Text

Proposed Tech Specs Increasing Setpoint Tolerance for Pressurizer & Main Steam Safety Valves from 1% to 2%
	ML20211H402
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	02/11/1987
From:	; PORTLAND GENERAL ELECTRIC CO.
To:	;
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ML20211H337	List: ML20211H333; ML20211H379; ML20211H402;
References
	TAC-64803, NUDOCS 8702260168
	Download: ML20211H402 (86)
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LCA 148 Attachment A 6 Page 1 of 5 REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG (12%).

APPLICABILITY: MODES 4 and 5.

ACTION:

a. With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

b. In the event a safety valve fails or is found inoperable, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances surrounding the failure, including the cause, if known.

SURVEILLANCE REQUIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4.3.

870:N260168 EF70211 PDR ADOCK 05000344 P PDR l

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TROJAN-UNIT 1 3/4 4-3 Amendment No. 97 l I

LCA 148 Attachment A 6 Page 2 of 5 REACTOR COOLANT SYSTEM 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.3.1 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG (12%). !

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

I a. With a pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In the event a safety valve fails or is found inoperable, a

, Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances surrounding the failure, including the cause, if known.

c. The provisions of Specification 3.0.4 may be suspended for one valve at a time for up to 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for entry into and during 4 operation in Mode 3 for the purpose of setting the pressurizer

! code safety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to heatup.

l SURVEILLANCE REQUIREMENTS t

4.4.3 No additional Surveillance Requirements other than those required by i specification 4.0.5.

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TROJAN-UNIT 1 3/4 4-4 Amendment No. 55, 7#, 84, 97

LCA 148 Attachment A 6 Page 3 of 5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-2; otherwise, be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, when tested pursuant to Specification 4.0.5.

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I TROJAN-UNIT 1 3/4 7-1 Amendment No. 7#

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TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING ALLOWABLE VALUE (12%) ORIFICE SIZE A B E D

a. PSV-2211, PSV-2231, PSV-2251, PSV-2271 1170 psig 1147-1193 psig Q = 11.05 sq. inches

b. PSV-2212, PSV-2232, PSV-2252, PSV-2272 1200 psig 1176-1224 psig Q = 11.05 sq. inches

c. PSV-2213, PSV-2233, PSV-2253, PSV-2273 1210 psig 1186-1234 psig R 16.0 sq. inches

d. PSV-2214, PSV-2234, PSV-2254, PSV-2274 1220 psig 1196-1244 psig = 16.0 sq. inches

e. PSV-2215, PSV-2235, PSV-2255, PSV-2275 1230 psig 1206-1254 psig R = 16.0 sq. inches t

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LCA 148 Attachment A 6 Page 5 of 5 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110%

(1305 psig) of its design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The safety valves are tested in accordance with the requirements of Section XI of the ASME Code. The lift setting allowable values are consistent with the safety analysis. The safety analysis is not as restrictive as the ASME Code. Therefore, in the event a safety valve lifts outside of the Section XI allowed tolerance, the Section XI provisions for adjusting the setpoint and testing addi-tional valves applies. Thetotagrelievingcapacityforallvalvesonall of the steam lines is 16.47 x 10 gbs/hr which is 109 percent of the total secondary steam flow of 15.07 x 10 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per cperable steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following bases:

For 4 loop operation SP =

~" x (109)

X For 3 loop operation 3p , (X) - (Y)(U) x (75)

Where:

SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam line TROJAN-UNIT 1 8 3/4 7-1

. . I m 148 Attachmnt B EXECUTIVE

SUMMARY

Trojan plant Technical specifications require that all safety relief valves be operable and lift at a pressure 1 percent of the nominal setpoint. The tolerance of only 1 percent can result l

in a significant number of occurrences of safety valves failing i surveillance . tests. Increased safety valve tolerances will result in fewer failures of safety valve surveillance tests and i less plant down time due to safety valve lift pressures out of l tolerance. This report summarizes a study performed to justify 4

an increase in the allowed tolerance on the safety valve setpoints.

j In evaluating the plant system response to the loss of electric load accident the acceptance criteria is that plant peak systems pressure not exceed 110 percent of design pressure, that is 2750 psia for the primary system and 1320 psia for the secondary. The RELAP5/ MOD 2 advanced systems analysis code was used to calculate l plant system response to the loss of electric load accident as described in the Trojan FSAR. A detailed input model, specific l to the Trojan plant design, was prepared and benchmarked against data taken during turbine trip testing of a similar plant. The ensuing evaluation shows that there is adequate margin in the Trojan plant design to allow pressurizar and steam generator safety valve setpoint tolerances of +4 percent and still meet the acceptance criteria for overpressure protection. Consideration of high pressurizer pressure trip setpoint and steam system I operational pressures supports increasing the magnitude of the negative tolerance to -3 percent.

It is recommended that the Technical Specification allowed value

. of pressurizer and steam generator safety valve setpoint tolerances be increased. This action should reduce the number of l

safety valve surveillance test failures and the time spent in 6

testing and performing maintenance on these valves.

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. TABLE OF CONTENTS EiLER 1.0 Introduction 1 2.0 Trojan RELAPS/ MOD 2 Model 4 2.1 Model Description 4 2.2 Model Initialization 7 2.3 Model Benchmark to Plant Data 7 3.0 Analysis of Loss of Electric Load Accident 10 3.1 Analysis Assumptions 10 3.2 Analysis Results 12 3.3 Consideration of Negative Setpoint Tolerances 13 4.0 Conclusions / Recommendations 14 5.0 Appendices Appendix A Benchmark Case Figures Al Appendix B Accident Analysis Figures B1 t

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1.d INTRODUCTION Trojan plant Technical Specifications require that all safety relief valves be operable and lift at a pressure of 1 percent of the nominai setpoint. The tolerance of only 1 percent can result in a significant number of occurrences of safety valves failing surveillance tests. Larger tolerances on the safety valve setpoints would result in fewer failures of safety valve surveillance tests and less plant down time for safety valve

' maintenance and ratesting. This study was performed to determine if enough margin exists in the overpressurization protection design of the Trojan nuclear plant to justify an increase in the safety valve tolerances. The criteria for success were reactor coolant system and main steam system peak pressures for the loss of electric load accident less than 110 percent of the respective i

system design pressure, that is less than 2750 psia for the j

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! primary and 1320 p.sia for the secondary system. The loss of electric load accident was chosen for study because, as described in section 15.2 of the Trojan FSAR, it conservatively bounds the l

analysis of all events classified by the American Nuclear Society as Condition II events which result in a decrease in heat removal

! by the secondary system.

The RELAPS/ MOD 2 advanced systems analysis code was used to calculate plant system response to the Loss of Electric Load

{

accident as described in the Trojan FSAR. A detailed input model

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was prepared based on the Trojan plant design, and benchmarked against data taken during the turbine trip startup test of the Diablo Canyon Unit 2 plant. The Diablo Canyon plant data is applicable to the Trojan model because of their similar design.

All relevant boundary systems and conditions, e.g. safety relief valves and turbine stop valves, were modelled according to Trojan design specifications.

For the cases of the Loss of Electric Load accident analyzed with

< different safety valve setpoint tolerances, the system response to the transient was essentially the same. The loss of load was characterized by an instantaneous closure of the turbine control valves which results in a less of heat sink for the steam generators. The steam system pressure and temperature increase and heat transfer from the primary to the secondary fluid decreases to less than the amount of heat being transferred from the core to the primary fluid. This results in heating and expansion of the primary fluid. A pressurizer in-surge occurs to match fluid expansion, system pressure increases, reactor trip i

occurs on high pressurizer pressure, the pressurizer safety i

i valves lift and limit primary pressure. On the steam system side the pressure and temperature increase until the steam generator I

safety valves (SGSVs) lift. With the opening of the SGSVs a heat sink is re-established and heat transfer from the primary to the secondary begins to increase. With the occurrence of reactor Ii trip, the amount of heat transfer required of the steam

, i generators to cool down the primary fluid is much less than at ll i

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full power conditions and the SGSV capacity is adequate to relieve the generated steam. The primary fluid cools and the pressura decreases, the pressurizar safety valves close, and the steam generator safety valve relief rate matches the steam production rate, and secondary pressure begins to decrease, thus ending the overpressurization event.

The results of this study show that there is acequate margin in the Trojan plant system overpressurization protection design to allow pressurizer and steam generator safety valve setpoint tolerances on the current nominal setpoints of +4 percent.

Consideration of high pressurizer reactor trip setpoint and steam system operational pressures support increasing the tolerances in the negative direction to -3 percent. It should be noted that a

the conclusions and recommendations of this study are based on the current nominal safety valve setpoints and a change in the setpoints may invalidate the results of this study.

The remainder of this report describes in more detail the work performed in the evaluation of the margin in overpressurization protection. Section 2 describes the RELAP5/ MOD 2 model and its

, initialization and benchmark. Section 3 describes the analysis of the loss of electric load accident. The results of the cases

, analyzed are presented and discussed and a discussion of negative setpoint tolerances is also included. Section 4 includes conclusions and recommendations based on the results of this i study. Tables and Figures are presented after section 4.0.

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. Plots generated in the benchmark and accident analysis cases are included in appendices.

M TROJAN RELAP5/ MOD 2 MODEL M Model Descrintion The RELAP5 model was based on Trojan plant parameters. It simulated the four reactor coolant system loops by integrating two separate loop models. One loop model (single loop) corresponds to one loop of the primary with a single steam generator and steam line, the other loop modeled the other three loops combining the parameters of those three loops (combined loop). Thus, in model development many parameters were determined for the single loop and simply increased by a factor of three for the combined loop. This technique was used in order to model the entire RCS with a reasonable number of RELAP5 volumes. This technique was also used in the original Trojan FSAR analysis of some events. The entire model, diagrammed in Figure 2.1, included 77 primary system volumes, 3 of which were used in the core region, 18 to model the vessel, and 8 in the tube region of each steam generator (single and combined) . The model included 44 secondary system volumes, wherein 4 volumes were used in the steam generator tube region, a total of 17 volumes were used for each steam generator (single and combined),

and 2 volumes were used to model the steam lines. Twenty heat f

structures modelling the core fuel rods, steam generator tubes 1

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l and pressurizer heaters were included. Other heat structures to model primary and secondary metal were not necessary for the analysis of the loss of electrical load transient. All relevant boundary conditions were modeled, e.g. feedwater source, steam header, and atmospheric conditions (for relief valves) according

to plant data. Models used for the core power, pressurizer, and main steam sa'fety valves are further discussed below.

Core Power i

Reactor core power was simulated using point kinetics to determine reactor power response to the transient. The point kinetics inputs were obtained from information presented in the Trojan FSAR Chapter 15 and correspond to the same beginning of life (BOL) kinetics parameters used in the FSAR analysis of the 1

loss of electric load accident. Following reactor trip the core decay heat output was calculated based on ANS5.1 1979 decay heat after infinite reactor operation.

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Pressurizer and Main Steam Safetv Valves l

The pressurizar and main steam safety valves were conservatively

[ modelled with the assumption that they lifted to 70 percent of capacity at the nominal setpoint plus the tolerance, and then linearly increased capacity to full capacity at 3 percent over the nominal setpoint. The valves were modelled as trip valves in f

the RELAPS input, therefore they could only be modeled as either i

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full open or full closed. To model the assumption above, a full open setpoint was determined that would result in the same integrated flow if the valve had opened to 70 percent and then l

linearly increased to full capacity. This setpoint is greater than the nominal plus tolerance setpoint. For all evaluated tolerances greater than or equal to 3 percent, the safety valves were assumed ' to lift to 100 percent capacity at the tolerance setpoint. No delay time was assumed in the valve lifting.

Figure 2.2 provides an illustration of the setpoint determination. Table 2.1 presents the valve nominal and analysis setpoints that were used in this study.

The flow capacity of a safety valve was modelled by assuming a RELAP5 junction area to model the valve and then determining an appropriate input discharge coefficient so that the desired flow would be calculated for the junction. The discharge coefficients were determined assuming saturated steam at 1320 psia for the SGSVs and 2650 psia for the pressurizar safety valves. Since the flow capacity of the SGSVs was determined at the overpressure limit of the steam system, the valves could not pass more than design flow. The pressurizar safety valves, however, could relieve greater than design flow at pressures higher than 2650 psia. Safety valve relief flow capacities that were used in the

, analysis are shown in Table 2.2.

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M MODEL INITIALIZATION The model was initialized to two different sets of conditions, 102 percent full power design conditions at beginning of life (BOL) , and 100 percent power using Diablo Canyon initial conditions for benchmark. The 102 percent BOL conditions are shown in Table 2.3 with the design values obtained from the Trojan FSAR. The model initialized conditions for the Diablo Canyon benchmark are shown with the Diablo Canyon actual plant conditions in Table 2.4. In each case, very good agreement between desired and modeled conditions was obtained.

M MODEL BENCHMARK TO PLANT DATA The data for the model benchmark was obtained from PGE and was taken during the Diablo Canyon Unit 2 turbine trip from high power. The Diablo Canyon plant data is applicable to the benchmark of the Trojan model because of the basic similarities in plant design. Both plants are Westinghouse 4 loop pressurized water reactors with a rated full power level of 3411 MWt and use the Westinghouse "Model 51" recirculating steam generators. Both plants have similar designs of reactor vessel, pressurizer, and primary piping. Therefore, the primary system volumes are similar. Since both plants have the same power level, steam generator design, and primary volumes, the heat transfer and primary pressure response characteristics should be similar for 7

. - _ -. _ ~ -. .- . -. . _. _ _

k both plants. The benchmark was performed to determine the l capability of the Trojan plant model to calculate the overall heat transfer and primary pressure response in a transient j

! condition. The exact design of the steam system of Diablo Canyon was not obtained for this study and the potential exists

, for differences in secondary side volumes and steam dump system operation bet' ween the Diablo Canyon plant and the Trojan plant.

! This was not significant to the benchmark of the Trojan plant model since the secondary side pressure response was not

benchmarked but input as a boundary condition to the model for i

calculation of primary conditions. Therefore the influence of i

steam system differences was minimized in the benchmark.

The event the Diablo Canyon data represents is a turbine trip from 100 percent full power. Automatic control ' systems i

determined the steam header pressure and feedwater flow for the duration of the event. A reactor trip signal was generated by automatic systems at the time of the turbine trip. To model this event with the Trojan RELAP5/ MOD 2 model, the steam header i pressure and feedwater flow rate data from the test were input to I

the model and the model allowed to calculate all other

parameters. Reactor trip was modeled by assuming that the rods l

l were free to fall at time of turbine trip. Figures showing calculated system parameters along with test data are included in Appendix A.

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f The model results and test data compara very well. Some slight !

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. - - - - - - - - - - - _ . - - . _ - _ - - , - ~ _ - - . . - . - . . . - . - . - . . . . - - -

. differences are apparent. In the benchmark run the RELAP5'model

+ . 7',

slightly over-predicts the loss of heat transfer at the time of ,

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turbine trip. This is apparent from Figure 2.3 whIch jshows;'

calculated steam flow going to' sero immediately after turbiner i trip while test data show a -drastic reduction, but ' neh a

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cessation of ' steam flow. This _is ' probably due to modeling the secondary system pressure ' increase faster than occurred during i

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the test. Test data was available only at three second, int,ervals

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so the actual pressure ~' increase rate was not known. The result t

of this overprediction in the loss of heat transfer can be seen in Figure 2.4 which shows pressurizer pressdre. Calculated pressure increases slightly due to the heat up' caused by loss'of .

1 heat transfer hafore responding to the reactor trip and i

decreasing, while the test. data shows pressure only decreasing'in response to reacter trip. The calculated pressure response after the initial overprediction of loss of heat transfer shows excellent agreement with the data. This indicates a good match

! between calculated and actual heat transfer in the steam I,'

] generators. ,

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From the results of the benchmark to plant data it is concluded that the Trojan RELAP5/ MOD 2 model is capable of representing j actual plant response to transient conditions. >

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M Analysis of Loss of Electric Load Accident M Analysis Assumetions The analysis of the loss of electric load accident was performed with the following assumptions:

o Initial core power, reactor coolant average temperature, and reactor coolant pressure assumes nominal values censistent with Lteady-state full power operation. Mode 2. initia1L conditions are give 'in Table 2.3.

o The transient begins with an instantaneous closure of the turbine stop valves.

o No credit is taken for direct reactor trip on turbine trip. The reactor ' is tripped on a high pressurizer pressure of 2410. psig with a two second delay until rods are free to fall. These are the same assumptions as used in the original Trojan FSAR analysis.

l o Moderator and Doppler coefficients of reactivity used in the cases of the loss of load accident correspond to i BOL conditions used in accident analysis as described in Chapter 15 of the Trojan FSAR. A positive moderator i

\

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. temperature coefficient and a conservatively large

_g (absolute value) negative Doppler power coefficient are used.

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f o No credit for automatic reactor control is taken.

l1 o No credit is taken for operation of the steam dump system or the steam generator PORVs. The steam generator pressure rises to the safety valve setpoint where steam release through safety valves limits secondary steam pressure at the setpoint value. For additional conservatism the highest flow steam safety valve, SGSV number 5, is assumed to be not available.

o No credit is taken for operation of the pressurizer PORVs, or pressurizer spray flow. Pressurizer pressure increases to the safety valve setpoint where steam release through the safety valves limits primary system pressure. One pressurizar safety valve is assumed to be not available. The pressurizer heaters are assumed i

k to operate during the transient.

3 p o Feedwater flow is maintained at its initial value for the duration of the transient.

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Three cases of the loss of electric load were analyzed. All of the cases used the assumptions described above. The cases were differentiated only by safety valve setpoint tolerances. Safety valve setpoint tolerances of 1 percent (FSAR case), 3 percent and 4 percent were investigated.

M Analysis Results Table 3.1 presents the time sequence of events and peak system pressure results for all three cases and for the case reported in the FSAR. Figures 3.1 through 3.5 show the transient response calculated by the RELAPS model for the FSAR case assumptions.

Additional plots of the RELAPS analysis results are included in Appendix B.

L In the FSAR case, i.e., 1 percent tolerances, two steam generator i

safety valves opened to control steam pressure. In the analysis of the 3 and 4 percent tolerance cases only 1 SGSV opens. This is because the lift pressure of the second SGSV has been increased significantly by accounting for the increased setpoint tolerance. The single SGSV is capable of relieving enough steam to control pressure because the steam generator is at a higher pressure and temperature. The higher steam generator temperature results in a smaller temperature difference across the steam generator tubes. The smaller temperature difference translates to a lower heat transfer rate for the higher tolerance cases

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. which in turn results in less steam production. Steam pressure control only requires relieving the amount of steam being produced. Therefore less relief capacity is required for the increased tolerance cases with higher associated temperature and pressure since less steam is produced.

Peak pressures in the primary and steam system occur in the bottom of the reactor vessel and steam generator respectively.

Peak pressure results, as shown in Table 3.1, were determined from the RELAP5 calculations by adding 25 psi to the peak primary system pressure and 20 psi to the peak steam system pressure.

These margins conservatively account for uncertainty in system pressure drops and initial pressure conditions. The plots of system pressures shown in Appendix B do not include these margins.

M Necative Setcoint Tolerances The RELAPS analysis of loss of electric load accident considered the overpressurization that would occur if the safety valve setpoint tolerance increased resulting in a higher opening pressure for the valve. If the safety valve tolerance were in i

the negative direction, the valves would open earlier during an j overpressurization event resulting in lower peak pressures and I

would not challenge the overpressurization limit. In determining the limits of negative safety valve setpoint tolerances, other criteria than overpressure protection were used.

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This study assumed reactor trip on high pressurizer pressure trip. Since it is desirable to trip the reactor before the pressurizar safety valves lift the negative tolerance of the pressurizer safety valve setpoint is limited by the analysis value of high' pressurizar pressure trip setpoint of 2425 psia.

This setpoint is equivalent to a limit of -3 percent tolerance on the nominal 2500 psia setpoint of the pressurizar safety valves.

For the steam generator safety valves the negative tolerance was limited by low power operation steam pressures. Figure 3.6 shows i

the design steam pressure versus power level and the lift pressure of the lowest setpoint SGSV at the negative three percent tolerance value. To provide a margin of about 35 psi between the zero power operating pressure of 1115 psia and the lowest lift pressure of a SGSV, a tolerance of -3 percent was chosen.

4.0 Conclusion /Recommendatign The following conclusions are based on the results of this study:

o The RELAP5/ MOD 2 Trojan model developed in this study

conservatively calculates system pressures resulting I*

from overheating transients.

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o There is sufficient margin in the overpressurization

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. protection design of the Trojan nuclear plant to allow Pressurizer and Steam Generator Safety Valve setpoint tolerances of +4 percent of their current nominal setpoint.

o Negative tolerances of -3 percent on the current Pressurizer and Steam Generator Safety Valve setpoints could be allowed without resulting in pressurizer safety valve lift before reactor trip or spurious openings of the SGSVs.

It is recommended that the Trojan Technical Specifications be amended to provide increased setpoint tolerances. The results of this engineering evaluation support increasing the pressurizer and steam generator safety valve setpoint tolerances to +4 and -3 percent. This action will result in a lower probability of safety valve surveillance test failure and less time spent testing and performing maintenance on safety valves.

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f TABLE 2.1 Ntzninal and Analysis Safetv Valve SetDoints Pressurizer Safety Valves Naninal Tolerance Analysis Setpoint, psia SetDoint, osia 1% 3% 4%

2500 2532.5 2575.0 2600.0 Steam Generator Safety valves (SGSVs)

SGSV Naninal Tolcrancx: Analysis Setpoints, psia

Ntznber Setooint, osia 1% 3% 4%

1 1185 1201.8 1222.0 1233.8 2 1215 1232.2 1252.8 1265.0 3 1225 1242.3 1263.2 1275.4

, 4 1235 1252.5 1273.4 1285.8 5 1245 m m m

SGSV tolerance analysis setpoints include a +1.4 psi adjustant for node elevation change and friction pressure drop.

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TABLE 2.2 Safetv Valve Flow Capacities Used in Analysis Pressurizer Safety valves Flow per Valve = 498,0001hn/hr at 2650. psia.

'two valves are assumed available.

Steam Generator Safety Valves SGSV Flow Capacity of Saturated Number Steam at 1320 osia, Ibn/hr 1 671,083 2 688,083 3 911, 779 4 919,226 5 0.0*

SGSV Ntznber 5 assumed not available.

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TABLE 2.3 RELAPS Model Initial Conditions and Plant Desim Conditions Parameter Desim Value* Model Initial Condition 1

Nminal Operating Pressure 2250 psia 2247. psia Reactor Power 3479.22 W 3479.81 M Reactor Coolant Pump Power 12 W 7.1 m

'Ibtal NSS Power 3491.5 E 3486.9 M Total Coolant Flowrate 132.6 x 10 6 lbn/hr 132.1 x 106 lbn/hr Reactor Vessel Inlet Temp. 552.5 F 555.2 F Outlet Temp. 616.8 F 620.42 F Average Temp. 584.7 F 587.8 F Pressure drop 42.6 psi 29. psi Steam Generator Inlet Temp. 616.8 F 620.3 F Outlet Tenp. 552.3 F 555. F Pressure drop 34.5 psi 32.9 psi Piping Pressure Drop 9.0 psi 7. psi Total Prinary Pressure Drop 86.1 psi 70 psi Steam Pressure 910 psia 909 psia Steam Flow 15.07 x 106 lbn/hr 15.3 x 106 lbn/hr Steam Tenparature 533.3 F 533.1 F 1

l Feedwater Temperature 440. F 440. F

Design values are frm ESAR Table 5.1-1 except RCP power, reactor power, and total NSS power. RCP power is fr m FSAR Table 15.0-1. Design reactor power is 102 percent of 3411 Et and total NSS power is design reactor power plus 12 Wt i

pump power.

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DELE 2.4 Diablo Plant Conditions and REIAPS Model Initial Conditions for Bencherk Diablo Canyon Petrameter Cordition Initial Condition Core Power 3411 Et 3411 mt Pressuri:eer Pressure 2248 psia 2245 psia Hot Leg Tamperature 600 F 601 F Cold Leg Temperature 540 F 535 F Average Temperature 570 F 568 F Steam Header Pressure 743 psia 743 psia f

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TABLE 3.1 Tine Secuence of Events and Peak Pressures Time, sec 1% 34 4%

FSAR Tolerance Tolerance Tolerance Event ggg Case Case Case Ioss of electrical. load 0 0 0 0 Initiation of steam release 9.5 8.0 9.5 10 from steam generator SVs High pressurizer pressure 5.1 5.3 5.3 5.3 reactor trip point reached Rod begin to drop 7.1 7.3 7.3 7.3 l Peak gessurizer pressure occurs 8.0 7.0 8.0 8.2 Peak Systen Pressures, psia Peak ginary gessure, psia

2621.4 2670.5 2698.6 Peak secondary pressure, psia

1261.5 1279.5 1282.5

Peak system pressures not reported in FSMt 6

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m cc Q 'M ru 8 0 w 2 o -

ggu w

b ,- u. w- b ~

w m

~ ~- w .

. ~ - ~ -

% % % w ~ ,~

~ ._r u

- t* h t~ t- - 9 .

d* ~? ~' " :r q

i,M%p T,,*

i e - ~.

e _. . . ,

a e d , I 3 o R I e w s _, ~ ~

-- Et L.

h 51 w

- % w 4 5 ~ N

-- s. 3- "T" T' b _ ._ _ .d. __ _ __

- 4 dE h b w

.c u w w

. . w .

m m ~ -

e . w

~ m ~ m m 1 m f mg *g m~

A m h =m 3 _ _ _ _,_ _ _

8 JE". .e . ' er .er s . e m .

. __ u. . -

L .._

a w o- a I#_ "

. II o s% 6 -- -

_ - f, w u-

tI' w : ~.w :'

o

, l_

l _,

1 o w e ww w ~.us.u._ o u g

, _.g _ w _ g _ o_ u .

2 : -

~ w _

N N b h N

$ l...,~,,,.,

1 Er3-

~ s s ~aT

~

M N

~

s 3 s t N. ~. s- ~i e

I w

o N N 21 s

..r-...__, . . . __ - - . _ . _ _ __,

FIGURE 2.2 Determination of Analysis Value of Safety Valve Setpoints D y f-

--.. C W

llt l v fD 8

t l

$ so E I af I

.

i E I C i e to

3 '

S A l$' D Pl & Ps PN Pressure, psi FN = Nominal Safety Valve Setpoint Pressure PT = Tolerance Safety Valve Setpoint Pressure P3 = Safety Valve Setpoint Used in Analysis PA = 3% Accumulation Pressure Area ABCD = A'B'C D 9

i I

i 22 i

~

1 FIGURE 2.3 l

i Trojan Model Benchmark Steam Flow .

1500 i x Model Single Loop

1400-

+ Test Data Loop 1 1200- O Test Data Loop 2 I 5 % A Test Data Loop 3 j g 1000 -

m

i --

X Test Data loop 4

n

~

j 1 -

800-El -

I 600 -

! E -

I 5 400 -

i s

m a x

+

+ 4

! 200- 5 8 3 g i -

^

V

M y + x .x

g +

0 - ' '

110 130 Time, sec. 150 '

l

O FIGURE 2.4

Tmjan Model Benchmark Pressurizer Pressure i

4 2400 RELAPS Model

+ Test Data i

N

" +

a c

2200 -

3 4 t +

, a l L n +

ro 0 2000 - +

g + + + +

n.

I 1800 , , , ,

110 130 150 Time, sec.

l

FIGURE 3.1 Loss of Electric Load Accident 1% Safety Valve Setpoint Tolerances (FSAR Case)

Total Reactor Power 7" I s i a a i S*

x .

o -

Transient occurs at 30 sec. -

k << -

.c

$ o.

' ~

5 p ea

~

N a

g i i I I I I I ca. 10. 20 30. 40 '.,0 60 70 80 Time, sec.

l l

FIGURE 3.2

Loss of Electric Load Accident 1% Safety Valve Setpoint Tolerances (FSAR Case)

Pressurizer Pressure i i i i i i i l 7ore. i M

w t Transient occurs at 30 sec.

eb

! E b w 3 -

l $ "

E m n.

N m -

c 3

e4 E

o. ,. _

ad N

i l i i E I I rb. 10 20 30 40 %. 60 70 90.

Time, sec.

FIGURE 3.3

Loss of Electric Load Accident 1% Safety Valve Setpoint Tolerances (FSAR Case)

Steam Temperature sw i i i i i i i

- Single SG g

Transient occurs at 30 sec. -

s-e Combined SG u.

h

~

B 2

N E

- a e

o 4/1 0 -

S

- w -

I I i I i i i ul). 10. 20 30. 40. s0 60 70- 60 Time, sec.

Q FIGURE 3.4 Loss of Electric Load Accident 1% Safety Valve Setpoint Tolerances (FSAR Case)

Pressurizer Liquid Level e

,; i i i i i i i Transient occurs at 30 sec.

9 -

o

M La.

> 9 .

$ C u

3 9 -

s. n N

L.

3 9 .

E o.

n 9 .

8 y i I I e i I i

O . 10. 20, 30 40. 'A 60 70. 60 -

Time, sec.

FIGURE 3.5 Loss of Electric Load Accident 1% Safety Valve Setpoint Tolerances (FSAR Case)

Primary Loop Temperatures g i i i i i i i w

- Combined Loop Hot Leg Transient occurs at 30 sec.

g _

o-e Single Loop Hot Leg C

- Combined Loop Cold Leg g: : : : : : : : : e -

x-* Single Loop Cold Leg 4 0 5

e .--. Loop Average tu

gg w

m : : : : : - ~

$. : : : : I -

.n f 1 I I I . _. .

- I I

Eb 10. 20 30 40 50- 60 70 80 Time, sec.

e

FIGURE 3.6 STEAM PRESSURE VERSUS LOAD 1300.

1200. 1185 psia Low Set Safety Valve

_________ _ _ _ _ _ _ _ _ _ _ _ _ _ _1149.45 psia

-3% Tolerance 1100.

.2 E

a E

$ 1000.

E SG Operating b Pressure m (FSAR Figure 10.2-1) 900.

800 0 20 40 60 80 100 i

LOAD, percent of full load i

I i

l l

\l.

30

< a Aenendix A Appendix A contains plots generated in the benchmark of the Trojan REIAP5/ MOD 2 model to the Diablo Canyon turbine trip test data. The benchmark was described in section 2 of this report.

The following figures are included:

Fiaura Ntmher Descrietion i Al Steam flow rate A2 Pressurizar pressure A3 Pressurizar level A4 Loop delta temperature A5 Loop average temperature A6 Hot leg temperature A7 Cold leg temperature

A8 Reactor and neutron power i

i!

l1 I

! 6 l

A-1

lJ' l

, - l+a D

Xt s

e

- T l

g-K -

R g -

0 A

T A +W 5 1

A D

M T S

H ! E T

C N

E !

[ A BWO C L l

E A S LF E R + 5 ,

EMEA U G

I E

MA DTS F

1o TTA I

0 D O +5 3

1 T S

M E T

+h 7 N o i.

M A x4 R

1 E

J X ,

O '

R

~

x

a

~

5 A T

A D

! 0 1 T 1 S 6 5 4 3 2 E

.- 5 4 3 2 '

1 9 B. 7 1 T

1 1 1 1 ' 0 O 0 O 0 O O 0 O

+

- 7V@m8Ev o$N23 :

3 O Ca. *

' ! l 1 ! l

1

~~ '

l ROJ AN h/ O J EL 3 E N C -

VAR <

PRESSURIZER PRESSURE 2.4 -

FIGURE A2 2.3 -

! +

2.2 -

mn +

a. m o

4 , eg +

i w

E se 2.1 - .,.

l a2 mo +

l mr i hlI-m +

a. +

2- + * + +

1

+ +

1 i

- 1.9 -

1.8 - -

, y-- --

110 130 150 TIME, SEC RELAPS + TEST DATA i

L TROJAN V ODEL BENC H V AR <

PRESSURIZER LEVEL i 100-FIGURE A3 90 -

80 -

70 -

z w 60 -

O y + + + ,

P a 50 -

J

- + -

40 - +

J 3 n -- + + + + + + +

20 -

10 -

O i i i i i 110 130 150 TIME. SEC RELAPS + TEST DATA

w ' ,"

i l

TROJAN MODEL BENCHMARK LOOP DELTA TEMPERATURE

, 70

FIGURE A4 60 - + + + +

l i

I

! 50 -

! g +

l Lai a: 40 -

l P m

1 a 30 -

l j

3__

+

20 -

l t

+

10 - +

+

4

+

A + - -

)

l O i i i i 110 130 150 j 11ME, SEC RELAP + TEST DATA '

l

~~ R O J A sl V ODE _ B EN CF V AR <

AVERAGE TEMPERATURE, F 580- -

FIGURE AS 570 - + +

+

h_

ui 560 -

p% +

x .

E g +

w tt +

2 55G -

N +

+

+ + + + +

540 -

530 r- , -, i 110 130 150 TIME, SEC RELAP + TEST DATA

D TROJAN V O D E _. 3 EN C -

VARK - -

HOTLEG TEMPERATURE 610 .

FIGURE A6~

i 600 - + +

'\

i 590 - s l

l 580 - yn c ele I S, }c L-of Ilot bj w + Te ge<'d re,F i nzi 570 - , ,

,\+~ ~'[

~

T- ,

560 - yn,de] c.mbi& - ..

~ * ^

~

n.

550 - L**/"e'~d,j Haf b f

+'

A~ 's"'- '? '

j

~

1

+

Te~ , .

+

x '

+ 'C'

'x 540 - I-

. s _x .=

\ ,

s 530 - .

1

- N- _' '1 "

~

l 3 '

520 - ,

7 ,_

t~

s 510.j i .

1 l 3 , , ,

~' '

-110 130 150 '_A 1 '

. TIME, SEC RELAPS + TEST DATA -

r m

" ^

l .

. s .. -

< %4 &

! TROJAN M ODEL BENCl- V ARK

! COLD LEG TEMPERATURE

! 550 549 -

l FIGURE A7 548 - '

~

^

547 - _+

546 -

+

j 545 -

i 544 -

i E 543 - +

~~

542 -.

ui _

+

l $ 541 - + 1

- + + + + +

$ 540 + + ,

i W 539 -

$ 538 - / M *N "gk l *f M 537 -

( 536 - p,,d e } c.mb,;,a A

~

l 535 - -

Looy ~~

534 -

l 533 -

l 532 -

I 531 -

l 530 ~

i

~

i, - i i

! 110 130 150 2'

l TIME, SEC -

RELAPS + TEST DATA

~,'

4

^

"w, a .

3

~

4

! TROJAN V O DEL 3EN CF V AR'<

i REACTOR / NUCLEAR POWER 110 -

FIGURE A8 l 100 _ + + -_+ -

D. ..

-' e r.

, go _ 1 l g 80 -

t h.

i o 70 -

t t-Z ia 60 - ~

? O

= x ,

y 50 - .,

E '

a 40 -

1 5 I O l 1 30 -

l 20 -

10 -

+

+ + + + A- *-

0 I I '

I= 3 110 130 150 TIME, SEC RELAPS + TEST DATA

o

Annandix B Appendix B contains plots generated in the analysis of the loss of electric load accident with 1, 3, and 4 percent tolerance on the nominal pressurizer and steam generator safety valve setpoints. The analysis of the loss of electric load accident was described in section 3 of this report. The following plots are included for each case:

a. Reactor power

, b. Pressurizar pressure

c. Hot leg pressure

d. Lower vessel pressure j

e. Pressurizer safety valve flow rate l

f. Pressurizar liquid level

g. Surge line flow rate

h. Primary loop temperatures

1. Core to primary fluid and primary to secondary fluid heat transfer rate

j. Steam generator outlet pressure t ,

k. Steam generatcr outlet temperature l

1 3

1. Feedwatar and steam flow rates i m. Steam generator safety valve flow rate I

i i

. B-1 li

.i l . !,: f ! 3 r7 o

HOeg Ea s UOs. n.35* $*dmym .

0o 9 9 r

mfr , o

- x ,o? "- .

0

- ~ - - - RF3 _

ES2 AA-CR1 -_ _

T 8 OC5 RA9 _

S7 81 PE0 6 0 I I O. 9

/ W -

0 EI 0 8 R 0

/ . P 0 ET 7

t RR TCO CfJ 2 ANA 0 i WT N i A I 1 S I 0A St F l E RT AY 3 N F 0 e CV I e

EA G L U V R

~

E T E I

T M O B E L l 4 F .

I 0 a i

R a S A E N C

C E

) S 5

0 e 6

0 a 7

0 e i

8 ~ -

0 _ .

~

1 a .

_ - ~ ~ ~

0 8

i i 0

7 i a 0 6

e e G

's 1

S E C C E

b. N . S A s i 0 I l R 4 B E E L 1 E O t i

R T U T G E I

V L .

F AE s e 0 VC 3 N

Y4 TR Ef Fl R0 S1 NTE i a 0

. RNR 2 JIU .

OCS 7 RRS 0 TEE /

PR 8 0 P 0 0I /

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- 1 S 1 RS AE

_ 2SR - - - - -

3FP .

Tc4

, 4 c

4

< 7n Nit 7c4 o'Dc 7 O_xw N2.Nh. J_ ~ oED)u uMu_ JrDJ_

t h @

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0 e2 1

d 0L n e Pi Pl g

_b

_n

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i ' 0 5

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i mmuo t t i e t .

I '0GF '007 '00E *0PZ '091 06 3 seS/mq[ '9498 MO[j GA LVA ASBJES JOZ pnSS8Jd [9401 l-,

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3FS ~ ~ .

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0 9r 4

$-. 9- 9 9o 9e 3 c 7 7 O_x * :

OOOO OOv7 t* mJ- EM 3OJ- mm 2O NUZJ7 Tm

FIGURE Bl.h 32-1159709-00 TROJAN SAFETY VALVE TOLERANCES F SAR CASE. I PERCENT 10t FRANCE

. REACTOR COOLANT Til0T. TCOLO. AND TAVE g i i i i i i i u,

_ TE WF 100010000 Combined toop ilot Leg Te 8

o_.e TE WF 200010000 Single Loop Hot Leg Temp s TE WF 125010000 Conbined Loop Cold Leg 2

[ NC 0 0 0 0 0 0 0 00 _

w E TWF 225010000 Single Loop Cold Leg Tem m -

E sCNTRLVAR 2

? Loop Average Temp hJ 0 0 0 0 : .

5 -

$ o = = = = = = = _

.h n

6 i i i n 1 i i hh - 10. 20 30. 40 !,0 . 60 70 80.

86/08/07.

TIME (SEC)

i l

e FIGURE Bl.i 4

32-1159709-00 TROJAN SRFETY VALVE TOLERANCES l

fSRR CASE. I PERCENT 101 FRANCE TOTAL CORE AND SG HEAT TRANSFER Te i i i S*

i i i i x

_ CNTRLVA8t 301 Core To Primary o -

e_.eCNTREVAR 3 Primary To Secondary ci m o

< ' e

~ vn

o s l l 3 co , -

I

. C's e

4a L

e so -

m L

l F-

&e ao

~

I i e i I I I I 00 10. 20 30 40. LO. 60 70 80.

86/08/07.

Tit 1E ISEC)

9 FIGURE Bl.j '

32-1159709-00 TROJAN SAFETY VALVE TOLERANCES f SAR CASE. I PERCEtJT TOLERANCE 7 ,," SG OUTLET PRESStEE i i i e i i 2-x

_ P 660010000 Single SG S -

3._e? 760010000 2

Combined SG 8 -

a N U

N ts m

J . '2 LLI E

D U) k *

[ -

m E

O

__a k

O O O O O O O

, I I i 1 1 I I

i. 10. 20 30 40. 50. 60 70 80.

86/08/07.

TIME (SEC1

__~ - . . .. .

FIGURE Bl.k 32-1159709-00 TROJAN SAFETY VALVE TOLERANCES f SAR CASE. I PERCENT TOLFRWE

. STEAt1 GENERATOR OUit ET STE@ TEt1P e i i i i i i i

_ TEt1PG 660010000 Single SG g _

u TEt1PG 760010000 Of Combined SG

-g -

u. w a -

w ro e 3 .

$2 11J a

t gd o -

a G

w r -

3a o o

> = = = = = = = -

g i e a e i e a

'h.

w 10. 20 30 40. 50. 60 70- 60 86/08/07.

TIME ISEC1

- - .- =

FIGURE Bl.1 32-1159709-00 TROJAN SAFET1r VALVE TOLERANCES FSAR CASE. I PERCENT T0t ERANCE e- -

SG FELOWRTER AND STERN FLOWRATES

%%e o a - .lo -

i i i i i i i

5 "

05 ____ WLOWJ 602000000 8 . : : - - : : -

Single SG FM Flow Rate 3o 5 /,

2

_e TLONJ 660010000 Single SG Steam Flow Rate E E 0 m ~ NLOWJ 702000000 Combined SG FW Flow Rate b7 - b<? -

_ w_,MLOWJ 760010000 m 0" m

0- Conbined SG Steam Flow Rat; -

a3 ca - _

No G

-U w 9 tr a:

2 x o o

-.J _s AA- 41.

~ -

m 't m e -

h~ q O '

W.--

w Z 2 0 _0 _

% -= = = = c p 9

N N,

b =x r =-q a u 1 3 -c=ccco n o -

o 8 e i n e i i ..

o 00 10. 20. 30 40 50 60- 70 80 86/08/07. -

TINE ISEC)

FIGURE Bl.m 32-1159709-00 TROJAN SAFETY VALVE TOL ERANCES

, FSA8t CASE. I PERCENT TOLfRA*JCE

. INDIVIOUAL NSSV FLOW RATES 3 g- g i i e i i i o

o m

_.NFLOWJ 666000000 o Single SG SGSV #1 u m Eg _ e g _

g NFLOWJ 667000000 o, n N Single SG SGSV #2 Eo -

c

" " % NFLOWJ 766000000 Combined SG SGSV #1 Mg m a

-Mgm "

w NFLOWJ 767000000 m Conbined SG SGSV #2

~

m m L O s _s ^- ^

^

-7 g m -

2 2 O O 2 2 k . k .

$k W

$h G

E E z 2 O O Yh o - ~?o of ~

2 Z .

3 3 7 7 O ': 1 : ':::: c': ' l x :: = 0 le-e_ , w o 00 10 20 30 40. '.,0 60 70 80 06/08/07.

Tir1E ISEC1 L

. s .

e.

6 e i i i i 5

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g m

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, 5 -

o

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8 E w -w -

= -

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l i e t

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'l 1

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91-9 1

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e 9

VOLUME PRESSURE (LSF/IN21 V410010000

( x!c*3 1 g.0 2.1 :. 23 : .4 2 . <, 2s m R n, E9 w I M

-a m-m - - N'S an, b shb B GE

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ap ,

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d C- E 5 P, R

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8 - -

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y. e

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FIGURE B2.c l

32-1159709-00 TROJAN SRIETY VHLVE 10t ERONCES TOLERANCE CASE . 3 PERff NT Tol[R4NCF

- 110i t EG FRESStEE

" '* i i i i i 9a x

i

_P 100010000 Conibined Loop o -

._ e 200010000 i

e Single Loop y -

- e4 l

m , _

J .

m r4 i tan j M

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1 $2 3

~

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<b. 10 M. 30 40. 50 50 70 SO j

86/08/07.

TIME ISEC) i

1 o

46 l

1 l

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8 S

m U w

m $ .

W g -

cJ $

= s W

-w g :

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r 15 - -

g 0$

w-m

M ufW gyB - -

R dm8 n'

%.~

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(

i I

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g f

e $ U

& w d

=

w w

j -

g-2 g -

8

-w w

' W&

d5 - -

f "s3

--a C-g5W ud gt>

s emC s 5." e 8.5 E ve - - de*

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m _d e, , , i . .

cse car c:t ce csi es i

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_ . _ _ _ _ _ . . _ _ .. . _ _ . _ _ . . . _ _ -_ . _ . . _.- .. .- ~ ._ . . .

F e

t 3

i 4

l l .

t FIGURE B2.f 1

' 32-1159709-00 TROJAN SAFETY VALVE TOLFRANCES TOLERANCE CRSE. 3 I'fRCENT TOLfRFNCE t

PRESSURIZER I LIQUID LEVEL

, s i I i i I j e 9 .

i 14.

I

! ~

9

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l 2

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e Q. ,

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i i

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! "0

- 10. 20. 30- 40. '.,0 . 60 M. 80

{ 86/08/07.

{

' Tit 1E I SEC 1 l

l .. .

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g u.

en w U N

= g _

-- d w w -

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h.

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l I

i B-21

. 4 FIGURE B2.h 32-1159709-00 TROJAN SAFETY VALVE T'" EEONCES TOLERANCE CASE. 3 PERCENT T0tERANCE

. REACTOR COOLANT THOT. TCOLO. 4NO TAVE g i i e i i i i

___TEWF 100010000 Combined Loop. Hot Leg Temp o -

o_e TEWF 200010000 0 Single Loop Hot Leg Temp

.g TE WF 125010000 Corrbined Loop Cold Leg Tem

= = = = = = =

- de*

m TL WF 2250\0000 m -

Single Loop Cold Leg Temp N y sCNTRLVAR 2 3 Loop Average Temperature E5 5 0 0 0 0 : : -

y' * <

s' M=4?4;g 2o -

g$ = = = = = = = _

3 o -

.o I I I I I I l 20 10. 20 30. 40 w. 60 70 s30 86/08/07.

TINE ISECI

t 4

FIGURE B2.i 32-1159709-00 TROJAN SAFETY VALVE TOLERANCES TOLERANCE CASE. 3 PERCf NT TOL FRANCE TOTAL CORE AND SG HEAT TRANSf ER Te i i i i i i i o_ ,

U ___ CNTRLVA8t 301 g, Core to primary 8tu/sec R o - _

w CNTRLVAR 3 3

Primary to secondary

" Btu /sec O O O O O O O g g4 - I -

" A I 5- f e : ,

E E , - -

h th m

E

- O - O G-6-4.-6 .

I i I t i I t I i g

ck). 10. 20 30 . 40. M1. 60. m. 80.

,! 86/08/07.

TINE ISEC)

t FIGURE B2.j 32-1IS9709-00 TROJAN SAFEiY VALVE TOL ERANCES TOLERANCE CASE . 3 PIRCINT TOL FR4KF 7 ,,,SG 7 0011ET PRESSURE i i i i i i i S-x

_ P G60010000 Single SG

,e _

P 760010000

_ g Combined SG g .

03 N 8

2 to -

A N gg te -

t.aJ W

D (n

~

ta ~

w -

(A.

LAJ I'

7 J

k

: : : : b

, I I I I f f i 3 10 20 30 40 W. %. 70 80 86/08/07.

Tit 1E (SEC)

e f

FIGURE B2.k 32-1159709-00 TROJAN SAFETY VALVE TOLERANCES TOLERFWCE CASE , 3 PERCfNT TOLER4NCF

' . STEAt1 GENERATOR OUit El 511 Art it F e

.o i i i i i i i

__ TEFG 660010000 Single SG 6 -

o__e iE W G 760010000 01 Combined SG a -

u. 51 co D

rr s ic y W

r N

8s a -

E

> \

3 #

= >

= = = = = = .

I I I t i t I Lil. 10. 20 30. 40 50 60 70 60 116/08/07.

TINE ISEC1

-- , o ~m - 7 '

., u h ('6k [ #

. .. g u. g_ .g 4 C ]* Re ', 2 , g .3o 8 >

r ,,

2 - fA e-e ,,

8 m 8m .

8 8 8 8 ~ ,g 8 7 ,

,.i.,e-e ,- 1 J'

8

-e5 !3s 7,, .sB - B s "f

,tgtgtjhe o '

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'; n' a ,i JL d) .'

,l' n O I

~

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4 ni O

, n' O j' _

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~ ~

n $

d 4

m -

o ..

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)

){- . ve

=

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det "

..~TA "s=

a o .

O-I se o o

$dt=G gg - o o -

g

, w .

O < t q) N

+ p-

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s .r

Ave - ,

o*

e -

Rug m o .

l Eeg m o Ms , , , , , .

ci si :i c- 5 t' o3

, i t 0lXI i

altos es aL6uts (31S/911 hb8 M0ld SSbW N011]Nor

( , i , , , , ;

st ct r ci vi s oo It .oix) i ateos es pautqwo] 'i J3S/911 hb3 M0ld SSbW N0113Nnr i

L B-26 1

L_____._._._.

t t

( .

FIGURE B2.m 32-1159709-00 TROJAN SAFETY VALVE TotERANCES TOLERANCE CHSE. 3 PERCENT 10tFRANCE

$* . . INDIVIDUAL NSSV FLOW RATES s ,a a i i i i i , i mm m n o

m

$ _ tifLOWJ 666000000 m S Single SG SGSV #1 e . .

5g ,* g -

o_.NFLOWJ 667000000 f" F" Single SG SGSV #2 O G '

s NFLOWJ 766000000 Cont)ined SG SGSV #1 Sg

-Mg* "

_ m NFLOWJ 767000000 l

m Combined SG SGSV #2 m m m L y

J _.J Q; -

_ _ .A I

g3 m

-gge

E ~K O O J .J t1- .

LL. .

<r w E E <

Z Z O O

a -;8 - -

t o

2 r o Z .

D '3 7 7

= =8 : : 6 0 'O : = = : : : = c'e-e-e-e-ele-e-e-e -ele--o-e-e-e-6 cs0 10. 20. 30

40. 50. 50 70 90 86/08/07.

TIME ISECl "O

s u1 .

\

i s

6 6 6 6 6 8

-i l

g g

b $

w l U V / W 7 (n

, ;E - -

f-y

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w W

cz:

3 w&

g o: .

2

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~ ~

fE c,,

9 e

E C t

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o n ce e

.e ocw e s

s"8 a w a.

de r- w Cn f us c5

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e w

m-g i -

t i i sr 0r :c rz sr s- c5 f t.Olx1 sqqvmeSaw *aamod acqovag I

t l

l l

B-28 I

l

l l i 1i -l. 'f;j. i - , I ,IJ , ~4

)oiw ut _

>ogarttJ QMttmOML nn 3

A J .Jn.iN.ZN._

.Et .

, >n Ooeaogo w x_O "-

ro g- <

at e r.m a

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a ag e r a9 h PT3

- ~ - _.

RO2 .

EL -

SE1 SR8 UA5 RN9 .

I C7 .

ZE0 .

81 0 i E 9 e

6 RC -

/ R0 0 PS0 8 RE

/ E. T 0 S R 8 SAO U J 2 RPA 0 a i EfN R

CS EA NF TE T

TY O

3 LV i 0 e i EA I' RL F

G NEV C

U T R i

ET E t O L

1 E E B 4 R 3 i

I 0 a A

N b S C E E C S 1

W e a 6

0 e i 7 a 0 a 8 - ~ - - -

0

. ( (

4

(

! FIGURE B3.c l

4 32-1159709-00 TROJAN SAFETY VALVE 10tERANCES TOLERANCE CASE. A l'fRCINT TOLERANCE

- H0i tEG PRESSURE

" r1 i i i i i i i Sa 5 __P 100010000 Combined Loop o - _ P 200010000 4 Yngle Loop y - _

- c4 e N w 2 o -

N tt.

~

_2 W ~

- et (11 i

D en

[j = -

e A

a (tJ

'O O O O O O O O O eg

~

g i l I i I I I

d. 10 20 30. 40. 50- 60 70. 80.

06/08/08.

Tit 1E ISEC1 1

4t ',. 4: ,lj ; .

3 c

cw3n J

1

>o.3rJ .OeCn3Et) 4 e t

-3.t i N ZN en

>n.nOmooao

.x9"-

e 4er m, c9 a

c 4g n9 d -

b. g -

- - - - - PT3 RO2 EL - -

SE1 SR1 -

UA5 RN9 EC7 81 E0 6 0 i

i A 9 -_

/ TC - _

0 A0 8 8S0

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8 0AO t

1 J 2

0 PA I

s 0EN FR CS VEA ENF STE S T ET Y l O F 3 LV I 0 I s

EA G RL U AV R T

NE E C

i ET B t

1 O L 3 E E 4

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S N E C C E 1 S 5

0 i i

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0 - -

t .

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Il ,i 1 I; e

9 nbcNm h'a gi:8"8 8 5oh 3C<Cp< ?

hm ,m

P95R n8m

AnE - g- m c . gm85 2 mnA- <p A , -r e 0 9 - - - ~ - - -

s 4

/

m b

l e .

t 0 i a 0 4

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_. 8 l

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u 0 6

a

_ s 1 s

e r

P l

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- - - ~ - - -

g- a. 8- g. 8 8 fa g.

=mS=5* .

H3m]A. -

i,; l,- i! i

< Il 2 1 :1 I wwe02

=

.DLomm s eL 3a:s

- ae>a . k+2 j s D9 89 c

~ *e Ro m,",

4

- ' ~ PT3 RO2 EL -

SE1 _

SRI UAS RN9 - _

I C7 -

81 0 ZE0 6 i

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0 A0 -

8 LS0 -

/ 1 E -

0 0 T 0 0 R 1

AO 2 0 J 0 e PA i LEN ER VCS EEA LNF TE T

TY 3 O 0 a LV F EA I RL G AV U T NE R C

I ET E M 0 E t B

4 E 3 I 0 e s R S A f E N C

C E 1

S 5

0 a i

6 0 e e

7 0 a a

8 _

0 ~ ~

1 .

( x.

,\ ( " -

. t

~ - - ~ . m i ,

m i ,

0 6

i W

1 S

g E C C E 3

N S A i I

B R e 0 4

E E L E R 0 N U 1 E I G C T I EN F VA LR ,

AE e i i

0 VL 3 O

YT T

ET FN AE SC R .

NE i n 0 AP 2 J .

OA 8 R 0

_ T .W EO /

0SL 8 0AF 0

/

C 6 9 E i f

10 8 0EN 7CI 9Nl SA I RE I EG

_ LR 2OU ~ -

3TS ~ -

0 9 c. 9- - n oa $, D7 7o X -

O()OO OQOv7 - *m L" acc XU ',r~

]

l<2 ZO Fu5

~

=m

8

(

FIGURE B3.h -

4 e

32-1159709-00 TROJAN SAFETY VALVE TOLERANCES TOLERANCE CASE. A PERCENT TOLERANCE

. REACTOR COOL ANT THOT. TCOLO. AND TAVE g i i i i i i :

_ TENPF 100010000 Conbined Loop Hot Leg g -

e_ e TEF1PF 200010000

= Single Loop Hot Leg s TENPF 125010000 Combined Loop Cold L.eg

- d: : : : : : : : - _ x__x TENPF 225010000 m ba Single Loop Cold Leg E3 ty s CNTRLVAR 2

'6 Loop Average Temperature Ir tr:

us m

$ ~? O O

^

0 h}. '

ee$ -

= = = = = = = _

e f, _

g i i e i a a e

80. 10. 20. 30. 40. ',0 60 70 80 06/08/08.

TINE ISEC1

_ . _ _ -. ~.

'i 4

(

FIGURE B3.1 32-1IS9709-00 TROJAN SAFETY VALVE TOLERANCES TOLERANCE CASE. A PERCfNT TOLERANCE TOTAL CORE AND SG IEHi TRANSFER Te i i i i i i i S~

U _ CNTRLVAR 301 Core To Primary o - _

o_.eCNTRt. VAR 3 4 Primary To Secondary g4 ..

\ .

= s a m W 7 ~

ta L.

y - .

2 -

D--

+J W

x - .

9 0 ; e : e -

1 I I I I I I o0. 10. 20 30. 40. 50 60. 70 80.

86/08/08.

TIME ISEC1

D

, L 4

{

FIGURE B3.j 32-1159709-00 TROJAN SAFETY VALVE TOLERANCES TOLERANCE CRSE. A PERCENT TOLERANCE SG OUllEi PRESSURE 7 ,e i i i i i i S-5 _.P 660010000 Single SG y - _

o_eP 760010000 Combined SG

~ ~

i * .

w z ti.

CD 3

14J l

Q' D

tn x -k i

b L1J E

'3 -

J O O O O O O O i

I I i i I l t

~

i 90 10 20. 30. 40. 50 60. 70 80.

86/08/08.

TINE ISEC)

J

Iil l '!I' ' ', j mSm u

>o3Eu >E J n *uEn1uaE>anu

' J J .u -

[g h os *E. 86 mf .e.

- - ~ - - S1 3 j T02 EL -

c RE1 -

NR1 A5 -

GN9

EC7 01 0 NE0 i i E 9 6 RC -

/ AA0 0 :

8 TS0

/

OE 0

R. T 8 e R OAO U J 2 l PA 0 i c i l

ER EN T CS EA

SNF TTE E T RTY

NO 3

0 LV I i TEA F ERL t

I 1 AV G PNE U T C R I ET E M O L

E E B 4 3

( 0 e i R

A S N k E C C E l

S 5

0 i i 6

0 e i 7

0 n i 8 - _ ~ _ _

0 o

C_

oe iS_

mT bE gE nT iN lN nP eP G eG d7 S6 6 G6 S0 0 G0 1 0

1 0 0 0 0 0 0 0 0 .

" T \

(

{

s f

FIGURE B3.1 32-IIS9709-00 TROJAN SAFETY VALVE TOLERANCES e TOLERANCE CASE. A l'ERCENT TOLERANCE SG FEE 0 WATER RNO STERN FLOWRATES g%=.

4 g.%= i i i i i i i

- o-S5 *5 _ NFLOWJ 602000000 o M : : - _ _ : : - -

Single SG FW Flow E e _ _ NFLOWJ 660010000 j ;

_ }2 $i891e SG Steam Flow 8 G % NFLOWJ 702000000 Combined SG FW Flow M, - d cf - _ x_xNFLOWJ 760010000 Combined SG Steam Flow m R ** Q-g uj c : ; ; ; ; c ;

We E 4

-UE 9 cc cc 2 3 O O

_J _.J LA. La.

cr y9 cr r r 2 z O O

~ . ~ _

C*

z 2

'? " x x x x x x = "w -x = x x 3 3 N o-cO-0-0OO-oq eee_c -

a a g i i t i i i i

~

ci a0. 10 20 30 40. *0 s 60 70. 80.

86/08/08.

TIME (SEC)

J b- .1 s 1 2 0 0 0 # 0 #

00 #2 0V 0

1 0 0 V

' 0#

0 0 0S 0S 0

0V 0V 0G 0G 0S 0S 0S 0S 6G 7G 6 7 6S 6

6S6 6G 6G 7S 7S J G J G S J d J d N

O S W WeWe LeL Oe l O n O n L i L i Wlg W g Wb Wb n n :n n i i o o

_S ._ S o :

C wC .

0

8 n-

=

i

=%-

1

=7 0

=

i

= .

'. 60

=

= -

i

' 0

=5

m. =

3 = )

B S E C C

= E E N L R

U A i

.0. S I R = 4 G E I L E F O = M TE =

I C T EN =

VA LR =

AF i = 0 VL 3 O =

YT T S ETE FNT =

AEA SCR R .

NEW i s 0 APO 2 J L .

O4F 8 R = 0 T .V /

ES 8 0SS 0

_ 0AM = /

C 9 L i

0 6

_. 0EA 7CU =

1 8 9ND 5AI 1 RV =

_ 1 EI LD

_ 2ON -

3TI _ - . .

.g gu .0 E 5$ ,$ . o 5

.

S * ,p M uM53_ - NEE 2OJ_L .A

$wE 2O~&MR

.g g r- $ g, ,k 0- d uE53_ - NGE 2OJL $GEzOHMR e%y S mec gO .

.L

?y0

_

v t e Letter Sequence Other
TAC:64803, (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers Results Approval Other: ML20211H333, ML20211H402
MONTHYEAR1987-02-11 [Table View]

ML20211H402

Text

SUMMARY

= = = = = = =

8 n-

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