ML19309C284

Forwards Revision 2 to 791130 Response to NRC 791025 Request on B&W Sys Sensitivity.Incorporates Addl Overcooling Sensitivity Analyses Into Apps a & B

ML19309C284
Person / Time
Site:	Midland
Issue date:	04/01/1980
From:	Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Harold Denton Office of Nuclear Reactor Regulation
References
HOWE-70-80, NUDOCS 8004080404
Download: ML19309C284 (250)

Text

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l G eneral Of fices - 1945 West Parnell Road, Jackson, Michigan 49201 * (517) 788 0453 April 1, 190:

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Enclosed erc tc n (10) copiec of Re;isior. 2 to C:nsumers Pcycr Cc par.y's rerpens e cf I;ovetter 32, 1979 to sour 10 CF" Sc.5h(f) recuest on 39 System Sensitivity drited OctcLc. 25, 1979 Eevision 2 incorporates additional overcooling scaritivity er.alyscs into Appendix A ar.d B. The chanted peces bear the r.otation "Ecvision 2 !./9C" and are narked in tne r.a:cin to indicat- where changes have been cede. Appendices C, D, E and F recain unchanced.

Consumers Power Ceepany Dited: April 1, 199? By %b ( b C7 NM $

StephenV) Howell, Senior Vice President l

Cworn and subscribcd to before e or tais let day of April 1930.

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RESPONSE TO 10 CFR 50.54(f)

APPENDIXES A AND B n

Ouestions

a. Identify the most severe overcooling events (considering both anticipated transients and accidents) which could occur at your facility. These should be the events which cause the greatest inventory shrinkage. Under the guidelines that no operator action occurs before 10 minutes and only safety systems can be used to mitigate the event, each licensee should show that the core remains adequately cooled.

b. 7:lentify whether action of the ECCS or RPS (or operator -

action) is necessary to protect the core following the most severe evercooling transient identified. If these systems are required, you should show that its decign criterion for the number of actuation cycles is adequate, considering arrival rates for excessive cooling transients.

Response

I. INTRODUCTION AND CONCLUSIONS A. Background O)

(- ' On October 25, 1979, the NRC issued a letter to utilities holding construction permits for B&W NSSSs. The utilities were requested to assess overcooling events on their plants, accounting for balance-of-plant (BOP) features.

B. Scope This report responds to the specific NRC requeste identified above. More than one transient type is analyzed to address the different frequency of occurrence classifications and to ensure that the most severe cases are indeed included in the evaluation. A quali-tative assessment of possible nonmitigative operator actions in the o to 10-minute time frame is also provided. This assessment provides indications of what operator action is anticipated during the initial phases of an overcooling transient.

The analyses identify the frequency of the reactor protection system (RPS), engineered safety features actuation system (ESPAS), and operator action for mitigation of the transient.

A summary of the results is given in Section II.

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Section III provides the details of the initial conditions, computer codes, and basic assumptions used in the analysis. The transient response data are given in Section IV.Section V demonstrates the' adequacy of the design criteria for each system.

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RESPONSE TO 10 CFR 50.54(f)

,-- APPENDIXES A AND B C. Conclusions Based on the analyses performed in this report, the following conclusions can be drawn.

1. The overcooling accidents [ main steam line breaks (MSLBs) and small steam line breaks (SSLBs)] and 2 the overcooling transients [ pressure regulator malfunction and main feedwater (MFW) overfill]

analyzed herein retain adequate core cooling even when analyzed with no operator action before 10 minutes and with only safety systems used to mitigate the event.

2. RPS and emergency core cooling system (ECCS) actuation are required to mitigate the most severe overcooling transients. However, operating data imply that the arrival rate of transients requiring RPS or ECCS actuation is within the design basis. 2 D. Applicability o'f Results The results presented in this report are applicable O specifically to this NSS with the parameters tabulated in Section III. Specific attention has been paid to the BOP equipment in the mitigative functions performed.

II.

SUMMARY

This section provides a detailed summary including identification of the safety concern and basis for selection of the transients to resolve the concern and principal results of the analysis. By reviewing this section, which is supported by the details given in Sections III, IV, and V, a concise overview can be obtained of the completed resolution of this safety l2 concern.

Section II.A addresses the selection of anticipated transient and accident conditions causing greatest l core shrinkage, and Section II.B discusses the phenomenon of void formation under inventory shrinkage conditions.

Section II.C summarizes the analyses. Section II.D summarizesSection V, demonstrating that the design criteria for the number of actuation cycles of the RPS and ESPAS are adequate.

A '

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RESPONSE TO 10 CFR 50.54(f)

APPEND 1XES A AND B p

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\ A. Limiting Overcooling Event Confirmation Maximum RCS coolant inventory shrinkage results from a decrease in the pressure and temperature of the coolant at a maximum rate, without a compensating coolant makeup addition. The double-ended steam line break (SLB) provides maximum cooldown rates and is analyzed in Section IV.B as the limiting accident.

Several sensitivities and differing conditions were analyzed to provide greater insight into the steam void formation and collapse which would occur and its subsequent effect on core cooling. These addi-tional studies were performed on the SLB because this accident was expected to result in RCS voiding.

In selecting the limiting anticipated transient, SAR and operating plant overcooling events were reviewed.

The most severe moderate-frequency event in the SAR is the steam pressure regulator malfunction. Additional analyses of this event are found in Section IV.A.2. 2 However, a review of plant transient data (see Section IV.A.1) has shown that overfeed by MFW af ter reactor trip has produced the most severe overcooling transients. Therefore, based on

((~N arrival rates for operating plants, the MFW overfeed

,) following a reactor trip / turbine trip is considered 2

the limiting anticipated transient and is analyzed in Section IV.B. l2 B. Shrinkage Effects Shrinkage of the reactor coolant system (RCS) coolant liquid volume occurs as temperature decreases during an overcooling event. The pressurizer volume of 1,500 cubic feet contains 800 cubic feet of mostly saturated water during normal operation. l2 Tb ;:3 liquid volume flows out of the pressurizer luco the system as the system inventory volume decreases. If the RCS coolant inventory volume decrease is greater than 800 cubic feet and continues to decrease, the pressurizer steam space can be transferred into the RCS. This type of steam voiding is limited by the inventory volume difference l between hot, full power, and the final pressure /

temperature achieved during the transient. Its effect is further mitigated by actuation of the ECCS.

l l The other mechanism which produces steam voids in l the RCS is flashing of RCS water. As the pressure

(~S rapidly decreases in the RCS, the liquid in the

() hotter portions of the system can become saturated A&B-3 Revision 2 4/80

RESPONSE TO 10 CPR 50.54(f)

APPENDIXES A AND D

, and steam can form. This effect can be aggravated by the hot metal in this area, flashing additional I water to steam. This process, in a non-loss-of-i coolant accident (non-LOCA) situation, is self-

' regulating. As the steam separates, or additional flashing occurs, the pressure decrease in the system lessens as the overcooling continues. The steam void formation is then reduced and the steam void will tend to collapse as a subcooled state is again established.

Examination of the SLB analysis indicates that a small amount of steam formation occurs in the upper hot leg region prior to the pressurizer emptying, occurring almost exclusively on the side with the af fected steam generator. If the affected i

steam generator is on the loop with the pressurizer, emptying the pressurizer contributes to the steam void formation. If the affected steam generator is on the opposite loop from the pressurizer, emptying the pressurizer has little effect on the steam voids on that side and they are quickly quenched. Therefore, the limiting accident, in terms of void volume formation, occurs for the SLB in the same loop that has the pressurizer.

C. Adequacy of Core Cooling

(

In this section, the results presented in Section IV are summarized and analyzed for determination of l

adequate core cooling.

The anticipated transient analyzed is the overfeed of the steam generators by MFW (Section IV.B). l2

! This overcooling transient, with no mitigative operator action for 10 minutes, resulted in the pressurizer emptying briefly. However, high-pressure injection (HPI) actuation is suf ficient to prevent any steam voiding in the RCS.

The design basis SLB accident (Section IV.C) l2 produces steam voiding in the upper hot leg regions of the RCS. Several sensitivity studies were performed to assess impact on steam void formation and subsequent core cooling flow. The sensitivity studies included the following:

1. Varying the time of loss-of-of fsite power (LOOP) from time of trip to time of ESPAS

2. With and without core decay heat A&B-4 Revision 2 4/80

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RESPONSE TO 10 CFR 50.54(f)

(T APPENDIXES A AND B U

3. Single-failure assumptions of stuck-open relief valve on unaffected steam generator or loss of one HPI pump

4. Moving the break from the steam generator with the pressurizer in its loop to the side without the pressurizer In all analyzed cases, core flow continued, or if interrupted, resumed immediately upon collapse of the void in the unaffected steam generator side of the RCS. The core region remained subcooled 2

throughout the transient for all cases analyzed.

SSLas, analyzed with similar conditions and single failures, show the same results. The pressure regulator malfunction (an SSLB with no single failure) shows no voiding for the worst conditions.

Mitigative operator action was not assumed in the analysis in the first 10 minutes. From a review of potential operator actions during this time, it is concluded that only two actions are of major y importance. Operator control of the steam generator ,

level would have reduced the extent of RCS inventory shrinkage for both MFW overfeed and SLB transients.

A nonmitigative operator action would result from the premature cutoff of the HPI flow. The indication to the operator during steam voiding situations, such as occurred during the SLB accident analyzed, 2

is to maintain HPI flow because pressurizer level and subcooled margin both indicate its necessity.

Adequate core cooling would necessitate that HPI makeup to the RCS be available at some point during the course of the overcooling transients.

D. Adequacy of Core Protective MeasuresSection V provides the details of the design basis for operating transient cycles. Operating plant data have shown the 40 cycles of actuation of HPI to be a sufficient design basis to cover automatic initiation arrival rates for this system. The analysis presented in Section IV confirms that the most severe overcooling events require ECCS actuation.

The operating plant data show that ESFAS automatic actuations occur less than once per year; therefore, 40 cycles per lifetime is an adequate design for

! transients not expected to occur greater than

! (; 40 times in the life of the plant.

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RESPONSE TO 10 CPR 50.54(f) r'N APPENDIXES A AND B III. ANALYTICAL TECHNIQUES A. Computer Codes The B&W-certified computer code TRAP 2 (Reference 1) has been used in the analyses presented in the following sections. This computer code is a nodal type, digital simulation (similar to CRAFT 2, 1 Reference 2), and is capable of handling rapid overcooling transients that may result in two-phase fluid conditions in the reactor coolant system.

The noding flowpath networks used in the TRAP 2 analysis of the plant are given in Piqures A&R-1 and A&B-2. A description of each node and the important flowpaths are given in Tables A&B-1 and l1 A&B-2. The more detailed noding shown in Figure A&B-1 (description in Table A&B-1) is referred to as maxi-TRAP. The less detailed model in Figure A&B-2 (description in Table A&B-2) is referred to as mini-TRAP. The more detailed maxi-TRAP model is used to compare with the mini-TRAP during the

[s) initial phase of the transient for model verification.

\d For the double ended SLB, the feedwater flow from 2 maxi-TRAP was used as input whereas the feedwater pump model shown in Figure A&B-2 was used in the other analyses presented in Section IV. The mini-TRAP model is used in the long-term solution where ,

system variables are more slowly varying and the additional noding is not required.

B. Transient Selection The types of overcooling events considered include those which constitute the initiating event, those which result from single failures following any initiating event, and those which are made more severe from single failures following the initiating overcooling event. The specific systems whose malfunction or failure are considered either as initiating events or single failures which enhance overcooling are:

1. Feedwater heater failure which causes a decrease in feedwater temperature

2. Feedwater flow control malfunction which causes an increase in feedwater flow A&B-6 Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

APPENDIXES A AND B

- 3. Steam pressure regulator malfunction which causes an increase in steam flow

4. Inadvertent opening or stuck-open steam relief  ;

valve which causes increased stean flow and/or depressurization of a steam generator

5. Steam system piping failure which causes excessive steam flow and depressurization of a steam generator The SAR analyses are referred to in order to narrow the most severe type of overcooling events for consideration. Specifically, these include:

1. Events which constitute an initiating event -

Items 1 through 4 above are moderate frequency, i i of which steam regulator malfunction is the l2 most limiting according to the SAR analyses.

Item 5 is a design basis event for which the double-ended rupture (DER) MSLB is limiting.

2. Events which result from single failure following any initiating event - This infrequent occurrence

/~~. \ is a combination of a moderate frequency event k-s plus one of Items 1 through 4 occurring as a single failure. The events chosen to be analyzed in this category are: 1) an immediate reactor trip on turbine trip signal (decrease the heat source) combined with a feedwater 2 flow control malfunction that allows continued MFW flow (increase the heat sink); and 2) an '

SSLB equivalent in break area to the steam regulator malfunction in Case 1, combined with various single failures as shown in Table A&B-3.

3. Events which are more severe from single _. failures following the initiating overcooling event - The limiting design basis overcooling transient is a double-ended SLB. .The single failure chosen to f maximize continued long-term cooling is a stuck-open relief valve on the unaffected steam generator.

The limiting or potentially limiting overcooling cases to be analyzed as discussed above are summarized in Table A&B-3.

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APPENDIXES A AND B h"' C. Basic Assumptions Key input parameters used in the plant analysis are given in Table A&B-4. These values represent as-built information, realistic setpoints, actuation times, flowrates, and valve closures. Other system parameters not listed are those applicable to the plant design - i described in the SAR. The assumption of a stuck rod l2 was removed from the shutdown rod worth, resulting in a more realistic, conservative direction for the over-cooling type events concerned with maximum RCS coolant shrinkage.

Single failures of active components assumed in the analysis are given in Table A&B-3. Some parameteri-zation of the single-failure assumption is done for the limiting overcooling case. Because only safety grade equipment is assumed to function, the single failures of mitigative equipment are limited to steam 2 relief valve, steac isolation valve, and HPI.

Table A&B-5 lists the equipment assumed to function-for each transient analyzec'.

No mitigative operator action is assuned for 10 minutes  !

(V ) in the analysis.

IV. RESULTS OF CORE COOLING STUDIES A. Anticipated Transients

1. Scope of Evaluation The anticipated transients analyzed in the SARs were reviewed for cooldown rates and consequences in order to select the most limiting case for shrinkage. Operating plant data was also reviewed. From this review, the transient with the highest frequency of occurrence and the potential for greatest overcooling was due to malfunctions resulting in overfeed of the steam generators by MFW.

Operating plant data shows that overcooling of the RCS has occurred from primarily two types of events: failure of a relief valve to reseat at the proper pressure, which limits the overcooling to the saturation temperature of the pressure at which the valve does reseat; and overfeed of the steam generators following a reactor trip, which has caused the greatest primary-cooldown observed.

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RESPONSE TO 10 CPR 50.54(f)

APPENDIXES A AND B (s-L.)\

Steam pressure regulator malfunctions that allow increased steam flow would represent overcooling by depressurizing the secondary system. Its effect is very similar to an SSLB analysis. The arrival rate for this transient has been zero at operating B&W plants. Therefore, the frequency classification for this transient is less than the MFW overfeed transient as presented in Section IV.B. The MFW overfeed represents the maximum cooling that can be achieved by feeding the once-through steam generators (OTSGs) with uncontrolled MFW.

2. Pressure Regulator Malfunction Analysis The pressure regulator malfunction transient was analyzed by assuming symmetric SSLBs on both steam generators downstream of the main steam isolation valves (MSIVs). The break size was chosen to yield approximately 28%

excess flow, corresponding to the excess steam flow which would occur if the turbine bypass system were opened and the governor valves were assumed to open to full open. This break

(N g ,) size is equivalent to 0.25 square feet per generator.

2 The sensitivity studies performed for the small SLB in Section IV.B indicate that the LOOP at ESPAS, with no decay heat, is the most severe voiding case. Therefore, these conditions were assumed for the pressure regulator malfunction transient. The sequence of events is given in Table A&B-6. The analysis was performed using the models and assumptions given in Section III.

The mini-TRAP model was used (Sections IV.B and IV.C contain mini- and maxi-TRAP comparisons).

Neither credit for integrated control system (ICS) nor operator action was assumed. No single failures, other than the initiating event, were assumed.

Figures A&B-3 through A&B-9 present the system parameters. The pressurizer empties in 24 seconds and starts to refill at 50 seconds. No voiding was calculated during this transient.

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RESPONSE TO 10 CFR 50.54(f)

APPENDIXES A AND B From the system responses shown, the only probable operator action would be to regulate HPI flow to maintain pressurizer inventory once level was reestablished. However, this particular action is not required for the first 10 minutes of the transient.

3. Conclusions The pressurizer empties and no steam void formation occurs in the RCS. Both loops remain subcooled. Loop flow for both loops is maintained throughout the transient, assuring adequate core cooling.

B. Anticipated Transient with Single Failure

1. Scope of Evaluation The two transients analyzed in this section 2 are SSLB and MFW overfeed transients. The SSLB uses the same break area as the pressure regulator malfunction presented in Section IV.A,

~N but assumes an MSIV failure to allow complete blowdown of one steam generator. The initiating event for the MFW overfeed transient is a turbine trip with simultaneous reactor trip and a control failure such that MFW continues to feed both steam generators at full capacity.

The worst coolant inventory event has been studied by analyzing a spectrum of different conditions. Table A&B-7 shows the various conditions for both SSLB and MFW overfeed and identifies these different analyses by case number for further reference in the discussion of results provided in the following sections.

2. MFW Overfeed Analysis The initiating event is a turbine trip with simultaneous reactor trip and a control failure such that main feedwater continues to feed both steam generators at full capacity. 2 The analysis was performed using the models and assumptions given in Section III. A comparison of the maxi-TRAP and mini-TRAP i 7-~g system parameters is shown in Figures A&B-10 2 I

) to A&B-12 for the first 100 seconds of the

\~/ transient. Based on the relatively good agreement of these results, the mini-TRAP 2 model was used for subsequent overfeeding l analyses. No credit for ICS nor operator A&B-10 Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

APPENDIXES A AND B actions was assumed. The sequence of events is given in Tables A&B-8 through.A&B-ll for each case analyzed. The figure numbers for each case are listed on the table corresponding to that case. The various conditions analyzed were: 1) RCS pumps running, 2) RCS pumps tripped at time of reactor trip, 3) RCS pumps tripped at time of ESPAS on low RCS pressure, and 4) the same as case 3 but assuming no decay heat is present.

For each case analyzed, the pressurizer empties briefly at about 3 minutes into the transient.

However, no steam void formation occurred in the RCS. The cooldown rate, i.e., RCS coolant inventory shrinkage, was not large enough to overcome the subcooled state of the RCS coolant inventory, nor the HPI rate. Without additional means to control OTSG levels, the steam generators will overfill with MFW. Subsequently, the steam lines will become filled with water.

This was found to occur prior to ESFAS isolation of the MFW. It was assumed for the analysis 2 s that no immediate operator action was taken

) and water relief out of the safety valves was permitted. After ESPAS isolated MFW, AFW flow was continued to the steam generators to maximize the cooling rate. No credit is taken for the safety-grade AFW level control and overfill protection systems incorporated in the Midland design which would prevent this occurrence. Thus, the analysis of this event is extremely conservative.

From the system response observed, two probable operator actions during the course of the transient are suggested. First, operator action would be needed to terminate the OTSG overfill by MFW early in the transient, which would stop the overcooling of the RCS. Second, because sufficient subcooled margin exists throughout most of the transient, the operator would regulate HPI flow to maintain pressurizer inventory. However, this particular action is not required for the first 10 minutes of the transient.

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RESPONSE TO 10 CFR 50.54(f) s APPENDIXES A AND B

3. Small SLB Analysis The SSLB was analyzed with a break area equivalent to that used in the pressure regulator malfunction (0.5 square feet). The analysis can be considered as a small break upstream of the MSIVs with no '

credit for building ESFAS signal or as a small break downstream of the MSIVs with an MSIV failure. The analysis was performed using the models and assumptions given in Section III.

No credit for ICS nor operator actions was assumed. The sequence of events is given in Tables A&B-12 through A&B-17 for each case analyzed. The figure numbers for each case are listed on the table en* responding to that case. The various conditions analyzed for the SSLB were: 1) RCS pumps running, 2) RCS pumps tripped at time of reactor trip, 3) RCS pumps ,

tripped at time of low RCS pressure ESFAS, and

4) the same as case 3 assuming no decay heat  :

is present. Case 4 showed the worst results and was selected for analysis with single failures. Two single failures were considered: 2

1) failure of one HPI pump and 2) stuck-open

(N

( ,) relief valve on the unaffected steam generator.

The results are very similar to the DER described in the following section. Significant steam voiding occurs in the loop with the affected steam generator. The steam void volume is 500 to 600 cubic feet, and occurs for the condition of RCS pump trip at time of low RCS pressure ESPAS with no decay heat. The single failure of a stuck-open relief valve on the unaffected steam generator does not aggravate the result significantly. The failure of one HPI pump causes the steam voiding to persist longer.

The analysis shows it still exists at the end of the analysis (10 minutes) but is decreasing as one HPI flow compensates for the shrinkage.

In all cases analyzed, core flow is maintained and, except for the case noted above, in 10 minutes a solid primary system is reestablished.

4. Conclusions r

The RCS coolant inventory remains subcooled i throughout the MFW overfeed transient, thus ,

l

-~ assuring core cooling. The pressurizer emptying

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1 RESPONSE To 10 CFR 50.54(f)

APPENDIXES A AND B

( was brief (less than 50 seconds) in duration before the HPI actuation started refilling the system. Only additional failures, such as bypass or relief valves stuck open, could increase the cooldown rate experienced during the transient. ESFAS terminates the excessive feedwater flow. With the fill rates of MFW assumed, the steam generators will overfill in 2 about 90 seconds.

The SSLB with the failures assumed produces steam voiding similar to the DER. The phenomenon is more completely described in Section IV.C.

In all cases analyzed, core cooling remains adequate, with subcooled conditions prevail-ing in the core region.

C. Accidents

1. Scope of Evaluation Maximum overcooling of the RCS results from an uncontrolled blowdown of the secondary plant (i.e., SLB accident). The DER from full power has been demonstrated in the SAR to result in maximum overcooling. Selection of the worst p)

( coolant inventory shrinkage case for this event has been studied by analyzing a spectrum of different conditions. Table A&B-7 shows the various conditions and identifies these different analyses by case number for further ,

reference in the discussion of results provided in the following sections.

2. SLB Analysis A double-ended guillotine break is assumed to occur in the 33.5-inch inside diameter steam line. The location of the break is outside of the reactor building. This analysis assumes the auxiliary feedwater (AFW) level control failure for all cases, except Case 9. Because 2 the system is safety grade on Midland, this assumption is extremely conservative. Other system parameters, models, and assumptions are as presented in Section III.

The sequence of events is given in Tables A&B-18 2 through A&B-26 for each case analyzed. The figures for each case are listed on the table corresponding to that case. The figures for [2

[~s \ Case 1, reactor coolant pumps running, also

\s l include a comparison of maxi-TRAP and mini-TRAP results, showing reasonably good agreement between the two models. Subsequent SLB analyses were performed using the mini-TRAP model.

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APPENDIXES A AND B The SLB accident was analyzed for 10 minutes g 'N assuming no mitigative operator action and 4

( ,) only safety-grade equipment for transient mitigation. The overcooling rate, as anticipated, is much higher for the SLR cases than that obtained for the MFW overfeed case presented in the previous section.

The case resulting in the most severe consequences of RCS shrinkage occurs with LOOP at the time of ESPAS actuation. The assumption of no decay heat aggravates this shrinkage effect.

A bubble rise velocity of 5 feet per second was used in the hot leg piping nodes. It is important to note that the void formation data presented include entrained, as well as separated, bubble mass. Therefore, with reactor coolant pumps running and during the start of flow coastdown, the bubble mass will be almost totally entrained. Comparing the single-failure assumption of a stuck-open relief valve on the unaffected steam generator versus failure of one HPI pump, the stuck-open relief valve (Case 4) results in the maximum steady steam void formation. However, for the one HPI failure (Case 7), the steam void remains in the RCS longer. The maximum steam void

[h

\m l occurs in the hot leg attached to the pressur-izer and is about 320 cubic feet for the cases analyzed.

The first steam void formation that appears during the SLB accident is due to flashing (i.e., reaching saturation) in both hot legs.

This occurs prior to the pressurizer emptying.

On the loop side opposite the pressurizer and analyzed with the unaffected steam generator, this effect is small and returns to a solid, subcooled state about the time the pressurizer empties. On the loop with the pressurizer and the affected steam generator, this steam void continues to increase as the pressurizer empties. ESFAS initiation also occurs at approximately this time and HPI injection, as well as isolation of the affected steam generator main steam and feedwater, tend to limit the size of the steam void formed. HPI flow is sufficient to overcome the shrinkage that is still occurring from the heat removal through auxiliary feedwater to the unaffected steam generator. As refill and repressurization of the RCS continue by the HPI, the steam void is quenched and

' ('"T collapsed. Core flow is maintained throughout l ( ,/ the transient. During LOOP cases, natural circulation is maintained by the cooling from l

! the unaffected steam generator side of the

! RCS.

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APPENDIXES A AND B If no credit is taken for the safety-grade AFW 1evel control, the AFW fills the unaffected l2 steam generator in 6 to 7 minutes. The pressurizer is filling, but has not completely filled in  :

the first 10 minutes of the accident. Thus, adequate time is available for operator action to prevent pressurizer overfill. The level control system on the unaffected steam generator (a fully safety-grade system at Midland) would 2 allow earlier repressurization of the RCS, thereby leading to earlier collapse of the void (refer to Case 9). .

3. Conclusions Steam void formation in the upper hot leg regions was found to occur during the steam line break accident. The magnitude and duration of the steam void formation varied with the conditions under which the analysis was performed. In all cases, core flow was maintained or, if interrupted, resumed upon collapse of the void in the unaffected 2 steam generator side of the RCS. In all cases, the core remained subcooled. Some of the specific phenomena noted for the various cases analyzed are

() as follows:

a. The LOOP assumption at ESFAS produces slightly worse consequences than at an earlier time.

This is because the pumps running maximize the overcooling such that the later the LOOP (up to ESPAS), the more shrinkage that has occurred.

LOOP after ESPAS should not continue to increase the severity, because isolation of the affected  ;

steam generator main feedwater supply occurs i at ESFAS and greatly reduces the overcooling rate.  ;

i

b. The assumption of no decay heat aggravates the steam voiding situation. However, as decay '

heat level decreases, the need for additional core flow decreases. In the extreme, no decay heat implies no core cooling is necessary.

c. Single failure of a relief valve on the unaffected steam generator to maximize cooling rate and a single failure of one HPI pump to maximize the refill repressurization effects were examined.

The larger magnitude of steam void occurred

' for the stuck-open relief valve case, whereas the steam void formation was of longer duration l

s for the HPI failure case.

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APPENDIXES A AND B r

O

%- d. The void formation in a given loop was large enough to create temporary flow blockage in that loop. However, the net core flow remains positive through most of the analyses, and is never 2 interrupted to the point that saturation occurs in the core region.

No mitigative operator action was assumed for 10 minutes in the analysis. With the fill rates of auxiliary feedwater assumed, and with l2 no credit taken for the safety-grade AFW level control system, the unaffected steam generator will overfill in 6 to 7 minutes. Core cooling appears adequate for all cases analyzed because subcooled conditions are maintained in the core region.

V. DESIGN BASIS FOR CORE PROTECTION Required ECCS and RPS actions necessary to protect the core have been summarized in Table A&B-5 and discussed No in more detail for each transient in Section IV.

operator action has been assumed within 10 minutes for mitigation in the analysis. This section demonstrates that the design criteria for the number of actuation

()N,

(, cycles are adequate.

Twenty-five different types of transient cycles (several l2 are SAR analyses) are used in evaluating the acceptable number of design cycles. These operating transients are listed in Table A&B-27, along with the number of l2 design cycles for each transient type. These data are the basis on which the stress evaluation is performed for the plant and will be contained in the technical specifications for the plant. The number of cycles for transient types listed in Table A&B-27 is not l2 meant to be an absolute limit, but was chosen on the basis of expected frequency (plus margin) and is shown to be acceptable in the stress evaluation.

Special transient analyses can be performed based on any actual transient data, thereby allowing categorization of the special case into one of the allowable transient design cycles.

The adequacy of the number of design cycles can be inferred from operating plant data. Table A&B-28 l2 compares the actual arrival rate for RPS and ESFAS actuation to date on plants of B&W design to the rates allowed by the design basis (Table A&B-27). The l2 operating data are less than the allowable actuation g-]w

(

rate for both systems, thereby supporting the adequacy i

\ of design.

A&B-16 Revision 2 4/80 l

l

RESPONSE TO 10 CFR 50.54(f)

APPENDIXES A AND B .

VI. REFERENCES i

1. J.J. Cudlin, P.W. Dagett, TRAP 2-FORTRAN Program for  !

Digital Simulation of the Transient Behavior of the Once-Through Steam Generator and Associated C861 ant [

System, BAW-10128 (August 1976), Babcock & Wilcox, Lynchburg, Virginia

2. R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, l CRAFT 2-PORTRAN Program for Digital Simulation of a s Multinode Reactor Plant During Loss of Coolant 7~ ~ ,

BAW-10092, Revision 2 (April 1975), Babcock & Wilcox, l Lynchburg, Virginia 4

j 1

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h

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O A&B-17 Revision 2 ,

4/80 i

l j

RESPONSE TO 10 CFR 50.54(f) .

TABLE A&B-1 f l

MAXI-TRAP fiODE AfiD PATl! DESCRIPTIO?l Description Node Number t l

i 1 Reactor Vessel Lower Plenum l 2 Core, Upper Plenum and Outlet Nozzles I 3, 16 Hot Leg Piping 4-13, 17-26 Primary, Steam Generatcr j

14, 27 Cold Leg Piping 15 Reactor Vessel Downccmcr 1 j 28, 55, 56 Pressurizer r 29 Containment i

30-39, 40-49 Secondary, Steam Generator  !

50, 51 Steam Risers l,. I 53, 54, 68, 69 Steam Generator Downcomer

,! F 64, 66 Feedwater Piping l

t 63 Turbine and Process Steam Plant 65, 67 Feedwater Piping and Feedwater Heater r

52, 57-62 Steam Piping  ;

f c

t l

2 ,

Revision 2 4/80

,-- -ww, .- -- - - - - - - ,,-, err w- - - - - , - --

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-1 (Cont'd)

Description Flow Path flumber ,

1 Core Core Bypass 2  ?

3, 4, 17, 18 Hot Leg Piping 5-13, 19-27 Primary, Steam Generator 14, 28 RC Pumps 15, 29 Cold Leg Piping 16 Reactor Vessel Downcomer 30 Pressurizer Surge Line 31-39, 41-49 Secondary, Steam Generator 40, 50 Steam Riser 7 51, 52, 56, 69, 70, 71 26 Inch Steam Piping 53, 58 Aspirator 54, 55, 85, 86 Steam Generator Downcomer i

59-62 Pressurizer 63, 66 feedwater Pumps 64, 65, 67, 68 Feedwater Piping 1 36 Inch Steam Piping 72 73, 76 MSIV i 74, 75, 77, 78 Process Steam and Turbine Piping Feedwater Piping Crossover 79 80 Steam Piping Crossover 57, 84 Break 83 - Aux. Feedwater 81 HPI 82 LPI (Not Used)

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) 2 Table A&B-2 MINI-TRAP NODE AND PATH DESCRIPTION Description Node Number

-1 Reactor Vessel, Lower Plenum 2

Reactor Vessel, Core 3

Reactor Vessel, Upper Plenum 4, 10 Hot Leg Piping (including " Candy Cane")

32, 33 " Candy Cane" and Upper S.G. Shroud 5-7, 11-13 Primary, Steam Generator Tube Region 8, 14 Cold Leg Piping 9 Reactor Vessel Downcomer 15 Pressurizer 16, 24 Steam Generator Downcomer 17, 25 Steam Generator Lower Plinum 18-20, 26-28 Secondary, Steam Generator Tube Region 21, 29 Steam Risers 22, 30 Main Steam Piping 23 Turbine 31 Containment Feedwater Heater 34 35, 36 Feedwater Piping l2 Description Path Number O- 1 2

Core Core Bypass 3 Upper Plenum, Reactor Vessel 4, 11 Hot Leg Piping 5, 12 Upper Steam Generator Shroud 45, 46, 47, 48 Top of Hot Leg " Candy Cane" 6, 7, 13, 14 Primary Heat Transfer Region, S.G. '

8, 15 RC Pumps 9, 16 Cold Leg Piping 10 Downconer, Pasctor Vessel 17 Pressurizer burge Line 18, 19, 26, 27 Steam Generator Dcwncomer and Plenum 20, 21, 28, 29 Secondary Heat Transfer Region, S.G.

22, 30 Aspirator 23, 31 Steam Riser, Steam Generator 24, 32 Main Steam Piping 25, 33 Turbine Piping i

34 Steam Crossover 36, 37 HP1 38*, 39, 43, 44 AFW 40, 41 Feedwater Piping 42 LP1 l

49 Stuck open Relief Yavle 2 35, 50 Leak Paths 51, 52 Feedwater Pump Paths 53, 54 Mrwiv g

Revision 2 4/80

O O RESPONSE TO 10 CFR 50.54(f)

• TABLE A&B-3 .

2

SUMMARY

OF EVENTS ANALY2ED Single Failure Sensitivity Studies Initiating Event A. Anticipated Event Steam Regulator Malfunction B. Anticipated Event Made More 2

Severe By Single Failure o LOOP

• at Reactor Trip

1. Reactor Trip / Turbine Trip Main Feedwater Overfeed o LOOP at Low RC* Pressure d5FASA Trip o Decay "est

),

o LOOP at Reactor Trip 2.- Small Steam Line Break Main Steam Isolation Valve o LOOP at Low RC Pressure ESFAS Trip o Decay fleat o HPI Single Failure C. Desian Basis Overcooling -

o LOOP at Reactor Trip Double Ended Steam Line Break Main Steam Relief Valve Stuck o LOOP at Low RC Pressure ESFAS Trip Open o Decay Heat o HPI Single Failure o Steam Generator Level Control o Break on Dif ferent OTSGs

a. W Ss D OD 4 U?

g-

• Definitions: LOOP - Loss of Offsite Power a

RC - Reactor Coolant ESFAS - Emergency Safeguards Features Actuation System y

RESPONSE TO 10 CFR 50.54(f)

OVERC00LitlG AflALYSIS IllPUT ASSUMPTIO.15 TABLE A&B-4 Parameter 177 FA 102%

Power Level T se, F 579 ,

RCS Operating Pressure (i.. Fressurizer tap),

psig 2155 PressurizerLevel(indicated),in. 180 l RPS Trip S.ignals High Flux, % FP 105.5 Low Pressure (core outlet), psig 1855 ESTAS Trip Setpoints Low RC Press., psig 1500 Low SG Press., psig 585 ESFAS Trip Delay, sec. 2.5 ,

MSIV Closure Time, sec. 5 MFWIV Closure Tire (linear ramped area), sec. 15 l1 Auxiliary Feedwater 1 Design Capacity Turbine, gpm 885 Motor, spm 885 Temperature, OF 40 Initiation Time After ESFAS, sec.

15

. With Offsite Power With Loss of Offsite Power 40 l

l Main Feedwater Tempera'ture, OF 430 HPI System Design Capacity per Pump, gpm' 2 pumps 0 500 each Temperature, UF 40 ,

Boron Concentration, ppm 2270 Initiation Time Af ter ESFAS, sec.

p I V With Offsite Power 25 With Loss of Offsite Power 30 I

OTSG Outlet Pressure, psig 910 l Revision 2 4/80 ,

f i

RESPONSE TO 10 CFR 50.54(f)

Table A&B-5 EQUIPMENT AND RELATED SYSTEMS l2 I

. ASSUMED TO FUNCTI'NO

.ESFAS MSLIS MSIV RC TURBINE FOGG AFW HPI LPI CFT FWIV PUNPS BYPASS TURRINE TRIP EVENT RPS/CRDCS Pressure Regulator Malfunction X X X X - -

X (a) - -

heactor Trip / Turbine Trip X -

X (a) X X with MFW Overfeed X X X -

2 6 X X X X -

x X (a) - -

Small Steam Line Break Steam Line Break (Double-Ended Rupture) X X X X -

X X (a) - -

1 Denotes systen used when needed in the analysis

- Denotes systen not used in the analysis (a) Loss of offsite power cases assume 4 pump coastdown Revision 2 4/80

i RESPONSE TO 10 CFR 50.54 (f)  !

TABLE A&B-6 l2 STEAM REGULATOR MALFUNCTION ,

SEQUENCE OF EVENTS I

Event Time (sec) 4 0.0 Rupture 4.8  :

Iligh flux trip setpoint Rod movement starts at 5.2 4

Low RC pressure ESFAS +

LOOP event initiation 16.53 Pressure emplies at 24.0 j MSIVs close 24.03 MFWIVs close 31.53

!!PI starts to flow at 46.53 Pressurized starts to refill 50.0 AFW initiation to both generators 56.53 (Refer to Figures A&B-3 through 9) l2 ,

i 1

i O Revision 2 4/80

c n- U

) RESPONSES TO 10 CFR 50.54(f)

' TABLE A&B-7 OVERCOOLING SENSITIVITY STUDIES l2 LOOP at LOOP at LOOP at ESFAS, RC Pumps reactor trip ESFAS with no decay heat MFW Overfeed running

. I.

Case 2 Case 3 Case 4 Reactor Trip / Turbine Trip Case 1 with MW Overfeed II. San 11 SLB Case 1 Case 2 Case 3 Case 4 SSLB with MSIV or RB ESFAS Failure Case 5 (b) 2 Case 6 (a)

III. Steam Line Break Case 1(c) Case 2 Case 3 Case 4 With stuck open relief valve on unaffected generator, 2 HPI pumps ,

available With failure of one HPI pump, no - Case 5 Case 6 Case 7(d) stuck open relief valve Case 8 (e)

. Case 9(f) a) With failure of one HPI' pump, no stuck open relief ,

valve.

b) With stuck open relief valve, 2 HPI pumps.

c) Maxi / Mini-TRAP comparison presented for this case.

d) The SLB occurs in the LOOP with the pressurizer.

e) The SLB occurs in the opposite LOOP from the pressurizer.

f) The Case 7 was re-analyzed without f ailure of the steam generator level control. Revision 2 4/80

l .

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RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-S l2 CASE 1  ;

i MAIN FEEDWATER OVERFEED SEQUENCE OF EVENTS l t

4  !

Event Time (sec)  !

i t 0.0 Turbine trip Turbine stop valves close 0.0 Reactor trip 0.4 [

l >

Turbine bypass valves open 3.0 -

Atmospheric dump valves open 4.0 i

170.0 }

Pressutizer empty Low RC pressure ESFAS 188.4 l

! l MSIVs close 200.9 {

203.4  ;

l MFWIVs close  ;

I HPI actuation 218.4 l Pressurizer starts to feed 220.0

[

(Refer Figures A&B-10 through 20) 2 1

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Revision 2 4/80

- . - - . . . . .c. .,_ .m - . . - . . _,.,,7. . _ . . . , _ . _ _ . ,e, _ . , ..y- . +-.v-- -

.. - - . . . . . _ - _ . . . . _ . . - _ . _ = . - . - - . - - . .. . . ... . -_- .- . .-.

i i 1 RESPONSE TO 10 CFR 50.54(f)

' TABLE A&B-9 CASE 2 [

[

MAIN FEEDWATER OVERREED. LOOP AT TRIP

. Event Time (sec) i r Turbine trip 0.0 {

! Turbine stop valve closed 0.01 ,

Reactor trip + LOOP initiation 0.4 l Turbine bypass valves open 3.0 l 4.0 2 ,

j Atmospheric dump vlaves open Pressurizer empty ~ 222.0 ,

Low RC pressure ESFAS 233.21 -

240.71 MSIVs close

,1 248.21

! MWIVs close i

HPI actuation 263.21 j

Pressuriter starts to feed ~268.00 ,

(Refer to Figures A&B-21 through 27) t 1

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Revision 2 4/80 l

J RESPONSE TO 10 CFR 50.54(f) ,

TABLE A&B-10 l J i

' CASE 3 MAIN FEEDWATER OVERFEED LOOP AT ESFAS Time (sec) i Event i L

Turbine trip 0. 0 ,

Turbine stop valves closed 0.01 Reactor trip 0.4 Turbine bypass opens 3.0 1

Atmospheric dump valves open 4.0 2 1 Pressurizer empty ~175  ;

Tow RC pressure ESFAS 188.45  !

+ LOOP initiation l MSIVs close 195.95 MFWIVs close 203.45 i HPI actuation 218.45 Pressurizer starts to feed ~220. 0  ;

I P

{ l

.l (Refer to Figures A&B 28 through 34) '

i e i i

0 i

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) i TABLE A&B-11 CASE 4 MAIN FEEDWATER OVERFEED LOOP AT ESFAS, NO DECAY HEAT Event Time (sec)

0. 0 Turbine trip

' Turbine stop valve closed 0.01 &

Reactor trip 0.4 i

Turbine bypass valve opens 3.0 l

4.0 2 Atmospheric dump valves open 147.0 Pressurizer empty Low RC pressure ESFAS 154.26

+ LOOP event initiation .

' 16.176 MSIVs close .

169.26 MFW1Vs close e 184.26 HP1 actuation 186.00 .

Pressurizer starts to feed (Refer to Figures A&B-35 through 41) f I

l t

2-Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) ,

O TABLE A&B-12

[

SMALL STEAM LINE BREAK i

(PUMPS RUNNING, CASE 1)

Time  !

Event (sec.)

Rupture 0.0 High flux trip setpoint 4.8 f

Rod movement starts 5.2 Low RC pressure ESFAS 19.0  !

Pressurizer empty 20.0 MSIV closes at 26.50 2 i Secondary low pressure ESFAS 27.67 MFWIV closes at 34 .0 HPI starts to flow at 44.0  ;

Auxiliary FW initiation to 48.70 O good SG starts at Pressurizer starts to refill at 260.0 ,

l l

(Refer to Figures A&B-51 through 59)

B I

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i Revision 2 4/80 1 I- -- -- .- . - . _ ___

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-13 SMALL STEAM LINE BREAK (PUMP TRIP AT REACTOR TRIP, CASE 2)

Time .

Event (sec.)

Rupture 0.0 High flux trip setpoint + 4.8 loop event initiation i Rod movement starts 5.2 ,

Low RC pressure ESFAS 23.48 Pressurizer anpties ~ 24.0 2 MSIV closes at 30.98 I

MFIV closes at 38.48 HPI flow starts at 53.48 Auxiliary FW initiation to good SG ~ 65.0 O Pressurizer starts to refill at ~ 320.0 (Refer to Tables A&B-60 through 68) i l r.

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-14 SMATL STEAM LINE BREAK (PUMP TRIP AT ESFAS, CASE 3) _

Time Event (sec.)

Rupture 0.0  :

High flux trip setpoint 4.8 Rod movement starts 5.2 Low RC pressure ESFAS + loop 19.0 i event initiation Pressurizer empties at 20.0 2 '

MSIV closes 26.50 MFWIV closes at 34.00 l O HPI flow starts at 49.00 Auxiliary FW initiation to good 78.50 SG starts at i Pressurizer starts to refill at 320.0  !

I (Refer to Tables A&B-69 through 77) j -

I  :

9 a '

Revision 2 I

- 4/80

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-15 SMALL STEAM LINE BREAK (PUMP TRIP AT ESFAS NO DECAY HEAT, CASE 4)

-' . l Time Event (sec.)

Rupture 0.0 High flux trip setpoint 4.8

~

Rod movement starts at 5.2  ;

Low RC pressure ESFAS + 16.53 loop event initiation Pressurizer closes at

• 20.0 2 MSIV closes at 24.03 t

MFWIV closes at 31.53 HPI starts to flow at 46.53 Auxiliary FW initiation to good 72.5 SG starts at ,

Pressurizer starts to refill at s 415.0 l

(Refer to Tables A&B-78 through 86) l l

Revision 2 l

4/80

RESPONSE TO 10 CFR 50.54(f) <

O TABLE A&B-16 SMALL STEAM LIN'I BREAK t

(PUMP TRIP AT ESFAS. NO DECAY HEAT. STUCK-OPEN RELIEF VALVE, CASE 5)

Time Event (sec.)

Rupture 0.0 High flux trip setpoint 4.48 l Rod movement starts at 4.88 j 15.0  !  !

Low RC pressure ESFAS + loop initiation 2 ;

Pressurizer empties at N 20.0  !

MSIV closes at 22.5 MFWIV closes at 30.0 l HPI starts to flow at 45.0 Auxiliary FW initiation to good 86.0 I  :

i  !

steam generator at Pressurizer starts to refill 450.0 l

4 (Refer to Tables A&B-87 through 95) ,

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1 O Revision 2 4/80 1

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-17 SMALL STEAV. LINE BREAK (PQfP TRIP AT ESFAS NO DECAY HEAT. ONE HPI FAILURE. CASE 6) i i

Time j Event (sec.)  : ,

Rupture 0.0 ,

High flux trip'setpoint.. 4.67 5.07 Rod movement starts i 2 j Low RC pressure ESFAS + loop 16.25 event initiation ,

23.75  :

MSIV closes at i

Pressurizer empties s 25.0 i

31 .25 i MFW1V closes at 46.25 HPI starts to flow at Auxiliary FW initiation to good 85.7 (

generator l l

1 .

I i (Refer to Tables A&B-96 through 104) l i

i I

Revision 2 4/80

RESPONSE TO 10 CPR 50.54(f)

I2 TABLE A&B-18 J DOUBLE ENDED STEAM LINE BRFA!;

CASE 1 - NO LOOP .

SEQUENCE OF EVENTS TIME, s EVENT Double Ended Rupture of 33.5" ID Steam Line Betueen SG and MSIV 0.0 Closure of Turbine Stop Valves 0.00 Reach Lou RC Pressure Setpoint 1.5 2.2 Control Rod Insertion Starts Reach Lou Steam Pressure ESFAS Sctpoint 1.9 Low RC Pressure ESFAS .

5.8 MSIV's Closed 9.4 MFW1V's Closed 16.9 ,

Unisolated SG Dry out 20.0 Pressurizer Empty 15.0 Auxiliary Feeduater Initiation to Good SG 26.9 HPI Injection Starts . . 30.8 Pressurizer Starts to Fill Up 215.0 SG Tube Region Full of liquid 430.0 (Refer Figures A&B-105 through 116) - 2

) '

Revision 2

. 4/80

"' "' ~~ ' - -

. *am m, ._,_,,,,..,.,,,..,.y.__. - - . .. . $

i

- RESPONSE TO 10 CFR 50.54(f) , ,

( ,

TABLE A&B-19 DOUBLE ENDED STEAM LISE BREAK ,

~

~

, CASE 2 - LOOP AT TRIP ,

i . SEQUENCE OF EVENTS EVENT .

k'IME,s t j

s I Doubic Ended Rupture of 33.5" ID Stea:a Line Ectueen SG and MSIV '

0.0

! Closure of Turbine Stop Valves O.00 Reach Lou RC Pressure Sctpoint + LOOP Initiation 1.5 2.2 f Control Rod Insertion Starts Reach Low Sten:a Pressure ESPAS Sctpoint 1.8 Low RC Pressure ESFAS 7.0  ;

j HSIV's Closed 9.3 i

MFWIV's Closed 16.8 Pressurizer Enpty ,

18.0 Unisolated SG Dry Out 20.0 HPI Injection Starts 37.0 I Auxiliary Feeduater Initiation 59.4 Pressurfzcr Starts to Fill Up .

60.0 .

SG Tube Region Full of Liquid 380.0 ,

(Refer Figures A&B-117 through 125)

Revision 2 4/80 i ,.

, er . , , , ,- -y-,_ .. -._ -_, . . .,y . - , _ _ -,..-c _._ u ,.-. -,_e .

y- - - . , __ ,-- _ _.. , - , --- ,

. RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-20 DOUBLE ENDED STEAM LINE EREAK CASE 3 - LOOP AT ESFAS CEQUENCE OF EVENTS

. EVENT TIME, s Double Er?.d Rupture of 33.5" ID Steam Line Ectween SG and !!SIV 0.0 Closure of Turbine Stop Valves. 0.0 Reach Low RC Pressure Setpoint l.5 Control Rod Insertion Starts 2.2 Reach Lew Steam Pressure ESFAS Setpoint 1.9 Lov RC Pressure ESFAS + LOOP Event Initiation 5.8 HSIV's Closed 9.4 MFh'IV's Closed 16.9 Pressurizer Empty 17.0 Unisolcted SG Dry Out 18.0 O'

~-- IIPI Injection Starts 35.8-Auxiliary Feedwater Initiation 55.5

' Pressurizer Starts to Fill Up 80.0 SG Tube Region Full of Liquid 380.0 l

l2 (Refer Figures A&B-126 through 134) 4

~

O' -

Revision 2 4/80

..m- . ,

• =a * * *

• t

RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-11 l2 DOUBLE ENDED STEAM L1::E BREAK

. CASE 4 - LOOP AT ESTAS,1;0 DECAY llEAT i

SEQUENCE OF EVEiTS EVI:!!T TIliE s 4

! Double Ended nupture of 33.5" .ID Stea:s Line Letsteen SG and t!SIV 0.0 i Closure of Turbine Stop Valves 0.0 Reach Low RC Pressure Sctpoint 1.5 Control Rod InscrLion Starts 2.2 Reach Low Stea.-i Pressure ESFAS Setpoint 1.9 Low RC Pressure ESFAS + LOOP Event. Initiation 5.8

}1SIV's Closed 9.4 ,

}1FWIV's Closed 16.9 _

i .

Pressurizer rapty 17.0 l

Unisolated SG Dry Out 20.0 HP1 Injection Starts '. 35.8 Auxiliary Feeds.ater Initiation to Good SG 55.5 -

Pressurizer Starts to Fill Up 330.0 SG Tube Rer, ion Full of Liquid 380.0 .

(Refer Figures A&B-135 through 143) .

j .

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Revision 2

• 4/80 f .

4 L_ _-_ __- _ _ _ _ . . . , . - . - . . _ . . .

RESPONSE TO 10 CFR 50.5'4(f)

DOUBl.E ENDED E, ! LI ;E EREAK CASE 5 - LOOP AT TRIP, HP1 FAILURE, NO STUCK RELIEF VALVE 4

SEQUENCE OF EVCiTS EVENT TIME, s Double Ended Rupture of 33.5" ID Steam Line between SG and !!SIV 0.0 Closure of Turb nc Stop Valves 0.0 i

Reach Low RC Pressure'Sctpoint + LOOP Event Initiation 1.5 Control Rod Insertion Starts 2.2 5

Reach Low Steam Pressure ESPAS Setpoint 1.8 Lott RC Pressure ESFAS 7.0 MSIV's Closed 9.3 HFWIV's Closed 16.8 Pressurizer E=pty 18.0 Unisolated SG Dry Out 22.0 HP1 Injection Starts 37.0

! Auxiliary Feedtrater Initiation to Good SG 57.0 SG Tube Region Full of Liquid 310.0 Pressurizer Starts to Fill Up 430.0

- (Refer Figures A&B 144 through 152) i' t

• 4 I

• Revision 2

,4/80

• w -

,n-,- ,., . - , , . - -.- , . _ . ,

, , , ,g y , mm mm e

+ - + ~

RESPONSE TO 10 CFR-50.54(f)

TABLE A&B 23 l2 DOUBLE E!!DED STFJJI LI:!E EREAR

( CASE 6 - LOOP AT ESFAS, HPI FAILURE,

. NO STUCK OPEN RELIEF VALVE EVENT ' TI!!E, s Double Ended Rupture of 33.5" ID Steam Line Between SG and MSIV 0.0 Closure of Turbine Stop Valves 0. 0 Reach Low RC Pressure Setpoint 1.5 Control Rod Insertion Starts 2.2 Reach Low Steam Pressure ESFAS Setpoint 1.9 Lou RC Pressure EsFAS + LOOP Event Initiation 5.8 MSIV's Closed 9,4  :

b MFl?IV's Closed 16.9 Pressurizer Empty 17.0 Unisolated SG Dry Out 20.0 HPI Ihjection Starts 35.8 Auxiliary Feeduater Initiation to Good SG 54.5 SG Tube Region Full of Liquid 320.0 Pressurizer Fill Up Starts 450.0 (Refer Figures A&B-153 through 161)

. ~. .

Rev ision 2 4/80

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

- =- - - - - - - - - - - - - ~ ~ ~ ' ' ~ ~ " * '

L _ . . . . - . . . . - . . . - .

~'

3 i

RESPONSE TO 10 CFR 5 '0.54(f) 2 TABLE A&B-24 l DOUltLE ENDED STE/Ji LINE BREAK CASE.7 - LOOP AT ESFAS,  ;

IIPI FAILURE, NO DECAY llEAT, NO STUCK OPEN RELIEF VALVE <

EVENT TI!!E , s .I U

t Doubic Ended Rupture of 33.5" ID Sten::: li Line Between SG and MSIV 0.0

. Closure of Turbine Stop Valves 0.0 Reach Low RC Pressure Setpoint 1.5  ;

Control Rod Insertion Starts 2.2 Reach Lou Steam Pressure ESTAS Setpoint 1.9

' Low RC Pressure ESTAS + LOOP Event Initiation 5.8 I

9.4

) MSIV's Closed . .

MFWIV's Closed 16.9 -

, Pressurizer Empty 17.0  ;

Unisolated SG Dry Out 20.0 i HPI Injection Starts - 35.8 Auxiliary Feedwater Initiation to Good SG 54.5 i SG Tuic Region Full of Liquid 290.0 2

{ReferFiguresA&B-162through170) l .

a O '

iQ)

Revision 2 i

4/80 I

I

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

g. 7 - . . .

, RESPONSE TO 10 CFR 50.54(f)

TABLE A&B-25 l2 DOUBLE E!!DED STEAM LI!iE EREAY, n

, CASE 8 - SLB ON OPPOSITE LOOP FROM PRESSURIZER, LOOP AT ESFAS, HPI TAILURE, l

NO STUCK OPE:1 RELIEF VALVES

• TIME, s

- EVE!;T i Doubic Ended Rupture of 33.5" ID Steam Line Between SG and MSIV 0.0 .

Closure of Turbine Stop Valves ' 0. 0 Reach Low RC Pressure Sctpoint 1.5 j Control Rod Insertion Starts 2.2 i Reach Lou Steam Pressure ESFAS Setpoint 1.9 Lou RC Pressure ESTAS + LOOP Event Initiation 5.8  :

MSIV's Closed 9.4 l

16.9 MFWIV's Closed Pressurizer Empty 17.0 Unisolated SG Dry Out 20.0 HPI Injection Starts 35.8 Auxiliary Feedwater Initiation to Good SG 54.5 SG Tube Region Full of Liquid 350.0 l

f

• (Refer Figures A&B-171 through 179) 9 '

e

+ e 4

1. 6 e

i -- ..

RESPONSE TO 10 CFR 50.54 (f) i TABLE A&B-26 l2  ;

I DOUBLE ENDED RUPTURE LEVEL CONTROL, CASE 9 [

EVENT TIME (SEC.)

Double Ended Rupture 0.0 of 33.5" ID Steam Line i

between SG and MSIV 0.001 [

Closure of Turbine Stop Valves ,

Reach Low RC Pressure 1.483 t

Setpoint  :

r 2.183 ,

Controls Rod Insertion Starts Reach Low Steam Pressure 1.9 ,

l ESFAS Setpoint Low RC Pressure ESFAS, 5.81 LOOP Event Initiation l 9.4 r MSIV'S closed .

t 16.9 MFWIV'S closed i 17.0 PZ Empty Unisloated SG Dry Out %20.0 EPI Injection Starts 35.81 Aux. FW to Good SG 54.5

/

l (Refer to Figures A&B-42 through 50) l 2l 1

( \

I Revision 2 4/80 l

RESPONSE TO 10 CPR 50.54(f) operating Transient Cveles 2 TABLE AGB- 27 O Transient .

Design Number Transient Description Cycles lA Heatup from 70*F to 8% Full Power (Normal) 240 IB Cooldown from 8% Full Power (Normal) 240 2 Power change O to 15% and 15 to 0% (Normal) 1440 r 3 Power Loading 8% to 100% Power (Normal) 48,000 4 Power Unloading 100% to 8% power (Normal) 4S,000 5 10% Step Load Increcse (Normal) 8,000 6

10% Step Load Decrease (Normal) 8,000 7 Step Load Reduction (100% to 8% Power) (Upset)

Resulting from turbine trip 160 Resulting from electrical load rejection 150 Total 310 I

Reactor Trip (Upset) 8 1 Type A 40 ,

,,) Type B 160 Type C 88 Tripr. included in transient numbers 11, 15,16,17, & 21 112

Total 400 9 Rapid Depressurization (Upset) 40 +

i 10 Change of Flow (Upset) 20 Rod Withdrawal Accident (Upset)

I 11 40 12 liydrotests (Tect) 20 13 Steady-State Pouer Variations (Normal) ~

14 Control Rod Drop (Upset) 40 '

15 Loss of Station Power (Upset) 40 16 Steam Line Failure (Paulted) 1 l l

17A Loss of Feeduater to One Steam Generator (Upset) 20 4

17B Stuck Open Turbine Bypass-Valve (Emergency) 10 LO Revision 2 4/80

i

- RESPONSE TO 10 CFR 50.54(f) i TABLE A&B-27 (Cont'd) Design

' Transient _ Cycl es gber Transient Description 40

' \ 18 Loss of Feedwater Heater (Upset Feed and Bleed Operations (Normal) 40,000 19 30,000 5

20 Miscellaneous A (Normal) 20,000 Miscellaneous B Misec11eneous C 4 x 10 6 ,

f 1

21 Loss of Coolant (Faulted) f Test Transients - High Pressure Injection System l 22 40 (Test) 1 Core Flooding Check Valve 240 23 Steam Generator Filling, Drcining, Flushing and Cleaning (Normal)

I Steam Generator Secondary Side Filling ,

i Condition 1 120 i Condition 2 120 Steam Generator Primary Side Filling i Condition 1 120 Condition 2 120 .

Flushing 40 Chemical Cle::ning 20 I Total 540 .

24 Ilot Functional Testing (Test) 1 i

I  !

i a

I I f

i Revision 1  !

12/79 [

t

, RESPONSE TO 10 CFR 50.54(f) i 2

TABLE A&B-28 l RPS/ESFAS FPIQUENCY i

i Actual Data A11oved

} Frequency Frequency i

Number No. of Reactor Trips (RPS) 310 8.9/yr 10/yr l 1

l.

No. of Automatic ESPAS Actuations 27 .816/yr 1.0/yr j No. of Plants Included 9 Approx. 35 1 Reactor Years 4

i i

i i

i i

e I

i 1

I-i I

l i

1 r

d Revision 2 4/80

• i 83

80 91 t,

g u _ _ @_ _ _ _ ,,

g. I 57 81 "

o ,,

o O @ g, ,

i, i.

-. 4 n g_

g ~

~g 54

_Q @~ "

~

a 88

@ is @

41 19 ft g a n l __

@ .1 p a n g, I n

i. ,, --

~~

s u 44 22 II 33 @

@ n u 43 23 T__@_ er n u

@ g g g g

_ ec y _e _

n- u -

- n n -

g7 19 n "liiT

" UnG uT st le n I u I I si I g "

st es Il

__ @l sati. fear usetus scutst. irt Fa Finste A&P-1

} t Revision 2 -

4/80 f

9

@ ~, 23  %

34 22 38 I~~9 10 dB I 31 h L _ _ .)

n @ 32 y g  % _

IS

• 21 a

101 FW 10 F

O._

e

]@ T C @[ -

~

. V T 3 i.

18 27 12 e e.

a, 18 13

@ g _

R_%'

i. .. o g

T D

@ g g

$$ sA 9

g 41

@h -

g @

i "g . l 53 S4 35 M l [

38 I i '

' ' Stel TRAP ismills ICII(M.177 FA fagers A&B-2 St 52 O

r-

.. N@R GRESL _ ... .

t a

_ , _ _u - - - - ,- --- ,- - - - -- --

l l

l RESPONSE TO 10 CFR 50.54(f)

STEAM REGULATOR WALFUNCTION, 177 FA l2 j FIGURE A&B-3 120 100 b O- -

I 2

80 - -

0 60 -

3 40 --

20 --

. L '

0 500 600 300 400 100 200 O

Time, Sec O Revision 2 4/80 1

O  !

l RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-4 STEAM REGULATOR MALFUNCTION,177 FA ,

2200 ,

i 2000 -  !

i 7

5 1800 - ,

5 O

e=8 O 2 1600 - .

, E ca S

1400 -} I f

1200 -

I 1000 500 600 300 400 0 100 200 Time, Sec

. O Revision 2 4/80

RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-5 STEAM REGUL ATOR MALFUNCTION, 177 FA l2 l

l 620 r

600 -

l 3

580 -

560 -

• 540 -

.i E

% HOT LEG

520

=

0 U=

500 -

1 I

480 -

460 -

440 -

COLD LEG l l

420 - l 1

. I f I e e I a fQQ 200 300 400 500 600 0 100 Time, Sec O

Revision 2 4/80

..7

, - - . ~ -

i RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-6 ' STEAN REGUL ATOR NALFUNCTION,177 FA 2 22.14 i

16.60 -

11.07 -

2 E

E 5.53 -

, L , . . , ,

600 200 300 400 500 0 100 Time, Sec F

I e 6

0 e

O Revision 2 t

4/80 1 . _ ,

E

'T RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-7 STEAM REGULATOR MALFUNCTION, 177 FA l2 900 800 -

l 5

5 700 -

O h E

y 600 -

w

500 -

2 5

w 2  ;

400 -

i i

l 300 -

200 200 300 400 500 600 O 100 Time, Sec Revision 2 4/80

._, -- --- - . . y. g.

W FIGURE A&B-8 STEAM REGUL ATOR NALFUNCTION,177 FA l2 ,

1000 .

900 -  :

f

800 -

5 5

700

=

-li 8

f

=

600 -

r 0  !

g 500 3 '

f,

" l 400 -

300 -

200 400 500 600 100 200 300 O

Time, Sec O .

Revision 2 4/80

--A-

RESPONSE TO 10 CFn 50.54 (f)

FIGURE A&B-9 STEAM REGUL ATOR WALFUNCTION,177 FA 2 O ,,

S.G. "B" lS FULL e 432 SEC 50 -

45 -

3 0 "A" S.G. "B" 1

IS FULL 3 40 -

438 SEC C 4

.f 35 - ,,

• l a

30 "A" 3" 30 E

25 -

O i 20 -

15 10 5

0 200 300 400 500 O 100 Time, Sec i

O' ~

Revision 2 4/80 l _ _ . . . -

l _

RES PONSE 'IO 10 CPR 50. 54 ( f )

FIGURE ACB-10 NFW DVERFEED, TURBINE TRIP, REACTOR TRIP-177 FA l2  ;

CORE AVERAGE TEMPERATURE VS TIME FOR NAXI MINI j TRAP COMPARISON 580 l -

l l I

. l 576 - i t

572 -

i l

568  ;

p -

{

N B

= i g 564 -

" l O;  ;

5 560 -

E '

e s N

3 g 556 - N

\

\ NINI-TRAP MODEL 552 -

\

NAXI-TRAP N0 DEL N s

! 548 - N l N ,

s N

\

A 544 -

s I  !  !  !

l l l 60 80 100 120 140 160 0 20 40 Time, see Revision 2 4/80

f RESPONSE 'IO 10 CFR 50.54( f)

FIGURE ASB-11 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP-177 FA l2 CORE OUTLET PRESSURE VS TINE FOR NAXI.NINI TRAP COMPARISON 2200 1

1 1

2150 I

l 1

I 2100 -[

l 2050 -

L El.

- \

E \

} 2000 -

5 m

O !

g 1950 -

5 I g

\

1900 -

g s

N WINI-TRAP N00EL N

1850 N NAXI-TRAP N00EL N N

\

\

1800 -

s

\

\

N N

1750 -

N N

N I I I I "I' I I 1700 O 0 20 40 60 80 100 120 140 Revision 2 4/80

RESPONSE 70 10 CFR 50.54(f)

FIGURE AGB-12 NFW OVERFEED, TURBINE TRIP, REACTOR TRIP-177 FA l2 f PRESSURIZER LEVEL VS TIME FOR MAXI MINI TRAP COMPARISON O i t

20  ;

[L ,

\ i

\

18 -

\ l

\  ;

\ >

16 -

\ f

\ i

\ i

\ .-

14 -

g r \

\  !

" l 12 - \

" \  !

" \ MINI-TRAP MODEL 10 -

N  :

0 2 N  !

\

8 -

N

\

N [

% i N i 6 - s NAXI TRAP N0 DEL s i

\ l

% l N  ;

N l 4 -

I l

l l

t 1

i i 1 I I

- -- I I 1 80 100 120 140 160 j O 20 40 60 I" ' "C Revision 2 4/80 1

RESPONSE TO 10 CFR 50.54 (f) l l l FIGURE A&B-13 NFW OVERFEED, TURBINE TRIP, REACTOR TRIP CASE 1, l2  !

O 177 FA RCS TENFERATURE VS TIME i

800 -

. I 590 580 -

l m 570 -

e M\ f 5 560 i

f

/ s N

O *E  :

i

\ HOT LEG U

= \  !

550 - f

\

l

\ '

\

COLO LEG \

540 l

\

\

\

\  :

530 -

g\  %  %~  %' -  % ,

520 -

l 510 -

l 500 -

i I 1 1 1 t 300 400 500 600 ,

O 100 200 Time, Sec Revision 2 6 4/80 l

- - - , , . . . . - - . - , - - - a .. . _. , - - . . - . .

RESPONSE TO 10 CFR 50.54 (f) 2 FIGURE A&B-14 NFt OVERFEEO. TURBINE TRIP. REACTOR TRIP-CASE 1 177 FA STEAM GENERATOR LEVEL VS TlWE l

FULL 50 -

/

/  !

45 -

l i 40 -

/

/ .

/ i

. 35 -

3

/

l E

30 -

/  ;

5 STEAM GENERiTOR B O e 0

a 25 -

/

STEAM GENERATOR A l

E /  ;

/  !

/ l 20 -

/ '

/

/  :

I 15 /

l 10 l

5 I '

O O 50 100 Revision 2 Time, sec 4/80

- -~

, O O O RESPONSE TO 10 CFR 50.54(f) l 2 FIGURE A&B-15 NFW OVERFEED. TUR8INE TRIP, REACTOR TRIP-CASE 1, 177 FA PRESSURIZER LEVEL VS TINE 38.74 33.20 -

0 27.67 -

t

- 22.14 -

O

~

16.60 -

11.07 -

5.53 -

kN m<

o p-I f l I 1 I i

p. g f 20.0 25.0 30.0 35.0 40.0 45.0 50.0 0 0.0 5.0 10.0 15.0 w Time, Sec X 10

RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-16 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP CASE 1, l2 177 FA SG A SECONDARY OUTLET PRESSURE VS TiuE 12.0

. I

@ 11.5 2 t

. 11.0 -

a i

1 h i

[ 10.5 -

5  %

5 _- a m

10.0 -

i

]. ,

E E

m o 9.5 -

i o  ;-

o' l "

9.0 8 ' I I I I I I '

" 20.0 25.0 30.0 35.0 40.0 45.0 50.0 0.0 5.0 10.0 15.0

' Time, sec X 10

RESPONSE TO 10 CFR 50.54 (f)

' FIGURE A&B-17 NFf DVERFEEO, TURBINE TRIP. REACTOR TRIP-CASE 1,177 FA l2 SG B SECONDARY OUTLET PRESSURE VS TINE 12.0 ,

! E 11.5 -

.=

E

- 11.0 -

i =

i E C

y 10.5 -

2

=

5 m 3 O 10.0 -

5 i u e

s, Rif co < <m s.5 -

op g- (

s I I I I I I I I I

u 9.0 25.0 30.0 35.0 40.0 45.0 50.0 I 0.0 5.0 10.0 15.0 20.0 Time, sec X 10

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

O ~

O O RESPONSE TO 10 CFR 50.54(f) .

4 FIGURE A&B-18 NFW OVERFEED, TURBINE TRIP, REACTOR TRIP-CASE 1, 177 FA l2 CORE DUTLET PRESSURE VS TINE i -

i 24 22 '

c3

\

D4 20 -

E ai

, - g i g 18 -

E U

3 c) 16 -

E

. 3

. 14 -

i D$

o

- , I O g i l l t l l gg 20.0 25.0 30.0 35.0 40.0 45.0 50.0 w 0.0 5.0 10.0 15.0 Time, Sec X 10

n

--. 1 I

i L

o '

o r

w

=

8 e i

o, w

m a

u.

4.

oc  !

ss:

o I e-o =

. o W -

n

_ =

u .m ,

- n. m n a w -

ae .- "

u

~

o m

w x

0 -

_w ,

g = = ._

sa: o ~

D B- = e

.w - o o o o. ~

a w =

Ww i wW

=

@ w w

- o m o == -

(n z =

w >

=

O ma D. i tn '

sq m g GI. H - o I

sa us ,

4  ;

m -

a:

o I O <

H ,

N

. , . , i ,

o o o o a o e o e a m w m ~ >

o a m e-e e e e

= m e e e J.'ajn}Jadesi alejaAy ajo3

. Revision 2 4/80 m 4

--7-,a@ 4 g y - - - - - . - , - - - .

1

. RESPONSE TO 10 CFR 50. 54 (f) ,

FIGURE A&B-20 NFW OVERFEED. TURBINE TRIP, RE ACTOR 1 RIP-CASE 1 l2 177 FA TOTAL POWER VS TINE

' l.2 I

l.0 Y t

.8 -

.E .

, i E *6 -

W

.4 -

a us No .2 _

m<

or

s I I I I ' ' ' ' '

u 0.0 25.0 30.0 35.0 40.0 45.0 50.0 O.0 5.0 10.0 15.0 20.0 Time, sec X 10 i

l i

RESPONSE TO 10 CPR 50.54(f)

FIGURE A&B-21 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP, LOOP l2 AT TRIP CASE 2, 177 FA 100 l

80 -

E

=

,- 60 -

E O =

3 AD .

20 -

0

( ' ' ' ' ' '

300 A00 500 600 O 100 200 Time, Sec I

Revision 2 4/80 e O

+-

--r -

O RESPONSE TO 10 CFR 50.54 (f) ,

FIGURE A&B-22 NFW OVERFEEO, REACTOR TRIP, TURBINE TRIP, l2 LOOP AT TRIP CASE *2, 177 FA 2200 2000 -

2 E .

E 1800 -

0

. a.

O  ;

g 1600 -

?

C 3

1400 -

1200 400 500 600 O 100 200 300 Time, Sec O Revision 2 4/80

e a z - + = .- --- s-r -- - w - - -- - - - - - . _ > ~-

RESPONSE TO 10 CPR 50.54(f)

FIGURE A&B-23 IIFW OVERFEED CASE #2,177 FA HOT LEG 550 -

k i

COLD LEG e

. 500 -

U l

i 450 -

G

, t I Time, Sec O Revision 2 4/80 i

l l

RESPONSE TO 10 CFR 50.54(f) [

O  :

f i

FIGURE A&B-24 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP, LOOP AT l2 [

l t

TRIP CASE #2, 177 FA i

27.67 22.14 - l i

1

=

16 . 60 -

l i

e .

3 11.07 -

O .".

I 5.53 -

f h  !

200 300 400 500 600  ;

0 100 Time, Sec i

l l

O Revision 2 4/80

l i

I RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-25 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP, LOOP AT TRIP CASE #2 177 FA 1150

%e .  ;

E i

1100 -

, l 2 .

E l o (

h j 1050 -

(

52

-u.

5* l  :

g 1000 - 7

% I 5

=

950 l g -

3 u,

i 900 ' '

300 400 500 600 -

0 100 200 Time, Sec -

l L

p i

Revision 2 4/80 l

l

O i RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-26 NFW DVERFEEO, TURBlNE TRIP, REACTOR TRIP, LOOP AT TRIP CASE #2, 171 FA 1150 N

1100

= I si l k

E  !

0 1050 -

L_

O= _

E

, 1000 -

f c

i

". 950 -

a M

900 500 600 300 400 O 100 200 Time, Sec O

~

. Revision 2 l i

4/80 l

l

FIGURE A&B-27 NFW OVERFEED. TUR8INE TRIP, REACTOR TRIP, LOOP AT TRIP CASE *2 O

FULL

/

/

50 -

j

/

/

/  :

/  :

40

,/

~

/

/

5 SG 2 /

a a 30 -

? /

w I l

= /

20 -

j f SG I 10 -

l l

l

' ' ' ' i 0

50 75 100

- 0 25 Time, Seconos O Revision 2 4/80 l

O RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-28 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP, t.00P

^

AT ESFAS CASE #3, 177 FA 120 100. -  :

80 3 -

e E

E 60 -

O i 40 -

20 -

Y , , i i .- ,

0  :

400 500 600 100 200 300 0

Time, Sec l0 i .

- Revision 2 4/80 I

i

-. .. ~.

O RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-29 IIFW OVERFEED, TURBINE TRIP, REACTOR TRIP, LOOP AT ESFAS CASE #3, 177 FA 2400 2200 -

2 E. 2000 -

i 5

O2 E 1800 -

t a

5 1600 -

u 1400 -

1200 100 200 300 400 W W O

. Time, Sec i

O Revision 2 4/80

. - - - . - - . - - , _ - , - - - , - - - , - - y , - - , . - _ - , ---

RESPONSE TO,10 CFR 50.54 (f) l NFW DVERFEED, TURBINE TRIP, REACTOR TRIP, LOOP FIGURE A&B-30 AT ESFAS CASE #3, 177 FA 600 -

l r

HOT LEG g 550 E

O  !

U

" COLD LEG l

l 500

- i 450 500 600 300 400 O 100 200 Time, Sec Revision 2  ;

3 4/80

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

r I

O RESPONSE TO 10 CFR 50. 54 (f)

FIGURE A&B-31 MFW OVERFEEO, TURBINE TRIP, REACTOR TRIP, LOOP AT ESFAS CASE #3, 177 FA 33.20 a

P 27.67 -  :

O 22.14 -

.- I O s a

$ 16.60 -

I 5 I O '

11.07 -

i 5.53 -

0 400 500 600

- 0 100 200 300 Time, Sec lO( Revision 2 4/80

,-r

- --- r -

b ll O

RESPONSE TO 10 CFR 50.54(f)

IIFW OVERFEED. TURBINE TRIP, REACTOR TRIP, LOOP FIGURE A&B-32 i AT ESFAS CASE #3, 177 FA 1150 l E 1100 - t b

a i

1050 -

t b D r

1000 5

i 5 ,

• 950 -

l m

900 i 400 500 600 100 200 300 O

Time, Ssc O

Revision 2

' 4/80

O  :

l RESPONSE TO 10 CFR 50.54(f)  :

I FIGURE A&B-33 NFf OVERFEED, TURBINE TRIP, REACTOR TRIP, LOOP AT ESFAS CASE #3. 177 FA 1150 E 1100 -

5 l U ~

2 1050

{ '

O s E

o l O 1000 0

5 M

950 -

M 900 100 200 300 400 500 600 0

Time, See l -

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

NFW OVERFEED, TURBINE TRIP, REACTOR TRIP, LOOP FIGURE A&B-34 AT ESFAS CASE #3, 177 FA I

/

/  !

50 -

/

/ l

/  !

?

/ '

/ ,

/

40 -

/

/

~ /

i /

/

> STEAM GENERATOR B 3" 30 -

/

/ '

/

) E G /

20 -

/

i

/  !

l

! STEAM GENERATOR A j

l 10 - / '

/

l

/ l l

, , i 0 i 15 100 0 25 50 l l

Time, Sec l

O Revision 2 4/80

i O .

l RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-35 NFW OVERFEED, LOOP AT ESFAS NO DECAY HEAT CASE #4,177 FA 120 p

100 -

80 3

so W

w E

t o;-O

>=

40 l

l l

l 20 -

Q t .

_ t I t l 0 100 200 300 400 500 600 Time, Sec O

Revision 2 4/80

l t

RESPONSE TO 10 CFR 50.54(f)

NFW OVERFEEO, LOOP AT ESTAS NO DECAY HEAT FIGURE A&B-36 CASE #4, 177 FA 2400 ,

f l

2200

2000 a.

J l E 1800 O  ;

E -

. 1600 u

a  ;

t l

1400 -

\

1200 -

t 1000 500 600 300 400 100 200 0

Time, Sec l

l i

O< Revision 2 4/80

- - - _ ,, = - -_

i RESPONSE TO 10 CFP. 50.54(f) 1 FIGURE A&B-37 IIFW OVERFEEO, LOOP AT ESFAS NO DECAY HEAT l O CASE #4, 177 FA I

l

e. l

\' .

600 -

M E

E

=

O y 550 ' -

HOT LEG

\

COLD LEG 500 -

I t . I t _

I '

300 400 500 600 0 100 200 Time, Sec O

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) ,

FIGURE A&B-38 NFW OVERFEED, LOOP AT ESFAS, NO DECAY HEAT.

CASE #4, 177 FA .'

N 33.20 t

27.67 -

l

; 22.14 -

a

16.60 -

a <

t O

t E~ 11.07 -

i 5.53 -

i

' -- ' i 0 600 300 400 500

' O 100 200 Time, Sec l I

l I

. I 1

i  !

l 0 Revision 2 4/80 j l

t i

O  !

RESPONSE TO 10 CFR 50.54(f)  !

i FIGURE A&B-39 NFW OVERFEED, LOOP AT ESFAS NO DECAY HEAT, f CASE #4, 177 FA 1100 i

a 1050 - -

5 a

1000

[- '

E E

o O r 0

950 -

5 w

= 900 -

i ,

I t i e i j 850 300 400 500 600 0 100 200 Time, Sec l

I O Revision 2 4/80 ,

l O

l

~

RESPONSE TO 10 CFR 50.54(f) t l

IIFW OVERFEED, LOOP AT ESFAS NO DECAY HEAT FIGURE A&B-40 CASE #4, 177 FA i

1100 .

i 1050 e

5 i

j 3 1000 - -

E -

o 7 O 8 g '950 E ,

m r

o ' ' ' ' l 900 300 400 500 600 0 100 200 ,

Time, Sec

. I O Revision 2 4/80 l

9 y y , ---r ,e-, - - -- - -

t l

RESPONSE TO 10 CFR 50.54 (f)  :

FIGURE A&B-41 NFW OVERFEED, LOOP AT ESFAS NO DECAY HEAT CASE #4, 177 FA l

(

P FULL '

i l

50 -

/ ,

/

/ i

/ l

/ l r 40 -

j j

. STEAM GENERATOR B ,

= ^

/  :

5 /

3 / l e

O

= 30 -

E 7 E

/

• /  !

! STEAM GENERATOR A

/ r 20 -

j

/

/ l

/

/

10 -

/

0 50 75 100 0 25 ,

Time Sec Revision 2 4/80

O .

RESPONSE TO 10 CFR 50.54 (f) 000BLE ENDED RUPTURE 1.00P AT ESFAS 1 HPI, FIGURE A&B-42 NO DECAY HEAT CASE #9, 177 FA 120 100 -

$ 80 -

e

.L E 60 O -

40 '-

gg ..

k 0 600 l 400 500 100 200 300 0

Time, See  :

1 I

O Revision 2 4/80  ;

RESPONSE TO 10 CFR 50.54 (f)

DOUBLE-ENDED RUPTURE LOOP AT ESFAS 1 HPI, FIGURE A&B-43 NO DECAY HEAT CASE #9, 177 FA 2500 I

I 2000 ,

2"

[

l 1500 -

1 ~

0 -

g

,000 .

=  :

8 ' ' ' ' '

500 '

400 500 600 100 200 300 ,

O I Time, Sec l

l l

. l 1

l Revision 2 l 4/80 j i

I l

O .

RESPONSE TO 10 CFR 50.54 (f) 000BLE-ENDED RUPTURE LOOP AT ESFAS 1 HPI, PIGURE A&B-44 NO DECAY HEAT CASE #9,177 FA 500

. 550 .

~

E -

O a

,a NOT LEG U

500 -

I l

COLO LEG i

i .

i 450 i 400 500 600 0 100 20 0 300 Time, $se o

0; Revision 2 f 4/80 l

RESPONSE TO 10 CFR 50.54 (f) 1 FIGURE A&B-l.5 DOUBLE-ENDED RUPTURE LOOP AT ESFAS 1 HPI, NO DECAY HEAT CASE 9, 177 FA 22.14

- 16.60 -

a i

11.07 -

E O E 5.53 -

O 200 300 400 500 600 0 100 i

Time, Sec ,

I l

l l

i l

l l

! l i

O l

Revision 2 l i

4/80 i

~

l l

l l

l RESPONSE TO 10 CFR 50.54(f) l i

DOUBLE-ENDED RUPTURE LOOP AT ESFAS 1 HPI, FIGURE A&B-46 NO DECAY HEAT CASE #9, 177 FA ,

?

. 1000  ;

E. 800 -

1

- r l

I 600 , -

• L ,

=

O 400 5

m

• 200 -

M 0<

20 0 300 400 500 600 ;

0 100 Time, See Revision 2 4/80

l RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-47 DOUBLE-ENDED RUPTURE LOOP AT ESFAS 1 HPI, M0 DECAY HEAT CASE #9, 177 FA  ;

4 1000 900 - .

EL

~

= '

800 - i Ei

.=  !

=

w l 700 -

l E Q i E. 600,.

500 -

l l

l t f I

[

200 300 400 20 W O 100

! Time, Sec l

l i

1 i

Ii l

l lO  ;

Revision 2 )

4/80 W" l

i RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-48 00U8LE.ENDEO RUPTURE LOOP AT ESFAS 1 HPI, NO DECAY HEAT CASE #9, 177 FA 20 STEAN GEN p . i 15 -

i

.i Oi ,

i STEAN GENERATOR A ENPTIES g

COMPLETELY AT 20 SEC .

, e i f i

g 300 400 500 600 0 100 200 Time, Sec O

Revision 2 4/80

i t

RESPONSE TO 10 CFR 50. 54 (f)

FIGURE A&B-49 DOUBLE ENDED RUPTURE LOOP AT ESFAS 1 HPI, {

NO DECAY HEAT CASE #9, 177 FA  ;

I i

(  !

300 -

i 250 -

=

f S

C 3

g 200 -

O <

E 1 50

=-

E SG UPPER PLENUlf 5 CANDY I

1. CANE 100 t i

50 ' ;-

CANDY CANE HOT L G i '

0 l 100 150 200 250 300 O 50 Time, Sec Rev'ision 2 4/80

RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-50 DOUBLE ENDED RUPTURE CASE #3,177 FA

O 80 -

SG UPPER PLENUM

= 50 O

f C

3 5

40 -

w J

B

=

30 -

NOTE: CANDY CANE &

=

HOT LEG ARE 3 CONSTANT AT "0" 20 -

10 -

f f i

g 15 10 0

5 Time, Sec O. Revision 2 4/80

A. _

O RESPONSE TO 10 CPR 50.54(f)

FIGURE A&B-51 SMALL SLB CASE #1,177 FA 120 100 [

80 . -

a w

60 ' -

E 40 .

20 . .

l ,

O i ,

300 400 500 600 0 100 200 Time, Sec .

k lO Revision 2 4/80 T ww-e--. v-y wrg,- - - y-9 - - ., . , , . ...-.- g -.-

e ---._,.__.m

..,. , - _.-_- ,_e w --- ,e- - 3

O -

RESPONSE TO 10 CFR 50.54 (f) i FIGURE A&B-52 SMALL SLB CASE #1, 177 FA  !

2500 2000 5

0 1500 -

E O i

=

1000 -

u 500 -

0 O 100 200 300 400 500 600 Time, Sec

\

!O l

Revision 2 l 4/80 l

i l

RESPONSE TO 10 CFR 50.54 (f) l O FIGURE A&B-53 SMALL SLB CASE #1, 177 FA 800-580 560 -

540 w

• 520 -

J I E  !

500 -

O if y 480 -

HOT LEG 460 -

440 -

420 -

COLD LEG 400 -

t t 4 f I t 400 500 600 100 200 300 0

Time, Sec O Revision 2 4/80

• e

• e mammasse -=

g- 6M y_ , -- - - - , _ -- - ., - - - , .

e RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-54 SMALL SLB CASE *1,177 FA ,

22.14 I i

- 16.60 -.

~

em E

11.07 -

E O S i

a. 5.53 -

i

' t i t t

g 300 400 500 600 0 100 200 l

Time, Sec O Revision 2 4/80 e

e m.

-j- - - - - . _ - . _ ,,- --

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-55 SMALL SLB CASE #1, 177 FA 1000 b e,

= 800 -

E C

" L a.

j 600 -

8 a

I_" '

p 3 400 -

V E a

80 G 200 -

0 ,

300 400 500 600 0 100 200 Time, Sec 5

Revision 2 4/80 F

a essee ee * * ** e

_e , se o e em e

_.m __,___

" - ' + - - - y

I M

l O

RESPONSE TO 10 CFR 50.54 (f) l FIGURE A&B-56 SMALL SLB CASE #1, 177 FA 1200 i

1000 -

E

{ 800 O

O m 5

U 600 -

CE3 400 -

g 200 200 300 400 500 600 -

O 100 Time, Sec 4

1

O, Revision 2

! 4/80 1

l - _ , _ _ _ - - - _ _ - _

l l

l l

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-57 SMALL SLB CASE #1,177 FA l 55 S. B.B FULL e 218 SEC 50 -

45 -

40 -

35 -

f f.

a 5 30 Sg g O "

i 25 -

E v,

20 15 -

10 SG A STEAM GENERATOR A EWPTIES 5

COMPLETELY e SS SEC 0

100 150 200 250 O 50 Time, Sec ,

lO I

Revision 2 4/80  ;

D

RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-58 SMALL SLB CASE #1,177 FA 350 I

300 -

u

?

0 E 250 -

c) e

$ CANDY CANE 200 -

O i d

fo 150 -

E i

4 1> 100 -

SG UPPER TUBE REGION 50 -

HOT LEG <

I

/ /, \ l '

1

~

A u - '~~'%e - -9 '* -

O 200 300 0 100 Time, Sec O

Revision 2 4/80

r O RESPONSE TO 10 CFR 50.54 (f) e SMALL SLB CASE dl, 177 FA FIGURE A&B-59 140  !

I l

120 en o

% 100 -

0 ,

5 o

E o 80 -

v2 ,

CANDY CANE d

60 -

S s

o

\ SG UPPER

> I I

S 40

/ T

/ ' UBE 3

I t REGION o i

> I \ HOT LEG l .

I \

I I  ;

20 -

I \

I

\

- I \

/ \ ,,

N O l t - i l. l.~.%I P.-f,w... s 200 300 100 0

Time, Sec

, Revision 2 I 4/80

( . - - - --.

l . - - , . -

a __ _ - - , - - - - - - - - - ---- --

RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-60 SilALL SLB CASE #2, 177 FA 120 I

100 -

80 .-

m E

< .e 60 <

2 40 -

20

' +

' 1 0

400 500 600 l 100 200 300 0

Time, Sec I

l Revision 2 4/80

. . - _ . . _ - _ . _ _ - _ . 4

, -- - a u_ u _s a_- -

O RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-61 SMALL SLB CASE #2, 177 FA 2200 .

( $

I l 7 l

2000 - -

1800 _  :

O E

or 3 1600 -

0 E

O =

~

=

1400 -

es  !

8  ;

1200 1000 - ,

800 600 300 400 500 O 100 200 Time, Sec I

O Revision 2 4/80

- - - - , ,p-,--m- , - , , , - - - - - -, , , - ,.,-,-,-w

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-62 SMALL SLB CASE #2.177 FA i

l l

600 -

1 i

l l

j 550 i 1

J l 1

B" ,

a

{ HOT LEG l j 500 -

O U a F l

i l

COLD LEG l l

450 -

l 400 t 300 400 500 600 .

O 100 200 .

Time, Sec O. Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) ,

FIGURE A&B-63 SMALL SLB CASE #2,177 FA 22.14 t r 16.60 -

l

~ i i

O  ; 11.07 -

E  !

E 5.53 - i i

i I

O 400 500 600 l O 100 200 300 i Time, Sec O

O Revision 2

.4/80

.y.-- - -_y _ _ _ _ - -

RESPONSE TO ..10 CFR 50.54(f) b FIGURE A&B-64 SMALL SLB CASE #2, 177 FA 1000-

. 800 i

l

=

e' E

y 600 -

E .

i i

'j 400 -

O M v,

=

8 200 -

I f

f 0 i I 400 500 500 100 200 ~300 0

Time, Sec O Revision 2 4/80

--en- _

\

. O .

1 RESPONSE TO 10 CFR 50.54(f)

?IGURE A&B-65 SilALL SLB CASE #2,177 FA i

1200 m

P 1000 - '

E

= 800 -

Oa g 600 - -

E so E

400 -

200 500, 600 l 300 400 0 100 200  !

l Time, Sec l

l i

/

lO I

Revision 2 4/80 l

t

) RESPONSE TO 10 CFR 50.54(f) l FIGURE A&B-66 SMALL SLB CASE #2, 177 FA O

• FULL I

, i l

. 50 -  :

i 40 - l

= t STEAM GENERATOR B I

a E -

30 5

a 4

5

• l

' t 20 STEAM GENERATOR A 10 i

i e i 0 i 150 200 250 0 50' 100 , l l

Time, Sac

, O Revision 2 4/80 1

w._ _ .. . . _ . .- . - . _ _ _ _ _. - - . -- _ .-.

RESPONSE TO 10 CFR 50.54(f)

O FIGURE A&B-67 SilALL SLB CASE #2, 177 FA 350 NOTE: NO VOIDING IN SG UPPER TUBE REGION 300 -

=

CANDY CANE 3 250 - -

E E

E ,

I u,

. 200 -

i .T O *

! 150 -

E '

E E ,

100 -

HOT LEG l I l 50 -

l e i

. O 200 300 0 100 Time, Sec O Revision 2 4/80

y .e- -*7y e t--- --

e e '" - -' --w' - -

- " " ' = * ' - - " - - - ' r"-'-"wN- r v-w'- 9T'"'ew

O RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-68 SNALL SLB CASE #2, 177 FA

\

NO YOIDING IN STEAM GENERATOR B r

O l

O -

Revision 2 4/80

- - . . - - _ = . - - . . _ . . - . - . - . . - . . - - - -

1 i

i l l l

{

i f

~

i I

1 RESPONSE TO 10 CFR 50.54(f)

' FIGURE A&B-69. SMALL SLB CASE #3, 177 FA I

120 ,

I l-100 l

l 80 . -

=  !

+

t o

& i

__ 60 i l

O

> f L

40 -

I l

20 -

i 400 500 600 100 200 300

- 0  :

Time, Sec. .

Revision 2 I 4/80

,,_ ,,,p.

se y g g agge ,,

6 S vga 6

- - , - ,n, - - - - . , , . , -

l RESPONSE TO 10 CFR 50.54 (f)

SMALL SLB CASE #3, 177 FA FIGURE A&B-70 i

2400 t f

2200 i-2000 - -

i k

b

= 1800 - -

E l

~

1600 O a

=

E I \

i 1400 I

t i

1200 r

1000 g

400 500 600 100 200 300 O

Time, Sec Revision 2 -

4/80

l RESPONSE TO 10 CFR 50.54 (f)  ;

O FIGURE A&B-71 SMALL SLB CASE #3, 177 FA l l

840 l

l 500  !

HOT LEG l

560 -

1 O

5=

520 -

i es i

t O

U

= 480 -

440 -

COLO LEG 400 -

• i ' ' ' '

360 t 100 200 300 400 500 600 0 l Time, see t

O Revision 2 4/80 eow w en m e 54e.* *

. - - _ -. _ ~

[

O -

1 1

I l

l l

RESPONSE TO 10 CFR 50.54 (f) i FIGURE A&B-72 SNALL SLB CASE #3, 177 FA l 27.67 22.14 -

1 16.60 -

O- 2 E,

y11.07 -

a.  ?

s.33 _I

- ' ' ' i 0

400 500 600 0 100 200 300 Time, Sec ,

O Revision 2 4/80

-.~-. .

. . . . . ._.,e

• e~~-

,+ , - _ - 22. - _ . -

O .

b RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-73 SMALL SLB CASE #3,177 FA 1000 5 800 h a >

0 600 -

E 5

m [

O 4 w

400 -

r 200 -

l 0 i 400 500 600 100 200 300 0

Time, Sec .

e 9

O Revision 2 4/80 p -- --

n - .---+

. O RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-74 SilALL SLB CASE #3,177 FA 1200

$ 1000 -

a.

800 -

E 5

a w 600 -

O 400 -

200 400 500 600 O 100 200 300 Time, Sec I

O i

O Revision 2 4/80

l RESPONSE TO 10 CFR 50.54 (f)

SMALL SLB CASE #3,177 FA FIGURE A&B-75 55

- NOTE: SG B FULL 9 254 SEC 50 -

45 -

40 -

SG B C

3 35 -

a i E 30 -

b y 25 -

20 -

l 15 -

t 10 -

NOTE: SG A EMPTIES SG A COMPLETELY e 112 SEC 5 -

e i i

e O

150 200 250 0 50 100 Time, Sec O

Revision 2 4/80

l RESPONSE TO 10 CFR 50.54 (f)

O FIGURE A&B-76 SMALL SLB CASE #3,177 FA 350 l i

300 -

i i

250 - NOTE: THE S.G. UPPER TUBE REGION REMAINS CONSTANT

• AT 0" M

200 -

E  ;

O

" l

" 150 -

l CANDY CANE [

/ '

100 -

HOT LEG 50 -

HOT LEG s

0 l i I' 300 400 50 100 200 0

Time, Sec O: .

Revision 2 4/80

1 RESPONSE TO 10 CFR 50.54 (f)

FIGURE A&B-77 SMALL SLB CASE #3, 177 FA 70 60 -

NOTE: THE HOT LEG 50 -

REMAINS CONSTAN1'AT "0" M

C 40 -

O 5 i SG UPPER TUBE REGION

' > l 3 30 -

E t

l CANDY CANE 20 -

10 -

. i i i e i 0 t i 60 70 80 90 100 10 20 30 40 50 0

Time, Sec O Revision 2 4/80

. - - . a. - - . . . __ h _ ra , a b

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-78 SMALL SLB CASE #4,177 FA 120 100 -

t 80 m

s,."

I d'

60 0

2 40 -

20 -

i E a 0

400 500 600 100 200 300 0

Time, Sec P

e Revision 2 l 4/80

__ = . . .

l o .

l O ,

RESPONSE TO 10 CFR 50.54 (f)

PIGURE A&B-79 SMALL SLB CASE #4, 177 FA .

2500 I

i

2000 a

I 1500 -

l

=

E E

o g 1000 -

0 0 '

500 ' '

400 500 600 O 100 200 300 Time, Sec j l

l O' Revision 2 l

4/80 l .

h I

i l

l l

l l

O RESPONSE TO 10 CPR 50.54(f)

SMALL SLB CASE #4, 177 FA I gIGURE A&B-80 i

g. l 600 I I

550 i

I

~

= 500 -

O  ! "

U

" HOT LEG '

450 -

I COLD LEG 400 -

i i i i 350 t 300 400 500 600 .

0 100 200 Time, See 1

Revision 2 4/80 4*6

O I

RESPONSE TO 10 CFR 50.54 (f) i FIGURE A&B-81 SMALL SL3 CASE #4,177 FA i

j l

22.14  :

9 E 16.60 -

I E

.'.e' E 11.07 -

O i U

5.53 -

= i i 0

400 500 600 100 200 300 0

Time, Sec l

'O Revision 2 4/80

-,--.-,.-a. . . _ . . , . , --

n ,

l RESPONSE TO 10 CFR 50.54(f) i

' FIGURE A&B-82 SMALL SLB CASE #4 177 FA 1000 I

800 -

a 3 600 -

b i

3 g 400 -

I l

l E m

=

w 200 -

0 300 400 500 600 O 100 200 Time, Sec O<

Revision 2 4/80

m_ m _ _,. w .a O .

O RESPONSE TO 10 CFR 50.54(f) h FIGURE ALB-83 SMALL SLB CASE #4, 177 FA i

l 1200 t i

r 1000 -

N w9 800 -

t 5

5 600 -

O i 04 D

5 400 - I

=

LD ,

M 200 -

1

, i e i ,

0 i 400 500 600 l 100 200 300 l 0 ,

j Time, Sec

! \

O Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-84 SMALL SLB CASE #4,177 FA O

FULL 50 -

STEAM GENERATOR B 40 -

_- i a

g 30 -

e O u, 20 -

10 - l

\

STEAM GENERATOR A I

e 0 i i i 100 200 300 0

Time, Sec Os *

- Revision 2 4/80 y - .. y _ ,, - _

, , __7, _ - _ , _ _ . , . . - . _

RESPONSE TO 10 CFR 50.54(f) ,

FIGURE M D-85 SMALL SLB CASE v4, 177 FA l l

O i i

l l

200 - .

i I

r l

=

l M 150 -

- i

~

a>

E 1 o I E ,

l

1. 100 -

CANDY CANE l

50 -

I HOT LEG l

\ SG UPPER USE REGION i

1

• e

' t ,

0 400  !

0 100 200 300 Time, Sec

~

O Revision 2  !

4/80 l

RESPONSE TO 10 CFR 50.54(f)

PIGURE ALB-86 SilALL SLB CASE #4, 177 FA l

400 -

i so E 300 -

l m,

.7 E

E CANDY CANE

> 200 -

SG UPPER TUBE \

REGION 100 - HOT LEG I

j 4 t

o 200 300 0 100 Time, Sec O

Revision 2 4/80

i i

t i

t RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-87 SNALL SLB CASE #5, 177 FA t

120 t

100 L

• 80 - l I

E D

R.

- I 60 ,

\

E  ;

l 40 20 0

100 200 300 400 0

t 1

' Time, Sec

• l .

6 Revision 2 4/80

-, ,--r- ----, , - . - - - - - - - , c- e- , w

O kk RESPONSE TO 10 CFR 50.54(f)

FIGURC A&& 88 SilALL SLB CASE #5, 177 FA .

2500 t

. 2000 E

3 1500 ,

5 o

O f

E u

1000 -

l t

500

' ' i 400 500 600 100 200 300 0

Time, Sec l

l l

l O .

Revision 2 4/80

RESPOtJSE TO 10 CFR 50.54(f)

FIGURE A& B- 89 SMALL SLB CASE v5, 177 FA 620  ;

580 540 -

m

• HOT LEG f

500 -

2 O -

h

• 420 -

380 COLO LEG 340 -

@ i , , e >

300 400 500 600 0 100 200 Time, Sec O

Revision 2 4/80 1

i i

l I

O i RESPONSE TO 10 CFR 50.54(f) f FIGURE A&B-90 SilALL SLB CASE 85,177 FA 22.14 9

16.60 -

~

. I J l w

3 11.07 -

I

)

E l

E 5.53 -

0 300 400 500 600 O 100 200 Time, Sec O,

Revision 2 4/80 l -

l l

l l

l l

i l

l l

RESPONSE TO 10 CFR 50.54(f) 1 i

FIGURE A&B-91 SMALL SLB CASE #5,177 FA l i

l 1000 i l

l l

i 800 h i i

3 '

i 600 -

O 5

i 400 -

M ,

200 -

x '

0  !

400 500 600 O 100 200 300 l

Time, Sec Revision 2 l 4/80 I

O RESPO!1SE TO 10 CPR 50.54(f)

FIGURE A&B-92 SMALL SLB CASE #5, 177 FA 1000 800 5

2 600 -

O r 3

5 U -

a 400 -

=

V1 200

! I i t t t Q

300 400 500 600 0 100 200 Time, Sec t

0 O. .

9 Revision 2 4/80

\

-- ~ --

RESPONSE TO 10 CFR 50.54 ( f )-

FIGURE A&B-9 3 SMALL SLB CASE #5, 177 FA  :

55 NOTE: SG B FILLS e 369 SEC.

50 45 - ,

i 40 -

j f 35 -

i E SG B 5- 30 -

E a

3 25 -

i 20 NOTE: SG A EMPTIES COMPLETELY I

AT 311 SEC J l

15 3G A 10 I

5 -

0 400 200 300 0 100 Time, Sec Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-94 SNALL SLB CASE #5, 177 FA ,

280 "0"

NOTE: THE SG UPPER TUBE REGION REMAINS CONSTANT AT l

240 -

l CANDY CANE l l

200 -

= ,

E I a., - \

l J 160 -

i E

O i

o

• i E >

e -

120 f i

80 -

I HOT LEG 40 - l e

0 l

400 500 l 100 200 300 O

Time, Sec O

Revision 2 4/80 i

O RESPONSE TO 10 CPR 50.54(f)

FIGURE A&B-95 SMALL SLB CASE #5,177 FA NO YOIDING IN STEAM GENERATOR B O

l O

Revision 2 4/80 I

l . . _ _ _ . _ _

f i

I

' RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-96 SMALL SLB CASE #6, 177 FA

. \ s.

120 100 I-80 e

E O

60 --

I o 40 20 l 0 300 400 500 600 0 100 200 Time, Sec i

Revision 2 4/80 I

r

O i i

I

. RESPONSE TO 10 CFR 50.54(f) i FIGURE A&B-97 SMALL SLB CASE #6,177 FA 2500 I  :

2000 P 2

"s' 5 i

' 5 1500 ,

0 E i

~ >

1000 -

o 5 N '

x 500 -

N I

l l

0 200 300 400 500 600 i 0 100 i Time, Sec l

Revision 2 4/80

RES PONS E 'IO 10 CFR 50. 5 4 ( f )

FIGURE A&B-98 SilALL SLB CASE #6, 177 FA i

600 -

l l

550 i

HOT LEG i 9

f I

E  :

~

- 500 -

O -

i  !

l U.c 450 - f COLD LEG  !

400 -

350 400 500 600 O 100 200 300 Time, Sec Revision 2  ;

4/80 i

+

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-99 SMALL SLB CASE #6, 177 FA 22,14 I

16.60 -

~

~

11.07 -

2- ,

3 i

M a- 5.53 -

i I 1 ]

f 400 500 600 0 100 200 300 Time, Sec i

1 e

4 Revision 2 4/80

_ u - -- -- ----

i l

l O

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-100 SMALL SLB CASE #6,177 FA i

1000 i

t 800 .

E g 800 E

p 400 -

w a

- 200 -

M X'

0 400 500 600 100 200 300 O

Time, Sec e

'O .

Revision 2 4/80

l i

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-101 SilALL SLB CASE #6. 177 FA 1400 ,

i 1200 - ,

V9 l 1000 -

a E

800 -

O W1 l

5 600 -

=

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! Revision 2 4/80 '

.24 _ _ m -

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-102 SMALL SLB CASE #6,177 FA FULL 50 ._

40 -

' - STEAW GENERATOR B E

a w

3 30 -

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20 -

l l

10 -

i l

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I

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150 200 250 300 l 0 50 100 Time, Sec l 1

O Revision 2 4/80 i

_ ._ . - , , _.,y _..-,,, -

I

[

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-103 SMALL SLB CASE #6,177 FA ,

I

400 l

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i

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< 5 O

j 200 O

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l O

400 500 600 100 200 300 0 .

Time, Sec  ;

j O \ .

Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-104 SilALL SLB CASE #6,177 FA 50 NOTE: NO V010 LNG IN HOT LEG l

i i

40 -

SG UPPER CANDY TUBE CANE

" REGION l l

a

(*3 30 -

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20 -

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j 10 -

I 100 150 0 50

. Time, See l

O.

Revision 2 4/80

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

RESPONSE 'IO 10 CFR 50,54 ( f) l l

l FIGURE A&B-105 SLB, CASE 1, 177 FA l2 2200 ,

1 i

l 2100 -

l l

l 2000 - - l 1900 - -

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1600 -

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i i

1 i

RESPONSE 'IO 10 CFR 50.54 ( f )

FIGURE A&B-106 SLB, CASE 1, 177 FA l2 i l

l

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l

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480 470 460 ,

g g g- i 1 I I~ I I 450 60 70 80 90 100 10 20 30 40 50 0

Time, see Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-107 SLB, CASE 1,177 FA l2 l O

18 -

16 14 -

12 -

10 _

5 8 -  ;

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\

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RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-110 SLB, CASE 1, 177 FA 600 -,

l 550 - -

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400 600 100 200 300 400 500 O

Time, see Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

O FIGURE A&B-111 St.B. CASE 1, 177 FA 33.2 27.67 -

r 22.14 -

I 2 16.60 -

0

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l l

l 11.07 -

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l

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200 300 400 500 600 0 100 Time, sec Revision 2 4/80 ,

RESPONSE TO 10 CFR 50.54(f)

) i i

FIGURE A&B-112 SLB, CASE 1, 177 FA i

1200 1000 -

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0

~

a.

u I

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a a 400 m .

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400 500 600 100 200 300 0

Time, see Revision 2 4/80

RESPONSE 'IO 10 CFR 50.54 ( f)

O FIGURE A&B-ll3 SLB, CASE 1, 177 FA 800 -

700 --

E E

600 - -

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ea -

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4/80 l i

i i

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-114 SLB, CASE 1, 177 FA 1

O FULL 50 J r

4 40 i

0

~

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a 30 -

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- Time, see Revision 2 4/80

i RESPONSE TO 10 CFR 50.54(f) '

i O  ;

t FIGURE A&B-ll5 SLB, CASE 1, 177 FA 350 300 -

ll h 1 11 li ll lg i CANDY CANE "A"

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l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-ll6 stb, CASE 1, 177 FA 80 l

l 70 p il l

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RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-119 SLB, CASE 2, 177 FA O ,

800 -

l i

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• 400 O O 100 200 Time, sec 300 400 500 600 l

Revision 2 i

4/80 I

FIGURE A&B-120 SLB. CASE 2, 177 FA 22.67 22.14 -

w E

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RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-123 SLB, CASE 2, 177 FA l

j 50 _  ;

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g  ;

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a

+

0;  !

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i NOTE: S.G. "A" BOILS.0P.Y WITHIN l

20.0 SECONDS 10 O _ _ _

____7__.._.

300 400 0 100 200 Time, Sec Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) i O

i I

FIGURE A&B-124 SLB. CASE 2, 177 FA 250 LOOP "A"

,s

. I N

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l

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i 0 25 50 75 100 j

\

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Revision 2 4/80

Ii I

t RESPONSE TO 10 CFR 50.54(f) l O FIGURE A&B-125 SLB, CASE 2, 177 FA l l l .

40  ;

t LOOP "B" i

l N .

f i

30 -

CANDY CANE  :

I W

Oi-1

1. 20 -

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i i

l l

_y 10 15 0 5 -

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RESPONSE TO 10 CFR 50.54(f) t i

i i

FIGURE A&B-126 SLB. CASE 3, 177 FA t

!.2  ;

I 1.0 -.

0.8 O

E  :

a. i

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e m

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I 0.4 -

I l

i 0.2 -

400 500 600 0 100 200 300 Time, see O

Revision 2 4/80

RESPO11SE 'IO 10 CFR 50.54 ( f) t i

SLB, CASE 3, 177 FA ,

FIGURE A&B-127 a

t 2000 -

f 6

1800 -- i en

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= I L

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l 600 800 100 200 300 400 500 l 0 l Time, sec l Revision 2 l l

4/80 l 1 I

e

. . _ . --. . .-..I

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-128 SLB. CASE #3, 177 FA 600 550 -

I

)

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500 -

it /

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j

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, 0 100 200 300 400 500 600 Time, Sec ig Revision 2 4/80 -

i RESPONSE TO 10 CFR 50.54(f)  :

i I

i FIGURE A&B-129 SLB, CASE 3, 177 FA  ;

33.20 '  :- >

27.67 -

22.14 -

O ,

t 16.60 -

E E

a.

11.07 -

5.53 -

i e  ! , ,

0 100 200 300 400 500 600 0

Time, Sec l

r 182 Revision 2 4/80

RESPONSE 'IO 10 CFR 50. 54 ( f ) ,

t i

i l

FIGURE A&B-130 SLB, CASE 3, 177 FA 1200 - l I

l 1000 -

l

_f43 M

3 0

=

800 '-

i w

L l

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l l

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400 500 600 100 200 300 0

Time, sec l l

i Revision 2 4/80 1

RESPONSE 'IO 10 CFR 50.54 ( f)

I t

FIGURE A&B-131 SLB, CASE 3, 177 FA i I

1200  :

w-r 1000 -' i i

I L

I m E 7

E. 1 800 6

[

, a N

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i 0 i 400 500 600 j 100 200 300 O )

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I l

l i

l

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r 4/80 l l

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-132 SLB, CASE 3, 177 FA ,

O FULL

! 50 -.  !

i i I

k 40 -

~

U 3

= .

a 30 -

% 1 O$

5 ,

20 4

NOTE: SG "A" BOILS ORY WITHIN 10

- 20.0 SECONOS l

i l

t f

l r

O a D 50 100 150 200 Time, sec 250 300 350 Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

O SLB CASE 3, 177 FA FIGURE A&B-133 300 - p A

I \

250 - l g l \

l i

\ CAN0Y CANE "A" m I 1 r 200 -

\

0: E E 150 -l g

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\

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l 50 l

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's s 0 < ,

/ ,

\x a rs _

100 150 0 50 Time, see O

Revision 2 4/80

i RESPONSE TO 10 CFR 50.54(f) j l

FIGURE A&B-134 SLB. CASE 3, 171 FA j O

l

- i 50 --

i l 40 - '

m,

~

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t2 30 -

L t

=

[

i 20 - j t

l l

10 l

i O

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I t

RESPONSE TO 10 CFR 50,54(f)  !

t i

l t

i i

FIGURE A&B-135 SLB. CASE 4, 177 FA ,

i 1.0 1 i I

0.8 - -  ;

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t==

as

= 0.6 - -

n.

N 0* '

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r 0.4 - -

i k

l 0.2 - -

s 1 o n

w - *ue-

• 100 200 300 400 500 O t

' Time, see Revision 2 4/80

RESPONSE '10 10 CPR 50.54 ( f)

O ,

FIGURE A&B-135 SLE, CASE 4, 177 FA 2190 1800 - -

l i

1600 -- t i

M  ;

E. t i 1400 -

l2 i

0

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i OEo 1200 -

O u  ;

1000 1

l 800 i

~ ~' - ~~ 1 i i 1 T ~ ~~ - ~ ~ i 500 600 300 400 0 100 200 Time, sec  ;

O Revision 2 4/80 t

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-137 SLB, CASE 4, 177 FA i

1 1

600 - l l

l I

525 y ,

\  :

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s O  :-

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m:

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l

\

s N

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l 375 -

s s .

f

's N i O 500 .

100 200 300 400 500 600 Time, see Revision 2 4/80

RESPONSE TO 10 CFR 50,54(f) l I

FIGURE A&B-138 SLB, CASE 4, 177 FA l WE 33.20 L

t i

27.67 i i

I 22.14 -

~

ene amm.

l M

> I N

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N b

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t )

1

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L

.400 500 600 l 100 200 300 \

O <

l Time, sec I

i I

Revision 2 4/80 l l

1 1

RESPONSE TO 10 CFR 50.54(f) f FIGURE A&B-139 SLB, CASE 4, 177 FA 1000 O

800' - i e

a  :

E 600 i

i E 400 --

t a I

?

u, a 200  ;

0 . - ,

0 100 200 300 400 500 600 Time, Sec ,

l Revision 2 4/80

--- ~ , , .-- , ----n--- -

---

w----

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-140 SLB, CASE 4, 177 FA 1000 i

. 800 .-

a e3 '

a

{ 600 -

O! a

,% 400 -

5 m

cm m

200 _

I I I I '

, 0

( 0 100 200 300 400 500 600 l

Time, Sec O Revision 2 4/80 l

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-141 I;B, CASE 4. 177 FA FULL  !

50 --

i i

~

l l  !

40 i

U  !

• i l 3  ;

l 3 30 -

6 l g

= STEAN GENERATOR "B" i

i 20 t

NOTE: SG "A" BOILS ORY WITHIN f f'

20.0 SECONDS 10 I

O O 50 100 150 200 250 300 350 400 ,

Time, Sec Revision 2 4/80

RESPONSE TO 10 CPR 50.54(f)

FIGURE A&B-142 SLB, CASE 4, 177 FA i

,\

300 -

\

I i

1 270 l 1

l 240 1

CANDY CANE 210 ,

I I

% 180 -

l LOOP "A" i 5 '

E 150 . . 1 E i a I i E i 120 -

j t

l f% >

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. REGION i 30 , -

\

l 1

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l ,

O ,

0 50 100- 150 200 250 300 350 Time, Sec Revision 2 4/80 .

RESPONSE 'IO 10 CFR 50.54 ( f) ,

FIGURE A&B-143 SLB. CASE 4. 177 FA 63- .

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l  ;

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I

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, - - , - ,-...-.,r, - -

RESPONSE TO 10 CFR 50.54(f) ,

1 t

i l

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FIGURE A&B-144 SLB, CASE 5,177 FA l e

eM 1.2  !,

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t i i I '

0.0 200 300 400 500 600 O 100 ,

Time, Sec l Revision 2 4/80

- - . _ .. . -_. =- _

M i

RESPONSE TO 10 CFR 50.54(f)

O FIGURE A&B-145 SLB. CASE 5, 177 FA j 2200 i

1 2000 -

r 1800  :

.su

a.  !

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l 1

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= 1400 - l a ,

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l 1200 -

1 I

i

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1000 -

i l

I I I I I l 500 800 300 400 500 i 0 100 200 I Time, Sec l

I I

Revision 2 j 4/80 l

l l

i

t RESPONSE TO 10 CFR 50.54(f) i O SLB, CASE 5, 177 FA  ;

FIGURE A&B-146 600_.

- -- [

f a

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550 -

l l

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N  :

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400 -

0 100 200 300 400 500 600 -

Time, Sec Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f)

O  !

i p

FIGURE A&B-147 SLB CASE 5, 177 FA 27.67  ;

i 22.14 i .

.e * .

i -

l  % 16.60 .  ;

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2 E

l 3 11.07 -

E  !

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5.53 -

6 o

9 O u 4 v  : i  ;

O 100 200 300 400, 500 600 ,

Time, Sec P

a Revision 2 4/80 .

l l

RESPONSE 'IO 10 CFR 50.54 ( f )

l i

i FIGURE A&B-148 St.B CASE 5,177 FA

{

1200 - l t

1000 Cl1

.7 E

800 5  !

r" 600

\

5 l

.n U 400 . .

l

<= i

)

i i

200 - l 1 -

0 200 300 400 500 500 0 100 1 Time, Sec l Revision 2 4/80 l

i i

RESPONSE TO 10 CFR 50.54(f) i i

FIGURE A&B-149 SLB CASE 5, 177 FA <

1400 I

i l

1200 -

n h'

R 1000 - ,

g

- i:

[:

y .

=

800 -

C i:

t:

E E

g .,

. 1 p 600 - ll m c I

t t

b 400 - t 1

• l

. I 200 I I I ' ' I O 100 200 300 400 500 600 l Time, see - - -

l l

l Revision 2  !

4/80 0 -n .

RESPONSE 70 10 CFR 50.54(f) l SLB CASE 5, 177 FA  :

FIGURE A&B-150 -

O FULL i

50 STEAN GENERATOR 8 i

I l

40 -

l i

i f

E N 30 [

• 1 5 O 3 i

i r

5 w .

t 5

NOTE: SG A BOILS ORY WITHIN U 20

- I 22.0 SECONOS i

i 10 - l l

I I i i 1

0 100 150 200 250 300 0 50 Time, Sec l Revision 2 I 4/80 i l

l  !

RESPONSE TO 10 CFR 50.54(f)  ;

i FIGURE A&B-151 SL,8 CASE 5, 177 FA O 264 -

252 240 -

228 -

lI (

I

- I 216 b

I ,

204 -

192 - /

l i 180 -

168 - l i "g 156 -

i CANDY CANE "A" 140 l O ~*[o

~

132 -

I i

l t

> 120 -

l l

I 108 -

l 96 - t '

1 L

I 84 -

t UPPER TUBE REGION, SG A l 72

• l 60 (l .

I I

48 . . ,I  ;

fi i i* HOT LEG "A" 36 . ,

e 24 .

.J

- l 12 I I '

L '

0 400 100 200 300 O

Time, sec

- Revision 2 ]

4/80 l l

L, RESPONSE 'IO 10 CFR 50.54( f )  !

i I

l c

l l

i FIGURE A&B-152 SLB, CASE 5, 177 FA .

[

f 1

l i

NO VOIDING IN LOOP B ,

l 1

i i

i  !

i t

l f

I l

l I

i 1

Revision 2  ;

4/80  !

1 I

I l

1 1

t RESPONSE TO 10 CFR 50.54(f) t FIGURE A&B-153 SLB CASE 6, 177 FA 1.0 -

0.8 -

{

l 0.6 - -  !

w -

e I a t Eb.

l

0.4 - - .

I l

i 0.2 , _.

i l __

0.0 0 100 200 300 400 500 600 -

Time, Sec l

l l

l l

i Revision 2 i 4/80 i l

l i

l t

RESPONSE TO 10 CFR 50.54(f)  !

t O  ;

l I

I FIGURE A&B-154 SLB CASE 6, 177 FA

' % 2190 t 2000 l

l 1800 J

b i

se 1600 -

= i G

t i

.3 E 1400 -. j I

M I

Cl3

• I O

1200 --

l l

l

\

1000 -

l l

I I l 1 800 300 400 500 600 0 100 200 Time, Sec Rsvision 2 l

4/80 l

RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-155 SLB CASE 6, 177 FA 600  ;

i 550 r-l l

i HOT LEG j l

t t l 6 '

[

O i soo 4f

's \  !

a A b \ l 8

\ COLO LEG

\ i I

\ l N

N 450 -

N '

N '

N N

N ,

N i N l

% == =  !

l l l I I I 400 o too 200 300 4ao Sao 600 (

l O

u . s.c  :

Revision 2 l 4/80 )

1 i

l

i l

t i

RESPONSE TO 10 CFR 50.54(f)  :

l l

t 6

i FIGURE A&B-156 SLB CASE 6, 177 FA 22.14  :

t

= 16.60 . -

_- l E

O r 2 11.07 _ _

k 2  !

L 5.53 ,_ _ ,

I l

1 1

l 0

ei .

0 100 200 300 400 500 600 l

Time, Sec i

I l

l Revision 2 f 4/80 i

l

t RESPONSE TO 10 CFR 50.54(f) l i

I i

FIGURE A&B-157 SLB CASE 6, 177 FA r

I 1000 i r

E. 800 .

E a i 2 >

Z  !

, 600 -

2 ,

5 a

h i 50

= r 5 400 --

E ,

m t

=  !

200 ,

i I  !

t 1 I '

, i 0

300 400 500 600 O 100 2La Time. Sec i Revision 2 4/80  ;

i I

i I

l RESPONSE TO 10 CFR 50.54(f) i i

L c

i t

i FIGUR3 A&B-158 SLB CASE 6, 177 FA  ;

P l

. 1000 -

=- ,

=

nn 800

.". I W

== t

= >

m l e.

= 600 -- 1

= ,

e '

O t

o. i

- c

.=

s 400 -

a i i

L I I I I I  !

, 200 i 0 100 200 300 400 500 600 i Time, Sec ,

t' t

l '

I l

i i

l Revision 2 4/80 f i

RESPONSE 'IO 10 CFR 50.54 ( f )

t i

l FIGURE A&B-159 SLB CASE 6, 177 FA 1 i

FULL ,

l  !

l 50 .

l 45 l

l 40 O 35 -

3 3

30 -

O

• E w

STEAN GENERATOR "B"

. 25 - '

2 I

20 -

NOTE: SG "A" BOILS ORY f WITHIN 20.0 SECONOS 15 -

t 10 - t 5 -

i l

O O 0

i 50 i

100 i

150 i

200 i

250 i

300 350 j

Time, Sec Revision 2 4/80

RESPONSE TO 10 CFR 50.54(f) i

\

l O l l

SLB CASE 6, 177 FA FIGURE A&B-160 300 t\ i.

lI

!D % LOOP "A"

\ \

250 -

g l

l l \

\ l 1

I l I CAN0Y CANE i

a, 200 -

g l '

i I  :

.e l

\

= 150 / \  ;

i 5 I

I ,

> l I

I  !

5 l 100 -l l I I l

\ SG UPPER l

l HOT LEG ' TUBE REGION f 50 e g

I f N I '

I

" e s% ,

9 0 (  ! l 5- '

I O 50 100 Time, Sec Revision 2 C4 4/80

RESPONSE 'IO 10 CFR 50.54 ( f)

FIGURE A&B-161 SLB CASE 6,177 FA l

50 l i I

40 CANDY CANE "B" I

( "r 30 O .

3 E

2 20 -

l l

l I

i l

10 l

O O 0

i i 10 15 Time, Se:

Revision 2 4/80 1

l l

i-RESPONSE TO 10 CFR 50.54(f) i FIGURE A&B-162 SLB CASE 7, 177 FA ,

1.2 1.0 i 0.8 --

t b" 0.6 - -  ;

t i

0.4 _ _

t i

0.2 - -

h O. 0 I l l l 0 100 200 300 400 500 ,

Time, Sec Revision 2 4/80

i i

l l

RESPONSE 'IO 10 CFR 50.54 ( f) f f

l FIGURS A&B-163 SLB CASE 7, 177 FA  ;

3000 l

[

i i

=

2500 -

=

a -

0 i l a- 2000 .._

l l

1 E .

f E  ;

O 1500 - -

[

1000 -

1 I I I I  :

500 O 100 200 300 400 500 600 i

Time, Sec l

\

l I

Revision 2 4/80 l

l l

r

RESPONSE TO 10 CFR 50.54(f) l FIGURE A&B-164 SLB CASE 7, 177 FA 600 570 l

i i

I l

520 u -

i l(

1l s y \

\

O i Y \

3 \

h 470 - \

~

\

y g HOT LEG

\ l

\ l

\ l

\ i

\ l 420 - \

COLD L G \ l

\ l

\ l N l N i N l N '

N

. O l l 1 1 1 1 I I I I )

i l

400 500 600 0 100 200 300 l Time, Sec Revision 2 4/80 I l

RESPONSE TO 10 CPR 50.54(f)  ;

i FIGURE A&B-165 SLB CASE 7,177 FA .

33.20 i

r 27.67 _

22.14 -

p 0 ,

= 16.60 - -

~

m 5 -

as.

11.07 - -

i 1

5.53 - -

l i

I

' I l O

0 100 200 300 400 500 600 Time, Sec i I

i i

Revision 2 4/80 i s

i i

RESPONSE TO 10 CFR 50.54(f) i l

i i

l FIGURE A&B-166 SLB CASE 7, 177 FA l l

1200  ;

i i

r i

1000 -

s

=  !

a 800 -  :

5  :

0 t

! E i j

~

600 .- l I

lii 1o -

400 --  ;

" 1 2

= ,

200  ;

1 l

l L I I I I i  !

O 0 100 200 300 400 500 600 l

Time, Sec j e

f

= i

! Revision 2 4/80 l

l

RESPONSE TO 10 CFR 50.54(f)

O I

l

~

l l

SLB CASE 7, 177 FA

~

FIGURE A&B-167

?

1200 l

a 1000 -

o". [

m M f a

l -

2 800  !

O$

E ,

8 -

a ,

= 600 - -

E' S

i 400 -

l 200 -

I I I O I 200 300 400 500 600 O 100 Time, Sec l

Revision 2 4/80

RESPONSE 'IO 10 CFR 50.54 ( f)

FIGURE A&B-108 SLB CASE 7, 177 FA l FULL 50 -

b 40 -

O STEAM GENERATOR "B" 3 i

, - 30 -

i 2"

~

a  !

m ,

20 -

1 l NOTE: SG "A" BOILS ORY WITHIN 20.0 SECONOS l

10 -

l I  !

O 0 100 200 300 Time, Sec Revision 2 4/80

l RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-169 SLB CASE 7, 177 FA

~

O  !  !

8  ;

11 250 -4{ f 11 11 11 11 Il 200 p -S j II / -

11 /

/ A l} / k/\

ll l

V \  ;

s il / \

- I l

/ \

g j 150 j

O a Q',

i i 2 I CANDY CANE "A" ,

lg, /j ,

I I /

100 L. I I '

I l I l\ I /

/ HOT LEG "A" l

I I

J

/

50 _l I

h?

d(

K \

i l ., i UPPER SG TUBE REGION l'

l Y\ i O .

0 I .

100 200 300 400 500 600 i

Time, Sec Revision 2 4/80 l

d RESPONSE 'io 10 CFR 50. 54 ( f )

FIGURE A&B-170 SLB CASE 7, 177 FA I

~~

l f

THERE IS NO VOIDING IN LOOP "B" l

l l

F v

b Revision 2 4/60

f RESPONSE TO 10 CFR 50.54(f) l I

l f

I i

i FIGURE A&B-171 SLB CASE 8, 177 FA  ;

1.2 ,

I I

1.0 - 3 l

f r

i 0.8 . -

O n E

e i i

d.6 - -

0= \

• I P

0.4 - -  ;

I 0.2  ;

j .--

I  !  !  !

0.0 100 200 ~300 400 500 O

Time, Sec i

Revision 2 4/80

k RESPONSE 'IO 10 CPR 50. 54 ( f )

i t

I l

FIGURE A&B-172 SLB CASE 8, 177 FA l

3000  ;

t 2500 -

t so o.

i -

I O5 2000 '

E '

i 5 1500 -j-e 0

1000 -

i

~

500 1 I I I O 100 200 300 400 500 600 Time, Sec O

U Revision 2 4/80

RESPONSE TO 10 CFR 50.54{f) i I

FIGURE A&B-173 SLB CASE 8. 177 FA j 600 O i l

\

l f

I

, i i

[

550 -

l r

[

l HOT LEG l f l 3 l i \ l 500 i N .

- i / s  !

i/ N O [E E!

' \

\

l i

i COLO LEG \

\  !

\  !

450 -

\  !

\  !

\ /T

! \ / l  !

N w/ l l r

400 - \

s N,

i i i iA O

O 100 200 .300 400 500 600 Time, Sec Revision 2 ,

4/80

RESPONSE TO 10 CFR 50.54(f)

O  !

i l

i i

i SLB CASE 8, 177 FA  :

FIGURE A&B-174 i

41.50 -

" 26.67 i

~

a s

- 1 t

13.84 -

1

~  !

a.

L t I _

l l l g

0 100 200 300 400 500 600 l 1

Time, Sec ,

i l

l 1

i Revision 2 O~ 4/80 l

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

- y - - -

--e ~ -

t i

RESPONSE 'IO 10 CFR 50.54 ( f) l l

l i

FIGURE A&B-175 SLB CASE 8, 177 FA l

1200  ;

l i

."m 1000 -

l

= ,

l E

M a- 800 -

% i

=

=

h  !

Ea 600 . -  :

m 2  ;

i I

c.o 400 -

l 1

1 1

200 -

l l

! 1 l

O - -

l 0 100 200 300 400 500 600 Time, Sec l

Revision 2 4/80 l

l l

RESPONSE TO 10 CFR 50.54(f) i O i FIGURE A&B-176 SLB CASE 8, 177 FA 1200 -

1000

. t E.  !

800 -

=

i N

=

i f

OE

• 600 .

i t

l

% l E 400 -

l 5

=

i

<n _

200 -  !

(

0 t i I ' I '

O 100 200 300 400 500 800 '

Time, Sec '

l l Revision 2 4/80

RESPONSE 'IO 10 CFR 50. 54 ( f ) i FIGURE A&B-177 SLB CASE 8, 177FA f

O FULL t

50 _

i i

l 45 -

t STEAM GENERATOR B  !

40 -

i C -

~

35 .

h

( 0 ,

l 3 30 _  !

O c%

o 3 25 -

l 3

65 NOTE: SG "A" BOILS ORY  !

u WITHIN 20.0 SECONOS l 20 -

e 15 -

l i

10 -

1 5

1 I

t i I 0

l 0 100 200 300 400 2 P

Time, Sec Revision 2 4/80

r RESPONSE TO 10 CFR 50.54(f)

FIGURE A&B-178 SLB CASE 8,177 FA j i

LOOP A g ,,

l\ /

400 _

jV HOT LEG

--- CANDY CANE g

I{ f

--- UPPER SG TUBING REGION 'l [f {

Jl  !

I l -l \ \\ll -

j'- l s f\ \li-f

/  :

i t

- I - 1 j

l .; .

Oa 8

I

'f f i

2 i 200 Ii i

ld  !

\

I i I .

i eI r i

/ '

100 -

/

1 i

4 /_-

og y -

.50 b I

\\ i i l

-0 n! Al I I I I O 100 200 300 400 500 800 Time, Sec Revision 2 l 4/80 l

RESPONSE TO 10 CFR 50.54(f)

O SLB CASE 8, 177 FA t FIGURE A&B-179 400 t

t l l '

l  :

300 -

l l1  !

i

• ?.,

ll CAN0Y CANE "B" f

l 1

O 200 -l1  !

f UPPER SG TUBE

, l REGION SG "B" j i l

l'

( l ,/ \

8

/ h g

l I

[

'/ k 100 -

\

I

{\ l \

50

\I t I

I HOT LEG "B"

\

g l i

\ , l\ 1 l 1

i I k

! l I i 30h O

O  ;

150 200 0 50 l

Time, Sec l

Revision 2 l i

4/80 e

. , _ . - - ..e. , - ,-.

RESPONSE TO 10 CFR 50.54(f)  ;

i APPENDIX C  ;

O Question c Provide a schedule of completion of installation of the iden'tified systems and components. l Response  ;

i Attached Table C-1 contains the construction status of the systems and components identified in Enclosure 3 of your letter  :

dated October 25, 1979, with the addition of the main feedwater system and integrated control system. Although this table includes the HPI system, DHR system, CFT system, quench tank, and .;

RCS piping , as you requested in your letter, Consumers Power

,- Company can find no relation or interaction between these items  ;

and the issue of OTSG sensitivity. Therefore, hardware and t 4

procedural changes for these systems are not addressed. ,

The table consists of a matrix for both units showing, for each l system and component, the percent complete by quantities and the  ;

estimated completion date. The percent complete figure for small pipe (2 inches and under), large pipe, major equipment (panels, switchgear, pumps, valves, heat exchangers, and tanks), i electrical equipment (cable, conduit, and trays,), and  ;

instrumentation (transmitters, controllers, tubing, and l

! indicators) is based on actual quantities installed as of l

- November 1, 1979, and compared to total quantities estimated in  !

February 1978. Some increases in total quantity are expected in  !

small pipe and-electrical quantities as a result of the present schedule review. Completion date is defined as the date when construction is '100% complete for all the equipment and 4 components within the scope of the system or component and when j the system or. component is ready for turnover for testing, ,

The completion dates reflected in the table are for the Consumers

- Power Company existing test schedule, which reflects fuel load dates of June 1981 and November 1981 for Units 2 and 1, respectively. They are currently under review and may be revised  !

4 in early 1980 when a revised project schedule will be_ issued.

Construction activities for large pipe and installation of major equipment for the listed. systems / components are currently "on schedule" to meet the above target . fuel load dates. However, i

i increased electrical quantities for raceway, small pipe, wire, and_ cable, plus constraints due to space limitations, are l . impacting small pipe and electrical construction activities for-the listed systems / components.

! Control room layout schedule activities are. essentially complete, '

with 72 of 90 panels already installed.. Panel ~ board-layout and control' panel . arrangement have been finalized, the indicating. or O readout devices and control-device have been fixed,:and most of the devices are already installed in the boards and wired _by the C-1 11/79

. - _ = _ . . - _ . _ - _ . . - - . - - -. . - . . . . . _ . - - - _ _ -

RESPONSE TO 10 CFR 50.54(f) i APPENDIX C panel fabricator. The remainder of the devices are currently being installed in the field. HVAC installation work is l

l complete, and electrical connections and terminations work is l

being performed on the installed panels. j t

i i

t i 1

! }

i f

I i

I

g. i i

J r

i i

i  !

. 4 t i

! i l

! l 1

I l

1 1

5 C-2 11/79 l

TABLE Cel CONSTRUCTION STATUS MIDLAND PLANT UNITS 1 AND 2

~,

Unit 2 % Complete (Installed) Unit I t Complete (Installed)

Completion Small"' La rg e'" Completion Smallm Large m Equip Elec* I ns t r Date l1 Equip Eleem I ns t r Date Pipem. Pipe System / Compt.nent Pipe

• Pipe 100 68 6 01/10/80 NA 65 100 68 6 04/24/80 HPI. NA 75 100 31 0 0/28/80 <10 75 100 26 0 08/08/80

.Afw <10 80 100 57 4 12/21/79 20 55 100 55 1 04/15/80 DRR 50 65 100 43 17 01/25/80 NA 90 100 43 0 06/06/80 CPT NA 95 RCS pressure 20 33 1 05/05/80 0 40 10 28 2 07/23/80 control 10 50 Makeup and 55 100 27 3 04/11/80 h le tdown 5 65 100 39 3 12/14/79 5

.h Steam generator 03/06/80 60 75 75

• 33 36 08/21/80 g pressure control 60 75 75* 50 36 NA NA 05/02/80 NA NA 90 NA NA 08/01/80 $

Steam generator NA NA 90 NA NA 05/05/80 NA 0 90 NA NA 07/23/80 o Pressurizer

• NA 50 90 NA NA 05/05/80 NA 10 90 NA NA 07/23/80 h Ouench tank' NA 50 90 O

control room By system NA NA 75 20 75 By system NA NA 75 20 75 layout RCS piping NA 90 NA NA 9 06/12/80 NA 75 NA NA 0 08/16/80 }

NA 02/06/81 NA NA 100 3'8 NA 03/18/81

-ICS* NA NA '100 5*

17 '02/25/80 55 92 100 36 2 06/04/80 70 95 100 47 MFW

• Pipe percent represents the summation of pipe spools, hangers, restraints, and field welds, wg mSmall pipe is primarily vents and drains.

.D

• Polling cable was weighted as 75% of ef fort and terminating cable was weighted as 25% of ef fort.

w- ,lubing, mounting, and electrical connections were weighted equally.

• S.

• Remaining 25% is installation of MAD valves, which are onsite.

o

" As scoped, the ICS includes only the control cabinet. Also, field transmitters are included in each respective system.

• Includes surge line.

" *About 50% of ICS system cables are pulled. Few are terminated. Cables in this system are mostly multi-(2) conductor.

This number of connection greatly influences total percent complete.

RESPONSE TO 10 CFR 50.54(f)

APPENDIX D

/~'\  ;

\~ 2 Question d Identify the feasibility of halting installation of these systems and components as compared to the feasibility of completing installation and ' then ef fecting significant changes in these systems and components.

Response

Consumers Power Company's review of the feasibility of halting

~

installation of these systems and components, as compared to completely installating and then ef fecting significant changes, may be divided into two primary parts: 1) review of overall plant construction status, and 2) assessment of schedule impact of possible changes.

With regard to overall plant construction status, the following information is provided with the percent complete figures based on quantities as discur. sed in Appendix C.

l a. Civil - Civil construction work is 82% complete; concrete ,

work is 94% complate. Primary and secondary shieJd wall construction is complete in both containment buildings. The i

f civil construction opening for the Unit 1 containment

(~'s building was closed in September 1979. The opening for the

( ,) Unit 2 containment building was closed earlier in 1979.

Within both containment buildings and the auxiliary building, essentially all temporary civil construction openings have been closed. Post-tensioning of Unit 2 is 67% complete.

Civil construction is essentially complete on the Midland plant.

b. Mechanical - All major mechanical equipment and components in both containment buildings and the auxiliary building have been set. All major NSSS components have been final set and the reactor coolant pump internals and motors have also been installed. The nuclear steam supply system (NSSS) for Unit 2 has all major loop piping welded and stress relieved.

Presently, the NSSS erector is working on the reactor vessel internals fit-up and head assembly in Unit 2. The NSSS for i

i Unit 2 is 3 months ahead of Unit 1. Eighty percent of the large pipe installation of NSSS support systems is installed, and it approaches that of a completed plant. Remaining mechanical work consists of small pipe, which is'45%

j complete, and both large and small pipe hanger installation, l

which is 60% complete.

c. Electrical - All major electrical equipment and components in both containment buildings and the auxiliary building have

()

been set. Installation of electrical' bulk commodities is as follows: cable tray is 96% complete; con 6uit is 72%  ;

D-1 11/79

RESPONSE TO 10CFR 50.54(f)

APPENDIX D complete; and electrical cable is 41% complete for both f- units.

\- d. Instrumentation - Installation of this portion of the plant is 13% complete for both units.

Consumers Power Company's assessment of halting installation of systems versus continuing installation and then effecting changes indicates that halting installation does not enhance the ability to modify the systems. Major component and large pipe installation is advanced beyond the point of effecting significant changes without major disruption. Small pipe, electrical, and instrumentation installation can accommodate changes, but halting installation is not deemed appropriate.

Continuing installation of small pipe, electrical, and instrumentation and then ef fecting changes will have less impact on project cost and schedule than halting installation because-electrical and instrumentation are on the present critical path for completion of the project. It is essential that Consumers Power Company continue with the installation of these items.

Halting construction on some of the systems impacts others in the same physical area because of installation of restraints and hangers. In addition to the schedule delay and increased costs asso:iated with a halt in construction, key personnel would be lost to the project.

Consumers Power Company has evaluated the Three Mile Island s

accid ent for impact on the Midland design and has closely s_,) monitored industry and NRC pertinent activities. Through this process, the effect of potential plant modifications on construction has been analyzed with, in general, a decision to continue these activities. In some circumstances, a conclusion has been to hold construction in specific areas pending final design determinations. A specific example of this is the MAD valves which are on an installation hold until completion of design review.

4 i

1 1

D-2 11/79 l

RESPONSE TO 10 CFR 50.54(f)-

APPENDIX E 9

Question e Comment on the OTSG sensitivity to feedwater transients. .

t Res ponse The Babcock & Wilcox nuclear steam supply system employs once-through steam generators (OTSGs) for heat transfer from the primary to the secondary system. The nuclear OTSG is a vertical straight shell and tube boiler in which the primary coolant (heat source) is on the tubeside and the secondary coolant is on the -

shellside. Main feedwater enters near the bottom of the tube bundle and flows upward. As it gathers heat, steam is generated and superheated before exiting to the steam piping system. Ti e overall primary-to-secondary heat transfer is controlled by the rate of feedwater introduction to the generator. This, in-turn, l

controls the area of the total tube bundle length which is exposed to liquid secondary coolant for a given input of primary power. Increasing feedwater flow increases the heat transfer and decreasing feedwater flow decreases heat transfer. l1 The design of the OTSG has yielded superior performance both in safety and efficiency in pressurized water reactors. The once-through design, with its superheated steam, exhibits a higher thermal efficiency than a recirculating steam generator, i

f' resulting in less waste heat rejected to the environment, better utilization of the uranium fuel, and a lower cost for electric power generation. The OTSG has exhibited an exceptional tube integrity record over its operating experience. This not only maximizes generator availability, but also minimizes the risk of radioactive release via a tube rupture. One inherent-feature of this design is the responsiveness to feedwater control mentioned above. This responsiveness makes possible an accuracy of control which has both operational and safety advantages. Safety analysis of limiting feedwater and secondary system pressure l

disturbances has demonstrated the ability to maintain safe core cooling without radioactive release _under the applicable licensing assumptions. However, the frequency of feedwater transients leading to disturbances of pressure and/or pressurizer level in the primary system of B&W plants has been higher than desired. This has been somewhat exacerbated by restrictions on plant operations which have been imposed since the TMI-2

. accident. Consumers Power Company supports the concept of l defen.se-in-depth; and existing plant features: accomplish

! this defense-in-depth as indicated on the attached i Figure E-1, which uses an overcooling event as an example.

l Additionally, through studies of the Three Mile Island accident and evaluation of the' operating history of B&W plants, Consumers Power Company has reviewed the Midland design for changes providing positive enhancement to the defense-in-depth concept. Plant modifications and areas O identified for further investigation as a result of this examination are presented in Appendix F.-

E-1 Revision 1 12/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX E Existing plant features, plus these changes, are sufficient to resolve many of the concerns expressed in Enclosure 1 of your (x}/

,, letter. In suppost of this conclusion, the following comments provide a point-by-point discussion of your Enclosure 1. In summary, we have concluded that is is neither necessary nor desireable to modify the fundamental operating characteristics 1 of the OTSG in view of its excellent performance record. i

a. Concern: " System modifications to increase the reliability of the AFW may have resulted in more frequent AFW initiation.

However, use of AFW results in introduction of cold (100 versus 400F) feedwater into the more sensitive upper section of the steam generators. This may act to enhance system sensitivity."

Comment: The Midland auxiliary feedwater system is a safety grade system' affording improved reliability as compared to older designs. AFW injection into the upper region of the OTSGs is an excellent design in that this configuration aids the initiation and It ismaintenance of reactor recognized that coolant upper head natural injection circulation.

serves to more closely couple AFW flow and temperature conditions with primary system response. Modifications to the AFW level control system, as discussed in Appendix F, will serve to alleviate this concern.

b. Concern: "Further system modifications provide control grade O reactor trips based on secondary system malfunctions such as turbine or feedwater pump trip. While these reactor trips do serve to reduce undercooling feedwater transiants by reducing reactor power promptly following LOMFW, they may amplify subsequent overcooling."

comment: Anticipatory reactor trip on loss of main feedwater has, in fact, yielded very smooth system response. This has been confirmed by recent field data. Use of anticipatory trips should be eliminated, however, for those disturbances ,

(such as turbine trip) which can be handled by plant control system action without challenging the plant safety systems.

This will reduce the number of plant trips (see Item d below). Discussion of the anticipatory reactor trip system to be incorporated in the Midland design can be found in Appendix F.

c. Concern: "A reexamination was made of small break and loss of feedwater events for B&W plants. This resulted in a modification of operator procedures for dealing with a small break which includes prompt RCP trip and raising the water level in the steam generators to 95% to promote natural circulation. Both these actions are taken when a prescribed low-pressure setpoint is reached in the reactor coolant system and for anticipated transients such as loss of E-2 Revision 1 12/79 l _ _ ._ _

RESPONSE TO 10 CFR 50.54(f)

APPENDIX E

() feedwater these (sic) actions may amplify undesireable primary system responses."

l1 Comment: The addition of an automatic reactor coolant pump trip, based upon the coincidence of signals indicating both low coolant system pressure and significant voids in the primary system, will eliminate the necessity for the operator to manually trip the reactor coolant pumps and raise OTSG water level. With the addition of this automatic function, reactor coolant pump trip should occur only for actual small breaks in the primary system. It should not occur for overcooling events initiated by feedwater transients. Such an automatic coincidence system is to be installed on Midland and is further discussed in Appendix F.

d. Concern: "In addition to the post-TMI changes discussed-above, actions were also taken to reduce the challengee to the power-operated relief valve (PORV) by raising the PCRV setpoint and lowering the high-pressure reactor trip. While these actions have been successful in reducing the frequency of PORV operation, they have resulted in an increased number of reactor trips. This occurs because the reactor will not trip for transients it previously would have ridden through '

by ICS and PORV operation."

/ Comment: The raising of the PORV setpoint and lowering of k_)N the high-pressure reactor trip have increased the number of s

reactor trips on the B&W operating plants. For Midland, ,

modifications are discussed in Appendix F which will restore the controlled relief capability of the PORV while providing an increased level of protection against PORV malfunction. A resetting of the reactor protection system high-pressure ,

setpoint and the PORV setpoint to the original values will restore the capability of the B&W nuclear steam systam to sustain a wide range of operational transients without a high-pressure reactor trip.

e. Concern: "It is felt that good design practice and maintenance of the defense-in-depth concept requires a stable, well-behaved system. Meticulous operator a?.tention and prompt manual action is used on these plants to compensate for the system sensitivity, rather than any inherent design features."

' Comment: The B&W OTSG and nuclear steam systems are designed

, to avoid reactor trip during various secondary system transients. This responsiveness is an inherent feature of the design. Some B&W operating plants do place reliance upon the operator to limit feedwater excursions which may result t

j from control system failure. However, Midland already includes a number of features to reduce this reliance upon the operator. Several additional modifications are being l

E-3 Revision 1 12/79

. RESPONSE TO 10 CFR.50.54(f)

APPENDIX E l

'\ investigated which will further reduce the requiremer.ts for the operator to act in response to a control system dailure  ;

and will thus improve our defense-in-depth against primary l system parameter excursions resulting from moderate secondary i system upsets. These changes are discussed in Appendix F. ]

\

f. Concern: "It appears that in many cases the main feodwater control system does not react quickly enough or is not sufficiently stable to meet feedwater requirements. Rather, j

' the system will often oscillate from underfeed to overfeed  !

conditions, causing a reactor trip and sometimes a high-pressure injection initiation. One undesirable element i

of this lack of stability is that overcooling transients on j the primary side proceed very much like a small break LOCA  !

(decrease in pressurizer level and pressure). Thus, for a i certain period of time, the operators may not know whether

they are having a LOCA c r an overcooling event." .

Comment: Overcooling transients in all PWR s} stems proceed initially like a small areak LOCA, and thus are not a unique problem of the GTSG. For example, on a recent reactor trip l i '

in a PWR with a recirctlating steam generator, a stuck-open turbine bypass valve with approximately 5% capacity caused an  ;

l excessive overcooling which resulted in a prompt loss.of  ;

reactor system pressure to the setpoint of the automatic safeguards injection system and contraction of the primary l coolant suf ficient to take pressurizer level.'below the range of indication. Consumers Power Company believes that proper .

design will result in a reduction in the frequency of such 4

events to the greatest degree practicable, which, when ,

i combined with satisfactory design mitigative capability,- 3

' adequate operator indications, and detailed training, will .

help the plant operators to actConsumers confidently and Company safely when l_1 l

these abnormal events occur. Power efforts  !

to improve identification and response to overcooling events  ;

are addressed in Appendix F. l i g. Concern: " A major area of concern arising from the B&W OTSG l

sensitivity is the response of pressurizer level indication. ,

4 Several B&W feedwate. transieats have' led to loss of pressurizer level indiation, Most notable was a

~

l f

November 1977 incident at Davis-Besse where 1.avel indication was lost for several mint'ces." t Comment: The loss of fressurizer level indication following l a, reactor trip is an operational concern and should be ~it However, minimized for expected abnormal occurences.

should be noted that the loss of indicated pressurizer level  !

on B&W operating plants is not synonymous with a loss of: l liquid'in the pressurizer. 7Certain B&W plants, such as the  ;

Davis-Besse Unit 1 reactor, have pressurizer level indicators '

l which'do not cover the full span of the pressurizer volume.

E-4 Revision 1 l i

12/79 I.

l

RESPONSE TO 10 CFR 50.54(f)

APPENDIX E In the case of Davis-Besse, more than 40 inches of pressurizer capacity remain below the zero point of the level indication system. Thus, at these plants a momentary loss of indicated level should not be confused with an emptying of the pressurizer and potential for loss of natural circulation. For Midland, the indicated pressura er level range will more closely relate to the full fluid volume of  !

the pressurizer and, therefore, the loss With ofthis indicated expanded pressurizer level will be minimized.

indication range, pressurizer level is expected to remain on-scale for feedwater upset transients such as those that- '

have occured at Davis-Besse.

h. Concern: "Some concerns also exist with regard to the operation of the pressurizer heaters when loss of level takes place. Nonsafety grade control circuitry If trips these the heaters nonsafety grade off when pressurizer level is low. l cutoffs should fail, the heaters would be kept on while uncovered." 1 r

comment: B&W operating plants include a control grade circu!t to remove power from the pressurizer heaters when liquid level is low, and in no instance have the pressurizer heaters on B&W operating plants been energized while

uncovered. The function of the heater interlock is being ,

reviewed for design adequacy.

l

i. Concer$: " Overfeed transient (MFW) (not uncommon to B&W) causei' overcooling; pressurizer level shrinks, pressure  ;

reachus 1,600 psi, RS actuates, RCP is tripped; AFW on l

(possible RCP seal failure)." ,

comment: For the Midland pla1.t, automatic equipment will be installed to eliminate the reacter coolant pump trip associated with low reactor coolant pressure only. In

• addition, main feedwater overfeed limiting equipment, independent of the integrated control system, will be l1 investigated as a means to terminate main feedwater flow This investigation is before excessive overcooling occurs.

further discussed in Appendix F.

.l . Concern: " Operator manually controls AFW (possible MFW instead or in addition if MFW is not isolated such that OTSG level comes up to 95% of operating range). This massive

addition of cold water may : lead to emptying of pressurizer

~

and interruption of natural circulation (or the hot leg may flash due to depressurization and interrupt natural circulation even if pressurizer doar r.st empty)." I i

i E-5 Revision 1 I

12/79

= - .

RESPONSE TO 10 CFR 50.54(f)

APPENDIX E Comment: The auxiliary feedwater system control circuitry will be modified for Midland to minimize the excessive addition of cold water which could lead to emptying of the pressurizer. Further discussion of this subject is focad in Appendix F.

k. Concern: "HPI delivers cold water, no heat transfer in OTSG, vapor from core leads to system repressurization; steam may condense or PORV may lift."

Comment: B&W calculations do not predict an interruption of core cooling or heat transfer to the OTSG as a result of the events sequence outlined. Delivery of the cold water by the high-pressure injection system will refill the reactor coolant system and quench any voids to provide additional i assurance of adequate core cooling.

1. Concern: "No pump restart criteria are available, and circulation may not be reestablished."

Comment: Criteria for restart of a reactor coolant pump are already provided in the current small break operating guidelines to permit forced flow to be reestablished promptly following repressurization of the reactor coolant system.

Further work in this area is proceeding under the abnormal

/~' transient operation guidelines (ATOG) program discussed in l_-) Appendix F in which Consumers Power Company is participating.

m. Concern: "It appears that an upgraded safety quality ICS which is designed to balance power to OTSG level in a better fashion could reduce the sensitivity illustrated in the above ,

sequence."

Comment: The integrated control system is designed to provide smooth and stable operation of the complete power plant during power operation. One of its functions is to l1; maintain the reactor plant online following various secondary system disturbances and eliminate unnecessary challenges to the reactor trip system. Following reactor trip, the ICS has a function in maintaining stable plant conditions within design limits. The recently completed ICS failure modes and effects analysis (FMEA) has identified measures which would improve the reliability of all controlConsumers functionsPower related to Company's the B&W operating plant ICS design.

response to the results of this FMEA is addressed in

. Appendix F. Control of auxiliary feedwater is provided by a l

safety grade system independent of the ICS. 1.s discussed in Item i, a system separate from the ICS to limit the main feedwater introduction which might occur as a result of primary control system failure is being investigated. The combination of improvements presently incorporated in the

, O,' Midland design and those under consideration should provide l

E-6 Revision 1 12/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX E

() substantial defense-in-depth against sequences of the sort discussed in this section. e Concern: "Regardless of the reasons, B&W plants are currently experiencing a number of feedwater transients which  ;

the NRC Staff feels are undesirable. The NRC Staff believes that modifications should be considered to reduce the plant sensitivity to these events and thereby improve the defense-in-depth which will enhance the safety of the plant."

n. Comment: Safety analysis has shown that adequate core cooling will be maintained and radioactive release will be avoided even for the most severe secondary system accidents within the plant's licensing basis. Midland already incorporates a number of design features which address the issuer, raised in this paper by improving system reliabilityIn and reducing the consequences of secondary system upsets.

addition to this, a carefully considered group of investigations and modifications discussed in Appendix F are being undertaken to reduce primary system response to feedwater disturbances and to reduce the magnitude and frequency of secondary system feedwater upsets. These modifications will improve plant performance and enhance safety through the defense-in-depth concept by terminating or mitigating transients early in their course before they result in seriously abnormal conditions.

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I CP R(P Ym

RESPONSE TO 10 CPR 50.54 ( f)

APPENDIX F

( ) Question f Provide recommendations on hardware and procedural changes related to the need for and methods for damping theInclude primarydetails system sensitivity to perturbations in the OTSG.

on any design adequacy studies you have done or have in progress.

Res ponse Much of the concern expressed about the " sensitivity" of the B&W OYSG PWR design is based on the operational experiences with the currently operating 177-FA plants, and particularly that experience accumulated since the accident at TMI-2. It is important to recognize that the normal evolution of design that has occurred on Mioland as a result of new regulatory requirements, improvements of the state-of-the-art in hardware, and the feedback of operating experience have resulted in the incorporation of several new features. These features serve to improve the reliability of the systems and equipment and thereby reduce the probability of challenges to the safety system, improve the response to the NSSS to those events that do occur, and provide better capability to mitigate the events that occur.

Some of the more significant pre-TMI-2 changes include:

Upgrade of required pressurizer heaters and controls to l1

(h a.

safety classification to ensure RCS subcooling (s_)

b. Addition of a two-channel, Class lE auxiliary feedwater control system

c. Initiation of auxiliary feedwater by the Class lE engineered safety features actuation system (ESFAS)

d. Addition of feed only good generator (FOGG) logic to the ESFAS to help ensure that auxiliary feedwater is delivered ,

only to the intact steam generator following secondary system breaks

e. Adoption of newer control systems hardware (NNI/ICS) which uses dual, auctioneered power supplies for the logic modules rather than individual power supplies for each logic module as in the earlier design In addition to the described design improvements over current operating plants afforded Midland as a result of state-of-the-art advancement, Consumers Power Company has initiated a reevaluation of plant design in light of the TMI-2 accident and its implications concerning further modifications. Some of these investigations have been initiated specifically to address to concern of overcooling in B&W type plants. The following material discusses changes that have been identified for l (()s F Revision 1 12/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F g_ ,/ incorporation in the Midland design pertinent to this issue and also addresses additional evaluations pertaining to overcooling In general, that are being conducted by Consumers Power Company.

the impact of presently identified modifications has been incorporated in the Midland project schedule now undergoing revision. Areas undergoing further evaluation have been reviewed for construction impact and, where appropriate, steps have been taken to ensure accommodation of expected changes within the construction schedule.

I. DAMPING THE PRIMARY SYSTEM SENSITIVITY TO PERTURBATIONS IN THE OTSG As an outgrowth of the Three Mile Island accident, IE Bulletin 79-05B outlined new requirements to enhance plant response to undercooling type transients. Plant modifications resulting from these requirements were expected to enhance the creation of natural circulation by inhibiting RCS voiding during these events and thus limit the impact on the RCS of anticipated transients  !

leading to undercooling events. In response to these requirements, the B&W operating plants, with the concurrence of the NRC, inverted the PORV and high RCS pressure trip setpoints and installed automatic reactor trip logicThese actuated by either changes addressed l1 turbine trip or loss of main feedwater.

f-~)

the NRC desire to minimize the operation of the PORV, thereby

\' ' reducing the probability of RCS blowdown caused by a stuck-open valve. Additionally, energy input into the RCS was reduced in through prompt reactor trip during transients that resulted .

primary system pressure increases.

Although these changes have succeeded in limiting the PORV actuations and RCS stored energy resulting from undercooling events, they have resulted in a significant increase in the  :

frequency of reactor trips. Specifically, post-Three Mile Island plant history compiled by B&W demonstrates as much as a doubling l1 of trip frequency of certain plants. Because overcooling events are most likely to occur following reactor trip, it can be concluded that the potential for these transients has increased.

Therefore, while the applicable concerns of the NRC Staff resulting from events at Three Mile Island seem valid, it appears that the prescribed solution is inadvisable because the method used to minimize the impact of an undercooling transient may increase the probability of an overcooling event.

Consdmers Power Company intends to adopt an alternative solution \1 that addresses TMI-2 concerns regarding primary system overpressure transients while maintaining a plant design resistant to overcooling events. This solution incorporates the following features:

A. Original B&W 177-FA PORV and high RCS pressure setpoints

[~'}

\_- (2,255 psig and 2,355 psig, respectively)

F-2 Revision 1 12/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F B. Safety grade anticipatory reactor trip on total loss of ON feedwater C. Fully qualified safety grade PORV D. Reliable safety grade indication of PORV position E. Dual safety grade PORV isolating block valves actuated by low RCS pressure ESFAS signal F. Test program to demonstrate PORV operability (EPRI)

Thus, for secondary transients (turbine trip / load rejection) , the original B&W design features of turbine bypass, ICS runback, and PORV actuation are retained to keep the reactor online.

Therefore, a critical reactor at power is available to minimize the probability of an overcooling event. To address the TMI-2 concern of actuation of an assumed unreliable PORV, design modifications are incorporated to ensure the proper operation of the valve, to display adequate indication of its position, and ,

provide automatic isolation of the valve if it fails.

Specifically, the commitment to provide an additional block valve and automatic isolation of the PORV addressed in Item e above is contingent upon restoration of the PORV pressure-reducing function as afforded by Item a. For loss of feedwater eV0nts

'N during which reactor trip is a certainty, specific additional trip circuitry is provided to anticipate the high RCS pressure trip and thus minimize energy input into the primary system.

In summary, the design features discussed above address' the ist2e of overcooling by minimizing unnecessary reactor trips while -

providing the capability to prevent undercooling transients or, if necessary, adequately mitigate their consequences. This ,

approach provides the most balanced and logical method for dealing with these two opposite, yet related, plant transients.

II. OTHER RELATED DESIGN ADEQUACY STUDIES A. Auxiliary Feedwater System A design review of the Midland AFW system since TMI-2 '

has resulted in several modifications to the original

! system design. The most significant was a modification of the AFW pump suction piping from one interconnected system for both Midland units to two systems operating independently to supply AFW for each unit.

Additionally, AFW flow indication is being upgraded to safety grade. Another potential modification previously identified is the addition of redundant flowpaths from the discharge of each AFW pump to each steam generator.

g ~g This potential change is being examined to gage its g j potential impact on system reliability in an in-depth F-3 11/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F reliability assessment of the Midland AFW system. To protect the Midland project schedule while this evaluation is being conducted, the additional AFW flow control valves that may be necessary are being ordered.

The AFW analysis is being conducted with the aid of an outside consultant, and is to identify both independent and dependent component failure modes, including the effects of equipment maintenance and operator errors, under the following three scenarios:

1. Loss of MFW with offsite ac power available

2. Loss of MFW with offsite ac power unavailable

3. DC power available only The probability of system failure for each scenarioNo will be calculated via fault and event tree techniques.

significant additional contributors to system failure having a substantial construction impact are expected to be identified in this formal reliability analysis.

For transients such as loss of main feedwater and loss of offsite power, the proper operation of the AFW level O control system is essential to the prevention of RCS overcooling. System control, as presently designed, is based solely on steam generator level error and allows essentially full AFW flow to the OTSGs until actual ,

level approaches the setpoint. Midland is currently investigating modifications to the AFW level control system which would limit the primary system cooldown rate following AFW actuation and incorporate multiple setpoints for final level. The level setpoint selected would be dependent on conditions of the reactor coolant pumps and would include the higher level setpoint required for mitigation of small break LOCAs. Limiting the cooldown rate of the primary system will allow time for the makeup system to recover pressurizer level and allow time for the operator to take actions necessary to prevent loss of indicated level. Analysis work is proceeding on an AFW level control scheme which limits primary system cooldown rates by limiting the rate of I

i l

steam generator level increase (i.e. , ' limiting AFW

! flowrates). The major impact of this change will be in obtaining the electronic hardware necessary to implement the new control scheme. Changeout of control electronics has a minimal impact

! on overall system construction because it can be accomplished within a few weeks following receipt of the necessary hardware. Therefore, there is no need to halt

(

F-4 11/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F

(

O construction on the AFW system based on expected changes to the AFW control scheme.

As discussed in the introduction to this section, the Midland design includes a FOGG system whose function is to identify the ruptured steam generator following a secondary systems line break and to prevent feeding AFW to this steam generator if the break is upstream of and cannot be isolated by the main steam and feedwater isolation valves. Blocking AFW flow to the ruptured steam generator prevents an uncontrolled cooldown of the '

primary system. The FOGG system accomplishes this without need for operator action. The current Midland FOGG logic permits AFW flow to the steam generator which repressurizes to 2725 psig following an actuation of the main steam line isolation system (MSLIS). This repressurization results in a permissive allowing AFW flow only to the intact steam generator. Recent steam line break analysis work by B&W has shown that for certain sizes of steam line breaks, the intact steam i generator will not repressurize to 725 psig and, therefore, neither steam genrator will receive AFW flow.

As a result of this potential for the presently designed FOGG system to block AFW flow to both steam generators, the logic for FOGG actuation is being reviewed. The

/~N k ,) following two options for correcting the identified s

problem with the FOGG system are being investigated.

1. Lowering the FOGG setpoint to a value less than 725 psig such that in all steam line break cases it can be shown that the intact steam genrator will repressurize above this value.

2. Modifying the FOGG logic to use differential pressure between the steam generators as a detecting parameter for FOGG actuation.

The latter option, actuating FOGG on steam generator differential pressure, appears to be the best method at present. Additional analytical work is required to determine the differential pressure setpoint to be utilized to actuate this system and to verify operability over the range of main steam line breaks.

Either change in the method of actuating FOGG described above will result in changes to control circuitry only.

No changes to piping are required to modify the FOGG system. Because control circuitry changes can be made within a few weeks of receipt of hardware, no halt in construction is justified based on potential modifications to the FOGG system.

l f (\ -

F-5 11/79

RES PONSE TO 10 CFR 50. 54 ( f ) ,

APPENDIX F

\ ,/ B. Pressurizer As mentioned in the introduction to this section, redundant portions of the pressurizer heaters and heater ,

controls have been upgraded to safety grade. This l upgrade provides the ensured capability of maintaining adequate subcooling after all anticipated transients i through proper RCS pressure control. Additionally ,

redundant pressurizer level and reactor coolant pressure indications have been upgraded to safety grade on the main control boards and auxiliary shutdown panel. To specifically address the concern of loss of pressurizer 1

level indication due to overcooling events, the ,

indication range is being extended to a scale of 0 to ,

400 inches. The original design range of 0 to l

320 inches creates a greater potential for an actual off-scale low level as demonstrated by operating plant ,

l history. The 40-inch increase at both ends of the present operating range, in conjunction with anticipated improvements in the AFW level control scheme discussed  ;

in Section II. A,, will provide additional assurance that i the pressurizer level will remain on-scale for all ,

anticipated operational occurrences. This design modification is currently being implemented.

()

In the event of loss of liquid inventory in the pressurizer, future availability of the pressurizer heaters may require their deenergization before uncovery. As a result, the existing heater low level interlock design is being reviewed to judge its ,

adequacy. Because modifications resulting from this  !

investigation would affect only control and/or ,

instrumentation design, no impact on construction schedule is expected.

C. Transient Identification t As discussed in Appendix E, overcooling events in all PWR systems proceed initially like a small break LUCX.

Therefore, it is important that any auotmatic plant t response required for either of these events be able to i differentiate between similar appearing transients in  !

order that system actuations only occur when desired.

Additionally, suf ficient indications are necessary to ,

allow the operator to follow the course of the transient and l1 verify proper safeguard features operation and adequate  ;

core cooling until the exact cause of the event can be positively identified. The current status of investigations by Consumers Power Company in these two ,

f3 areas is discussed below. In general, specifically i

(x --) identified design changes can be accommodated within the F-6 Revision 1 12/79 f - _ .

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F present construction schedule and modifications resulting from ongoing investigations are expected to effect only control and instrumentation design and therefore can be instituted during system construction with no anticipated impact. ,

1. Automatic Plant Response Additional investigations initiated as a result of the small break LOCA analyses submitted to the NRC by B&W (Evaluation of Transient Behavior and Small '

Reactor Coolant System Breaks in the 177 Fuel Assembly Plant, May 7, 1979) have demonstrated that under conservative conditions, tripping of the RCPs is l1 necessary for mitigation of certain size small breaks. Tripping of the reactor coolant pumps for overcooling type transients is, howevgr, undesireable because this action is not necessary '

for proper plant recovery and, in fact, sacrifices enhanced controlability afforded by forced circulation. To prevent automatic RCP tripping due to ESFAS actuation initiated by overcooling events, the Midland pump trip logic will include coincidence circuitry sensing RCP motor current.

This input will actuate on degraded pump current indicative of significant RCS void formation characteristic of a LOCA. For overcooling events, the extent cf void formation will not reach a point where degraded pump current will actuate RCP trip. (

i

2. Plant Indications t

a. Psat/Tsat Meter - Consumers Power Company is committed to providing a subcooling meter with redundant safety grade hot leg temperature and reactor coolant system pressure input,

b. Core Exit Thermocouples - Consumers Power Company l1 is assessing a technical proposal to utilize core exit thermocouples as a means of determining adequate core cooling. Specifically, the use of nonsafety grade core exit thermocouples in conjunction with the plant computer is being i

assessed for possible use in providing core map temperature trending, margin to' saturation of the average incore thermocouple temperature, j and margin to saturation of the hottest valid r thermocouple temperature.

1 i

( O)

F-7 Revision 1 l 12/79

RESPONSE TO 10 CFR 50.54 ( f)

APPENDIX F O Primary Coolant Level Indication - Consumers

( j c.

Power Company is participating in a B&W engineering study to determine the most appropriate method for operator recognition of inadequate coolant level. The method presently being considered to provide the information is hot leg water level in lieu of reactor vessel water level. Differential pressure on the hot leg (from the top of the candy cane to the bottom of the hot leg piping) is the technique being evaluated for measuring the level.

d. Natural Circulation Flow Indication -

Consumers Power Company is reviewing the r

technical feasibility of providing a low flow indication as a means of confirming core cooling during natural circulation modes of cooldown.

Control room panel space is available for the hard l1 wire display of the above-mentioned plant indications, if implemented. However, core exit thermocouple information is intended to be displayed in the control room through the plant computer-printer and/or CRT.

}

D. Integrated Control Systems (ICS) FMEA In response to the TMI-2 event, B&W performed a failure mode and effects analysis (FMEA) of the ICS. The results of this analysis, supported by Consumers Power  :

Company through the B&W owners group, have been supplied to the NRC (BAN-1564, August 1979). As a result of this i effort, several areas have been identified as warranting additional review on a plant-specific basis for evaluation of possible changes which may result in the enhancement of reliability and safety. Consumers Power Company is evaluating the recommendations contained within the ICS FMEA and intends to make modifications necessary to elicit improvements based on the results of this evaluation. Design changes resulting from this investigation would affect only control and/or instrumentation and, therefore, could be accommodated at any time during system construction.

1 0

.U F-8 Revision 1

! 12/79 L

RESPONSE TO 10 CFR 50.54( f)

APPENDIX F E. Main Feedwater

\ As a result of your 10 CFR 50.54(f) request of October 25, 1979, B&W has undertaken several reviews designed to further evaluate causes of overcooling events and mitigative plant response. One of these studies consisted of examination of operating plant experience aimed at identifying the sources of overcooling transients and assessing the effect of possible modifications. Additionally, B&W has reviewed the typical sequence of events for overcooling transients, identified the existing plant design  :

features which provide defense-in-depth against the occurrence of inadequate core cooling, and investigated the need for additioanl changes where suggested to improve these defenses. These studies, the ICS FMEA discussed in Section II.D, and the Consumers Power Company review of B&W operating plant experience and Midland plant design have identified various MFW f aults which could lead to secondary system upsets. Consumers Power Company intends to bring together information from these sources in a detailed review and analysis of the MFW system. The outcome of this task is expected to be an identification of changes which would significantly decrease the frequency of feedwater upsets.

From the analysis that was presented on turbine / reactor

\. trip and main feedwater overfill, it has become apparent that the OTSG may become filled and water will enter the main steam lines. Consumers Power Company believes that this condition is not desirable and will provide ,

protection against this occurance upon evaluation of the options available. This change and any other changes to the feedwater system which may result from our continuing review are expected to impact controls and instrumentation only. Therefore, it is expected that any resulting changes can be accommodated at any time during system construction.

F. Miscellaneous Studies In May 1979, Consumers Power Company contracted with EDS Nuclear to conduct a design review of selected Chapter 15 accidents and selected plant systems (safety and non-safety). Sixteen safety and operational sequence diagrams and fifteen auxiliary systemThe diagrams are being methodology for used as a vehicle for this review.

i the analysis is identical to that referenced in Chapter 15 (Page 15-5) of Regulatory Guide 1.70, Rev 3.

All identified potential design inadequacies are being formally documented and resolved through joint action resulting from Consumers Power Company, Bechtel, and B&W

{v}

F-9 11/79

RESPONSE TO 10 CFR 50.54(f)

APPENDIX F review. The analysis has presently identified some

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( ,) deficiencies (e.g., improper main steam line isolation signal initiation logic) . This deficiency, which is believed to be representative of others which may be uncovered, will be corrected by modifications to the system controls and therefore can be accommodated during system construction.

Following the. release of the Short Term Lessons Learned l1 Report (NUREG-0578), the B&W 177-FA owners group, including Consumers Power Company, embarked on the anticipated transient operating guidelines ( ATOG) program. The basic input, with minor modifications in 1 scope, is the safety sequence analysis performed by EDS This analysis Nuclear for Consumers Power Company.

provides the design input for the ATOG program which involves construction of event trees, dynamic analyses, and development of operating guidelines. Because the safety sequence analysis provides the basic design input to the program, no hardware changes are expected to result from ATOG that will not first be identified by the EDS prog ram. Expected changes resulting from ATOG will be procedural in nature and therefore do not impact cons truc tion .

Finally, B&W has performed an analysis of the dynamic post-trip response of the NSSS to overcooling f-~)

t

\ transients. This post-trip responsiveness study investigated primary system senstivity, measured by fluctuations in pressurizer level, to changes in various plant parameters. Consumers Power Company is presently reviewing this analysis to determine its implications on plant design. Preliminary review indicates most concerns have already been addressed or are identified for further study.

/"\

L)

F-10 Revision 1 12/79

RESPONSE TO 10 CFR 50. 54 ( f)

APPENDIX F j III.

SUMMARY

The defense-in-depth approach to maintaining adequate core cooling is accomplished through system design and plant procedures. In the specific case of overcooling events, Figure E-1 depicts the implementation of this approach '

through provisions to minimize the frequency of the transient, indications to allow the operator to detect the event and evaluate the adequacy of the response, and procedures to ensure proper actions are taken and to identify additional steps to counteract failures or degraded conditions as identified. While current plant design fulfills the requirements for defense-in-depth, Consumers Power Company believes that additional measures can and should be taken to strengthen the Midland plant's resistance to overcooling events and to ensure acceptable Design mitigation changes describedfor in those transients which do occur. 1 this appendix, along with additional modifications and procedural changes which may result from the studies that have been addressed, are expected to accomplish this goal.

Table F-1 summarizes these items and Figure F-1 shows how  :

they will compliment existing plant features to provide additional overcooling defense-in-depth. Consumers Power Company believes that this approach will adequately resolve current concerns.

i In conclusion, Consumers Power Company is autively pursuing

\_s/ the issue of overcooling and has committed to modifications which, coupled with the outcome of various ongoing plant l1 studies, will result in a finalWe design which provides adequate believe that most changes which de'fenses to these events.

are significantly impacted by construction status have already been initiated and that any remaining modifications -

resulting from completion of design studies can be accommodated during system construction.

\._)

i t

F-11 Revision 1 f 12/79

~

in- - , . . . . . . . ~ , - -

RES PONSE TO 10 CFR 50. 54 ( f)

APPENDIX F TABLE F-1 DEFENSE-IN-DEPTH FEATURES FOR MAINTAINING ,

ADEQUATE CORE COOLING

1. Automatic Closure of PORV Block Valves on ESFAS Actuation

2. PORV Upgrade and Qualification

3. Pressurizer Heater Upgrade

4. Extended Pressurizer Indication Range and Upgraded Qualification

5. Fully Safety Grade AFW System

6. FOGG System

7. Total Loss of Main Feedwater Anticipatory Reactor Trip

8. Safety Grade PORV Position Indication

' 9. APW System Improvements (Reliability Analysis, Flow Indication Upgrade, Piping Modifications)

(

10. Improved AFW Flow Control

11. Pressurizer Heater Interlock

12. Automatic RCP Trip Circuitry Featuring Low Motor Current Coincidence Logic l1

13. Instrumentation to Detect Inadequate Core Cooling

14. ICS FMEA i

15. MFW System Review

16. ATOG Program

17. Restoration of Original PORV and High RCS Pressure Trip Setpoints l

t O

F-12 Revision 1 12/79

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