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.,-70 60 >, .µ *r-> *r-.µ u '° Q) 50 0::: E '° u V') 40 30. 20 10 0 0 . _.,* -" .. *. ,, "B" Curve 0.4 0.6 0.8 . Fraction of Core Fully Controlled Figure, 1: Scram Reactivity Curves 2. 'f.'* .. . ***.* .... . *' , c .' EOCCurve 1.0 | .,-70 60 >, .µ *r-> *r-.µ u '° Q) 50 0::: E '° u V') 40 30. 20 10 0 0 . _.,* -" .. *. ,, "B" Curve 0.4 0.6 0.8 . Fraction of Core Fully Controlled Figure, 1: Scram Reactivity Curves 2. 'f.'* .. . ***.* .... . *' , c .' EOCCurve 1.0 | ||
. \ '* .i 2.0 PVRPOSE AND SCOPE This report is intended to provide the nee es sa ry justifications for adoption and implem.entation of the proposed plant and analysis modifications. While this report is considered a specific submittal on these modifications, the contents relate to, and are based on, the nuclear and thermal-hydraulic parameters derived for mixed 7x7 and 8x8 fuel bundle cores to be loaded into the Dresden 3 and Quad Cities 1 reactor systems at the next rrntage Because the history of scram reactivity insertion function changes has been reported in great detail in previous submittals ( 1), a *bac;kgrouna aiscuss*ion 'is not req\iire*a in this report. The earlier reports a re, however, included by reference. The proposed plant modifications are: Replacement of one electromatic relief valve with a combina:.. tion safety/relief valve, and Raise the setpoints on the eight spring safety valves** The proposed analytical modifications are: Use of more realistic control rod scram times, and, similarly, ---.. -----**----------*----** --* ' ---------------------Reference 1 presented a review of the abnormal operational transients to determine those which might be significantly affected by the scram reactivity function changes and thus require re-analysis. Those transients determined to require re-analysis were the steam flow disturbances, feedwater system transients, and recirculation system transients. These same transients required re-analysis due to the new changes and the results are presented herein .. The transients were re-analyzed for both the "B" and the EOC scram reactivity curves. Also included in this document for completeness are the analyses of the pump trip transients. The, intent of this report is to provide the description and analytical results using the .above modifications in conjunction with the particular scram reactivity function curve for the affected failure*-caused transients. This information represents the. technical bases supporting plant operation for the next refueling cycle. 3 | . \ '* .i 2.0 PVRPOSE AND SCOPE This report is intended to provide the nee es sa ry justifications for adoption and implem.entation of the proposed plant and analysis modifications. While this report is considered a specific submittal on these modifications, the contents relate to, and are based on, the nuclear and thermal-hydraulic parameters derived for mixed 7x7 and 8x8 fuel bundle cores to be loaded into the Dresden 3 and Quad Cities 1 reactor systems at the next rrntage Because the history of scram reactivity insertion function changes has been reported in great detail in previous submittals ( 1), a *bac;kgrouna aiscuss*ion 'is not req\iire*a in this report. The earlier reports a re, however, included by reference. The proposed plant modifications are: Replacement of one electromatic relief valve with a combina:.. tion safety/relief valve, and Raise the setpoints on the eight spring safety valves** The proposed analytical modifications are: Use of more realistic control rod scram times, and, similarly, ---.. -----**----------*----** --* ' ---------------------Reference 1 presented a review of the abnormal operational transients to determine those which might be significantly affected by the scram reactivity function changes and thus require re-analysis. Those transients determined to require re-analysis were the steam flow disturbances, feedwater system transients, and recirculation system transients. These same transients required re-analysis due to the new changes and the results are presented herein .. The transients were re-analyzed for both the "B" and the EOC scram reactivity curves. Also included in this document for completeness are the analyses of the pump trip transients. The, intent of this report is to provide the description and analytical results using the .above modifications in conjunction with the particular scram reactivity function curve for the affected failure*-caused transients. This information represents the. technical bases supporting plant operation for the next refueling cycle. 3 | ||
.. ..... ,' ...... __ -I . 3.0 ,)' -. SUMMARY OF RESULTS I 'The analyses of the single-failure-caused abnormal. ! operational transients were performed considering the modifications of relief valve rep la cement, raised *safety valve set points and faster scram times in junction with the "B".and EOC scram reactivity curves .. The above, when combined with the nuclear and hydraulic input parameters consistent with the mixed bundle f7x7 plus 8x8') cor*es, *re*s\llte*a *in"a-n op*erating plan for the next cycle which maintained the consequeoces of the transient analyses within the established limits i for Dresden 3 and Quad Cities 1. ! :_ ____ --* --------**-------.:. The cycle operating pla.n for each reactor established by the analysis contained herein is to operate at 100% of licensed rated power to that exposure in the cycle at which the scram reactivity function is equivalent to the "B" scram reactivity curve. At this point the power . will be reduced to 9 3% of licensed rated power for operation during the remainder of the cycle. The scram reactivity insertion function for the first pa rt of the cycle is described by the "B" curve and for the last part of the cycle by the EOC curve (Figure 1 ). The estimated exposure into the cycle to reach the "B" scram reactivity shape is approximately 3400 Mwd /T for Dresden 3 and approximately 4000 Mwd /T_ for Quad Cities 1. The exact*value will be derived well before it is approached and submitted approximately 30 days before tl:i.e cycle reaches that value. \ * * " 4. 0 DISC USS ION 4. 1 Scram Reactivity During the past year the Dresden 2, 3 and the Quad Cities 1, 2 systems have been analyzed for several exposure intervals based on different scram reactivity insertion curves (Refs. 1, 2 and 3). These analyses have been based on the "A" curve (FSAR), the "B" curve, and the end of cycle "C" curve, (This end of cycle "C" curve was calculated for D3 EOC 2 and submitted as Dresden Special Report No. 29, Supplement A.) Each of these curves described a different average core* | .. ..... ,' ...... __ -I . 3.0 ,)' -. SUMMARY OF RESULTS I 'The analyses of the single-failure-caused abnormal. ! operational transients were performed considering the modifications of relief valve rep la cement, raised *safety valve set points and faster scram times in junction with the "B".and EOC scram reactivity curves .. The above, when combined with the nuclear and hydraulic input parameters consistent with the mixed bundle f7x7 plus 8x8') cor*es, *re*s\llte*a *in"a-n op*erating plan for the next cycle which maintained the consequeoces of the transient analyses within the established limits i for Dresden 3 and Quad Cities 1. ! :_ ____ --* --------**-------.:. The cycle operating pla.n for each reactor established by the analysis contained herein is to operate at 100% of licensed rated power to that exposure in the cycle at which the scram reactivity function is equivalent to the "B" scram reactivity curve. At this point the power . will be reduced to 9 3% of licensed rated power for operation during the remainder of the cycle. The scram reactivity insertion function for the first pa rt of the cycle is described by the "B" curve and for the last part of the cycle by the EOC curve (Figure 1 ). The estimated exposure into the cycle to reach the "B" scram reactivity shape is approximately 3400 Mwd /T for Dresden 3 and approximately 4000 Mwd /T_ for Quad Cities 1. The exact*value will be derived well before it is approached and submitted approximately 30 days before tl:i.e cycle reaches that value. \ *-1-* " 4. 0 DISC USS ION 4. 1 Scram Reactivity During the past year the Dresden 2, 3 and the Quad Cities 1, 2 systems have been analyzed for several exposure intervals based on different scram reactivity insertion curves (Refs. 1, 2 and 3). These analyses have been based on the "A" curve (FSAR), the "B" curve, and the end of cycle "C" curve, (This end of cycle "C" curve was calculated for D3 EOC 2 and submitted as Dresden Special Report No. 29, Supplement A.) Each of these curves described a different average core* | ||
* exposure condition, the net effect being a gradual shifting of the curve to a function that tended toward less negative reactivity in the early portion of the scram stroke. This 4 | * exposure condition, the net effect being a gradual shifting of the curve to a function that tended toward less negative reactivity in the early portion of the scram stroke. This 4 | ||
.1 ;)'. tendency has been noted in all analyses done for operating BWRs. The shifting of the scram reactivity curve is contributed to by the Ha ling principle of reactivity control.. *Over the course of several fuel cycles, the core exposure will increase until an equilibrium end . of cycle value is attained. Subsequent replacement of high exposure fuel with new fuel will lower the average *wfre*re**upo*n*the p*rogr*essive 'buildup of exposure on the refueled core will again approach the equilibrium value. As the core average exposure progresses toward the equilibrium value, the end of cycle variations in *exposure from cycle to cycle become less and less. The re'sult of this from a scram reactivity standpoint is a smaller and smaller variation in the scram reactivity insertion function. The predominant change in currently operating plants is the initial shift from the FSAR curve to* the first end of cycle scram reactivity curve. Subsequent end of cycle curves show this trend to smaller and smaller variations. The variations in the EOC curvves between Dresden 3 and Quad Cities 1 for the next reload are small and do not result in the reduction of the recommended margins. The EOC curve is used in the analyses which is applicable to the last pa rt of the cycle. As stated above, the replacement of high exposure fuel with. new fuel lowers the core average exposure; increases core reactivity, and improves the scram reactivity insertion rate when compared to the end of cycle/equilibrium value. This has the effect of restoring scram reactivity insertion capability as shown by the comparison of the 11B11 and EOC curves in Figure 1. Analysis of the reloaded cores has determined that the first ""3400 Mwd /T for Dresden 3 (-4000 Mwd /T for Quad Cities 1} has a scram reactivity | .1 ;)'. tendency has been noted in all analyses done for operating BWRs. The shifting of the scram reactivity curve is contributed to by the Ha ling principle of reactivity control.. *Over the course of several fuel cycles, the core exposure will increase until an equilibrium end . of cycle value is attained. Subsequent replacement of high exposure fuel with new fuel will lower the average *wfre*re**upo*n*the p*rogr*essive 'buildup of exposure on the refueled core will again approach the equilibrium value. As the core average exposure progresses toward the equilibrium value, the end of cycle variations in *exposure from cycle to cycle become less and less. The re'sult of this from a scram reactivity standpoint is a smaller and smaller variation in the scram reactivity insertion function. The predominant change in currently operating plants is the initial shift from the FSAR curve to* the first end of cycle scram reactivity curve. Subsequent end of cycle curves show this trend to smaller and smaller variations. The variations in the EOC curvves between Dresden 3 and Quad Cities 1 for the next reload are small and do not result in the reduction of the recommended margins. The EOC curve is used in the analyses which is applicable to the last pa rt of the cycle. As stated above, the replacement of high exposure fuel with. new fuel lowers the core average exposure; increases core reactivity, and improves the scram reactivity insertion rate when compared to the end of cycle/equilibrium value. This has the effect of restoring scram reactivity insertion capability as shown by the comparison of the 11B11 and EOC curves in Figure 1. Analysis of the reloaded cores has determined that the first ""3400 Mwd /T for Dresden 3 (-4000 Mwd /T for Quad Cities 1} has a scram reactivity | ||
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, . peak neutron flux to 76. 2 per cent of the full power level and the peak average surface heat flux to 65. 6 per cent. As core inlet sub-cooling is increasing, there is no significant decrease in thermal margins. The action of the bypass system and the reactor scrarn limits the vessel pressure rise to about 66 psi. Peak steamline pressure is about 1014 psig which is well below the setpoint of the first relief valve. The analysis performed assumed that the reactor feedwater pumps were also tripped on high water level, thus terminating the increase in reactor water level. 5. 2. 3. Pump Trips An abrupt reduction in core flow causes an increase in the core void fraction and thereby *decreases reactor power. The fuel surface heat flux decreases more slowly than the* coolant flow because of the lag due to the fuel time constant, so thermal margins momentarily decrease. Therefore of . primary concern are the fuel thermal margins which are experienced throughout these transients.* They finally will *.lead to steady-state power/flow characteristics with thermal margins greater than the initial high power condition. The rotating inertias of the recirculation flow control system are chosen to ir ovide acceptable flow reductions for all pump trip possibilities. The worst cases, where the fuel thermal margins are of greatest concern, occur for maximum initial power. 5. 2. 3. 1. Two Pump Trip The two loop trip provides an evaluation of the thermal margins provided by the rotating inertia of the recirculation drive equipment. The decrease in flow causes additional . void formation in the core which decreases reactor power. , The time constants of the fuel cause the surface heat flux to lag behind the flow decay, and the mismatch between reactor thermal power and recirculation flow brings about a decrease in the critical heat flux ratio of the reacto.r. When necessary, the rotating inertia can be increased such that the flow coastdown following the drive motor trip becomes slower and the power/flow mismatch less severe. This analysis used only the rotating inertia available from the recirculation drive equipment. The minimum critical heat flux ratio (MCHFR) was analyzed, first: having a thermal limiting 7x7 fuel channel and second: having a thermal limit.., ing 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 70 was found to occur at about 2. 02 seconds .after the trip of both recirculation drive motors. 15 | , . peak neutron flux to 76. 2 per cent of the full power level and the peak average surface heat flux to 65. 6 per cent. As core inlet sub-cooling is increasing, there is no significant decrease in thermal margins. The action of the bypass system and the reactor scrarn limits the vessel pressure rise to about 66 psi. Peak steamline pressure is about 1014 psig which is well below the setpoint of the first relief valve. The analysis performed assumed that the reactor feedwater pumps were also tripped on high water level, thus terminating the increase in reactor water level. 5. 2. 3. Pump Trips An abrupt reduction in core flow causes an increase in the core void fraction and thereby *decreases reactor power. The fuel surface heat flux decreases more slowly than the* coolant flow because of the lag due to the fuel time constant, so thermal margins momentarily decrease. Therefore of . primary concern are the fuel thermal margins which are experienced throughout these transients.* They finally will *.lead to steady-state power/flow characteristics with thermal margins greater than the initial high power condition. The rotating inertias of the recirculation flow control system are chosen to ir ovide acceptable flow reductions for all pump trip possibilities. The worst cases, where the fuel thermal margins are of greatest concern, occur for maximum initial power. 5. 2. 3. 1. Two Pump Trip The two loop trip provides an evaluation of the thermal margins provided by the rotating inertia of the recirculation drive equipment. The decrease in flow causes additional . void formation in the core which decreases reactor power. , The time constants of the fuel cause the surface heat flux to lag behind the flow decay, and the mismatch between reactor thermal power and recirculation flow brings about a decrease in the critical heat flux ratio of the reacto.r. When necessary, the rotating inertia can be increased such that the flow coastdown following the drive motor trip becomes slower and the power/flow mismatch less severe. This analysis used only the rotating inertia available from the recirculation drive equipment. The minimum critical heat flux ratio (MCHFR) was analyzed, first: having a thermal limiting 7x7 fuel channel and second: having a thermal limit.., ing 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 70 was found to occur at about 2. 02 seconds .after the trip of both recirculation drive motors. 15 | ||
: 5. 2. 3. 1 Two Pump Trip (continued) With the thermal limiting 8x8 fuel channel a MCHFR of 1. 61 was 'found to occur at about 2. 33 seconds after the trip of both recirculation drive motors. Figure 5 *shows the results of a two pump trip. In both cases no damage to the fuel barrier occurs. No scram is initiated directly by the simultaneous pump trip and the power settles out at part load and natural cl.rculation .. c.onditi0ns. Nuclear* .system pr.es sure deer.eases throughout the transient such that the nuclear systen1 process barrier is not threatened by over pres sure. 5. 2. 3. 2 One Pump Trip Normal trip of one recirculation loop is accomplished through the drive motor breaker. However, a worse coastdown transient occurs if the generator field excitation breaker is opened, separating the pump and its motor from the inertia of the MG set. Results of this transient is shown in Figure 6. Diffuser flow on the tripped side reverse at about five seconds, however, M ratio (pump suction/drive flow ratio) in the active jet pumps increases greatly, producing about 142 per cent of normal diffuser flow and about 60 per cent of rated core flow. MCHFR was calculated the same way as in the two pump trip,* first with ... a thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 74 was found to occur at about 1. 59 seconds. While with the thermal limiting 8x8 fuel channel a MCHFR of 1. 67 was found to occur about 1. 59 seconds. Even for this case in which instantaneous coolapse of the generator field was assumed, no. damage to the fuel barrier occurs. 5. 2. 3. 3 Pump Seizure This case represents the instantaneous stoppage of one pump motor shaft. It produces the most rapid decrease of core . flow. The.reactor is assumed to be operating at maximum power conditions. Figure 7 shows the results of this transient. Note the fast decrease. in drive flow in the seized loop due to the large loss introduced by the stopped rotor. Jet pump diffuser flow on the seized loop reverse at about 800 milliseconds. Core flow reaches its minimum value in about 1. 2 seconds. MCHFR is again analyzed by, first with a* thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With a thermal limiting 7x7 fuel channel a MCHFR of 1. 21 was found to occur at about 2. 00 seconds. With the thermal limiting 8x8 fuel channel a MCHFR of 1. 12 was found to occur at about 1. 90 seconds. Nucleate boiling is maintained since MCHFR is greater than 1. 0, fore,. no damage occurs to the fuel clad barrier. The initial 16 -**- | : 5. 2. 3. 1 Two Pump Trip (continued) With the thermal limiting 8x8 fuel channel a MCHFR of 1. 61 was 'found to occur at about 2. 33 seconds after the trip of both recirculation drive motors. Figure 5 *shows the results of a two pump trip. In both cases no damage to the fuel barrier occurs. No scram is initiated directly by the simultaneous pump trip and the power settles out at part load and natural cl.rculation .. c.onditi0ns. Nuclear* .system pr.es sure deer.eases throughout the transient such that the nuclear systen1 process barrier is not threatened by over pres sure. 5. 2. 3. 2 One Pump Trip Normal trip of one recirculation loop is accomplished through the drive motor breaker. However, a worse coastdown transient occurs if the generator field excitation breaker is opened, separating the pump and its motor from the inertia of the MG set. Results of this transient is shown in Figure 6. Diffuser flow on the tripped side reverse at about five seconds, however, M ratio (pump suction/drive flow ratio) in the active jet pumps increases greatly, producing about 142 per cent of normal diffuser flow and about 60 per cent of rated core flow. MCHFR was calculated the same way as in the two pump trip,* first with ... a thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 74 was found to occur at about 1. 59 seconds. While with the thermal limiting 8x8 fuel channel a MCHFR of 1. 67 was found to occur about 1. 59 seconds. Even for this case in which instantaneous coolapse of the generator field was assumed, no. damage to the fuel barrier occurs. 5. 2. 3. 3 Pump Seizure This case represents the instantaneous stoppage of one pump motor shaft. It produces the most rapid decrease of core . flow. The.reactor is assumed to be operating at maximum power conditions. Figure 7 shows the results of this transient. Note the fast decrease. in drive flow in the seized loop due to the large loss introduced by the stopped rotor. Jet pump diffuser flow on the seized loop reverse at about 800 milliseconds. Core flow reaches its minimum value in about 1. 2 seconds. MCHFR is again analyzed by, first with a* thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With a thermal limiting 7x7 fuel channel a MCHFR of 1. 21 was found to occur at about 2. 00 seconds. With the thermal limiting 8x8 fuel channel a MCHFR of 1. 12 was found to occur at about 1. 90 seconds. Nucleate boiling is maintained since MCHFR is greater than 1. 0, fore,. no damage occurs to the fuel clad barrier. The initial 16 -**- | ||
}* I I-' -...J 0 w I-a: a: u.. D I-z w u a: w a.. 150. 100. 50. 1 NEUTfltlN FLUX I 2 PEAK FUEL CENTI:R TEMP 3 AVE SURFACE Hl'm' FLUX II FEEOHATER FLCH VESSa STEAM F,LCH I I . .... -***-8. --*---. -. 12. **. **--*-***-.. .... 16. TIME tSECl. -so ............... ____ _..,...__ ____ .,..,_ _____ ._ ___ _ fl. 1,1. 8. 12. 16. TIME CSECl Figure 5. --TRIP CJF HJCJ DRIVE MCJTCJAS 100 PC PCJWEA 160.t------so_ * ._._ .................. _._._...,,._ ____ 1,1. 8. . 12. TIME tSECl 1 NEUTRdN FLUX 2 SURF HEAT. FUJX 120.1-------+-------+------ -------....,---Cl w l-a: a: LL. D 1-z w 1,1(). u a: w 0.... o. 25. --*So. 75. 100. CCJAE FLCJW (/.) *- | }* I I-' -...J 0 w I-a: a: u.. D I-z w u a: w a.. 150. 100. 50. 1 NEUTfltlN FLUX I 2 PEAK FUEL CENTI:R TEMP 3 AVE SURFACE Hl'm' FLUX II FEEOHATER FLCH VESSa STEAM F,LCH I I . .... -***-8. --*---. -. 12. **. **--*-***-.. .... 16. TIME tSECl. -so ............... ____ _..,...__ ____ .,..,_ _____ ._ ___ _ fl. 1,1. 8. 12. 16. TIME CSECl Figure 5. --TRIP CJF HJCJ DRIVE MCJTCJAS 100 PC PCJWEA 160.t------so_ * ._._ .................. _._._...,,._ ____ 1,1. 8. . 12. TIME tSECl 1 NEUTRdN FLUX 2 SURF HEAT. FUJX 120.1-------+-------+-------1--------....,---Cl w l-a: a: LL. D 1-z w 1,1(). u a: w 0.... o. 25. --*So. 75. 100. CCJAE FLCJW (/.) *- | ||
v:-:* I-' CD 1 NEUTRCIN FLUX 2 PEAK FUEL CEN R TE!1P 3 RVE SUFf'ACE T FLUX ij FEEOa<lTEA fl ..................................... TIME (SECl ............................................. TIME * (SECl Cl LLJ >-a: 0: LL 0 >-z LLJ u 0: LLJ 0... Figure 6 --TRIP OF ONE GEN FIELD BREAKER 100 PC POWER -, ;* ...................................... ** TIME CSECJ 1 NEUT N FLUX 2 SURF HERT FLUX 80. ILO. I so. 75. CORE FLOW C/.l 100. 25. r Cl LLJ t-a: a: LL o. t-z LLJ L) a: LLJ 0.. ,,, ,1 l> I 100. 50. ................. _._.._._._,_...__ ____ ....,.... _____ ...__ ____ ob: 8. 12. 16. TIME (SEC) 100. 1 LEVEL IINCH-AE SEP-SKIRT 2 SENSED LEVEL INCHES! 3 TURBINE STEAM OH l7. I ij OH:: INLET Fl t % l DRIVE FLOH 1 7. l o.1--------+------ ------+----- ----TIME CSEC) Figure 7 -..; ONE PUMP SEIZURE 100 PC PC!WER .. **;-........................................ TIME (SECl 1N£ Nl'l.UX 2 SURF HEAT FLUX 120.1------ ----- ----- ----- ---'---Cl 80. LLJ t-a: a: LL 0 t-z w ij(). L) a: LLJ 0.. CC!RE FLC!W C/.l '\" t:* : I' I r:* / .. ; *r, .t' 1, 1. -* r 5.2.3.3 Pump Seizure (continued) pressure regulator maintains pressure control as the reactor settles out at the final lower power condition. No scram occurs. Because the nuclear system pressure decreases throughout the transiEn t the nuclear system process barrier is not threatened by over pressure. 5. 3 "B" Scram Reactivity Curve Analyses A ssuniing the previously described changes (Section 4. 0), the transient analyses verifies operation at 100% of licensed power with the "B" scram reactivity curve. An acceptable pressure margin ( 25 psi) between the peak pressure at the turbine trip without bypass with trip scram transient and the first safety valve setpoint is maintained. The Target Rock safety/relief valve plus the eight spring safety valves at the new higher setpoints were shown to provide adequate overpressure protection in accordance with ASME Code requirements. 5. 3. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient The relief valve sizing transients were reanalyzed at the exposed core conditions to verify the adequacy of the four electromatic relief valves plus one Target Rock safety I relief valve (at reduced flow) to terminate the pressure transient when the reactor is subjected to a rapid pressurization evert such that satisfactory margins are maintained between the peak pressure resulting and the first safety valve setpoint. The event analyzed is the turbine trip with simultaneous reactor scram but with a failure of the turbine bypass system. The results for this transient are illustrated by Figure 8. Initial conditions prio°r to the event a re reactor power at the 100 per cent level, core recirculation flow at 98 million lb. per hour, and reactor steam dome pressure at 1005 psig. The first relief valve setpoint is 1125psig and the remaining four valves have setpoints of 1130.and 1135 psig, two valves at each setpoint. The sudden closure of the turbine stop valves with no initial bypass flow causes a rapid rise in system pressure at a rate essentially double that which results when the bypass system functions. Position switches on the stop valves initiate immediate reactor scram. The rapid pressurization causes core void collapse and neutron flux increases, pea king at a bout 143 per cent of the initial value before the scram becomes effective. Co re average surface heat flux dips initially and then increases to slightly over 100 per cent of its initial value at about L 36 seconds. No significant 20 a w I-a: a: LL. D I-:z: w Ll a: w CL I\) ....... 1 VESSEL PRES CPSil 2 snt LINE PRESf\ISE \RSI! 3 TUH81KE PRES RiSE \PSIJ ll BrPf;SS v;1LVE FUll1 l7.l, 350 | v:-:* I-' CD 1 NEUTRCIN FLUX 2 PEAK FUEL CEN R TE!1P 3 RVE SUFf'ACE T FLUX ij FEEOa<lTEA fl ..................................... TIME (SECl ............................................. TIME * (SECl Cl LLJ >-a: 0: LL 0 >-z LLJ u 0: LLJ 0... Figure 6 --TRIP OF ONE GEN FIELD BREAKER 100 PC POWER -, ;* ...................................... ** TIME CSECJ 1 NEUT N FLUX 2 SURF HERT FLUX 80. ILO. I so. 75. CORE FLOW C/.l 100. 25. r Cl LLJ t-a: a: LL o. t-z LLJ L) a: LLJ 0.. ,,, ,1 l> I 100. 50. ................. _._.._._._,_...__ ____ ....,.... _____ ...__ ____ ob: 8. 12. 16. TIME (SEC) 100. 1 LEVEL IINCH-AE SEP-SKIRT 2 SENSED LEVEL INCHES! 3 TURBINE STEAM OH l7. I ij OH:: INLET Fl t % l DRIVE FLOH 1 7. l o.1--------+-------1-------+------1-----TIME CSEC) Figure 7 -..; ONE PUMP SEIZURE 100 PC PC!WER .. **;-........................................ TIME (SECl 1N£ Nl'l.UX 2 SURF HEAT FLUX 120.1-------1-------1-------1-------1----'---Cl 80. LLJ t-a: a: LL 0 t-z w ij(). L) a: LLJ 0.. CC!RE FLC!W C/.l '\" t:* : I' I r:* / .. ; *r, .t' 1, 1. -* r 5.2.3.3 Pump Seizure (continued) pressure regulator maintains pressure control as the reactor settles out at the final lower power condition. No scram occurs. Because the nuclear system pressure decreases throughout the transiEn t the nuclear system process barrier is not threatened by over pressure. 5. 3 "B" Scram Reactivity Curve Analyses A ssuniing the previously described changes (Section 4. 0), the transient analyses verifies operation at 100% of licensed power with the "B" scram reactivity curve. An acceptable pressure margin ( 25 psi) between the peak pressure at the turbine trip without bypass with trip scram transient and the first safety valve setpoint is maintained. The Target Rock safety/relief valve plus the eight spring safety valves at the new higher setpoints were shown to provide adequate overpressure protection in accordance with ASME Code requirements. 5. 3. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient The relief valve sizing transients were reanalyzed at the exposed core conditions to verify the adequacy of the four electromatic relief valves plus one Target Rock safety I relief valve (at reduced flow) to terminate the pressure transient when the reactor is subjected to a rapid pressurization evert such that satisfactory margins are maintained between the peak pressure resulting and the first safety valve setpoint. The event analyzed is the turbine trip with simultaneous reactor scram but with a failure of the turbine bypass system. The results for this transient are illustrated by Figure 8. Initial conditions prio°r to the event a re reactor power at the 100 per cent level, core recirculation flow at 98 million lb. per hour, and reactor steam dome pressure at 1005 psig. The first relief valve setpoint is 1125psig and the remaining four valves have setpoints of 1130.and 1135 psig, two valves at each setpoint. The sudden closure of the turbine stop valves with no initial bypass flow causes a rapid rise in system pressure at a rate essentially double that which results when the bypass system functions. Position switches on the stop valves initiate immediate reactor scram. The rapid pressurization causes core void collapse and neutron flux increases, pea king at a bout 143 per cent of the initial value before the scram becomes effective. Co re average surface heat flux dips initially and then increases to slightly over 100 per cent of its initial value at about L 36 seconds. No significant 20 a w I-a: a: LL. D I-:z: w Ll a: w CL I\) ....... 1 VESSEL PRES CPSil 2 snt LINE PRESf\ISE \RSI! 3 TUH81KE PRES RiSE \PSIJ ll BrPf;SS v;1LVE FUll1 l7.l, 350 | ||
* 6 5GFEll' VMLVE l'lM \7.J 100.' so. o. o. !!. e. TIME 100. 16. en a: a: ....J ....J D D 200. -1oob. I!. s. TIME lSECl Figure 8 Turbine Trip Without Bypass 6. TIME 12. 16. lSECl 1 NET AE?.CTIVIrGrr | * 6 5GFEll' VMLVE l'lM \7.J 100.' so. o. o. !!. e. TIME 100. 16. en a: a: ....J ....J D D 200. -1oob. I!. s. TIME lSECl Figure 8 Turbine Trip Without Bypass 6. TIME 12. 16. lSECl 1 NET AE?.CTIVIrGrr | ||
* 2 SCRfil'1 RERCT!V ll 3 rol'Pl..ER ll VO!O AEflC VI 1 6. 12. 16. TIME CSECl l l l I r | * 2 SCRfil'1 RERCT!V ll 3 rol'Pl..ER ll VO!O AEflC VI 1 6. 12. 16. TIME CSECl l l l I r | ||
Revision as of 02:11, 2 May 2018
| ML17252B182 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 03/29/1974 |
| From: | Commonwealth Edison Co |
| To: | US Atomic Energy Commission (AEC) |
| References | |
| Download: ML17252B182 (43) | |
Text
l * ........ .:/ r*--. \ DRESDEN STATION REPORT NO. 29, SUPPLEMENT B TRANSIENT ANALYSES FOR DRESDEN-3 CYCLE 3 AND. QUAD-CITIES-1 CYCLE 2 / COMMONWEALTH EDISON COMPANY
."J' . I . , 1. 0 ' 2.0 3.0 4;0 5.0 .. ., . Introduction Purpose and Scope Summary of Results Discussion 4. 1 Scram *Reactivity 4.2 Control Rod Scram Times 4.3 Safety and Relief Valves Analyses 5. 1 Abnormal Operational Transients 5. 2 Transients Not Affected by Scram 5. 2. 1 Flow Control Malfunction -Full Coupling Demand 5. 2. 2 .. Feedwater Controller Failure -Maximum Demand 5. 2. 3 :Pump Trips 5. 2. 3. 1 Two Pump Trip 5. 2. 3. 2 One Pump Trip ,5, 2. 3. 3 Pump Seizure .5. 3 "B" Scram Reactivity Curve Analyses 5. 3. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient 5. 3. 2.
- Ma in Steam Isolation Valve Closure with Indirect Scram -Safety Valve Adequacy Transient 5. 3. 3 Nominal Turbine Trip 5. 3. 4 Loss of Generator Load *. 5. 3. 5 Loss of Ma in Condenser Vacuum 5. 3. 6 Main Steam Isolation Valve Closure -Position Scram 5. 4 EOC Scram Reactivity Curve Analysis 5. 4. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient 5. 4. 2 Main Steam Isolation Valve Closure With Indirect Scram -Safety Valve Adequacy Transient 5. 4. 3 Nominal Turbine Trip 5. 4. 4 Loss of Generator Load 5. 4. 5 Loss of Main Conde.nser Vacuum 5. 4. 6 Main Steam Isolation Valve Closure -Position Scram
... *. .. . -... l. 6. 0 Technical Specification Changes 6. 1 Scope of Changes 6. 2 *Specific Changes 7. 0 References ( -**-.
' . 1. 0 INTRODUCTION Analytical models have been effectively used for many years to describe the various operations, transients and hypothetical accidents associated with GE BWRs. The development of
- increasingly more sophisticated analytical methods has provided *a higher level of understanding of BWR dynamics which in turn has permitted the adjustment of operating techniques with the ultimate goals of greater plant safety and improved plant performance. Dne such development has led to a better definition of the scram reactivity insertion function in exposed cores, and the effects of changes in core dynamic characteristics resulting from the e*xposure dependent change in the scram reactivity function. Previous analyses (References 1, 2 and 3), have shown that reactor power restrictions were necessary to ensure continuation of plant operation within previously established limits. Meanwhile, further evaluations were in progress to determine what action could be te1:ken to allow operation at the highest reactor power within FSAR established limits. As a result of this evaluation it was determined that the end of cycle curve for the next reload cycle would be used. Based on this end of cycle scram re*activity curve (which has been defined generally as the 11C11 curve) a combination of plant and analytical fications have been determined which minimizes the consequences of single-failure-caused abnormal operational transients yet
- maintains the established limits and margins. In conjunction with the end of cycle curve, the "B" curve, previously described in *reference 1, has been used for the beginning of cycle evaluation with the same combination of plant and analytical modifications. Figure 1 is a comparison of the two curves. Together, these evaluations form the transient analyses bases for the proposed plant modifications and technical specification changes discussed
- in this report. 1 --
.,-70 60 >, .µ *r-> *r-.µ u '° Q) 50 0::: E '° u V') 40 30. 20 10 0 0 . _.,* -" .. *. ,, "B" Curve 0.4 0.6 0.8 . Fraction of Core Fully Controlled Figure, 1: Scram Reactivity Curves 2. 'f.'* .. . ***.* .... . *' , c .' EOCCurve 1.0
. \ '* .i 2.0 PVRPOSE AND SCOPE This report is intended to provide the nee es sa ry justifications for adoption and implem.entation of the proposed plant and analysis modifications. While this report is considered a specific submittal on these modifications, the contents relate to, and are based on, the nuclear and thermal-hydraulic parameters derived for mixed 7x7 and 8x8 fuel bundle cores to be loaded into the Dresden 3 and Quad Cities 1 reactor systems at the next rrntage Because the history of scram reactivity insertion function changes has been reported in great detail in previous submittals ( 1), a *bac;kgrouna aiscuss*ion 'is not req\iire*a in this report. The earlier reports a re, however, included by reference. The proposed plant modifications are: Replacement of one electromatic relief valve with a combina:.. tion safety/relief valve, and Raise the setpoints on the eight spring safety valves** The proposed analytical modifications are: Use of more realistic control rod scram times, and, similarly, ---.. -----**----------*----** --* ' ---------------------Reference 1 presented a review of the abnormal operational transients to determine those which might be significantly affected by the scram reactivity function changes and thus require re-analysis. Those transients determined to require re-analysis were the steam flow disturbances, feedwater system transients, and recirculation system transients. These same transients required re-analysis due to the new changes and the results are presented herein .. The transients were re-analyzed for both the "B" and the EOC scram reactivity curves. Also included in this document for completeness are the analyses of the pump trip transients. The, intent of this report is to provide the description and analytical results using the .above modifications in conjunction with the particular scram reactivity function curve for the affected failure*-caused transients. This information represents the. technical bases supporting plant operation for the next refueling cycle. 3
.. ..... ,' ...... __ -I . 3.0 ,)' -. SUMMARY OF RESULTS I 'The analyses of the single-failure-caused abnormal. ! operational transients were performed considering the modifications of relief valve rep la cement, raised *safety valve set points and faster scram times in junction with the "B".and EOC scram reactivity curves .. The above, when combined with the nuclear and hydraulic input parameters consistent with the mixed bundle f7x7 plus 8x8') cor*es, *re*s\llte*a *in"a-n op*erating plan for the next cycle which maintained the consequeoces of the transient analyses within the established limits i for Dresden 3 and Quad Cities 1. ! :_ ____ --* --------**-------.:. The cycle operating pla.n for each reactor established by the analysis contained herein is to operate at 100% of licensed rated power to that exposure in the cycle at which the scram reactivity function is equivalent to the "B" scram reactivity curve. At this point the power . will be reduced to 9 3% of licensed rated power for operation during the remainder of the cycle. The scram reactivity insertion function for the first pa rt of the cycle is described by the "B" curve and for the last part of the cycle by the EOC curve (Figure 1 ). The estimated exposure into the cycle to reach the "B" scram reactivity shape is approximately 3400 Mwd /T for Dresden 3 and approximately 4000 Mwd /T_ for Quad Cities 1. The exact*value will be derived well before it is approached and submitted approximately 30 days before tl:i.e cycle reaches that value. \ *-1-* " 4. 0 DISC USS ION 4. 1 Scram Reactivity During the past year the Dresden 2, 3 and the Quad Cities 1, 2 systems have been analyzed for several exposure intervals based on different scram reactivity insertion curves (Refs. 1, 2 and 3). These analyses have been based on the "A" curve (FSAR), the "B" curve, and the end of cycle "C" curve, (This end of cycle "C" curve was calculated for D3 EOC 2 and submitted as Dresden Special Report No. 29, Supplement A.) Each of these curves described a different average core*
- exposure condition, the net effect being a gradual shifting of the curve to a function that tended toward less negative reactivity in the early portion of the scram stroke. This 4
.1 ;)'. tendency has been noted in all analyses done for operating BWRs. The shifting of the scram reactivity curve is contributed to by the Ha ling principle of reactivity control.. *Over the course of several fuel cycles, the core exposure will increase until an equilibrium end . of cycle value is attained. Subsequent replacement of high exposure fuel with new fuel will lower the average *wfre*re**upo*n*the p*rogr*essive 'buildup of exposure on the refueled core will again approach the equilibrium value. As the core average exposure progresses toward the equilibrium value, the end of cycle variations in *exposure from cycle to cycle become less and less. The re'sult of this from a scram reactivity standpoint is a smaller and smaller variation in the scram reactivity insertion function. The predominant change in currently operating plants is the initial shift from the FSAR curve to* the first end of cycle scram reactivity curve. Subsequent end of cycle curves show this trend to smaller and smaller variations. The variations in the EOC curvves between Dresden 3 and Quad Cities 1 for the next reload are small and do not result in the reduction of the recommended margins. The EOC curve is used in the analyses which is applicable to the last pa rt of the cycle. As stated above, the replacement of high exposure fuel with. new fuel lowers the core average exposure; increases core reactivity, and improves the scram reactivity insertion rate when compared to the end of cycle/equilibrium value. This has the effect of restoring scram reactivity insertion capability as shown by the comparison of the 11B11 and EOC curves in Figure 1. Analysis of the reloaded cores has determined that the first ""3400 Mwd /T for Dresden 3 (-4000 Mwd /T for Quad Cities 1} has a scram reactivity
- insertion function which is equal to or greater than the 11B11 curve. Thus, the initial part of the cycle uses the "B" curve for the analyses. The 100 per cent scram reactivity worth (see Figure 1), in dollars is -43. 125 for the "B" curve and -38. 05 for the EOC curve. 4. 2 Control Roel Scram Times The Technical Specifications presently require the average scram times for all operable control rods to be as tabulated below: 5
. .,*. *r* -------------.,.---- 4.3 5* 20 50 90 0. 375 0. 900 2.00 5.00 These scram times show a large reduction in insertion rate after the 50 per cent inserted point. Experience has shown that the insertion rate is more nearly constant for the entire --*,st>ro*ke ... BpecHi0a*Hy, the*r.e +s no "knee" at -the 50% point. Also, experience has shown that the actual control rod formance is better than these values. Thus, the implementation *of a faster control rod scram time to result in a linear insertion rate is feasible. By using a-900/o insertion time of 3. 50 seconds, the full insertion becomes a linear function similar to the actual observed rod performance. This faster control rod scram time has a small, though measurable, effect on the transients analyzed. The primary impetus for this change is the more realistic evaluation of control rod performance than that presently assumed. The new Technical Specification is shown on Figure 2 along with other Technical Specifications and the range of typical plant experience. In conjunction with .. this change, the Technical Specification for the fastest three rods will be changed to 3. 8 seconds for the 9 0% insertion point. Safety and Relief Valves . The primary concerns resulting from the changing of the scram reactivity curve have.been in satisfying required pressure margins and minimizing fuel thermal duty as a result of the occurrence o.f abnormal operational transients, particularly events associated with turbine and generator trips with concurrent failure of the pass valves. While other transient events are affected, the sequences are such that recommended margins are maintained. GE recommends the mai'l1tenance of a minimum margin of 25 psi between the peak pressure resulting from the worst case tion event (turbine trip without bypass) and the setpoint of the lowest sef spring safety valve throughout the lifetime of the plant. Previous analyses have revealed possible difficulty in maintaining this margin. Resetting the safety valve setpoints and addition of one Target Rock combination safety/relief valve will help maintain this margin. The replacement of of the electromatic relief valves with the combination safety/relief valve, set to actuate at the relief valve setpoint of 1125 psig, satisfies the ASME Code requirement for the lowest safety valve to actuate at a pressure such that the peak design pressure anywhere in the vessel is not exceeded. With this requirement met, the spring safety valve setpoints can be raised to a value not to exceed the pressure that is equivalent to._110% of the vessel design pressure at the highest point in the vessel. The spring safety will be re-set in pairs to setpoints of 1240, 1250 and 1260 psig. The Target Rock valve has been 80 70 60 -:: 50 c B 8. w 5 40 a: V> ..J . ..J :::> *U.. 30 20 10 RANGE OF TYPICAL EXPERIENCE / / / ORIGINAL TECflNI CAL SPECIF I CATION BWR 4 & BWR 2/3 CURRENT
- TECHNICAL SPEC! FI CATION PROPOSEtl NEW* TECHNICAL FI CATION * .. 0 0.2 0.4 0.6 0.8 1.0 2.0 3.0 4.0 . 5.0 ELAPSED TIME AFTER SCRAM SIGNAL (sec) Figure 2 --CONTROL ROD DRIVE SCRAM TIMES
,' .* _ ... * *evaluated based on the same performance characteristics as* -the electromatic valves because of discharge pipe limitations. The recommended 25 psi margin is intended to preclude lifting *of the spring safety valves during transients, the result of which would be the admission of reactor steam to the drywell air space. This does not present a plant safety problem but could introduce operating inconveniences. The required reactor vessel overpressure *prote*ction as defined by the ASME Boiler and Pressure Vessel Code are satisfied through self actuation of the combination safety/relief valve. Analyses have shown that ASME overpressure protection ments are fully met with the proposed change. Sections 5. 3. 2 and 5. 4. 2.) __ . . . --The substitution of a Target Rock valve for an Electromatie valve was accomplished by reducing the valve throat area proportionately to restrict t.he valve discharge flow to a value equivalent to the electromatic valve discharge piping flow capability in .order not to exceed the original design conditions of the discharge piping. The vendor's nameplate will read 622, 000 pounds per hour with a set-point of 112 5 psig. _
- Because the Target Rock valve discharges horizontally and the elect romatic valves discharged vertically a slight modification *of the discharge piping orientation is required. This modified piping arrangement will be .e_valuated by dynamic analysis of the piping stresses. Since the major input parameters, i.e., steam flow, and discharge piping configuration, remain the same there is no. change in the effect of relief valve operation en the torus. The Target Rock valve is substituted directly for the electromatic valve in the Automatic Blowdown Systems with no change in system function or 'performance. This is accomplished electrically by connecting the air operator sol.enoid valve in the electrical circuit in a manner similar to the electromatic solenoid. The testability and automatic discharge functions remain unchanged. 8 --
............. ' " .. .. . '*-*-.. ----------'l!-he-subst-i-t.u-t.i0n-0f-a-'F-a-:r'get;--R0ck-v-a-lve-for-an-E-lectro matic valve was accomplished by reducing the valve throat *area proportionately to restrict the valve discharge flow to a value equivalent to the electromatic valve discharge piping flow capability in order not to exceed the original design conditions of the discharge piping. The vendor's nameplate will read 622,000 pounds per hour with a setpoint of 1125 psig. * * "'Bec*aus-e*-the 'T-arget *Rock -di-scha*r:g-es *hori--zontally and the electromatic valves discharged vertically a slight modification of the discharge piping orientation is required. This modified piping arrangement will be confirmed by dynamic analysis of the piping *stresses. Since the major input parameters, i.e., steam flow, and discharge piping configuration, remain the same there is no change in the effect of relief valve operation on the torus. The Target Rock valve is substituted directly for the electromatic valve in the Automatic Discharge Systems with no change in system function or performance. This is accomplished electrically by connecting the air operator solenoid valve in the electrical circuit . in a manner similar to the electromatic solenoid. The testability and automatic discharge functions remain unchanged. 9 \
_____ _..._ I i . ' -.* ---* ....... -' .. --**-------.. --**-*-*-**-**-****--* ..... *----*-.:--**-____ ... ___ , ... _ .. --*--.-:..---* __ .:: __ . ---. 5-:-0 :A-N:KL:JYSES 5. 1 A bnc;>rmal Operational Transients A complete range of single-failure-caused events which are abnormal but reasonably expected to occur during the life of the plant were analyzed as part of the original licensing of the plant. These analyses were described in the FSAR. Subsequent sub-mittals _I:iave, where appropriate, included additional tion of those events of significance to the concept being reported (i.e., reloads, changes to transient analysis parameters, or -pla'nt *_m-odifrcati:ons-). *See Refere.nces 1, 2 and 3. The transient re-analyses can be categorized into the following *nuclear system parameter variations: Nuclear system pressure increases; Reactor vessel water (moderator) temperature decreases; Positive reactivity insertions; Reactor vessel coolant inventory decreases; Reactor core coolant flow increases; and, Reactor core coolant flow decreases. These categories were all reviewed in detail in Dresden Station Special Report No. 29 (Reference 1 ), to determine those which might be significantly affected and therefore needed reanalyses . . A subsequent review of these categories for the current analyses agreed with the Special Report 29 (SR 29} results. The following transients were found to require re-analyses in order to determine the specific changes that might occur compared to the previous analytical results: "*** ..... . 10 ........ _
.... _____ _ Turbine trip without bypass with trip scram -relief valve adequacy transient. Main steamline isolation valve closure with indirect scram -safety valve adequacy transient. No_minal turbine trip. Loss of generator load. Loss of**ma*in-*condens*e*r -*vac-uum. Main steamline isolation valve closure with position scram. Flow controller malfunction -full coupling demand. Feedwater controller malfunction -maximum demand .. The last two transients in the above list are not greatly affected by small changes in the initial part of the scram reactivity function due to the time delay in scram actuation. The change from the "B" curve to the ECC curve has an insignificant effect on these transients and they are both covered by the single evaluation contained in the next section.
- Also included in the next section are the core coolant flow decrease transients. These transients are the: One pump trip, Two pump trip, Pump seizure. These transients are included for completeness, since a scram does not occur as a direct result. The two subsequent sections contain the analytic;:al results for the remainder of the transients delineated above. These transients are affected to a significant degree by changes discussed in* Section 4. 0. Of these two sections, the first one contains the _analytical results ba.sed on the use of the "B" scram reactivity curve. * \, The second section contains the analytical results based on the use of the ECC scram reactivity curve. 5. 2 Transients Not Affected by Scram 5. 2. 1 Flow Control Malfunction -Full Coupling Demand. The flow control malfunction of failure of one of the M/G set speed controllers was re-analyzed at exposed core scram conditions. The case re-analyzed was the failure of a speed controller in the direction to demand full generator speed, starting from a typical low power initial condition with power at 65 per cent of full power and with core recirculation flow at AA I. 5. 2. 2 approximately 49 per cent of design. The failure of the controller causes the scoop positioner to move at its maxi1num rate to establish full coupling between the drive motor and the generator. The results of the transient a re illustrated by Figure 3. Both speed loops are initially operating at about 41 per cent of rated speed. Upon the failure of the controller, the generator and pump speeds of the failed loop rapidly increase to about 102 per cent of that speed which produces rated drive flow. The other loop remains at its initial speed conditions. Drive flow in the failed loop also increases, causing the jet pump diffuser flows associated with that loop to increase from 49 per cent to approximately 140 per cent of rated flow. The drive flow in the other. (non-failed) loop remains I essentially constant, (decreasing only 2 percent); however, the jet pumps associated with that loop experience decreasing diffuser flows* because of the greater core .AP, with flows essentially zero at 20 seconds. Total core flow increases rapidly to about 74 percent of rated flow causing reactor power to increase such that a reactor scram from high neutron flux results at about 3. 8 seconds. Neutron flux peaks at 170 per *cent of the value at fall p0wer. Peak surface heat flux however, increases to only 76. 5 per cent and thus thermal margins are maintained throughout t.he transient. Feedwater Controller Failure -Maximum Demand The response of the plant to an excess feedwater flow transient *was re-analyzed and is illustrated in Figure 4. The transient :was initiated from a typical low power condition with reactor power at 65 per cent of full power and core recirculation flow at 49 per cent of rated. The feedwater controller was assumed to fail such as to demand maximum feedwater valve opening resulting in about 110 per cent of design feedwater flow. Because of the low initial power level, a large mismatch between steamflow and feedwater flow results and. the level transient is more severe. The maximum rate of level rise is about 2. 5 inches per second. The insurge of excess feedwater initially causes a slight reactor *pressure and power increase. At about 5 seconds, the core inlet flow becomes cooler and a faster increase in reactor power results. Sensed water level reaches the high level turbLne trip setpoint at 7. 25 seconds causing fast closure of the turbine stop valves and simultaneous opening of the bypass valves and a reactor scram. As the bypass capacity is smaller than the steaming rate at the time of the turbine trip occurs, reactor pressure increases. The stop valve dosure scram limits the 12*
1 NEUTRON FLUX 2 PERK FUEL TEMP 3 AVE SURFACE T FLUX ij FEEOWATER F Cl 100. w I-a: a: LL. 0 I-z w 50. u a: w a.. -ob. ij. 8. 12. TIME CSECl / i ..... w 100. ................. ....... TIME CSECl Figure 3 350. 200. so. 120. Cl BO. w a: a; LL. Q 1-z w ijQ. u a; w a.. ,_ .... . -. SPEED CONT II I I FAIL 1 VEssa m:s A:!sc tPSil 2 STH LINE P!r<::S riJSE tPSU 3 TURBINE PRES FUSE Cf1Sll '" BIPASS (7.J ::. RELIEF VALVE* CW t7.J B SAFETY, VALVE t:tl ;. A 8 R I . '.-! .. I -7 '.> I ,: j ij. 8. TIME 12. 16. CSECl 1 2 SUR!' FLUX HERT FLUX I / !J__/ I \ .. .. 25. so. 75. 100 * *CORE FLOW (/.) INCREASE FLOW I I j
- 1 ., j i I 1 i d I f I ., i I \ *l ,_., I I I I *1 *.j i i l I \ *I I 1 :\ '] \ j 1 . *1 ' I . *$. ,, 1 .j A '..:.'.{ 1 . _., 0 LIJ er a: lL. 0 1-z: w u a: LIJ a.. .......................... TIME CSECJ *Figure 4 200. so. ' -1000 120. Cl 80. LIJ . I--er a: lL. 0 1-z LIJ ij(). u a: LIJ Cl.. FEEDWATEA o-* , ,. I. l VESSEL P!lES RI c tPSIJ 2 STM U NE ffiES !SE !PSU 3 TURBINE P!'ES : !SE tPSil *' B'l"PRSS VAL VE LCJw tXl :> FEUEF VAL VE : S SllFETI' VR1. VE OW !Xl OW tZl ;.: \,,. /,,; ,, """""' VI ' RI It !\ u R u .1 I 2--'l ., ? \ -*I . 10. 30. ij(). TIME CSECJ l r-.£:1Jil_1 2 SURfA IN FLUX :E HERT FLUX .... -( I /, . 25. so
- 75. 100. CDRE FLOW (%) CONTROL FAIL MAXIMUM DEMAND
, . peak neutron flux to 76. 2 per cent of the full power level and the peak average surface heat flux to 65. 6 per cent. As core inlet sub-cooling is increasing, there is no significant decrease in thermal margins. The action of the bypass system and the reactor scrarn limits the vessel pressure rise to about 66 psi. Peak steamline pressure is about 1014 psig which is well below the setpoint of the first relief valve. The analysis performed assumed that the reactor feedwater pumps were also tripped on high water level, thus terminating the increase in reactor water level. 5. 2. 3. Pump Trips An abrupt reduction in core flow causes an increase in the core void fraction and thereby *decreases reactor power. The fuel surface heat flux decreases more slowly than the* coolant flow because of the lag due to the fuel time constant, so thermal margins momentarily decrease. Therefore of . primary concern are the fuel thermal margins which are experienced throughout these transients.* They finally will *.lead to steady-state power/flow characteristics with thermal margins greater than the initial high power condition. The rotating inertias of the recirculation flow control system are chosen to ir ovide acceptable flow reductions for all pump trip possibilities. The worst cases, where the fuel thermal margins are of greatest concern, occur for maximum initial power. 5. 2. 3. 1. Two Pump Trip The two loop trip provides an evaluation of the thermal margins provided by the rotating inertia of the recirculation drive equipment. The decrease in flow causes additional . void formation in the core which decreases reactor power. , The time constants of the fuel cause the surface heat flux to lag behind the flow decay, and the mismatch between reactor thermal power and recirculation flow brings about a decrease in the critical heat flux ratio of the reacto.r. When necessary, the rotating inertia can be increased such that the flow coastdown following the drive motor trip becomes slower and the power/flow mismatch less severe. This analysis used only the rotating inertia available from the recirculation drive equipment. The minimum critical heat flux ratio (MCHFR) was analyzed, first: having a thermal limiting 7x7 fuel channel and second: having a thermal limit.., ing 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 70 was found to occur at about 2. 02 seconds .after the trip of both recirculation drive motors. 15
- 5. 2. 3. 1 Two Pump Trip (continued) With the thermal limiting 8x8 fuel channel a MCHFR of 1. 61 was 'found to occur at about 2. 33 seconds after the trip of both recirculation drive motors. Figure 5 *shows the results of a two pump trip. In both cases no damage to the fuel barrier occurs. No scram is initiated directly by the simultaneous pump trip and the power settles out at part load and natural cl.rculation .. c.onditi0ns. Nuclear* .system pr.es sure deer.eases throughout the transient such that the nuclear systen1 process barrier is not threatened by over pres sure. 5. 2. 3. 2 One Pump Trip Normal trip of one recirculation loop is accomplished through the drive motor breaker. However, a worse coastdown transient occurs if the generator field excitation breaker is opened, separating the pump and its motor from the inertia of the MG set. Results of this transient is shown in Figure 6. Diffuser flow on the tripped side reverse at about five seconds, however, M ratio (pump suction/drive flow ratio) in the active jet pumps increases greatly, producing about 142 per cent of normal diffuser flow and about 60 per cent of rated core flow. MCHFR was calculated the same way as in the two pump trip,* first with ... a thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With the thermal limiting 7x7 fuel channel a MCHFR of 1. 74 was found to occur at about 1. 59 seconds. While with the thermal limiting 8x8 fuel channel a MCHFR of 1. 67 was found to occur about 1. 59 seconds. Even for this case in which instantaneous coolapse of the generator field was assumed, no. damage to the fuel barrier occurs. 5. 2. 3. 3 Pump Seizure This case represents the instantaneous stoppage of one pump motor shaft. It produces the most rapid decrease of core . flow. The.reactor is assumed to be operating at maximum power conditions. Figure 7 shows the results of this transient. Note the fast decrease. in drive flow in the seized loop due to the large loss introduced by the stopped rotor. Jet pump diffuser flow on the seized loop reverse at about 800 milliseconds. Core flow reaches its minimum value in about 1. 2 seconds. MCHFR is again analyzed by, first with a* thermal limiting 7x7 fuel channel and second with a thermal limiting 8x8 fuel channel. With a thermal limiting 7x7 fuel channel a MCHFR of 1. 21 was found to occur at about 2. 00 seconds. With the thermal limiting 8x8 fuel channel a MCHFR of 1. 12 was found to occur at about 1. 90 seconds. Nucleate boiling is maintained since MCHFR is greater than 1. 0, fore,. no damage occurs to the fuel clad barrier. The initial 16 -**-
}* I I-' -...J 0 w I-a: a: u.. D I-z w u a: w a.. 150. 100. 50. 1 NEUTfltlN FLUX I 2 PEAK FUEL CENTI:R TEMP 3 AVE SURFACE Hl'm' FLUX II FEEOHATER FLCH VESSa STEAM F,LCH I I . .... -***-8. --*---. -. 12. **. **--*-***-.. .... 16. TIME tSECl. -so ............... ____ _..,...__ ____ .,..,_ _____ ._ ___ _ fl. 1,1. 8. 12. 16. TIME CSECl Figure 5. --TRIP CJF HJCJ DRIVE MCJTCJAS 100 PC PCJWEA 160.t------so_ * ._._ .................. _._._...,,._ ____ 1,1. 8. . 12. TIME tSECl 1 NEUTRdN FLUX 2 SURF HEAT. FUJX 120.1-------+-------+-------1--------....,---Cl w l-a: a: LL. D 1-z w 1,1(). u a: w 0.... o. 25. --*So. 75. 100. CCJAE FLCJW (/.) *-
v:-:* I-' CD 1 NEUTRCIN FLUX 2 PEAK FUEL CEN R TE!1P 3 RVE SUFf'ACE T FLUX ij FEEOa<lTEA fl ..................................... TIME (SECl ............................................. TIME * (SECl Cl LLJ >-a: 0: LL 0 >-z LLJ u 0: LLJ 0... Figure 6 --TRIP OF ONE GEN FIELD BREAKER 100 PC POWER -, ;* ...................................... ** TIME CSECJ 1 NEUT N FLUX 2 SURF HERT FLUX 80. ILO. I so. 75. CORE FLOW C/.l 100. 25. r Cl LLJ t-a: a: LL o. t-z LLJ L) a: LLJ 0.. ,,, ,1 l> I 100. 50. ................. _._.._._._,_...__ ____ ....,.... _____ ...__ ____ ob: 8. 12. 16. TIME (SEC) 100. 1 LEVEL IINCH-AE SEP-SKIRT 2 SENSED LEVEL INCHES! 3 TURBINE STEAM OH l7. I ij OH:: INLET Fl t % l DRIVE FLOH 1 7. l o.1--------+-------1-------+------1-----TIME CSEC) Figure 7 -..; ONE PUMP SEIZURE 100 PC PC!WER .. **;-........................................ TIME (SECl 1N£ Nl'l.UX 2 SURF HEAT FLUX 120.1-------1-------1-------1-------1----'---Cl 80. LLJ t-a: a: LL 0 t-z w ij(). L) a: LLJ 0.. CC!RE FLC!W C/.l '\" t:* : I' I r:* / .. ; *r, .t' 1, 1. -* r 5.2.3.3 Pump Seizure (continued) pressure regulator maintains pressure control as the reactor settles out at the final lower power condition. No scram occurs. Because the nuclear system pressure decreases throughout the transiEn t the nuclear system process barrier is not threatened by over pressure. 5. 3 "B" Scram Reactivity Curve Analyses A ssuniing the previously described changes (Section 4. 0), the transient analyses verifies operation at 100% of licensed power with the "B" scram reactivity curve. An acceptable pressure margin ( 25 psi) between the peak pressure at the turbine trip without bypass with trip scram transient and the first safety valve setpoint is maintained. The Target Rock safety/relief valve plus the eight spring safety valves at the new higher setpoints were shown to provide adequate overpressure protection in accordance with ASME Code requirements. 5. 3. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient The relief valve sizing transients were reanalyzed at the exposed core conditions to verify the adequacy of the four electromatic relief valves plus one Target Rock safety I relief valve (at reduced flow) to terminate the pressure transient when the reactor is subjected to a rapid pressurization evert such that satisfactory margins are maintained between the peak pressure resulting and the first safety valve setpoint. The event analyzed is the turbine trip with simultaneous reactor scram but with a failure of the turbine bypass system. The results for this transient are illustrated by Figure 8. Initial conditions prio°r to the event a re reactor power at the 100 per cent level, core recirculation flow at 98 million lb. per hour, and reactor steam dome pressure at 1005 psig. The first relief valve setpoint is 1125psig and the remaining four valves have setpoints of 1130.and 1135 psig, two valves at each setpoint. The sudden closure of the turbine stop valves with no initial bypass flow causes a rapid rise in system pressure at a rate essentially double that which results when the bypass system functions. Position switches on the stop valves initiate immediate reactor scram. The rapid pressurization causes core void collapse and neutron flux increases, pea king at a bout 143 per cent of the initial value before the scram becomes effective. Co re average surface heat flux dips initially and then increases to slightly over 100 per cent of its initial value at about L 36 seconds. No significant 20 a w I-a: a: LL. D I-:z: w Ll a: w CL I\) ....... 1 VESSEL PRES CPSil 2 snt LINE PRESf\ISE \RSI! 3 TUH81KE PRES RiSE \PSIJ ll BrPf;SS v;1LVE FUll1 l7.l, 350
- 6 5GFEll' VMLVE l'lM \7.J 100.' so. o. o. !!. e. TIME 100. 16. en a: a: ....J ....J D D 200. -1oob. I!. s. TIME lSECl Figure 8 Turbine Trip Without Bypass 6. TIME 12. 16. lSECl 1 NET AE?.CTIVIrGrr
- 2 SCRfil'1 RERCT!V ll 3 rol'Pl..ER ll VO!O AEflC VI 1 6. 12. 16. TIME CSECl l l l I r
- 5. 3. 1 Turbine Trip Without Bypass -Relief Valve /\dequacy Transient (continued) 5. 3. 2 reduction in MCHFR occurs as the core recirculation flow increases due to the reduction in core voids resulting from the pressure increase and the reactor scram and the heat flux increase is small. Reactor pressure rises to the setpoint of the first relief valve in_a.bo.ut .1.. 8 _s_e.conds a.nd _a.11 relief .v.alve.s ar.e tripped by 2. 7 seconds. The peak pressure at the location of the safety valves is 1185 psig which is 5 5 psi below the 1240 psig setpoint of the first two safety valves. Ma in Steam Isolation Valve Closure with Indirect Scram -Safety Valve Adequacy Transient Safety valve sizing transients were also reanalyzed at the exposed core conditions to verify that the eight spring safety valves plus one Target Rock valve (at reduced flow) will provide sufficient margin to the ASME Code limit when the reactor is subjected to its worse main steam isolation transient. Both the turbine trip without bypass and the main steam isolation valve closure events result in large increases in reactor pressure. Recent analyses have determined that when direct reactor scrarn.s are ignored, i.e., those originating from position switches on the stop and isolation valves, slightly higher pressure peaks result from the closure of all main steam isolation valves. The transient analyzed to verify satisfactory margins to the ASME Code limit is_ thus the three second closure of all main steam isolation valves. -It is assumed that (a) the reactor is at full power when the steam line isolation occurs, (b) the relief valves fail to open, (c) direct reactor scram from the position switches fails, (d) the backup scram due to high neutron flux shuts down the reactor. Figure 9 shows the results of this transient. As the steam flow vs. position characteristics of the isolation valves is non-linear, the effects of closing the main stea1n isolation valves are not immediate. However; by about 1. 5 seconds after the valves start to close, pressure has increased to the point where core void collapse results in a significant increase in neutron flux. Flux scram occurs at about 1. 8 seconds and limits the neutron flux peak to approximately 450 per cent of initial level. Steamline pressure reaches the Target Rock safety/relief valve setpoint of 1125 psig at about 2. 9 seconds. The first spring safety valve setpoint of 1240 psig is reached at about 4. 7 seconds. Vessel dome pressure reaches a peak of !l_68 psig and the peak pressure at the bottom of the vessel is 1297 psig which is 78 psi below the peak pressure allowed by the ASME Section III Code, *(110% of vessel design pressure of 1250 psig or 1375 psig). 22 0 w I-a: a: LL 0 I-:z: w u a: w a_ 1 NEUTRON FLUX 2 PEAK FUEL CEN A TEMP 3 AVE SUAFRCE H-T FLUX q FEEOi'ATEfl FLO 100. so. 50. o. TIME CSECJ Figure 9 350. . h 1 VESS::l. PRES A CPS!l . 2 SHI PRES CPS!J 3 TUR6!NE PRES AiS8 !P$!J BIPP.SS VALVE f!_Oil1 C7.J !:> REUEF VALVE f c;:1 6 SAFE.TT VALVE f (j)ll 17.l , I 200. I h-.. . -0 w I-a: a: . LL 0 I-:z: w u a: w a_ so. >---. 120. 80. iw. o. o. / I\ o;. 2. 1 FLUX 2 SUAFA ':E HERT FLUX I 25. MSIV CLCJSURE-FLUX SCRRK -* ----' IR q., 6. 8
- TIME CSECJ I I f li?' ) so. 75. CCJRE FLCJW (/.) / /. 1-1* ., :=;
--.. --------. *-. .:.,. ..... ---. .,, . ----* ...... _ -.: ..... _,__ ..... ;,,.:.-.':' ... 5. 3. 3 Nominal Turbine
- A turbine trip is the primary turbine protection mechanism and is initiated whenever various turbine or reactor system malfunctions occur which may threaten turbine operation. turbine trip initiates fast closure (approximately 0. 1 s_ec_0nd) .: of the turbine stop valves. Reactor scram is initiated imr:iediately from ** ** **position sv1itches mounted on the turbine stop valves. Figure 10 illustrates the results of this tr.ansient when occurring .at the maximum reactor power level ( 100 per cent of licensed). The sudden closure of the stop valves causes a rapid pressurization of the steam line and reactor vessel with subsequent void collapse and a small reactor power increase. Because of the fast action of the turbine trip scram, the neutron flux peak is held to 120 per cent of initial value. Core average surface heat flux decreases out the transient and no decrease in MCHFR occurs.. As the pass valves open almost immediately, the vessel pressure increase rate is limited to less than 50 psi/second. The action of the bypass system alone limited the peak pressure rise in the steam line to 1098 psig which is 142 psi below the first spring safety valve setpoint. Vessel dome pressure. peaks at a bout 1100 psig. 5. 3. 4 Loss of Generator Load Loss of generator load is quickly sensed by the power-load unbalance circuitry in the Turbine Electro-hydraulic Control (EHC) system. This circuitry energizes the fast-acting solenoid operated valves which open disk dump valves on the turbine control valves such that the control valves close rapidly and excessive turbine-generator overspeed is prevented. Bypass valves are opened rapidly when the load demand is stepped to zero and reactor pressure control is transferred from the control valves to the bypass valves. Reactor scram is initiated immediately from position switches mounted on the fast-acting solenoid valves. As the closure of-the control valves is almost as fast as the closure of the turbine stop valves, the transient is almost jdentical to that of the Turbine Trip tra nsi.ent described in 5. 3. 3 but less severe. 5. 3. 5 Loss of Main Condenser Vacuum If condenser vacuum is lost while the unit is in operating the following trips will occur: 24 0 UJ I-a: a:: LL. 0 I-z: UJ so. u a:: UJ a.. r(, ob. it. \Jl 16. 16. 8. 12. TIME CSECl a. , 12. TIME CSECl " so. o. Cf) a:: a: _I _I 0 D o. -s. ...................................... a. 12. 16. TIME CSECl TIME CSECl Figure 10 Turbine Trip With Bypass
- 5. 3. 5 Loss of Ma in Condenser Va cu um (continued) Alarm at 2411 Hg vacuum Scram at 22 Hg vacuum Turbine Trip at 2011 Hg vacuum Closure of Bypass Valve at 711 Hg vacuum. The worse case for this type of event would be instantaneous loss of vacuum with the unit operating at 100 per cent of licen,sed ,power. The transient for this case becomes identical to the turbine trip without bypass transient described in 5. 3. 1. Slower losses *of condenser vacm;im will produce less severe transients because the scram will precede the stop valve *closure resulting from the Turbine Trip and some bypass flow will be permitted to remove stored heat. 5. 3. 6 Main Steam Isolation Valve Closure -Position Scram The closing time for the main steam isolation valves can be as fast as 3 seconds. The effects on the reactor are not immediate as the flow characteristics of the valves are linear; however, a significant vessel pressure rise results when the reactor is suddently cut off from its primary heat sink. Position switches on the isolation valves initiate reactor scram before the valves reach the 10 per cent closed position. The transient resulting from inadvertent closure of all main steam isolation valves from an initial power level of 100 per cent of licensed is illustrated in Figure 11. There is no increase in neutron flux or core average surface heat flux as the position scram becomes effective before significant .valve area reduction occurs. MCHFR therefore increases throughout the transient. Core pressure does not rise significantly until about 1. 5 seconds after the isolation valves start to close. By 2. 5 seconds the pressure rise rate is approximately 60 psi/sec and pressure reaches the first relief valve setpoint at about 3. 6 seconds. The relief.valve flow removes the excess stored heat and holds peak steamline pressure to 115.5 psig which is 8 5 psi below the setpoint of the first safety valve. Vessel pressure peaks at 1155 psig. The isolation condenser will be initiated after about 15 seconds, time delay required for actuation, to handle the long-term decay heat removal. 26 CJ w r-a: a: l.J... 0 I-z w u a: w CL N '1 100. 50. ob. 12. 16. CSECJ 100. so. en a: a: ..J ..J D CJ o. . . TIME CSECJ 1 VESSEL Pf\ES R! !PSIJ 2 STK LINE P!'GS !SE lPSil 3 T'JriBH,'f: Prr.:S !SE tPSil
- ij Sl1'flSS Vl'l.'fE fl.J3H l7.J 6 ::f!FETY VI'!.. YE FLOM lZl 200. *:;. so. ij. 8. 12. 15; TIME CSECJ 1 tfl FF....Acrrvrnl 2 SCM11 REHCT H UTY 3 DOi'!'lER f\:JlCT 5. Ii VOID !YI 0. -5
- 5. 4 ECC Scram Reactivity Curve Analysis Assuming the previously described changes 0), the transient analyses verifies operation at 93% of licensed power with the EOC scram reactivity curve. An acceptable pressure . margin ( 25 psi) between the peak pressure of the turbine trip without bypass with trip scram transient and the first safety valve setpoint is maintained. The Rock safety/relief valve plus the eight spring safety valves at the new higher setpoints were shown to provide adequate overpressure protection in accordance with ASME code ments for operation at the 93% power level. All of the transien.ts presented in this section were evaluated at the 93% power level. 5. 4. 1 Turbine Trip Without Bypass -Relief Valve Adequacy Transient The relief valve sizing transients were reanalysed with the design basis scram conditions to verify the adequacy of the four electromatic relief valves plus one Target Rock safety/relief valve (at reduced flow) to terminate the pressure transient when the reactor is subjected to a rapid pressurization event such that satisfactory margin.s are maintained between the peak pressure resulting and the first safety valve setpoint. The event analysed is the turbine trip with taneous reactor scram but with a failure of the turbine bypass system. The results for this transient are illustrated by Figure 12. Initial conditions prior to the event are reactor power at 93% of the license rated power level, core recirculation flow at 98 million lb per hour, and reactor steam dome pressure at 1005 psig. The . first relief valve setpoint is 1125 psig and the remaining four valves have setpoin_ts of 1130. and 1135 psig, two valves at each setpoint. The sudden closure of the turbine stop valves with no initial bypass flow causes a rapid rise in system pressure at a rate essentially double that which results when the bypass systerri functions. Position switches on the stop valves initiate-immediate reactor scram. The rapid pressurization causes core void collapse and neutron flux increases, peaking at about 176 percent of the initial value before the scram becomes effective. Core average surface heat flux dips initially and then increases to slightly over 99 percent of its initial value at about 1. 7 sec ands. No significant reduction in MCHFR occurs as the core recirculation flow increases due to the reduction in core voids resulting from the pressure increase and the reactor scram and the heat flux increase is small.
! N l.O 0 UJ f-LL D I-:z: UJ u a: w 0... 100. so.H Q.. 3. B. 1IME lSECJ .................................... 1IME lSECJ Figure 12 .350. I l VESSEL Rik. ll'S rJ 2 STM PRES FHSE ll'S!l 3 TU19 W: PR:S FITSE tPS Il 3Y!'P:3S V'i!.. 'IE f1..Cl-I L':l FE.IE" -Ci<z.,-,l=7.1=----6 SllFET'f m_ VE OH (7.J * . ......................... 0 UJ ...... IL 0 I-:z: UJ <....) a: UJ 0... -.J 120. BO. >-,_. 1--o. o. l 2 SJR!'A! iE HEAT R.UX r 25. 1URB TRIP NCT BYPASS TRIP SCRAM 1IME lSECJ !)(). 75. 100. CORE FLOW (/.J / i' I I . r I '. I I l I l I I ! i ! i f. i I;,
.* ., ' .Reactor pressure rises to the setpoint of the first relief valve in about 1. seconds and all relief valves are tripped by 2. 1 seconds. The peak pressure at the location of the safety valves is 1198 psig which is 42 psi below the 1240 psig setpoint of the first two safety valves. 5. 4. 2 Main Steam Isolation Valve Closure With Indirect Scram -Safety Valve Adequacy Transient Safety valve sizing transients were also reanalysed at the design basis scram conditions to verify that the eight spring safety valves plus the Target Rock Valve (at reduced flow) will provide sufficient margin to the ASME Code limit when the reactor is subjected to its worse main stream isolation transient. Both the turbine trip without bypass and the main steam isolation valve closure events result in large increases in reactor pressure. Recent analyses have determined that when direct reactor scrams are ignored, i.e., those originating from position switches on the stop and isolation valves, slightly higher pressure peaks result from. the closure of all main steam isolation valves. The transient analysed to verify satisfactory mar to the ASME Code limit is thus the three second closure of all main steam isolation valves. It is assumed that (a) the reactor is at 93% of full power when the.* steam line isolation occurs, (b) the relief valves fail to open, (c) direct reactor scram from the position switch fails, (d) the backup scram due to high neutron flux shuts down the reactor*. Figure 13 shows the results of this transient. As the steam flow vs. position characteristic of the isolation *valv.es is non-linear, the effects of closing the main steam isolation . valves are not immediate. However, by 1. 5 seconds after the *valves start to close, pressure has increased to the point where core void collapse results in a significant increase in neutron flux. Flux scram occurs at about 1. 8 seconds and limits the neutron flux peak to approximately 400 percent of initial level. Streamline sure reaches the Target Rock safety I relief valve setpoint of 1125 psig at about 2. 99 seconds. The first spring . safety valve setpoint of 1240 psig is reached at about 4. 73 seconds. Vessel dome pressure reaches a peak of 127 7 psig and the peak pressure at.the; bottom of the vessel is approximately 1301 psig which is 74 psi below the peak pressure allowed by the ASME Section III Code, (110% of vessel design pressure of 1250 psig or 1375 psig) . . 5. 4. 3 Nominal Turbine Trip A turbine trip is the primary turbine protection mechanism and 30
- . : l.50. C) lOO. UJ ...... LL 0 ...... z so. a: UJ Q.. I s -. ob. G. 6. b. 11. e. 6. ; w lSECJ l IME lSECJ I-' 1 2. SU<'FfN: HEAT R.UX 120. ! so. C) 60. lLl ....... a: a:: lL 0 ....... Q z o. UJ lo(). (..) a:.: UJ 0.. -J r . r ' \I. . e. 8. 25. 75. 100. 1IME lSEC) CORE FLClr.I Ci:'.J Figure 13 MSIV CLOSURE-FLUX SCRRM*
' * .. .. -i.1 '< .. ....... f\'4W:" ** '-__ l -'*: .. :'.:.: ........ .;._.4'-_.:,: is whenever various turbine or reactor system functions occur which may threaten turbine operation. The turbine trip initiates fast 0. 1 second) *.**.* .-..
- of the turbine stop valves. Reactor scram is initiated from position switches mounted on the turbine stop valves. Figure 14 illustrates the results of this transient when occuring at 93% the maximum reactor power level. The sudden closure of the stop valves causes a rapid pressurization of -the .. s.te.am.line.and rea.c.to.r v.essel w.ith subsequent void collapse and a small reactor power increase. Because of the fast action of the turbine trip scram, the neutron flux peak is held to 130% of initial value. Core average surface heat flux peaks at 95% of initial value then decreases throughout the transient and no decrease in MCHFR occurs. As the bypass valves open almost immediately, the vessel pressure increase rate is limited to less than 50 psi/second. The action of the bypass system alone limited the peak pressure rise in the streamline to 1120 psig which is 120 psi below the first spring safaty valve setpoint. Peak pressure in the vessel is about 1127 psig. 5. 4. 4 Loss of Generator Load Loss of generator load is quickly sensed by the power-load unbalance circuitry in the Turbine Electro-hydraulic Control (EHC) system. This circuitry energizes the fast-acting solenoid ope;rated valves which open disk dump valves on the turbine control valves such that the control valves close rapidly and excessive *turbine-generator overspeed is prevented. Bypass valves are opened rapidly when the load demand is stepped to zero and reactor pressure control is transferred from the control valves to the bypass valves. Reactor scram is initiated iIT1mediately from position switches mounted on the fast-acting solenoid valves. As the closure of the control valves is almost as fast as the closure of the turbine stop valves, the transient is almost identical to that of the Turbine Trip transient described in 5. 4., 3 but less severe. 5. 4. 5 Loss of Main Condensor Vacuum If condenser vacuum is lost while the unit is m operation the following trips will occur:
l I W w u.. 0 :z a:: LLJ Cl . ..J . lo. G. 6. 1 IME (SECl -robL
- 1IME (SECl l VfSSEL PFES Rb: tPSTJ 2. S TH Lr1E P!'1'.S nTSE tPSTJ 3 Tl.PilihE PRES !'\ iSC lPS !J ... ----** ... : 35().1-------+------+-------1....-li._,O!'=.PSSS,-:, .. 'IF.=.,....,!"=-,..=c;.;.,..* =c.::,,...J __ \lf!..'-'F. FUli t..'i;J G SR'Elr' 'R.. \IE . Gi L".J * ........... .._._ ..................... TIME CSECl 120. 0 60. UJ t--LL D :z llO. l.IJ 0... ._, fo f 2 SUl'l'R FLUX -t5!T FLUX r. 25. K; .. ) &O. , 75. 100. CORE FLOrl (/.) -**-* Figure 14 --TURBTRIP BYPASS TRIP SCRAM-1* /. f
- I
- I f
..... I Alarm at 2411 Hg vacuum 9cram at 2211 Hg vacuum Turbine Trip at 2011 Hg vacuum Closure of Bypass Valves at 7.11 Hg vacuum The worse case for this type of event would be instantaneous loss of vacuum with the unit operating at full The transient for this case becomes identical to the turbine trip without bypass transient de'scribed in 5. 4. 1. Slower losses of condensor vacuum will produce less severe transients because the scram will proceed 'the srop valve *closure *:re*sulting.*'from the Tur*bine Trip and some bypass flow will be permitted to remove stored heat. 5. 4. 6 Main Steam Isolation Valve Closure -Position Scram **The closing time for the main steam isolation valves can be as fast as 3 seconds. The effects on the reactor are not immediate as the flow characteristics of the valves are non-linear; however, a significant vessel pressure rise results when the reactor is suddenly cut off from its primary heat sink. Position switches on the isolation valves initiate reactor scram before the valves reach the 10 percent closed position. The transient resulting from inadvertant closure of all inain steam isolation valves from an initial power level of of full power is illustrated in Figure 15. There is no increase in neutron flux or core average surfact heat flux as the position scram becomes effective before significant valve area reduction occurs. MCHFR therefore increases out the transient. Core pres sure does not rise significantly until 'about 2. 5 seconds after the isolation valves start to close. By 3. 5 *seconds the pres sure rise rate is approximately 60 psi/ sec and . pressure reaches the first relief valve setpoint at about 4. 06 seconds. The relief valve flow removes the excess stored heat and holds peak streamline pressure to 1160 psig which is 80 psi below the setpoint of the first safety valve. Vessel pressure peaks at 1164 psig. isolation condenser will be initiated to handle the long-term delay heat removal. 34 Cl 100. UJ .. -.. :* I-a: er: *LL 0 I-z UJ so. u er: w <.:; ob. '** 1IME 1 IME (SECJ 1l.Jl l.n' 1 t-elITFi FLUX 2 SUR!'R HEAT FLUX 100. 120. so. 80. 0 UJ I-a: er: LL D I-z: o. UJ u er: \.....S I I w a_ ..... r . -25. so. 75. 100. TIME lSECJ CORE FLOW (/.l Figure 15 --MSIV CLCJSURE-POS. SCRRM
- 6. 0 TECHNICAL SPECIFICATION CHANGES
- 6. 1 Scope of Changes The principal changes of interest concern the improved control rod scram times and the increased safety valve setpoints. These changes are needed to be sistent with the new assumptions used in the transient re-analyses and are discussed in detail in Section 4. 0 above. Other changes are those associated with the *r*esults *of the transient *r-e*-analyse*s *di*s cus s,ed in Section 5. 0 above. The locations of the changes are presented *as D3/QC-l for application to the respective documents. 6. 2 Specific Changes Item Location Basis statement for 2. 1 Basis statement for 2. l.E &F Bottom of page 12I14 End of para E and F *. page 18 /21 36 Change Add a reference to this report in footnote ( 3) Add a reference this report Reason The. statements a re supported by the analyses. The statements are supported by the transient re-analyses
-.:. . . . . -. . .. *-,,i*,t ' -k#'bL -****J\' ,..,.,. ?ft,j:"'*1****;.,.( q,1,)*.J1r3 6. 2 Specific Changes (continued) Item Basis statement for 2. 1. I Location Page 18A I 21 Change In the sixth tence take out the words pres sure or" Reason The re-analysis shows that a significant pressure increase will occur for this transient. Limiting system 2.2.B Page 19 /24 Change pressure .setpoints to 1240, 1250, 1260 and *---***----*--* -**----ig,*-The transient analyses were done with these new points. :*-.---*-*-* *-* -**. -* -----***-'Add to 1 ist of ---*-----*--*--_____ ; This accounts for the use of the Target Rock valve as a safety Basis statement for 1 .2 Basis statement for 1. 2 Basis statement for 2. 2 Para 5 on Page 20/25 Refs. at bottom of page 20/25 Third tence on . page 21 /26 Pa.ge 21/ 26 37 '.'I valve at 1125 psig" At the bottom of the page add: "Target Rock combination safety/ relief valve" The fourth sen-* tence change to read : ". . limit the reactor pressure to a value ( 5) which is 25 psi below II .... Add appropriate sections from this _report Add appropriate sections from this report In the fifth sentence change 1240 and 1280 to 1163 and tively. This is in conjunction. with the new assumptions. for, and results of the transient re-analyses. The re-analyses need to *be referred to. The re-analyses need to be referred to. This is result of the new analysis using the Target Rock valve and the increased spring -safety valve setpoints.
.. , .. ,. 6. 2 Specific Changes (continued) Item Basis statement for 2.2 Spec. 3. 3. c. 1 Basis *statement 3. 3. c. Spec. 3. 5. D. 2 & 3. 5. D. 3 Spec. 4. 5. D. 3 Spec. 4.6.E Location Page 21/ 26 Change In the eighth sen-tence change 1305 and 1330 to <. 1277. and Reason ** -< '.l-3*0 l --re-s*pectiveJ.:y Tables on Change the two The transient re-analyses page 58 / tables, in sequence, were done with these new 75-7b to the following: scram time requirements. % Inserted From Fully Average Scram Insertion* Withdrawn Time (secs) 5 20 50 90 0. 375 0 .. 900 2.00 3. 50 % Inserted From Fully Withdrawn Average Scram Insertion -------( s_e_c_s...:..) __ _ 4th, 5th, 6th tences on pages 63 and 64/83 Page 78/ 100-101 5 20 50 90 Change ences from * . Fig. 3. 5. 2 of the SAR to Fig. 1 of this report -Delete the word electromatic. Same as above. Page 90/ 119 Change safety valve setpoints to 1240, 1250, 1260 and 1260 -psig. r:*. 0. 398 0.954 2. 120 3.80 A new scram activity insertion curve was used in the transient analyses. The re-analysis was done with 5 relief valves; one Target Rock and 4 electromatics. The re-analysis was done with these new setpoints.
6 .* 2 Specific Changes (continued) . Item Spec. 4.6.E Location*. Page 90/ 119 Change Reason Add the fol-This includes lowing to the the use of the safety valve Target Rock valve list: the re-analysis. Number of valves: l* Setpoint {psig); 1125 the value 1 under number of 'valves in the relief valve list. At the bottom of the page add: "Target Rock combination safety/ relief valve .... 39 in
- 7. 0 iEFERENCES 7. 1 Dresden Station Special Report No. 29, AEC Dkt. 50-249, July 2, 1973.
- 7. 2 J.S. Abel letter to D. J. Skovholt, "Scram Reactivity Limitations for Dresden Units 2 and 3 and Quad Cities Units 1 and 2 -AEC Dkts. 50-237, 50-249, 50-254 and 50-265, 11 October 18, 1973. *7. 3 Dresden Station Special Report No. 29, -Supplement A, A'EC Dkt '50 ..;.2t49, *F*ehrua-ry '4, l9 7*4. 40