, and reaction forces.
J. Structural requirements covering such items as equipment foundations, electrical raceway supports, and pipe supports.
- k. Hydraulic requirements such as pump net positive suction head (NPSH), allowable l
pressure drops, allowable fluid velocities, maximum pump discharge pressure, and j
minimum or maximum flow rates necessary to assure design functions are met.
- l. Chemistry requirements such as limitations on water chemistry.
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- m. Electrical requirements such as capacity / capability of power supplies, power supply availability, maximum / minimum voltage, voltage quality, raceway cable fiH, raceway ventilation, cable derating, and maximum / minimum load necessary to assure design functions are met.
- n. Instrumentation and control requirements for normal operation, reactor protection, and transient and accident mitigation.
- o. Requirements for redundancy, independence, and testability to meet single failure requirements.
- p. Analysis and procedural requirements to demonstrate operability of SSCs that lack testability.
- 6. Design bases do not include additional design information such as; design analyses, design output documents, and information regarding implementation of the design bases that is contained in other documents, some of which are docketed and some of which are retained by the licensee.
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L PWR Auxillary Foodwater System (AFWS) 10 CFR 50.2 Design Bases I
- The function of the AFWS is to supply water to the steam generators for reactor decay heat removal and for removal of other residual and process heat associated with reactor operation (such as pump heat) during plant startup, shutdown (including post fire safe shutdown) and layup operations; during anticipated operational occurrences (such as reactor trip with loss of offsite power); during postulated accident conditions; during station blackout; and in response to anticipated transients without scram.
[Most of the functional design requirements and criteria for the AFWS are included in the General Design Criteria and NUREG-0800, Standard Review Plan, Section 10.4.9. Any Information contained in the FSAR that describes how these criteria are satisfied,
.ncluding any exceptions that were allowed by the staff, is considered design basis information]
Examoles of Soecific AFWS Desian Basis Information:
The AFWS is configured into [three] trains, with the auxiliary feedwater (AFW) pumps taking a suction through separate and independent suction lines from the condensate storage tank (CST) as the primary safety-grade water supply, and pumping the water to the steam generator secondary side via separate and independent connections to the main feedwater lines downstream of the feedwater isolation valves. A minimum water level of [X ft.] at a temperature less than [Y *F] must be maintained in the CST in order to satisfy the AFWS design-basis feedwater requirements. The backup safety-grade water supply for the AFWS is the essential service water system. Automatic transfer to the backup source is provided on low AFW pump suction pressure.
The AFWS has [two] electric-motor-driven pumps and one turbine-driven pump. Each of the electric-driven pumps supplies (two] different steam generators, and the turbine-driven pump supplies all (four] steam generators. All three pumps wjll deliver at least
[200%) of the minimum required feedwater flow, while each motor-dnven pump will deliver at least [50%) and the turbine-driven pump will deliver at least [100%] of the required feedwater flow, as assumed in the accident analysis.
The AFWS possesses diversity in motive power sources such that system pedormance requirements may be met _with either of the assigned power sources (e.g., a system with ao a-c subsystem and a redundant steam /d-c subsystem).
Following any of the postulated accident conditions (or other events that involve a loss of normal feedwater, including station blackout), the AFWS will start automatically and supply enough water to the steam generators to prevent the release of reactor coolant through the pressurizer safety valves, at a discharge head that exceeds the lowest steam generator safety valve setting by [3%). The system design will include necessary instrumentation and controls to prevent overcooling of the reactor coolant system and overfilling of the steam generators.
b The AFWS can be operated from the control rcom, using instrumentation and control a
features that are sufficient to permit operation at hot shutdown conditions for at least
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[four] hours, followed by cooldown to [Z 'F] (the maximum allowed cut-in temperature
' for the residual heat removal system), using only safety related equipment and assured safety-grade water supplies, assuming the worst-case single active failure.
Based on an analysis using methods and data presented in NUREG-0611 and NUREG-0635, the AFWS has an unreliability that is less than [10 ] per demand.
d The AFWS will continue to satisfy its functional requirements following a postulated break in the AFW piping inside containment together with a single electrical failure.
The AFWS will automatically terminate AFW flow to a depressurized steam generator and automatically provide feedwater flow to the intact steam generators.
The turbirie driven pump will operate with steam produced in the steam generators (i.e.,
i from one ot' two steam supplies located upstream of the main steam isolation valves) l and deliver sufficient feedwater flow to a common header that is capable of feeding all steam generators to safely cool down the reactor coolant system without relying on AC power for a period of [two] hours.
The capability exists to detect, collect, and control system leakage and to isolate portions of the AFWS in case of excessive leakage or component malfunctions.
The following conditions will automatically start the turbine-driven AFW pumps:
low low steam generator level in [2 of 4] steam generators a
loss of one main feed pump if power is greater than [80%]
e loss of both main feed pumps a
Engineered Safety Features actuation signal station blackout The following conditions will automatically start the motor-driven AFW pumps:
low-low steam generator level in any single steam generator a
loss of one main feed pump if power is greater than [80%)
a loss of both main feed pumps
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Engineered Safety Features actuation signal station blackout (when power is restored to the emergency busses) a The turbine-driven AFW pump shall be designed to meet the requirements of ASME Code for Pumps and Valves, Class 3.
The AFW system is Safety Category I and Quality Group C to the motor-operated AFW stop valve; the motor-operated stop valve and the downstream check valve and piping are Quality Group B.
The pumps shall be designed tc ri set the requirements of ASME Section lil, Class 3.
The pump motors are Category ILI and shall meet the requirements of NEMA MG1 1969.
The design of the AFW system satisfies the requirements of 10 CFR 50, Appendix A, General Design Criteria 2,4,5,19,34,44,45 and 46.
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t Examoles of Tooical AFWS Desian Basis information:
Application of general system design and quality requirements, such as seismic, ASME Code Class, Quality Group designation, and required load combinations; and disposition
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of special interest design topics ihot have general applicability, such as water hammer.
Requirements for protection from natural phenomena, flooding, internal and external missiles, pipe whip, jet impingement, etc.
J General accident scenario assumptions, such as application of the single active and
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passive failure criteria, the effects of failure of non-essential equipment, and assumptions associated with availability of off-site power.
1 General requirements associated with operational testing and the installation of
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instrumentation and control features for confirrring operational readiness and system capability.
Fire protection and environmental qualification requirements.
Examples of Succortina Desian Information for the AFWS:
j Separate Engineered Safety Features quality power subsystems (i.e., independent j
Class 1E power supplies) and control air subsystems serve each AFW pump and its
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associated valves. The valves associated with the turbine-driven pump are served by both electric and control air subsystems, with appropriate measures precluding any interactions between the two subsystems. The turbine-driven pump receives control power from a third DC electric channel that is distinct from the channels serving the electric pumps. The eight air-operated flow control valves automatically control steam generator level to a preselected setpoint.
The motor-driven AFW pumps shall be of a horizontal centrifugal design, driven by a continuous duty electric motor to meet the system operating requirements during normal and emergency conditions. The pump bearings shall be air cooled and equipped with constant level oilers.
The AFW pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.
The condensate storage tanks are lined to prevent corrosion; other components are protected by chemical additions to the water.
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t Turbine Generator 10 CPR 50.2 Design Basse
' The functum of the turbine generator system is to convert the energy from the nuclear steam j
supply system into electrical energy. The turbine generator system is designed in a way that protects structures, systems, and components important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles.
Examples of Soecific Turbine G1nerator Desian Basis information:
The turbine generator will be operated primarily as a base loaded unit with an output of
[XXXX] MWe net. The turbine generator is capable of load following when required (i.e., capable of following generator-demanded load changes when these load changes occur within the automatic control range of 15-to 100-percent full power).
The turbine generator unit and associated systems is capable of a [XX%) load reduction without producing a reactor trip by dumping steam into the main condenser through the turbine bypass system. The maximum rate of load reduction without steam bypass will i
be a [XX percent] step change or a [X percent per minute] ramp as dictated by the reactor coolant system.
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l The turbine generator provides input to the reactor protection system i.e., a reactor trip j
on turbine trip due to low hydraulic control oil pressure (2-out-of-3) when power is above
[XX] percent, and a reactor trip when all four turbine stop valves close when power is above [XX] percent.
The turbine control and overspeed protection system controls turbine action under normal and abnormal operating conditions and assure that a full-load turbine trip will not cause the turbine to overspeed beyond acceptable limits. Under these conditions, the i
control and protection system permits an orderly shutdown by use of the turbine bypass system and main steam relief system. The overspeed protection system is redundant j
(meets the single failure criterion) and diverse and is testable when the turbine is in operation. The overspeed protection function is provided by the two-channel electro-hydraulic controls (EHC), backed up by a load-rejection relay (LRR) and the mechanical-hydraulic controls (MHC), plus two overspeed trip devices. In the event of a load rejection, the EHC normally acts to close the control valves. In addition, an LRR is provided to monitor generator output; if it senses that a negative load gradient over a certain load range exceeds its settings, it acts to open solenoid valves in the hydraulic control lines, thus initiating fast closure of the control valves for a preset period of time.
For normal speed-load control, the speed governor action of the electro-hydraulic control system fully cuts off steam at approximately 103 percent of rated turbine speed by closing the control and intercept valves.
A mechanical overspeed trip device is provided that will actuate the stop, control,
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reheat steam stop, and intercept valves at approximately [XXX] percent of rated speed.
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An independent and redundant backup electrical overspeed trip circuit is provided that senses the turbine speed by magnetic pickup and closes all valves associated with speed control at approximately [XXX] percent of rated speed.
The turbine high pressure stop and control valves and reheat steam stop and intercept valves are arranged and have valve closure times such that a failure of any single valve to close will not result in excessive turbine overspeed in the event of a turbine generator trip signal. The high pressure stop and control valves have closing times of less than
[XXX] milliseconds and the low pressure stop and control valves less than [XXX]
milliseconds.
The extraction steam check valves provided at extraction connections shall be beated close to the turbine and capable of closing within one second from the initiation of the reverse flow to maintain stable turbine speeds in the event of a turbine generator trip signal.
The turbine generator automatically trips on any of the following signals: turbine overspeed, steam generator high-high level or safety injection, low condenser vacuum, thrust bearing failure, excessive vibration, low bearing oil pressure, moisture separator drain system high level, prolonged loss of generator stator coolant, loss of hydraulic control fluid pressure, manual turbine trip from the control room, manual turbine trip at the turbine, generator electrical trips, and reactor trip.
After an automatic turbine trip, the generator breaker is delayed to furnish uninterrupted power to the reactor coolant pump motors for at least [XX] seconds without relying on the success of a bus transfer, provided the generator conditions permit this. Likewise, following a manual turbine trip or a turbine trip due to certain turbine faults or certain generator protection signals, the generator trip is delayed approximately [XX] sec.
Generator trip (after a turbine trip) is also conditional upon detection of reverse power, except for certain generator faults, to minimize the probability and the degree of,
overspeed after a turbine trip.
The turbine generator system is designed to permit periodic testing of components important to safety while the unit is operated at rated load.
Connection joints between the low pressure turbine exhaust and the main condenser are be arranged to prevent adverse effects on any safety-related equipment in the turbine room in the event of rupture. Safety-related systems or portions of systems located in the vicinity of the turbine generator system (e.g., components of the anticipatory reactor trip channels) are designed to withstand the affects of pipe cracks or breaks in high-and moderate-energy piping systems outside containment.
Topical desian bases elements for the Turbine Generator:
Components, piping, and structures in the turbine generator system are designed in accordance with applicable codes and standards.
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Containment isolation Motor Operated Valve (MOV)
- 10 CFR 50.2 Design Bases The function of a Containment isolation MOV is to isolate fluid flow through the process line in accordance with leakage limits upon receiving a closure signal during plant startup, shutdown, and layup operations; during anticipated occurrences (such as reactor trip with loss of offsite i
power); during postulated accident conditions; and during station blackout (if applicable). The MOV might also have a function to reopen following its isolation function to allow re-establishment of fluid flow in the process line.
Examples of Desian Basis information for the Containment Isolation MOV Flow rate, differential pressure, system pressure, fluid type and quality, and temperature a
of the process fluid under which the Containment Isolation MOV must be capable of operating under full range of plant conditions.
Ambient pressure, temperature, radiation, humidity and seismic conditions for the environment of the Containment Isolation MOV for equipment qualification (EQ) under the full range of plant conditions.
Leakage limits for isolation of the process fluid.
ASME Code Class and applicable requirements.
Stroke-time requirements for the valve to perform its functions under full range of plant conditions.
Valve thrust and torque operating requirements to perform its functions under full range of plant conditions.
Motor actuator output thrust and torque capability under full range of plant conditions including degraded voltage conditions.
Weak link limitations for the valve, actuator, and motor.
Short-term and lifetime operating cycle requirements.
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Nuclear Energy Institute Project No. G89 cc:
Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy institute Nuclear Energy Institute Suite 400 Suite 400 1776 l Street, NW 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708 Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy Institute ABB-Combustion Engineering, Inc.
Suite 400 12300 Twinbrook Parkway, Suite 330 1776 l Street, NW Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Engineering
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Nuclear Energy Institute Suite 400
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1776 i Street, NW Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy institute Suite 400 1776 i Street, NW Washington, DC 20006-3708 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. Jim Davis, Director Operations.
Nuclear Energy Institute Suite 400 1776 l Street, NW l
Washington. DC 20006-3708 f
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Ps DISTRIBUTION: Ltr to NEl re Design Bases Dated May 1a 1000 ntral File UBLIC RGEB r/f OGC ACRS SCollins/RZimmerman WKane BSheron GHolhan JStrosnider CACarpenter j
FAkstulewicz TBergman SMagruder GTracy, EDO WLanning, RI BMallett, Ril JGrobe, Rill AHowell, RIV e-mail -
JTatum EMcKenna
- DFischer TScarbrough CJackson RWeisman MMarkley, ACRS DAllison I