ML16127A477
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TMI-1 UFSAR APPENDIX 14A 14A-i REV. 18, APRIL 2006 APPENDIX 14A - DESIGN REVIEW FOR CONSIDERATION OF EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENT TABLE OF CONTENTS SECTION TITLE
1.0 INTRODUCTION
2.0 REVIEW CRITERIA 3.0 GENERAL DISCUSSION 3.1 DESIGN 3.2 QUALITY ASSURANCE 3.3 SHUTDOWN CAPABILITY 4.0 DISCUSSION OF BREAKS 4.1 CONTROL ROOM AND CONTROL BUILDING 4.2 INTERMEDIATE AND TURBINE BUILDING 4.3 AUXILIARY, FUEL, AND INTERMEDIATE AREA (PERSONNEL ACCESS)
BUILDING (AUXILIARY AREA) 4.4 DIESEL GENERATOR BUILDING 4.5 COLD SHUTDOWN CAPABILITY 5.0
SUMMARY
AND CONCLUSIONS 6.0 INSPECTION REQUIREMENTS 7.0 METHODS OF ANALYSIS AND DESIGN 7.1 THERMAL-HYDRAULIC ANALYSIS 7.1.1 BLOWDOWN ANALYSIS 7.1.1.1 MAIN STEAM BLOWDOWN ANALYSIS 7.1.1.2 FEEDWATER BLOWDOWN ANALYSIS 7.1.2 COMPARTMENT PRESSURIZATION ANALYSIS 7.1.2.1 MAIN STEAM LINE RUPTURES 7.1.2.2 FEEDWATER LINE RUPTURES 7.1.3 PIPE THRUST AND JET IMPINGEMENT 7.2 STRUCTURAL ANALYSIS AND DESIGN 7.2.1 WALL AND SLAB CAPACITIES
TMI-1 UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE APPENDIX 14A 14A-ii REV. 18, APRIL 2006 7.2.1.1 GENERAL YIELD-LINE THEORY AND SHEAR CAPACITY 7.2.1.2 COMPARTMENT DIFFERENTIAL PRESSURE VERSUS CAPACITIES 7.2.1.3 JET IMPINGEMENT AND COMBINED DIFFERENTIAL PRESSURE 7.2.2 RESTRAINT DESIGN PROCEDURE 7.2.2.1 ASSUMPTIONS 7.2.2.2 ANALYTICAL METHOD 7.2.2.3 MATERIAL PROPERTIES 7.2.2.4 SECTION PROPERTIES 7.2.2.5 EXAMPLE MODEL 7.2.2.6 DETAIL DESIGN
8.0 REFERENCES
TMI-1 UFSAR APPENDIX 14A 14A-iii REV. 18, APRIL 2006 LIST OF TABLES TABLE TITLE 14A-1 HIGH-ENERGY LINES INVESTIGATED FOR CONSEQUENCES OF POSTULATED PIPE RUPTURES 14A-2 SYSTEM REQUIRED FOR SAFE SHUTDOWN FOLLOWINGA POSTULATED MAIN STREAM OR FEEDWATER PIPE RUPTURE 14A-3 TABULATION OF SYSTEMS AND BREAK LOCATIONS 14A-4 MAXIMUM COMPARTMENT PRESSURES 14A-5 COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE 14A-6 CHRONOLOGY OF EVENTS FOR HIGH ENERGY PIPE BREAK 14A-7 THREE MILE ISLAND NUCLEAR STATION - UNIT 1 INTERMEDIATE BUILDING ENVIRONMENT CONDITIONS 14A-8 MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK 14A-9 CONCRETE PROPERTIES
TMI-1 UFSAR APPENDIX 14A 14A-iv REV. 18, APRIL 2006 LIST OF FIGURES FIGURE TITLE 14A-1 MAIN STEAM, FEEDWATER & EMERGENCY FEEDWAT LINES IN THE INTERMEDIATE BUILDING 14A-2 MAIN FEEDWATER & AUXILIARY STEAM LINES IN THE 14A.FIG-2 TURBINE BUILDING NEAR THE REACTOR BUILDING 14A-3 DECAY HEAT, MAKEUP & PURIFICATION, AUXILIARY STEAM LINES, AND HIGH ENERGY SAMPLE LINES IN THE FUEL HANDLING AND AUXILIARY BUILDINGS BELOW EVALUATION 305 FT 14A-4 MAKEUP & PURIFICATION AND AUXILIARY STEAM LINES ABOVE ELEVATION 305 FT IN THE AUXILIARY AND FUEL HANDLING BUILDINGS 14A-5 PHYSICAL ARRANGEMENT OF BUILDINGS 14A-6 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 112 TO STOP VALVE SV-4 14A-7 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 113 TO STOP VALVE SV-3 14A-8 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 114 TO STOP VALVE SV-2 14A-9 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 419 TO STOP VALVE SV-1 14A-10 MAIN STEAM FROM STEAM GENERATOR-A TO EMERGENCY FEEDWATER PUMP TURBINE 14A-11 MAIN STEAM FROM STEAM GENERATOR-B TOHEADER AT ELEVATION 297 FT 0 IN.
14A-12 FEEDWATER FROM 30 IN HEADER TO PENETRATION NO 103 14A-13 FEEDWATER FROM 30 INCH HEADER TO PENETRATION NO 227 14A-14 MAINSTEAM BLOWDOWN THRUST BREAK AT CONTAINMENT 56 (AT ISOLATION VALVE) 14A-15 MAIN FEEDWATER BLOWDOWN THRUST BREAK AT CONTAINMENT 3T24 14A-16 INTERMEDIATE BUILDING SUBCOMPARTMENT DESIGNATIONS
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE APPENDIX 14A 14A-v REV. 18, APRIL 2006 14A-17 RESTRAINT FOR BREAKS IN 12 INCH HEADER 14A-18 RESTRAINT FOR BREAKS IN MAIN FEEDWATER 14A-19 MOMENT VS CURVATURE AND SHEAR VS SHEARING STRAIN 14A-20 DYNAMIC PIPE BREAK MODEL 14A-21 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A MAIN STEAM LINE BREAK (HEAT SLAB INCLUDING, EL 295')
14A-22 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A EFW PUMP TURBINE STEAM LINE BREAK HEAT SLAB INCLUDED, EL 295')
14A-23 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A MAIN STEAM LINE BREAK (HEAT SLAB INCLUDED, EL 322')
14A-24 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A EFW PUMP TURBINE STEAM LINE BREAK (HEAT SLAB INCLUDED, EL 322')
14A-25 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING STEAM LINE BREAK (HEAT SLAB INCLUDED, LONG TERM COOLDOWN, EL 295')
14A-26 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A STEAM LINE BREAK (HEAT SLAB INCLUDED, LONG TERM COOLDOWN, EL 322')
TMI-1 UFSAR APPENDIX 14A 14A-1 REV. 21, APRIL 2012 APPENDIX 14A DESIGN REVIEW FOR CONSIDERATION OF EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENT
1.0 INTRODUCTION
A design review was performed in response to the United States Atomic Energy Commission (AEC) request, "General Information Required for Consideration of Effects of Piping System Break Outside Containment," of December 15, 1972, for Three Mile Island Nuclear Power Station, Unit-1. Also in response to the AEC letter of June 1, 1973, an additional design review was performed. The review covers all piping systems in the unit in accordance with the criteria presented in the AEC request. All of the possible effects of postulated pipe failures as outlined in the AEC request have been considered in the review, and it is concluded that the unit can be safely shut down. This report summarizes the results of the design review and outlines the methods of achieving safe shutdown.
Design modifications will be implemented to provide assurance that the Engineered Safeguard Systems and the Emergency Feedwater System will be operable after postulated high energy pipe breaks. Inservice inspection of certain specified postulated break locations will provide assurance that rupture will not occur at those respective points. Operation of the Emergency Feedwater System will permit a more rapid shutdown capability than is possible with Engineered Safeguard Systems alone. Thus, the primary means of effecting a cooldown after a postulated break outside containment would be the Emergency Feedwater System with high pressure-low pressure injection cooldown serving as a backup.
TMI-1 UFSAR APPENDIX 14A 14A-2 REV. 21, APRIL 2012 2.0 REVIEW CRITERIA The sets of criteria used for the review are as follows: (1) considering the type of affects that can result from a pipe break and (2) criteria addressing the degree of system operability required following a break. The criteria used are as follows:
Pipe Break Criteria Pipes Evaluated Effects Considered____________
Fluid above 200 F and 275 psig Longitudinal and circumferential at terminal ends and high stress breaks including pipe whip, jet locations on lines 4 inches or impingement, flooding, and greater environmental conditions Fluid above 200 F and 275 psig Circumferential breaks including at terminal ends and high stress pipe whip, jet impingement, locations on lines 1 to 4 inches flooding, and environmental conditions Pipe Break Criteria Pipes Evaluated Effects Considered___________
Fluid above 200 F and/or 275 psig Crack breaks including jet at the most adverse locations impingement, flooding, and for all pipes environmental conditions Fluid below 200 F and 275 psig None System Operability Criteria System Operability Required_________
Systems required to bring the unit No loss of required redundancy to safe shutdown following the break permitted Reactor Protection System and the Loss of redundancy but no loss Engineered Safeguards System of function permitted Table 14A-1, "High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures,"
indicates the various high energy lines investigated under the above criteria and presents appropriate data or information concerning them. Figures 14A-1, 14A-2, 14A-3, and 14A-4 supplement the location and elevation data given in Table 14A-1 for the various high energy lines investigated.
TMI-1 UFSAR APPENDIX 14A 14A-3 REV. 21, APRIL 2012 3.0 GENERAL DISCUSSION 3.1 DESIGN A rupture of the high energy piping is considered highly unlikely due to the low seismic and operating stress levels. All these systems have been conservatively designed and all the systems except auxiliary steam to the Auxiliary Building have been analyzed in accordance with USAS B31.1.0, Code For Power Piping. The auxiliary steam system to the Auxiliary Building has been designed and analyzed as described in Section 5.4.4.2. This includes all portions of the auxiliary steam system located in the Control, Fuel Handling and Auxiliary Buildings.
Results of these analyses show that the maximum stress levels from combined operating and seismic conditions are well below those limits designated as potential pipe rupture stress levels.
Piping systems are designed to USAS B31.1.0. In addition, portions of the auxiliary steam system piping are analyzed in accordance with the CDFM methodology as described in Section 5.4.4.2. Quality assurance was applied to USAS B31.1.0 requirements for the non nuclear piping and to USAS B31.7 requirements for nuclear piping. (Nuclear piping is defined as piping that normally contains radioactivity.) The analysis of the auxiliary steam piping is based on the configuration and conditions of the piping system at the time of the walkdown and evaluation.
To ensure that the existing condition is maintained, the quality classification of the auxiliary steam piping system has been upgraded to Regulatory Required with QA.
On non-nuclear piping, welders qualified to ASME Section IX requirements were used. Piping system leakage testing was performed in accordance with piping code (USAS B31.1) requirements. All welds were visually inspected.
a.
The main steam piping welds 4 inches and over were 100 percent radiographed from steam generators to the turbine generator.
b.
The main feedwater piping welds 4 inches and over were 100 percent radiographed from pumps to steam generator.
c.
The emergency feedwater piping welds were 100 percent radiographed from the steam generators up to the first isolation valve (which is in the Intermediate Building).
d.
The steam supply (to the emergency feedwater pump turbine) piping welds were 100 percent radiographed.
All the NDT required by USAS B31.7 was applied to nuclear piping systems, i.e., decay heat, makeup and purification, sampling, and so forth.
3.2 QUALITY ASSURANCE The design and construction phase Quality Assurance Program was a three-level program.
The first level of the program was performed by the equipment manufacturer or site contractor, the second level by Met-Ed's main contractor (i.e., B&W--NPGD, GAI, or UE&C, as appropriate), and the third level by Met-Ed itself and/or its agent, MPR Associates.
TMI-1 UFSAR APPENDIX 14A 14A-4 REV. 21, APRIL 2012 3.3 SHUTDOWN CAPABILITY The unit can be brought to a cold shutdown condition by utilization of either the emergency feedwater system and atmospheric dump valves, or HPI cooling and the Reactor Building emergency coolers.
The emergency feedwater system and the atmospheric dump valves are located in the Intermediate Building. The emergency core cooling systems are located in the Auxiliary Building area, separated from the Intermediate and Turbine Buildings by the Reactor and Fuel Handling Buildings.
The highest energy lines are located in the Intermediate and Turbine Buildings. For major breaks in these lines, unit shutdown will be accomplished through utilization of the Emergency Feedwater System and atmospheric dump valves or HPI Cooling and the Reactor Building Cooling Systems. For crack breaks, a normal unit shutdown will be achieved.
TMI-1 UFSAR APPENDIX 14A 14A-5 REV. 22, APRIL 2014 4.0 DISCUSSION OF BREAKS The layout of equipment within the buildings and the physical arrangement of the buildings themselves provide protection for the shutdown equipment from high energy line breaks. See Figure 14A-5 and Table 14A-2 for relative building and equipment locations.
Locations of postulated breaks have been determined for each of the high energy piping systems that might endanger the Emergency Feedwater System if rupture should occur. The selection of breaks is based on the results of stress analyses previously performed on the as-built piping systems. This review considered effects of pressure, deadweight, and thermal expansion during normal operating, upset, test conditions, and the operating basis earthquake (OBE). The stress levels obtained by this review for main steam and feedwater were found to be lower than those of the AEC pipe rupture criteria. To provide a conservative criterion for selecting break locations, the two intermediate points of highest stress are postulated as break locations.
Design basis breaks in straight or curved pipes 4 inches in diameter or greater are assumed to be either longitudinal or circumferential with the break area equal to the flow area of the pipe.
Design basis breaks at branch and longitudinal in the run with the break area equal to the flow area of the branch. The criteria used to select design basis break locations are as follows:
a.
Postulated breaks at all terminal points (anchors or rigid attachment to equipment or anchor extensions).
b.
Postulated breaks at all branch points.
c.
Postulated intermediate breaks between terminal points whenever the primary stress (pressure, weight, OBE) plus secondary stress (thermal) exceeds 80 percent of Sh +
SA), or where secondary stress alone exceeds 80 percent of SA.
d.
As a minimum, two intermediate breaks between terminal points were selected at locations of highest stress.
The above pipe break location criteria were applied to the high energy piping systems, and the results are tabulated in Table 14A-3. This tabulation shows the pipe isometric drawing numbers associated with each of these systems, the potential break point identification number, and the primary and secondary stresses proportional to the 80 percent allowable limit. The break locations are shown on the isometrics on Figures 14A-6 through 14A-13 and in plan on Figures 14A-1 through 14A-4.
In addition, crack breaks were postulated at adverse locations and assumed to be one half the pipe diameter in length and one half the pipe wall thickness in width.
The specific thrust versus time curves used in designing the restraints defined in this supplement are shown on Figures 14A-14 and 14A-15.
Item 15 of the Atomic Energy Commission document titled "General Information Required for Consideration of the Effects of a Piping System Break Outside Containment," Reference 12, required that a discussion should be provided for the potential for flooding of safety related equipment in the event of failure of a feedwater line or any other line carrying high energy fluid.
TMI-1 UFSAR APPENDIX 14A 14A-6 REV. 22, APRIL 2014 4.1 CONTROL ROOM AND CONTROL BUILDING The Control Building equipment, electrical power and control, chilled water system, and ductwork systems are contained within the structure of the Control Building. In this isolated location they would not experience adverse effects from any high energy pipe break. Access to the Control Building structure is either through the Turbine Building or the Fuel Handling Building. Outside air to the Control Building is ducted to the Control Room from a remote underground intake terminal and would not be adversely affected by a high-energy pipe break.
The Turbine Building would experience momentary overpressure if the break occurred in this area, but this would be dissipated through numerous wall and roof openings. Steam leakage from the turbine hall to the Control Room or Control Building during this period is minimized as it is forced to travel through the west Turbine Building wall, through multiple doors in series, before entering the Control Building areas. Also, the doors have automatic closers. Any steam leakage into the corridor space outside the Control Room or the Control Room space will be condensed and dissipated by the ventilation systems, and no significant ambient changes would be anticipated in these areas.
Investigation indicates that there are no high-energy lines larger than 1 inch other than the auxiliary steam pipe in or near the Control Building, and thus, postulated pipe whip and steam jet impingement are not able to damage the Control Room. Due to the low operating pressure of the auxiliary steam system, rupture of this line is not a consideration.
High-energy sample lines (under 1 inch) are discussed in Section 4.3.
4.2 INTERMEDIATE AND TURBINE BUILDINGS a.
The Intermediate and Turbine Buildings contain all of the lines over 1 inch with internal fluid exceeding both 200oF and 275 psig.
The pressure and temperature response in the Intermediate Building following steam line breaks are based on plant specific mass and energy release rates as shown in Table 14A-8. The thermal-hydraulic results are summarized on Table 14A-7 and supporting Figures 14A-21 through 14A-26 which indicate the temperature and pressure time history profiles for each level or compartment. As shown, the Main Steam Line Break is the limiting and enveloping break in the Intermediate Building from a pressure/temperature viewpoint. This is due to the much larger mass of steam being released compared to the EFW pump turbine steam supply line break.
The Reactor Building will maintain its containment integrity when subjected to the resultant external pressurization of a main steam or feedwater break within the Intermediate Building. The Reactor Building will not be subjected to main steam pipe whip because the 3 ft thick interior walls of the Intermediate Building effectively restrain the pipe.
b.
Electrical equipment is required to function subsequent to a High Energy Break (HELB) inside or outside of containment and must meet environmental qualification requirements. The qualification conditions considered include post accident pressure and temperature conditions in Table 14A-7. Electrical equipment which is required to
TMI-1 UFSAR APPENDIX 14A 14A-7 REV. 22, APRIL 2014 function following a postulated HELB is located at the 295 ft elevation of the Intermediate Building.
c.
Isolation valves in the Intermediate Building are on lines that might be open to the containment atmosphere during normal operation (purge valve) and have been reviewed. It was found that for any postulated high energy line break, the valve will not be damaged and will close upon receiving an electrical (deenergize) signal or loss of control air.
d.
The systems that will be used to bring the plant to a safe shutdown after the postulated major break in the Intermediate Building are listed in Table 14A-2. A detailed accident review was made to resolve the effect of each postulated break defined by Section 4.0 on the operability of the Emergency Feedwater Train. The objective of this review was to establish those breaks that would ultimately prevent the operation of both steam generators or both Emergency Feedwater Trains and to determine the design modifications necessary to assure emergency feedwater operation.
1)
Breaks at the containment penetrations in the small compartments 2 and 5 (refer to Figure 14A-16) could produce pressures in excess of wall and/or slab capacities. Portions of the Emergency Feedwater System below elevation 322 ft 0 inch could be damaged by the resulting debris. To provide reasonable assurance that the postulated ruptures in those compartments will not occur, the associated welds are to be inspected in accordance with Technical Specification 4.15.1.
The breaks in compartments 3 and 4 are similar but do not produce differential pressures that would produce incremental collapse of the Intermediate Building interior structures. The method of calculating slab and wall capacity (Yield Line Theory) is reviewed in Section 7.2.1.1 of Appendix 14A and a comparison of capacities with expected differential pressures is summarized in Section 7.2.1.2 of Appendix 14A.
2)
The 12 inch main steam header below elevation 322 ft 0 inch and shown on Figure 14A-1 is positioned directly opposite emergency feedwater valve EF-V1B.
The restraint/shield schematically represented on Figure 14A-17 is provided to protect valve EF-V1B from damage due to postulated breaks in the header and in the 12 inch main steam line that connects to the end of the header nearest EF-V1B.
3)
The feedwater line from containment penetration No. 103 runs approximately parallel to and above a 12 inch main steam line, which is also in the overhead of the same compartment as the turbine driven EFW Pump. To prevent both Steam Generators from becoming inoperable and the loss of the turbine driven EFW pump, the postulated breaks in this section of the pipe have been reevaluated with respect to References 15 and 16. The results of this evaluation indicate that this section of pipe can be classified as "superpipe" and that no breaks have to be postulated in this area (i.e., Break Locations 1 and 7 have been changed to 13 and 28 on Figure 14A-12). The pipe whip restraints
TMI-1 UFSAR APPENDIX 14A 14A-8 REV. 22, APRIL 2014 installed on the feedwater line to reduce pipe deflections due to a pipe break at locations 1 and 7 shall remain to reduce pipe deflections from pipe breaks at locations 13 and 28. Figure 14A-18 illustrates the protection provided.
The design procedure applied to the design of this restraint and the one discussed in Section 4.2.d.2. of Appendix 14A conform to the procedures presented in Section 7.2.2 of Appendix 14A.
4)
Section 4.5 of this appendix provides an analysis of cooldown capability without using the Emergency Feedwater System.
5)
An evaluation was made in Reference 13 for different alternatives available in order to mitigate flooding in the Intermediate Building in case of a postulated feedwater line break. The flood protection modifications were implemented to mitigate the effects of flooding due to a postulated main feedwater line break (MFLB) in the Intermediate Building by allowing water to flow into the tendon access gallery and the alligator pit. By removing the upper half of the western water "stop wall" on the alligator pit and opening the doors at entrance "A" and "B" to the tendon access gallery, there will be approximately 25 minutes before flooding in the Intermediate Building adversely affects the emergency feedwater system components not qualified for submergence would be adversely affected (Reference 14). Intermediate Building flood detection alarm system has been added to alert the operator in the Control Room to flooding conditions as a result of a MFLB. The alarm will provide the operator with sufficient time, approximately 20 minutes, to take corrective action to prevent damage to the EFW pumps.
e.
Postulated breaks in the Turbine Building cannot adversely affect equipment utilized for the safe shutdown of the reactor. Since there is no affected reactor safety equipment in this area, the review of breaks in the Turbine Building is complete.
4.3 AUXILIARY, FUEL, AND INTERMEDIATE AREA (PERSONNEL ACCESS)BUILDINGS (AUXILIARY AREA)
Those reactor protection systems and engineered safeguards systems that could be affected by postulated pipe breaks are all located in the Auxiliary Building.
The effects of breaks in the Intermediate or Turbine Building on the Auxiliary Building ambient atmosphere will be minimal and momentary as this steam leakage to the Auxiliary Building area must pass through the west turbine building wall, through the controlled access Hot Tool Room door, and travel approximately 150 feet before entering the Auxiliary or Fuel Handling Buildings.
This leakage would be continuously condensed and dissipated by the outside air ventilation exhaust systems in the corridor and in the Auxiliary and Fuel Handling Buildings.
The postulated breaks in the auxiliary area will not require protective action because they do not deplete primary system inventory or impair the normal heat removal systems (main steam and feedwater).
Another characteristic of pipe breaks in this area is that they are substantially lower in energy than breaks in the Intermediate or Turbine Building. No lines in this area larger than 1 inch
TMI-1 UFSAR APPENDIX 14A 14A-9 REV. 22, APRIL 2014 carry fluids that exceed both 200ºF and 275 psig. The largest line in this area with postulated crack breaks is 14 inches versus 24 inches in the Intermediate or Turbine Building.
All break locations in the auxiliary area will meet the established criteria (one string of each engineered safeguards system will remain operable). The break will not damage both strings of any engineered safeguards system due to the separation of engineered safeguard components. In a few cases where cabling for both trains of an engineered safeguards system runs in the vicinity of a postulated break, the effects on the cabling will be analyzed and, if necessary, protected by a barrier or rerouted as required.
The capability to detect and sustain the flooding as a result of breaks in the auxiliary area is discussed in Section 6.4.5 of the FSAR.
Postulated high energy sample line breaks will not affect any engineered safeguard/reactor protection system equipment nor render any vital building areas permanently uninhabitable.
The HVAC has been designed to confine the consequences of reactor coolant small line breaks to relatively small areas of the Fuel Handling Building at elevation 281 ft 0 inch and the radioactive sample station in the Control Room tower which are of no immediate importance in the event of a high energy line rupture. The radiation monitoring system and Control Room controls of the Reactor Building isolation valves will be used to terminate any postulated high energy sample line breaks.
The letdown line to the makeup system normally operates below 200F and is therefore subject only to crack breaks according to the criteria prescribed in Section 2.0 of this Appendix.
However, a full diameter break would result in flow sufficiently high to render the coolers ineffective and the temperature of the flow from the break would exceed 200F. To assure sufficient conservatism, full size breaks have been postulated in the letdown line. For a break downstream of the breakdown orifice, the flow would be slightly above normal and the effectiveness of the cooler inside containment would not be significantly reduced.
Analyses performed by GPUN (Reference 17) determine postulated line break locations and address the effects of those breaks on the letdown line and nearby equipment due to jet impingement. The result of the evaluation show that the letdown line is structurally adequate to ensure that pipe whip is not an issue during a postulated line break event and that the pipe will remain stationary. Furthermore, jet impingement from the postulated rupture points does not pose a threat to any other safety related system in the plant. The compartment pressurization resulting from the break has been analyzed and found to be below that which would cause damage to the building.
The temperature, pressure, and humidity environment in the vicinity of the break has been determined by analysis (References 18, 19). The safeguard equipment in the vicinity of the break consists of valves, cabling, and pumps. The valves and cabling have been qualified for the environment expected to result from the postulated break. The pumps are in separate compartments and are not affected by the local environment. The potential flooding from a postulated letdown line break does not result in a water level high enough to impact safety related equipment or circuits.
A postulated break will be automatically isolated by closure of the Reactor Building isolation valves when high temperature is sensed in the letdown line.
TMI-1 UFSAR APPENDIX 14A 14A-10 REV. 22, APRIL 2014 Since the Makeup System in the auxiliary building is not a High Energy Line per the criteria of Section 2.0, a High Energy Line Break (HELB) need not be postulated. Therefore, this is not a Design Basis Event for TMI-1 and 10 CFR 50.49 does not apply. The environmental evaluation was performed to respond to initial licensing questions but is not in the scope of 10 CFR 50.49 (Reference 20, 21).
4.4 DIESEL GENERATOR BUILDING The Diesel Generator Building has no direct doorway or access connection to either the Intermediate Building or the Turbine Building. However, ventilation air comes from the machine shop. Steam leakage into the machine shop would be minimal as closed doorways separate the machine shop and Turbine Building. Also, the machine shop has a separate air supply and exhaust system that would dissipate any steam leakage.
Prolonged steam conditions in this space are not to be expected. The north diesel room has no direct doorway, access, or ventilation connection to either the Intermediate Building or Turbine Building or any other area expected to become steam laden following the pipe break.
Therefore, the north diesel room ambient atmosphere will not experience any change as a result of the pipe break.
4.5 COLD SHUTDOWN CAPABILITY The adequacy of the borated water storage tank as an interim heat sink for the Three Mile Island Nuclear Station Unit-1 Reactor Coolant System has been evaluated for the following set of assumptions:
a.
Steam line break occurs inside the Intermediate or Turbine Building during rated power operation b.
Loss of all feedwater to both steam generators occurs d.
Loss of offsite power occurs In addition to this set of assumptions, this evaluation is valid for any situation where Reactor Coolant System energy removal through the steam generators is no longer available because of a HELB in containment.
There are three primary areas of concern for this condition. These areas are prevention against core uncovering, protection against excessive Reactor Building pressure, and the ability to achieve cold shutdown conditions.
The B&W digital computer code CRAFT, Reference 10, has used to determine the characteristics of this accident with regard to core uncovering and mass energy releases to the containment. The mass and energy release data from CRAFT were used in the digital computer code CONTEMPT, Reference 11, for Reactor Building pressure calculations. The assumptions and results of the analysis are summarized in Table 14A-6. A single steam generator blowdown was considered as the most conservative case since, for a double
TMI-1 UFSAR APPENDIX 14A 14A-11 REV. 22, APRIL 2014 blowdown, the high pressure injection (HPI) pump would be started almost instantaneously on low Reactor Coolant System pressure actuation, meaning a lower probability of core uncovering.
Core uncovering is prevented by pumping water from the borated water storage tank via the Makeup and Purification System (HPI) into the Reactor Coolant System. With one makeup and purification (HPI) pump started 15 minutes after the break, the minimum coolant level in the reactor vessel occurs at approximately 140 minutes and at no time falls below the top of the core. Operator action is assumed to occur 15 minutes after the break in starting the makeup and purification pump (HPI).
The building pressure increases during the transient as boiloff occurs through the pressurizer safety valves (2515 psia). Assuming the boiloff goes directly to the building atmosphere with no credit for the quench tank, the building pressure reaches the Reactor Building cooler and high pressure injection set point (4 psig) 38 minutes after the break. With one building cooler operative at this time, the building pressure reaches a maximum value of 24 psig and never exceeds the design pressure limit. Furthermore, the reactor spray actuation set point (30 psig) is not reached, and a single building cooler provides adequate protection throughout the transient against excessive Reactor Building pressure.
High pressure injection of BWST water continues until the BWST is depleted (approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, assuming one HPI pump is operating). At this time, further cooldown is achieved by using the decay heat (low pressure injection) pumps drawing from the Reactor Building sump to supply suction to the makeup and purification (HPI) pumps. The sump recirculation continues until the Decay Heat Removal System (LPI) can be actuated to reduce the system to cold shutdown. Cold shutdown is then achieved by venting the system pressure and actuating the Decay Heat Removal System to recirculate the reactor coolant through the decay heat coolers.
TMI-1 UFSAR APPENDIX 14A 14A-12 REV. 22, APRIL 2014 5.0
SUMMARY
AND CONCLUSIONS The results of this design review are summarized as follows:
a.
A rupture of the high energy piping systems is considered highly unlikely. The systems other than auxiliary steam piping have been conservatively designed in accordance with the criteria in USAS B31.1.0, Code for Power Piping. Materials, fabrication, and quality assurance requirements of the Code have been utilized. In addition, the main steam piping has been subject to 100 percent radiography of welds from the steam generators to the turbine stop valves, and the feedwater piping has been subject to 100 percent radiography from the steam generators to the feedwater pumps. Quality assurance provisions of USAS B31.7, Code for Nuclear Piping, have been implemented for nuclear systems. The auxiliary steam piping system has been shown to be adequate for combined effects of operating conditions and the SSE. Quality assurance provisions are implemented to ensure configuration of the auxiliary steam piping system is maintained.
b.
The Atomic Energy Commission criteria had been implemented in identifying postulated break locations in high energy piping systems.
c.
All of the equipment required for shutdown is protected from the postulated ruptures by virtue of location (remote from high energy lines), restraints, inspection, or barriers.
d.
The Control Room will remain habitable and operable following a postulated high energy pipe break due to its remote location from such breaks.
e.
The plant can be brought to cold shutdown conditions by utilizing either feedwater system or the Emergency Core Cooling System and Reactor Building cooling.
f.
This analysis shows that the borated water storage tank serves as an adequate interim heat sink from the standpoint of core covering, Reactor Building pressure considerations and achieving cold shutdown.
g.
Further, the analysis shows that the core remains covered, the Reactor Building is protected against excessive pressure, the plant can be taken to cold shutdown, and containment integrity can be maintained.
After Cycle 17, the original OTSGs were replaced with enhanced OTSGs. The existing HELB-related analyses and evaluations contained in Appendix 14A were reviewed in Reference 124 and confirmed to remain applicable following the TMI SG replacement.
TMI-1 UFSAR APPENDIX 14A 14A-14 REV. 20, APRIL 2010 6.0 INSPECTION REQUIREMENTS It is intended to ultrasonically inspect the welds identified on Figures 6 and 9 in compartments 2 and 5 to the extent possible. This inspection will be performed periodically as required by Technical Specification 4.15.
TMI-1 UFSAR APPENDIX 14A 14A-15 REV. 21, APRIL 2012 7.0 METHODS OF ANALYSIS AND DESIGN 7.1 THERMAL-HYDRAULIC ANALYSIS A detailed thermal-hydraulic analysis was performed in order to evaluate the consequences of a postulated high energy line rupture in the Intermediate Building. The analysis included the calculation of system transients and blowdown loads as well as compartment pressurization for use in the structural evaluation.
7.1.1 BLOWDOWN ANALYSIS 7.1.1.1 Main Steam Blowdown Analysis Selection of Line Breaks A total of Two (2) double ended guillotine pipe breaks in the Main Steam System assumed at the worst locations from an equipment qualification viewpoint; i.e., breaks are located in areas where the environmental response was most severe. Piping stress levels were not considered as an input to break location.
The selected break located at the 322 ft elevation is a guillotine break of a 24 inch main steam line upstream of the main steam line isolation valve. Guillotine breaks give the maximum total flow and most severe conditions in the building.
The location of the worst case break for the steam supply line to the EFW pump turbine is at the 295 ft elevation and is a guillotine break of the steam supply line.
Computer Model An EDSFLOW Computer model was constructed for the Intermediate Building and was utilized to evaluate the main steam line breaks. The building was nodalized as volumes. Portals such as doorways and stairwells were modeled as connecting pathways between the volumes.
The Intermediate Building was nodalized into sixteen volumes. The 295 ft and 305 ft elevations were more precisely modeled since most of the safety related electrical equipment is located in these spaces, and a more accurate prediction of the pressure and temperature response is required. The area of the walls and ceiling of each compartment was calculated and modeled as an exposed concrete surface to simulate their heat sink effect.
The heat transfer coefficient (h) to the walls was calculated by the EDSFLOW Computer Code to be 5 Btu/F-hr-ft2 throughout the transient for the 24 inch main steam line break case. The heat transfer coefficients calculated internal to the code as a function of transient time and is based upon fundamental natural and forced convection heat transfer correlations. This value is conservative since the effect of wall condensation was not considered. In accordance with the Uchida Condensing Heat Transfer Correlations, heat transfer coefficients substantially higher (up to a maximum of 280 Btu/F-hr-ft2 at an air to steam ratio of 0.1) than the value calculated could have been used shortly after break initiation. For the steam supply to the emergency feed pump turbine line break case a heat transfer coefficient of 50 Btu/F-hr-ft2 was utilized during blowdown, and 5 Btu/F-hr-ft2 after blowdown was terminated. These values are well
TMI-1 UFSAR APPENDIX 14A 14A-16 REV. 21, APRIL 2012 below the actual condensing coefficients. The results of the analysis, therefore, are considered conservative in nature.
The thermal properties of concrete used in this analysis are listed in Table 14A-9.
The blowdown mass and energy release rates for the 24 inch line break were based on a circumferential break. The mass and energy release rates are included in Table 14A-8.
The blowdown mass and energy release rates for the 12 inch line break were calculated utilizing a modified Darcy's equation. The mass and energy release rates are 512 1bm/sec-ft2 and 6.154 x 105 Btu/1bm-ft2, respectively. It is conservatively assumed that this blowdown is constant for ten minutes without depressurizing the steam generators. After ten minutes, it is assumed emergency feedwater to the faulted OTSG would be terminated by the operator.
Key Analytical Assumptions The following assumptions and approximations were used in this analysis:
1.
Initial Intermediate Building pressure, temperature and humidity were assumed to be 14.7 psia, 90F, and 60 percent, respectively.
2.
All steam lines were assumed to be at their maximum operating pressures and temperatures.
3.
Forward and reverse flow shock losses were accounted for where junction flow areas were reduced or expanded in relation to the adjacent volume flow areas.
4.
For the EFW pump turbine steam supply line break, doors are assumed to be in their most conservative position at the initiation of the event. If the door is closed, it is assumed to open at a differential pressure of 2 psid when opening away from the door jamb, and 4 psid when opening against the door jamb.
5.
The analysis was performed without the 92 ft2 vent area at the 322 ft elevation available.
There is 75 ft2 total vent area from the building when the opening at the 322 ft elevation is not assumed available.
6.
Operator action to terminate EFW flow to the faulted OTSG feedwater is required to mitigate the EFW pump turbine steam supply line break within ten minutes.
7.
The steam vent pathway out of the Intermediate Building is assumed available after a 2 psid pressure buildup in the building.
7.1.1.2 Feedwater Blowdown Analysis The RELAP-3 digital computer code, Reference 2, was used in analyzing the feedwater blowdown transients for the determination of thrust loads. The system was represented by an assemblage of control volumes connected by flow paths or junctions. The effects of valves, pumps, heat exchangers, and check valves are included in the code.
TMI-1 UFSAR APPENDIX 14A 14A-17 REV. 21, APRIL 2012 The steam generators were modeled so that the feedwater inlet nozzles were above the steam generator water level. In this representation, backflow through the inlet nozzles would be steam.
The feedwater lines were divided into several volumes for each case. The volumes were selected so that volume size and junction location would provide optimum system representation for the particular case being analyzed.
It was assumed that for the durations of these analyses, the feedwater pumps would continue to operate and that flow would be a function of head. It was further assumed that for the duration of these analyses, an unlimited constant pressure supply of water was available at the feedwater pump suction. Both main feedwater pumps were combined and modeled as a single pump.
In modeling flow nozzles, the actual nozzle throat area was used if flow to the leak was in a forward direction through the flow nozzles. Where flow to the leak was in a reverse direction through the nozzle, an effective flow area was calculated and used in the model.
Basic Assumptions a.
Reactor operation at full load conditions.
b.
Steam generator nominally at 925 psi with feedwater inlet at 1000 psi and 462oF.
c.
Feedwater line check valve fails to close.
d.
Pumps do not trip.
e.
Circumferential and longitudinal breaks were considered.
f.
Break volumes were selected to account for the segment of piping up to the first elbow on either side of the break.
7.1.2 COMPARTMENT PRESSURIZATION ANALYSIS 7.1.2.1 Main Steam Line Ruptures The pressure temperature transients resulting from the postulated rupture of a main steam line in the Intermediate and Turbine Buildings were investigated.
The transients for subcompartment integrity evaluation were calculated by extending the short term main steam blowdown model to include control volumes representing compartments of the Intermediate Building with their interconnecting vent area. Figure 14A-16 presents the Intermediate Building compartment designations. Double-ended circumferential ruptures were considered as the limiting case. Table 14A-4 presents the maximum wall and slab pressure differentials obtained for the Intermediate Building subcompartments where over-pressurization was investigated as a potential problem.
TMI-1 UFSAR APPENDIX 14A 14A-18 REV. 21, APRIL 2012 7.1.2.2 Feedwater Line Ruptures With its significantly lower energy content, the feedwater line rupture does not represent a problem with respect to compartment pressurization.
7.1.3 PIPE THRUST AND JET IMPINGEMENT The thrust forces developed by a jet flow being expelled from a ruptured pipe and calculated from the internal pipe pressure and the density of the mass being accelerated out of the rupture area. The conditions used to calculate this thrust assume that the rupture occurs over 1 millisecond with the maximum thrust force being established by using the corresponding pressure and mass flow conditions. As the blowdown of the system progresses, the pipe stagnation pressure is assumed to be equal to the static pressure at the rupture location for both longitudinal breaks and for circumferential breaks. The thrust calculation used for these analyses includes the integrated pressure and momentum effects and does not take advantage of upstream flow restrictions to reduce the pipe pressure.
Thrust forces on those lines not affected by blowdown characteristics during the first 15 seconds were assumed as 1.26 PA for steam and 2.0 PA for subcooled fluid.
The forces on targets in the path of escaping fluids are dependent on the size and shape of the target and its distance from the rupture area. Since the actual shape of the rupture dictates the flow field shape being generated, it was assumed that a typical rupture is circular and the free stream expansion of the jet to be conical with an included angle of 30 degrees. Calculation of the force on an object was determined by assuming that the dynamic pressure developed at the rupture exit is for the maximum mass flow and pressure conditions in the high energy line and is inversely proportional to the cross section area of the conical expansion being generated. This dynamic pressure was applied on the targets assuming a target drag coefficient of 2.0, i.e.,
complete stagnation of the escaping fluid.
7.2 STRUCTURAL ANALYSIS AND DESIGN 7.2.1 WALL AND SLAB CAPACITIES Postulated ruptures of the main steam lines at the containment penetrations were investigated to determine differential pressures in compartments 2 through 5. Peak differential pressures are presented in Section 7.1.2. To investigate the retention or loss of structural integrity due to breaks within the small compartments, the load carrying capacity of slabs and walls was calculated and compared to expected differential pressure.
7.2.1.1 General Yield-Line Theory and Shear Capacity Yield-Line Theory (References 3, 4, and 5) takes into consideration the inelastic behavior of the reinforced concrete structural element (wall or slab) in developing a mechanism prior to loss of structural integrity. In brief, the steps involved in the evaluation of the uniform pressure capacity of slabs and walls are as follows:
TMI-1 UFSAR APPENDIX 14A 14A-19 REV. 21, APRIL 2012 a.
The ultimate moment capacity of the cross section is calculated at various negative and positive moment regions, i.e., at the boundaries and at midspan. The sections are typically doubly reinforced and the moment capacity is given conservatively by:
Mu = 0.9AsFy (1d - a/2) where:
Mu = moment capacity/ft As = area of tension steel/ft Fy = yield strength of reinforcing steel d = effective depth from steel centroid to extreme compression fiber a = depth of equivalent compression concrete The estimate is conservative in that it neglects the increase in strength of the tension steel after yielding occurs. Also, section capacities are reduced by a factor of 0.9.
b.
The support conditions for the slab or wall under consideration are determined. Where the supports of a given wall or slab are connected to slabs or walls of approximately equal thickness and when the reinforcing steel is sufficiently anchored to develop its strength, the support condition is taken as being fixed. Where the above conditions are not satisfied, the support condition is assumed to be simple support or free in the case of free edges adjacent to containment.
c.
The correct yield-line pattern is established by trial and error using the Principle of Virtual Displacements or The Equilibrium Method. Sufficient trails are executed to determine the uniform load-carrying capacity (absolute minimum of all trials). A 10 percent reduction is applied to all results to allow for corner effects.
The uniformly distributed load corresponding to the correct yield-line pattern is taken to be the differential pressure capacity of the wall or slab in bending.
d.
Each wall or slab is checked for punching shear (two way action) around its periphery (the perimeter is defined at a distance d/2 from the support lines) and for local shear (one way action) at a distance d from support faces.
The concrete shear strength for two way action is taken as 4 f c and the slab capacity is calculated by:
Wps = 4 f c bod Ap where:
Wps = pressure-producing punching shear capacity
TMI-1 UFSAR APPENDIX 14A 14A-20 REV. 21, APRIL 2012 b8 = perimeter length d = effective depth Ap = slab or wall surface area bounded by b The concrete shear strength for one way action is taken as:
Vc = 1.9 c
f '
+ 2500 P Mu Vd 3.5 c
f '
as defined in Reference 6. The slab capacity is then calculated by considering the most critical section at d distance from a support face.
The wall or slab capacity as controlled by shear is taken as the lesser of the two values described above.
e.
The maximum differential pressure capacity is then the lesser of the two values from c.
(as controlled by moment) and d (as controlled by shear).
7.2.1.2 Compartment Differential Pressures Versus Capacities The differential pressure results defined in Section 8.1.2 for circumferential breaks in compartments 2 through 5 are presented in Table 14A-5. A comparison of peak differential pressures to the wall or slab capacities indicates the following:
a.
Structural slabs and walls in compartments 2 and 5 may experience excessive differential pressures due to postulated breaks. Protection will be provided as described in Section 4.2.d.1 of this Appendix.
b.
Postulated breaks in compartments 3 and 4 should not result in wall or slab failure and no protection is required.
7.2.1.3 Jet Impingement And Combined Differential Pressure Jet impingement is a direct result of either a longitudinal or circumferential pipe break.
Longitudinal pipe rupture, which can occur at any orientation about the circumference of the pipe at the break point, results in a jet axis perpendicular to the longitudinal axis of the pipe.
Circumferential (guillotine) breaks result in a jet axis parallel to the longitudinal axis of the pipe.
The jet is assumed to diverge from the break at a 30 degree conical angle, and the total integrated force on an object is determined by assuming that the dynamic pressure developed at the rupture exit is inversely proportional to the cross sectional area of the conical expansion cone at any station. Thus, the walls and slabs are evaluated for the effects of the jet.
After investigating various postulated pipe ruptures, the most severe case of jet impingement from a main steam break in compartments 3 and 4 is a longitudinal break acting on the wall between compartments 4 and 5. This is due to the fact that the main steam pipe is approximately 2 feet away from the wall, and the jet impingement force from a postulated side split break at the containment vessel penetration strikes the wall at its unsupported edge. Jet
TMI-1 UFSAR APPENDIX 14A 14A-21 REV. 21, APRIL 2012 loads in compartments 2 and 5 were not investigated since protection against those breaks is provided as stated in Section 4.2.d.1 of this Appendix.
The capability of the wall between compartments 4 and 5 to resist the jet impingement force was analyzed using Yield-Line Theory. The jet impingement load was conservatively taken as a point load acting at the free edge of the wall. This analysis assumed a semicircular fan-shaped crack pattern and resulted in an ultimate capacity approximately three times higher than the jet force.
The wall was then evaluated for its ability to resist the punching shear caused by the jet impingement. The punching shear capacity of the wall was found to be approximately four times the punching shear stress caused by the jet impingement force.
If the impingement force is assumed to be instantaneous and a dynamic load factor of two is used, both the moment and shear checks stated above are more than satisfactory. The jet load in this case was not combined with pressure since the peak jet force occurs at the instant of break and the peak pressure occurs later.
The wall between compartments 4 and 5 was also analyzed for the combined effects of jet impingement and pressurization at a time subsequent to pipe break. The ultimate moment capacities of the wall needed to resist the pressurization load and the jet impingement was less than the available ultimate moment capacity of the wall. The technique for combining the two loadings is described in Reference 5. In a similar manner, the combined shear stress caused by pressurization and jet impingement was found to be less than the available shear capacity of the wall. All other cases of combined pressurization and jet impingement in compartments 3 and 4 were less critical than the case discussed.
7.2.2 RESTRAINT DESIGN PROCEDURE 7.2.2.1 Assumptions The assumptions involved in designing restraints are as follows:
a.
Guaranteed minimum yield strength of pipe steel reduced in accordance with operating temperature.
b.
Guaranteed minimum ultimate strength of the pipe steel is unaffected by temperature.
c.
Ultimate strain of both piping and restraint material is one half of guaranteed minimum percent elongation.
d.
Guaranteed minimum values of yield strength and ultimate strength for restraint material A 36 are taken from applicable ASTM specifications.
e.
A 10 percent increase in material properties is applied to allow for strain rate effect.
The design procedure discussed in the following section utilizes computer program DYREC (Reference 7). Final designs of restraints were investigated using the techniques presented.
TMI-1 UFSAR APPENDIX 14A 14A-22 REV. 21, APRIL 2012 7.2.2.2 Analytical Method The dynamic analysis of lumped mass models of the rupturing pipe and restraint system was performed by direct numerical time integration of the equations of motion. The computer program DYREC includes the following capability:
a.
Element types; bilinear beam, bilinear axial or rotational spring, "special" axial or rotational spring.
b.
Plastic, elastic, or elasto-plastic impact after closing specified gaps.
c.
Constant zero or non-zero nodal boundary conditions.
d.
Piecewise linear force-time histories.
7.2.2.3 Material Properties Pipe Material - A 106 Grade B Fy = Yield strength at 600F
= 25.9 ksi x 1.1 = 28.5 ksi Fult = Ultimate strength
= 60 ksi x 1.1 = 66.0 ksi
% Elongation
= 22 percent E = Modulus of elasticity
= 25.7 x 103 ksi Structural Steel - A36 Fy = 36 ksi x 1.1
= 39.6 ksi Fult = 60 ksi x 1.1
= 66.0 ksi
% Elongation
= 20 percent E = Modulus of elasticity
= 30 x 103 ksi 7.2.2.4 Section Properties a.
Pipe Cross Section The bilinear moment versus curvature relationship is defined by the following points:
My
= FyS Mult
= FyZ + (Fult-Fy )S Where:
TMI-1 UFSAR APPENDIX 14A 14A-23 REV. 21, APRIL 2012 S
= Elastic Section Modulus Z
= Plastic Section Modulus Ey = F/E = Yield Strain Eult = (% elongation)/2 = Ultimate Strain
Øy = 2 Ey/d = Yield Curvature
Ø ult = 2 Eult/d = Ultimate Curvature Where:
d = O.D. of pipe cross-section The shear versus shearing strain relationship for pipe cross section is defined by:
Fsy
= Shear Yield Stress* = Fy/3 As
= Effective Shear Area 8 = 0.53 A pipe Yy
= Shear Yield Force
= As Fy G = Shear Modulus
= 0.4E Yy
= Shear Yield Strain = Yy/(As G)
Yult = Ultimate Shear Force
= 2Yy Yult = Ultimate Shear Strain
= 309
- Von Mises Criteria b.
Structural Wide Flange Cross Section In defining the moment versus curvature relationship and shear versus shearing strain, the moment is assumed to be carried by the flanges and the shear by the web.
My
= Fy Af (d-tf)
Where:
Af
= Area of One Flange d
= Depth of Cross-Section tf
= Flange Thickness
TMI-1 UFSAR APPENDIX 14A 14A-24 REV. 21, APRIL 2012 Mult
= Fult a (d-tf)
Ey
= 2 Ey/d Eult
= 2 Eult/d Yy
= Fsy A Web Vy
= Vy/(Aweb G)
Vult
= 2 Vy Vult
= 30%
The bilinear moment versus curvature and shear versus shearing strain curves are illustrated on Figure 14A-19.
7.2.2.5 Example Model The pipe condition at rupture, shown on Figure 14A-20 part A, is representative of a postulated break in the 12 inch main steam header. Figure 14A-20 B part illustrates the dynamic model of the same condition.
The restraint is positioned so as not to interfere with normal piping operation. Therefore, the model has a gap of 4 inches between nodes 3 and 8 at the instant F(t) is applied, i.e., at the time postulated rupture occurs. Node 8, in this case, is representative of the mass of the header that is assumed to break away. The modeling is conservative in that credit is not taken for the restraint offered by the 12 inch branch line coming into the 12 inch header.
Nodes 1 through 7 and bilinear beam elements 1 through 6 represent two structural wide flange sections embedded in the existing slab.
The fineness of lumping as well as the magnitude of time increment for numerical integration are selected to ensure a reasonable approximation of the dynamic transient. The total time of execution of the mathematical model on the computer is set to allow multiple impacts of the rupturing pipe and observance that the selected restraint is indeed bringing the rupturing pipe to a stable steady state condition.
7.2.2.6 Detail Design The details of the restraint design are considered in two parts, steady state and transient effects.
The restraint is proportioned such that after the transient occurs, the shears, moments, reactions, and so forth, are within the allowable values of applicable codes (e.g., AISC, manual of Steel Construction, 7th Edition). During the transient, the maximum element curvatures and shears are limited to approximately one half of their ultimate values.
TMI-1 UFSAR APPENDIX 14A 14A-25 REV. 18, APRIL 2006
8.0 REFERENCES
1.
Redfield, J. A., Murphy, J. H., and Davies, V. C., "FLASH-2 A Fortran IV Program for the Digital Simulation of a Multimode Reactor Plant During Loss-of-Coolant,"
WAPD-TM-666 (April 1967).
2.
Rettig, W. H., Jayne, G. A., Moore, K. V., Slater, C. E., Uptmor, M. L., "RELAP 3 - A Computer Program for Reactor Blowdown Analysis," IN-1321 (June 1970).
3.
Ferguson, P. M., "Reinforced Concrete Fundamentals," Second Edition, John Wiley &
Sons, Inc., New York, New York, 1965, pp. 318-348.
4.
Hognestad, E., "Yield-Line Theory for the Ultimate Flexural Strength of Reinforced Concrete Slabs," Journal of the American Concrete Institute, Volume 24, No. 7, March 1953.
5.
Wood, R. H., "Plastic and Elastic Design of Slabs and Plates," The Ronald Press Company, New York, 1961.
6.
"Building Code Requirements for Reinforced Concrete, ACI Standard 318-71, American Concrete Institute, Detroit, Michigan, 1970.
7.
DYREC, Dynamic Response Calculator, January 1973, Gilbert Associates Computer Program (Company Proprietary).
8.
Cowper, G. R., "The Shear Coefficient in Timoshenko's Beam Theory," Journal of Applied Mechanics, June 1966.
9.
Hall, W. J. and Newmark, N. M., "Shear Deflection of Wide Flange Steel Beams and the Plastic Range," Journal of Engineering Mechanics Division, ASCE, Vol. 81, October 1955.
10.
CRAFT - Description of Model for Equilibrium LOCA Analysis Program, B&W Topical Report BAW-10030.
11.
Richardson, L. C., Finnegan, L. J., Wagner, R. J., and Waage, J. M., "CONTEMPT," A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of-Coolant Accident, Phillips Petroleum Co., Atomic Energy Division, Idaho Falls, Idaho, AEC Research and Development Report TID-4500, issued June 1967.
12.
Atomic Energy Commission (AEC) Request, General Information Required for Consideration of Effects of Piping System Break Outside Containment," December 15, 1972.
13.
GPU Service Technical Data Report - TDR 250 "Review of Intermediate Building Flooding Following a Feedwater Line in the Intermediate Building of TMI Unit 1.
14.
GPUNC letter to NRC, #5211-84-2193 dated August 1, 1984.
Subject:
Intermediate Building Flooding Modification.
TMI-1 UFSAR APPENDIX 14A 14A-26 REV. 18, APRIL 2006 15.
USNRC Standard Review Plan Section 3.6.2 - Determination of Rupture Locations and Dynamic Effects associated with Postulated Rupture of Piping, Rev. #1, July 1981.
16.
USNRC Branch Technical Position MEB3 Postulated Rupture Locations in Fluid System Piping Inside and outside Containment, Rev. #1, July 1981.
17.
GPUN Calculation C-1101-211-E540-076, Letdown Line Break Structural and Dynamic Assessment, Rev. 1.
18.
GPUN Calculation C-1101-211-E540-077, RELAP5 Analysis of Letdown Line Break in the Auxiliary Building, Rev. 0.
19.
GPUN Calculation C-1101-211-E540-078, GOTHIC Analysis of Auxiliary Building EQ Environment from a Letdown Line Break, Rev. 0.
20.
NRC Internal Memorandum (Marsh to Milano) dated February 6, 1997,
Subject:
Three Mile Island, Unit 1 - Licensing Basis for Letdown Line Pipe Break Outside Containment.
21.
NRC Letter (Wiggins to Langenbach) dated October 15, 1998,
Subject:
NRC Inspection Report No. 50-289/98-06 and Notice of Violation. GPUN File No. 1920-98-30622.
TMI-1 UFSAR APPENDIX 14 A14A-27 REV. 21, APRIL 2012 TABLE 14A-1 (Sheet 1 of 2)
High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures Description of Line
- of lines Nom Line OD, in.
Type of Fluid Contained Normal Operation Main Bldg Areas Occupied by these lines Remarks Temperature oF Pressure psig Intermediate Bldg Elev.
Turbine Bldg Elev.
Auxiliary Bldg Elev.
Fuel Handling Bldg Elev.
Main Steam (MS) 4 24 Superheated Steam 570 900 337 7 & 347 1 347 1 None None None Main Steam Dump 2
12 Superheated Steam 570 900 346 0 345 8 & 342 9 None None None MS supply to EFW Pump Turbine and MS Dump Lines 4
8 indr takeoffs Superheated Steam 570 900 338 8 None None None 8 inch takeoffs rise off MS Line in small Intermediate Building cubicles and drop vertically into 10 inch headers that terminate at end of main headers 2
10 Sub header
~ 314' 1
12 header 297 0 Main Feedwater (FW) 2 20 FW Pump Disch Hot Water (subcooled) 455
>950 346 0 to 324 0 &
316 2 30 - 324 3 None None Pressure varies due to HD loss in piping system. One 20 inch discharge from common 30 inch manifold is in Int.
Building.
The rest are in Turbine Building 1
30 Common Man.
20 - 346 0 & 324 0
2 20 Outlets from common man.
Aux Steam (AS) to EFW Pump Turbine 1
4 Saturated steam 382 185 344 0 to 303 6 None None None None Emergency Feedwater Pump Discharge (EFW) 1 6
Cold Water 90
>1000 304 0 & 306 6 None None None The 3 EFW pump discharge lines tie into common 6 inch manifold and two 6 inch outlet lines feed out of this into the Reactor Building.
2 4
297 7 & 306 6 2
6 298 3 & 318 0 Aux Steam (AS) to Auxiliary Building 1
4 Saturated steam 227 5
None 365 10 289 2 to 288 8 363 6 & 362 6 to 329 0 In passageway between Fuel Handling to Building and Control Tower.
One 2 inch branches off within Fuel Handling Building at 339 ft.
1 6
366 7 & 356 6 288 8 to 294 1 329 0 to 328 8 1
8 356 6 & 363 7 294 1 to 296 0 301 3 to 289 2
TMI-1 UFSAR APPENDIX 14 A14A-28 REV. 21, APRIL 2012 TABLE 14A-1 (Sheet 2 of 2)
High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures Description of Line
- of lines Nom Line OD, in.
Type of Fluid Contained Normal Operation Main Bldg Areas Occupied by these lines Remarks Temperature oF Pressure psig Intermediate Bldg Elev.
Turbine Bldg Elev.
Auxiliary Bldg Elev.
Fuel Handling Bldg Elev.
Makeup Pump Discharge 3
4 Pump Disch Cold Water 135
>2250 None None 287 9 to 284 3 291 0 Refer to Figure 3 for details of pipe routing of discharge piping from makeup pumps 1
4 Comm Manf 3
4 Manf Outlet Letdown to Makeup &
Purification 1
2 1/2 Cold Water 120
~ 2200 None None 291 6 289 0 282 0 Splits into 3 branch manifolds where pressure is broken down to between 25 and 75 psig Decay Heat Suction 1
12 RC LETDOWN Water varying from hot to cold 250 to between 110 & 140 300 to atmospheric plus static head None None 291 6 263 0 None 2
12 BRANCH 2
14 PUMP SUCTION Decay Heat Discharge 2
10 Water varying from hot to cold 210 to between 100
& 130
~ 400
~ 150 None None 272 9 274 0 291 6 None None Reactor Coolant Sample 1
3/8 Hot water (subcooled) and steam Between 540 &
605 Steam is saturated
~ 2200 None None 312 0 Between Reactor Bldg and FH Bldg Line passes through passageway between FH Building and Control Tower and enters Control Tower as same place as Steam Gen. sample, below Steam Generator Secondary Side Sample 2
3/8 Hot water (subcooled)
~ 500
~ 900 None 315 6 None 315 6 In passageway between FH Bldg and Control Tower Line enters radioactive sample station in room at elevation 306 feet 0 inch near northwest corner of Control Tower
TMI-1 UFSAR APPENDIX 14 A14A-29 REV. 21, APRIL 2012 TABLE 14A-2 (Sheet 1 of 2)
SYSTEMS REQUIRED FOR SAFE SHUTDOWN FOLLOWING A POSTULATED MAIN STEAM OR FEEDWATER PIPE RUPTURE System Building Equipment Location Elevation Area Makeup and Purification System Auxiliary 281-0 Northeast section Decay Heat Removal Systems D.H. Pumps and Coolers Auxiliary 261-0 North-central section Closed Cycle Cooling System Coolers Auxiliary 271-0 West section Closed Cycle Cooling Pumps section Auxiliary 305-0 North-central River Water Pumps Screen House 308-0 East-central section Borated Water Storage Tank (BWST)
Outdoors 306-0 Near northeast corner of Auxiliary Building Nuclear Services Cooling Water System Closed Cycle Cooling Coolers Auxiliary 271-0 West section Closed Cycle Cooling Water Pumps Auxiliary 305-0 North-central section River Water Pumps Screen House 308-0 East-central section Air Handling (ES)
Reactor Building Cooling Units Reactor 287-0 East-central section Cooling Water Piping Intermediate 295-0 North-central section Reactor Building Coolers Reactor 291-0 West end Diesel Generators Diesels Diesel 306-0 Most of Bldg.
area Fuel Tanks Underground Outside Below grade North of Diesel Generator Building TABLE 14A-2
TMI-1 UFSAR APPENDIX 14 A14A-30 REV. 21, APRIL 2012 (Sheet 2 of 2)
SYSTEMS REQUIRED FOR SAFE SHUTDOWN FOLLOWING A POSTULATED MAIN STEAM OR FEEDWATER PIPE RUPTURE Equipment Location System Building Elevation Area Spent Fuel Cooling System Fuel Handling 281-0 Central west side Pressurizer Safety Valves and PORV Reactor 354-0 On top pressurizer RPS and ES Actuation Reactor and Auxiliary RPS and ES Actuation Racks Control Emergency Feedwater Intermediate 295-0 East section Atmospheric Dump Valves Intermediate 295-0 EF-P-1 &
Bypass Header Room
TMI-1 UFSAR APPENDIX 14 A14A-31 REV. 21, APRIL 2012 TABLE 14A-3 (Sheet 1 of 1)
TABULATION OF SYSTEMS AND BREAK LOCATIONS Primary +
Secondary Secondary Stress, % of Figure No.
Break
- Street, Allowable and Locations
% of (deadweight Reference Indicates on Allowable
+ Thermal System Isometrics Isometrics (Thermal)
+/-Seismic)
Generator "B" Figure 14A-11 29 63.9 57.5 to Header 43 end 7.4 21.2 at EL 297'-0" 13 end 6.7 59.3 12 20.6 2.8 10 29.7 43.6 9 end 32.4 45.5 Feedwater ISO 13 4.3 36.5 from 30" Header 28 27.0 49.1 to Penetration Figure 14A-12 30 28.9 50.6 103 44 end 22.3 55.4 Feedwater ISO 1 end 50.7 49.4 from 30" Header 12 27.8 46.1 to Penetration Figure 14A-13 8
36.7 44.2 227 53 end 42.3 51.7 Main Stream Figure 14A-1 1 end 00.7 21.2 Dump to 116 89.0 95.8 Intermediate 115 87.3 88.8 Wall 117 end 44.2 59.7 Main Stream Figure 14A-1 1 end 43.5 53.8 Dump to 29 72.3 77.7 Intermediate 28 71.0 77.2 Wall 43 end 27.8 46.0
TMI-1 UFSAR APPENDIX 14 A14A-32 REV. 21, APRIL 2012 TABLE 14A-4 (Sheet 1 of 2)
MAXIMUM COMPARTMENT PRESSURES Circumferential Break Steam Line Rupture in Compartment 2 Maximum Pressure in Compartment 2 = 82.1 psia = 67.4 psig maximum Differential Pressure, Subcompartment 2 to Subcompartment 3 = 65.7 psid 6 = 63.0 psid 1A = 66.4 psid 12 = 66.5 psid Reactor Bldg = 67.4 psid Turbine Bldg
= 67.4 psid Steam Line Rupture in Compartment 3 Maximum Pressure in Compartment 3 = 45.1 psia = 30.4 psig Maximum Differential Pressure, Subcompartment 3 to Subcompartment 2 = 27.4 psid 4 = 29.8 psid 1B = 26.6 psid 7 = 28.8 psid 13 = 3.8 psid Reactor Bldg = 30.4 psid Steam Line Rupture in Compartment 4 Maximum Pressure in Compartment 4 = 36.3 psia = 21.6 psig Maximum Differential Pressure, Subcompartment 4 to Subcompartment 3 = 20.8 psid 5 = 20.8 psid 1C = 12.9 psid 14 = 2.4 psid 8 = 18.8 psid Reactor Bldg = 21.6 psid
TMI-1 UFSAR APPENDIX 14 A14A-33 REV. 21, APRIL 2012 TABLE 14A-4 (Sheet 2 of 2)
MAXIMUM COMPARTMENT PRESSURES Circumferential Break Steam Line Rupture in Compartment 5 Maximum Pressure in Compartment 5 = 70.1 psia = 55.4 psig Maximum Differential Pressure, Subcompartment 5 to Subcompartment 11 = 54.3 psid 4 = 54.3 psid 1D = 53.5 psid 15 = 54.3 psid 9 = 52.3 psid Reactor Bldg = 55.4 psid Longitudinal Break Steam Line Rupture in Compartment 3 Maximum Pressure in Compartment 3 = 33.07 psia = 18.37 psig Maximum Differential Pressure, Subcompartment 3 to Subcompartment 2 = 15.99 psid 4 = 17.27 psid 1B = 16.05 psid 7 = 16.62 psid 13 = 1.32 psid Reactor Bldg = 18.37 psid Steam Line Rupture in Compartment 4 Maximum Pressure in Compartment 4 = 21.67 psia = 6.97 psig Maximum Differential Pressure, Subcompartment 4 to Subcompartment 3 = 3.57 psid 5 = 3.52 psid 1C = 1.62 psid 14 = 1.02 psid 8 = 2.87 psid Reactor Bldg = 6.97 psid
TMI-1 UFSAR APPENDIX 14 A14A-34 REV. 21, APRIL 2012 TABLE 14A-5 (Sheet 1 of 2)
COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE Peak Break In Differential Structural Compartment Capacity Pressure Member No.
(psi)
(psi)
Slab, Compartment 2 2
47.8 66.4 Wall, Compartment 2 to 3 2
44.4 65.7 Wall, Compartment 2 to 6 2
39.0 63.0 Roof, Compartment 2 2
54.0 66.5 Slab, Compartment 3 2
32.3 26.6 Wall, Compartment 3 to 4 2
43.7 29.8 Wall, Compartment 3 to 7 3
39.0 28.8 Wall, Compartment 3 to 12 3
44.4 27.4 Roof, Compartment 3 3
36.5 3.8 Slab, Compartment 4 4
9.3 12.9 Wall, Compartment 4 to 3 4
43.7 20.8 Wall, Compartment 4 to 5 4
53.2 20.8
TMI-1 UFSAR APPENDIX 14 A14A-35 REV. 21, APRIL 2012 TABLE 14A-5 (Sheet 2 of 2)
COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE Peak Break In Differential Structural Compartment Capacity Pressure Member No.
(psi)
(psi)
Wall, Compartment 4 to 8 4
40.2 18.8 Roof, Compartment 4 4
33.5 2.4 Slab, Compartment 5 5
19.2 53.5 Wall Compartment 5 to 4 5
53.2 54.3 Wall, Compartment 5 to 9 5
47.0 52.3 Roof, Compartment 5 5
50.1 54.3
TMI-1 UFSAR APPENDIX 14 A14A-36 REV. 21, APRIL 2012 TABLE 14A-6 (Sheet 1 of 1)
CHRONOLOGY OF EVENTS FOR HIGH ENERGY PIPE BREAK TIME (SEC)
EVENT 0
Double-ended break of a 24 inch diameter steam line on the secondary side 1
Reactor trip on variable low pressure; turbine stop valves close, isolating the unaffected steam generator 47 Damaged steam generator blows dry 450 Unaffected steam generator provides no more heat sink; minimum system pressure of about 1550 psia is reached 900 Operator action starts one HPI pump 1200 Primary loop becomes solid with subcooled water; pressurizer code relief valve opens at set point of 2515 psia 2300 Reactor Building cooler actuation set point of 4 psig is reached 5700 Steam first appears in the core 8500 Minimum coolant level in reactor vessel is reached; core remains covered 8800 Containment Building pressure reaches the maximum value of 24 psig
TMI-1 UFSAR APPENDIX 14 A14A-37 REV. 21, APRIL 2012 TABLE 14A-7 (Sheet 1 of 1)
THREE MILE ISLAND NUCLEAR STATION - UNIT-1 INTERMEDIATE BUILDING ENVIRONMENTAL CONDITIONS Main Steam Line Break at 322'-0" Elevation Peak (F)
Peak (psia)
Pressure & Temperature Elevation (ft)
Temperature Pressure Profiles 355 187.0 24.2 322 326.0 24.2 Figure 14A-23 295 322.0 23.6 Figure 14A-21 Main Steam to EFW Pump Turbine Line Break 355 95.0 16.7 322 205.0 16.7 Figure 14A-24 295 273.0 16.7 Figure 14A-22
TMI-1 UFSAR APPENDIX 14 A14A-38 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 1 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 0.0 0.0 0.0 0.0 0.0 0.05 4.973+3 1241.3 6.476+3 1234.0 0.1 5.962+3 1234.9 4.715+3 1234.3 0.1001 5.000+2 1234.9 4.715+33 1234.3 0.2 5.000+2 1234.9 4.411+3 1204.6 0.3 5.000+2 1234.9 4.214+3 1199.7 0.4 5.000+2 1234.9 4.038+3 1200.5 0.5 5.000+2 1234.9 3.876+3 1201.2 0.75 5.000+2 1234.9 3.530+3 1202.5 1.0 5.000+2 1234.9 3.252+3 1203.5 1.5 5.000+2 1234.9 2.923+3 1204.3 2.0 5.000+2 1234.9 2.700+3 1204.9 3.0 5.000+2 1234.9 2.224+3 1205.4 4.0 5.000+2 1234.9 1.948+3 1205.0 5.0 3.518+2 1174.6 2.186+3 1214.1 6.0 3.00 +2 1172.4 2.299+3 1206.3
TMI-1 UFSAR APPENDIX 14 A14A-39 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 2 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 7.0 3.10+2 1172.9 2.465+3 1205.8 8.0 3.29+2 1173.7 2.556+3 1205.9 9.0 3.38+2 1174.0 2.630+3 1205.4 10.0 3.46+2 1174.4 2.689+3 1205.1 11.0 3.50+2 1174.6 2.702+3 1204.9 12.0 3.49+2 1174.5 2.693+3 1204.9 13.0 3.47+2 1174.4 2.669+3 1205.0 14.0 3.43+2 1174.3 2.637+3 1205.0 15.0 3.39+2 1174.1 2.604+3 1205.1 17.0 3.21+2 1203.8 2.534+3 1205.2 20.0 3.07+2 1205.0 2.440+3 1205.3 25.0 2.88+2 1205.0 2.286+3 1205.4 30.0 2.69+2 1205.6 2.150+3 1205.3 35.0 2.50+2 1205.4 2.001+3 1205.2 40.0 2.33+2 1205.2 1.864+3 1204.8
TMI-1 UFSAR APPENDIX 14 A14A-40 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 3 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 45.0 2.18+2 1204.8 1.748+3 1204.5 50.0 2.05+2 1204.4 1.642+3 1204.0 55.0 1.93+2 1203.9 1.543+3 1203.5 60.0 1.81+2 1203.3 1.452+3 1202.9 70.0 1.57+2 1201.9 1.200+3 1201.3 75.0 5.79+1 1197.2 6.128+2 1189.9 75.0-3600 0.0 0.0 8.700+1 1234.4
TMI-1 UFSAR APPENDIX 14 A14A-41 REV. 21, APRIL 2012 TABLE 14A-9 (Sheet 1 of 1)
CONCRETE PROPERTIES Thermal Volumetric Temperature Conductivity Heat Capacity (F)
K, (Btu/hr - Ft20F)
(Btu/ft30F) 20 0.80 30.0 200 0.81 30.0 300 0.82 30.0
TMI-1 UFSAR APPENDIX 14A 14A-i REV. 18, APRIL 2006 APPENDIX 14A - DESIGN REVIEW FOR CONSIDERATION OF EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENT TABLE OF CONTENTS SECTION TITLE
1.0 INTRODUCTION
2.0 REVIEW CRITERIA 3.0 GENERAL DISCUSSION 3.1 DESIGN 3.2 QUALITY ASSURANCE 3.3 SHUTDOWN CAPABILITY 4.0 DISCUSSION OF BREAKS 4.1 CONTROL ROOM AND CONTROL BUILDING 4.2 INTERMEDIATE AND TURBINE BUILDING 4.3 AUXILIARY, FUEL, AND INTERMEDIATE AREA (PERSONNEL ACCESS)
BUILDING (AUXILIARY AREA) 4.4 DIESEL GENERATOR BUILDING 4.5 COLD SHUTDOWN CAPABILITY 5.0
SUMMARY
AND CONCLUSIONS 6.0 INSPECTION REQUIREMENTS 7.0 METHODS OF ANALYSIS AND DESIGN 7.1 THERMAL-HYDRAULIC ANALYSIS 7.1.1 BLOWDOWN ANALYSIS 7.1.1.1 MAIN STEAM BLOWDOWN ANALYSIS 7.1.1.2 FEEDWATER BLOWDOWN ANALYSIS 7.1.2 COMPARTMENT PRESSURIZATION ANALYSIS 7.1.2.1 MAIN STEAM LINE RUPTURES 7.1.2.2 FEEDWATER LINE RUPTURES 7.1.3 PIPE THRUST AND JET IMPINGEMENT 7.2 STRUCTURAL ANALYSIS AND DESIGN 7.2.1 WALL AND SLAB CAPACITIES
TMI-1 UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE APPENDIX 14A 14A-ii REV. 18, APRIL 2006 7.2.1.1 GENERAL YIELD-LINE THEORY AND SHEAR CAPACITY 7.2.1.2 COMPARTMENT DIFFERENTIAL PRESSURE VERSUS CAPACITIES 7.2.1.3 JET IMPINGEMENT AND COMBINED DIFFERENTIAL PRESSURE 7.2.2 RESTRAINT DESIGN PROCEDURE 7.2.2.1 ASSUMPTIONS 7.2.2.2 ANALYTICAL METHOD 7.2.2.3 MATERIAL PROPERTIES 7.2.2.4 SECTION PROPERTIES 7.2.2.5 EXAMPLE MODEL 7.2.2.6 DETAIL DESIGN
8.0 REFERENCES
TMI-1 UFSAR APPENDIX 14A 14A-iii REV. 18, APRIL 2006 LIST OF TABLES TABLE TITLE 14A-1 HIGH-ENERGY LINES INVESTIGATED FOR CONSEQUENCES OF POSTULATED PIPE RUPTURES 14A-2 SYSTEM REQUIRED FOR SAFE SHUTDOWN FOLLOWINGA POSTULATED MAIN STREAM OR FEEDWATER PIPE RUPTURE 14A-3 TABULATION OF SYSTEMS AND BREAK LOCATIONS 14A-4 MAXIMUM COMPARTMENT PRESSURES 14A-5 COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE 14A-6 CHRONOLOGY OF EVENTS FOR HIGH ENERGY PIPE BREAK 14A-7 THREE MILE ISLAND NUCLEAR STATION - UNIT 1 INTERMEDIATE BUILDING ENVIRONMENT CONDITIONS 14A-8 MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK 14A-9 CONCRETE PROPERTIES
TMI-1 UFSAR APPENDIX 14A 14A-iv REV. 18, APRIL 2006 LIST OF FIGURES FIGURE TITLE 14A-1 MAIN STEAM, FEEDWATER & EMERGENCY FEEDWAT LINES IN THE INTERMEDIATE BUILDING 14A-2 MAIN FEEDWATER & AUXILIARY STEAM LINES IN THE 14A.FIG-2 TURBINE BUILDING NEAR THE REACTOR BUILDING 14A-3 DECAY HEAT, MAKEUP & PURIFICATION, AUXILIARY STEAM LINES, AND HIGH ENERGY SAMPLE LINES IN THE FUEL HANDLING AND AUXILIARY BUILDINGS BELOW EVALUATION 305 FT 14A-4 MAKEUP & PURIFICATION AND AUXILIARY STEAM LINES ABOVE ELEVATION 305 FT IN THE AUXILIARY AND FUEL HANDLING BUILDINGS 14A-5 PHYSICAL ARRANGEMENT OF BUILDINGS 14A-6 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 112 TO STOP VALVE SV-4 14A-7 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 113 TO STOP VALVE SV-3 14A-8 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 114 TO STOP VALVE SV-2 14A-9 MAIN STEAM FROM REACTOR BUILDING PENETRATION NO 419 TO STOP VALVE SV-1 14A-10 MAIN STEAM FROM STEAM GENERATOR-A TO EMERGENCY FEEDWATER PUMP TURBINE 14A-11 MAIN STEAM FROM STEAM GENERATOR-B TOHEADER AT ELEVATION 297 FT 0 IN.
14A-12 FEEDWATER FROM 30 IN HEADER TO PENETRATION NO 103 14A-13 FEEDWATER FROM 30 INCH HEADER TO PENETRATION NO 227 14A-14 MAINSTEAM BLOWDOWN THRUST BREAK AT CONTAINMENT 56 (AT ISOLATION VALVE) 14A-15 MAIN FEEDWATER BLOWDOWN THRUST BREAK AT CONTAINMENT 3T24 14A-16 INTERMEDIATE BUILDING SUBCOMPARTMENT DESIGNATIONS
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE APPENDIX 14A 14A-v REV. 18, APRIL 2006 14A-17 RESTRAINT FOR BREAKS IN 12 INCH HEADER 14A-18 RESTRAINT FOR BREAKS IN MAIN FEEDWATER 14A-19 MOMENT VS CURVATURE AND SHEAR VS SHEARING STRAIN 14A-20 DYNAMIC PIPE BREAK MODEL 14A-21 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A MAIN STEAM LINE BREAK (HEAT SLAB INCLUDING, EL 295')
14A-22 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A EFW PUMP TURBINE STEAM LINE BREAK HEAT SLAB INCLUDED, EL 295')
14A-23 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A MAIN STEAM LINE BREAK (HEAT SLAB INCLUDED, EL 322')
14A-24 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A EFW PUMP TURBINE STEAM LINE BREAK (HEAT SLAB INCLUDED, EL 322')
14A-25 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING STEAM LINE BREAK (HEAT SLAB INCLUDED, LONG TERM COOLDOWN, EL 295')
14A-26 TIME HISTORY FOR THE INTERMEDIATE BUILDING FOLLOWING A STEAM LINE BREAK (HEAT SLAB INCLUDED, LONG TERM COOLDOWN, EL 322')
TMI-1 UFSAR APPENDIX 14A 14A-1 REV. 21, APRIL 2012 APPENDIX 14A DESIGN REVIEW FOR CONSIDERATION OF EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENT
1.0 INTRODUCTION
A design review was performed in response to the United States Atomic Energy Commission (AEC) request, "General Information Required for Consideration of Effects of Piping System Break Outside Containment," of December 15, 1972, for Three Mile Island Nuclear Power Station, Unit-1. Also in response to the AEC letter of June 1, 1973, an additional design review was performed. The review covers all piping systems in the unit in accordance with the criteria presented in the AEC request. All of the possible effects of postulated pipe failures as outlined in the AEC request have been considered in the review, and it is concluded that the unit can be safely shut down. This report summarizes the results of the design review and outlines the methods of achieving safe shutdown.
Design modifications will be implemented to provide assurance that the Engineered Safeguard Systems and the Emergency Feedwater System will be operable after postulated high energy pipe breaks. Inservice inspection of certain specified postulated break locations will provide assurance that rupture will not occur at those respective points. Operation of the Emergency Feedwater System will permit a more rapid shutdown capability than is possible with Engineered Safeguard Systems alone. Thus, the primary means of effecting a cooldown after a postulated break outside containment would be the Emergency Feedwater System with high pressure-low pressure injection cooldown serving as a backup.
TMI-1 UFSAR APPENDIX 14A 14A-2 REV. 21, APRIL 2012 2.0 REVIEW CRITERIA The sets of criteria used for the review are as follows: (1) considering the type of affects that can result from a pipe break and (2) criteria addressing the degree of system operability required following a break. The criteria used are as follows:
Pipe Break Criteria Pipes Evaluated Effects Considered____________
Fluid above 200 F and 275 psig Longitudinal and circumferential at terminal ends and high stress breaks including pipe whip, jet locations on lines 4 inches or impingement, flooding, and greater environmental conditions Fluid above 200 F and 275 psig Circumferential breaks including at terminal ends and high stress pipe whip, jet impingement, locations on lines 1 to 4 inches flooding, and environmental conditions Pipe Break Criteria Pipes Evaluated Effects Considered___________
Fluid above 200 F and/or 275 psig Crack breaks including jet at the most adverse locations impingement, flooding, and for all pipes environmental conditions Fluid below 200 F and 275 psig None System Operability Criteria System Operability Required_________
Systems required to bring the unit No loss of required redundancy to safe shutdown following the break permitted Reactor Protection System and the Loss of redundancy but no loss Engineered Safeguards System of function permitted Table 14A-1, "High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures,"
indicates the various high energy lines investigated under the above criteria and presents appropriate data or information concerning them. Figures 14A-1, 14A-2, 14A-3, and 14A-4 supplement the location and elevation data given in Table 14A-1 for the various high energy lines investigated.
TMI-1 UFSAR APPENDIX 14A 14A-3 REV. 21, APRIL 2012 3.0 GENERAL DISCUSSION 3.1 DESIGN A rupture of the high energy piping is considered highly unlikely due to the low seismic and operating stress levels. All these systems have been conservatively designed and all the systems except auxiliary steam to the Auxiliary Building have been analyzed in accordance with USAS B31.1.0, Code For Power Piping. The auxiliary steam system to the Auxiliary Building has been designed and analyzed as described in Section 5.4.4.2. This includes all portions of the auxiliary steam system located in the Control, Fuel Handling and Auxiliary Buildings.
Results of these analyses show that the maximum stress levels from combined operating and seismic conditions are well below those limits designated as potential pipe rupture stress levels.
Piping systems are designed to USAS B31.1.0. In addition, portions of the auxiliary steam system piping are analyzed in accordance with the CDFM methodology as described in Section 5.4.4.2. Quality assurance was applied to USAS B31.1.0 requirements for the non nuclear piping and to USAS B31.7 requirements for nuclear piping. (Nuclear piping is defined as piping that normally contains radioactivity.) The analysis of the auxiliary steam piping is based on the configuration and conditions of the piping system at the time of the walkdown and evaluation.
To ensure that the existing condition is maintained, the quality classification of the auxiliary steam piping system has been upgraded to Regulatory Required with QA.
On non-nuclear piping, welders qualified to ASME Section IX requirements were used. Piping system leakage testing was performed in accordance with piping code (USAS B31.1) requirements. All welds were visually inspected.
a.
The main steam piping welds 4 inches and over were 100 percent radiographed from steam generators to the turbine generator.
b.
The main feedwater piping welds 4 inches and over were 100 percent radiographed from pumps to steam generator.
c.
The emergency feedwater piping welds were 100 percent radiographed from the steam generators up to the first isolation valve (which is in the Intermediate Building).
d.
The steam supply (to the emergency feedwater pump turbine) piping welds were 100 percent radiographed.
All the NDT required by USAS B31.7 was applied to nuclear piping systems, i.e., decay heat, makeup and purification, sampling, and so forth.
3.2 QUALITY ASSURANCE The design and construction phase Quality Assurance Program was a three-level program.
The first level of the program was performed by the equipment manufacturer or site contractor, the second level by Met-Ed's main contractor (i.e., B&W--NPGD, GAI, or UE&C, as appropriate), and the third level by Met-Ed itself and/or its agent, MPR Associates.
TMI-1 UFSAR APPENDIX 14A 14A-4 REV. 21, APRIL 2012 3.3 SHUTDOWN CAPABILITY The unit can be brought to a cold shutdown condition by utilization of either the emergency feedwater system and atmospheric dump valves, or HPI cooling and the Reactor Building emergency coolers.
The emergency feedwater system and the atmospheric dump valves are located in the Intermediate Building. The emergency core cooling systems are located in the Auxiliary Building area, separated from the Intermediate and Turbine Buildings by the Reactor and Fuel Handling Buildings.
The highest energy lines are located in the Intermediate and Turbine Buildings. For major breaks in these lines, unit shutdown will be accomplished through utilization of the Emergency Feedwater System and atmospheric dump valves or HPI Cooling and the Reactor Building Cooling Systems. For crack breaks, a normal unit shutdown will be achieved.
TMI-1 UFSAR APPENDIX 14A 14A-5 REV. 22, APRIL 2014 4.0 DISCUSSION OF BREAKS The layout of equipment within the buildings and the physical arrangement of the buildings themselves provide protection for the shutdown equipment from high energy line breaks. See Figure 14A-5 and Table 14A-2 for relative building and equipment locations.
Locations of postulated breaks have been determined for each of the high energy piping systems that might endanger the Emergency Feedwater System if rupture should occur. The selection of breaks is based on the results of stress analyses previously performed on the as-built piping systems. This review considered effects of pressure, deadweight, and thermal expansion during normal operating, upset, test conditions, and the operating basis earthquake (OBE). The stress levels obtained by this review for main steam and feedwater were found to be lower than those of the AEC pipe rupture criteria. To provide a conservative criterion for selecting break locations, the two intermediate points of highest stress are postulated as break locations.
Design basis breaks in straight or curved pipes 4 inches in diameter or greater are assumed to be either longitudinal or circumferential with the break area equal to the flow area of the pipe.
Design basis breaks at branch and longitudinal in the run with the break area equal to the flow area of the branch. The criteria used to select design basis break locations are as follows:
a.
Postulated breaks at all terminal points (anchors or rigid attachment to equipment or anchor extensions).
b.
Postulated breaks at all branch points.
c.
Postulated intermediate breaks between terminal points whenever the primary stress (pressure, weight, OBE) plus secondary stress (thermal) exceeds 80 percent of Sh +
SA), or where secondary stress alone exceeds 80 percent of SA.
d.
As a minimum, two intermediate breaks between terminal points were selected at locations of highest stress.
The above pipe break location criteria were applied to the high energy piping systems, and the results are tabulated in Table 14A-3. This tabulation shows the pipe isometric drawing numbers associated with each of these systems, the potential break point identification number, and the primary and secondary stresses proportional to the 80 percent allowable limit. The break locations are shown on the isometrics on Figures 14A-6 through 14A-13 and in plan on Figures 14A-1 through 14A-4.
In addition, crack breaks were postulated at adverse locations and assumed to be one half the pipe diameter in length and one half the pipe wall thickness in width.
The specific thrust versus time curves used in designing the restraints defined in this supplement are shown on Figures 14A-14 and 14A-15.
Item 15 of the Atomic Energy Commission document titled "General Information Required for Consideration of the Effects of a Piping System Break Outside Containment," Reference 12, required that a discussion should be provided for the potential for flooding of safety related equipment in the event of failure of a feedwater line or any other line carrying high energy fluid.
TMI-1 UFSAR APPENDIX 14A 14A-6 REV. 22, APRIL 2014 4.1 CONTROL ROOM AND CONTROL BUILDING The Control Building equipment, electrical power and control, chilled water system, and ductwork systems are contained within the structure of the Control Building. In this isolated location they would not experience adverse effects from any high energy pipe break. Access to the Control Building structure is either through the Turbine Building or the Fuel Handling Building. Outside air to the Control Building is ducted to the Control Room from a remote underground intake terminal and would not be adversely affected by a high-energy pipe break.
The Turbine Building would experience momentary overpressure if the break occurred in this area, but this would be dissipated through numerous wall and roof openings. Steam leakage from the turbine hall to the Control Room or Control Building during this period is minimized as it is forced to travel through the west Turbine Building wall, through multiple doors in series, before entering the Control Building areas. Also, the doors have automatic closers. Any steam leakage into the corridor space outside the Control Room or the Control Room space will be condensed and dissipated by the ventilation systems, and no significant ambient changes would be anticipated in these areas.
Investigation indicates that there are no high-energy lines larger than 1 inch other than the auxiliary steam pipe in or near the Control Building, and thus, postulated pipe whip and steam jet impingement are not able to damage the Control Room. Due to the low operating pressure of the auxiliary steam system, rupture of this line is not a consideration.
High-energy sample lines (under 1 inch) are discussed in Section 4.3.
4.2 INTERMEDIATE AND TURBINE BUILDINGS a.
The Intermediate and Turbine Buildings contain all of the lines over 1 inch with internal fluid exceeding both 200oF and 275 psig.
The pressure and temperature response in the Intermediate Building following steam line breaks are based on plant specific mass and energy release rates as shown in Table 14A-8. The thermal-hydraulic results are summarized on Table 14A-7 and supporting Figures 14A-21 through 14A-26 which indicate the temperature and pressure time history profiles for each level or compartment. As shown, the Main Steam Line Break is the limiting and enveloping break in the Intermediate Building from a pressure/temperature viewpoint. This is due to the much larger mass of steam being released compared to the EFW pump turbine steam supply line break.
The Reactor Building will maintain its containment integrity when subjected to the resultant external pressurization of a main steam or feedwater break within the Intermediate Building. The Reactor Building will not be subjected to main steam pipe whip because the 3 ft thick interior walls of the Intermediate Building effectively restrain the pipe.
b.
Electrical equipment is required to function subsequent to a High Energy Break (HELB) inside or outside of containment and must meet environmental qualification requirements. The qualification conditions considered include post accident pressure and temperature conditions in Table 14A-7. Electrical equipment which is required to
TMI-1 UFSAR APPENDIX 14A 14A-7 REV. 22, APRIL 2014 function following a postulated HELB is located at the 295 ft elevation of the Intermediate Building.
c.
Isolation valves in the Intermediate Building are on lines that might be open to the containment atmosphere during normal operation (purge valve) and have been reviewed. It was found that for any postulated high energy line break, the valve will not be damaged and will close upon receiving an electrical (deenergize) signal or loss of control air.
d.
The systems that will be used to bring the plant to a safe shutdown after the postulated major break in the Intermediate Building are listed in Table 14A-2. A detailed accident review was made to resolve the effect of each postulated break defined by Section 4.0 on the operability of the Emergency Feedwater Train. The objective of this review was to establish those breaks that would ultimately prevent the operation of both steam generators or both Emergency Feedwater Trains and to determine the design modifications necessary to assure emergency feedwater operation.
1)
Breaks at the containment penetrations in the small compartments 2 and 5 (refer to Figure 14A-16) could produce pressures in excess of wall and/or slab capacities. Portions of the Emergency Feedwater System below elevation 322 ft 0 inch could be damaged by the resulting debris. To provide reasonable assurance that the postulated ruptures in those compartments will not occur, the associated welds are to be inspected in accordance with Technical Specification 4.15.1.
The breaks in compartments 3 and 4 are similar but do not produce differential pressures that would produce incremental collapse of the Intermediate Building interior structures. The method of calculating slab and wall capacity (Yield Line Theory) is reviewed in Section 7.2.1.1 of Appendix 14A and a comparison of capacities with expected differential pressures is summarized in Section 7.2.1.2 of Appendix 14A.
2)
The 12 inch main steam header below elevation 322 ft 0 inch and shown on Figure 14A-1 is positioned directly opposite emergency feedwater valve EF-V1B.
The restraint/shield schematically represented on Figure 14A-17 is provided to protect valve EF-V1B from damage due to postulated breaks in the header and in the 12 inch main steam line that connects to the end of the header nearest EF-V1B.
3)
The feedwater line from containment penetration No. 103 runs approximately parallel to and above a 12 inch main steam line, which is also in the overhead of the same compartment as the turbine driven EFW Pump. To prevent both Steam Generators from becoming inoperable and the loss of the turbine driven EFW pump, the postulated breaks in this section of the pipe have been reevaluated with respect to References 15 and 16. The results of this evaluation indicate that this section of pipe can be classified as "superpipe" and that no breaks have to be postulated in this area (i.e., Break Locations 1 and 7 have been changed to 13 and 28 on Figure 14A-12). The pipe whip restraints
TMI-1 UFSAR APPENDIX 14A 14A-8 REV. 22, APRIL 2014 installed on the feedwater line to reduce pipe deflections due to a pipe break at locations 1 and 7 shall remain to reduce pipe deflections from pipe breaks at locations 13 and 28. Figure 14A-18 illustrates the protection provided.
The design procedure applied to the design of this restraint and the one discussed in Section 4.2.d.2. of Appendix 14A conform to the procedures presented in Section 7.2.2 of Appendix 14A.
4)
Section 4.5 of this appendix provides an analysis of cooldown capability without using the Emergency Feedwater System.
5)
An evaluation was made in Reference 13 for different alternatives available in order to mitigate flooding in the Intermediate Building in case of a postulated feedwater line break. The flood protection modifications were implemented to mitigate the effects of flooding due to a postulated main feedwater line break (MFLB) in the Intermediate Building by allowing water to flow into the tendon access gallery and the alligator pit. By removing the upper half of the western water "stop wall" on the alligator pit and opening the doors at entrance "A" and "B" to the tendon access gallery, there will be approximately 25 minutes before flooding in the Intermediate Building adversely affects the emergency feedwater system components not qualified for submergence would be adversely affected (Reference 14). Intermediate Building flood detection alarm system has been added to alert the operator in the Control Room to flooding conditions as a result of a MFLB. The alarm will provide the operator with sufficient time, approximately 20 minutes, to take corrective action to prevent damage to the EFW pumps.
e.
Postulated breaks in the Turbine Building cannot adversely affect equipment utilized for the safe shutdown of the reactor. Since there is no affected reactor safety equipment in this area, the review of breaks in the Turbine Building is complete.
4.3 AUXILIARY, FUEL, AND INTERMEDIATE AREA (PERSONNEL ACCESS)BUILDINGS (AUXILIARY AREA)
Those reactor protection systems and engineered safeguards systems that could be affected by postulated pipe breaks are all located in the Auxiliary Building.
The effects of breaks in the Intermediate or Turbine Building on the Auxiliary Building ambient atmosphere will be minimal and momentary as this steam leakage to the Auxiliary Building area must pass through the west turbine building wall, through the controlled access Hot Tool Room door, and travel approximately 150 feet before entering the Auxiliary or Fuel Handling Buildings.
This leakage would be continuously condensed and dissipated by the outside air ventilation exhaust systems in the corridor and in the Auxiliary and Fuel Handling Buildings.
The postulated breaks in the auxiliary area will not require protective action because they do not deplete primary system inventory or impair the normal heat removal systems (main steam and feedwater).
Another characteristic of pipe breaks in this area is that they are substantially lower in energy than breaks in the Intermediate or Turbine Building. No lines in this area larger than 1 inch
TMI-1 UFSAR APPENDIX 14A 14A-9 REV. 22, APRIL 2014 carry fluids that exceed both 200ºF and 275 psig. The largest line in this area with postulated crack breaks is 14 inches versus 24 inches in the Intermediate or Turbine Building.
All break locations in the auxiliary area will meet the established criteria (one string of each engineered safeguards system will remain operable). The break will not damage both strings of any engineered safeguards system due to the separation of engineered safeguard components. In a few cases where cabling for both trains of an engineered safeguards system runs in the vicinity of a postulated break, the effects on the cabling will be analyzed and, if necessary, protected by a barrier or rerouted as required.
The capability to detect and sustain the flooding as a result of breaks in the auxiliary area is discussed in Section 6.4.5 of the FSAR.
Postulated high energy sample line breaks will not affect any engineered safeguard/reactor protection system equipment nor render any vital building areas permanently uninhabitable.
The HVAC has been designed to confine the consequences of reactor coolant small line breaks to relatively small areas of the Fuel Handling Building at elevation 281 ft 0 inch and the radioactive sample station in the Control Room tower which are of no immediate importance in the event of a high energy line rupture. The radiation monitoring system and Control Room controls of the Reactor Building isolation valves will be used to terminate any postulated high energy sample line breaks.
The letdown line to the makeup system normally operates below 200F and is therefore subject only to crack breaks according to the criteria prescribed in Section 2.0 of this Appendix.
However, a full diameter break would result in flow sufficiently high to render the coolers ineffective and the temperature of the flow from the break would exceed 200F. To assure sufficient conservatism, full size breaks have been postulated in the letdown line. For a break downstream of the breakdown orifice, the flow would be slightly above normal and the effectiveness of the cooler inside containment would not be significantly reduced.
Analyses performed by GPUN (Reference 17) determine postulated line break locations and address the effects of those breaks on the letdown line and nearby equipment due to jet impingement. The result of the evaluation show that the letdown line is structurally adequate to ensure that pipe whip is not an issue during a postulated line break event and that the pipe will remain stationary. Furthermore, jet impingement from the postulated rupture points does not pose a threat to any other safety related system in the plant. The compartment pressurization resulting from the break has been analyzed and found to be below that which would cause damage to the building.
The temperature, pressure, and humidity environment in the vicinity of the break has been determined by analysis (References 18, 19). The safeguard equipment in the vicinity of the break consists of valves, cabling, and pumps. The valves and cabling have been qualified for the environment expected to result from the postulated break. The pumps are in separate compartments and are not affected by the local environment. The potential flooding from a postulated letdown line break does not result in a water level high enough to impact safety related equipment or circuits.
A postulated break will be automatically isolated by closure of the Reactor Building isolation valves when high temperature is sensed in the letdown line.
TMI-1 UFSAR APPENDIX 14A 14A-10 REV. 22, APRIL 2014 Since the Makeup System in the auxiliary building is not a High Energy Line per the criteria of Section 2.0, a High Energy Line Break (HELB) need not be postulated. Therefore, this is not a Design Basis Event for TMI-1 and 10 CFR 50.49 does not apply. The environmental evaluation was performed to respond to initial licensing questions but is not in the scope of 10 CFR 50.49 (Reference 20, 21).
4.4 DIESEL GENERATOR BUILDING The Diesel Generator Building has no direct doorway or access connection to either the Intermediate Building or the Turbine Building. However, ventilation air comes from the machine shop. Steam leakage into the machine shop would be minimal as closed doorways separate the machine shop and Turbine Building. Also, the machine shop has a separate air supply and exhaust system that would dissipate any steam leakage.
Prolonged steam conditions in this space are not to be expected. The north diesel room has no direct doorway, access, or ventilation connection to either the Intermediate Building or Turbine Building or any other area expected to become steam laden following the pipe break.
Therefore, the north diesel room ambient atmosphere will not experience any change as a result of the pipe break.
4.5 COLD SHUTDOWN CAPABILITY The adequacy of the borated water storage tank as an interim heat sink for the Three Mile Island Nuclear Station Unit-1 Reactor Coolant System has been evaluated for the following set of assumptions:
a.
Steam line break occurs inside the Intermediate or Turbine Building during rated power operation b.
Loss of all feedwater to both steam generators occurs d.
Loss of offsite power occurs In addition to this set of assumptions, this evaluation is valid for any situation where Reactor Coolant System energy removal through the steam generators is no longer available because of a HELB in containment.
There are three primary areas of concern for this condition. These areas are prevention against core uncovering, protection against excessive Reactor Building pressure, and the ability to achieve cold shutdown conditions.
The B&W digital computer code CRAFT, Reference 10, has used to determine the characteristics of this accident with regard to core uncovering and mass energy releases to the containment. The mass and energy release data from CRAFT were used in the digital computer code CONTEMPT, Reference 11, for Reactor Building pressure calculations. The assumptions and results of the analysis are summarized in Table 14A-6. A single steam generator blowdown was considered as the most conservative case since, for a double
TMI-1 UFSAR APPENDIX 14A 14A-11 REV. 22, APRIL 2014 blowdown, the high pressure injection (HPI) pump would be started almost instantaneously on low Reactor Coolant System pressure actuation, meaning a lower probability of core uncovering.
Core uncovering is prevented by pumping water from the borated water storage tank via the Makeup and Purification System (HPI) into the Reactor Coolant System. With one makeup and purification (HPI) pump started 15 minutes after the break, the minimum coolant level in the reactor vessel occurs at approximately 140 minutes and at no time falls below the top of the core. Operator action is assumed to occur 15 minutes after the break in starting the makeup and purification pump (HPI).
The building pressure increases during the transient as boiloff occurs through the pressurizer safety valves (2515 psia). Assuming the boiloff goes directly to the building atmosphere with no credit for the quench tank, the building pressure reaches the Reactor Building cooler and high pressure injection set point (4 psig) 38 minutes after the break. With one building cooler operative at this time, the building pressure reaches a maximum value of 24 psig and never exceeds the design pressure limit. Furthermore, the reactor spray actuation set point (30 psig) is not reached, and a single building cooler provides adequate protection throughout the transient against excessive Reactor Building pressure.
High pressure injection of BWST water continues until the BWST is depleted (approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, assuming one HPI pump is operating). At this time, further cooldown is achieved by using the decay heat (low pressure injection) pumps drawing from the Reactor Building sump to supply suction to the makeup and purification (HPI) pumps. The sump recirculation continues until the Decay Heat Removal System (LPI) can be actuated to reduce the system to cold shutdown. Cold shutdown is then achieved by venting the system pressure and actuating the Decay Heat Removal System to recirculate the reactor coolant through the decay heat coolers.
TMI-1 UFSAR APPENDIX 14A 14A-12 REV. 22, APRIL 2014 5.0
SUMMARY
AND CONCLUSIONS The results of this design review are summarized as follows:
a.
A rupture of the high energy piping systems is considered highly unlikely. The systems other than auxiliary steam piping have been conservatively designed in accordance with the criteria in USAS B31.1.0, Code for Power Piping. Materials, fabrication, and quality assurance requirements of the Code have been utilized. In addition, the main steam piping has been subject to 100 percent radiography of welds from the steam generators to the turbine stop valves, and the feedwater piping has been subject to 100 percent radiography from the steam generators to the feedwater pumps. Quality assurance provisions of USAS B31.7, Code for Nuclear Piping, have been implemented for nuclear systems. The auxiliary steam piping system has been shown to be adequate for combined effects of operating conditions and the SSE. Quality assurance provisions are implemented to ensure configuration of the auxiliary steam piping system is maintained.
b.
The Atomic Energy Commission criteria had been implemented in identifying postulated break locations in high energy piping systems.
c.
All of the equipment required for shutdown is protected from the postulated ruptures by virtue of location (remote from high energy lines), restraints, inspection, or barriers.
d.
The Control Room will remain habitable and operable following a postulated high energy pipe break due to its remote location from such breaks.
e.
The plant can be brought to cold shutdown conditions by utilizing either feedwater system or the Emergency Core Cooling System and Reactor Building cooling.
f.
This analysis shows that the borated water storage tank serves as an adequate interim heat sink from the standpoint of core covering, Reactor Building pressure considerations and achieving cold shutdown.
g.
Further, the analysis shows that the core remains covered, the Reactor Building is protected against excessive pressure, the plant can be taken to cold shutdown, and containment integrity can be maintained.
After Cycle 17, the original OTSGs were replaced with enhanced OTSGs. The existing HELB-related analyses and evaluations contained in Appendix 14A were reviewed in Reference 124 and confirmed to remain applicable following the TMI SG replacement.
TMI-1 UFSAR APPENDIX 14A 14A-14 REV. 20, APRIL 2010 6.0 INSPECTION REQUIREMENTS It is intended to ultrasonically inspect the welds identified on Figures 6 and 9 in compartments 2 and 5 to the extent possible. This inspection will be performed periodically as required by Technical Specification 4.15.
TMI-1 UFSAR APPENDIX 14A 14A-15 REV. 21, APRIL 2012 7.0 METHODS OF ANALYSIS AND DESIGN 7.1 THERMAL-HYDRAULIC ANALYSIS A detailed thermal-hydraulic analysis was performed in order to evaluate the consequences of a postulated high energy line rupture in the Intermediate Building. The analysis included the calculation of system transients and blowdown loads as well as compartment pressurization for use in the structural evaluation.
7.1.1 BLOWDOWN ANALYSIS 7.1.1.1 Main Steam Blowdown Analysis Selection of Line Breaks A total of Two (2) double ended guillotine pipe breaks in the Main Steam System assumed at the worst locations from an equipment qualification viewpoint; i.e., breaks are located in areas where the environmental response was most severe. Piping stress levels were not considered as an input to break location.
The selected break located at the 322 ft elevation is a guillotine break of a 24 inch main steam line upstream of the main steam line isolation valve. Guillotine breaks give the maximum total flow and most severe conditions in the building.
The location of the worst case break for the steam supply line to the EFW pump turbine is at the 295 ft elevation and is a guillotine break of the steam supply line.
Computer Model An EDSFLOW Computer model was constructed for the Intermediate Building and was utilized to evaluate the main steam line breaks. The building was nodalized as volumes. Portals such as doorways and stairwells were modeled as connecting pathways between the volumes.
The Intermediate Building was nodalized into sixteen volumes. The 295 ft and 305 ft elevations were more precisely modeled since most of the safety related electrical equipment is located in these spaces, and a more accurate prediction of the pressure and temperature response is required. The area of the walls and ceiling of each compartment was calculated and modeled as an exposed concrete surface to simulate their heat sink effect.
The heat transfer coefficient (h) to the walls was calculated by the EDSFLOW Computer Code to be 5 Btu/F-hr-ft2 throughout the transient for the 24 inch main steam line break case. The heat transfer coefficients calculated internal to the code as a function of transient time and is based upon fundamental natural and forced convection heat transfer correlations. This value is conservative since the effect of wall condensation was not considered. In accordance with the Uchida Condensing Heat Transfer Correlations, heat transfer coefficients substantially higher (up to a maximum of 280 Btu/F-hr-ft2 at an air to steam ratio of 0.1) than the value calculated could have been used shortly after break initiation. For the steam supply to the emergency feed pump turbine line break case a heat transfer coefficient of 50 Btu/F-hr-ft2 was utilized during blowdown, and 5 Btu/F-hr-ft2 after blowdown was terminated. These values are well
TMI-1 UFSAR APPENDIX 14A 14A-16 REV. 21, APRIL 2012 below the actual condensing coefficients. The results of the analysis, therefore, are considered conservative in nature.
The thermal properties of concrete used in this analysis are listed in Table 14A-9.
The blowdown mass and energy release rates for the 24 inch line break were based on a circumferential break. The mass and energy release rates are included in Table 14A-8.
The blowdown mass and energy release rates for the 12 inch line break were calculated utilizing a modified Darcy's equation. The mass and energy release rates are 512 1bm/sec-ft2 and 6.154 x 105 Btu/1bm-ft2, respectively. It is conservatively assumed that this blowdown is constant for ten minutes without depressurizing the steam generators. After ten minutes, it is assumed emergency feedwater to the faulted OTSG would be terminated by the operator.
Key Analytical Assumptions The following assumptions and approximations were used in this analysis:
1.
Initial Intermediate Building pressure, temperature and humidity were assumed to be 14.7 psia, 90F, and 60 percent, respectively.
2.
All steam lines were assumed to be at their maximum operating pressures and temperatures.
3.
Forward and reverse flow shock losses were accounted for where junction flow areas were reduced or expanded in relation to the adjacent volume flow areas.
4.
For the EFW pump turbine steam supply line break, doors are assumed to be in their most conservative position at the initiation of the event. If the door is closed, it is assumed to open at a differential pressure of 2 psid when opening away from the door jamb, and 4 psid when opening against the door jamb.
5.
The analysis was performed without the 92 ft2 vent area at the 322 ft elevation available.
There is 75 ft2 total vent area from the building when the opening at the 322 ft elevation is not assumed available.
6.
Operator action to terminate EFW flow to the faulted OTSG feedwater is required to mitigate the EFW pump turbine steam supply line break within ten minutes.
7.
The steam vent pathway out of the Intermediate Building is assumed available after a 2 psid pressure buildup in the building.
7.1.1.2 Feedwater Blowdown Analysis The RELAP-3 digital computer code, Reference 2, was used in analyzing the feedwater blowdown transients for the determination of thrust loads. The system was represented by an assemblage of control volumes connected by flow paths or junctions. The effects of valves, pumps, heat exchangers, and check valves are included in the code.
TMI-1 UFSAR APPENDIX 14A 14A-17 REV. 21, APRIL 2012 The steam generators were modeled so that the feedwater inlet nozzles were above the steam generator water level. In this representation, backflow through the inlet nozzles would be steam.
The feedwater lines were divided into several volumes for each case. The volumes were selected so that volume size and junction location would provide optimum system representation for the particular case being analyzed.
It was assumed that for the durations of these analyses, the feedwater pumps would continue to operate and that flow would be a function of head. It was further assumed that for the duration of these analyses, an unlimited constant pressure supply of water was available at the feedwater pump suction. Both main feedwater pumps were combined and modeled as a single pump.
In modeling flow nozzles, the actual nozzle throat area was used if flow to the leak was in a forward direction through the flow nozzles. Where flow to the leak was in a reverse direction through the nozzle, an effective flow area was calculated and used in the model.
Basic Assumptions a.
Reactor operation at full load conditions.
b.
Steam generator nominally at 925 psi with feedwater inlet at 1000 psi and 462oF.
c.
Feedwater line check valve fails to close.
d.
Pumps do not trip.
e.
Circumferential and longitudinal breaks were considered.
f.
Break volumes were selected to account for the segment of piping up to the first elbow on either side of the break.
7.1.2 COMPARTMENT PRESSURIZATION ANALYSIS 7.1.2.1 Main Steam Line Ruptures The pressure temperature transients resulting from the postulated rupture of a main steam line in the Intermediate and Turbine Buildings were investigated.
The transients for subcompartment integrity evaluation were calculated by extending the short term main steam blowdown model to include control volumes representing compartments of the Intermediate Building with their interconnecting vent area. Figure 14A-16 presents the Intermediate Building compartment designations. Double-ended circumferential ruptures were considered as the limiting case. Table 14A-4 presents the maximum wall and slab pressure differentials obtained for the Intermediate Building subcompartments where over-pressurization was investigated as a potential problem.
TMI-1 UFSAR APPENDIX 14A 14A-18 REV. 21, APRIL 2012 7.1.2.2 Feedwater Line Ruptures With its significantly lower energy content, the feedwater line rupture does not represent a problem with respect to compartment pressurization.
7.1.3 PIPE THRUST AND JET IMPINGEMENT The thrust forces developed by a jet flow being expelled from a ruptured pipe and calculated from the internal pipe pressure and the density of the mass being accelerated out of the rupture area. The conditions used to calculate this thrust assume that the rupture occurs over 1 millisecond with the maximum thrust force being established by using the corresponding pressure and mass flow conditions. As the blowdown of the system progresses, the pipe stagnation pressure is assumed to be equal to the static pressure at the rupture location for both longitudinal breaks and for circumferential breaks. The thrust calculation used for these analyses includes the integrated pressure and momentum effects and does not take advantage of upstream flow restrictions to reduce the pipe pressure.
Thrust forces on those lines not affected by blowdown characteristics during the first 15 seconds were assumed as 1.26 PA for steam and 2.0 PA for subcooled fluid.
The forces on targets in the path of escaping fluids are dependent on the size and shape of the target and its distance from the rupture area. Since the actual shape of the rupture dictates the flow field shape being generated, it was assumed that a typical rupture is circular and the free stream expansion of the jet to be conical with an included angle of 30 degrees. Calculation of the force on an object was determined by assuming that the dynamic pressure developed at the rupture exit is for the maximum mass flow and pressure conditions in the high energy line and is inversely proportional to the cross section area of the conical expansion being generated. This dynamic pressure was applied on the targets assuming a target drag coefficient of 2.0, i.e.,
complete stagnation of the escaping fluid.
7.2 STRUCTURAL ANALYSIS AND DESIGN 7.2.1 WALL AND SLAB CAPACITIES Postulated ruptures of the main steam lines at the containment penetrations were investigated to determine differential pressures in compartments 2 through 5. Peak differential pressures are presented in Section 7.1.2. To investigate the retention or loss of structural integrity due to breaks within the small compartments, the load carrying capacity of slabs and walls was calculated and compared to expected differential pressure.
7.2.1.1 General Yield-Line Theory and Shear Capacity Yield-Line Theory (References 3, 4, and 5) takes into consideration the inelastic behavior of the reinforced concrete structural element (wall or slab) in developing a mechanism prior to loss of structural integrity. In brief, the steps involved in the evaluation of the uniform pressure capacity of slabs and walls are as follows:
TMI-1 UFSAR APPENDIX 14A 14A-19 REV. 21, APRIL 2012 a.
The ultimate moment capacity of the cross section is calculated at various negative and positive moment regions, i.e., at the boundaries and at midspan. The sections are typically doubly reinforced and the moment capacity is given conservatively by:
Mu = 0.9AsFy (1d - a/2) where:
Mu = moment capacity/ft As = area of tension steel/ft Fy = yield strength of reinforcing steel d = effective depth from steel centroid to extreme compression fiber a = depth of equivalent compression concrete The estimate is conservative in that it neglects the increase in strength of the tension steel after yielding occurs. Also, section capacities are reduced by a factor of 0.9.
b.
The support conditions for the slab or wall under consideration are determined. Where the supports of a given wall or slab are connected to slabs or walls of approximately equal thickness and when the reinforcing steel is sufficiently anchored to develop its strength, the support condition is taken as being fixed. Where the above conditions are not satisfied, the support condition is assumed to be simple support or free in the case of free edges adjacent to containment.
c.
The correct yield-line pattern is established by trial and error using the Principle of Virtual Displacements or The Equilibrium Method. Sufficient trails are executed to determine the uniform load-carrying capacity (absolute minimum of all trials). A 10 percent reduction is applied to all results to allow for corner effects.
The uniformly distributed load corresponding to the correct yield-line pattern is taken to be the differential pressure capacity of the wall or slab in bending.
d.
Each wall or slab is checked for punching shear (two way action) around its periphery (the perimeter is defined at a distance d/2 from the support lines) and for local shear (one way action) at a distance d from support faces.
The concrete shear strength for two way action is taken as 4 f c and the slab capacity is calculated by:
Wps = 4 f c bod Ap where:
Wps = pressure-producing punching shear capacity
TMI-1 UFSAR APPENDIX 14A 14A-20 REV. 21, APRIL 2012 b8 = perimeter length d = effective depth Ap = slab or wall surface area bounded by b The concrete shear strength for one way action is taken as:
Vc = 1.9 c
f '
+ 2500 P Mu Vd 3.5 c
f '
as defined in Reference 6. The slab capacity is then calculated by considering the most critical section at d distance from a support face.
The wall or slab capacity as controlled by shear is taken as the lesser of the two values described above.
e.
The maximum differential pressure capacity is then the lesser of the two values from c.
(as controlled by moment) and d (as controlled by shear).
7.2.1.2 Compartment Differential Pressures Versus Capacities The differential pressure results defined in Section 8.1.2 for circumferential breaks in compartments 2 through 5 are presented in Table 14A-5. A comparison of peak differential pressures to the wall or slab capacities indicates the following:
a.
Structural slabs and walls in compartments 2 and 5 may experience excessive differential pressures due to postulated breaks. Protection will be provided as described in Section 4.2.d.1 of this Appendix.
b.
Postulated breaks in compartments 3 and 4 should not result in wall or slab failure and no protection is required.
7.2.1.3 Jet Impingement And Combined Differential Pressure Jet impingement is a direct result of either a longitudinal or circumferential pipe break.
Longitudinal pipe rupture, which can occur at any orientation about the circumference of the pipe at the break point, results in a jet axis perpendicular to the longitudinal axis of the pipe.
Circumferential (guillotine) breaks result in a jet axis parallel to the longitudinal axis of the pipe.
The jet is assumed to diverge from the break at a 30 degree conical angle, and the total integrated force on an object is determined by assuming that the dynamic pressure developed at the rupture exit is inversely proportional to the cross sectional area of the conical expansion cone at any station. Thus, the walls and slabs are evaluated for the effects of the jet.
After investigating various postulated pipe ruptures, the most severe case of jet impingement from a main steam break in compartments 3 and 4 is a longitudinal break acting on the wall between compartments 4 and 5. This is due to the fact that the main steam pipe is approximately 2 feet away from the wall, and the jet impingement force from a postulated side split break at the containment vessel penetration strikes the wall at its unsupported edge. Jet
TMI-1 UFSAR APPENDIX 14A 14A-21 REV. 21, APRIL 2012 loads in compartments 2 and 5 were not investigated since protection against those breaks is provided as stated in Section 4.2.d.1 of this Appendix.
The capability of the wall between compartments 4 and 5 to resist the jet impingement force was analyzed using Yield-Line Theory. The jet impingement load was conservatively taken as a point load acting at the free edge of the wall. This analysis assumed a semicircular fan-shaped crack pattern and resulted in an ultimate capacity approximately three times higher than the jet force.
The wall was then evaluated for its ability to resist the punching shear caused by the jet impingement. The punching shear capacity of the wall was found to be approximately four times the punching shear stress caused by the jet impingement force.
If the impingement force is assumed to be instantaneous and a dynamic load factor of two is used, both the moment and shear checks stated above are more than satisfactory. The jet load in this case was not combined with pressure since the peak jet force occurs at the instant of break and the peak pressure occurs later.
The wall between compartments 4 and 5 was also analyzed for the combined effects of jet impingement and pressurization at a time subsequent to pipe break. The ultimate moment capacities of the wall needed to resist the pressurization load and the jet impingement was less than the available ultimate moment capacity of the wall. The technique for combining the two loadings is described in Reference 5. In a similar manner, the combined shear stress caused by pressurization and jet impingement was found to be less than the available shear capacity of the wall. All other cases of combined pressurization and jet impingement in compartments 3 and 4 were less critical than the case discussed.
7.2.2 RESTRAINT DESIGN PROCEDURE 7.2.2.1 Assumptions The assumptions involved in designing restraints are as follows:
a.
Guaranteed minimum yield strength of pipe steel reduced in accordance with operating temperature.
b.
Guaranteed minimum ultimate strength of the pipe steel is unaffected by temperature.
c.
Ultimate strain of both piping and restraint material is one half of guaranteed minimum percent elongation.
d.
Guaranteed minimum values of yield strength and ultimate strength for restraint material A 36 are taken from applicable ASTM specifications.
e.
A 10 percent increase in material properties is applied to allow for strain rate effect.
The design procedure discussed in the following section utilizes computer program DYREC (Reference 7). Final designs of restraints were investigated using the techniques presented.
TMI-1 UFSAR APPENDIX 14A 14A-22 REV. 21, APRIL 2012 7.2.2.2 Analytical Method The dynamic analysis of lumped mass models of the rupturing pipe and restraint system was performed by direct numerical time integration of the equations of motion. The computer program DYREC includes the following capability:
a.
Element types; bilinear beam, bilinear axial or rotational spring, "special" axial or rotational spring.
b.
Plastic, elastic, or elasto-plastic impact after closing specified gaps.
c.
Constant zero or non-zero nodal boundary conditions.
d.
Piecewise linear force-time histories.
7.2.2.3 Material Properties Pipe Material - A 106 Grade B Fy = Yield strength at 600F
= 25.9 ksi x 1.1 = 28.5 ksi Fult = Ultimate strength
= 60 ksi x 1.1 = 66.0 ksi
% Elongation
= 22 percent E = Modulus of elasticity
= 25.7 x 103 ksi Structural Steel - A36 Fy = 36 ksi x 1.1
= 39.6 ksi Fult = 60 ksi x 1.1
= 66.0 ksi
% Elongation
= 20 percent E = Modulus of elasticity
= 30 x 103 ksi 7.2.2.4 Section Properties a.
Pipe Cross Section The bilinear moment versus curvature relationship is defined by the following points:
My
= FyS Mult
= FyZ + (Fult-Fy )S Where:
TMI-1 UFSAR APPENDIX 14A 14A-23 REV. 21, APRIL 2012 S
= Elastic Section Modulus Z
= Plastic Section Modulus Ey = F/E = Yield Strain Eult = (% elongation)/2 = Ultimate Strain
Øy = 2 Ey/d = Yield Curvature
Ø ult = 2 Eult/d = Ultimate Curvature Where:
d = O.D. of pipe cross-section The shear versus shearing strain relationship for pipe cross section is defined by:
Fsy
= Shear Yield Stress* = Fy/3 As
= Effective Shear Area 8 = 0.53 A pipe Yy
= Shear Yield Force
= As Fy G = Shear Modulus
= 0.4E Yy
= Shear Yield Strain = Yy/(As G)
Yult = Ultimate Shear Force
= 2Yy Yult = Ultimate Shear Strain
= 309
- Von Mises Criteria b.
Structural Wide Flange Cross Section In defining the moment versus curvature relationship and shear versus shearing strain, the moment is assumed to be carried by the flanges and the shear by the web.
My
= Fy Af (d-tf)
Where:
Af
= Area of One Flange d
= Depth of Cross-Section tf
= Flange Thickness
TMI-1 UFSAR APPENDIX 14A 14A-24 REV. 21, APRIL 2012 Mult
= Fult a (d-tf)
Ey
= 2 Ey/d Eult
= 2 Eult/d Yy
= Fsy A Web Vy
= Vy/(Aweb G)
Vult
= 2 Vy Vult
= 30%
The bilinear moment versus curvature and shear versus shearing strain curves are illustrated on Figure 14A-19.
7.2.2.5 Example Model The pipe condition at rupture, shown on Figure 14A-20 part A, is representative of a postulated break in the 12 inch main steam header. Figure 14A-20 B part illustrates the dynamic model of the same condition.
The restraint is positioned so as not to interfere with normal piping operation. Therefore, the model has a gap of 4 inches between nodes 3 and 8 at the instant F(t) is applied, i.e., at the time postulated rupture occurs. Node 8, in this case, is representative of the mass of the header that is assumed to break away. The modeling is conservative in that credit is not taken for the restraint offered by the 12 inch branch line coming into the 12 inch header.
Nodes 1 through 7 and bilinear beam elements 1 through 6 represent two structural wide flange sections embedded in the existing slab.
The fineness of lumping as well as the magnitude of time increment for numerical integration are selected to ensure a reasonable approximation of the dynamic transient. The total time of execution of the mathematical model on the computer is set to allow multiple impacts of the rupturing pipe and observance that the selected restraint is indeed bringing the rupturing pipe to a stable steady state condition.
7.2.2.6 Detail Design The details of the restraint design are considered in two parts, steady state and transient effects.
The restraint is proportioned such that after the transient occurs, the shears, moments, reactions, and so forth, are within the allowable values of applicable codes (e.g., AISC, manual of Steel Construction, 7th Edition). During the transient, the maximum element curvatures and shears are limited to approximately one half of their ultimate values.
TMI-1 UFSAR APPENDIX 14A 14A-25 REV. 18, APRIL 2006
8.0 REFERENCES
1.
Redfield, J. A., Murphy, J. H., and Davies, V. C., "FLASH-2 A Fortran IV Program for the Digital Simulation of a Multimode Reactor Plant During Loss-of-Coolant,"
WAPD-TM-666 (April 1967).
2.
Rettig, W. H., Jayne, G. A., Moore, K. V., Slater, C. E., Uptmor, M. L., "RELAP 3 - A Computer Program for Reactor Blowdown Analysis," IN-1321 (June 1970).
3.
Ferguson, P. M., "Reinforced Concrete Fundamentals," Second Edition, John Wiley &
Sons, Inc., New York, New York, 1965, pp. 318-348.
4.
Hognestad, E., "Yield-Line Theory for the Ultimate Flexural Strength of Reinforced Concrete Slabs," Journal of the American Concrete Institute, Volume 24, No. 7, March 1953.
5.
Wood, R. H., "Plastic and Elastic Design of Slabs and Plates," The Ronald Press Company, New York, 1961.
6.
"Building Code Requirements for Reinforced Concrete, ACI Standard 318-71, American Concrete Institute, Detroit, Michigan, 1970.
7.
DYREC, Dynamic Response Calculator, January 1973, Gilbert Associates Computer Program (Company Proprietary).
8.
Cowper, G. R., "The Shear Coefficient in Timoshenko's Beam Theory," Journal of Applied Mechanics, June 1966.
9.
Hall, W. J. and Newmark, N. M., "Shear Deflection of Wide Flange Steel Beams and the Plastic Range," Journal of Engineering Mechanics Division, ASCE, Vol. 81, October 1955.
10.
CRAFT - Description of Model for Equilibrium LOCA Analysis Program, B&W Topical Report BAW-10030.
11.
Richardson, L. C., Finnegan, L. J., Wagner, R. J., and Waage, J. M., "CONTEMPT," A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of-Coolant Accident, Phillips Petroleum Co., Atomic Energy Division, Idaho Falls, Idaho, AEC Research and Development Report TID-4500, issued June 1967.
12.
Atomic Energy Commission (AEC) Request, General Information Required for Consideration of Effects of Piping System Break Outside Containment," December 15, 1972.
13.
GPU Service Technical Data Report - TDR 250 "Review of Intermediate Building Flooding Following a Feedwater Line in the Intermediate Building of TMI Unit 1.
14.
GPUNC letter to NRC, #5211-84-2193 dated August 1, 1984.
Subject:
Intermediate Building Flooding Modification.
TMI-1 UFSAR APPENDIX 14A 14A-26 REV. 18, APRIL 2006 15.
USNRC Standard Review Plan Section 3.6.2 - Determination of Rupture Locations and Dynamic Effects associated with Postulated Rupture of Piping, Rev. #1, July 1981.
16.
USNRC Branch Technical Position MEB3 Postulated Rupture Locations in Fluid System Piping Inside and outside Containment, Rev. #1, July 1981.
17.
GPUN Calculation C-1101-211-E540-076, Letdown Line Break Structural and Dynamic Assessment, Rev. 1.
18.
GPUN Calculation C-1101-211-E540-077, RELAP5 Analysis of Letdown Line Break in the Auxiliary Building, Rev. 0.
19.
GPUN Calculation C-1101-211-E540-078, GOTHIC Analysis of Auxiliary Building EQ Environment from a Letdown Line Break, Rev. 0.
20.
NRC Internal Memorandum (Marsh to Milano) dated February 6, 1997,
Subject:
Three Mile Island, Unit 1 - Licensing Basis for Letdown Line Pipe Break Outside Containment.
21.
NRC Letter (Wiggins to Langenbach) dated October 15, 1998,
Subject:
NRC Inspection Report No. 50-289/98-06 and Notice of Violation. GPUN File No. 1920-98-30622.
TMI-1 UFSAR APPENDIX 14 A14A-27 REV. 21, APRIL 2012 TABLE 14A-1 (Sheet 1 of 2)
High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures Description of Line
- of lines Nom Line OD, in.
Type of Fluid Contained Normal Operation Main Bldg Areas Occupied by these lines Remarks Temperature oF Pressure psig Intermediate Bldg Elev.
Turbine Bldg Elev.
Auxiliary Bldg Elev.
Fuel Handling Bldg Elev.
Main Steam (MS) 4 24 Superheated Steam 570 900 337 7 & 347 1 347 1 None None None Main Steam Dump 2
12 Superheated Steam 570 900 346 0 345 8 & 342 9 None None None MS supply to EFW Pump Turbine and MS Dump Lines 4
8 indr takeoffs Superheated Steam 570 900 338 8 None None None 8 inch takeoffs rise off MS Line in small Intermediate Building cubicles and drop vertically into 10 inch headers that terminate at end of main headers 2
10 Sub header
~ 314' 1
12 header 297 0 Main Feedwater (FW) 2 20 FW Pump Disch Hot Water (subcooled) 455
>950 346 0 to 324 0 &
316 2 30 - 324 3 None None Pressure varies due to HD loss in piping system. One 20 inch discharge from common 30 inch manifold is in Int.
Building.
The rest are in Turbine Building 1
30 Common Man.
20 - 346 0 & 324 0
2 20 Outlets from common man.
Aux Steam (AS) to EFW Pump Turbine 1
4 Saturated steam 382 185 344 0 to 303 6 None None None None Emergency Feedwater Pump Discharge (EFW) 1 6
Cold Water 90
>1000 304 0 & 306 6 None None None The 3 EFW pump discharge lines tie into common 6 inch manifold and two 6 inch outlet lines feed out of this into the Reactor Building.
2 4
297 7 & 306 6 2
6 298 3 & 318 0 Aux Steam (AS) to Auxiliary Building 1
4 Saturated steam 227 5
None 365 10 289 2 to 288 8 363 6 & 362 6 to 329 0 In passageway between Fuel Handling to Building and Control Tower.
One 2 inch branches off within Fuel Handling Building at 339 ft.
1 6
366 7 & 356 6 288 8 to 294 1 329 0 to 328 8 1
8 356 6 & 363 7 294 1 to 296 0 301 3 to 289 2
TMI-1 UFSAR APPENDIX 14 A14A-28 REV. 21, APRIL 2012 TABLE 14A-1 (Sheet 2 of 2)
High Energy Lines Investigated for Consequences of Postulated Pipe Ruptures Description of Line
- of lines Nom Line OD, in.
Type of Fluid Contained Normal Operation Main Bldg Areas Occupied by these lines Remarks Temperature oF Pressure psig Intermediate Bldg Elev.
Turbine Bldg Elev.
Auxiliary Bldg Elev.
Fuel Handling Bldg Elev.
Makeup Pump Discharge 3
4 Pump Disch Cold Water 135
>2250 None None 287 9 to 284 3 291 0 Refer to Figure 3 for details of pipe routing of discharge piping from makeup pumps 1
4 Comm Manf 3
4 Manf Outlet Letdown to Makeup &
Purification 1
2 1/2 Cold Water 120
~ 2200 None None 291 6 289 0 282 0 Splits into 3 branch manifolds where pressure is broken down to between 25 and 75 psig Decay Heat Suction 1
12 RC LETDOWN Water varying from hot to cold 250 to between 110 & 140 300 to atmospheric plus static head None None 291 6 263 0 None 2
12 BRANCH 2
14 PUMP SUCTION Decay Heat Discharge 2
10 Water varying from hot to cold 210 to between 100
& 130
~ 400
~ 150 None None 272 9 274 0 291 6 None None Reactor Coolant Sample 1
3/8 Hot water (subcooled) and steam Between 540 &
605 Steam is saturated
~ 2200 None None 312 0 Between Reactor Bldg and FH Bldg Line passes through passageway between FH Building and Control Tower and enters Control Tower as same place as Steam Gen. sample, below Steam Generator Secondary Side Sample 2
3/8 Hot water (subcooled)
~ 500
~ 900 None 315 6 None 315 6 In passageway between FH Bldg and Control Tower Line enters radioactive sample station in room at elevation 306 feet 0 inch near northwest corner of Control Tower
TMI-1 UFSAR APPENDIX 14 A14A-29 REV. 21, APRIL 2012 TABLE 14A-2 (Sheet 1 of 2)
SYSTEMS REQUIRED FOR SAFE SHUTDOWN FOLLOWING A POSTULATED MAIN STEAM OR FEEDWATER PIPE RUPTURE System Building Equipment Location Elevation Area Makeup and Purification System Auxiliary 281-0 Northeast section Decay Heat Removal Systems D.H. Pumps and Coolers Auxiliary 261-0 North-central section Closed Cycle Cooling System Coolers Auxiliary 271-0 West section Closed Cycle Cooling Pumps section Auxiliary 305-0 North-central River Water Pumps Screen House 308-0 East-central section Borated Water Storage Tank (BWST)
Outdoors 306-0 Near northeast corner of Auxiliary Building Nuclear Services Cooling Water System Closed Cycle Cooling Coolers Auxiliary 271-0 West section Closed Cycle Cooling Water Pumps Auxiliary 305-0 North-central section River Water Pumps Screen House 308-0 East-central section Air Handling (ES)
Reactor Building Cooling Units Reactor 287-0 East-central section Cooling Water Piping Intermediate 295-0 North-central section Reactor Building Coolers Reactor 291-0 West end Diesel Generators Diesels Diesel 306-0 Most of Bldg.
area Fuel Tanks Underground Outside Below grade North of Diesel Generator Building TABLE 14A-2
TMI-1 UFSAR APPENDIX 14 A14A-30 REV. 21, APRIL 2012 (Sheet 2 of 2)
SYSTEMS REQUIRED FOR SAFE SHUTDOWN FOLLOWING A POSTULATED MAIN STEAM OR FEEDWATER PIPE RUPTURE Equipment Location System Building Elevation Area Spent Fuel Cooling System Fuel Handling 281-0 Central west side Pressurizer Safety Valves and PORV Reactor 354-0 On top pressurizer RPS and ES Actuation Reactor and Auxiliary RPS and ES Actuation Racks Control Emergency Feedwater Intermediate 295-0 East section Atmospheric Dump Valves Intermediate 295-0 EF-P-1 &
Bypass Header Room
TMI-1 UFSAR APPENDIX 14 A14A-31 REV. 21, APRIL 2012 TABLE 14A-3 (Sheet 1 of 1)
TABULATION OF SYSTEMS AND BREAK LOCATIONS Primary +
Secondary Secondary Stress, % of Figure No.
Break
- Street, Allowable and Locations
% of (deadweight Reference Indicates on Allowable
+ Thermal System Isometrics Isometrics (Thermal)
+/-Seismic)
Generator "B" Figure 14A-11 29 63.9 57.5 to Header 43 end 7.4 21.2 at EL 297'-0" 13 end 6.7 59.3 12 20.6 2.8 10 29.7 43.6 9 end 32.4 45.5 Feedwater ISO 13 4.3 36.5 from 30" Header 28 27.0 49.1 to Penetration Figure 14A-12 30 28.9 50.6 103 44 end 22.3 55.4 Feedwater ISO 1 end 50.7 49.4 from 30" Header 12 27.8 46.1 to Penetration Figure 14A-13 8
36.7 44.2 227 53 end 42.3 51.7 Main Stream Figure 14A-1 1 end 00.7 21.2 Dump to 116 89.0 95.8 Intermediate 115 87.3 88.8 Wall 117 end 44.2 59.7 Main Stream Figure 14A-1 1 end 43.5 53.8 Dump to 29 72.3 77.7 Intermediate 28 71.0 77.2 Wall 43 end 27.8 46.0
TMI-1 UFSAR APPENDIX 14 A14A-32 REV. 21, APRIL 2012 TABLE 14A-4 (Sheet 1 of 2)
MAXIMUM COMPARTMENT PRESSURES Circumferential Break Steam Line Rupture in Compartment 2 Maximum Pressure in Compartment 2 = 82.1 psia = 67.4 psig maximum Differential Pressure, Subcompartment 2 to Subcompartment 3 = 65.7 psid 6 = 63.0 psid 1A = 66.4 psid 12 = 66.5 psid Reactor Bldg = 67.4 psid Turbine Bldg
= 67.4 psid Steam Line Rupture in Compartment 3 Maximum Pressure in Compartment 3 = 45.1 psia = 30.4 psig Maximum Differential Pressure, Subcompartment 3 to Subcompartment 2 = 27.4 psid 4 = 29.8 psid 1B = 26.6 psid 7 = 28.8 psid 13 = 3.8 psid Reactor Bldg = 30.4 psid Steam Line Rupture in Compartment 4 Maximum Pressure in Compartment 4 = 36.3 psia = 21.6 psig Maximum Differential Pressure, Subcompartment 4 to Subcompartment 3 = 20.8 psid 5 = 20.8 psid 1C = 12.9 psid 14 = 2.4 psid 8 = 18.8 psid Reactor Bldg = 21.6 psid
TMI-1 UFSAR APPENDIX 14 A14A-33 REV. 21, APRIL 2012 TABLE 14A-4 (Sheet 2 of 2)
MAXIMUM COMPARTMENT PRESSURES Circumferential Break Steam Line Rupture in Compartment 5 Maximum Pressure in Compartment 5 = 70.1 psia = 55.4 psig Maximum Differential Pressure, Subcompartment 5 to Subcompartment 11 = 54.3 psid 4 = 54.3 psid 1D = 53.5 psid 15 = 54.3 psid 9 = 52.3 psid Reactor Bldg = 55.4 psid Longitudinal Break Steam Line Rupture in Compartment 3 Maximum Pressure in Compartment 3 = 33.07 psia = 18.37 psig Maximum Differential Pressure, Subcompartment 3 to Subcompartment 2 = 15.99 psid 4 = 17.27 psid 1B = 16.05 psid 7 = 16.62 psid 13 = 1.32 psid Reactor Bldg = 18.37 psid Steam Line Rupture in Compartment 4 Maximum Pressure in Compartment 4 = 21.67 psia = 6.97 psig Maximum Differential Pressure, Subcompartment 4 to Subcompartment 3 = 3.57 psid 5 = 3.52 psid 1C = 1.62 psid 14 = 1.02 psid 8 = 2.87 psid Reactor Bldg = 6.97 psid
TMI-1 UFSAR APPENDIX 14 A14A-34 REV. 21, APRIL 2012 TABLE 14A-5 (Sheet 1 of 2)
COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE Peak Break In Differential Structural Compartment Capacity Pressure Member No.
(psi)
(psi)
Slab, Compartment 2 2
47.8 66.4 Wall, Compartment 2 to 3 2
44.4 65.7 Wall, Compartment 2 to 6 2
39.0 63.0 Roof, Compartment 2 2
54.0 66.5 Slab, Compartment 3 2
32.3 26.6 Wall, Compartment 3 to 4 2
43.7 29.8 Wall, Compartment 3 to 7 3
39.0 28.8 Wall, Compartment 3 to 12 3
44.4 27.4 Roof, Compartment 3 3
36.5 3.8 Slab, Compartment 4 4
9.3 12.9 Wall, Compartment 4 to 3 4
43.7 20.8 Wall, Compartment 4 to 5 4
53.2 20.8
TMI-1 UFSAR APPENDIX 14 A14A-35 REV. 21, APRIL 2012 TABLE 14A-5 (Sheet 2 of 2)
COMPARISON OF WALL AND SLAB CAPACITY TO DIFFERENTIAL PRESSURE Peak Break In Differential Structural Compartment Capacity Pressure Member No.
(psi)
(psi)
Wall, Compartment 4 to 8 4
40.2 18.8 Roof, Compartment 4 4
33.5 2.4 Slab, Compartment 5 5
19.2 53.5 Wall Compartment 5 to 4 5
53.2 54.3 Wall, Compartment 5 to 9 5
47.0 52.3 Roof, Compartment 5 5
50.1 54.3
TMI-1 UFSAR APPENDIX 14 A14A-36 REV. 21, APRIL 2012 TABLE 14A-6 (Sheet 1 of 1)
CHRONOLOGY OF EVENTS FOR HIGH ENERGY PIPE BREAK TIME (SEC)
EVENT 0
Double-ended break of a 24 inch diameter steam line on the secondary side 1
Reactor trip on variable low pressure; turbine stop valves close, isolating the unaffected steam generator 47 Damaged steam generator blows dry 450 Unaffected steam generator provides no more heat sink; minimum system pressure of about 1550 psia is reached 900 Operator action starts one HPI pump 1200 Primary loop becomes solid with subcooled water; pressurizer code relief valve opens at set point of 2515 psia 2300 Reactor Building cooler actuation set point of 4 psig is reached 5700 Steam first appears in the core 8500 Minimum coolant level in reactor vessel is reached; core remains covered 8800 Containment Building pressure reaches the maximum value of 24 psig
TMI-1 UFSAR APPENDIX 14 A14A-37 REV. 21, APRIL 2012 TABLE 14A-7 (Sheet 1 of 1)
THREE MILE ISLAND NUCLEAR STATION - UNIT-1 INTERMEDIATE BUILDING ENVIRONMENTAL CONDITIONS Main Steam Line Break at 322'-0" Elevation Peak (F)
Peak (psia)
Pressure & Temperature Elevation (ft)
Temperature Pressure Profiles 355 187.0 24.2 322 326.0 24.2 Figure 14A-23 295 322.0 23.6 Figure 14A-21 Main Steam to EFW Pump Turbine Line Break 355 95.0 16.7 322 205.0 16.7 Figure 14A-24 295 273.0 16.7 Figure 14A-22
TMI-1 UFSAR APPENDIX 14 A14A-38 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 1 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 0.0 0.0 0.0 0.0 0.0 0.05 4.973+3 1241.3 6.476+3 1234.0 0.1 5.962+3 1234.9 4.715+3 1234.3 0.1001 5.000+2 1234.9 4.715+33 1234.3 0.2 5.000+2 1234.9 4.411+3 1204.6 0.3 5.000+2 1234.9 4.214+3 1199.7 0.4 5.000+2 1234.9 4.038+3 1200.5 0.5 5.000+2 1234.9 3.876+3 1201.2 0.75 5.000+2 1234.9 3.530+3 1202.5 1.0 5.000+2 1234.9 3.252+3 1203.5 1.5 5.000+2 1234.9 2.923+3 1204.3 2.0 5.000+2 1234.9 2.700+3 1204.9 3.0 5.000+2 1234.9 2.224+3 1205.4 4.0 5.000+2 1234.9 1.948+3 1205.0 5.0 3.518+2 1174.6 2.186+3 1214.1 6.0 3.00 +2 1172.4 2.299+3 1206.3
TMI-1 UFSAR APPENDIX 14 A14A-39 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 2 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 7.0 3.10+2 1172.9 2.465+3 1205.8 8.0 3.29+2 1173.7 2.556+3 1205.9 9.0 3.38+2 1174.0 2.630+3 1205.4 10.0 3.46+2 1174.4 2.689+3 1205.1 11.0 3.50+2 1174.6 2.702+3 1204.9 12.0 3.49+2 1174.5 2.693+3 1204.9 13.0 3.47+2 1174.4 2.669+3 1205.0 14.0 3.43+2 1174.3 2.637+3 1205.0 15.0 3.39+2 1174.1 2.604+3 1205.1 17.0 3.21+2 1203.8 2.534+3 1205.2 20.0 3.07+2 1205.0 2.440+3 1205.3 25.0 2.88+2 1205.0 2.286+3 1205.4 30.0 2.69+2 1205.6 2.150+3 1205.3 35.0 2.50+2 1205.4 2.001+3 1205.2 40.0 2.33+2 1205.2 1.864+3 1204.8
TMI-1 UFSAR APPENDIX 14 A14A-40 REV. 21, APRIL 2012 TABLE 14A-8 (Sheet 3 of 3)
MASS AND ENERGY RELEASE RATES FOR 24 INCH MAIN STEAM LINE BREAK Time After Piping OTSG Rupture Blowdown Enthalpy Blowdown Enthalpy (sec)
(lbm/sec)
(Btu/lbm)
(lbm/sec)
(Btu/lbm) 45.0 2.18+2 1204.8 1.748+3 1204.5 50.0 2.05+2 1204.4 1.642+3 1204.0 55.0 1.93+2 1203.9 1.543+3 1203.5 60.0 1.81+2 1203.3 1.452+3 1202.9 70.0 1.57+2 1201.9 1.200+3 1201.3 75.0 5.79+1 1197.2 6.128+2 1189.9 75.0-3600 0.0 0.0 8.700+1 1234.4
TMI-1 UFSAR APPENDIX 14 A14A-41 REV. 21, APRIL 2012 TABLE 14A-9 (Sheet 1 of 1)
CONCRETE PROPERTIES Thermal Volumetric Temperature Conductivity Heat Capacity (F)
K, (Btu/hr - Ft20F)
(Btu/ft30F) 20 0.80 30.0 200 0.81 30.0 300 0.82 30.0