ML20141F301
| ML20141F301 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/25/1984 |
| From: | STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20141F293 | List: |
| References | |
| PROC-840925, NUDOCS 8601090172 | |
| Download: ML20141F301 (76) | |
Text
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J.O.No. 12241 November 15, 1973 2BVM-45 Rev. October 21, 1976 f
Responsible Lead En r SJdmA./u7 /
Rev. July 8, 1977 Originating Engineer
&,21ume 46 Rev. May 5, 1978
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Project Engineer tA4a,,uA Rev. December 10, 1980 Rev. March 17, 1982 Rev. September 7, 1982 Rev. January 25, 1983 Rev. June 6, 1983 Rev. 9, March 29, 1984 Rev. 10, Sept.25, 1984 i
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Procedure for PREPARATION OF SYSTEM DESIGN INFORMATION g
0,6 REQUIRED FOR PIPE STRESS ANALYSIS
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Beaver Valley Power Station - Unit No. 2 b
Duquesne Light Company Pittsburgh, Pennsylvania F. 366 EK 8 40 9 2 5 0CO2 IEDIPENDENT RETIEVER (EAP 3.1)
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PovEn
.c UNCOMTROLW) COPY Stone & Webster Engineering Corporation Boston, Massachusetts w
8601090172 860103 PDR ADOCK 05000412 A
Tcble of Contents Page 1.0 PURPOSE AND SCOPE 1
2.0 REFERENCES
1
3.0 DESCRIPTION
OF PLANT AND SYSTEM OPERATING CONDITIONS 4
3.1 ASME III Systems 4
3.2 Non-ASME III Systems (Class 4) 6 4.0 PREPARATION OF INFORMATION REQUIRED FOR ASME III, CODE CLASS 1 PIPING SYSTEMS 8
4.1 General 8
5.0 PREPARATION OF INFORMATION REQUIRED FOR ASME III, CODE CLASS 2 AND 3 PIPING SYSTEMS 8
5.1 General 8
5.2 Procedure 9
5.2.1 Preparation of Stress Analysis Data Package 9
5.2.1.1 Specification of System Conditions 9
'5.2.1.1.1 Categorizing and Identifying High and Moderate Energy Lines 12 5.2.1.1.2 Break and Crack Exclusion Identification 12 5.2.1.2 Flow Transient Data 12 5.2.1.3 Safety / Relief Valve Information 13 5.2.1.4 Input References 13 5.2.1.5 Miscellaneous Information 14 5.2.2 Review and Approval of Stress Analysis Data Package 15 5.2.3 Revisions to Stress Analysis Data Package 15 l
6.0 PREPARATION OF INFORMATION REQUIRED FOR NON-ASME III PIPING SYSTEMS (CLASS 4) 15 6.1 General 15 6.2 Procedure 16 6.2.1 -Preparation of Stress Analysis Data Package 16 6.2.1.1 Specification of System Conditions 16 6.2.1.2 Flow Transient Data 17 6.2.1.3 Safety / Relief Valve Information 17 6.2.1.4 Input References 17 6.2.1.5 Mis.cellaneous Information 17 6.2.2 Review and Approval of Stress Analysis Data Package 17 6.2.3 Revisions to Stress Analysis Data Package 17 I
7.0 PROCEDURE FOR PREPARATION OF LINE DESIGNATION TABLE (LDT) 18 7.1 General 18 7.2 Preparation 18 7.3 Review and Approval of LDT 21 7.4 Distribution 21 8.0 PROCEDURE FOR PROCESSING PIPE STRESS DATA 21 8.1 General 21 u
Table of Contents (Cont'd)
Page 8.2 System Engineer Develops and Maintains Stress Information Data 21 8.2.1 Develop and Maintain Stress Analysis Data Package 21 8.3 Pipe Stress Coordinator (PSC) Processes and Transmits Stress Analysis Data Package to Pipe Stress and Supports Engineering (PSAS) 22 8.4 PSAS Completes Stress Analysis 22 8.5 Supplemental Action by System Engineer /PSC 22 8.6 Figure 1 - Relationship of Various Plant Conditions 23 8.7 Figure 2 - Relationship Along Plant Conditions for Various System Conditions 24 Appendix I
- Plant Conditions Used As A Basis For BVPS-2 ASME III Piping System Design Appendix II - Not used Appendix III - Forms
- Appendix IV - Flow Transient Table (NUREG-0582)
Appendix V - Additional Flow Transients 9
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t PREPARATION OF SYSTEM DESIGN INFORMATION REQUIRED FOR PIPE STRESS ANALYSIS 1.0 PURPOSE AND SCOPE This procedure provides instructions for preparing, reviewing, distributing, and revising system stress analysis data packages which contain. system condition descriptions and temperature, pressure, and equipment information required for pipe stress analysis for piping 4-systems in Beaver Valley Power Station - Unit No. 2 (BVPS-2).
Section 2 provides a listing of appropriate documents referenced throughout this procedure.
Section 3 provides definitions and clarification of various terms used in deriving the input for pipe stress analysis.
Sections 4, 5,
and 6 provide the-detailed instructions for the preparation of information required for the analysis of ASME III, Code Class 1 systems, ASME III Code Class 2 and 3 systems, and Non-ASNE III (Class 4) systems, respectively.
Section 7 provides instructions for completing Line Designation. Tables.
Section 8 provides the detailed instructions for processing the above information. This procedure meets the intent of Power Division Technical Procedure PTP-26.1.15-0 for specification of system information.
2.0 REFERENCES
2.1 ASME III American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section'III, 1971 Edition, including all addenda thereto issued through the Winter 1972 addendum.
2.2 ANSI B31.1 American National Standards Institute, Code for Pressure Piping.
ANSI B31.1.0, 1967 Edition and all addenda thereto including B31.1.0d addendum dated June 30, 1972.
2.3 PTP40.4.1 i
Power Division Technical Procedure, PTP-0.4.1-0 Classification of Structures, Systems, and Components for Nuclear Projects dated July 17, 1980, and all' issued revisions.
2.4 PTP-26.1.15 Power Division Technical Procedure, PTP-26.1.15-0 Information For Pipe Stress Analysis and Pipe Supports Design / Analysis For Nuclear Power Plants, dated April 14, 1981, and all issued revisions.
2.5 PTP-0.8.1 Power Division Technical Procedure, PTP-0.8.1-1, Definitions and Terminology for Pressurized Water Reactors, dated August 14,
- 1980, and all issued revisions.
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2 2.6 2BVS-939 Design Specification for Piping Engineering and Design, ASME III, Code classes 1, 2;
and 3 and ANSI B31.1 Class 4, dated June 29, 1984 l
including all issued addenda and revisions.
2.7 2BVS-939A Specification for Stone & Webster Pipe Classes, dated December 21, 1983, l
and all issued revisions.
2.8 2BVS-60 Specification for Thermal Insulation, dated April 21, 1983, and all l
issued revisions.
2.9 ASME III Code Case 1606-1 Stress Criteria,Section III, Classes 2 and 3, Piping Subject to Upset, Emergency, and Faulted Operating Conditions, dated December 16, 1974.
2.10 EMTP 9.12 Engineering Mechanics Technical Procedure, EMTP 9.12-0, Flow Transient Loads For Piping System Design, dated October 5, 1983.
2.11 EMTP 9.22 Variable Spring Hanger Design, dated July 28, 1982.
2.12 2BVM-1 Purchase Order and Specification Numbering System - Parts 1 and 2, dated October 7, 1980 and all issued revisions.
2.15'2BVM-6 Manufacturers' Drawing Handling System, latest revision dated October 29, 1982, and all issued revisions.
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2.14 2BVM-12 Instructions. for Preparation of Flow Diagrams, dated June 18, 1983 and all issued revisions.
2.15 2BVM-13 Instructions for Drawing Handling, and Review, dated September 8, 1972, and all issued revisions.
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~ 2.16 2BVM-17 Personnel Assignments, latest revision dated September 30, 1981 and all issued revisions.
2.17 2BVM-23 Schedules, Reports, and Controlled Indices, dated July 9, 1979, and all issued revisions.
2.18 2BVM-29 Vendor-Supplied Documentation Control Procedure, dated April 23, 1973, and all issued revisions.
2.19 2BVM-81 Tabulation of SWEC Mark Numbers for Valves, Steam Traps, and Strainers, latest revision dated June 22, 1983, and all issued revisions.
2.20 2BVM-85
- Criteria for Postulating Pipe Breaks and Cracks and Analyzing the Dynamic and Environmental Effects Inside and Outside Containment, latest revision May 31, 1984, and all issued revisions.
2.21 2BVM-94 Handling of Boston-Generated Engineering and Design Coordination Reports
' (E&DCR's), dated November 7,1977, and all issued revisions.
2.22 2BVM-116.
Seismic Classification for structures, Systems, and Components, dated February 12, 1979, including.all issued revisions.
2.23 Reference Deleted 2.24 2BVM-119 Environmental Conditions for Equipment Qualification Requirements, dated April 19, 1984, including all issued revisions.
2.25 2BVM-121 ASME III Class 1 System Operating Conditions, dated September 28, 1983, including all issued revisions.
2.26 2BVM-146 Project Equipment System, latest revision June 11, 1982, and all issued revisions.
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-2.27 ASME XI American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, " Rules for Inservice Inspection of Nuclear Reactor Coolant Systems," 1977 Edition, including all addenda thereto issued through the Summer 1978 addendum.
2.28 2BVM-2 Filing Instructions and Job Book Index, latest revision August 30, 1983, and all issued revisions.
2.29 2BVM-150 Procedure for AnalysisofBuhied Piping, dated July 16, 1984, and all 4
issued revisions.
2.30 2BVM-157 Design Criteria Document for Pipe Stress, Pipe Supports, and Duct Supports, dated May 9, 1984, and all issued revisions.
2.31 NUREG-0582 4
Water Hammer in Nuclear Power Plants, issued July 1979.
3.0 DESCRIPTION
OF PLANT AND SYSTEM OPERATING CONDITIONS 3.1 ASME III Systems The following terms are used to define the allowable stress levels and the appropriate loading combinations applicable for the design of ASME III piping systems:
3.1.1 Plant Condition Plant Condition is defined as the condition of the power plant as a whole resulting from a postulated event. Reference Figure 1.
A plant condition is subdivided into either a Plant Operating Condition or a Plant Test Condition as described below.
4 3.1.1.1 Plant Operating Condition Plant Operating Condition is defined as the condition of.the power plant as a whole.to a postulated event other than testing. A plant operating condition is further subdivided into cne of the following conditions as defined in ASME III, Subsection NB-3113:
- 1) Normal Condition,
- 2) Upset Condition, 3) Emergency Condition, and 4) Faulted Condition.
Note: These four categories of operating conditions apply only to Plant Operating Conditions and not to System Operating l
Conditions as defined in Section 3.2.
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5 The specific Plant Operating Conditions to be used as the basis for the design of ASME III Code Class 2 and 3 systems on BVPS-2 are listed in Appendix I, and in 2BVM-121 for ASME III Code Class 1 Piping Systems.
3.1.1.2 Plant Test Condition Plant Test' Condition is defined as the condition of the power plant as a whole to a postulated testing event. The specific Plant Test Conditions to be used as the bases for the design of ASME III Code Class 2 and 3 systems on BVPS-2 are listed in Appendix I, and in 2BVM-121 for ASME III Code Class 1 Piping.
3.1.2 Syr. tem Condition System condition is defined as the condition of a particular system to a postulated event.
System conditions are subdivided into several categories as defined below and shown on Figure 2.
3.1.2.1 System Opera' ting Condition System operating Condition is defined as the condition of a particular system resulting from a postulated Plant Operating Condition.
3.1.2.2 System Test Condition During Plant Test Condition System Test Condition During Plant Test Condition is defined as the condition of a particular system during a postulated Plant Test Condition.
3.1.2.3 System Test Condition During System Testing System Test condition During System Testing is defined as the condition of a particular system while only the system and not the entire power plant is in a test condition.
During this testing, the unit is considered to be in a Normal Plant -Operating Condition.
Examples of this system condition are system hot and cold hydrostatic tests, periodic testing required by the Technical Specifications, and Inservice Inspection Operability Tests.
Although not a true test condition, chemical cleaning cf a system is also considered in this category.
For hydrostatic test conditions, the test pressure as defined in Table 3.5-5 of 2EVM-157 is to be used:
a) For.ASME Class 1, 2 and 3 Test Pressure = 1.25 x Design Pressure b) For ASME Class 4 Ter,t Pressure = 1.50 x Design Pressure u
6 3.1.2.4 System Design Condition System Design condition is defined as the condition of a particular system or portion of the system resulting from the Design Pressure and Design Temperature for that system as defined below.
3.1.2.4.1 Design Pressure The specified design pressure shall not be less than the maximum difference in pressure between the inside and outside of the piping system, sWbsystem, or component which exists during any System operating Condition associated with a Normal or Upset Plant Operating Condition.
Consideration should be given to those buildings whose ambient pressures could be greater or less than atmospheric.
Except as noted below, the system Design Pressures shall be derived from the maximum pressure occurring in the system or portion of the system during the Normal or Upset Plant Operating Condition.
In some cases, it may be necessary, however, to define the System Design i
conditions from System conditions which exist during Plant Conditions other than the Normal or Upset Plant Operating Condition. This is particularly relevant for systems which do not operate during the Normal or Upset Plant Operating Condition (e.g., most safeguards systems).
ASME III Code case 1606-1 provides a limit of 1.5 P for the Emergency Plant operating Condition, and 2.0 P for the Faulted Plant Operating Condition for all Code Class 2 and 3 systems (P is the Design Pressure).
These limitations on Design Pressure may require that the System Design
. Conditions be selected from System operating conditions which exist during the Emergency or Faulted Plant Operating Condition. This is consistent with the definitions of Design Pressure and Temperature.
These definitions only require that these design param=,
s not be less than those which occur during the Normal or Upset Plant Operating Conditions.
i 3.1.2.4.2 Design Temperature
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The specified design temperature shall not be less than the maximum metal temperature which exists during any system Operating Condition associated with a Normal or Upset Plant Operating Condition for the system, subsystem, or component considered.
Consideration should be given to those buildings whose ambient temperatures could be greater than the system temperatures.
3.1.2.4.3 Design Mechanical Loads i
The specific combinations of values of Mechanical Loadings which exist coincident with the above Design Pressures and Temperatures.
3.2 Non-ASME III (Class 4) Systems Class 4 piping systems are generally analyzed on the basis of conditions l
which exist within the system during no rmal plant operation.
The i
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following applies 'to Class 4 systems when defining parameters for normal plant operation.
3.2.1 Normal Plant Operation or Normal Plant Operating Condition The definition of Normal Plant Operation or Normal Plant Operating Condition is the same as the definition of the Normal Plant Operating 1
Condition defined for ASME III. Appendix I provides the Normal Plant Operating Conditions considered in the BVPS-2 design.
The stress analysis of a class 4 system is based primarily on the Normal Plant Operating Condition and considers the Maximum Operating Pressure and the Maximum Operating Temperature Range which can exist in a system i -
or portion of the system during the Normal Plant Operating Condition.
The exception to this is when a piping system does not function during the Normal Plant Operating Condition (e.g.,
relief valve discharge lines). For these exceptions, the normal operating condition is defined as the condition of the system when it operates to perform its intended function.
3.2.2 Maximum Operating Pressure (Design)
The Maximum Operating Pressure is the maximum. pressure the system or portion of the system will experience during the Normal Plant Operating Condition.
3.2.3 Maximum Operating Temperature Range (Design)
The Maximum Operating Temperature ~ Range is specified by selecting the maximum and the minimum temperature the system or portion of the system will experience during the Normal Plant Operating Condition.
Some Class 4 systems, however, may require analysis for transients other than those which are expected during Normal Plant Operating Conditions.
These systems fall into one or both of the following categories:
3.2.4 Class 4 Seismically Analyzed Systems (Seismic Category II)
Project Procedure 2BVM-116 provides a definition of Seismic Category I and Seismic Category II systems. All Seismic Category II piping (except as noted below) which is defined by 2BVM-116 shall be seismically stress analyzed to meet the requirements of NC-3600 of ASME III, 1971 Edition, through winter 1972 Addenda, except that seismic piping stresses need only be equated to faulted allowables (2.4 Sh) only to ensure that piping structural and pressure boundary integrity is maintained. Pipe l
stresses due to SSE anchor movements are to be omitted. The thermal and deadweight analyses shall meet the normal requirements of either ANSI B31.1 Power Piping Code, 1967 Edition, including Addenda through June 30, 1972 or ASME III as shown above. However, the allowable stress values (Se and Sh) shall always be taken from ANSI B31.1.
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- Seismic Category II piping which is directly connected to any ASME III class 2, or 3 piping shall satisfy the same criteria as the ASME III class 2, or 3 piping'up to and including the first anchor (s) beyond the ASME/non-ASME class break.
For Seismic Category II piping, which is directly. connected to ASME III class 1 piping, refer to Note 3 of Table 3.5-1.in 2BVM-157.
3.2.5 Class 4 Systems Which Experience Transients Which Are Expected to Exceed the Normal Operating Conditions Class 4 systems are analyzed to the rules of ANSI B31.1.
ANSI B31.1 recognizes that a piping system will' occasionally experience operating conditions which exceed those temperatures and pressures which are experienced during Normal Plant Operation. This situation is somewhat analogous to the Upset Plant Operating Condition defined for ASME III.
These System operating Conditions must be specified so that stresses from these transients can be analyzed.
4.0 PREPARATION OF INFORMATION REQUIRED FOR ASME III_
CODE CLASS 1 PIPING SYSTEMS 4.1 General The specific operating conditions and all other information required for ASME III Code Class 1 piping systems will be included in 2BVM-121 and Section 7 of 2BVS-939.
See also Appendix I herein.
5.0 PREPARATION OF INFORMATION REQUIRL3 FOR ASME III CODE CLASS 2 AND 3 PIPING SYSTEMS 5.1 General The design of ASME III, Code Class 2 or 3 systems and connecting non-ASME III piping systems up to and including the first anchor (s) must consider system conditions which are associated with postulated Plant conditions. Appendix I specifies the Plant Conditions which must be considered in the design of code Class 2 cr 3 systems: the appropriate loading combinations and the allowable stresses associated with a specific Plant Condition are specified in 2BVM-157.
.The responsibility of specifying the System conditions (i.e., System Operating, System Test, and System Design Conditions) associated with these specified Plant conditions is the System Engineer's.
(Refer to 2BVM-17 for System responsibilities.) Each System Engineer is also responsible to check all interfaces between systems to ensure that the system conditions defined are compatible.
The Stress Analyst is responsible for evaluating the calculated stresses in the system piping for each System Condition using the loading combinations and allowable stresses specified by the associated Plant condition. The calculational methods used for stress analysis are specified in 2BVM-157 which is a part of 2BVS-939 by reference.
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- 5.2 Procedure 5.2.1 Preparation of Stress Analysis Data Package The System Engineer is responsible for compiling a Stress Analysis Data Package consisting of the information required by the Stress Analyst to complete the stress analysis for each Code Class 2, or 3 system. This Stress Analysis Data Package is a controlled document which shall consist of two parts:
- 1) Specification of System conditions and
- 2) Stress Analysis Information. The completed stress package including attachments shall have consecutively numbered pages, and accountability of all page numbers shall be shown on either the cover sheet or the table of contents.
5.2.1.1 Specification of System Conditions The System Engineer shall specify the System Conditions using:
1.
The Plant Condition / System Condition Matrix (Attach-ment 2BVM-45A10) 2.
An appropriate number of System Condition Description Sheets (Attachment 2BVM-45A4) 3.
An appropriate number of Pipe Stress Data Tables (Attachment 2BVM-45A9)
The System Engineer must specify the System conditions which can occur during a Normal, Upset, Emergency or Faulted Plant Condition.
Hydrostatic Test, System Test (Pre-op, ISI/ Tech. Spec.) and Containment Pressure Test conditions must be included.
Each System Condition shall be numbered sequentially and shall be described in sufficient detail by the System Engineer on the System condition Description Sheet (Attachment 2BVM-45A4),
so that the Stress Analyst clearly understands the condition of the system. A brief system description of the function of the system is to be included under General Notes (Attachment 2BVM-l~
45A3). A flow path sketch for each system condition may be included as an attachment to the stress analysis data package for clarity.
References for all the data specified in the System Condition Caeets should be included (FSAR not an acceptable reference).
In addition, the System Engineer shall indicate which system condition shall be used as the Reference Operating Condition (ROC) to be used by the stress analyst as the reference position for dead weight design of variable spring hangers as. required by ENTP-9.22.
The ROC is usually the system condition which exists during normal plant operation.
In the section designated " operating components", the System Engineer shall list those i
components which must perform a mechanical motion (e.g., pumps, control valves) during the specific System condition.
The frequency of operation may also be included to describe if the condition is normal - sustained, occasional - 1 percent, or occasional -
10 percent as defined below
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10 Normal - sustained: loads where the duration exceeds 10 percent of j.
the operating period.
1 occasional 1 percent loads whose duration does not exceed 1 percent of the operating period.
Occasional 10 percent loads whose duration does not exceed 10 percent of the operating period.
1 once the system condition description sheets have been completed, the System Engineer shall match the numbers of the specific System e
l Conditions which apply to the specific Plant Condition specified on the Plant Condition / System Condition Matrix.
Appendix I provides a
description of all plant conditions which correspond to the numbers on the matrix.
After all System conditions have been specified on the matrix sheets, the System Engineer must complete the Pipe Stress Data Table (Attachment 28VM-45A9) for all Code Class 2, 3, and connecting non-ASME III lines in the system. The System Engineer shall enter the line numbers on the form in numerical order, and, in the appropriate columns, enter the System operating Condition pressures and temperatures.
He f
must also determine the anticipated total number of full temperature cycles over the total number of years of expected system operation and note this number in the stress analysis data package.
If the total j
number is greater than 7000, the value will be specified on the page for l
general notes (Attachment 2BVM-45A3), otherwise it will be indicated as 7000 or less. This information is required to determine the stress range reduction factor for cyclic conditions as specified in the code.
In the-case where the pressure and/or temperature may vary during a System operating condition, the following applies:
i 1.
If the pressure varies, enter the maximum pressure which may exist during the System condition 2.
If the operating temperature varies, and the variation is:
a.-
Less than 50*F, and the temperature range that may exist t
during the system, condition (i) is above the normal ambient temperature for that l
area, enter the maximum temperature which may exist
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during the system condition.
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(ii) is below the normal ambient temperature for that l
area, enter the minimum temperature which may exist during the system condition.
(iii) envelops the normal ambient temperature for that area, enter the temperature range (i.e. the maximum i
and minimum temperature).
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b.
Greater than 50*F, enter the temperature range (i.e. the maximum and minimum temperature).
3.
Where empty or filled lines remain static, the building ambient i
temperature range must be entered in the Data Table and a j
reference to 2BVM-119 or calculations 'added to the System condition Description Sheets.
For all cases, when the temperature range is entered in the Pipe Stress Data Table, the temperature variation must be explained on the System condition Description Sheet so that the Stress Analyst clearly understands the conditions of the system.
After all System Operating Conditions have been entered on the Pipe l
Stress Data Table, the System Engineer shall derive the Systei Design l
Conditions.
- Except, as noted below, the Design Pressure and Design Temperature shall be derived by selecting the maximum pressure and the maximum temperature which can exist for that line for all System conditions which are associated with the Normal or Upset Plant Operating Conditions.
These pressures and temperatures shall be entered in the columns designated " Design Pressure" and " Design Temperature" in the Line Designation Table (See Section 7.0).
The " Design Pressure" and
" Design Temperature" are not included in the Pipe Stress Data Table. As indicated in Section 3.1.2.4.1, ASNE III Code Case 1606-1 provides limits of 1.5 and 2.0 times Design Pressure for system pressures during the Emergency and Faulted Plant Operating Conditions, respectively. The pressures which occur in System Operating conditions associated with these Plant Conditions should be verified to be within these limits. If not,.the Design conditions should be derived from System Conditions corresponding to Emergency or Faulted Plant Operating Conditions as required.
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l Note: If the system or portion of the system is within the Westinghouse scope, the Design Conditions as derived above shall be verified to be the same as provided by Westinghouse.
If they are not the same but lower, the Westinghouse Design i
conditions should be specified.
If they are higher, the actual Design conditions shall be resolved with Westinghouse.
After all the system conditions have been specified, the System Engineer shall complete the analysis section on the title page by placing a "Yes" or "No" before each category, therefore identifying the types of analyses the Stress Analyst should use.
- Also, a Table of Contents (Attachment 2BVM-45A13) shall be included as page 2 of the stress package.
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5.2.1.1.1 Categorizing and Identifying High and Moderate Energy Lines NRC SRP's 3.6.1 and 3.6.2 require that safety related equipment be protected from high and moderate energy pipe failures such that plant shutdown / accident mitigation can be achieved. NRC SRP's 3.5.1.1 and 3.5.1.2 also require a similar protection requirement for internally generated missiles originating from lines with high internal energy.
In order to perform the accident evaluations for these SRP's, all line s,
including facilities piping, must be categorized and identified as either high energy, moderate energy, or moderate energy because of a system's short operational period.
The requirements by which piping lines can be categorized and identified in the Pipe Stress Data Table in the applicable stress analysis data package are provided in BVM-45A14.
It should be noted that 2BVM-85 requires that all Code Class 1 piping be defined as high energy including piping
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between normally closed Reactor Coolant Pressure Boundary isolation valves.
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5.2.1.1.2 Break and Crack Exclusion Identification l
NRC SRPs 3.6.1 and 3.6.2 require that safety-related equipment be protected from high and moderate energy pipe failures such that plant shutdown can be achieved and offsite doses are maintained below acceptable limits. Project procedure 2BVM-85 provides direction on how
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protection should be provided. One method available is to design the piping system to be " break exclusion" or " crack exclusion" i.e., due to low stress levels, as defined in NRC MEB 3-1, no breaks or cracks need be postulated. One specific exception to these criteria is the main steam and feedwater lines in the main steam valve house for which a 1 sq.
ft.
crack must be postulated for equipment environmental qualification in accordance with NRC ASB 3-1.
These systems are designed and classified as " break exclusion" up to and including this first isolation valve from the containment penetration.
Other examples are the piping from the containment penetrations to the first isolation valves outside containment on the HCS and CVS lines which are moderate energy and designed to " crack exclusion" requirements.
These and all lines requiring break or crack exclusion shall be identified by a note, I
which clearly defines the exclusion area, on the pipe stress data tables in the stress analysis data packages.
l 5.2.1.2 Flow Transient Data
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Flow transient loadings. associated with specific system operating conditions shall be evaluated for potential occurrence and significance.
i EMTP-9.12 identifies piping systems that may require evaluation of flow transients and defines the required input to complete analyses.
Additional guidance is contained in NUREG-0582.
Appendix IV lists the Pressurized Water Reactor fluid flow transient events identified in NUREG-0582 and EMTP 9.12.
A short description of the applicability of j
these transients to BVPS-2 is included.
Appendix V lists additional j
flow transient conditions analyzed by BVPS-2.
A flow transient data l
sheet (FTDS, Attachment 2BVM-45AS) shall be completed with the information described below for each applicable flow transient.
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13 Each FTDS.shall *contain a description of the transient occurring and sufficient data or references, as applicable, to completely describe the l
transient (e.g., valve opening / closing times, reference to pump curves, pump start times, voids in systems, etc.).
The FTDS shall also specify the system condition description number (s) during which the transient may occur. The FTDS shall be included in the system stress analysis data package after the System condition Description Sheets. Also, a reference to the flow transient data sheet shall be included on the
. applicable system condition description sheets.
The following ' valve and pump data or a reference to the data should be l
included on the FTDS as applicable: valve-opening / closing time, orifice
- area, flow rate, discharge coefficient (Cv), and fluid; Pump curves and flow rates.
5.2.1.3 Safety / Relief Valve Information Each safety / relief valve in the system shall be identified on a safety / relief valve data sheet (Attachment 2BVM-45A6). The sheet shall include:
1.
A short written description of the relief valve's operation, specifying whether it is a pressure or thermal relief valve 2.
The relief valve mark number 3.
The relief valve set pressure 4.
Valve characteristics or a reference for the following additional informations opening / closing time, flow
- rate, discharge coefficient (Cv),
orifice area, accumulation, temperature, and fluid.
An effective opening time of 0.020 seconds should be used for the
. purpose of transient analysis of the pressure relief valves.
The condition of downstream piping (filled or empty) should also be indicated for each pressure relief valve.
f If there are no relief valves in the system, "Not Applicable" shall be included in'the description section for completeness.
5.2.1.4 Input References The System Engineer shall provide references for all data included in the stress analysis data package. All references shall be identified with a revision number and/or date.
2BVM-119 or calculations shall be l
referenced for building ambient conditions for Normal, Anticipated operational occurrences and Accident Conditions.
The maximum post accident temperature for fluid systems within the containment is assumed to be 275'F; since the initial temperature spike to 370'F occurs over a short period of time, the piping and fluid would never experience the greater temperatures.
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- To identify lines to be seismically analyzed the stress analyst shall refer to 2BVM-116 for the building seismic category. All piping within Seismic Category. I structures must be at least Seismic Category II unless-specifically exempted in 2BVM-116. All QA Category I equipment and piping are Seismic category I.
5.2.1.5 Miscellaneous Information The System Engineer shall complete the miscellaneous information sheets (Attachment 2BVM-45A7) with equipment mark numbers and SWEC File Numbers for manufacturers' drawings._
Source documents other than vendor drawings shall be noted on the miscellaneous information sheets.
l The following includes a method for the System Engineer to obtain the SWEC File Number for equipment and instrumentation manufacturer drawings.
Section II of 2BVM-6 (Manufacturer's Drawing Handling System
- MDMS) provides references to the MDHS reports which are available for assisting in locating SWEC File Numbers. Depending upon the information available on the component (i.e., mark number, PO number, etc.),
the j
System Engineer may locate the SWEC Manufacturer's Drawing File Number by using the appropriate MDHS report and the description provided in 2BVM-6 on the information included in the-SWEC File Number. For example:
1.
With the SWEC equipment Mark Number, use MDHS report 3.7.1 (ID cross-reference report) 2.
With the account number (Attachment 2BVM-6A) and PO number (2BVM-1), use MDHS report 3.2.2 (by SWEC File Number).
To obtain manual and motor-operated valve weights:
1.
List the valve mark number and size as it appears on the flow diagram (e.g., 3/4" VBS015-D-4),
2.
Refer to 2BVM-81 for the SWEC File No. for each valve mark number and size, j
3.
Refer to MDHS ReportJ3.. 2 to find the latest SWEC revision and 2
approval status.
- 4. - Find the valve weignt on the drawing (valve body weight, actuator weight, assembly weight all included).
All Restriction Orifices (RO),
Flow Elements (FE), Expansion Joints (EJ), and Flexible Hoses must be identified by mark number to ensure l
that the proper qualification reports and calculations are performed.
For Ros and FEs, a reference to the orifice specification for additional data must be included.
u
__.__.-.___._,-.,,____.-.__.___._-,_,_.__,,__.,_,.._,_.___.,,..__._m
t 15 5.2.2 Review and Approval of Stress Analysis Data Package The System Engineer shall have the completed stress analysis data
. package typed by ATMS, and checked by a checker and the cognizant Principal Engineer.
The checker shall be a technically qualified individual other than the preparer. Also, an independent review of the preparation methods and data in accordance with 2BVM-45 shall be performed by a Stress Analysis Data coordinator.
- Then, the System i
Engineer and Stress Analyst shall meet to review the inputs and identify and discuss the revisions before the package is approved. This finished package shall be reviewed by the Lead Power and Lead EMD Engineers and approved by the Project Engineer before being transmitted to the Power Division Pipe Stress Coordinator for processing in accordance with Section 8.0 of this procedure. QA Category I packages shall be marked NUCLEAR SAFETY RELATED. The cover sheet shown as Attachment 2BVM-45Al should be used.
5.2.3 Revisions to S. tress Analysis Data Package The System Engineer will be responsible for assuring that the data included in the Stress Analysis Data Package is both correct and current.
Any revisions to the system information must be documented by 3
revision of the Design Specification for Piping Engineering and Design.
To revise a package, the System Engineer shall fill out the Stress Analysis Data Package Revision Change / Reason Sheet (Attachment 2BVM-45A2), make the necessary changes to the data package, and submit it with a new cover sheet (Attachment 2BVM-45A1) for approval in 1
accordance with Section 5.2.2.
The revision may be in a replacement page format, or may be issued in its entirety as a complete revision.
Note: Each item changed by the revision shall be keyed by a line in the margin and the revision number in a triangle next to the change.
6.0 PREPARATION OF INFORMATION REQUIRED FOR NON-ASME III (CLASS 4)
SYSTEMS i
6.1 General The design of Class 4 systems is based primarily on the condition of the system or portion of a system in response to Normal and Upset Plant Operating Conditions. The System Engineer is responsible for specifying the Maximum Operating Pressure and the Maximum Temperature Range (maximum operating and minimum operating temperatures) which exist during Normal Operation.
The Stress Analyst is responsible for evaluating the stresses in the piping system based on these parameters, and is also responsible for evaluating seismic loadings on those systems designated Seismic Category II.
Note: SWEC Class 4 (Seismic Category II) systems shall be analyzed in accordance with Section 3.2.4.
N
....-.,_.-__v..-
3 16 The System Engineer is also responsible for identifying and specifying upset system conditions which exceed normal conditions.
6.2 Procedure 9
4 (6.2.1 Preparation of Stress Analysis Data Package
,The System Engineer is responsible for compiling a Stress Analysis Data
(\\ Package consisting of the information required by the Stress Analyst to
complete the stress analysis of each Class 4 system.
The Stress Analysis Data Package shall consist of two parts: 1) Specification of System conditions and 2) Stress Analysis Information.
6.2.1.1 Specification of System conditions
. I The System Engineer shall specify the system conditions using:
1 1.
An appropriate number of System condition Description Sheets (Attachment 2BVM-45A4) 2.
An appropriate number.
of Pipe Stress Data Tables (Attachment 2BVM-45A9).
[
The System Engineer must specify all normal and upset (abnormal) system conditions including Hydrostatic test (with temperature specified),
System test, and Containment Pressure test. The Upset Plant Conditions described in Appendix I should be used as a guide to identify all the system conditions which could occur.
Each System Condition shall be described in the same manner as ASME, Class 2 or 3 systems (see section 5.2.1.1),
with references, operating components identified, the l
anticipated total number of temperature cycles, and a flow path sketch ifI necessary. A brief system description of the function of the system is(to be included under General Notes (Attachment 2BVM-45A3).
A g
Reference Operating Condition muss also be identified. The System Engin,eer is not required to match the specific class 4 System conditions by n, umber in the Plant Condition / System Condition Matrix, unless the class 4 piping is directly connected to ASME III piping and therefore becomes subject to the requirements of the ASME III piping (See section 3.2.4).
However, the frequency of operation shall be indicated to describe if the condition is normal - sustained, occasional - 1 percent, or occasional -10 pere'ent as defined below:
Normal - sustained:. loads where the duration exceeds 10 percent of the operating period.
's occasional 1 percent loads whose duration does not exceed 1 percent of the operating period.
Occasional 10 percent: loads whose duration does not exceed 10 percent of the operating period.
After all System Conditions have been specified, the System Engineer must complete the Pipe. Stress Data Table for all Class 4 lines in the
17 same manner as an ASME, Class 2 or 3 system (see section 5.2.1.1).
Identification of high energy lines, consecutive numbering of all pages including attachments, and a table of contents shall be included in accordance with Section 5.2.1.1.
The Design Pressure and Design Temperature shall be' derived by selecting at least the maximum pressure and temperature that can exist during Normal and Upset plant conditions for each line for all system conditions specified on the System condition Description Sheets.
-These i
prersures and temperatures shall be entered in the columns designated
" Design Pressure" and " Design Temperature" in the Line Designation Table (LDT).
6.2.1.2 Flow Transient Data Same as Section 5.2.1.2 6.2.1.3 Safety / Relief Valve Information Same as Section 5.2.1.3.
6.2.1.4 Input References Same as Section 5.2.1.4'.
6.2.1.5 Miscellaneous Information Same as Section 5.2.1.5.
6.2.2 Review and Approval of Stress Analysis Data Package The System Engineer shall have the completed stress analysis data package typed by ATMS and checked by a checker and the cognizant Principal Engineer.
The-checker shall be a technically qualified individual other than the preparer. Also, an independent review of the preparation methods and data in accordance with 2BVM-45 shall be i
i performed by a Stress Analysis Data Coordinator. Then the package shall l
be reviewed and signed by _ the Lead Power Engineer before being transmitted.to the Power Division Pipe Stress Coordinator for processing in accordance with Section 8.0 of this procedure. The cover sheet shown I
as Attachment 2BVM-45A8 should be used.
For some systems, the review and approval procedure applicable to ASME III systems as described in Section 5.2.2 may also be followed.
6.2.3 Revisions to Stress Analysis Data Package The System Engineer will be responsible for assuring that the data included in the Stress Analysis Data Package is both correct and current.
The Design Specification for Piping Engineering and Design will be updated periodically as required to reflect these revisions. To revise a package, the System Engineer shall fill out the Stress Analysis
-Data Package Revision Change / Reason Sheet (Attachment 2BVM-45A2), make the necessary changes to the data package, and submit it with a new 4
N
18
- cover sheet (Attachment 2BVM-4SA8) for approval in accordance with section 6.2.2.
The revision may be in a replacement page format, or may be issued in its entirety as a complete revision.
NOTE: Each item changed by the revision shall be keyed by a line in the margin and the revision number in a triangle next to the change.
7.0 PROCEDURE FOR PREPARATION OF LINE DESIGNATION TABLE (LDT) i 7.1 General The purpose of the LDT is to compile pertinent data for each line depicted on BVPS-2 flow diagrams in one document.
This document will be used as the reference by both Headquarters and Site personnel in both engineering and construction.
The LDT is issued by systems according to the system designation codes.
All the lines.with the same three letter system designation code are issued as one LDT with a cover sheet as shown in Attachment 2BVM-45A12.
The LDT for each system shall be updated as required.
7.2 Preparation Line Designation Tables shall be prepared for all flow diagrams with the exception of facilities ventilation schematics.
Input to the computerized LDT data base shall be prepared using BVM-45A11. The input sheet information should be arranged in sequential order by RM or RB, system designation, and line number in
'that order.
The' field justification for each input field is indicated by the asterisks in the first line of the input data in Attachment 2BVM-45A11.
i (e.g.
" pipe schedule" and " insulation type" are left-justified, as indicated by the asterisks in the first column of the field.
For the right-justified fields, the input for the field ends at the last column l-of the field.)
i Data from a field for any 'line number can be deleted by placing an j
asterisk (*) in the field to be deleted.
i Instructions ^for compiling the data are listed below Note: It is the System Engineer's responsibility to assure that the i
data common to both the LDT, and the Stress Analysis Data Package are consistent.
l r
6 a
~
19 Column 1 " Revision Symbol" This column should only be filled in with a minus sign (-) when an entire line is to be deleted. The addition sign (+) when an entire line is added, or an asterisk (*) when a line's information is revised is not required as input since these symbols are printed automatically in the output as part of the update program.
Column 2 " Unit No" A No. 2 will be placed in this column indicating BVPS Unit No. 2.
Columns 3 through 5 " System Designation" These columns will contain the system designation as established by 2BVM-146.
Columns 6 through 8 "Line size" All' line sizes will be represented by three digits as defined in 2BVM-12.
Columns 9 through 11 "Line Number" Line numbers will also be expressed by three digits (001-999), or by a letter followed by two digits (A01-Z99) when the number of lines requires it.
Once deleted, a line number will not be reused.
Columns 12 and 13 " Code Class" These columns will have one of four numbers in them:
01, 02, 03 04.
Nos. 01 through 03 represent the ASME III Class 1, 2 or 3.
No.
04 indicates that the line is in accordance with ANSI B31.1 code.
Columns 14 through 21 " Fluid" These columns are used.to describe the media in the line; i.e.,
air, water, steam, etc.
Columns 22 through 33 "From" These columns are used to describe the origin of this line in relation to other lines, valves, or equipment.
Columns 34 through 45 "To" These columns are used to describe the termination of this line in relation to other lines, valves, or equipment.
d 20 Columns 46 through 49 "Line Class" These columns are for the Stone & Webster line class (e.g., 152, 302). Definitions are in the piping engineering and design specification.
Columns 50, 51, and 52 " Pipe Schedule" These columns are used to. designate pipe schedule, where applicable.
These are also obtained from the Pipe Class Specification 2BVS-939A.
Columns 53 through 61 " Operating" These columns will designate the applicable stress analysis data package series, e.g., SI-RM(RB)-Number.
Since the stress analysis data packages describe several different operating conditions, and the range of values is a more accurate portrayal of operating temperatures and pressures, the stress analysis data packages are a more definitive source than the LDT's.
Columns 62 through 70 " Design" These columns are for the design pressure, and temperature in psig, and degrees F,
respectivaly.
For ASME III systems refer to section 3.1.2.4 for definition of Design Conditions.
For non-ASME III systems, the Design Pressure and Temperature of a system or portion of a system are derived from the maximum pressures and temperatures experienced by the system or portion of a system during the Normal and Upset Plant Operating Conditions.
Columns 71 through 74 " System Release No."
These columns are to identify the system release number for each i
system and its associated hydrostatic pressure testing boundary diagrams.
Column 75 " Test Media" This column will show the media with which this line is tested; i.e., "A" for air, "W"
for' water, etc.
Columns 76 and 77 " Insulation Thickness" These columns will show the number of inches of insulation, if required, that will be put on the line.
Insulation requirement will be obtained from the insulation specification number 2BVS-60.
o m
21 Columns 78 and 79'" Insulation Type" These columns will show the type of insulation (or no insulation) l that will be put on the line.
Insulation types will be obtained from the insulation specification (2BVS-60). Letter "S" will be used if there.is no insulation requirement.
Column 80 " Quality Assurance Category" 4
This column is to indicate quality assurance category of the line.
(See PTP-0.4.1).
Revising Information If any questions should arise that cannot be answered, please contact the LDT-Coordinator for BVPS Unit 2.
j 7.3-Review and Approval of LDT The LDT's for more than one system may be issued / revised and distributed at the same time. The LDT coordinator shall compile the latest LDT's after receiving them from the computer department and give them to the appropriate system engineer for review.
For each LDT to be issued or revised,.the system engineer shall review the output and sign the cover sheet of the LDT under his responsibility in the space provided.
The LDT's are then reviewed and signed by the cognizant Principal Engineer and the Lead Test Engineer. The Lead Power Engineer shall review and approve the completed LDT package for distribution, by signing the cover sheet of each LDT.
7.4 Distribution Distribution of the Line Designation Tables will be in accordance with 2BVM Schedules, Reports, and Controlled Document Indices.
j 8.0 PROCEDURE FOR PROCESSING PIPE STRESS DATA 8.1 General This section outlines major individual responsibilities in handling Pipe Stress Data for BVPS-2.
8.2 System Engineer Develops and Maintains Stress Information Data
-8.2.1 Develop and Maintain Stress Analysis Data Package The System Eng'ineer is responsible for maintaining the system stress information up to date. A Stress Analysis Data Package for each system shall be processed by the System Engineer (approved as indicated on BVM-45A1 or 2BVM-45A8) and sent to the Power Division Pipe l
Stress Coordinator (PSC) in accordance with Sections 5.0, or 6.0 of this
. procedure, as applicable. The Stress Analysis Data Package will be i
e I
o
i 22
- specifically referenced in the ASME III Specification for Piping Engineering and Design and become part thereof.
8.3 Pipe Stress Coordinator (PSC) Processes and Transmits Stress Analysis Data Package to Pipe Stress and Supports Engineering (PSAS)
The PSC will be responsible for maintaining a file (Job Book 211-SI) of all Stress Analysis Data Packages (including all revisions) transmitted in accordance with Sections 5.0 and 6.0 of this procedure.
The Job Book will be microfilmed for future traceability and record retention.
In addition the PSC will maintain an index of all Stress Analysis Data Packages in Job Book 211-SI, and distribute the index to the Lead EMD Engineer, as required.
l Instruction for piping drawing and E&DCR handling and review will be in accordance with 2BVM-13 and 2BVM-94, respectively, and Job Book filing shall be in accordance with 2BVM-2.
8.4 PSAS Completes Stress Analysis PSAS.will. be responsible for completing the stress analysis for piping systems in accordance with the methods described in 2BVS-939.
The Stress Analysis results will be documented on the following forms.
1.
Stress calculations (See Job Book 211-NP) 2.
Stress reports (SSR, for Class 1 piping only, see Job Book 227-3.4) 3.
Nozzle load summaries (NL, for equipment as required, see Job Book 211-NP) l
- 8. 5' Supplemental Action By the System Engineer /PSC The System Engineer shall transmit when appropriate the completed nozzle load summary to the applicable equipment manufacturer. A copy of this transmittal must be sent to the Pipe Stress Job Book 227-4 and a copy to the EMD Lead Engineer.
It is the responsibility of the System Engineer to make sure that a response is made for all data transmitted to the manufacturers.
I t
U
i l
PLANT. CONDITION l C N
g.
l PLANT OPERATING CONDITION PLANT TEST CONDITION i
NORMAL UPSET EMERGENCY FAULTED i
i 1
i 2BVM-45 FIGURE I RELATIONSHIP OF
)
VARIOUS PLANT CONDITIONS
PL ANT OPERATING CONDITION PLANT TEST CONDITION INORMAL, UPSET, EMERSENCV,FAULTEDI SYSTEM CONDITION s
- C n
i i
r 3
r SYSTEM OPERATING SYSTEM TEST CONDITION SYSTEM TEST CONDITION CONDITION DURING SYSTEM TEST DURING PLANT TEST i NORMAL PL ANT OPERATING CONDITIONI CONDITION i
4 i
SYSTEM DESIGN CONDITION
)
(NORMALLY, Tite MAX. TEMP. Att0
}
PHESS EXISTING IN SYSTEM OR i
j PONTION OF SYSTEM let A SYSTEM i
OPEh ATING CONosTION ASSOCIATED waTit THE NORMAL OR UPSET PLANT OPERATING CON 0sTIONI 28vM-45 FIGURE 2 q
RELATIONSHIP ALONG PLANT CONDITIONS i
FOR VARIOUS SYSTEM CONDITIONS i
)
i
APPENDIX I j
i j
PLANT CONDITIONS USED AS A BASIS FOR BVPS-2, ASME III, PIPING SYSTEM DESIGN Listed -in Sections I and II below are the Plant Operating conditions (categorized as Normal,
- Upset, Emergency, and
^
Faulted), and thelPlant Test Conditions to be used for the design of piping systems for BVPS-2.
The definitions of these conditions are given in Section III of this Appendix.
The postulated Plant Conditions a unit could experience over its operational life can be infinite. Because of this, the below listed Plant Conditions have been selected, based on judgment and
. experience, to be sufficiently severe and diverse to provide an adequately conservative design basis which will envelope all credible Plant Conditions.
I.
PLANT OPERATING CONDITIONS A.
Normal Plant Operating Conditions 1.
Reactor Coolant System heatup at 100*F/hr i
2.
Reactor Coolant System cooldown at 100*F/hr 3.
Unit loading at 5 percent of full power per minute 4.
Unit unloading at 5 percent of full power per minute
.5.
Step load increase of 10 percent of full power 6.
Step load decrease at 10 percent of full power 7.
Large step load decrease with steam dump 8.
Steady state fluctuations 9.
Feedwater cycling at hot shutdown 10.
Loop out of service - normal loop shutdown 11.
Loop out of service - normal loop startup 12.. Loop out of service - steady state 13.
Unit loading between 0 and 15 percent of full power 14.
Unit unloading between 0 and 15 percent of full power 15.
Boron concentration equalization I
I-1 o
._.--..-....._....__,,.....___,..,...,_,_,_.,___.,._..,_,,,__._..m.,_.__.___..._,.,
__-__,,,...__.,_...-.m_..
16.
Refualing B.
Upset Plant Operating Conditions 1.
Loss of load (without immediate reactor trip) 2.
Loss of power 3.
Partial loss of flow 4.
Reactor trip from full power - no cooldown 5.
Reactor trip from full power - cooldown, no safety injection 6.
Reactor trip from full power - cooldown with safety injection 7.
Inadvertent RCS depressurization 8.
Inadvertent startup of an inactiva loop 9.
Control rod drop 10.
Inadvertent safety injection actuation 11.
Inadvertent rod control cluster assembly withdrawal 12.
Load transients initiated by spurious operation of active elements
'o f steam and power conversion system 13.
Loss of normal feedwater 14.
Loss of condenser cooling 15.
Single failure in the electrical / control systems makeup is provided by charging 16.
Minor RCS leak pumps
'17.
Single error of operator 18.
Single failure of control component 19.
Minor secondary system leaking which would not
- prevent orderly shutdown and cooldown assuming normal makeup.
C.
Emergency Plant Operating Conditions 1.
Small loss of coolant accident makeup rate greater than charging pump capacity I-2 o
2._
S2all st:am lina brsck - msk:up rato grooter than normal makeup capacity D.
Faulted Pl' ant Operating Condition 1.
Double-ended rupture of large reactor coolant pipe 2.
Large steam line break 3.
Feedwater line break 4.
Reactor coolant pump locked rotor 5.
Control rod ejection 6.
Steam generator tube rupture II. PLANT TEST CONDITIONS 1.
Turbine Roll Test 2.
Primary System Hydrostatic Test 3.
Secondary System Hydrostatic Test 4.
Primary System Leakage Test o
5.
Secondary System Leakage Test 6.
Steam Generator Tube Leakage Test 7.
Containment Pressure Test I-3 s,
III.
DESCRIPTION OF CONDITIONS The specific System Conditions associated with the Reactor Coolant System for the Plant Conditions defined below are included in 2BVM-121.
A.
Normal Plant Operating Conditions 1,2.
Heatup and Cooldown at 100*F/hr The design heatup and cooldown cases are conservatively represented by continuous operations performed at a unifonn temperature rate of 100*F per hour.
(These operations can take place at lower rates approaching the minimum of 0*F per hour.)
The expected normal rates are 50*F/hr.
For these cases, the heatup occurs from ambient (assumed to be 120*F) to the no-load temperature and pressure condition and the cooldown represents the reverse situation. Reactor coolant system temperature can be as low as 70*F if the system is depressurized.
In actual practice, the rate of temperature change of 100*F per hour will not be attained because of other limitations such as:
a.
Material ductility considerations which establish maximum permissible temperature rates of change, as a function of plant pressure and temperature, which are below the design rate of 100*F per hour.
b.
Slower initial heatup rates when using pump energy only.
c.
Interruptions in the heatup and cooldown cycles due to such factors as drawing a pressurizer steam bubble, rod withdrawal,
- sampling, water chemistry, and gas adjus tments.
The number of such complete heatup and cooldown operations la specified as 200 each, which corresponds to five such occurrences per year for the 40-year plant design life.
3,4. Unit Loading and Unloading at 5 Percent of Full Power / Min.
The unit loading and unloading cases are conservatively represented by a continuous and uniform ramp power change of 5 percent per minute between 15' percent load and full load.
This load swing is the maximum possible consistent with operation under automatic reactor control.
(It should be noted that in actual practice, changes in power level may take place at rates less than 5 percent per minute. This would be due to turbine loading limitations at low power levels.) The reactor temperature will vary with load as prescribed by the Reactor Control System.
The number of loading and unloading operations is defined as 18,300.
One loading operation per day yields 14,600 such operations during the 40-year design life of the plant.
I-4 u
_-... = -
5,6. Stcp Lecd Incratsa and D2cressa of 10 Percent of Full Power
- The 110 percent step change in load demand is a transient which is assumed to be a change in turbine, control valve opening due to disturbances in the electrical network into which the plant output is tied.
The reactor control system is designed to restore plant equilibrium without reactor trip following a 110 percent step change in turbine load demand initiated from nuclear plant equilibrium conditions in the range between 15 percent and 100 percent full load, the power range for automatic reactor control.
In effect, during load change conditions, the reactor control system attempts to match turbine and reactor outputs in such a manner that peak reactor coolant temperature is minimized and reactor coolant temperature is restored to its programmed set point at a sufficiently slow rate to prevent excessive pressurizer pressure decrease.
Following a step decrease in turbine load, the secondary side steam pressure and temperature initially increase since the decrease in nuclear power lags behind the step decrease in turbine load. During the same increment of time, the reactor coolant system average temperature and pressurizer pressure also initially increase.
Because of the power mismatch between the turoine and reactor and the increase in reactor coolant temperature, the control system automatically inserts the contrcl rods to reduce core power. With the load decrease, the reactor coolant temperature will ultimately be reduced from its peak value to a value below its initial equilibrium value at the inception of the transient.
The reactor coolant average i
temperature set point change is made as a function of turbine-generator load as determined by first stage turbine pressure measurement.
The pressurizer pressure will also decrease from its peak pressure value and follow the reactor coolant decreasing temperature trend. At some point during the decreasing pressure transient, the saturated water in the pressurizer begins to flash which reduces the rate of pressure decrease. Subsequently, the pressurizer heaters come on to restore the plant pressure to its normal value.
Following a step increase in turbine load, the reverse situation occurs, i.e., the secondary side steam pressure and temperature initially decrease and the reactor coolant average temperature and pressure initially decrease. The control system aut'omatically withdraws the control rods to increase core power.
The decreasing presture transient is reversed by actuation of the pressurizer heaters and eventually the system pressure is restored to its normal value.
The reactor coolant average temperat,ure will be raised to a value above its initial equilibrium value at the beginning of the transient.
The number of each operation is specified at 2,000 times or 50 per year for the 40-year plant design life' The 10 percent step load increase transient can be initiated at any power level between 15 percent and 90 percent of full load.
The 10 percent I-5 O
step load decrease transient can be initiated at any power level between 25 percent and 100 percent of full load.
- 7. Large Step Load Decrease With Steam Dump This transient applies to a step decrease in turbine load from full power, of such magnitude that the resultant rapid -increase i-in reactor coolant average temperature and secondary side steam pressure and temperature will automatically initiate a secondary side steam dump that will prevent both reactor trip and lifting of steam generator safety valves. Thus, when a plant is designed to accept a step decrease of 95 percent from full power (complete loss of outside load but retaining the plant auxiliary load of
. 5 percent),
the steam dump system provides the heat sink to accept 85 percent of the turbins load. The remaining 10 percent of the total step change is assumed by the reactor control system (control rods).
If a steam dump system was not provided to cope with this transient, there would be such a strong mismatch between what the turbine is asking for and what the reactor is delivering that a reactor trip and lifting of steam generator safety valves would occur. BVPS-2 is designed to accept step changes of 100 percent load rejection (with 85 percent steam dump capability).
The number of occurrences of this transient is specified at 200 times or 5 per year for the 40-year plant design life.
4
- 8. Steady State Fluctuations It is assumed that the reactor coolant temperature and pressure at any point in the system vary around the nominal (ste dy state) values.
These local variations can occur at many frequencies, except for design purposes two cases should be considered:
a.
Initial Flactuations These are due to control rod cycling during the first 20 full power months of reactor operation.
Temperature is assumed to vary by 13*F and pressure by 125 psi, once during each 2 min period. The total number of occurrences is limited to 1.5 x 105 These fluctuations are assumed to occur consecutively, and not simultaneously with the random fluctuations.
4 b.
Random Fluctuations - Temperature is assumed to vary by 20.5'F and pressure by i6 psi, once every 6 min. With a l
6 min' period, the total number of occurrences during the plant design life does not exceed 3.0 x 108
- 9. Feedwater Cycling at Hot Shutdown I
These transients can occur when the plant is at "no load" j
conditions, during which intermittent (slug) feeding of 30*F feedwater into the steam generators is assumed.
Due to l
fluctuations arising from this mode of operation, the reactor i
j coolant average temperature decreases at a rate of approximately i
t I-6 u
..-.s.
- m..
.. ~ -.
2*F/ain far o total d:croaso of 24'F.
Tha tcmparature knmediately begins to retu"n to normal no-load temperature at a hestup rate of approximately 0.7*F/ min.
This transient is assumed to occur 2000 times over the life of the plant.
10,11,12. Loop out of Service The plant may be operated at a reduced power level with a single loop out of service for limited periods of time.
This is accomplished by reducing power level and tripping a single RC pump (as opposed to tripping a pump at full power).
It 'is assumed that this transient occurs twice per year or 80 times in the life of the plant.
Conservatively, it is assumed that all 80 occurrences can occur in the same loop.
In other words, it must be assumed that the whole RCS is subjected to 80 transients while each loop is also subjected to 80 inactive loop transients.
When an inactive loop is brought back into service, the power level is reduced to approximately 10 percent and the pump is started.
It is assumed that an inactive loop is inadvertently started up at maximum allowable power level 10 times over the life of the plant.
(This transient is covered under Upset Plant Conditions, Inadvertent Startup of an Inactive Loop.) Thus, the normal startup of an inactive loop is assumed to occur 70 times during the life of the plant.
13,14. Unit Loading and Unloading Between 0 and 15 Percent of Full Power The unit leading and unloading cases between zero and 15 percent power are represented by continuous and uniform ramp power
- changes, requiring 30 min for loading and 5 min for unloading.
During loading, reactor coolant temperatures are increased from the no-load value to the normal load program temperatures at the 15 percent power level. The reverse temperature change occurs during unloading.
Prior to loading, it is assumed that the plant is at hot shutdown conditions, with 70*F feedwater cycling.
During the two hour l
period fol. lowing the beginning of loading, the feedwater temperature increases from 70*F to approximately 300*F due to l
steam dump and turbine startup heat input to the feedwater.
Subsequent to unloading, feedwater heating is terminated, steam dump is reduced to residual heat removal requirements, and feedwater temperature decays from approximately 300*F to 70'F.
l The number of these loading and unloading transients is assumed to be 500 each during the 40-year plant design life, which is equivalent to about one occurence per month.
I-7
- 15. Boron concentration Equalization Following any large change in boron concentration in the RCS, spray is initiated in order to equalize concentration between the loops and the pressurizer.
This can be done by manually operating the pressurizer backup heaters, thus causing a pressure increase, which will initiate spray at a compensated pressurizer pressure of approximately 2275 psia.
The proportional sprays return the pressure to 2250 psia and maintain this pressure by matching the heat input from the backup heater until the concentration is equalized. For design purposes, it is assumed that this operation is performed once after each load change in the design load follow cycle. With two load changes per day and i
a 90 percent plant availability factor over the 40-year design life, the total number of occurrences is 26,400.
The only effects of these operations on the primary system are that a.
the reactor coolant pressure varies in step with the pressurizer pressure, and b.
the pressurizer surge line nozzle at the hot leg will l
experience the temperature shocks associated with outflow from the pressurizer.
j
- 16. Refueling At the end of plant cooldown, the fluid in the reactor coolant system is at 140*F.
At this time, the vessel head is removed and the refueling canal is filled. This is done by pumping water from the refueling water storage tank, which is outside and conservatively assumed to be at 32*F, into the loops by means of the low head safety injection pumps.
It should be conservatively assumed that the cold water flows directly into the vessel and that all the fluid in the RCS is replaced with the colder water within 10 min.
This operation is assumed to occur twice per year for 80 times over the life of the plant.
i
- 8. Upset Conditions
- 1. Loss of Load (Without Immediate Reactor Trip)
This transient applies to a step decrease in turbine load from full power (turbine trip) without immediately initiating a reactor trip, and represents the most severe pressure transient on the reactor coolant system under upset conditions.
The reactor eventually trips as a consequence of a high pressurizer level trip initiated by the reactor protection system.
Since redundant means of tripping the reactor are provided as a part of the reactor protection system, transients of this nature are not t
~
expected but are included to ensure a conservative design.
r I-8 u
--- - = --...
- Th2 numbar of cccurrcnces of this transient is spscified at 80 times, or 2 times per year.for the 40-year plant design life.
~
2.
Loss of Power This transient applies to a blackout situation involving the loss of outside electrical power to the station, assumed to be operating initially at 100 percent power, followed by reactor and turbine trips. The reactor coolant pumps are deenergized, as are all electrical loads connected to the turbine-generator bus, including the main feedwater and condensate pumps.
Following coastdown of the reactor coolant pumps, natural circulation builds up in the system to some equilibrium value.
This condition permits removal of core residual heat through the steam generators which by this time are receiving feedwater, assumed to be at 32*F, from the auxiliary feedwater system operating from diesel generator power.
For equipment design purposes, it is f
conservatively assumed that the single turbine driven auxiliary l
feedwater pump, equal in capacity to the two motor driven pumps, j
also operates.
Steam is removed for reactor cooldown through atmospheric relief valves provided for this purpose.
i Tne number of occurrences of this transient is specified at 40 times or one per year for the 40-year plant design life.
- 3. Partial Loss of Flow This transient applies to a partial loss of flow from full power, in which a reactor coolant pump is tripped out of service as the result of a loss of power to that pump. The consequences of such an accident are a reactor and turbine trip, on low reactor coolant flow, followed by automatic opening of the steam dump system and flow reversal in the affected loop. The flow reversal i
causes reactor coolant at cold leg temperature to pass through the steam generator and be cooled still further.
This cooler water then flows through the hot leg piping and enters the i
reactor vessel outlet nozzles.
The net result of the flow reversal is a sizeable reduction in the hot leg coolant temperature of the affected loop.
The number of occurrences of this transient is specified at 80 times or 2 times per year for the 40-year plant design life.
4.5,6. Reactor Trip From Full Powco i
A reactor trip from full power may occur from a variety of causes resulting in temperature and pressure transients in the reactor coolant system and in the secondary side of the steam generator.
This is the result of continued heat transfer from the reactor coolant in the steam generator. The transient continues until the reactor coolant and steam generator secondary side i
temperatures are in equilibrium at zero power conditions.
A continued supply of feedwater and controlled dumping of steam remove the core residual heat and prevent the steam generator
)
i I-9 N
..~-,--__.---,..m
._-_.--__.-,_-m.
-..m _ _.-_,,
.____.,,------_.m.-.,-----
safoty valvas from lifting. The reactor coolant temperature and pressure undergo a rapid decrease from full power values as the reactor protection. system causes the control rods to move into the core.
Various moderator cooldown transients associated with reactor trips can occur as a result of excessive feed or steam dump after trip or large load increase.
For design purposes, reactor trip is assumed to occur a total of 400 times or 10 times per year over the life of the plant. The various types of trips and the number of occurrences for each are as follows:
Condition 4 - Reactor trip with no inadvertent cooldown -
230 occurrences Condition 5 - Reactor trip with cooldown but no safety injection - 160 occurrences
~
Condition 6 - Reactor trip with cooldown actuating safety injection - 10 occurrences.
- 7. Inadvertent Reactor Coolant System Depressurization several events can be postulated as occurring during normal plant operation which will cause rapid depressurization of the reactor coolant system. These includes a.
Actuation of a single pressurizer safety valve.
b.
Inadvertent opening of one pressurizer power operated relief valve, due either to equipment malfunction or operator error.
c.
Malfunction of a single pressurizer pressure controller causing one power operated relief valve and two pressurizer spray valves to open.
l d.
Inadvertent opening of one pressurizer spray valve, due either to equipment malfunction or operator error, e.
Inadvertent auxiliary spray.
of these events, the pressurizer safety valve actuation causes l
the most severe transients, and'is used as an " umbrella" case to conservatively represent the reactor coolant pressure and temperature variations arising from any of them.
(Item e, inadvertent auxiliary spray, imposes especially severe temperature shock on the pressurizer spray nozzle and pressurizer i
vessel interior.
This case is treated separately as a Class I transient
.'m it7M-121 for purposes of analyzing the spray nozzle and press u.x t - vessel.)
l I-10 l
...~,,___..,_~~m_.--
_,,,_.-_,,,,,_-.--m
-.,m
1 Wh:n o prescurizar safety valvo cp:ns, cnd rcmsins cpsn, th2 system rapidly depressurizes, the reactor trips, and the safety injection system is actuated. Also, the passive accumulators of
- the SIS are actuated when RCS pressure decreases by approximately 1600 psi, about 12 min after the depressurization begins. The depressurization and cooldown are eventually terminated by operator action.
All of these effects are completed within approximately 18 min.
It is conservatively assumed that none of the pressurizer heaters are energized.
With pressure constant and safety injection in operation, boiloff of hot leg liquid through the pressurizer and open safety valve will continue.
For design purposes, this transient is assumed to occur 20 times during the 40-year design life of the plant.
- 8. Inadvertent Startup of an Inactive Loop This transient can occur when a loop is out of service. With the plant operating at maximum allowable power level the reactor coolant pump in the inactive loop is started as a result of i
operator error. Reactor trip occurs on high nuclear flux.
This transient is assumed to occur 10 times during the life of the plant.
- 9. Control Rod Drop This transient occurs if a bank of control rods (worth I percent reactivity) drops into the fully inserted position due to a j
single component failure. The reactor is tripped on either low pressurizer pressure or negative flux rate, depending on time in core life and magnitude of the reactivity insertion.
It is assumed that this transient occurs 80 times over the life of the plant.
i
- 10. Inadvertent Safety Injection Actuation A spurious safety injection signal results in an immediate i
reactor trip followed by actuation of the high head centrifugal i
charging pumps.
These pumps deliver borated water from the i
refueling water storage tank to the RCS cold legs. The initial portion of this transient is similar to the reactor trip from full power with no cooldown.
Controlled steam dump and feedwater i
flow after trip removes core residual heat.
Reactor coolant temperature and pressure decrease as the control rods move into the core.
i Later in the transient, the injected water causes the RCS pressure to increase to the pressurizer power operated relief l
valve set point and the primary and secondary temperatures to j
decrease gradually. The transient continues until the operator stops the charging pumps.
It is assumed that the plant is then i
i I-11
_ ~. _
- _ _ _... - - -. _.. - -. - ~.. -. =.
returned to no-load conditions, with pressure and temperature changes controlled within normal limits.
For design purposes this transient is assumed to occur 60 times during the 40-year design life of the plant.
t
- 11. Inadvertent Rod Control Cluster Assembly Withdrawal I
An inadvertent control rod withdrawal due to operator error, equipment failure, etc, causes a sudden reactivity increase. RCS pressure / temperature increases with a reactor trip occurring on high neutron flux or high T across the core.
Following reactor i
i trip, RCS heat is removed via the steam generators by the addition of feedwater and controlled steam dump to the condenser.
This transient does not initiate safety injection and controlled heat removal via the steam generators returns RCS parameters to normal hot standby values.
Cooldown of the RCS continues i
according to normal operating procedures as described in Normal Plant Operating Condition No. 2 - RCS cooldown at 100*F/hr.
- 12. Load Transients Initiated by Spurious operation of I
Active Elements of Steam and Power Conversion System System transients may occur due to Upset Plant Conditions caused by events such as individual component failures, multiple pump operation, incorrect control due to inaccurate sensing of fluid system parameters, system misalignments, etc. Each system must i
be evaluated by the the System Engineer for any potential l
transients not covered by other Upset Plant Conditions.
- 13. Loss of Normal Feedwater i
A loss of normal feedwater (from pump failures, valve malfunctions, or loss of ac power) results in a reduction in capability of the secondary system to remove heat generated in I
the RCS. Reactor trip is initiated on either low-low water level in the steam generators or by a steam flow / feed flow mismatch in coincidence with low SG water level. This transient assumes that all normal power supplies are available (if normal power is lost, l
Upset Plant Condition No. 2 Loss of Offsite Power wculd envelope this condition).
The reactor coolant pumps continue to l
operate following reactor trip.
Auxiliary feedwater is automatically initiated to the steam generators to remove RCS heat and maintain hot shutd6wn conditions.
Steam dump is directed to the condenser through control of the condenser dump valves.
Cooldown of the RCS continues in this mode for i
approximately 4 hr until the residual heat removal system (RHS)
)
is initiated at RCS temperature of 350*F. The RHS removes heat until cold shutdown is reached approximately 24 hr after reactor i,
trip.
i i
I-12 i
u
- 14. L*ss af Cindensar C aling A loss of circulating water to the condenser results in a loss of vacuum in the condenser. Turbine trip occurs due to a loss of vacuum'.
All normal and emergency power supplies are assumed available during this transient.
Therefore, main feedwater continues to the steam generators for RCS heat removal. With the condenser unavailable, main steam is dumped to the atmosphere through the steam dump valves. Following reactor trip, cooldown of the RCS continues using main feedwater and steam dump until the residual heat removal system is started at RCS temperature of 350*F.
- 15. Single Failure in Electrical / Control Systems For determining fluid system transients, failures in electrical / control systems consider the following cases:
a.
Loss of one Class 1E bus (orange or purple) b.
Loss of one Class 1E vital channel (red, white, blue, or yellow) c.
Loss of one of two essential buses The loss of normal (black) power is not included in these transients since this plant condition is covered by the " Loss of offsite Power," Upset Plant Condition No.*2.
The Class 1E and essential buses and associated control systems are powered from the _ emergency diesel generators and would not experience failure due to a loss of offsite power.
- 16. Minor RCS Leak - Makeup Provided by Charging Pumps A minor RCS leak is considered to be a 3/8 in or smaller hole in the reactor coolant pressure boundary. This size is based on the charging pump capability to provide makeup to the RCS to replace the volume lost from the leak. The pump capacity is sufficient to maintain normal RCS pressure / temperature and no significant fluid system transients occur in the RCS. However, other systems associated with the RCS must be evaluated for resulting transients. For example, the CHS system may require increased charging and decreased letdown flows due to the effects of the leak. Plant shutdown is not required during this transient since normal RCS conditions are maintained.
- 17. Single Error of Operator Determination o' system transients should include the worst System operating condition due to an operator error.
For example, valve misalignment could result in excessive pressure buildup in one portion of the system and opening of relief valves.
The System Engineer must evaluate each system for operator induced transients.
1-13
- 18. Single Failure of Control Component These transients consider system failures incorrectly actuating
- fluid system components. For example, failure in the feedwater control system could cause excessive feedwater flow through a wide open flow control valve.
The control circuits for each fluid system must be evaluated by the System Engineer to determine the most severe event.
- 19. Minor Secondary System Leaking Which Would Not Prevent Orderly Shutdown and Cooldown Assuming Normal Makeup A minor secondary system leak is considered as a 2.5 ina or smaller hole in either the main steam, feedwater, or secondary.
system lines associated with the steam generator pressure boundary. A leak of this size is not sufficient to prevent orderly cooldown of the reactor within the 100*F/hr limit. Both the leaking system and all other systems which could be affected by transients in the leaking systems must be evaluated for this Upset Plant Condition.
C. Emergency Conditions t
- 1. Small Loss-of-Coolant Accident For design transient purposes, the small loss-of-coolant accident is defined as a break equivalent to the severance of a 1 in id branch connection.
(areaks smaller than 0.375 in id can be handled by the normal makeup system and produce no significant fluid systems transients.) Breaks which are much larger than 1 in will cause accumulator injection soon after the accident and are regarded as a Faulted Plant Condition.
For design purposes it is assumed that this transient occurs five times during the life of the plant.
It should be assumed that the safety injection system is actuated immediately after the break occurs and delivers water from the RWST at a minimum temperature of 45'F to the RCS.
- 2. Small Steam Line Break i
For design transient purposes, a small steam line break is defined as a break equivalent in effect to a steam safety valve opening and remaining open. This transient is assumed to occur five times during the life of the plant.
The following conservative assumptions were mide a.
The reactor is initially in a hot, sero power condition.
b.
The small steam break results in immediate reactor trip and SI actuation.
c.
A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transient.
i i
1-14 o
d.
The sOf:ty injecticn system op rct:s et design cap: city and repressurizes the reactor coolant system within a relatively short time.
D. Faulted Conditions
- 1. Double-Ended Rupture of Large Reactor Coolant Pipe Following rupture of a reactor coolant pipe resulting in a large loss of coolant, tha primary system pressure decreases causing the primary system temperature to decrease. Because of the rapid blowdown of coolant from the' system and the comparatively large heat capacity of the metal sections of the components, it is likely that the metal will still be at or near the operating temperature by the end of blowdown.
It is conservatively assumed that the safety injection system is actuated to introduce water at a minimum temperature of 45'T into the reactor coolant system.
The safety injection signal will also result in reactor and turbine trips.
- 2. Large steam Line Break This transient is based on the complete severance of the largest steam line. The following conservative assumptions were made:
a.
The reactor is initially in a hot, zero-power condition.
b.
The steam line break results in benediate reactor trip and SI actuation.
l c.
A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transients.
d.
The safety injection system operates at design capacity and repressurizes the reactor coolant system within a relatively short time.
The above conditions result in the most severe temperature and pressure variations which the primary system uill encounter during a steam break accident.
- 3. Feedw'ter Line Break s
This accident involves a double ended rupture of the main
{
feedwater piping from full power, resulting in the rapid blowdown of one steam generator and the termination of main feedwater flow to the others.
The feedwater line break in a two loop plant causes more severe transients than in 3 and 4 loop plants.
Therefore, for design purposes, the results of a two loop plant analysis are used.
The blowdown is completed in approximately l
27 sec. Conditions were conservatively chosen to give the most nevere primary side and secondary side transients. All auxiliary feedwater flow is assumed to exit at the break. The incident is I-15 l
l o
_ -.. -,,. _...... _ - _ = _. - _ _,.. _ _ _. _ -..
I terminated when the operator manually realigns the auxiliary feedwater system to isolate the break and to deliver auxiliary feedwater to the intact steam generator (s). This realignment is i
assumed to occur within 10 min after the initiation of the i
incident.
i
- 4. Reactor Coolant Pump Locked Rotor This accident is based on the instantaneous seizure of a reactor coolant pump with the plant operating at full power. The locked rotor can occur in any loop.
Reactor trip occurs almost isusediately, as the re ult of low coolant flow in the affected loop. conservatively, the transient is based on a locked reactor coolant pump for a two loop plant.
t
- 5. Control Rod Ejection i
is i
This accident ~
based on the single most reactive control rod j
being instantaneously ejected from the core.
This reactivity insertion in a particular region of the core causes a severe pressure increar.e in.the reactor coolant system such that the pressuriser safety valves will lift and also causes a more severe 2
temperature transient in the loop associated with the affected l
region than in the other loops. For conservatism, the analysis is based on the reactivity insertion and does not include the mitigating effects (on the pressure transient) of coolant blowdown through the hole in the vessel head vacated by the i
ejected rod.
t
- 6. Steam Generator Tube Rupture i
t This accident postulates the double-ended rupture of a steam generator tube resulting in a decrease in pressurizer level and j
reactor coolant pressure.
Reactor trip will occur due to the j
resulting safety injection signal.
In addition, safety injection l
actuation automatically isolates the feedwater lines, by tripping all feedwater pumps and closing the pump discharge valves.
When this accident occurs, some of the reactor coolant blows 'down into l
the affected steam generator causing the shell side level to rise.
If the level rises sufficiently, a high level alarm will occur and the feedwater regulating valve will close.
Approximately 10 min after the rupture, the primary system l
pressure is reduced melow the secondary safety valve setting.
At j
this time, the planned procedure for recovery from this accident calls for isolation of the steam line leading from the affected
{
Therefore, this accident will result in a l
transient in the RCS which is no more severe than that associated l
with a reactor trip from full power.
(Upset Plant condition j
No. 6 - Reactor trip with cooldown actuating safety injection.)
i I
i l
l
!=16 i
i
E. Pirnt Tcst Crndititns
- 1. Turbine Roll Test This transient is imposed upon the plant during the hot functional test period for turbine cycle checkout.
Reactor coolant pump power is used to heat the reactor coolant to operating temperature (no-load conditions) and the steam generated is used to perform a turbine roll test. However, the plant cooldown during this test exceeds the 100*F per hour design rate.
The number of such test cycles is specified at 20 times, to be performed at the beginning of plant operating life prior to irradiation.
This transient occurs before plant startup and the number of cycles is therefore independent of other operating transients.
- 2. Primary System Hydrostatic Test The pressure tests covered by this section include both shop and field hydrostatic tests which occur as a result of component or system testing.
This hydro test is performed at a water temperature which is compatible with reactor vessel material ductility requirements and a test pressure of 3107 psig (1.25 times design pressure).
In this test, the reactor coolant system is pressurized to 3107 psig coincident with steam generator secondary side pressure of 0 psig. The reactor coolant system is designed for 10 cycles of these hydrostatic tests, which are performed prior to plant startup. The number of cycles is independent of other operating transients.
Additional hydrostatic tests will be performed to meet the in-service inspection requirements of ASME Section XI. A total of four such tests is expected. The increase in the fatigue usage factor caused by these tests is easily covered by the conservative number (200) of primary side leakage tests that are considered for design and no additional specification is required.
- 3. Secondary System Hydrostatic Test The seconda'ry side of the steam generator is pressurized to 1.25 design pressure with a minim,um water temperature of 120'F coincident with the primary side at 0 psig.
For design purposes it is assumed that the steam generator will experience 10 cycles of this test.
These tests may be performed either prior to plant startup or subsequently following major repairs, or both.
The number of cycles is independent of other operating transients.
I-17 i
1
- 4. Primary System Leaktge Tost subsequent to each time the primary system has been opened, a a
leakage test will be performed. During this test the primary system pressure is, for design purposes, raised to 2500 psia, with the system temperature above the minimum temperature imposed by reactor vessel material ductility requirements, while the system is checked for leaks.
1 In actual practice, the primary system will be pressurized to approximately 2235 psig, as measured at the pressurizer, to prevent the pressurizer safety valves from lifting during the leakage test. This also avoids application of pressure in excess of normal operating pressure to the outside of the fuel.
During this leakage test the secondary side of the steam generator must be pressurized to at least 635 psig so that the pressure differential across the tube sheet does not exceed 1600 psi. This is accomplished with the steam, feedwater, and blowdown lines closed off.
For design purposes, it is assumed that 200 cycles of this will occur during the 40-year design life of the plant.
- 5. Secondary System Leakage Test During the life of the plant it may be necessary to check the secondary side of the steam generator, particularly the manway
- closure, for leakage.
For design purposes, it is assumed that the steam generator secondary side is pressurized to just below its design pressure, to prevent the safety valves from lifting.
{
In order not to exceed a secondary side to primary side pressure differential of 670 psi, the primary side must also be pressurized. The primary system must be above the minimum temperature imposed by reactor vessel material ductility requirements, i.e., between 120*F and 250*F.
It is assumed that this test is performed 80 times during the 40-year life of the plant.
- 6. Steam Generator Tube Leakage Test During the, life of the plant it may be necessary to check the i
steam generator for tube leakage and tube-to-tube sheet leakage.
This is done by visual inspection of the underside (channel head i
i side) of the tube sheet for water leakage, with the secondary side pressurized. Tube leakage tests are performed during plant cold shutdowns.
For these tests, the secondary side of the steam generator is i
(
pressurized with water, initially at a relatively low pressure, and the primary system remains depressurized. The underside of the tube sheet is examined visually for leaks.
If any are observed, the secondary side is depressurized and repairs made by tube plugging. The secondary side is then repressurized (to a l
I-18 1
4
higher'pressura) cnd tha undarsid2 cf tha tub 2 sh st is cgnin checked for leaks. This process is repeated until all the leaks are repaired.
The maximum (final) secondary side test pressure
' reached is 840 psig.
The total number of tube leakage test cycles is defined as 800 during the 40-year life of the plant.
Following is a breakdown of the anticipatednumberofoccurrencesagachsecondaryside test pressure:
m-Number 8f M e Test Pressure, psig Occurrences
'QI 200 400 400 200 600 120 840 80 Both the primary and secondary sides of the steam, generators will be at ambient temperatures during these tests.
2.38/1 7.
Containment Pressure Test In addition to the Preoperational Containment Structural Integrity Pressure Test, Containment Integrated I,eak Rate Pressure Tests are conducted at the first refueling shutdown but not more than 3 years subsequent to the preoperational test and at intervals not to exceed 5 years thereafter during refueling (normal plant condition No. 16).
During the pre-operational
- test, containment pressure will be held at 52 psig for a period of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Operating conditions for systems with lines inside containment, lines penetrating the containment, and lines or equipment located outside the containment in one of the adjacent buildings and supported off the containment wall, must be specified since some movement of the containment wall is expected causing additional pipe stresses.
For pipe stress considerations, the maximum test pressure (structural integrity test) will be used to evaluate affected systems.
I-19 48
App:ndix III Forms Attachments Title 2BVM-45A1 Stress Analysis Data Package Title Page (for ASME III Code Class 2 and 3 Systems) 2BVM-45A2 Pipe Stress Data Package Revision Change / Reason Sheet 2BVM-45A3 General Notes Sheet 2BVM-45A4 System Condition Description Sheet 2BVM-45A5 Flow Transient Data Sheet 2BVM-45A6 Safety / Relief Valve Data Sheet 2BVM-45A7 Miscellaneous Information Sheet 2BVM-45A8 Stress Analysis Data Package Title Page (for non-ASME III Code Class 4 System) 2BVM-45A9 Pipe Stress Data Table 2BVM-45A10 Plant Condition / System Condition Matrix 2BVM-45A11 Line Designation Table Input Data 2BVM-45A12 Line Designation Table Cover Page 2BVM-45A13 Table of Contents 2BVM-45A14 Flow Chart:
Categorization and Identification of High and Moderate Energy Piping e
- b
. l,'
. BVM-45A1
,J.O. No. 12241 SI-Date Beaver Valley Power Station - Unit No. 2 Stress Analysis Data Package QA Category __
NUCLEAR SAFETY RELATED System:
Flow Diagram Number and Revision:
Prepared by Checked by (Principal Engineer)
Reviewed by (Lead Power Engineer)
Reviewed by (Lead EMD Engineer)
Approved by' (Project Engineer)
Analysis:
ASME Class 2 rjqdfor 3 ANSI B31.1 Seismic Cat. I Seismic Cat. II Fluid Flow Transients Confirmation Required other 1 of __
u
Attechm:nt 2BVM-45A2 J.0. No. 12241 O',
BVPS-2 Systc..
STRESS ANALYSIS DATA PACKAGE REVISION CHANGE / REASON SHEET Change Reason 1.
1.
i 4
D
+
e 4
0 e
of O
Attcchmint 2BVM-45A3 J.0, No. 12241 SI-
'BVPS-2 Systems GENERAL NOTES 1.
The anticipated total number of full temperature cycles for Conditions 1 through __ is 7,000 or less.
2.
The Plant Condition / System Condition Matrix (Attachment 1) shows which system conditions could exist during postulated plant conditions.
3.
Flow sketches for Conditions __ through are included as Attachment __.
4.
System
Description:
of o
Attachmint 2BVM-45A4 J.O. No. 12241 SI-BVPS-2 System:
)
SYSTEM CONDITION DESCRIPTION SHEET Condition No. __
Title:
Description of Operation:
References:
1.
Operating Components of
Attachm2nt 2BVM-45A5 J.O. No. 12241 SI-
- BVPS-2 System:
FLOW TRANSIENT DATA SHEET ASME Class 2 and/or 3 ANSI B31.1 Description of Transient Event Pump start /stop Check valve operation S/RV open/close Other (describe)
Isolation valve closure Flow control valve operation Condition Description Additional Information:
References e
of
Atttchmint 2BVM-45A6 J.O. No. 12241 SI-BVPS-2 System:
SAFETY / RELIEF VALVE DATA SHEET ASME Class 2 and/or 3 ANSI B31.1 Description Mark No.
Set Pressure (psig)
Additional Information:
Reference:
9 Of l
. o
4 AttachmInt 2BVM-45A7 Page 1 of 2
- J.O. No. 12241 SI-BVPS-2 System:
MISCELLANEOUS INFORMATION SHEET A.
Manual Valves Size WT (In.)
Mark No.
SWEC File No.
(lbs)
B.
Operating Valves Size WT (In.)
-Mark No.
SWEC File No.
(lbs)
A = Approved AR = Approved as Revised
- = Confirmation Required
- = Manufacturer's Approximate Weight
- ## = Total weight of valve and operator Center of gravity noted on SWEC drawings of
Attcchmint 2BVM-45A7 Page 2 of 2 J.O. No. 12241 SI-BVPS-2 System:
MISCELLANEOUS INFORMATION SHEET (Cont)
C.
Equipment Mark No.
Name SWEC File No.
D.
Miscellaneous 9
A - Approved AR = Approved as Revised
- = Confirmation Required a
of a
r AttechmInt 2BVM-45A8 J.O. No. 12241 SI-Date Beaver Valley Power Station - Unit No. 2 Stress Analysis Data Package QA Category __
System:
Flow Diagram Number and Revision:
Prepared by Checked by (Principal 2ngineer)
Reviewed by (Lead Power Engineer)
Analysis: No ASME Class 2 and/or 3 ANSI B31.1 No Seismic Cat. I Seismic Cat. II Fluid Flow Transients Confirmation Requ, ired other 1 o f __
u
. 'l 8vM-45A9 Page 1 of 2 J.O. No. 12241 SI-System:
BEAVER VALLEY POWER STATION - UNIT NO. 2 PIPE STRESS DATA TABLE Ope ra t i ng Ope ra t ing Ope ra ting Opera ting Ope ra t ing Ope ra t ing Condition 1 Condition 2 Cond i t ion 3 Condition 4 Condition 5 Condition 6 High Press Temp Press Temp Press Temp Press Temp Press Temp Press Temp line Number Ene rgy psig
- F p514.
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_- BVM-45A10 (pg 1 of 2)
BVPS-2 SI-J.O. No. 12241-SYSTEM:
l PLANT CONDITION /SYSTkM CONDITION Mt.TRIX (REF. APPENotX ICF 28vM -45 FOR PLANT CONDITIONS)
A. NORMAL PLANT OPERAT,1NG CONDITIONS Plent Condition 1
8 3
4 5
6 7
8 9
10 11 la la le IS 14 P
Sy-*em Condition (
- 3. UPSET PLANT OPERATING CONDITIONS Pir.nt Condition 1
2 3
4 8
8 7
8 9
10 18 12 13 14 iS le iF 18 19 e
System Condition 4 of u
i Attachm:nt 2BVM-45A10 (pg 2 of 2)
BVPS-2 SI.
J.O. No. 12241 SYSTEM:
l PLANT CONDITION / SYSTEM CONDITION MATRIX (REF. APPENOlX 10F 2SVM-45 FOR PLANT CONDITIONS)
C. EMERGENCY PLANT OPERATING CONDITIONS Plant Condition 1
8 r
System Condition i 5
O.
F AULTED PLANT QPERATING CON 0lTIONS Plant Condition 1
2 3
4.J G
r System Condition f sL 1
l E.
PLANT TEST CON 0lT10NS Plant Condition 1
2 3
4 5
6 7
System Conditionf 5
F. SYSTEM TEST CON 0lTIONS DURING SYSTEM TESTING of LP
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DUQUESNE LIGH T COM PANY - BEAVER VALLEY UNIT 2 J. O.
N O.12 2 4 i S T' O N E AND WE8 STER E N GIN E E RING COR POR A T O O N 80STON, MASSACHUSETTS LIN E D ESIG N ATION^ TABLE
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FOR LDT-ORIGINAL LINE DESIGNATION TABLE ISSUE DATE II/ 20 /72 3
CURRENT REVISION DATE FOR g
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SYSTEN ENGINEER PRINCIPAL ENGINEER LEAD TEST ENGINEER LE AD POWER ENGINEER
Attcchm2nt 2BVM-45A13 J.O. No. 12241 SI-BVPS-2 System:
TABLE OF CONTENTS Page(s)
Stress Analysis Data Package Revision Change / Reason Sheet General Notes System Condition Description Sheets Flow Transient Data Sheets Safety / Relief Valve Data Sheet Miscellaneous Information Sheets Pipe Stress Data Tables :
6 of a
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i ATTACHMENT 2BVM-45A14 (P. 1 o f 2)
FLOW CHART: CATEGORIZATION cnd IDENTIFICATION of HIGH and MODERATE ENERGY PIPING IDENTIFY LINE TO BE CATEGORIZED
[
IESI #1: Does line of concern b
operate or is maintained E
pressurired durins
- a. reactor start-up?
Identify lines as
- b. operation at power?
moderate energy -
NO TEST #1
- c. hot standby?
see note alb
(
- d. reactor cooldown to cold f
shutdown but not p
including refueling?
- YES h
h TEST #2: Does any of the Identify line as b
specified operating condi-moderate energy -
NO TEST #2 tions for the line of concern see note alb
(
meet the pressure / temperature C
criteria for a high energy h
line as defined in note #27 f-YES I
Identify line as E
TEST #3: Are the specified moderate energy -
NO TEST #3 operating conditions, which see note alb T
meet the high energy line C
pressure / temperature criteria,
[
for a " NORMAL" system operation?
That is, they are "NOI"
[
- a. for an Upset, Emergency
[
or Faulted system condition.
,5
- b. for a system testing
(
system condition.
t
- c. for a system maintenance L
system condition.
BYES L
TEST #4: Can the line be classified as moderate energy because the normal b
system operation (s), which Identif* line as NO TEST #4 meet the high energy line high energy -
(
pressure / temperature cri-see note ela
[
teria, experience these F
high internal energies for E
less than 2% of the time that I YES the systems operate?
Refer to note #3 for additional information.
C Identify line as moderate energy because of system's short operational period -
see no'te alc u
ATTACHMENT 2BVM-45A14 (P.2of2)
NOTES:
- 1. Identify line categories as follows
- a. High energy Lines so categorized shall be identified by a "YES" in the column titled "High Enersy" in the Pipe Stress Data Table for the particular line number.
4
- b. Moderate energy - Lines so categorized shall be identified by a "NO" in the column titled "High Enersy" in the Pipe Stress Data Table for the particular line number.
- c. Moderate energy because of system's short operational period - Lines so categorized shall be as identified as follows:
(1) A "N0 *" shall be identified in the column titled "High Enersy" in the Pipe Stress Data Table for the particular line number.
(2) The following note shall be added to the Pipe Stress Data Table.
- lines so identified have been re-classified as mode-rate energy because the line experiences high energy operating conditions for a short operational period.
(3) A single asterisk (*) is only sussested, any symbol may be used.
- 2. High energy line pressure / temperature conditions are system conditions where either or both of the following are met:
- s. Max. operating temperature exceeds 200
- F, or
- b. Max operating pressure exceeds 275 psis.
I NOTE: A line is not high energy if the operating pressure is Opsis or less, regardless of the operating temperature.
- 3. A line meeting TESTS ~1, 2 & 3 can still be classified as moderate energy, for pipe ruptura purposes only, if the normal system operating condition (s), meets the high energy line
~
pressure / temperature criteria for less than 2 percent of the time that the line/ system operates as moderate energy (i.e.
all lines not high energy with P>0 pois). For example systems such as RHS qualifies as moderate energy.
It should be noted that this re-categorization test applies and should be used DNLY when a line cannot be categorized as moderate energy using the first three tests. This rule has nothing to do with classifying what system or plant operating conditions exist.
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APPENDIX IV (Page 1 of 10)
BVPS-2 STUDY OF FLOW 1
TRANSIENT ANALYSES IDENTIFIED IN NUREG-0582 i
Title / Description Applicability A.' Pump Start vith Inadvertently Voided Discharge Lines 1.
RHS Applicable only during emergency cooldown (during normal cooldown system is vented before operation and kept full by RWST).
2.
MHSI NA - Discharge lines are normally filled and operating as the chemical and volume control system.
3.
LHSI NA - RWST minimum water level is above the elevation of the cold les injection noz-zles, therefore, the lines are kept full.
4.
SWS Applicable-Enveloped by tran-sient of restart of pumps.
(See Section F, Transient No. 40).
I Vacuum breakers were installed to reduce water hammer from restart of pumps. Voids would form as a result of pump stop and SWS water drain back to ultimate heat sink.
i 5".
CCP NA - RHS coolers and RHS pump seal water coolers are the only two coolers not normally in operation.
The portion of the CCP system to the RHS coolers is vented before normal cooldown.
During emergency cool-down air pockets would be trapped in the RHS coolers J
- To be analyzed by BVPS-2 u
-.,.-----..___.-,..--_._....,.n
APPENDIX IV (Pege 2 of 10)
Title / Description Applicability and would not pose a water hammer problem.
6.
FWE NA - Voids are not expected due to filling and venting procedures af ter ISI tests.
7.
RSS Applicable - RSS discharge lines to spray rings are normally empty. Enveloped by system fill of empty lines (See Section B, Transient No. 14).
8.
QSS Applicable - QSS discharge lines to spray rings are normally empty. Enveloped by system fill of empty lines (see Section B, Transient No. 15).
B. Expected Flow Discharge into Initially Empty Lines 9.
RCS Safety and Applicable Relief Valve Discharge 10.
RHS Suction Applicable Relief Valve Discharge 11.
HHSI NA - Lines normally filled and operating as CVCS.
12.
LMSI - SIS NA - Lines normally filled Lines from with water from the RWST.
RWST to RCS Voids are not expected due Cold and to filling and venting Hot Legs procedures.
- To be analyzed by BVPS-2 o
APPENDIX IV (Pega 3 of 10)
Title / Description Applicability
- 13.
LHSI - Cross-Applicable over from RSS for Longterm Safety In-jection from the Containment Sump
- 14.
RSS - Normal Applicable Pump Start in-to Empty Lines and Full Flow Pump Test
- 15. OSS Normal Applicable Pump Start into Empty Lines
- 16.
NSS Safety Applicable Relief Valves and Residual Heat Removal Valves Opening
- 17. MSS Steam Sup-Applicable - Initially ply to Aux.
empty lines at ambient Feedwater Pump temperature are filled Turbine with main steam.
18.
CCP Flow into NA - All coolers supplied by Empty Coolers CCP system are normally filled except when down for maintenance.
Venting procedures during gradual fill eliminate the possibility of this tran-sient occurring.
- 19.
SWS Flow into Applicable - When the RSS Initially Es-system starts af ter accident, pty RSS service water valves isolating Coolers the RSS cooler are opene*
filling the empty lines and coolers.
- To be analyzed by BVPS-2 o
=-
APPENDIX IV (Paga 4 of 10)
Title / Description Applicability
- 20.
FWE Opening of Applicable - Relief valve RV101 on the opens filling empty line Turbine Pump (relieves excess pressure Discharge due to overspeed of the turbine).
C. Valve Opening, Closing, and Instability 21.
RHR - Closing /
NA - Suction and discharge Opening of motor operated isolation Suction or valves have an opening /
Discharge closing time of 120 sec.
Isolation No water hammer will occur Valves due to valve closing or opening. Also, since the HCVs fail open and a recire loop is provided
^
around the pumps, a flow path for the pumps would always exist after failure of any one of the control valves. No water hammer would occur.
22.
HHSI - Closing /
NA - Discharge motor-Opening of Dis-
- operated isolation valve charge Isola-has an opening / closing tion Valves time of 10 sec. Also, a flow path would always exist after a failure of the control. valves. No water hammer should occur.
I 23.
LHSI Opening /
NA - Discharge motor-Closing of Dis-operated isolation valves charge Isola-have an opening / closing tion Valves time of 12.2 sec. No water hammer should occur.
- 24. EFWS - Closure Applicable of' Flow Con-trol Valves
- To be analyzed by BVPS-2 u
T APPENDIX IV (Pcga 5 of 10)
Title / Description Applicability
- 15.
FWE - Closure Applicable due to modulation of Flow Con-of control valves causing trol Valves pressure waves.
Control valve closing not con-sidered to produce signi-ficant dynamic loads.
- 26. MSS Steam Sup-Not Applicable - The steam ply to Aux hammer event of closing Feedwater Pump the isolation valves would Turbine Isola-be enveloped by the steam tion Valve fill of empty line transient Closure discussed in Section B, Transient No. 17.
- 27.
MSS Closure of Applicable Turbine Stop Valves
- 28.
MSS Opening of Applicable Turbine Bypass Valve
I,.td developed would be enveloped by those from the closing of the turbine stop valves (150 milli-seconds).
30.
RSS Isolation NA - Discharge motor-Valve Closure operated isolation valves (2RSS*MOV156A-have an opening / closing time D) of 60 sec.
31.
QSS Isolation NA - Discharge motor-Valve Closure operated isolation valves (2QSS*MOV101A, have an opening / closing time B) of 60 sec.
D. Check Valve Closing.and Delayed Opening s
- To be analyzed by BVPS-2 o
APPENDIX IV (Paga 6 of 10)
Title / Description Applicability 32.
RSS Check NA - If only 2 of 4 RSS pumps Valve Closure start, RSS water will also During Minimum fill the risers up to the Safeguards spray headers in the safety Operation trains that are not operating.
No water slug is expected due to the displacement of air in the riser. This transient would be enveloped by the startup transient in Section B, No. 14.
- 33. QSS Check NA - Same as above for RSS Valve Closure except that only 1 of 2 QSS During Minimum pumps start.
Safeguards Operating i
- 34.
FWS Main Feed-Applicable - Check valve water Check must maintain its structural Valve Slam integrity only.
- 35. MSS Check Applicable - Check valve Valve Closure closure due to postulated MSS in the Steam line break.
This transient Supply Lines should be analyzed in con-to the Aux.
junction with steam fill Feed Pump of the empty line (Section B, Turbine Transient No. 17) to determine impact.
E. Water Entrainment in Steam Lines
- 36. MSS Main Steam NA - Entrained water in main Lines steam line should be limited by the collection and drain-ing of any condensate formed in the lines.
- 37. MSS Water En-Nk-Sameasabovefor trainment in main steam lines.
Turbine Bypass Piping
- To be analyzed by BVPS-2
APPENDIX IV (Page 7 of 10)
Title / Description Applicability
- 38. MSS Water En-NA - Same as above for trainment in main steam lines. Should Steam Lines to be included as part of the Aux. Feed-steam fill of aux. feed-water Pump water pump turbine.
Turbine F. Transient Cavita-tion (Column Separation)
- 39. CCP Water NA - BVPS-2 has a surge tank Column Separa-which keeps the system filled tion Effects (no drain back after pump Following Pump stop.)
Stopping
- 40s. SWS Water Applicable - Analyzed in EMD Column Separa-Stress Calc 173. Vacuum tion Effects at Breakers were located to Pump Discharge eliminate water hammer.
Following Pump Stop and Restart 40b. SWS Water Column NA - Check valves added Separation Effects,'(([,[ ','to eliminate water
~
at Main Steam column separation.
Valve House Cool-ing Coils Follow-ins Pump Stop and Restart 41.
RSS Water NA - A study of the contain-Column Separa-ment sump as conducted by tion Effects Alden Labs to set the following pump minimum water level before Stopping or pump operation to preclude Inadequate cavitation of the RSS pumps.
NPSH Also, pumps have been tested with inadequate NPSH for 13 minutes without failure.
The RSS pumps and safety trains are separated and redundant. A flow transient I
analysis for column separa-tion due to drain back to i
- To be analyzed by BVPS-2
%J J
APPENDIX IV (Page 8 of 10)
Title / Description Applicability cont. sump after pump stop and subsequent restart is not required since a single failure in one train would not affect the others.
G. Steam Bubble Col-lapse Due to Rapid Condensation 42.
HHSI Collapse NA - Cold water injection of Steam Bub-into the RCS from safety bles Formed as injection could cause rapid a Result of condensation of steam with Local or resulting water hammer back System through safety injection lines.
Depressurization Loads are not considered large enough to be of concern (pg.
A-21, NUREG-0582).
43.
LHSI Collapse NA - Same as NHSI above.
of Steam Bub-bles due to Local or System Depressurization 44.
Safety Accum-NA - Same as NHSI above.
ulator Collapse of Steam Bubble Due to Local or System Depressurization 1
45.
RHR Collapse NA - Steam bubble formation of Steam between check and closed Bubble isolation valves due to higher temperature in the RCS avoided by startup procedures.
Also, the heat transfer from pipe to containment atmosphere i
l would constantly condense any steam present.
I I
APPENDIX IV (Page 9 of 10) s Title / Description Applicability 46.
FWS Slug Im-NA - Unlikely due to in-pact Due to verted
'J' tube design of Rapid Conden-feedring in steam generators sation in on BVPS-2.
Steam Generator 47.
CCP Steam Bub-NA - Unlikely since system ble Formation is filled and vented be-and Collapse fore being placed into in Heat operation.
Exchangers 48.
SWS Steam NA - Same as CCP above.
Bubble Forma-tion and Col-lapse in Heat Exchangers H. Pump Start and Postulated Seizure with Full Lines 49.
HHSI NA - Pump start /stop negligible due to adequate venting and priming procedures before start, (Pg. A-30, NUREG-0582).
Pump seizure is not analyzed as a transient mode on BVPS-2 since motor overload protec-tion is provided which should trip the motor as rotational resistance increases prior to complete seizure.
Also rotational momentum of the pump and motor will privent an instantaneous pressure spike as a result of pump seizure.
I I
47 s
APPENDIX IV (Page 10 of 10) i Tit e Description Applicability l/
50 LHSI NA - Pump start /
stop should not produce large dynamic loads due to adequate venting and priming procedures before start.
Postulated pump l'
,I '
... seisure is considered.a single failure on BVPS-2.
Since LHSI pump suction and discharge lines are separated and redundant on BVPS-2, no flow transient analysis is required.
51.
RHS NA - Same as NHSI above (Transient No. 49).
52.
RSS NA- Pump start /stop are t
analyzed as part of tran-sient No. 14 (pump start and fill of empty Postulated 3
pump seizure is considered a single. failure on BVPS-2.)
Since the RSS pump suction and discharge lines are separated and redundant, no flow transient analysis is required.
- 53. QSS
,NA - Same as RSS above.
54.
CCP NA,- Same an ENSI above (Transi'ai. 'Jo. 49)
(
tNA - Same as HHSI above 56.
FWE-(Transient No. 49).
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)
1 4r
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a APPENDIX V (Page 1 of 1)
ADDITIONAL FLOW N SIENTS ANALYZED BY BVPS-2 Title / Description 1.
Moisture separator safety valve discharge into initially empty lines.
\\;
2.
Feedwater heater vent safety valve discharge into initially empty lines.
3.
Impact of inlet water on tubes in the RSS coolers.
4.
Feedwater pump discharge check valve closure after feedwater jj u
pump trip.
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