ML19208D738
| ML19208D738 | |
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|---|---|
| Site: | Crystal River  |
| Issue date: | 08/24/1979 |
| From: | FLORIDA POWER CORP. |
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| NUDOCS 7909290456 | |
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Text
4 e
9 ANALYSIS
SUMMARY
IN SUPPORT OF AN EARLY RC PUMP TRIP e
9 e
e 9
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CONTENTS Page I.
INTRODUCTION.
1 II.
SMALL BREAK ANALYSIS.
2 A.
Introduction.
2 B.
System Response With RC Pumps Running 2
C.
Analysis Applicability to Davis-Besse 1 11 D.
Effect of Prompt RC Pump Trip on Low Pressure ESFAS Signal.
13 E.
Conclusions 13 e
III.
IMPACT ASSESS!ENT OF A RC PUMP TRIP ON NON-LOCA EVENTS.
15 A.
Introduction.
15 B.
General Assessment of Pump Trip in Non-LOCA Events.
15 C.
Analysis of Concerns and Results.
16 D.
Conclusions and Summary 18
ANALYSIS SU> DIARY IN SUPPORT OF AN EARLY RC PUMP TRIP I.
INTRODUCTION B&W has evaluated the effect of a delayed RC pump trip during the course of small loss-of-coolant accidents and has found that an early trip of the RC pumps is required to show conformance to 10CFR50.46. A summary of the LOCA analyses performed to date is provided in Section II.
This discussion includes:
1.
A description of the models utilized.
2.
Break spectrum results with continuous RC Pump Operation.
3.
Break spectrum results with delayed RC pump trips including estimates of peak cladding temperatures.
4.
Justification that a prompt pump trip following ESFAS actuation on low RC pressure provides LOCA mitigation.
An impact assessment of the required pump trip on non-LOCA events has also been completed and is presented in Section III. This evaluation supports the use of a pump trip following ESEAS actuation for LOCA mitigation since no detrimental consequences on non-LOCA events were identified.
en m
II.
SMALL BREAK ANALYSES A.
Introduction Previous small break analyses have been performed assuming a loss-of-offsite power (reactor coolant pump coastdown) coincident with re-actor trip.
These analyses support the conclusion that an early RC pump trip for a LOCA is a saic condition. However, a concern has been identified regarding the consequences of a small break transient in which the RC pumps remain operative for some time period and then are lost by some means (operator action, loss-of-offsite power, equipment failure, etc.).
This section contains the results of a study to further understand how the small break LOCA transient evolves with the RC pumps operative. Cpecifically, section B.
describes the system response with the RC pumps running for B&W's 177-EA lowered-loop plants.
In-cluded in this section is the development of the model used for the analysis, a break spectrum sensitivity study, and peak cladding tem-perature assessments for cases where the RC pumps trip at the worst time.
6 Section An1I demonstrates the applicability of the conclusions 6
drawn in section ATT to a 177-EA raised-loop plant (Davis-Besse 1).
The effect of a prompt tripping of tue RC pumps upon receipt of a o
lowprgssureESEASsignalisdiscussedinsectionWYs7.
Finally, sec-tion Ac5Fsummarizes the conclusions of this analysis.
B.
System Response With RC Pumps Running 1.
Introduction Recent evaluations have been performed to examine the primary system response during small breaks with the RC pumps operative.
During the transient with the RC pumps available, the forced circulation of reactor coolant will maintain the core at or near the saturated fluid temperature. However, for a range of break sizes, the reactor coolant system (RCS) will evolve to high void fractions due to the slow system depressurization and the high liquid (low quality fluid) discharge through the break es a re-sult of the forced circulation.
In fact, the RCS void fraction will increase to a value in excess of 90% in the short term.
In 1054 06 k.
the long term, the system void fraction will decrease as the RCS RCS depressurizes, HPI flow increases, and decay heat diminishes.
With the RCS at a high void fraction, if all RC; pumps are postu-lated to trip, the forced circulation vill no Jpnger be available and the residual liquid would not be sufficient to keep the cot.
covered. A cladding temperature excursion would ensue until core cooling is reestablished by the ECC systems.
The following para-graphs summarizes the results of the analyses which were performed for the 177-FA lowered-loop plants, to develop the consequences of this transient.
2.
Method of Analysis The analysis method used for this evaluation is basically that de-scribed in section 5 of BAW-10104, Rev. 3, "B&W's ECCS Evaluation Model"I and the letter J.H. Taylor (B&W) to S. A. Varga (NRC), dated 2
July 18, 1978, which is applicabic to the 177-FA lowered-loop plants for power levels up to 2772 MWt.
The analysis uses the CRAFT 23 code to develop the history of the RCS hydrodynamics.
However, the CRAFT 2 model used for this study is a modification of the small break evaluation model described in the above ref-er9nces.
Figure 2-1 shows the CRAFT 2 noding diagram for small breaks from the above referenced letter. The modified CRAFT 2 model consists of 4 nodes to simulate the primary side,1 node for the secondary side of the steam generator, and 1 node representing the reactor building.
Figurc 2-2 shows a schematic diagram of this rodel. Node 1 contains the cc3d leg pump discharge piping, downcomer, and lower plenum. Node 2 is the primary side of the SG and the pump suction piping. Node 3 contains the core, upper ple-num, and the hot legs. Node 4 is the pressurizer and nodes 5 and 6 represent the reactor building and the SG secondary side, re-spectively. This 6 node model is highly simplified compared to those utilized in past ECCS analyses.
It does, however, maintain RCS volume and elevation relationships which are important to properly evaluate the system response during a small break with the RC pumps running.
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The breaks analyzed in this section are assumed to be located in the cold leg piping between the reactor coolant pump discharge and the reactor vessel.
Section B.7 demonstrates that this is the worst break location.
Key assumptions which differ from those de-scribed in the July 18, 1978, letter are those cdncerning the equip-ment availability and phase separation.
These are discussed below.
a.
Equipment Availability Tha analyses which were performed assumed that the RC pumps re-main operative after the reactor trips.
For select cases, after the system has evolved to high void fractions (approxi-mately 90%) the RC pumps were assumed to trip. Also, the im-pact of 1 versus 2 HPI systems for pump injection were examined.
The majority of the analyses performed assumed 2 HPI pumps.
However, as is demonstrated later, even with 2 HPI pumps avail-able, cladding temperatures will exceed the criteria of 10 CFR 50.46 using Appendix K evaluation techniques.
Therefore, fur-ther analysis with only 1 HPI pump would only be academic, b.
Phase Separation The present ECCS evaluation model created to evaluate small breaks without RC pumps operative,(quiescent RCS) uti-lizes the Wilson 4 bubble rise correlation for all primary sys-tem control volumes in the CRAFT evaluation.
In this analysis, for the time period that the RC pumps are operative, the pri-mary system coolant is assumed to be homogeneous, i.e., no phase separation in the system.
In reality, the flow rates in the core and hot legs are low enough that slip will occur.
This will cause an increased liquid inventory in the reactor vessel compared to that calculated with the homogeneous model.
With the homogeneous assumption, core fluid is continuously circulated throughout the primary system and a portion of that fluid is lost via the break.
During the later stages of the transient, e slip model will result in fluid being trapped in the reactor vessel and the hot legs. The only method of losing liquid during this period will be by boiling caused by the core decay heat.
Thus, the assumption of homogenicty for the period with the RC purps operative is conservative.
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Following tripping of the RC put.ps and the subsequent loss-of-oo forced circulation, the system will collapse and separate.
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The residual liquid will then collect in the reactor vessel and g 1_
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the loop seal in the cold leg suction piping, For this period of the transient, the Wilson bubble rise model is utilized.
The homogeneous assumption for the period with the RC pumps operating applies to nodes 1, 2, and 3 in the CRAFT model.
Node 4, the pressurizer, and node 6, the secondary side of the steam generators, utilize the Wilson bubble rise model throughout the transient as these nodes are not in the direct path of the forced circulation.
3.
Benchmarking of the 6 Node CRAFT Model Studies were performed to compare the results of the 6 node model to the more extensive evaluation model for B&W's 177-FA lowered-loop plants as described in the letter J.H. Taylor (B&W) to S.A.
Varga (NRC), dated July 18, 1978.
The break size selected for 2
this comparison is a 0.025 ft break at pump discharge.
This break represents the largest single-ended rupture of a high energy line (2-1/2 inch sch 160 pipe) on the operating plat.s.
The break can be viewed as " realistic" or the worst that would be ex-pected on a real plant. Figures 2-3 and 2-4 are the results of this comparison.
System pressure and percent void fraction shown in Figures 2-3 and 2-4, respectively, compare very well with those from the more extensive (23 nodes) CRAFT 2 small break model. As seen in these figures, the difference is not significant and is less than a few percent. The computer time for this 6 node model is, however, significantly decreased. The model utilized for this study is thus justified based on comparison of results to the more extensive small break model and desirable because of its economical run time.
4.
Analysis Results The break sizes examined for this analysis rarged from 0.025 ft2 to 0.2 ft2 in area and are located in the pung discharge piping.
Breaks of this size do not result in a rapid system depressuri-zation and rely predominantly upon the HPIs fer mitigation.
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Table 2-1 summarizes the analyses performed for thi. evaluatien.
The majority of the analyses performed utilized 2 HPI pumps through-out the transient. The effect of utilizing 1 HPI pump is discussed in this section.
Figures 2-5 and 2-6 show the system pressure and average system void fraction transients for the break spectrum analyzed assuming continuous RC pump operation and 2 HPI's available.
In Figure 2-6, the average system void fraction is defined as V
-V 1
2 Average system void, % =
x 100 y
1 V = total primary liquid volume excluding the pressuri-3 zer at time = 0, V = total primary liquid volume excluding the pressuri-2 zer at time = t.
This parameter was utilized in place of the mixture height in that the coolant will tend to be homogeneously mixed with the RC pumps operative. Under these assumptions, the core is cooled by forced circulation of two-phase fluid and not by pool boiling as in the case where the RC pumps are not running and separation of steam and water occurs. As shown in Figure 2-5, the system pressure re-sponse is basically independent of breck size during the first several hundred seconds into the transient. This occurs because the forced circulation of reactor coolant maintains adequate heat transfer in the steam generators; the primary system thus depres-surizes to a pressure (about 1100 psia) corresponding to the sec-ondary control pressure (i.e., set pressure of SG safety relief valves). After some time (250 seconds for the 0.1 ft2 break), the system pressure will decrease as the break alone relieves the core energy.
Figure 2-6 shows the evolution of the system void fraction; values in excess of 90% are predicted very early (300 seconds) into the transient.
For the larger breaks the system high void fractions occur early in time.
For the smaller breaks it takes in the order of hours before the system evolves to high void fraction.
Core cooling is maintt'.ned during a small break with continuous RC pump I-1054 06(0
operatien regardless of void fraction.
In the long term, the sys-tem vili depressurft and the enhanced performance of the ECCS (RPI and LPI) will result in reduced system void fraction.
Figure 2-7 illustrates this long term system behavior for a 0.10 ft2 break.
For this case, the LPIS are operative at approximately 2300 seconds, and a substantial decrease in system void fraction results.
An arbitrary pump trip after approximately 2700 seconds would not result in core uncovery.
The potential for core uncovery due to an RC pump trip is thus limited to a discrete time period during which the natural evolution of the system produces high void frac-tions and prior to LPI actuation.
For a 0.1 ft2 break, this time period is on the order of 2000 seconds.
For smaller breaks, this critical time could be a few hours even if the operator initiated a controlled cooldown and system depressurization as recommended in the small break guidelines.
Although the analyses described above used 2 HPI pumps, the effect of only 1 HPI pump available on the system void fraction evolution while the RC pumps are operating is not significant.
Figures 2-8 and 2-9 show the impact of one versus two HPI pumps on systen pres-sure and average void fraction transients for a 0.05 ft2 break with the RC pumps operative. As seen from these figures, the results with one HPI pump are not significantly different to the two HPI pump case and are bounded by the spectrum appr se's utilized. With one HPI pump, the system does depressurize more slowly (less steam condensation) and a higher short term equilibrium vo'd fraction is achieved. Also, recovery of the core following a loss of the RC pumps would be significantly longer with only 1 HPI pump avail-able.
The majority of the.tnalyses provided in this report uses two HPI pumps and demonstrat:s a core cooling problem with worst time pump trip given that assumption. As analysis of one HPI available cases would only show a larger problem, such cases have not been exten-sively considered. As demonstrated in section B.4, the resolution of this problem, forced early pump trip, provides assurance of core cooling for both one or two HPIs available cases.
Therefore, 1054 061.
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there is no need for further pursuit of the single HPI available Case.
The effect of the RCP tripping during the transient was studied by assuming that the pumps are lost when the system, reaches 90% void fraction. Loss of the RC pumps at this void fraction is expected to produca essentially the highest peak cladding temperature.
After the RC pumps are tripped, the fluid in the RCS separates and liquid falls to the lowest regions, i.e.,
tha lower plenum of the RV and the pump suction piping. At 90% void fraction, the core will be totally uncovered following the RC pump trip.
Thus, the time required to recover the core is longer than that for RC pump trips initiated at lower system void fractions.
System void frac-tions in excess of 90% cc.a possibly result in slightly higher tem-peratures due to the longer core refill times that may occur.
However, the peak cladding temperature results are not expected to be significantly different as the system pressure and core de-cay heat, at the time that a higher void fraction is reached, will be lower.
Table 2-2 shows the core uncovery time for the cases analyzed with the RC pumps tripping at 9C% void fraction with 2 HPI pumps avail-able for core recovery. As shown, the core will be uncovered for approximately 600 seconds for the breaks analyzed.
Figures 2-10 and 2.11 show the system pressure and void fraction response for the 0.075 ft2 break with a RC Amp trip at 90% void fraction. As seen in these figures, the system depressurizes faster after the RC pump trip, due to the change in leak quality, and the void fraction decreases indicating that the core is being refilled.
Figure 2-12 shops the core liquid level response fol.iowing the RC pump trip. The core is refilled to the 9 foot level with collapsed liquid approximately 625 seconds after the assumed pump trip.
Once the core liquid level reaches the 9 foot elevation, the core is expected to be covered by a two-phase mixture and the elsdding temperature excursion would be terminated.
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5.
Effect of 1.0 ANS versus 1.2 ANS Decav Curve An analysis was performed using the more realistic 1.0 ANS decay curve instead of 1.2 ANS decay curve.
The study was done for a 0.05 ft2 break with 2 HPI;s available and pumps tripped at 90%
system void fraction.
Figures 2-13 and 2-14 show a comparison of system pressure and average system void fraction for 1.0 and 1.2 ANS decay curves. As seen in Figure 2-13, the system pressure for 1.0 ANS case begins to drop from scturation pressure (41100 psia) about 200 seconds earlier than the case with 1.2 ANS as a result of reduced decay heat. Also, the system will evolve to a lower average void fraction as shown in Figure 2-14.
After the pumps trip at 90% system void fraction, the case with 1.0 ANS decay ct.rve has a shorter core uncovery time by approximately 200 sec-onis compared to 1.2 ANS case. This case demonstrates that the efi'ect of a delayed RC pump trip may be acceptable when viewed realistically.
A peak cladding temperature assessment for this case will be provided in a supplementary response planned for September 15th, to the I&E Bulletin 7905-C.
6.
Effect of No Auxiliary Feedwater Analyses have also been performed with the RC pumps available and no auxiliary feedwater.
These analyses all assumed 2 HPI pumps were available.
The system void fraction evolutions for these calculations were not significantly different from those discussed with auxiliary feedwater.
Thus the conclusions of the cases with auxiliary feedwater apply.
m
.2 break Location Sensitivity Study A otudy was conducted to demonstrate that the break location utiliz;-d for the preceeding analyses is indeed the worst break location. As stated previously, the analyses were performed assuming that the break 2
was 16phted in the bottom of the pump discharge piping. A 0.075 ft hot leg break was analyzed to provide a direct amparison to a similar case in the cold leg.
For this evaluation, the RC pumps were assumed to trip after the RCS void fraction rea.hes 90%.
Figure 2.15 shows the average system vcid fraction transient and the core uncovery times for both the 0.075 ft hot and cold leg breaks. As shown, the cold leg break reaches 90% void fraction approximately 150 seconds earlier than the hot leg break. Also, the cold leg break yields a core uncovery time of 175 seconds longer than the hot leg break.
The quicker core racovery time for the hot leg break is caused by th<: greater penetration of the HPI fluid for this break. For a cold leg break in the pump discharg';
piping, a portion of the HPI fluid is lost directly out the break and is not available for core refill.
For a hot leg break, the full HPI flow is available for core refill.
Thus, as shown by direct comparison and for the reasuns given above, hot leg breaks are less severe than breaks in the pump discharge piping.
8 Peak Cladding Temperature Assessment As described previously, a RC pump trip, at the time the RCS void fraction is 90%, will reselt in core uncovery times of approximately 600 seconds. The peak cladling temperatures for these cases were evaluated using the small break evaluation model core power shape used to demonstrate compliance with Appendix K and 10CFR50.46. Also, an adiabatic heatup assumption during the time of core uncovery was utilized.
This approach is extremely conservative in tbst the power shape and 1054 0#,
local power rate (kw/ft) analyzed is not expected to occur during normal plant operation. Furthermore, use of an adiabatic heatup assumption neglects any credit for the steam cooling that will occur during the core refill phase and also neglects the effect of any radiation heat transfer. Using a decay heat power level based on 1.2 ANS at 1500 seconds, the cladding vill heatup at a rate will be 6.5 F/S under the adiabatic assumption. With a core uncovery period of 600 seconds and the adiabatic heatup assumption, cladding temperatures will exceed the criteria of 10CFR50.46. Use of a more realistic heat transfer approach with the extreme power shape utilized for this eval-uation is also expected to result in cladding temperature in exceas of the criteria.
In order to ensure compliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prompt tripping of the RC pumps is required.
Section B. demonstrates that a prompt trip of the RC pumps upon receipt of a low pressura ESEAS signal will result in compliance to the criteria.
An evaluation of the peak cladding temperature using a power shape encountered during normal operation for a realistic transient response with delayed RC pump trip will be provided by September 15, 1979.
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C.
Analysis Applicability to Davis-Besse I The significant parametric differences between the raised-loop Davis-Besse I plant and the precceding generic lowered-loop analysis are in the high pressure injection (HPI) delivery rste and the amount of liquid volume which can effectively be used to cool the core.
The liquid volume differential is due to the basic design difference; raised versus lowered loops. Because of the raised design, system water available after the RC pumps trip will drain into the reactor vessel. For the lowered loop designs, the available water is split between the reactor vessel and the pump suction piping. Thus, for the same average system void fraction, the collapsed core liquid level following an RC pump trip is higher for the raised loop design than for the lowered loop design.
Figure 2-16 shows a comparison of the delivered HPI flow for the Davis-Besse I plant and the lowered loop plants. As shown, for a similar number of HPI pumps available, the Davis-Besse I pumps will deliver more flow. For the delayed pump trip cases presented in section B.4 of this report, the Davis-Besse I plant will take approximately 450 seconds to recover the core as opposed to :600 seconds for the lowered-loop plants. However, it is noted that the core recovery time is based on using two HPI's rather than one, as required by Appendix K.
Use of only one HPI pump for Davis-Besse I will result in core uncovery times in excess of 600 seconds.
The Davis-Besse I plant cannot be shown to be in compliance with 10CFR50.46 for a delayed RC pump trip.
Prompt reactor coolant pump trip is, therefore, necessary to ensure compliance of the Davis-Besse I plant with 10CFR50.46.
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.D.
Effect of Prompt RC Pump Trip on Low Pressure ESFAS Signal As demonstrated by the previous sections, the ECC system can not be demonetrated to comply with 10CFR50.46 using present evaluation techniques and Appendix K assumptions under the assumption of a delayed RC pump trip.
Thus, prompt tripping of the RC pumps is necessary to ensure conformance.
Operating guidelines for both LOCA and non-LOCA events have been developed which require prompt tripping of the RC pumps upon receipt of a low pressure ESFAS signal.
Because no diagnosis of the event is required by the operator and ESFAS initiation is alarmed in the control room, prompt tripping of the RC pumps can be assumed.
e The effect of a prompt reactor coolant pump trip on an ESFAS signal has been examined to ensure that the consequences of a small LOCA are bounded by previous small break analysed 2which assume RC pump trip on reactor trip.
As shown by Table 2-3 at the time of low pressure ESFAS initiation, keeping the RC pumps running results in a lower average system void fraction.
This occurs because the availability of the RC pumps results in lower hot leg temperatures and thus less flashing in the RCS at a given pressure. Thus, a prompt trip upon receipt of an ESEAS signal will result in a less severe system void fraction evolution thcn cases previously analyzed assuming RC pump on reactor trip.
E.
Conclusions The results of the analyzes described in this section can be summarized as follows:
1)
If the RC pumps remain operative, core cooling is assured regardless of system void fraction.
- 2) For breaks greater than 0.025 ft, the RCS may evolve to system void fractions in excess of 90%.
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2
- 3) At 40 minutes, the 0.025 ft break has evolved to only a 47% void fraction. Thus, a delayed RC pump trip for breaks less than 0.025 ft will not result in core uncovery.
- 4) The potential for high cladding temperatures for a small break transient with delayed RC pump trip is restricted to a time period between that time where the system has evolved to a high void fraction and the time of LPI actuation.
- 5) Even with 2 HP1 pumps available, tripping of the RC pumps at the worst time (90% void fraction) results in a core uncovery period which cannot be shown to comply with 10CFR50.46, if Appendix K assumptions are utilized.
- 6) A prompt RC pump trip upon receipt of a low pressure ESFAS signal will provide compliance to 10CFR50.46.
- 7) The above conclusions are apn11 cable to both the B&W 177 FA lower 2d and raised loop NSS designs.
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III. IMPACT ASSESSMENT OF A RC PUMP TRIP ON NON-LOCA EVENTS A.
Introduction Some Chapter 15 events are characterized by a primary system response similar to the one following a LOCA. The Section 15.1 events that result in an increase in heat removal by the secondary system cause a primary system cooldown and depressurization, much like a small break LOCA.
Therefore, an assessment of the conse-quences of an imposed RC pump trip, upon initiation of the low RC pressure ESFAS, was made for these events.
B.
General Assessment of Pump Trip in Non-LOCA Events Several concerns have been raised with regard to the effect that an early pump trip would have on non-LOCA events that exhibit LOCA characteristics. Plant recovery would be more difficult, dependence.
on natural circulation mode while achieving cold shutdown would be highlighted, manual fill of the steam generators would be required, and so on.
However, all of these drawbacks can be accommodated since none of them will on its own lead to unacceptable consequences.
- Also, restart of the pumps is not precluded for plant control and cooldown once controlled operator action is assumed. Out of this search, three major concerns have surfaced which have appeared to be sub-stantial enough as to require analysis:
1.
A pump trip could reduce the time to system fill /repressurization or safety valve opening following an overcooling transient.
If the time available to the operator for controlling HPI flow and the margin of subcooling were substantially reduced by the pump trip to where timely and effective operator action could be questionable, the pump trip would become unacceptable.
2.
In the event of a large steam line break (maximum overcooling), the blowdown may induce a steam bubble in the RCS which could impair natural circulation, with severe consequences on the core, es-pecially if any degree of return to power is experienced.
3.
A more general concern exists with a larEe steam line break at EOL conditions and whether or not a return to power is experienced following the RC pump trip.
If a return to critical is experienced, natural circulation flow may not be sufficient to remove heat and to avoid core 6amage.
1054073,,,.
Overheating events were not considered in the impact of the RC pump trip since they do not initiate the low RC pressure ESFAS, and therefore, there would be no coincident pump trip.
In addi-tion, these events typically do not result in an empty pressurizer or the formation of a steam bubble in the primary system.
Reactivity transients were also not considered for the same reasons.
In addi-tion, for overpressurization, previous analyses have shown that for the worst casc conditions, an RC pump trip will mitigate the pressure rise. This results from the greater than 100 psi reduction in pressure at the RC pump exit which occurs after trip.
C.
Analysis cf Concerns and Results 1.
System Repressurization In order to resolve this concern, an analysis was performed for a 177 FA plant using a MINITRAP model based on the case set up for niI-2, Figure 3.1 shows the noding/ flow path scheme used and Table 3,1 provides s description of the nodes and flow paths. This case assumed that, as the result of a small steam line break (0.6 ft.
split) or of some combination of secondary side valve failure., secondary side heat demand was increased from 100% to 138% at time zero. This increase in secondary side heat demand is the smallest which results in a (high flux) reactor trip and is very similar to the worst moderate frequency overcooling event, a failure of the steam pressure regulator.
In the analysis, it was assumed that following HPI actuation on low RC pressure ESFAS, main feedwater is ramped down, MSIV's shut, and the auxiliary feedwater initiated with a 40-second delay.
This action was taken to stop the cooldown and the depressurization of the system as soon as possible after HPI actuation, in order to minimize the time of refill and repressurization of the system.
Both HPI pumps were assumed to function.
The calculation was performed twice, once assuming two of the four RC pumps running (one loop), and once assuming RC pump trip right after HPI initiation. The analysis shows that the system behaves very similarly with and without pumps.
In both cases, the pressurizer refills in about 14 to 16 minutes from initiation of the transients, with the natural circula-105407h,
tion case refilling about one minute before the case with two of four pumps running (See Figures 3.2,3.3). In both cases, the systen is highly subcooled, from a minimum of 30 F to 120*F and increasing at the end of 14 minutes (refer to Figure 3.4).
It is concluded that an RC pump trip following HP1 actuation will not increase the probability of causing a LOCA through the pressurizer code safeties, and that the operator will have the same lead time, as well as a large margin of subcooling, to control HPI prior to safety valve tapping. Although no case with all RC pumps was made, it can be inf erred from the one loop case (with pumps running) that the subcooled margin will be slightly larger for the all pu=ps running case.
The pressurizer will take longer to fill but should do so by 16 minutes into the transient. Figure it shows the coolant temperatures (hot leg, cold leg, and core) as a function of time for the no RC pumps case.
2.
Effect of Steam Bubble on Natural Circulation Coolins; For this concern, an analysis was perfomed for the same generic 177 FA plant as outlined in Part 1, but assuming thct as a result of an unmitigated large SLB (12.2 f t.
DER), the excessive cooldown would produce void formation in the primary system.
The intent of the analysis was to also show the extent of the void formation and where it occurred. As in the case analyzed in Part 1, the break was symmetric to both generators such that both would blow down equally, maximizing the cooldown (in this case there was a 6.1 ft.2 break on each loop). There was no MSIV closure during the transient on either steam generator to maximize cooldown. Also, the tur-bine bypass system was assumed to operate, upon rupture, until isolation on ESFAS.
ESFAS was initiated on low RC pressure and also actuated HPI (both pumps), tripped RC pumps (when applicable) and isolated the HFWIV's. The AFW was initiated to both generators on the low SG pressure signal, with minimum delay time (both pt"nps operating).
This analysis was performed twice, once assuming all RC pumps running, once with all pumps being tripped
- on the HPI w
actuation (nf ter ESFAS), with a short (45 second) delay.
In both cases, voids were fomed in the hot legs, but the dura-105407'$.
tion and size were smaller for the case with no RC pump trip (refer to Figure 3.7).Although the RC pump operating case had a higher cooldown rate, there was less void forma-tion, resulting from the additional system mixing. The coolant temperatures in the pressurizer loop hot and cold legs, and the core, are shown for both' cases in Figures 3.5, 3.6.
The core outlet pressure and SG and pressurizer levels versus time are given for both cases in Figures 3.8, 3.9.
This analysis shows that the system behaves very similarly with and without pumps, although maintaining RC pump flow does seem to help mitigate void formation.
The pump flow case shows a shorter time to the start of pres-surizer refill than the natural circulation case (Figure 3.9),
although the time difference does not seem to be very large.
3.
Effect of Return to Power There was no return to power exhibited by any of the BOL cases analyzed above'.
Previous analysis experience (ref.
Hidland FSAR, Section 15D) has shown that a RC pump trip will mitigate the consequences of an EOL return to power condition by reducing the cooldown of the primary system. The reduced cooldown substantially increases the suberitical margin which, in turn, reduces or eliminates return to power.
D.
Conclusions and Summary A general assessment of Chapter 15 non-LOCA events identified three areas that warranted further investigation for impact of a RC pump trip on ESFAS low RC pressure signal.
1.
It was found that a pump trip does not significantly shorten the time to filling of the pressurizer and approximately the same time interval for operator action exists.
2.
For the maximum overcooling case analyzed', the RC pump trip increased the amount of two-phase in the primary loop; however, the percent void formation is still too small to affect the ability to cool on natural circulation.
3.
The suberitical return-to-power condition is alleviated by the RC pump trip case due to the reduced overcooling effect.
Based upon the above assessment and analysis, it is con-cluded that the consequences of Chapter 15 non-LOCA events are not -
1054 078
increased due to the addition of a RC pump trip on ESFAS low RC pressure signal, for all 177 FA lowered loop plants.
Although there were no specific analyses performed for TECO, the conclusions drawn from the analyscs for the lowered loop
~P ants are applicable.
l 5
a 6
4 e
e' O
105407$
19 -
v
Table 2-1.
Analysis Scope With AFW Available Continuous RC Break location pump operation RC pump trip @ 90% void Break size, 2
2 HPI (ft )
Cold leg Hot leg 2 HPI 1 HPI 0.025 X
X 0.05 X
X*
X X*
0.075 X
X X
X 0.10 X
X X
0.20 X
X
- Analyzed with both 1.0 and 1.2 ANS decay curves.
1054 0 9
_ 20 _
Table 2-2.
Impact Assessment of Break Spectrum With RC Pump Trip at 90% Void Break size (ft )
Core uncovery time (sec) 2 0.10 550 0.075 625 0.05 575 Notes:
1.
Two HPIs available dering the transient.
2.
Core uncovery time is the time period following pump trip re-quired to fill the inner RV with water to an elevation of
- 9. ft in the core which is ap-proximately 12.ft when swelled.
~ 21 1054 08
Table 2-3.
Comparison of System Void Fractions at ESFAS Signal System void fraction a
AS Break size, (ft2)
Pumpe on Pumps tripped 0.02463 0.0 0.04 4.47 0.05 0.04 0.055 6.74 0.07 8.06 0.075 0.90 0.085 8.45 0.10 2.17 7.97 0.15 10.70 0.20 6.78 105408U
_ 22 _
MINITRAP2 NODE DESCRIPTION NODE NUMBER DESCRIPTION 1,33 Reactor Vessel, Lower Plenua 2,34 Reactor Vessel, Core 3,35 Reactor Vessel, Upper Plenum 4,10 Hot Leg Piping 5-7,11-13 Primary, Steam Generator 8,14 Cold Leg Piping 9,32 Reactor Vessel Downcomer 15 Pressurizer 16,24 Steam Generator Downcomer 17,25 Steam Generator Lcwcr Plenum 18-20,26-28 Secondary, Steam Generator 21,29 Steam Riscrs 22,30 Main Steam Piping 23 Turbine 31 Containment MINITRAP2 PATH DESCRIPTION PATH NUMBER DESCRIPTION 1,2 Core 45,46 Core Bypass 3,5,5,11,12,44 hot Leg Piping 6,7,13,14 Pitmary, steam Generator 8.15 RC Pumps 9,16 Cold Leg Piping 10,43 Downcomer, heacto:. Vessel 17 Pressurizer Surge Line 18,13,26,27 Steam Generator Downcomer 20,21,28,29 Secondary, Steam Generator 22,30 Aspirator 23,31 Steam Riser 74,32 Steam Piping 25,33 Turbine Piping 34,35 Break (or Leak) Path 36,37 HPI 38,39,43,44 A7W 40,41 Main Feed Pumps 42 LPI Table 3.1 n
1054 082 Figure 2-1. CRAFT 2 Noding Diagram ior small aren Q1 23 Il 3
7.
14 g
@d 11 13 o
cc a,
O
<s
( A 3
- [ f #
e $
3 p
i T
s e
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's g_
. g.....,_.....
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h h-b,.. a.... u.i..
us rath
.s c...to i.
Qn
- a......i. c... s. :
p.n.
5ps.; 6..t
)tode No.
Identification Path No.
Identiffcatien 1
Douncomer 2,2 Core 2
lever Plenum 3,1,18,19 Bot Les Piping 3
Core, Core Bypass. Upper 5,20 Bot leg. Upper Plenum, Upper Ecad 6,21 SC Tubes 4,14 Bot les Paping 7.22 SC, lover Uezd 3,15 -
Steam Generator Upper B
Core Fjpass Bead SC Tubes (Upper Balf) 9,13.24 Cold t.eg Piping 6.16 SG Tubes (1.over Ealf) 10,14.25 Pt=ps 8,18 SG Lover Bead 11,12,15,16,26,27 Cold Leg Piping 9.11.19 Cold Leg Piping (Pu=p Suction) 17,31
~
Dovr. o=er 10.12,20 Cold Les Piping (Pump Discharge) 23 LP1 13 Upper Dovnconer 28,29 Upper Devneo=er
-(Above the q,of Nozzle Belt) 30 Preseuriser 21 Pressuriser 32 Yeat Valve 22 Cootsinnent 33,34 trak & Return Path 35,36 EP1 37 Containment Sprays O
D O
vo
.ald.3L)B 19 L,T" u-a 1054 08F
Figure 2-2.
CRAFT 2 N0 DING DIAGRAM FOR SMALL BREAKS (6.N0DE MODEl)
CFT l
G 4
3 2
6 5
g L
1 G
s m
a n,
n@
s M@
(
LEAK PATHS 8 & 9 M@1 M ao M
Node No.
Identification Path No.
Identification 1
Core 2
Primary SG 2
LPI 3
Core, UP, Hot Legs
-3,10,11 HPI 4
Pressurizer 4
Not Legs 5
Containment 5
Pumps 6
Secondary SG 6
Vent valve 7
Pressurizer 8,9 Leak & Return Path oo
]D~
1u
_ ]_
_a 105408[.
CORE PRESSURE VS TIME,177-LL, 2772 MWt, PUMPS ON 2
0.025 FT BREAK 23 H00E N00EL 2
20 O.025 FT BREAK 6 N00E N00EL
~~~~
18 n
"o;
".2 16 E.
E 3
14 S
o 12 10 t
i f
f g
0 500' 1000 1500 2000 2500 Time, sec Figure 2-3 lD *
- M T a a lt a
. ]_
1cs4 086 2-
PERCENT SYSTEM VOIDS VS TIME, PUMPS ON 100 80 e
ed."
E SD t;
fh0 $
y 40 e.02g T E (M -' ')
t' E
y *t e ts I
9 p 'g,015 20 - / ',#,
/
0 0
400 800 1200 1600 2000 2400 Time, sec Figure 2-4 "fg?
eD
'y J
5 W M UL1 3
1054 08'
- 2, _
BREAK SPECTRUM-RC PRESSURE WITH THE RC PUMPS OPERATIVE AND 2 HPl PUMPS 2500 2000
=
i 1500
.i m '.
N p
1000 2
s s s
- 0.025 FT N. N
'~ s N
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,~~
- x-
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- N..
0.2 FT2 0.10 FT2 0
t i
O 500 1000 1500 2000 2500 3000 3500 Time, see Figure 2-5 c
D
- [1 aa1 0q 9 i o Jul a 1054 08%
_a BREAK SPECTRUM. AVERAGE SYSTEM V010 FRACTION WITH THE RC PUMPS OPERATIVE AND 2 HPI PUMPS 100 9,- o _ p o,o s
.,,,,o o
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./,,,,,
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0 0
400 800 1200 1600 20 %
M M
Time, sec Figure 2-6 D@n D
go P
U T)
I D
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' 05408{
2 BREAK WITH CONTINUOUS RC PUMP 0.1 FT OPERATION AND 2 HPI PUMPS 2500 100 g
/
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~ ~.
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0 0
i i
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0 500 1000 1500 2000 2500 3000 Time, sec Figure 2-7 D
- D oo 07 9]
7,-
10540$8 o
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_ 5
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" a RC FRESSURE FOR 0.05 FT2 BREAK AVAILABLE 1 HPl VS 2 HPI'S
/
30 2 HPI'S, PUMP DN, HOMOGENEDUS 25 1 HPI, PUMP DN,HOMOGENE0US 20
=
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15 i
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1 0
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500 1000 1500 2000 2500 3000 Time, sec Figure 2-8 e
PDO D
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I).
eqgw ofu a 1!2,2 105409h 2
AVERAGE SYSTEM V010 FRACTION FOR 0.05 FT AVAILABLE 1 HPI VS 2 HPI'S 100 o-o-
-o--o 7 /*f 80 s'
/
l/
/
f
2 HPI'S, PUMP DN,HOM0 GENE 00S f
60
/
-o-1 HPI, PUMP DN, HOMOGENEOUS w
O /
O
/
m
/
/
40 o,/
2 s
- ~
/
/
20 0
i i
i 0
400 800 1200 1600 2000 2400 2800 3200 Time, sec Figure 2-9 e
S
- ]D D
o o JU
) h a
JU aJ > L : a 1054 09 k 2
RC PRESSURE FOR O'.075 FT, PUMPS OFF @ 90% SYSTEM V010 1
30
_ _ _ _ 2 HPl*S, PUMP DN, HOMOGENEOUS 25 2 HPI'S, PUMP OFF @
90% V010, 2 PHASE e2' 20 E
15 E
i
=>
i a
10
'~ss s%
5 9.N*sQg[~
N.
,, ~ ~ ~
0 O
500 1000 1500 2000 2500 3000 Time, sec' Figure 2-10 I
cO{D D
D 19 1054093
_ 33 _
AVERAGE SYSTEM V010 FRACTION FOR 2
0.075 FT, PUMPS OFF @ 905 SYSTEM V010 100
,s"~~~~...
80 f
- N,%
m'
/
N
/
E l
60
/
3
/
~
/
6
/
2 HPl'S, PUMP DN,HOMOGENE005 e
1 y
40
,/
2 HPI'S, PUMP OFF G 90% V010 j
,/
2 PHASE
/
20 g
0 400 800 1200 1600 2000 2400 3200 Time, sec Figure 2-11 D
D 01 iS 1
h[
J 1054 096 AVAILABLE LIQUID VOLUME VS TINE FOR 0.075 FT2 BREAK WITH 1.2 ANS DECAY HEAT CURVE 3000
^
m
{
2500 E
2 5
2000 3
9' LEVEL OF ACTIVE CORE
=
1500 2
E:
1000 i
RC PUMPS OFF 500 0
400 000 1200 1600 2000 Time, see Figure 2-12 e
D D
n1 0 p) 9 7(
b dL
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_a 105409g
9 2
RC PRESSURE VS TIME FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUMP TRIP 2
,.., 0.05 FT, 2 HPI'S 3000 1.2 ANS, PUMP DN 2
0.05 FT, 2 HPI'S 1.0 ANS, PUMP DN 2
2500
_, _ 0.05 FT, 2 HPI'S 1.2 ANS, PUMP OFF 2
_ 0.05 FT, 2 HPI'S
~
1.0 ANS, PUMP OFF 5 2000 et
.I a
m 1500 n.
i
's
~
- D '.'~~.
O
1000
~. q u s.
500
'4'.-
M 0
0 500 1000 1500 2000 2500 3000 Time, sec q
ol 0 o
c, 0 fb t
s ul)
_ #L 1 a
105409h
-u-
2 PERCENT SYSTEM V010 FRACTION FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUMP TRIP 100
- , M'-;.* k -.q%*,., %
80 sY 60 0.05 FT, 2 HPI*S, PUMP DN*
2 6"
1.2 ANS y
2 0.05 FT, 2 HPI'S, PUMP DN, E
40
/
1.0 ANS
-.~
f 2
/
-o 0.05 FT, 2 HPl'S, PUMP 0FF' e
/
1.2 ANS
/
2 20
_ z _ 0.05 FT, 2 HPl'S, PUMP OFF,
. 0 ANS 0
0 400 800 1200 1600 2000 2400 2800 Time, sec Figure 2-14 e
d a o J1
$," @d }{
~
a LUL)a JL
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1054098.
2 AVERAGE SYSTEM V010 FRACTION VS TIME FOR A 0.075 FT BREAK, BREAK LOCATION COMPARISON PUMPS OFF @ 90% V010 100 UNC0VERY TIME = 625 B0 s
h/#
UNC0VERY TIME =
j
[/@
f 60 450 SEC
,/@e
/r ~
u>
y 40
,/
T/
20 - f/
E 0
0 400 800 1200 1600 2000 2400 Time, see Figure 2-15 4
e v:,1 mpal[iDQ 7 b. [
lucuedu ou b
\\
105409f
-r-
4 COMPARISON OF DELIVERED HIGH PRESSURE INJECTION FLUID TO RV FOR PUMP OISCHARGE BREAK t
\\
1200 -g 1100 gf #
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900 g
m E
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3 800 g' 700
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I 300 N
- LOOP I yp, 200 100 0
i e
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200 600 1000 1400 1300 2200 2600 3000 Pressure, (psia)
Figure 2-16 e
J O Q Ju F
m cu mu o
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HINITRAP2 Noding and Flow Path Scheme
Pi;i SClZEi: 1. ] SIEfd' Cd.Elial[i: L IC'JID I EVEl. Vil SUS Ti't!;SILHT ll!'E 2
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S.G. 1UBE REf:10!1 ooeogo 0
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STEAM GENERATOR 'A' 6:
STEAli! GENERATOR 'B' I
I I
t t
-t D
2 4
6 8
10 12 14 16 Transient Time (Minutes)'
Tf9 0
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Figure 3.5 ao.
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Figure 3.7 KEY O:
HOT LEG (PRZR) - RC PUI1P TRIP 0:
HOT LEG 'B' LOOP-RC PuntP TRIP A:
Il0T LEC (l'RZR LO0l') - NO TRIP 0:
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REFERENCES I
B.M. Dunn, et al., "B&W's ECCS Evaluation Model," BAW-10164, Rev. 3, August 1977.
2 Letter, J.H. Taylor (B&W to S.A. 5arga (NRC), July 18, 1978.
3 R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, " CRAFT 2 - Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant,"
BAW-10092, Rev. 2, April 1975.
J.F. Wilson, R.J. Grenda, and J.F. Patterson, "The Velocity of Rising Steam in a Bubbline Two-Phase Mixture," ANS Transactions, 5, (1962).
/ o dv 9 in w T
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. 49 -
GUIDELINES FOR OPERATOR ACTION I.
htroduction Guidance for operator action, during both LOCA and non-LOCA events, to account for the impact of the RC pump trip requirement of IE bulletin No.79-05C, have been developed and are presented below. The general intent of these additional instructions is as follows:
1.
To establish the basis and criteria for a RC pump trip and 2.
To identify plant conditions for which a restart of the RC Pumps, if tripped, is permissable.
Section VI provides the " Operating Guidelines for Small Breaks" updated to include the impact of the RC pump trip requirements. These guidelines, in general, apply to any abnormal event where a RCP trip is required and will be used as the basis for revisions to emergency operating procedures and operator training.
II. Ba is and Criteria for a RC Pump (RCP) Trip B&W analyses of small loss-of-coolant accidents, with the RC pumps operative, indicated that the primary reactor coolant conditions evolve to high void fractions during the initial stages of the transient when the system pressure is still relatively high. The consequences of these postulated events with continouous RC pump operation are acceptable as effective core cooling is maintained due to the forced circulation of reactor coolant.
For a certain range of small breaks, however, a RCP trip (by any means such as loss of power or operator action) at a time when the coolant void fraction is excessively high can lead to core uncovery and a potential for cladding temperatures in excess of 2200F, lObd l
To preclude the potential consequence of an untimely RCP trip, the RCP's will be promptly shutdown when RCS conditions indicate a small break in this size range may be in progress.
This action ensures safe plant con-ditions as demonstrated by past small break analyses, undIr Appendix K assumptions, wherein the RC pumps were assumed inoperative early during the transient.
In the interim, until design changes can be made to automate the RCP trip, operating procedure will require that the operator trip the RCP's imediately following ESFAS actuation due to low RC pressure (1 1600 psig).
Table 1 outlines the general diagnostic and confirmatory actions which will be required in addition to other immediate actions in present procedures.
These imediate actions apply to any abnormal event which results in automatic ESFAS actuation on low RC pressure and will be memorized by reactor operating personnel during training programs.
As indicated above, a prompt trip of the RC pumps is required in order to maintain demonstrated conformance to 10CFR50.46. To provide good assurance that the operator will trip the RC pumps when required, the pump trip criteria (low pressure ESFAS actuation) was chosen over other possible candidates because it is a clear, simple, and early indication that a small LOCA may be in progress. The visual indication and alanns in the control room following ESFAS actuation also alert the operator to the status of the plant, and no decision process or continuous monitoring by the operator is required to decide that an RC pump trip is necessary. With procedure changes consistent with Table 1 and additional training, failure of the operate to, initiate an RC pump trip when required is believed to be remote.
1054lik
Table 1:
IMMED'IATE ACTIONS REQUIRED FOLLOWING ESFAS ACTUATION 1.
Criteria for RCP Trip Upon automatic actuation of the ESFAS due to low reactor coolant system pressure, RC pump operation shall be promptly terminated.
2.
Imediate Action A.
Upon receipt of an ESFAS actuation (indicated via audiable and visual alarms within the control room) the operator shall immediately verify that RC pressure is less than the low pressure ESFAS setpoint via examination of wide range RC pressure instrumentation or ESFAS Trip Status Indication, if available.
B.
If RC pressure is less than the low pressure ESFAS setpoint, RC pump operation shall be immediately terminated by manual depressing the in-dividual RC pump trip switches in the control room.
NOTE:
If the ESFAS has been actuated due to high RB pressure, the operator shall monitor RC pressure and trip the RC pumps if pressure decreases below the ESFAS setpoint.
C.
The operator shall imediately verify that the RC pumps are tripped by visual examination of RC pump status indications (status lights, motor current,etc.).
D.
Followi., a trip of the RC pumps, the operator shall verify that the auxiliary feedwater system has been acteated and that SG level is controlled to the emergency high level control setpoint to ensure establishment of natural circulation.
1054 li b
III. Criteria for RCP Restart Plant control following abnormal events, including small breaks, is greatly improved if the RC pumps are operative. Witt: forced circu-lation of reactor coolant, the steam generators and associated auxiliary systems are more effective in removing the primary system stored energy and decay heat. The plant is also placed in a more " normal" mode of operation where more familiar pressure / temperature control procedures can be employed by operating personnel. Therefore, to compliment the RC pump trip criteria provided in Section II, conditions under which an RC pump restart is allowed have also been identified. These conditions cover bnth LOCA and non-LOCA events and have been carefully chosen to preclude the development of excessive void fractions for small breaks where an RC pump restart is allowed.
Table 2 lists the conditions under which a RC pump restart is allowed.
For each condition, typical events for which they apply and a brief discussion of the basis for the RC pump restart is provided.
It should be noted that a RC pump restart is not allowed unless feedwater is available to at least one steam generator. A cross-reference to the appropriate sections of the small break guidelines where specific informatica can be found is also given.
Furthermore, the criteria given in Table 2 are g
not new as each was previously issued in past small break guideline submittals.
Q B&W has reviewed the guidelines in light of the break size and system conditions mb for which a RC pump trip is required and has confirmed that the RC pump
>Cza Qg ret, tart guidance is still appropriate.
ca, A 1054 11F c=a As indicated in Table 2, system repressurization and the establishment of g
subcooled conditions are specified for use on non-LOCA events as criteria for which a RC pump restart is allowed.
For these abnormal events, restart of the RC pumps is recommended by B&W when the Pump Restart criteria is satis-fi d t id 'r i
t r and control. Emer ene rocedures for non-
LOCA events, for which a RC pump trip may be initiated, will thus be revised to include the pump restart criteria.
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10541if w
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ammet TABLE 2: RC PLMP RESTART CRITERIA I 2,3 TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUSSION CONDITION FOR WHICH A PUPF RESTART IS ALLOWED WHICH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION)
Regain Coolant Subcooling
- 1. Small Leak 4.3.4.3.2 Following any reactor trip event during
- 2. Small Break within which the RC pumps become inoperative 1.
P-T conditions indicate capacity of HPI sys.
(loss of power due to natural causes/
coolant is > SOF subcooled.
- 3. Isolated Small Break equipment failures or due to a deliberate
- 4. Non-LOCA Overcooling /
trip initiated by the operator), the RC depressurizing event rumps can be restarted if RC conditions are stabilized and at least 50F of sub-
- 5. Loss-of-Offsite Power Event cooling is indicated for the existing P-
{_
T state. If subcooled conditions are b
indicated, the primary and secondary sys-L ~y tems are directly coupled (ie, decay
_/
C) heat removal via natural circulation);
F and if a breach of the primary pressure f-Shh 9
boundary is present also, the resulting leak will be within the capacity of the ECCS systems. The operator should restart b.Q {w]
the RC pumps (1 in, each loop) return to low SG Level control, and proceed with I
i b cp I_
a plant cooldown or maintain the plant 3'~
at hot shutdown if the initiating event is correctable and a return to power g
lj operation possible.
NOTE: The subcooling criteria will be the principle indicator for a RCP restart for non-LOCA events.
Repressurization 1.
Stable or increasing
- 1. Small Break within capacity 4.3.'.4.1 Certain small breaks will result in a pressure with PRCS >
of HPIS system repressurization due to momentary 1600 psig.
- 2. Overcooltr3/Depressurization loss of the SG as a condensgr for primary event system steam (ie, the HPIS is refilling
- 3. Isolated Small Break the system and a steam bubble is trapped within the hot legs above the SG tubes condensing surface). Small breaks which produce this primary system behavior O
are sufficiently small such that high J1 void fractions will not evolve if the RC pumps are restarted. A RCP restart is thus allowed; this action will equal-ize primary and secondsry pressures and temperatures and couple the primary and secondary systems such that an orderly "k
cooldown and depressurization of the RCS can be accomplished. Section 4.3.4.4.1 of the small break guidelines would
TABLE 2 CONT'D 2
CONDITION FOR WHICH '
TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUSSION A PUPP RESTART IS ALLOWED 1AllCH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION) akply to a very small break where a sys-tem repressurization would occur early (C)
(ie, prior to initiation of the second-r ~~',
ary system depre?surization). A RCP OC restart and resulting drop in the primary b ' "!
system pressure to that of the second-L ary side may allow the HPl$ to establish Q]> h]
a subcooled primary system. System L--
repressurization above the low pressure Ci G
]
ESFAS setpoint for non-LOCA events is k
also an acceptable condition for an RC
{
pump restart. In most cases, increasino a
RC pressure will also tend to re-establish I
the reactor coolant subcooled margin L-which, as indicated above, is the principle
{
indicator for a RCP restart for non-LUCA event. A pump restart, when system pressure is above the ESFAS setpoint when the 50F subcooled margin is not yet established, is permissable since small breaks for which a RC pump trip is re-quired will not produce the system behanfor.
- 2. Increasing system pressure Small Break 4.3.4.4.2 4.3.4.4.2 of the small break guidelines dere PRCS > + 600 (psig) applies during the cooldown process where during coollown process, the secondary pressure has been manually reduced below normal control (hot shut-down) setpoints. A pump bump procedure is stipulated. The intent of this action
,'is to mix the system so that steam can be condensed to allow a system refill.
'If a refill and subcooled conditions are not established, the 600 psi decrease in primary system pressure will prevent high RCS void fractions witi an RCP k
restart per the guidance provided.
~
Final Transition to LPI O
Coolino LD Stab 11 red pressure with Small Break 4.3.4.4.3 For certain small breaks, a primary system PSS< 100 psig and PRCS refill may not be possible until low
> 250 psig primary system pressures are achieved.
'n Complete depressurization may be impeded due to steam trapped within the upper hot leg piping. A bump of an RCP will depressurize the RCS such that a transition to LPI cooling per Appendix A of the small break guidelines is possible
TABLE 2 CONT'D 2.3 CONDITION FOR WHICH TYPICAL. EVENTS FOR INSTRUCTION LOCATION DISCUSSION A PUPF RESTART IS ALLOWED WHICH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION)
Continued operation of an RCP is also allowed since the LPI system will elimi-nate the potential for further increase in the system void fraction.
m Inadequate Core Cooling Small Break N/A Current considerations of the indications of and mitigating actions for inadequ-ate core cooling may result in the potential use of the RC pumps under certain condi-tions. Criteria for use of the RC pumps, if required, will be developed consistert with the schedule requirement of Item 5 (short term) of 79-05C.
NOTE:
1.
An RC Pump restart is allowed on1y if feedwater
~
is available to at least one steam generator.
2.
Standard precautions to be observed prior to pump restart.
~
A. CCW has been maintained or will be reinstated prior to r-starting the RC pumps.
cM)
B. Seal injection flow h s been maintained to all RC pumps.
r:m ]
C. Seal return is maintained or is reinstated prior to starting et-ESP) h pumps.
D. Pres r 250 psig.
r_ a 3.
Emergency operating limits for continued pump operation.
CD I%
m A. Shaft runout (vibration) shall not exceed 30 mils.
23 y
B. Frame vibration as measured on the lower motor mounting rm flange shall not exceed 5 -ils.
U M
IV. Operating Guidelines for Small Break Part I and Part II of the " Operating Guidelines for Small Breaks" have been revised to include the RC pump trip requirement of IE Bulletin 79-05C and are attached. This information will serve as the basis for revisions to emergency procedures and additional operator training.
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REVISION NO. 3 8/21/79 PART I'- OPERATING GUIDELINES FOR SMALL EREAKS 1.0 SYMPTOMS AND INDICATIONS (IMMEDIATE INDICATIONS) 1.1 EXCESSIVE REACTOR COOLANT SYSTDI (RCS) MAKEUP
- 1.2 DECREASING RCS PRESSURE 1.3 REACTOR TRIP 1.4 DECREASING PRESSURIZER LEVEL
- 1.5 ESFAS ACTUATION
- 1.6 LOW MAKEUP TANK LEVEL *
- MAY NOT OCCUR ON ALL SMALL BREAKS 2.0 IMMEDIATE ACTIONS o
2.1 IF THE ESFAS HAS BEEN INITIATED AUTOMATICALLY BECAUSE OF LOW RC PRESSURE, IMMEDIATELY SECURE ALL RC PUMPS.
2.2 VERIFY CONTROL ROOM INDICATIONS SUPPORT THE ALARMS RECEIVED, VERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS.
2.3 BALANCE HIGH-PRESSURE INJECTION (HPI) FLOW BETWEEN ALL INJECTION LINES WHEN HPI IS INITIATED.
o 2.4 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM GENERATOR (OTSG) LEVEL IS MAINTAINED BY FEEDWATER CONTROL (LOW LEVEL LIMIT WITH RC PUMPS OPERATING, EMERGENCY LEVEL WITHOUT RC PUMPS OPERATING).
o 2.5 MONITOR SYSTEM PRESSURE AND TEMPERATURE.
IF SATURATED CONDITIONS OCCUR, INITIATE HPI.
3.0 PRECAUTIONS o
3.1 IF THE ESFAS HAS BEEN INITIATED ON LOW RC PRESSURE, TERMINATION OF RC PUMP OPERATION TAKES PRECEDENCE OVER ALL OTHER IMMEDIATE ACTIONS.
NOTE:
IF ESFAS HAS BEEN ACTUATED ON HIGH RB PRESSURE, THEN MONITOR RC PRESSURE AND TRIP RC PUMPS ONCE PRESSURE DECPEASES BELOW THE ESFAS LOW PRESSURE SETP(
.T.
105411$
Revioed
REVISION NO. 3 8/21/79 PART I'- OPERATING GUIDELINES FOR SMALL BREAKS
?
1.0 _ SYMPTOMS AND INDICATIONS (IMMEDIATE INDICATIONS) 1.1 EXCESSIVE REACTOR COOLANT SYSTEM (RCS) MAKEUP
- 1.2 DECREASING RCS PRESSURE OO O
n 1.3 REACTOR TRIP o
CO 1.4 DECREASING PRESSURIZER LEVEL
- 1.5 ESFAS ACTUATION *
~
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D{
1 i
1.6 LOW MAKEUP TANK LEVEL
- O J
_J
- MAY NOT OCCUR ON ALL SMALL BREAKS 2.0 IMMEDIATE ACTIONS o
2.1 IF THE ESFAS HAS BEEN INITIATED AUTOMATICALLY BECAUSE OF LOW RC PRESSURE, IMMEDIATELY SECURE ALL RC PUMPS.
2.2 VERIFY CONTROL ROOM INDICATIONS SUPPORT THE ALARMS RECEIVED, VERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS.
2.3 BALANCE HIGH-PRESSURE INJECTION (HPI) FLOW BEIWEEN ALL INJECTION LINES WHEN HPI IS INITIATED.
o 2.4 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM GEKERATOR (OTSG) LEVEL IS MAINTAINED BY FEEDWATER CONTROL (LOW LEVEL LIMIT WITH RC PUMPS OPERATING, EMERGENCY LEVEL WITHOUT RC PUMPS OPERATING).
o 2.5 MONITOR SYSTEM PRESSURE AND TEMPERATURE.
IF SATURATED CONDITIONS OCCUR, INITIATE HPI.
3.0 PRECAUTIONS o
3.1 IF THE ESFAS HAS BEEN INITIATED ON LOW RC PRESSURE, TERMINATION OF RC PUMP OPERATION TAKES PRECEDENCE OVER ALL OTHER IMMEDIATE ACTIONS.
NOTE: IF ESFAS HAS BEEN ACTUATED ON HIGH RB FRESSURE THEN MONITOR RC PRESSURE AND TRIP RC PUMPS ONCE PRESSURE DECREASES BELOW THE ESFAS LOW PRESSURE SETPOINT.
1054 l'f
- Revised
_... _ 3.2 IF ESFAS HAS BEEN INITIATED, Tile RC PUMP TRIPPED, AND THE RCS DETERMINED TO BE AT LEAST 50 F SUBC00 LED. TIIE OPERATOR S110ULD ESTABLISil AS QUICKLY AS POSSIBLE IF THE CAUSE FOR Tile DEPRESSURIZA-TION IS DUE TO EITHER A LOCA OR NON-LOCA (OVERCOdLING) EVENT.
PROCEED TO STEP 4.4 FOR NON-LOCA EVENTS.
3.3 IF THE HPI SYSTDI !!AS ACTUATED BECAUSE OF LOW PRESSURE CONDITIONS, IT MUST RDIAIN IN OPERATION UNTIL ONE OF THE FOLLOWING CRITERIA IS SATISFIED:
1.
THE LPI SYSTDI IS IN OPERATION AND FLOWING AT A RATE IN EXCESS OF 1000 GPM IN EACH L7NE AND THE SITUATION HAS BEEN STABLE FOR 9 ObD 20 MINUTES.
h O
o OR oE_bJ
- 5. ALL HOT AND COLD LEC TEMPERATURES ARE AT LEAST 50 F BELOW THE SATURATION TDfPERATURE FOR THE EXISTING RCS PRESSURE, THE HOT LEG TDTERATURES ARE NOT MORE THAN 50 F HOTTER THAN THE SECOND-ARY SIDE SATURATION TDIPERATURE, AND Tile ACTION IS NECESSARY TO PREVENT THE INDICATED PRESSURIZER LEVEL FROM COING OFF-SCALE HIGH.
IF 50 F SUBC00 LING CANNOT BE MAINTAINED, THE HPI SHALL BE REACTIVATED. THE DEGREE OF SUBC00 LING BEYOND 50 F AND THE LENGTH OF TDIE HPI IS IN OPERATION SHALL BE LD!ITED BY THE PRESSURE /TD!PERATURE CONSIDERATIONS FOR THE VESSEL INTEGRITY (SEE SECTION 3.4).
3.4 WHEN THE REACTOR COOLANT IS > 50 F SUBC00 LED, THE REACTOR VESSEL DOWNCOFER PRESSURE /TDIPERATURE (P-T) COMBINATION SHALL BE KEPT BELOW AND TO THE RIGHT OF Tile LDIIT CURVE SHOWN IN FIGURE 1.
THE DOWNCOMER TDFERATURE SHALL BE DETERMINED AS FOLLOWS:
3.4.1 WITH ONE OR MORE RC PUMPS OPERATING USE ANY COLD LEG RTD AS AN INDICATION OF REACTOR VESSEL './0WNCOMER TDIPERATURE.
3.4.2 WITH NO RC PUMPS OPERATING THE RV DOWNCDMER TDIPERATURE Si!ALL BE DETERMINED BY AVERAGING THE FIVE LOWEST INCORE THERMOCOUPLE TDIPERATURE READINGS AND SUBTRACTING 150 F FROM THE AVERAGE INCORE THERMOCOUPLE TDIPERATURE VALUE.
1054124 Revised
-m pp s
d ca1 ET M
l
.a' T
=
t - 150 F g
R j
DWN 5
O I
I J
LP I
o WHERE T
= AVERAGE RV DOWNCOMER TEMPERATURE, F DWN 5
= SUM OF THE 5 LOWEST INCORE THERMOCOUPLE TEMPERATURE te READINGS.
NOTE: FIGURE I IS APPLICABLE ONLY UNDER LOCA CONDITIONS. THE P/T CURVE IN THE TECHNICAL SPECIFICATION IS VALID FOR ALL OTHER OPERATING CONDITIONS.
NOTE: WHEN THE REACTOR COOLANT IS LESS THAN 50 F SUBC00 LED, THE REACTOR VESSEL DOWNCOMER PRESSURE TEMPERATURE COMBINATION WILL INHERENTLY BE BELOW AND TO THE RIGHT OF THE LIMIT CURVE. THEREFORE, NO OPERATOR ACTION WILL BE REQUIRED TO PREVENT EXCEEDING THE REACTOR VESSEL INTEGRITY LIMITS UNTIL AFTER A > 50 F SUBC00 LED MARGIN EXISTS.
NOTE: WHEN THE REACTOR COOLANT IS 2 50'r SUBC00 LED, RC PRESSURE CAN BE REDUCED BY REDUCING :iE HPI FLOW RATE TO AVOID EXCEEDING THE RV INTEGRITY.LDfITS.
35 PRESSURIZER LEVEL MAY BE INCREASING DUE TO RCS READHING SATURATED CONDITIONS OR A BREAK ON TOP OF THE PRESSURIZER.
3.6 IF HIGH ACTIVITY IS DETECTED IN A STEAM GENERATOR, ISOLATE THE LEAKING GENERATOR.
IT IS RECOMMENDED THAT BOTH STEAM GENERATORS NOT BE ISOLATED.
3.7 OTHER INDICATIONS WHICH CAN CONFIRM THE EXISTENCE OF A LOCA:
- 3. 7.1 RC DRAIN TANK (QUENCH TANK) PRESSURE (RUPTURE DISK MAY BE BLOWN).
- 3. 7. 2 INCREASING REACTOR BUILDING SUMP LEVEL.
3, 7. 3 INCREASING REACTOR BUILDING TEMPERAIURE.
- 3. 7. 4 INCREASING REACTOR BUILDING PRESSURE.
- 3. 7.5 INCREASING RADIATION MONITOR READINGS INSIDE CONTAINMENI
- 3. 7. 6 REACTOR COOLANT SYSTEM TEMPERATURE BECOMING SATURATED RELATIVE TO THE RCS PRESSURE.
- 3. 7. 7 HOT LEG TEMPERATURE EQUALS OR EXCEEDS PRESSURIZER TEMPERATURE.
- 3. 8 HPI COOLING REQUIREMENTS COULD DEPLETE THE BORATED WATER STORAGE TANK, AND INITIATION OF LPI FLOW FROM THE REACTOR BUILDING SUMP TO THE HPI PUMPS WOULD BE REQUIRED.
3.9 ALTERNATE INSTRUMENT CHANNELS SHOULD BE CHECK E AS AVAILABLE TO CONFIRM KEY PARAMETER READINGS (IE, SYSTEM TEMPERATURES, PRESSURES AND PRESSURIZER LEVEL).
3.10 MAINTAIN A TEMPERATURE VERSUS TIME PLOT AND A CORRESP0rDING TEMPERATURE PRESSURE PLOT ON A SATURATION DIAGRAM. THESE PLOTS WILL MAKE IT POSSIBLE
'IO TRACK THE PLANT'S CONDITION THROUGH PLANT C00LDOWN. PRIMARY TEMPERA-TURE AND PRESSURE WILL DECREASE ALONG THE SATURATION CURVE UNTIL SUBC00 LED CONDITIONS ARE ESTABLISHED. THIS WILL BE INDICATED BY PRIMARY SYSTEM PRESSURE NO LONGER FOLLOWING THE SATURATION CURVE, AS PRIMARY SYSTEM TEMPEPiTURE DECREASES. WHEN THIS OCCURS, PRIMARY SYSTEM PRESSURE SHOULD BE CfNTROLLED BY ADJUSTING HPI FLOW, TO MAINTAIN 50'F SUBC00 LING. THE DEGREE OF SUBC00 LING BEYOND 50*F SHALL BE CONTROLLED WITHIN THE LIMITS DEFINED IN SECTION 3.4.
2 o
3.IL COMPONENT COOLING WATER (CCW) AND SEAL INJECTION SHOULD BE MAINTAINED TO THE RC PUMPS TO INSURE CONTINUED SERVICE OR THE ABILITY TO RESTART THE PUMPS AT A LATER TIME.
1054129 o
3.11.1 IF CCW IS LOST AND Tile RC PUMPS ARE OPERATIVE, CCW MUST BE RESTORED 1:ITilIN 30 MI!!UTES OR THE RC PUMPS MUST BE MANUALLY TRIPPER.
o 3.11.2 IF Tile RC PUMPS ARE TRIPPED FOR ANY REASON, SEAL INJECTION SHOULD BE MAINTAINED TO ENSURE LONG TERM SEAL INTEGRITY.
4.0 FOLLOWUP ACTIONS 4.1 IDENTIFICATION AND EARLY CONTROL 4.1.1 IF HPI HAS INITIATED BECAUSE OF LOW PRESSURE, CONTROL HFI IN ACCORDANCE WITil STEP 3.2.
4.1.2 IF BOTH HPI TRAINS HAVE NOT ACTUATED ON ESFAS SICNAL, START SECOND HPI TRAIN IF POSSIBLE. BALANCE HPI FLOWS.
4.1.3 IF RC PRESSURE DECREASES CONTINUOUSLY, VERIFY THAT CORE FLOOD TANKS (CFTs) AND LOW PRESSURE INJECTION (LPI) HAVE, ACTUATED AS NEEDED, AND BALANCE LPI.
o 4.1.4 IF CAUSE FOR COOLDOWN/DEPRESSURIZATION IS DETERMINED TO BE DUE TO A NON-LOCA OVERC00 LING EVENT AND THE RCS IS AT LEAST e
50 F SUBC00 LED THEN PROCEED TO SECTION 4.4.
4.1.5 ATTD!PT TO trCATE AND ISOLATE LEAK IF POSSIBLE.
LETDOWN WAS ISOLATEE IN STEP 2.1.
OTHER ISOLATABLE LEAKS ARE PORV (CLOSE BLOC". VALVE) AND BETWEEN VALVES IN SPRAY LINE (CLOSE SPRAY AND Br.0CK VALVE).
4.1.6 DETERMINE AJAILABILITY OF REACTOR COOLANT PUMPS (RCPs) AND MAIN AND ArXILIARY FEEDWATER SYSTDIS.
IF FEEDWATER IS NOT AVAILALLE 30 TO 4.2.
IF FEEDWATER IS AVAILABLE GO TO 4.3.
4.2 ACTIONS IF FEEDWATER IS NOT AVAILABLE 4.2.1 THROUGHOUT THE FOLLOWING STEPS MAINTAIN MAXDfUM HPI FLOW AND RESTORE FEEDWATER AS SOON AS POSSIBLE.
o 4.2.2 IF RCPs ARE OPERATING, GO TO ONE PUMP PER LOOP.
IF RCPs ARE NOT OPERATING, CO TO STEP 4.2. 5 BELOW.
4.2.3 IF RCS PRESSURE INCREASES, OPEN PORV AND LEAVE OPEN.
NOTE:
IF THE PORV CANNOT BE ACTUATED, THE SAFETIES WILL RELIEVE PRESSURE.
o Revised 1054126 o
CLOSE PORV OR BLOCK VALVE, IF POSSIBLE, AND PROCEED TO STEP 4.3.2.
NOTE: IF TIE PORV CANNOT BE ACTUATED, THE SAFETIES WILL RELIEVE PRESSURE.
ON Tile OPERATE RANGE, CLOSE PORY OR BLOCK VALVE, IF POSSISLE.
NOTE: OTSG LEVEL SHOULD BE MONITORED PERIODICALLY DURING THE FILL PROCESS.
LEVELS > 95% ON THE OPERATING RANGE MUST BE AVOIDED TO PRECLUDE FEEDWATER CARRYOVER TO THE STEAMLINES.
- 4. 2.7 VERIFY NATURAL CIRCULATION IN THE RCS BY OBSERVING:
g 4.2.7.1 COLD LEG TEMPERATURE IS SATURATION TEMPERATURE OF SECONDARY SIDE PRESSURE WITHIN APPROXIMATELY 5 MINUTES.
4.2.7.2 PRIMARY AT (T110T - TCOLD) BEC01ES CONSTANT 4.2.8 GO TO STEP 4.3.4.1.
4.3 ACTIONS WITH FEEDWATER AVAILABLE TO ONE OR BOTH GEN'ERATORS 4.3.1 MAINTAIN ONE RCP RUNNING PER LOOP (STOP OTHER RCPs).
IF NO RCPs OPERATING (DUE TO A LOSS OF OFFSITE POWER OR DUE TO MANUAL SECURDENT PER SECTION 2.0), CO TO STEP 4.3.4 BELOW.
4.3.2 ALLOW RCS PRESSURE TO STABILIZE.
4.3.3 ESTABLISH AND MAINTAIN OTSG COOLING BY ADJUSTING STEAM PRESSL"-E VIA TURBINE BYPASS AND/0R ATMOSPIERIC DUMPS.
C00LDOWN AT 100 F PER HOUR TO ACHIEVE AN RC PRESSURE OF 250 PSIG. REFER TO PRE-CAUTION 3.10 FOR DEVELOPMENT OF TEMPERATURE AND PRESSURE PLOIS.
ISOLATE CORE FLOOD TANKS WHEN 50 F SUBC00 LING IS ATTAINED E;D RC PRESSURE IS LESS THAN 700 PSIG.
GO INTO LPI COOLING PER APPENDIX A.
o 4.3.4 IF RCPs ARE NOT OPERATING:
4.3.4.1 ESTABLISH AND CONTROL OTSG LEVEL TO 95% ON THE OPERATE RANGE. VERIFY TIE CONDITIONS IN STEP 4.2.7..
NOTE: OTSG LEVELS GREATER TilAN 95% ON THE OPERATING RANGE MUST BE AVOIDED TO PRECLUDE FEEDWATER CARRYOVER INTO TIIF. STEAMLINES.
541$
IF IT BEGINS INCREASING, so TO STEP 4.3.4.4.
~
4.3.4.3 PROCEED WITH A CONTROLLED C00LDOWN AT 100 F/HR BY CONTROLLING STEAM GENERATOR SECONDARY SIDE PRESSURE.
MONITOR RC PRESSURES AND TEMPERATURES DURING C00LDOWN AND PROCEED AS INDICATED BELOW:
4.3.4.3.1 IF RC PRESSURE CONTINUES TO DECREASE, FOLLOWING SECONDARY SIDE PEESSURE DECREASES AND WITH PRIMARY SYSTEM TEMPERATURES INDICATING SATURATED CONDITIONS, CONTINUE COOLDOWN UNTIL AN RC PRESSURE OF 150 PSI IS REACHEE, AND PROCEED TO STEP A.4 0F APPENDIX A.
4.3.4.3.2 IF RC PRESSURE STOPS DECREASING IN RESPONSE TO SECONDARY SIDE PRESSURE DECREASE AND REACTOR SYSTEM BECOMES SUBC00 LED, CHECK TO SEE THAT TE FOLLOWING CONDITIONS ARE BOTH SATISFIED:
A) ALL HOT AND COLD LEG TEMPERATURES ARE BELOW THE SATURATION TEMPERATURE FOR TE EXISTING RCS PRESSURE.
AND B) ECS HOT LEG TEMPERATURES ARE NOT MORE THAN 50 F HOITER THAN ThE STEAM GENERATOR SECONDARY SIDE SATURATION TEMPERATURE.
IF THESE CONDITIONS ARE SATISFIED AND REMAIN SATISFIED, CONTINUE COOLDOWN TO ACHIEVE AN RCS TEMPERATURE (COLD LEG) 0F 280 F, AND PROCEED TO STEP A.1 0F APPENDIX A.
NOTE: IF TE CONDITIONS ABOVE ARE MET BELOW 700 PSIG, THE CORE FLOOD TANKS SHOULD BE ISOLATED.
NOTE: IF THE PRIMARY SYSTEM IS 50 F SUBC00 LED IN BOTH HOT AND COLD LEGS AND PRIMARY 105412h
8-SYSTEM PRESSURE IS ABOVE 250 PSIC, STARTING A REACTOR COOLANT PUMP IS PER-HISSIBLE. IF SYSTEM DOES NOT RETURN TO AT LEAST 50 F SUBC00 LING IN TWO MINUTES, TRIP PUMPS.
IF FORCED CIRCULATION IS ACHIEVED, PROCEED TO STEP 4.3.
e 105412f 4.3.4.3.3 IF RC PRESSURE STOPS DECREASING AND THE CONDITIONS OF 4.3.4.3.2 ARE NOT MET OR CEASE TO BE MET OR IF RC PRESSURE BEGINS TO INCREASE, THEN PROCEED TO STEP 4.3.4.4 BELOW.
4.3.4.4 RESTORE RCP FLOW (ONE PER LOOP) WHEN POSSIBLE PER THE INSTRUCTIONS BELOW.
IF RC PUMPS CANNOT BE OPERATED AND PRESSURE IS INCREASING, GO TO STEP 4.3.4.6.
4.3.4.4.1 IF PRESSURE IS INCREASING, STARTING A PUMP IS PERMISSIBLE AT RC PRESSURE GREATER THAN 1600 PSIG.
4.3.4.4.2 IF REACTOR COOLANT SYSTEM PRESSURE EXCEEDS STEAM GENERATOR SECONDARY PRESSURE BY 600 PSIG OR MORE " BUMP" ONE REACTOR COOLANT PUMP FOR A PERIOD OF APPROXIMATELY 10 SECONDS (PREFERABLY IN OPERABLE STEAM GENERATOR LOOP). ALLOW REACTOR C00LAh7 SYSTEM PRESSURE T0STABILi.'2.
CONTINUE C00LDOWN.
IF REACTOR COOLANT SYSTEM PRESSURE AGAIN EXCEEDS SECONDARY PRESSURE BY 600 PSI, WAIT AT LEAST 15 MINUTES AND REPEAT THE PUMP " BUMP".
BUMP ALTERNATE PUMPS SO THAT NO PUMP IS BUMPED h0RE THAN ONCE IN AN HOUR.
THIS MAY BE REPEATED, WITH AN INTERVAL OF 15 MINUTES, UP TO 5 TIMES. AFTER THE FIFTH
" BUMP," ALLOW THE REACTOR COOLANT PUMP TO CONTINUE IN OPERATION.
4.3.4.4.3 IF PRESSURE HAS STABILIZED FOR GREATER THAN ONE HOUR, SECONDARY PRESSURE IS LESS THAN 100 PSIG AND PRIMARY PRESSURE IS GREATER THAN 250 PSIG, BUMP A PUMP, WAIT 30 MINUTES, AND START AN ALTERNATE PUMP.
105412f 4.3.4.5 IF FORCED FLOW IS ESTABLISHED, GO TO STEP 4.3.3.
4.3.4.6 IF A REACTOR COOLANT PUMP CANNOT BE OPERATED Ah3 REACTOR COOLANT SYSTEM PRESSURE REACHES 2300 PSIG.
OPEN PRESSURIZER PORV TO REDUCE RE'CTOR COOLANT A
SYSTEM PRESSURE. RECLOSE PORV WHEN RCS PRESSURE FALLS TO 100 PSI ABOVE THE SECONDARY PRESSURE.
REPEAT IF NECESSARY.
IF PORV IS NOT OPERABLE, PRESSURIZER SAFETY VALVES WILL RELIEVE OVFRPRESSURE.
4.3.4.7 MAINTAIN RC PRESSURE AS INDICATED IN 4.3.4.6 IF PPISSURE INCREASES. MAINTAIN THIS COOLING MODE UNTIL AN RC PUMP IS STARTED OR STEAM GENERATOR COOLING IS ESTABLISHED AS INDICATED BY ESTABLISHINC CONDITIONS DESCRIBED IN 4.3.4.3.1 OR 4.3.4.3.2.
VHEN THIS OCCURS, PROCEED AS DIRECTED IN THOSE STEPS. GO TO STEP 4.3.2 IF FORCED FLOW IS ESTABLISHED.
105412q 4.4 NON-LOCA OVERC00 LING TRANSIENT WITH FEEDWATER A VAILABLE 4.l 1 IMMEDIATELY RESTART A RC PUMP IN EACH LO)P IF THE RCS IS 50 F SUBC00 LED.
4.4.2 RESTORE CONTROL OF FEEDWATER FLOW AND GENERATOR LEVEL AND CONTROL STEAM PRESSURE VIA TURBINE BYPASS OR ATMOSPHERIC DUMP VALVES TO STABILIZE OR CONIROL PLANT HEATUP.
NOTE: CONSIDERABLE HPI MAY HAVE BEEN ADDED TO THE RCS.
THEREFORE, TO PREVENT RCS FROM GOING S LID, THE ABOVE ACTION MAY BE NECESSARY.
4.4.3 AS LONG AS THE RCS IS MAINTAINED 50 F SUBC00 LED, THROTTLE HPI AND LETDOWN FLOW TO MAINTAIN PZR LEVEL AT 100 INCHES.
4.4.4 USING TURBINE BYPASS VALVES AND FEEDWATER SYSTEM, CONTROL STEAM GENERATORS AS NEEDED TO LIMIT PLANT HEATUP UNTIL RC PRESSURE CONTROL CAN BE RE-ESTABLISHED WITH THE PRESSURIZER.
NOTE: COLD RCS WATER MAY tiAVE BEEN ADDED TO THE PRESSURIZER; THEREFORE, A PERIOD OF TIME MAY ELAPSE BEFORE NORMAL RC PRESSURE CONTROL CAN BE ESTABLISHED WITH THE PRESSURIZER BEATERS.
4.4.5 ONCE PRESSURE CONTROL IS RE-ESTABLISHED, USE NORMAL HEATUP/.
COOLDOWN PROCEDURE TO ESTABLISH DESIRED PLANT CONDITIONS.
APPENDIX A LPI COOLING A.1 DETERMINE IF PRIMARY COOLANT IS AT LEAST 50 F SUBC00 LED.
IF NOT GO TO STEP A.3.
A.l.1 START LPI PUMPS.
IF BOTH PUMPS ARE OPERABLE GO TO STEP A.2.
FOR ONE LPI PUMP OPERABLE MAINTAIN OTSG COOLING AS FOLLOWS. THE OPERABLE LPI PUMP WILL BE USED TO MAINTAIN SYSTEM INVENTORY.
A.1.2 OBTAIN PRIMARY SYSTEM CONDITIONS OF 280 F AND 250 PSIG.
A.l.3 ALIGN THE DISCHARGE OF THE OPERABLE LPI PUMP TO THE SUCTIONS OF THE HPI PUMPS AND TAKE SUCTION FROM THE BWST.
IF THE BWST IS AT THE LOW LEVEL ALARM, ALIGN LPI SUCTION FROM THE RB SUMP AND SHUT SUCTION FROM BWST.
A.l.4 START THE OPERABLE LPI PUMP SPECIFIED ABOVE. THE HPI-LPI SYSTEMS WILL NOW BE IN " PIGGY BACK" AND HPI FLOW IS MAINTAINING SYSTEM PRESSURE.
A.l.5 GO TO SINGLE RC PUMP OPERATION.
A.l.6 WHEN THE SECOND LPI PUMP IS AVAILABLE ALIGN IT IN THE DECAY HEAT HODE AND COMMENCE DECAY HEAT REMOVAL.
(DECAY HEAT SYSTEM FLOW GREATER THAN 1000 GPM). SECURE REMAINING RC PUMP WHEN DECAY HEAT REMOVAL IS ESTABLISHED.
CAUTION: VERIFY THAT ADEQUATE NPSH EXISTS FOR THE DECAY HEAT PUMP IN THE DH REMOVAL MODE.
IF INADEQUATE, TRANSFER TO LPI MODE.
A.I.7 REDUCE REACTOR COOLANT PRESSURE TO 150 PSIG BY THROTTLING HPI FLOW.
CONTROL RC TEMPERATURE USING THE DECAY HEAT SYSTEM COOLER BYPASS TO HAINTAIN SYSTEM PRESSURE AT LEAST 50 PSI ABOVE SATURATION PRESSURE, TO ASSURE THAT NPSH REQUIREMENTS FOR THE DECAY HEAT PUMP ARE MAINTAINED.
105413$
A.1. f.;
SECURE THE HPI PUMP AND SHIFT THE LPI PUMD SijPPLT ING IT TO THE LPI INJECTION MODE.
A.1.9 REDUCE REACTOR COOLANT TEMPERATURE TO 100 F BY CONTROLLING THE DECAY HEAT SYSTEM COOLER BYPASS.
NOTE:
IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, RETURN TO OTSG COOLING USING NATURAL CIRCULATION OR ONE REACTOR C001 ANT PUMP (A1).
df..COOLDOWN ON TWO LPI PUMPS A.2.1 MAINTAIN RCS PRESSURE AT 250 PSIG AND REDUCE RCS TEMPERATURE TO 280 F.
A.2.2 ALIGN ONE LPI PUMP IN THE DECAY HEAT REMOVAL MODE.
A.2.3 SECURE ONE RC PU:!P IF IWO ARE OPERATING.
.A.2.4 START THE DECAY HEAT PUMP IN THE DECAY HEAT REMOVAL MODE, AhT WHEN DECAY HEAT SYSTEM FLOW IS GREATER THAN 1000 GPM, SECURE THE RUNNING RC PUMP.
A.2.5 REDUCE RC PRESSURE TO 150 PSIG BY THROTTLING HPI FLOW.
CONTROL RC TEMPERATURE TO MAINTAIN AT LEAST 50 PSI MARGIN TO SATURATION PRESSURE.
A.2.6 START THE SECONP LPI PUMP IN THE LPI INJECTION MODE.
SECURE HPI PUMP.
A.2.7 SHIFI LPI SUCTION FROM THE BWST TO THE REACTOR BUILDING SUMP WHEN SUFFICIENT NPSH IS AVAILABLE.
NOTE: THIS IS DESIRABLE TO AVOID UNNECESSARY QUANTITIES OF WATER IN CONTAINMENT.
A.2.8 REDUCE REACTOR COOLANT TEMPERATURE TO 100 F BY CONTROLLING THE DECAY HEAT SYSTEM COOLER BYPASS.
NOTE: IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, RETURN TO OTSG COOLING USING NATURAL CIRCULATION OR CNE RC PUMP PER A.1.
1054 13L
A.3 COOL DOWN RC SYSTEM AT SATURATION A.3.1 MAINTAIN RC PRESSURE AT 250 PSIG.
A.3.2 ALIGN ONE LPI PUMP TO SUCTION OF THE HPI PUMPS AND THE SUCTION TO THE REACTOR BUILDING SUMP.
(SHUT BWST SUCTION VALVE FOR THIS PUMP.)
A.3.3 WHEN THE BWST LEVEL REACHES THE LO-LO LEVEL LIMITS, START THE LPI PUMP AND SHUT THE HPI PUMP SUCTION FROM THE BWST.
A.3.4 WHEN PRIMARY SYSTEM TEMPERATURE BECOMES SUBC00 LED BY AT LEAST 50 F, GO TO A.l.l.
A.4 COOLDOWN WITHOUT REACTOR COOLANT PUMPS A.4.1 RCS INITIAL CONDITIONS ARE: PRESSURE 150 PSI, TEMPERATURE AT SATURATION.
A.4.2 ALIGN LOW PRESSURE INJEC' ION SYSTEM FOR SUCTION FROM REACTOR BUILDING SUMP AND PLACE INTO SERVICE.
A.4.3 BALANCE LPI INJECTION AND CONTROL RC TEMPERATURE WITH DECAY HEAT COOLERS.
A.4.4 ISOLATE CORE FLOOD TANKS.
A.4.5 GO TO STEP A.1.1 AND FOLLOW THE PROCEDURE GIVEN THERE, IGNORING THE INSTRUCTIONS RELATING TO RC PUMP OPERATION.
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Part II: Small Break Phenomena - Description of Plant Behavior 1.
Introduction A loss-of-coolant accident is a condition in which liquid inventory is lost from the reactor coolant system.
Due to the loss of mass from the reactor coolant system, the most significant short-term symptom of a loss-of-coolant eccident is an uncontrolled reduction in the reactor coolant system pressure.
The reactor protection system is designed to trip the reactor on low presscre.
This should occur before the reactor coolant system reaches saturation conditions.
The existence of saturated conditions within the reactor system is the principal longer-term indication of a LOCA and requires special consideration in the development of operating procedures.
Following a reactor trip, it is necessary to remove decay heat from the reactor core to prevent damage.
However, so long as the reactor core is kept covered with cooling water, core damage will be avoided. The ECCS systems are designed to respond automatically to low reactor coolant pressure conditions and take the inital actions to protect the reactor core. They are sized to provide sufficient water to keep the reactor core covered even with a single failure in the ECCS systems.
Subsequent operator actions are required ultimately to place the plant in a long-term cooling mode. The overall objective of the automatic emergency core cooling system and the followup operator actions is to keep the reactor core cool.
A detailed discussion of the small break LOCA phenomenalogy is presented in this section. This discussion represents Part II of the operating procedure guidelines for the development of detailed operating procedures. Part I presents the more detailed step-by-step guidelines.
1054 13 The response of the primary system to a small break will greatly depend on break size, its location in the system, operation of the reactor coolant pumps, the number of ECCS trains functioning, and the availability of secondary side cooling.
RCS pressure ani! pressurizer level histories for various combinations of parameters are presented in order to indicate the wide range of system behavior which can occur for small LOCA's.
2.
Impact of RC Pump Operation on a Small LOCA With the RC pumps operating during a small break, the steam and water will remain mixed during the transient. This will result in liquid being discharged out the break continuously.
Thus, the fluid in the RCS can evolve to a high void fraction as shown in Figure 1.
The maximum void fraction that the system evolves to, and the time it occurs, is dependent on the break size and location.
Continued RC pump operation, even at high system void fractions, will provide sufficient core flow to keep cladding temperatures within a few. degrees of the saturated fluid temperature.
Since the RCS can evolve to a high void fraction for certain small breaks with the RC pumps on, a RC pump trip by any means (i.e., loss of offsite power, equipment failure, etc.) at a high void fraction during the small break transient may lead to inadequate core cooling. That is, if the RC pumps trip at a time period when the system void fraction is greater than approximately 70%, a core heatup will occur because the amount of water left in the RCS would not be sufficient to keep the c6re covered. The cladding temperature would increase until core cooling is re-established by the ECC systems.
For certain break sizes and times of RC pump trip, acceptable peak cladding temperatures during the event could not be assured and the core could be damaged. Thus, prompt operator action to trip the RC pumps upon receipt of a low pressure ESFAS signal is required in order 1054138
' to ensure that adequate core cooling is provided.
Following the RC pump trip, the small break transient will evolve as described in the subsequent sections.
3.
Small Breaks with Auxiliary Feedwater There are four basic classes of break response for small breaks with auxiliary feedwater.
These are:
1.
LOCA large enough to depressurize the reactor coolant system 2.
LOCA which stabilizes at approximately secondary side pressure 3.
LOCA which may repressurize in a saturated condition 4.
Sr..all LOCA which stabilizes at a primary system greater than secondary system pressure The system transients for these breaks are depicted in Figure 2.
3.1 LOCA Large Enough to Depressurize Reactor Coolant System: Curves 1 and 2 of Figure 2 show the response of RCS pressure to breaks that are large enough in combination with the ECCS to depressurize the system to a stable low pressure.
ECCS injection easily exceeds core boil-off and ensures core cooling.
Curves 1 and 2 of Figure 3 show the pressurizer level transient.
Rapidly falling pressure causes the hot legs to saturate quickly. Cold leg temperature reaches saturation somewhat later as RC pumps coast down or the RCS depressurizes below the secondary side saturation pressure. Since these breaks are capable of depressurizing the RCS without aid of the steam generators, they are essentially unaffected by the availability of auxiliary feed;iater. Upon receipt of a low pressure ESFAS signal, the operator must trip all RC pumps and erify that all ESFAS actions have been completed. The operator must also balance HPI flows such that flow is available through all HPI injection nozzles even if only one HPI is available. The operator should also balance LPI flows, should the system be actuated, to ensure flow through both lines.
The operator needs to take no further actions to bring the system to a safe shutdown 1 F4 138 condition. Rapid'depressurization of the steam generators would only act to accelerate RCS depressurization.
It is, however, not necessary.
Restarting of the RC pumps is not desirable for this class of break.
Long-term cooling will require the operator to shift the LPI pump suction to the reactor building sump.
Curve 3 3.2 LOCA Which Stabilizes at Approximately Secondary Side Pressure.
of Figure 2 shows the pressure transient for a break which is too small in combination with the operating HPI to depressurize the RCS. The steam generators are, therefore, necessary to remove a portion of core decay heat. Although the system pressure will initially stabilize near the secondary side pressure, RCS pressure may eventually begin falling as the decay heat level decreases. Curve 3 of Figure 3 shows pressurizer level behavior. The hot leg temperature quickly equalizes to the saturated temperature of the secondary side and controls primary system pressure at saturation. The cold leg temperature may remain slightly subcooled.
If the HPI refills and repressurizes the RCS, the hot legs can become subcooled. The immediate operator action is to trip the RC pumps upon receipt of the low pressure ESFAS signal and then verify ESFAS functions.
The operator must then balance HPI in order to ensure flow through each high pressure injection line.
Followup action by the operator is to raise the emergency feedwater level to 95% on the operating range and check for established natural circulatioa. This is done by gradually depressurizing the steam generators.
If this test fails, intennittent bumping of a RC pump should be perfonred as soon as one is available. Continued depressurization of the steam generators with natural circulation leads to cooling and depressurization of the RCS. The operator's goal is to depressurize the RCS to a pressure that enables the ECCS to exceed core boil-off, possibly refill the RCS, and to ultimately establish long-term cooling.
10S4 iM 3.3 LOCA Which May Repressurize in a Saturt M i Condition.
Curve 4 of Figure 2 shows the behavior of a small break that is too small, in combination with the HPI, to depressurize the primary system, Although steam generator feedwater is available, the loss of primary system coolant and the resultant RCS voiding will eventually lead to interruption of natural circulation.
This is followed by gradual repressurization of the primary system.
It is possible that the primary system could repressurize as high as the pressurizer safety valve setpoint before the pressure stabilizes.
This is shown by the dashed line in Curve 4.
Once enough inventory has been lost from the primary system to allow direct steam condensation in the regions of the steam generators contacting secondary side coolant, the primary system is forced to depressurize to the saturation pressure of the secondary side.
Since the cooling capabilities of the secondary side are needed to continue to remove decay heat, RCS pressure will not fall below that on the secondary side.
HPI flow is sufficient to replace the inventory lost to boiling in the core, and condensation in the steam generators removes decay heat energy. The RCS is in a stable thermal condition and it will remain there until the operator takes further action.
The pressurizer level response is characterized by Curve 3 of Figure 3 during the depressurizatior.,
and Curve 4 of Figure 3 during the temporary repressurization phase. The dashed line indicates the level behavior if pressure is forced up to-the pressurizer safety valve setpoint. During this transient, hot leg temperature will rapidly approach saturation with the* initial system depressurization, and it will remain saturated during the whole transient.
C Id leg temperature will approach saturation as circulation is lost, but may remain slightly subcooled during the repressurization phase of the transient.
Later RCS depressurization could cause the cold leg temperatures to reach saturation, Subsequent refilling of the primary 105413$
. system by the HPI might cause temporary interruption of steam condensation in the steam generator as the primary side level rises above the secondary side level.
If the depressurization capability of the break and the HPI is insufficient to offset decay heat, the primary system *will once more repressurize.
This decreases HPI flow and increases loss through the break until enough RCS coolant is lost to once more allow direct steam condensation in the steam generator.
This cyclic behavior will stop once the HPI and break can balance decay heat or the operator takes some action.
The operator's inmediate action is to trip the RC pumps upon receipt of the low pressure ESFAS signal and verify the completion of all ESFAS functions.
The operator should then balance HPI flow.
Following that, he should raise the steam generator level to 95% of the operating range and check for natural circulation.
If it is positive, he should depressurize the steam generators, cool and depressurize the primary system, and attempt to refill it and establish long-term cooling.
If the system fails to go into natural circulation, he should open the PORV long enough to bring and hold the RCS near the secondary side pressure.
Once natural circulation is established or a RC pump can be bumped, he will be able to continue depressurizing the RCS with the steam generators and establish long-term cooling.
3.4 Small LOCA Which Stabilizes at p> Psec. Curve 5 of Figure 2 shows the behavior of the RCS pressure to a break for which high pressure injection isbeingsuppliedandexceedstheleakflowbeforeth[epressurizerhas emptied. The primary system remains subcooled and natural circulation to the steam generator removes core decay heat. The pressurizer never empties and continues to control primary system pressure. The operator needs to trip the RC pumps and ensure that ESFAS actions have occured. Throttling of HPI 0
is permitted only after RCS subcooling of 50 F has been established, the pressurizer has refilled, and natural or forced circulation has been } {j 5 4 14#
verified.
A restart of the RC pumps under these conditions is desirable for plant control.
3.5 Small Breaks in Pressurizer. The system pressure transient for a small break in the pressurizer will behave in a manner similarsto that previously discussed.
The initial depressurization, however, will be more rapid as the initial inventory loss is entirely steam.
The pressurizer level response for these accidents will initially behave like a very small break without auxiliary feedwater.
The initial rise in pressurizer level shown in Figure 4 will occur due to the pressure reduction in the pressurizer and an insurge of coolant into the pressurizer from the RCS. Once the reactor trips, system contraction causes a decreasing level in the pressurizer.
Flashing will ultimately occur in the hot leg piping and cause an insurge into the pressurizer.
This ultimately fills the pressurizer.
For the remainder of the transient, the pressurizer will remain full. Toward the later stages of the transient, the pressurizer may contain a two-phase mixture and the indicated level will show that the pressurizer is only partially full.
Except for closing the PORV block valve, operator actions and system response are the same for these breaks as for similar breaks in the loops.
4.
Small Greaks Without Auxiliary Feedwater There are three basic classes of break response for small breaks without auxiliary feedwater.
These are:
1.
Those breaks capable of relieving all decay heat via the break.
2.
Breaks that relieve decay heat with both the HPI injection and via the break.
I 3.
Breaks which do not automatically actuate the HPI and result in system repressurization.
The system pressure transients for these breaks are depicted in Figure 5.
105414$
4.1 LOCA's Large Enough to Depressurize Reactor Coolant System.
For Class 1 (curve 1 of Figure 5), RC system pressure decreases smoothly throughout the transient.
For the larger breaks in this class, CFT actuation and LPI injection will probably occur.
ForthSsmaller breaks of this class only, CFT actuation will occur.
Auxiliary feedwater injection is not necessary for the short-term stabilization of these breaks.
The pressurizer level for this transient rapidly falls off scale. Operator action and plant response are similar to those described for this class of breaks with a feedwater supply.
4.2 LOCA's Which Reach a Semi-Stabilized State.
For Class 2 (Curve 2 of Figure 5) breaks, the RC pressure will rapidly reach the low pressure ESFAS trip signal (about two to three minutes). With the HPI's on, a slow system depressurization will be established coincident with the decrease in core decay heat.
No CFT actuation is expected.
Auxiliary feedwater is not necessary for the short-term stabilization of these breaks. The pressurizer level for this transient rapidly falls off scale.
The operator needs to trip the RC pumps upon the low pressure ESFAS signal, verify completion of all ESFAS functions, and try to establish secondary side cooling.
Balancing of the HPI must also be perfonned.
If steam generator feedwater cannot be obtained and RCS pressure is increasing, the operator should open the PORV and provide all the HPI and makeup capability possible.
The goal is to depressurize and cool the core with the ECCS, the PORV, and the break.
If secondary side cooling is again established, the operator should verify natural circulation, and 1[f unavailable, bump a RC pump to complete RCS cooldown with the steam generators.
At this point, the PORV can be closed, the system refilled, and long-term cooling established.
1054 140~
4.3 Small LOCA's Which do not Actuate the ESFAS. Automatic ESFAS actuation will not occur for Class 3 (Curve 3 of Figure 5) breaks.
Once the SG secondary side inventory is boiled off, system repressurization will occur as the break is not capable of removing all the decay heat being generated in the core.
System repressurization to the PORV or the pres-surizer safety valves will occur for smaller breaks in this class.
For the "zero" break case, repressurization to the PORV will occur in the first five minutes. Operator action is required within the first 20 minutes to ensure core coverage throughout the transient.
For the 177-FA 1cwerad-1990 plants, this action can be either manual actuation of the auxiliary feedwater system or the HPI system.
The establishment of auxiliary feedwater will rapidly depressurize the RCS to the ESFAS actuction pressure, and system pressure will stabilize at either the secondary side SG pressure or at a pressure where the HPI equals the leak rate.
Upon receipt of the low pressure ESFAS signal, the operator must trip the RC pumps.
For the raised loop Davis-Besse plant (which has a safety-grade auxiliary feedwater system) operator actio:. is necessary at some time greater than 20 minutes (approximately 40 minutes) as there is increased inventory in the loops that is available to drain into the reactor vessel.
However, because the plant is equipped with low shut-off head HPI pumps, the operator must establish auxiliary feedwater in order to depressurize the RCS.
For the Class 3 breaks, pressurizer level response will be as shown in Figure 6.
The minimum refill time for the pressurizer is that for the "zero" break and is shown in Figure 6.
After initially drawing inventory from the pressurizer, the system repressurization will cause the pressurizer level to increase, possibly to full pressurizer level. Once the operator action to restore auxiliary feedwater has been taken, the systT0 5 4 1 d depressurization will result and cause an outsurge from the pressurizer.
Complete loss of pressurizer level may result.
For the smaller breaks in Class 3 which result in a system repressurization following the actuation of the HPI system, pressurizer level will increase and then stabilize.
Without auxiliary feedwater, both the hot and cold leg temperatures will saturate early in the transient and, for the Class 1 and 2 breaks, will remain saturated.
For the Class 3 breaks, once auxiliary feedwater is established, the cold leg temperatures will rapidly decrease to apprcximately the sautration temperature corresponding to the SG secondary side pressure and will remain there throughout the remainder of the transient.
Hot leg temperatures will remain saturated throughout the event.
The operator needs to manually initiate all ESFAS actions, balance HPI flow, and attempt to restore secondary' side cooling.
In the meantime, he should actuate the makeup pump and open the PORV in order to cool the core and limit the RCS repressurization.
Once feedwater is available, he can close the PORV and continue the RCS cooldown and depressurization with the steam generators.
If natural circulation has not been established, he can bump a RC pump to cause forced circulation. The goal is to depressurize to where the ECCS can refill the RCS and guarantee long-tenn cooling.
4.4 Small Breaks in Pressurizer. See the writeup for small breaks in pressurizer with feedwater.
Small breaks in the pressurizer will differ from those in the loops in the same manner as those previously described in the section addressing small breaks in the pressurizer with auxiliary feed.,
5.
Transients with Initial Response Similar to a Small Break Several transients give initial alarms similar to small breaks. These transients will be distinguished by additional alarms and indications or subsequent system response.
Overcooling transients such as steam line breaks, increased feedwater 105414f flow, and steam generator overf'11 can cause RCS pressure decreases with low-pressure reactor trip and ESFAS actuation.
But steam line breaks actuate low steam pressure alams for the affected steam generator, and steam generator overfills result in high steam generator level indications.
The overcooling transients will repressurize the primary system because of HPI actuation, and will return to a subcoolcd condition during repres-surization.
The imediate actions for both overcooling and small break transients are the sana, including tripping of the RC pumps.
The operator will recognize overcooling events during repressurizatf or.,
if not sooner, and is instructed to throttle HPI and restart the RC pumps, if subcooled conditions are established, by the small break operating instructions.
A loss-of-feedwater transient will result in a high reactor system pressure alam but does not give an ESFAS actuation alarm.
A loss of integrated control system power transient starts with a high RC pressure trip. After the reactor trip, this becomes an overcooling transient and will give low reactor system pressure and possible ESUS actuation.
Steam generator levels remain high and the system becomes subcooled during repressurization.
Design features of the B&W NSS provide automatic protection during the early part of small break transients, thereby providing adequate time for small breaks to be identified and appropriate action taken to protect the system. The only prompt manual operator action required is to trip the RC pumps once the low pressure ESFAS signal is reached.
6.
Transients that might Initiate a LOCA There are no anticipated transients that might initiate a 1.0CA since the PORV has been reset to a higher pressure and will not actuate during anticipated transients such as loss of main feedwater, turbine trip, or loss of offsite p ver.
1054 14 However, if the PORV should lift and fail to reset, there are a number of indications which differentiate this transient from the anticipated transients identified above. These include:
o ESFAS actuation o Quench tank pressure / temperature alanns o Saturated primary system o Rising pressurizer level These additional signals will identify to the operator that in addition to the anticipated transient, a LO A has occurred.
In the unlikely event that small breaks other than a malfunctioning PORV occur after a transient, they can be identified by initially decreasing RCS pressure Small and convergence to saturation conditions in the reactor coolant,,
break repressurization, if it occurs, will follow saturation conditions.
By remaining aware of whether the reactor coolant remains subcooled or becomes saturated af ter transients, the operator is able to recognize when a small break has occurred.
7.
HPI Throttling For small LOCA's, the HPI system is needed to provide makeup to the RCS and must remain operable unless specific criteria are satisfied.
The basis for these criteria are described below.
For certain small breaks, system depressurization will result in LPI actuation. Since the LPI is designed to provide injection at a greater capacity than the HPI, termination of the HPI is allowed.
However, this action should only be taken if the flow rate through each line is at least 1000 gpm and the situation has been stable for 20 minutes.
The 20-minute time delay is included to ensure that the system will not repressurize and result in a loss of the LPI fluid.
In the event of a core flooding line break, the LPI fluid entering the broken core flooding line will not reach the vessel. Thus, in order to ensure that fluid is continually being injected to the RV for all breaks, the LPI must be providing fluid through both lines. The 1000 gpm is equivalent to the flow from 105414h
13-two HPI pumps and ensures that upon termination of the HPI pumps, adequate flow is being delivered to the RV.
Throttling or termination of the HPI flow is also allowed if all the following criteria are met:
0 A.
Hot and cold leg temperatures are at least 50 F below the saturation temperatures for the existing RCS pressure.
B.
Hot leg temperatures are no more than 50 F hotter than the secondary side saturation temperature (This ensures that heat is being removed via the SG.)
C.
The action is necessary to prevent the indicated pressurizer level from going off-scale high.
Under thase conditions, the prima y system is solid.
Continued HPI flow at full capacity may result in a solid pressurizer and would result in a lifting of the PORV and/or the pressurizer code safety valves.
This may in turn lead to a LOCA. Thus, HPI flow should be throttled to maintain a stable inventory in the RCS.
However, if the 50 F subcooling cannot be maintained, the HPI shall be innediately reactivated.
HPI flows should also be throttled to prevent violation of the nil ductility temperature (NDT) for the reactor vessel.
This concern can only arise if the fluid temperature within the reactor vessel is at least 50 F subcooled. A curve of th allowable downcomer temperature for a given RCS pressure is provided within the operating guidelines. The downcomer temperature is determined by one of two methods:
1.
If one or more RC pumps are operative, the cold leg RTD reading will be essentially the same as the reactor vessel downcomer temperature.
2.
Without the RC pumps operating, the cold leg RTD's may not provide 105414f temperature readings indicative of the actual RV downcomer temperature, as a stagnant pool of water may exist at these locations.
The incore thermocouples will provide the best indicator of the downcomer temperature and should be; utilized if no RC pumps are available.
In order to account for heat added to the fluid from the core,150 F must be subtracted from the incore thermocouple readings to reflect the downcomer temperature.
This method will result in temperatures which will be 1mver than the expected downcomer temperature.
Thus, use of this methodology assumes that NDT will not be a problem.
O O
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V.
Guidelines for Non-LOCA Events Because of the broad spectrum of system conditions covered by the small break guidelines, the operator actions and precautions identified to bring the plant to a long term cooling mode apply, in general, to any abnormal event which results in a decrease in RCS pressure. The small break guidelines will thus be utilized to update the emergency procedures for non-LOCA events; at a minimum, the following pertinent sections of the small break guidelines will be incorporated:
1.
RC Pump Trip Criteria and SG Level Control actions to promote natural circulation.
2.
RC pump Restart Criteria 3.
HPI Control Criteria 4.
The need to monitor system subcooling limits.
The items will be supplemented by the additional instructions / precautions to the effect that:
- 1. For non-LOCA events, a restart of the RC pumps (1 per loop) and termination
.of SG fill is prudent to minimize system overcooling due to addition of cold AFW to the OTSG's.
Note: The establishment of a subcooled condition (>50F) is a clean indication that a non-LOCA event or a LOCA for which a RCP trip is not required is not in progress.
- 2. HPI should be throttled, when 50F subcooling is established, to avoid a pressurizer overfill.
- 3. During severe overcooling events, sufficient HPI water may be added, prior to achieving a subcooled condition (> 50F) and a pressurizer level (on-scale), such that the system may evolve to water solid state when the RC temperature recovers to a hot shutdown condition (s 530F).
1054156
- m.
4 Operator action to control primary temperature (via secondary steam pressure-control using the turbine bypass valves and/or atmospheric dumps) may be required to maintain pressurizer level on scale.
NOTE: The Operating Guidelines For Small Breaks have been modified to include Item 3 above.
With operator training in the post-LOCA recovery methods in con" junction with modification of existing emergency procedures based on the small break guidelines, plant recovery and control can be achieved for any abnormal event for which an RCP trip is required.
e e
e
REACTOR COOLANT PUMP TRIP ON HICH PRESSURE INJECTION SIGNAL la response to NRC Bulletins79-05C and 79-06C, dated July 26, 1979, Florida Power Corporation wishes to respond to Short-term Action Item 1 A on page 2 of 3.
This item requires that "upon reactor trip a9d initiation of RPI caused by low reactor coolant system pressure, immediately trip all operating reactor coolant pumps (RCP's)".
Florida Power Corporation proposes that a high pressure injection signal (HP1) be provided to immediately trip all operating RCP's.
The proposed design is outlined as follows:
1.
A new Clark type PM relay and a solid state switch, Hamlin type P/N7, in parallel with existing relays 63Yl/RCl and 63Y2/RC1.
These relays and solid state switches are the same types that are presently used in the Engineered Safety Features Actuation System.
2.
The new components described above would be installed in both HPI ES Actuation "A" and "B" and operate by becoming de-energized when the 1500 psi bistable becomes de-energized below 1500 psi and is not bypassed.
3.
One normally closed contact from each of the relays in the HPI ES Actuation "A" and "B" would be seriesed and placed in the trip circuits of each of the RCP's.
The purpose of using contacts in series is to permit testing of a channel without tripping the RCP's.
4.
In addition, two (2) alarms would be added to the main control board to alert the operator that (1) an HPI signal has tripped the RCP's or (2) an alarm would alert the operator that one of the ES Actuation "A" or B" signals has been actuated.
5.
The schedule for implementation of this trip function is September 30, 1979.
EMGemhFa D6 105415('