ML20247G928

App 5A, Overpressure Protection for C-E Sys 80 Pwrs, to CESSAR Sys 80+ Std Design

ML20247G928
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Site:	05200002, 05000470
Issue date:	03/30/1989
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
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ML20247G537	List: LD-89-035, Forwards Vols 1-17,consisting of Chapters 1-18,to CESSAR Design Certification, for Approval,Per 10CFR52 ML20247G666 ML20247G678 ML20247G714 ML20247G851 ML20247G862 ML20247G868 ML20247G913 ML20247G928 ML20247G983 ML20247G989 ML20247H033 ML20247H291 ML20247H308 ML20247H316 ML20247H336 ML20247H353 ML20247H360 ML20247H394 ML20247H397 ML20247H403 ML20247H417 ML20247H427 ML20247H443 ML20247H453 ML20247H470 ML20247H478 ML20247J059 ML20248B002 ML20248C380 ML20248C392
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NUDOCS 8904040347
Download: ML20247G928 (16)

Text

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O APPENDIX SA OVERPRESSURE PROTECTION FOR COMBUSTION ENGINEERING SYSTEM 80 - PRESSURIZED WATER REACTORS O

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I APPENDIX 5A O

OVERPRESSURE PROTECTION FOR COMBUSTION ENGINEERING SYSTEM 80 - PRESSURIZED WATER REACTORS ',

ABSTRACT This Appendix documents the adequacy of overpressure protection provided for Combustion Engineering's (C-E's) System 80 pressurized water reactor, steam generators, and Reactor Coolant System.

Overpressurization of the Reactor Coolant System and steam generators is precluded by means of primary safety valves, secondary safety valves and the Reactor Protection System. Pressure relief capacity for the steam i generators and Reactor Coolant System is conservatively sized to satisfy the overpressure requirements of the ASME Boiler and Pressure Vessel Code,

, Section III. The safety valves in conjunction with the Reactor Protection l

System, are designed to provide overpressure protection for a loss-of-load incident with a delayed reactor trip.

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\ CHAPTER 5 APPENDIX 5A TABLE OF CONTENTS Page Amendment Abstract i 9 ii Text Page Amendment 5A-1 SA-2 9 5A-3 9 5A-4 10 SA-5 10 O Figures Amendment 5A-1 5A-2 5A-3 5A-4 O

Amendment No. 10 June 28, 1985

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l*sg  : APPENDIX 5A l OVERPRESSURE PROTECTION FOR COMBUSTION ENGINEERING 1.

I SYSTEM 80 - PRESSURIZED WATER REACTORS-TABLE'0F CONTENT 7; SECTION TITLE PAGE NO.

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1.O INIR0aufl10N SA 2.O ANALYSIS SA-1 2.i METHOD SA-1 2.2 ASSUMPTIONS SA 2.2.1 SECONDARY SAFETY VALVE SIZING 5A-2 2.2.2 PRIMARY SAFETY VALVE SIZING 5A-3 3.0 CONELUS10HS 5A-4

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LIST OF FIGURES FIGURE TITLE SA-1 Optimized Safety Valve Capacities 5A-2 Steam Generator Pressure Complete Loss of Turbine Generator Load 5A-3 Maximum Reactor Coolant System Pressure vs.

Time for Worst Case Loss of Load Incident SA-4 Maximum Reactor Power vs. Time for Worst Case Loss of Load Incident O

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1.0 .lNIRODEUON Overpressure protection for 2-E's System 80 pressurized water reactor, steam generators, and Reactor Coolant System is in accordance with the requirements set forth in the ASME Boiler and Pressure Vessel Code,Section III. Overpressure protection is considered by means of primary safety valves, secondary safety valves, and the Reactor Protection System. Analysis of all reactor and steam plant transients causing pressure excursions is conducted. The worst case transient, loss-of-load, in conjunction with a delayed reactor trip, is the design basis for the primary safety valves.

The primary safety valves, secondary safety valves, and Reactor Protection System maintain the Reactor Coolant System below 110% of design pressure during worst case transients. The secondary safety valves are sized conservatively to pass steam flow at greater than the proposed licensed power level of 3817 Mwt. Steam generator pressure is limited to less than 110% of steam generator design pressure during worst case transients.

2.0 ANAlYj])

2.1 METHOD C-E has performed a parametric study to determine the design basis incident for sizing the primary safety valves. The design basis incident is a loss-of-load with a delayed reactor trip. The analysis is performed using digital computer codes which accurately model the thermal, hydraulic, and nuclear performances of the Reactor Coolant and Steam Systems. The digital G codes used in the transient analysis include reactor kinetics, thermal and hydraulic performance of the Reactor Coolant System, and the thermal and hydraulic performance of the steam generators. The computer simulation includes effects of reactor coolant pump performance, elevation heads, inertia of surge line water and friction drop in the surge line. Worst case initial conditions and nuclear parameters are assumed for the parametric analysis. The reactor is assumed to trip at a RCS pressure of 2450 psia, while the primary safety valves are assumed to lift at a pressure of 2525 psia, which is 25 psi above the system design pressure. During the analysis, the throat area associated with these valves is increased parametrically until the above design basis incident analysis indicates that a further increase in throat area will r.c result in a significant decrease in RCS peak pressure. The performance of the digital codes employed in the analysis have been verified by transient data from operating plants.

2.2 ASSUMPTIONS

a. At the onset of the loss-of-load transient, the Reactor Coolant and Main Steam Systems are at maximum rated output plus a two percent uncertainty. By choosing the highest possible power output, the heatup rate of the primary loop is maximized, hence the rate of pressurization is also maximized.

b. Moderator temperature coefficient is zero. Analytical stedies supported by core data show that the moderator SA-1

gemperaturecoefficientcanvarybetweenzeroand-3.5x10" for various phases of core life. Therefore, a coefficient of zero is chosen to maximize the power / pressure transient.

c. Doppler coefficient of .8 x 10 -5 AK/K/F is used in the loss-of-load analysis. Actual opgrating coefficients can be exegeted to range from -1.4 x 10 at zero power to -1. x 10 AK/K?F at full power. By choosing a relatively small Doppler coefficient, the reduction in reactivity with increasing fuel temperature is minimized, thereby maximizing the rate of power rise.

d. No credit is taken for letdown, charging, pressurizer spray, turbine bypass, or feedwater addition after turbine trip in the loss-of-load analysis. Letdown and pressurizer spray both act to reduce primary pressure. By not taking credit for these systems, the rate of pressurization is increased.

Including these systems in the analysis will delay tripping the reactor on high primary pressure by 1.1 seconds. The peak primary pressure is not affected. By not taking credit for the addition of feedwater, the steam generator secondary inventory will be depleted at a faster rate. This in turn reduces the capability of the steam generator to remove heat from the primary loop, thereby maximizing the rate of primary pressurization.

e. The analysis reflects consideration of plant instrumentation error and safety valve setpoint errors. For example, all '

safety valves are assumed to open at their maximum popping pressure. This extends the period of time before energy can be removed from the system. The reactor trip setpoint errors are always assumed to act in such a manner that they delay reactor trip, again resulting in maximum pressurization.  ;

f. Pressurizer pressure at the onset of the incident is 2200 l psi. By using the lower limit of the normal plant operating ,

pressure, the time required to trip the plant on high pressure  !

is increased. j 2.2.1 SECONDARY SAFETY VALVE SIZING The discharge piping serving the secondary safety valves is designed to accommodate rated relief capacity without imposing unacceptable backpressure on the safety valves. l The secondary safety valves are conservatively sized to pass excess steam flow. This limits steam generator pressure to less than 110% of steam generator design pressure during worst case transients. A plant's secondary safety valves consist of three banks of valves with staggered set pressures.

The vavles are spring loaded-bellows type safety valves procured in accordance with ASME Boiler and Pressure Vessel Code,Section III.

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Figure 5A-2 depicts the steam generator pressure transient for this worst G case loss-of-load incident. As can be seen in Figure SA-2, the steam generator pressure remains below 110 percent of design pressure during the incident.

2.2.2 PRIMARY SAFETY VALVE SIZING The reactor drain tank, inlet and discharge piping are sized to preclude unacceptable pressure drops and backpressure which would adversely affect valve operation.

Primary safety valve backpressure is limited by the design pressure of the valve bellows. These bellows prevent any accumulated backpressure from being imposed on the valve spring, thus allowing valve operation at its design setpoint rather than at its setpoint plus backpressure.

The design basis incident for sizing the primary safety valves is a loss of

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turbine-generator load in which the reactor is not immediately tripped. No credit is taken for any pressure-reducing devices except the primary and secondary safety valves. In reality, the incident would be terminated by a number of reactor trips. These include:

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a. Steam generator low level trip;

b. High pressurizer pressure trip; e Manual trip.

9 If the high primary pressure trip were to become inoperative, other reactor trips would proceed to shut the reactor down as their setpoints are exceeded.

A series of loss-of-load studies are run with various size; of primary safety valves. As can be seen in Figure 5A-1, after the safety valve capacity increases to a certain size, additional increase in capacity has negligible effect in reducing the maximum system pressure experiences during the loss-of-load transient. C-E's primary safety valves are chosen so as to minimize the maximum pressure experienced during the loss-of-load transient. The minimum specified safety valve capacity is identified on Figure SA-1.

Figures SA-2, 5A-3 and 5A-4 present curves of steam generator pressure, maximum Reactor Coolant System pressure and core power versus time for the worst case loss of turbine-generator load. As can be seen on Figures SA-2 and SA-3, the maximum steam generator pressure and reactor coolant loop pressures remain below l'.% of design during this worst case transient.

The first, second, and third banks of secondary safety valves open at approximately 3.7, 5, and 6.2 seconds, respectively. The secondary safety valves remove energy from the Reactor Coolant System and thus mitigate the pressure surge. The primary safety valves are conservatively assumed to open at 1 percent above the normal Reactor Coolant System design pressure 5.7 seconds after the initiation of the upset condition.

SA-3 Amendment No. 9 February 27, 1984

The analysis of a complete loss of load ' incident is described in Chapter 15, 10 Section 15.2. As demonstrated in this analysis, if a complete loss of load occurs without a simultaneous reactor trip, the protection provided by the high pressurizer pressure trip, primary safety valves and secondary safety valves is sufficient to assure that the integrity of the RCS and main steam system is maintained and that the minimum DNB ratio is not less than 1.19.

2.2.3 ACCEPTABILITY OF SAFETY VALVE BLOWDOWN 2.2.3.1 Background Full scale, full pressure prototgcal testing of pressurizer safety valves The blowdown settings required to insure was performed by EPRI in 1981.

stable valve operation during the blowdown from the set pressure were above  ;

the 5% setting specified in the ASME Code. In order to insure that the extended blowdown would not adversely affect overpressure protection or plant s

operation, analyses were performed to evaluate the NSSS response. The analyses described below demonstrate that a blowdown setting, including associated uncertainties, of 18.5% is acceptable. '

2.2.3.2 Results of Evaluation An extended blowdown of the safety valves could result in swelling of the pressurizer liquid level due to flashing and possible liquid carryover through the safety valves. since the safety valve design specification specifies dry saturated steam flow conditions, it is desirable to show that these conditions are maintained during the extended blowdown. It is also desirable to verify that the RCS remains in a subcooled condition in order that the steam bubble 9 formation in the RCS is precluded.

A computer analysis was performed of the Loss-of-load event with delayed reactor trip, similar to that used in safety va've sizing, except that a conservative 20% safety valve blowdown and initial conditions biased to maximize pressurizer liquid level were assumed. The purpose of this analysis was to determine the pressurizer liquid level response and the RCS subcooling under these conservative conditions. For cdditional conservatism, an additive adjustment was made to the computer-calculated pressurizer levels on the basis of a very conservative pressurizer model. This model assumed that the initial saturated pressurizer liquid did not mix with the cooler insurge liquid, that the initial liquid remained in equilibrium with the pressurizer steam space, and that the steam which flashed during blowdown remained dispersed in the liquid phase and caused the liquid level to swell water level vs time curve showed a maximum of 98%(p)The (1730adjug'ed pressurizer ft ), below the safety valve nozzle elevation of 100%, so that dry saturated steam flow to the l 10 safety valves is assured throughout the blowdown. The computer analysis also showed that adequate subcooling was maintained in the RCS during the blowdown, so that steam bubble formation is precluded.

(1) CEN-227, " Summary Report on the Operability or Pressurizer Safety Relief g Valves in C-E Designed Plants", December 1982.

(2) Water level expressed as the percentage of the distance from the lower l Level nozzle to the upper level nozzle.

SA-4 Amendment No. 10 l June 28, 1985 l

9 In addition, the System 80 safety analyses of pressurization events were re-evaluated to determine the impact of assuming an 18.5% blowdown below nominal set pressure (to 2040 psia) for the pressurizer safety valves in lieu of the 5% specified by the ASME Code. The evaluation indicated that, for the FWLB event analysis, which produces the greatest increase in pressurizer level, the increased blowdown would not result in the pressurizer liquid level g reaching the safety valve nozzle elevation and thus normal safety valve operation would be assured. Further, subcooling in the RCS was maintained during the blowdown.

In summary, analyses show that adequate plant overpressure protection and RCS subcooling are ensured during a blowdown of 18.5% below rominal pressurizer safety valve set pressure.

3.0 CONCLUSION

S C-E's System 80 pressurized water reactor, steam generators, and Reactor Coolant System are protected from overpressurization in accordance with the guidelines set forth in the ASME Boiler and Pressure Vessel Code,Section III.

Peak Reactor Coolant System and Secondary System pressures are limited to less 10 than 110% of design pressures during worst case loss of turbine-generator load. Overpressure protection is afforded by primary safety valves, secondary safety valves, and the Reactor Protection System.

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SA-5 Amendment No. 10 June 28, 1985

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