ML19260B239
| ML19260B239 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 12/04/1979 |
| From: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| HOWE-315-79, NUDOCS 7912070451 | |
| Download: ML19260B239 (150) | |
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General offices: 1945 West Parnell Road, Jackson, Michigan 49201 * (517) 788-0453 December 4, 1979 Hove 315-79 US Nuclear Regulatory Commission Attn: Mr Harold R Denton Office of Nuclear Reactor Regulation Washington, DC 20555 '
MIDLAND PROJECT DOCKET Ho 50-329, 50-330 RESPONSE TO 10CFR50 54 REQUEST ON DESIGN ADEQUACY OF B&W SYSTE!G FILE: Oh85 19 SERIAL: 8026 Enclosed are ten (10) copies of Revision 1 to Consumers Power Company's response of November 30, 1979 to your 10CFR50 5h(f) request on B&W System Sensitivit,y dated October 25, 1979 Revision 1 is administrative in nature and provides replacement for certain preliminary graphs and tables and corrects minor errors and omissions. Since the majority of the pages are affected, a ecmplete response is being provided.
The changed pages bear the notation " Revision 1 12/79" and are marked in the margin to indicate where changes have been made. All other pages remain unchanged.
The technical review of Appendixes A and B referred to in my original transmittal letter dated November 30, 1979 continues and if our review uncovers additional thanges, we vill notify your staff and provide a revised report.
Consumers Power Company Dated: December 4, 1979 By Stephen Q Howell, Denior Vice President Sworn and subscribed to before me on this 4th day of December 1979 Y. &t.s4)
HotaryPublicyJacksonCountp, Michigan My commission expires September 21, 1982 \
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j Stephen H. Howell Senior Vice President oeneral offices: 1945 West Pernell Road, Jackson, Michigan 49201 e (517) 788-0453 December k, 1979 Hove 315-79 US Nuclear Regulatory Comission Atta: Mr Harold R Denton Office of Nuclear Reactor Regulation Washington, DC 20555 MIDLAND PROJECT DOCKET No 50-329, 50-330 RESPONSE TO 10CI" ISO.54 REQUEST ON DESIGN ADEQUACY OF B&W SYSTEM 3 FILE: 0485.19 SERIAL: 8026 Enclosed are ten (10) copies of Revision 1 to Consumers Power Company's response of November 30, 1979 to your 10CFR50 54(f) request on B&W System Sensitivity dated October 25, 1979 Revision 1 is administrative in nature and provides replacement for certain preliminary graphs and tables and corrects minor errors and omissions. Since the ma,jority of the pages are affected, a complete response is being provided.
The changed pages bear the notation " Revision 1 12/79" and are marked in the margin to indicate where changes have been made. All other pages remain unchanged.
The technical review of Appendixes A and B referred to in my original transmittal letter dated November 30, 1979 continues and if our review uncovers additional changes, we vill notify your staff and provide a revised report.
Consumers Power Company Dated: December 4, 1979 By -
J M~
Stephen %) Howell, Senior Vice President Sworn and subscribed to before me on this hth day of December 1979 YYA)
Notary Publicy Jackson CountyV, Michigan My comission expires September 21, 1982 1514 254
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parlar k (Q@l]3] Stephen H. Howell senior Vice President Generat of fices: 1945 West Parnsit nood, Jackson, Michigan 49201 * (517) 788-0453
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November 30, 1979 Howe-306-79 US Nuclear Regulatory-Commission - ~ * -
Attn: Mr Harold R Denton - - -- ' - -
Office of Nacicar Reactor Regulation Washington, DC 20555 .
. MIDLAND PROJECT . . .
DOCKET NO 50-329, 50-330 RESPONSE TO 10CFR50.54 REQUEST ON DESIGN ADEQUACY OF B&W SYSTEMS FILE: 0485 19 SERIAL: 7999 Enclosed are ten (10) copies of Censumers Power Company's response to your 10CFR50 54(f) request dated October 25, 1979 regarding the Design Adequacy of Babcock & Wilcox Nuclear Steam Supply Systems Utilizing Once Through Steam Generators for Midland Unit 1 and 2.
~
The attached response consists of Appendix A through F which correspond to .
your questions a through f. Appendixes A and B represent the B&W analysis input on overcooling events. Due to time constraints and our desire to meet your schedule, Censumers Power Company has not completed a detailed technical review of this c:sterial. Certain obvious modifications to cover the specific design details of the Midland Plant have been cade. If our review uncovers additional changes, we vill notify your staff and provide a revised report.
Consumers Power Company has studied the concern of sensitivity of the reactor coolant tempemture and volume to perturbations in the secondary system and has concluded that there are steps that can and vill be taken to reduce these secondary perturbations and to address the concern for sensitivity.
The discussion of plant changes is presented in Appendix F. It should be noted that these modifications do not involve major changes to large pieces of equip-ment such as vessels, heat exchangers or pumps.
Considering the magnitude of changes and for the reasons presented in Appendixes C and D, Consumers Power Company believes that installation of the affected syctcms can and should continue and that necessary modifications can be accommodated during continued construction. Hence, we believe there is no benefit in halting all or parts of construction of the Midland Plant Unit 1 and 2. d
2 Consumers Power Company Dated: November 30, 1979 B
- r3 d StephenIQovell.,SeniorVicePresident Sworn and subscribed to before me on this 30th day of November 1979
- f. Y b t 4&
Notary Publicy Jackson County, Wichigan My commission expires September 21, 1982 k
O 1514 256
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APPENDIXES A AND B 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. Identify whether action of the ECCS or RPS (or operator action) is necessary to protect the core following the most severe overcooling transient identified. If these systems are required, you should show that its design criterion for the number of actuation cycles is adequate, considering arrival rates for excessive cooling transients.
Response
I. INTRODUCTION AND CONCLUSIONS A. Background 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 pla.its, accounting for balance-of-plant features.
B. Scope This report responds to the specific NRC requests identified above. More than one transient type is analyzed to address different frequency of occurrence classifications and to ensure that the most severe cases are indeed included in the evaluation. A qualitative assessment of possible nonmitigative operator actions in the O to 10-minute time frame is also provided. This assessment provides indication of what operator action is anticipated during the initial phases of an overcooling transient.
The analyses identify the frequency of the RPS, ESFAS, and operator action for mitigation of the transient.
A summary of the results is given in Section II.
Section III provides the details of the initial conditions, computer codes, and basic assumptions used in the analysis. The transient response data are given 1514 257 f
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APPENDIXES A AND B in Section IV.Section V demonstrates the adequacy of the design criteria for each system.
C. Conclusions Based on the analyses performed in this report, the following conclusions can be drawn.
- 1. The overcooling accident (main steam line break) and the overcooling transient (main feedwater 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 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.
It should be noted that this report could not
- exhaustively determinc 'he most severe overcooling transient in the allotted time; the reasons for selecting main feedwater overfill are discussed in Section IV.A.l.
D. Applicability of Results The results presented in this report are applicable specifically to this NSS with the parameters tabulated in Section III. Specific attention has been paid to the balance-of-plant equipment in the mitigative functions performed.
II.
SUMMARY
This section provides a detailcd 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 concern.
Section II. A addresses the selection of anticipated transient and accident conditions causing greatest 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 1514 258 A&B-2 11/79
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APPENDIXES A AND B that the design criteria for the number of actuation cycles of the RPS and ESFAS are adequate.
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 additiona) studies were performed on the SLB because this accident was expected to result in RCS voiding. Howev t.r , it will be shown that the limiting moderate-frequency event analyzed does not produce voiding as a result of RCS coolant inventory shrinkage.
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. Review of plant transient data (see Section IV. A.1) has shown that overfeed by main feedwater af ter reactor trip has produced the most severe overcooling transients.
Therefore, based on arrival rates for operating plants and the cooldown rate associated with this transient, main feedwater overfeed following a reactor trip /
turbine trip is considered the limiting anticipated transient and is analyzed in Section IV.A.
B. Shrinkage Effects Shrinkage of the 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 saturated water during normal opera tio n . This liquid volume flows out of the pressurizer into the system as the system inventory valume 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 between hot, full power, and the final pressure / temperature achieved during the transient. Its effect is further mitigated by actuation of the emergency core cooling system (ECCS).
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APPENDIXES A AND B The other mechanism which produces steam voids in the RCS is flashing of RCS water. As the pressure rapidly decreases in the RCS, the liquid in the hotter portions of the system can become saturated by the hot metal in this area, flashing additional water to steam. This process, i,n a non-LOCA situation, is self-regulating.
As the steam separatea, 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 affected steam generator. If the affected steam generator is on the loop with the pressurizer, emptying the pressurizer contributes to the steam void formation. If the af fected 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 adequate core cooling.
The anticipated transient analyzed is the overfeed of the steam generators by main feedwater (Section IV. A) .
- This overcooling transient, with no mitigative operator action for 10 minutes, resulted in the pressurizer emptying briefly. However, HPI actuation is sufficient to prevent any steam voiding in the RCS.
The design basis steamline break accident (Section IV.B) 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:
- 2. With and without core decay heat 1514 260 A&B-4 11/79
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APPENDIXES A AND B
- 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. The core region remained subcooled throughout the transient for all cases analyzed. The remainder of SLB cases l1 pre,unced in Section IV.B all satisfy this criterion.
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 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. Adequate indications are available to the operator during steam voiding situations that would exist during the SLB accident analyzed to ensure the continuation of HPI flow. Pressurizer level and subcooled margin both indicate the necessity of HPI. Adequate core cooling would necessitate that HPI be available at some point during 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 stystem. 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.
III. ANALYTICAL TECHNIQUES A. Computer Co6es The B&W-cettified 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 3
simulation (similar to CRAFT 2, Reference 2), and is 3 capable of handling rapid overcooling transients that A&B-5 Revision 1 12/79 1514 261
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APPENDIXES A AND B 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 Figures A&B-1 and A&B-2. A description of each node and the important flowpaths are l1 given in Tables A&B-1 and 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 during the initial phase of the transient while the primary and secondary variables are rapidly changing. In the interest of computer calculational timesaving, 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 I result from single failures following any initiating event, and those which are made more severe from single failcares 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
- 3. Steam pressure regulator malfunction which causes increased steam flow
- 4. Inadvertent opening or stuck-open steam relief valve which causes increased steam 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:
A& B-6 Revision 1 12/79 1514 2t2
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APPENDIXES A AND B
- 1. Events which constitute an initiating event -
Items 1 through 4 above are moderate frequency, and steam regulator malfunction is the 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 occurence is a combination of a moderate frequency event plus one of Items 1 through 4 occurring as a single failure. The event chosen to be analyzed in this category is an inmediate reactor trip on turbine trip signal (decrease the heat source) combined with a feedwater flow control malfunction that allows continued main feedwater flow (increase the heat sink).
- 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 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.
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.
The assumption of a stuck rod was removed from the shutdown rod worth, resulting in a more realistic, conservative direction for the overcooling 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 parameterization 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.
Table A&B-5 lists the equipment assumed to function for each transient analyzed.
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APPENDIXES A AND B No mitigative operator action is assumed for 10 minutes 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 were also reviewed. For 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 main feedwater.
Operating plant data show 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 ggg generators following a reactor trip, which has caused the greatest primary cooldown observed.
Steam pressure regulator malfunctions that allow increased steam flow would represent overcooling by depressurizing the secondary system. Its effect is very similar to a small SLB analysis. The arrival rate for this transient has been zero at cperating B&W plants. Therefore, in the limited time frame for the preparation of this report, the MFW overfeed transient is presented. The MFW overfeed represents the maximum cooling t!!at can be achieved by feeding the OTSGs.
- 2. Main Feedwater 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.
The sequence of events for this transient is given in Table A&B-6. The analysis was performed using the models and assumptions givec in Section III. A comparison of the maxi-TRAP and mini-TRI? system parametere is shown in Figures A&B-3 to A&B-5 for the first 2 miuntes of the transient. Based on the relatively good agreement of these results, the A&B-8 11/79 1514 264
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APPENDIXES A AND B mini-TRAP run was extended to 10 minutes. No credit for ICS and operator actions was assumed.
For this analysis, it was assumed that only safety grade equipment functioned. Because the ESFAS signal on low RC pressure does not directly actuate AFW, the steam generators must boil down to the lov level actuation cetpoint to initiate flow. AFW actuation would also occur when the main feed pumps are tripped due to actuation of a main steam line isolation signal (MSLIS) upon ESFAS actuation.
However, AFW injection would not occur until the OTSG level setpoint was reached, which would be af ter the 10-minute time period encompassed by this analysis. Therefore, actuation of AFW at an earlier time w^uld not increase che severity of the overcooling transient.
Figures A&B-6 through A&B-13 ptesent system parameters. For this transient, the pressurizer empties briefly at about 3 minutes. However, during the 30- to 50-second duration before HPI starts to increase RCS inventory and refill the pressurizer, no steam void formation occurred in the RCS. The cooldown rate (i.e., RCS coolant inventory shrinkage) was not largo enough to overcome the subcooled state of the RCS coolant inventory or the HPI flowrate.
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 main feedwater early in the transient, which would stop the overcooling of the RCS. Also, 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.
- 3. Conclusions The RCS coolant inventory remained subcooled through the transient, thus ensuring adequate core cooling. The pressurizer emptying was brief (less than 50 seconds) in duration before 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 1514 265 A&B-9 11/79
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APPENDIXES A AND B main feedwater assumed, the steam generators will overfill in approximately 90 seconds.
The reactor coolant pumps running case represents the maximum cooling rate. Therefore, no voiding for this case ensures that the reactor coolaat pump trip case, which would reduce the cooldown rate, also would not produce voids in the reactor coolant system.
B. Accidents
- 1. Scope of Evaluation Maximum overcooling of the RCS results from an uncontrolled blowdown of the secondary plant (i.e.,
SLB accidenu). The double-ended rupture from full power has been demonstrated in the SAR to result in maximum overcooling. Selection of the worst 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
'7 the 33.5-inch inside diameter steam line. The location of the break is outside of the reactor building. This analysis assumed that the safety grade AFW level control system did not function and that overcooling was maximized by continuous AFW injection at its design flowrate. Other system parameters, models, and assumptions are as presented in Section III.
The sequence of events is given in Tables A&B-8 through A&B-15 for each case analyzed. The figures for each case are listed on the table for that case. The figures for Case 1, reactor coolant pumps running, also 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.
The SLB accident was analyzed for 10 minutes assuming no mitigative operator action and only safety grade equipment for transient mitigation. 3c3 9gg The overcooling rate, as anticipated, is much IJl CUV A&B-10 11/79
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APPENDIXES A AND B
/ higher for the c 3 cases than that obtained for the MFW overfeed cast presented in the previous section.
The case resulting in the most severe consequences of RCS shrinkage occurs with LOOP at the time of ESFAS 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 occurs in the hot leg attached to the pressurizer 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 collapsed. Core flow is maintained throughout the transient.
During LOOP cases, natural circulation is maintained by the cooling from the unaffected steam generator side of the RCS.
If no credit is taken for the safety grade AFW level control system, the unaffected steam A&B-ll 11/79
RESPONSE TO 10 CFR 50.54(f)
APPENDIXES A AND B generator fills 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. Taking credit for the safety grade level control system on the unaffected steam generator would allow earlier repressurization of the RCS , thereby leading to earlier collapse of the void.
C. 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 and the core remained subcooled. Some of the specific phencmena noted for the various cases analyzed are as follows:
- 1. The LOOP assumption at ESFAS produces slight.ly worse consequences than at an earlier time. This is because the pumps running maximize the overcooling such that the later the LOOP (up to ESFAS), the more shrinkage that has occurred. LOOP af ter ESFAS should not continue to increase the severity, because isolation of the affected steam generator main feedwater supply occurs at ESFAS and greatly reduces the overcooling rate.
- 2. 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.
- 3. 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 for the HPI failure case.
- 4. The void formation in a given loop was large enough to create temporary flow blockage in tha t loop.
However, the net core flow remains posit'.ve throughout the transient, and is never interrupted to the point that saturation occurs in the core region.
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APPENDIXES A AND B No mitigative operator action was assumed for 10 minutes in the analysis. With the fill rates of auxiliary feedwater assumed, the unaffected steam generator will overfill in 6 to 7 minutes if credit for the safety grade AFW level control system is not taken. 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 in more detail for each transient in Section IV. No 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 cycles are adequate.
Twenty-four different types of transient cycles (several are l1 SAR analyses) are used in evaluating the acceptable number of design cycles. These operating transients are listed in Table A&B-16, along with the number of 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-16 is not meant to be an absolute limit, but was c5neer an ~ the basis ot 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-17 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-16). The operating data are less than the allowable actuation rate for both systems, thereby supporting the adequacy of design.
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APPENDIXES A AND B VI. REFERENCES
- 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 Coolant System, BAW-10128 (August 1976), Babcock & Wilcox, Lynchburg , Virginia
- 2. R. A. Hedrick, J.J. Cudlin, and R.C. Foltz, CRAFT 2-FORTRAN Program for Digital Simulation of a Multinode Reactor Plant During Loss of Coolant, BAW-10092, Revision 2 (April 1975), Babcock & Wilcox, Lynchburg, Virginia 1514 270 A&B-14 11/79
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TABLE A&B-1 MAXI-TRAP fl0DE AND PATH DESCRIPTIO!1 Node Number Description 1 Reactor Vessel Lower Plenum 2 Core, Upper Plenum and Outlet Nozzles 3, 16 Hot Leg Piping 4-13, 17-26 Primary, Steam Generator 14, 27 Cold Leg Piping 15 Reactor Vessel Downcomer 28, 55, 56 Pressurizer 29 Containment 30-39, 40-49 Secondary, Steam Generator 50, 51 Steam Risers -
53, 54, 68, 69 Steam Generator Downcomer 64, 66 Feedwater Piping 63 Turbine and Process Steam Plant 65, 67 Feedwater Piping and Feedwater Heater 52, 57-62 Steam Piping 1514 271 ASB-15 11/79
RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-1 (Cont'd)
Flow Path flumber Description 1 Core 2 Core Bypass 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, ,1- 0 Secondary, Steam Generator 40, 50 Steam Riser 51, 52, 56, 69, 70, 71 26 Inch Steam Piping 53, 58 Aspirator 54, 55, 85, 86 Steam Generator Downcomer 59-62 Presstrizer 63, 66 Feedwate" Pumps 64, 65, 67, 68 Feedwater Piping 72 36 Inch Steam Piping
- 73. 76 MSIV 74, 75, 77, 78 Process Steam and Turbine Piping 79 Feedwater Piping Crossover 80 Steam Piping Crossover -
57, 84 Break 83 Aux. Feedwater 1514 272 81 HPI 82 LPI (Not Used)
A&B-16 11/79
RESPONSE TO 10 CPR 50.54(f)
TABLE A&B-2 MIflI-TRAP fl0DE Atl0 PATH DESCRIPTIO!! l t
Node Number Description 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 Plenum 18-20, 26-28 Secondary, Steam Generator Tube Region 21, 29 Steam Risers 22, 30 Main Steam Piping 23 Turbine 31 Containment Path Number Description 1 Core 2 Core Bypass 3 Upper Plenum, Reactor Vessel 4, 11 Hot Leg Piping 5, 12 Upper St am Generator Shroud 45, 46, 47, 48 Top of Hot Lec Candy Cane" 6, 7, 13, 14 Primary Heat Transfer Region, S.G.
8, 15 RC Pumps 9, 16 Cold Leg Piping 10 Downcomer, Reactor Vessel 17 Pressurizer Surge Line 18, 19, 26, 27 Steam Generator Downcomer 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 34 Steam Crossover 36, 37 HPI 38, 39, 43, 44 AFW 40, 41 Main Feed Pumps 42 LPI ,
49 Stuck Open Relief Valve 35, 50 Leak Paths 1514 273 A&B-17 11/79
_ _ _ _ .. . _ _ _ . _ _ _.____ .. _ _ _ _ ..s . ~ . . . _ _ , .._
i l
1 TABLE A&B-3
SUMMARY
OF EVENTS ANALYZED t .
Initiating Event Single Failure Sensitivity Studies A. Anticipated Event Made More E!
m Severe By Single Failure
{
m Reactor Trip / Turbine Trip Main Feedwater Overfeed m
! E3
>. ~
E5 B. Design Basis Overcooling m Double Ended Steam Line Main Steam Relief Valve Stuck e LOOP at Reactor Trip E!
Break Open e LOOP at Low RC Pressure ESFAS Trip-E e Decay Heat g e HPI Single Failure
{
e Steam Generator Level Control e Break on Different OTSGs D%)
%J ~%J 45.
RESPONSE TO 10 CFR 50.54(f)
_0VERC00LIttG AftALYSIS IflPUT ASSUMPTI0ftS TABLE A&B-4 Parameter 177 FA Power Level 102%
T,yg, OF 579 RCS Operating Pressure (at Pressurizer tap),
psig 2155 Pressurizer Level (indicated), in. 180 RPS Trip S.ignals High Flux, % FP 105.5 Low Pressure (core outlet), psig 1855 ESFAS Trip Setpoints Low RC Press., psig 1500 Low SG Press., psig 585 ESFAS Trip Delay, sec. 2.5 MSIV Closure Time, sec. 5 M WIV Closure Time (linear ramped area), sec. 15 l1 Auxiliary Feedwater Design Capacity ~
Turbine, gpm 885 Motor, gpm 885 Temperature, OF 40 Initiation Time After ESFAS, sec.
With Offsite Power 15 With Loss of Offsite Power 40 Main Feedwater Tempera'ture, F 430 HPI System Design Capacity per Pump, gpm 2 pumps 0 500 each Temperature, F 40 ,
Boron Concentration, ppm 2270 Initiation Time After ESFA5, sec.
With Offsite Power 25 1514 275 With Loss of Offsite Power 30 OTSG Outlet Pressure, psig 910 Revision 1 A&B-19 12/,79
[
i TABLE A&B-5 EQUIPMENT AND RELATED SYSTDIS ASSUMED TO FI'NCTION ESFAS MSLIS MSIV RC FV E'!T TURBINE TURBINE TRIP RPS/CRDCS FOCG AFW MPI LPI CPT 4
- FWIV PUMPS BYPASS i i i
1 1 l'
.I
- I Reactor Trip / Turbine Trip g
l,t ,j g
vich MFU Overfecd X m
y
'8 X X X - -
X X X X l
8 z
'I un M
j Sten:2 Line Break A (Doubic-Ended Rupture) O I. >
ai
- e X X X X - -
X i
(a) - -
o l :ii tob o n
m i I! o X un Denotes system used when needed in the analysis ^
A a -
Denotes system not used in the analysis C i .
5'
.f g
r., (a) ' Loss of offsite power cases assu:ne 4 pump coastdown -
'l
! le LTI 1 -
m a
a I \
i .' M N il* N
RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-6
}!AIN FECY.JATER OVERFEED SEOUENCE OF EVENTS EVENTS TI!!E (SEC,-)
Turbine Trip 0.0 Turbine Stop Valves Close 0.0 Reactor Trip 0.4 Turbine Bypass Valves Open 3.0 Atmospheric Dump Valves Open 4.0
' Pressurizer Empty 170.0
~
Low RC Pressure ESFAS 188.4
}!SIV's Close 200.9 MFWIV's Close 203.4 HPI Actuation 218.4 Pressurizer Starts to Fill 220.0 (Refer Figures A&B-3 to A&B-13) 1514 277 11/79 A&B-21
- - - ~ ~ ~ . - - - -_ _ _ . . _ _ , . . . , _ _ . ,__..,_;.. , , _ _ _ _ _ _ _
t TABLE A&B-7 SLB SEllSITIVITY STUDIES i i
RC pumps LOOP at LOOP at LOOP at ESFAS, ;
Steam line break running reactor trip ESFAS with no c'ecay heat j i
With stuck open relief valve on un- Case 1(*) Case 2 Case 3 Case 4 affected generator, 211PI pumps
[-
available l l
With failure of one llPI pump, no - Case 5 Case 6 ," c stuck open relief valve m
8 z
un t1 j
$ (" .t xi-/ Mini-TRAP camparison presented for this case. )
$ The SLB occurs in the LOOP with the pressurizer.
" n I
" (c)The SLB occure in the opposite LOOP from the pressurizer.
]
o
.8=
m
~
b !
i
$ N x r
$ U t
6
RESPONSE TO 10 CPR 50.54(f)
TABLE A&B-8 DOUBLE ENDED STEMI LINE BREAK CASE 1 - NO LOOP .
SEQ'JENCE OF EVENTS ..
EVENT TIME, s Double Ended Rupture of 33.5" ID Steam Line Between SG and MSIV 0.0 Closure of Turbine Stop Valves 0.00 Reach Low RC Pressure Setpoint 1.5 Control Rod Insertion Starts 2.2 Reach Lev Steam Pressure ESFAS Setpoint 1.9 Low RC Pressure ESFAS . 5.8 MSIV's Closed 9.4 MFWIV's Closed 16.9 .
Unisolated SG Dry Out 20.0 Pressurizer Empty 15.0 Auxiliary Feedvater 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-14 to A&B-25) 0 O
1514 279 e
11/79 A&B-23
"*'O%W4 9 W%h. pp gg 6 m b p-+ e% m me. ,,_ g g ,
RESPONSE TO 10 CFR 50.54 ( f)
, TABLE A&B-9
~
DOUBLE ENDED STEAM LINE BREAK
. CASE 2 - LOOP AT TRIP .
SEQUENCE OF EVENTS EVENT . 'hIME, s Double Ended Rupture of 33.5" ID Sten:: Line Between SG and MSIV 0.0 Closure of Turbine Stop Valves 0.00 Reach Low RC Pressure Setpoint + LOOP Initiation 1.5 Control Rod Insertion Starts 2.2 Reach Low Steam Pressure ESFAS Setpoint 1.8 Low RC Pressure ESFAS 7.0 .
MSIV's Closed 9.3 MFWIV's Closed 16.S Pressurizer Empty , 18.0 Unisolated SG Dry Out 20.0 HPI Injection Starts 37.0 Auxiliary Feedwater Initiation 59.4 Pressurfzcr Starts to Fill Up . 60.0 .
SG Tube Region Full of Liquid 380.0 (Refer Figures A&B-26 to A&B-34) 1514 280
.- 11/79 A&B-24 9 6.. h e- phaeMymW P* 4 eg , e ger eg o e *** * * * * * * * * *
- RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-10 DOUBLE ENDED STEAM LINE BRE/X CASE 3 - LOOP AT ESFAS SEQUENCE OF EVENTS -
. EVENT TIME, 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 Low RC Pressure ESFAS + LOOP Event Initiation 5.8 MSIV's Closed 9.4 MFWIV's Closed 16.9 Pressurizer D::pty 17.0 Unisolated SG Dry Out 18.0 ,
HPI Injection Starts 35.8 ,
Auxiliary Feedwater Initiation 55.5
~Pressuriser Starts to Fill Up f30.0 SG Tube Region Full of Liquid 380.0 (Refer Figures A&B-35 to A&B-43)
S
~
1514 281 O
e A&B-25 -
RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-ll DOUBLE ENDED STEMI LINE EREAK i CASE 4 - LOOP AT ESFAS, NO DECAY llEAT SEQUENCE OF EVENTS EVENT - TIME, s Double Ended Rupture of 33.5" lD 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 Low RC Pressure ESFAS + LOOP Event Initiation 5.8 MSIV's Closed 9.4 ,
MFWIV's Closed 16.9 ,
Pressurizer Empty 17.0 Unisolated SG Dry Out 20.0 PPI Injectica Starts -
35.8 Auxiliary Feedwater Initiation to Good SG 55.5 Pressurizer Starts to Fill Up 330.0 SG Tube Region Full of Liquid 380.0 .
(Refer Figures A&B-44 to A&B-52) _
1514 282 A&B-26 Il/79
RESPONSE TO 10 CFR 50.54(f)
GN4BLE A&B-12 DOUBLE ENDED STEMI LINE BREAK CASE 5 - LOOP AT TRIP, HPI FAILURE, NO STUCK RELIEF VALVE SEQUENCE OF EVENTS WET TE 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 Sctpoint + LOOP Event initiation 1.5 Control Rod Insertion Starts 2...
Reach Low Steam Pressure ESFAS Setpoint 1.8 I Low RC Pressure ESFAS 7.0 MSIV's Closed 9.3 IDPRIV's Closed 16.8 Pressuriser Empty 18.0 Unisolated SG Dry out 22.0 HPI Injection Starts 37.0 Auxiliary Feedwater 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-53 to A&B-61) e 1514 283 A&B-27 11/79
RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-13 DOULLE ENDED STEAM LINE BREAK CASE 6 - LOOP AT ESFAS,
, HPI FAILURE,
. NO STUCK OPEN RELIEF VALVE EVENT
' TIME, 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 Reach Low Steam Pressure ESFAS Setpoint 1.9 Low RC Pressure ESFAS + LOOP Event Initiation 5.8 MSIV's Closed 9,4 !
MFk'IV's Closed 16.9 Pressurizer Empty 17.0 Unisolated SG Dry Out 20.0 llPI Ihjection Starts 35.8 Auxiliary Feedwater Initiation to Good SG 54.5 SG Tube Region Full of Liquid 320.0 Pressurizer Fill Up Starts 450.0 (Refer Figures A&B-62 to A&B-70) d
. 1514 284
~
11/79 A&B-28
, , , % ,,v - m ,_ * * * *
- * * * * ' " * " * * " " ^ ~ "~^# '
RESPONSE TO 10 CFR 50.54(f) ,
Tm '.F A&B-14 DOUBLE EIGED STEAM LINE BREAK CASE.7 - LOOP AT ESFAS, HPI FAILURE, NO DECAY HEAT, NO STUCK OPEN RELIEF VALVE
_ EVENT TIME, s l
Double Ended Rupture of 33.5" ID Steam l
. 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 Low RC Pressure ESFAS + LOOP Event Initiation 5.8 MSIV's Closed 9.4 - I MFWIV's Closed 16.9 n
Pressurizer Empty 17.0 Unisolated SG Dry Out 20.0 HPI Injection Starts 35.8 Auxiliary Feedvater Initiation to Good SG 54.5 SG Tuin Region Full of Liquid 290.0 (ReferFiguresA&B-71toA&B-79) 1514 2B5 O
A&B-29 III '
. RESPONSE TO 10 CFR 50.54(f)
TABLE A&B-15 DOUBLE ENDED STEAM LINE BREAK CASE 8 - SLB ON OPPOSITE LOOP FROM PRESSURIZER, LOOP AT ESFAS, HPI FAILURE, NO STUCK OPEN RELIEF t LVES EVENT TIME, 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 Lou Steam Pressure ESFAS Setpoint 1.9 Low RC Pressure ESFAS + LOOP Event Initiation 5.8 .
MSIV's Closed 9.4 MFWIV's Closed 16.9 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 (Refer Figures A&B-80 to A&B-88)
[514286' V
A&B-30 .
$$j79
., -. . . , . . - _ ~ . . . . . . , . - - , . - _ - . . _ . . . ~ _ _ . . . . --- - - . . . . _
1 , .
RESPONSE TO 10 CFR 50.54(f)
TA3LE ASB-16 Operating Transient Cveles Transient . Design Numbcr Transient Description Cycles 1A Heatup from 70*F to 8% Full Power (Normal) 240 1B Cooldown from 8% Full Power (Normal) 240 2 Power change 0 to 15% and 15 to 0% (Normal) 1440 3 Power Loading 8% to 100% Power (Normal) 48,000 4 Power Unloading 100% to 8% power (Normal) 48,000 5 10% Step Load Increase (Normal) 8,000 6 10% Step Load Decrease (Normal) 8,000 7 Step Load Reduction (100% to 8% Power) (Ups et)
Resulting from turbine trip 160 Resulting from electrical load rejection 150 Total 310 8 Reactor Trip (Upset) 1 Type A 40 Type B 160 Type C 88 Trips included in transient numbers 11, 15,16,17, & 21 112 Total 400 9 Rapid Depressurization (Upset) 40 10 Change of Flow (Upset) 20 11 Rod Withdrawal Accident (Upset) '
40 12 Hydrotests (Test) 20 13 Steady-State Power Variations (Normal) =
- 14 Control Rod Drop (Upset) 40 15 Loss of Station Power (Upset) 40 16 Steam Line Failure (Faulted) 1 17A Loss of Feedwater to One Steam Generator (Upset) 20 17B Stuck Open Turbine 3ypass-Valve (Emergency) 10 1514 287 Revision 1 A&B-31 12/79
RESPONSE TO 10 CFR 50.54(f)
TABLE ASB-16 (Cont'd) Design Number Transient Description Cycles 18 Loss of Feedwater Heater (Upset 40 19 Feed and Bleed Operations (Normal) 40,000 20 Miscellaneous A (Normal) 30,000 Miscellaneous B 20,000 Miscelleneous C 4 x 10 6 21 Loss of Coolant (Faulted) 1 22 Test Transients - High Pressure Injection System (Test) 40 Core Flooding Check Valve 240 23 Steam Generator Filling, Draining, Flushing and Cleaning (Normal)
Steam Generator Secondary Side Filling Condition 1 120 Condition 2 120 Steam Generator Primary Side Filling Condition 1 120 Condition 2 120 Flushing 40 Chemical Cleaning 20 Total 540 24 Ilot Functional Testing (Test) 1 1514 288 Revision 1 12/79 A&B-32 .
. RESPONSE TO 10 CFR 50.54(f) i TABLE A&B-17 RPS/ESFAS FREOUENCY Actual Data Allowed Nun:ber Requency Frequency No. of Reactor Trips (RPS) 310 8.9/yr 10/yr l 1 No. of Automatic ESFAS Actuations 27 .816/yr 1.0/yr No. of Plants Included 9 Approx. 35 1 Reactor Years 1514 289 Revision 1 A&B-33 12/79
RESPONSE TO 10 CFR 50.54(f) 63 o o o
82 -
gg n o
@ T i
j ag u _ _ @_ _ _ _ ,,
g o 57 i at
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RESPONSE TO 10 CFR 50.54(f) .
- 23
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g 38 I~~1 54
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e -
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1514 291 0 D p ?n 'r EINI TIAF N0 DING SQlEEE, 177 FA
'> #JJ ((d g ASB-35 Reviston 1 FCGURE ASB-2 12/79
RESPONSE TO 10 CFR~50.54(f)
FIGURE ASB-3 MFW OVERFEED, TURBINE TRIP, REACTOR TRIP-177 FA CORE AVERAGE TEMPERATURE VS TIME FOR M/XI-MINI TRAP COMPARISDN I !
580 l-I l
i 576 -
i 572 -
p 568 -
ru mau Y
- 564 -
li 3 560 -
g 2 N 5 N g
556 -
N
\
\ MINI-TRAP MODEL 552 -
\
MAXI-1 RAP MODEL \
548 -
N N
N N
N 544 -
N I I I I I I I 540 O 20 40 60 80 100 120 140 160 Time, see ASB-36 Revision 1 1514 < 92 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-4 NFW OVERFEED, TURBINE TRIP, REACTOR TRIP-177 FA CORE OUTLET PRESSURE VS TIME FOR MAXI-MINI TRAP COMPARISON 2200 1
1 2150 1 1
1 2'aa i
i L9mm39 lmq;IL m
x 1
, 2050 -
5 4
- \
~
\
y 2000 -
5 5
j 1950 -
5 I g
\
1900 -
x N
N MINI-TRAP MODEL 1850 - N MAXI-TRAP MODEL N
\
\
1800 - \
s N
N N
1750 - N N
N N
1700 I I I I I f 0 20 40 60 80 100 120 140 Time, sec Revision 1 ASB-37 12/79 1514 293
RESPONSE 'IO 10 CFR 50.54( f)
FIGURE A&B-5 NFW OVERFEEO, TURBiHE TRIP, REACTOR TRIP-177 FA PRESSURIZER LEVEL VS ilME FOR MAXI-MINI TRAP COMPARISON 20
[k
\
\
18 g
\
\
\
14 -
g
- \
_- \
i 12 \
\
b \ MINI-TRAP MODEL 10 -
- N 2 N N
N B -
N
\
\
N 6 - \
MAXI-TRAP MODEL s
s N
N 4 N 2
7-7 - - - , y 7 ,
0 20 40 60 80 100 120 140 160 Time, see ASB-38 Redsion 1 12/79 1614 7ad
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-6 MFW OVERFEEO, TURBINE TRIP, REACTOR TRIP 177 FA RCS TEMPERATURE VS TIME 600 -
- ~ ~ -
590 -
580 -
570 -
o O
d ft 560
[\ \
" I .\
~
\
$ 550 - \
\
\
\
540 - \
\
\
\
530
\ N
%-~ ~
520 51 0 1514 295 500 -
r~-- I 'r -~ !- - r O 100 200 300 400 500 600 ACB-39 Time, sec Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-7 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP- 177 FA STEAN GENERATOR LEVEL VS TIME FULL __.
50 45 -
/
40 -
/
/
r /
. 35 -
E a
w
$ 30 -
5 STEAN GENERATOR B a STEAN GENERATOR A y 25 -
/
/
/
/
20 -
/
/
/
15 /
' ~
bk Ok3hd 5
1514 296 I I 0
O 50 100 see Revision 1 Time' A&B- 40 12/79
I FIGURE A&B-8 MFW OVERFEEO, TURBINE TRIP, REACTOR TRIP- 177 FA PRESSURIZER LEVEL VS TIME 38.74 l
33.20 27.67 -
g r 1 %
t e
m 5 5 22.14 4 T, - \
8
" N H
-- O y 18.60 , Q e w
' m m
11.07 i t O
UE 5.53 h5 e ta -
E cn o -
A I I '
- 0.00 I I I N 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50 0 m
N Time, sec X 10
FIGURE ASB-9 NFW OVERFEEO, TURBitiE TRIP, REACTOR TRIP- 177 FA SG A SECONDARY OUTLET PRESSURE VS TIME 12.0
@ 11.5 N
.5 to E o 11.0 z M $
m g k E h
[ 10.5 - 5 5 N O o _
a 0 10.0 g ,
E ^
5 9.5 -
UE E D N.
em-- 3 H- Q 0- #
9.0 I I e i 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 N
e Time, sec X 10 co
FIGURE ASB-10 NFW DVERFEEO, TURBINE TRIP, REACTOR TRIP- 177 FA SG B SECONDARY OUTLET PRESSURE VS TIME 12.0
@ 11.5 5
- m
- o'u a
- II.0 $
> m E E g i
- 10.5 5
= n E D E m m
$ I0.0 -
E $'
m m
9.5 -
M Ci? q M.
' $. - 9.0 l i I I I I I I o
" tn
- 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0
- Time, sec X 10 w
W W
PIGURE ASB-11 NFE OVERFEEO, TURBINE TRIP, REACTOR TRIP- 177 FA CORE OUTLET PRESSURE VS TIME 24 22
= %
5 o z
. 20 N 8
Y a <
t lis I i
j m
m 16 ,
o a
gx 14 -
50 a tt
>d-0~ 12 I I I i t l t
{
H
- 0. 0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 A
Time, sec X 10 u
O O
FIGURE AEB-12 NFW OVERFEEO, TURBINE TRIP, REACTOR TRIP CORE AVERAGE' TEMPERATURE VS TIME 600 590 -
580 f t E m
.T o a 5 g 570 -
n E b
$ ~560 -
o na 2 n m
E
- 2 550 -
E E
2 540 -
d UN D $. 530 -
- .g -
w ~
g ~ 520 i i i i i i l n e 0 100 200 300 400 500 Time, $se
PIGURE A&B-13 NFW OVERFEED. TURBINE TRIP, REACTOR 1 RIP- 177 FA TOTAL POWER VS TIME I.2 1.0 3 g
0 en
, .0 -
% E b
= _ -
E
- .6 Q w
.4 -
I C
g, .2 -
40 EE k 0
0,0 1 I - , ,
un "
- 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0
- Time, sec X 10 u
CD N
RESPONSE 'IO 10 CFR 50.54(1 :
FIGURE A&B-14 SLB, CASE 1, 177 FA 2200 2100 2000 1900 - -
L 1800 l l
1700 I E i f0 1600 * -
E 1500 -
a 1400 -
. , 1 o
1300 -
1200 - I i
1100
\
\
i 1000 NAXI-TRAP MODEL g
\
900 \
~
\ MINI-TRAP MODEL 800 -
\
~
l % % ^ N ^g p f 700 s,
w 600 0 10 20 30 40 50 60 70 80 90 100 Time, sec ,
ASB-47 Revision 1 1514 303 12/79
RESPONSE TO 10 CFR 50.54(f)
FIC,URE ASB-15 stb, CASE 1, 177 FA 580 570 -1 1
560 -
\
550 -
m 540 - $
E \
{ 530 -
}
k g 520 - NAXI-TRAF MODEL g
=
\
5 510 g ulNI-TRAP 5
" \ %-
500 -
g / ,,'~~~,
490' -
480 470 460 ,
_ p- - p - - _
I~
0 20 30 40 50 60 70 80 90 100 1.0 Time, sec ASB-48 Revision 1 12/79 1514 304
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-16 SLB CASE 1, 177 FA 20 i
18 -
16 14 12 -
=
, 10 -
=
5 r
8 -
6 -
g
\ NAXI-TRAP N00EL
\
4 -
g MINI-TRAP N00E \
2 -
\
\
, \ - ,. _ _.
0 10 20 Time, Sec Revision 1 ASB-49 12/79 1514 305
RESPONSE TO 10 CFR 50.54(f)
=
~
e
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W r==
Pan e==
e 3
=
CD
=
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7 E m
A W
< c: o C
C cs" H
fL o
N
_ o g -, - - - _- g C W W W. g
- . . N. .
samod 1E101 Revision 1 ASB-50 12/79 1514 306
RESPONSE TO 10 CFR 50.54(f)
_ 8
=
f E
n O
J -
8 w
E o
- a O
ro N
m k -N e--
l ca e
M E
8 a
t H
p.
8 o a o a E
m E .E., 3 alsd 'aJnssaJd taltnD a.'og Revision 1 12/79 A&B-51 -
1514 307
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-19 SLB, CASE 1, 177 FA 600 550 - -
i
!! I
\
i.
.s 1 3
1 500 '.
g HOT LEG
\
\
\
\
COLO LEG \
450 -
g
\
\
N
, N N
N N
N 4GO O 100 200 300 400 500 600 Time, set.
Revision 1 ASB-52 12/79 1514 308
RESPONSE'TO 10 CFR 50.54(f)
FIGURE ASB-20 SLS, CASE 1, 177 FA 33.2 27.67 -
r 22.14 -
h b '
16.60 -,
5 U
t 11.07 5.53 -
A i _ _ . _
100 ,
200 300 400 500 600 T!ae, sec 1514 309 Revision 1 ASB-53 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-21 SLB, CASE 1, 177 FA 1200 1000 -
.2 E.
J
= 800 _
=
2 600 2 E
o ,
M ' .
< l
= 400 -
200 b
- + - -
0 l 0 100 200 300 400 500 600 Time, see Revis 4 310 ASB-54 12/79
RESPONSE 20 10 CFR 50.54(f)
FIGURE 75B-22 gtg, gggg ;, y77 pA 800 .-
I 700 --
=ca.
.i E
p 600 - -
=
5 O
g 500 2
w
=
M 400 300 1
r ---- -
200 r- r -- -
0 100 200 300 400 500 800 Time, see 1514 311 Revision 1 ASB-55 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-23 SLB, CASE 1, 177 FA FULL 50 M 40 E
m j 30 -
5 E
5 NOTE: SG "A" BOILS ORY WITHIN 20.0 SECONOS 20 -
10 -
O r I r
,}14 s
j}2 0 100 200 300 400 i Revision 1 AEB-Sge, sec 12/79
RESPONSE 'IO 10 CFR 50. 54 ( f )
FIGUPI ASB-24 SLB, CASE 1, 177 FA 350 l
300 4 l
!l11 1
- 1 11
,l S
CANDY CANE "A"
"?
- I i g'-
-l \
{ 200 w
-l lA
- o. l/ \
h ! \
- l \
\
\
\
l i I \
100 4 \
\
l g l
s
\
l \.
N
\
HOT LEG "A" \
N___. ., ,
0 100 200 300 Time, sec 1514 31.3 Revision 1 ASB-57 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-25 stb, CASE 1, 177 FA 80 -
1 il 70 g
Il II 60 - ll il CAN0Y CANE "B"
!l 50 II
%~ 'll g 11 aa 11 m 40 g g-5 II *
.,I 30
,i iI
'l I i n ; i
/t I\
l i f \
l I i \
'" 1 i i g
\
i i [
g I
I I
L_._J I
(- ---
' 7% 3 _. -
/N N-L ,
C 50 100 150 Time, see 1514 314 Revision 1 ASB-58 12/79
H C D 0 m < ,. x , n d 3= =e R
E S
P A O S N B J S
- E 5 a 9 n o . I' d O 1
2 1 1 0 0
1 C F
R
? 5 0
5 4
(
f
)
1R 2e
/v 7i 9s i 1 o
n 5 1
1 O r -
, i i-4
. E =c . e= == .=
3 1
E. ::o 5 .
RESPONSE TO 10 CFR 50.54(f)
=
E o
5 -
3 0
d W
< =
C3 -
O d
d E r- .i m a k
o C
H
=
2 o
r o e o o a 8 E
3 3
2 8
E alsd 'ajnssajd taltn0 ajo3 1514 31o, Revision 1 12/79 ASB-60
RESPONSE TO 10 CFR 50.54(f)
FIGURE A&B-28 SLB, CASE 2, 177 FA 600 550
.- I
~
a l HOT LEG 3 k U
= 500 1 f I / s s/ \
\
\
\
\ COLO LEG
\
450 -
N
\
N
\
g -
I N
N
- N l N
' s s
N N
400 ' -
0 100 200 300 400 500 600 Time, sec 1514 317
~
Revision 1 12/79 AGB-61
FIGURE A&B-29 SLB, CASE 2, 177 FA 22.67 22.14 -
a
- 5
" m
~ o
> , 18.60 5 m
E 5 i .E s
u ,
" ~
o
.2 g
= 11.07 m a
N
."_ ' w
- O N
.Ds
^
5.53 D HM N (D
\<
-J >*-
@ (n 6 - i-- i.._ ,..._
o en 0 -
0 10p 200 300 400 500
$- 600
" ilme, sec
\
FIGURE ASB-30 SLB, CASE 2, 177 FA I
1000 '
- 800 E .
. m E o
> R =
., a m E 600 "
if m '
" g a
= w w
3" o E
=
a 400 - w (D O e
N N a
64 O w
200
~ :o -
~
-4 >*-
- f. A L . . . . .
o 0
[ _"
m 0 100 200 300 400 500 600 Time, sec
RESPONSE TO 10 CFR 50.54(f)
=
5 o
2 3 0
d 3
- - o u =
- E
=
- E E
. m i _ 8~
se l0 o
C E -
E E
7 ,
o g
=
=
g
=
=
a a
sisd 'ajnsssJd iia,, Josejauas meats 1514 320 Revision 1 12/79 ASB-64
RESPONSE TO 10 CFR 50.54(f)
FIGURE A&B-32 SLB, CASE 2, 177 FA 50 -
40 -
- STEAM GENERATOR "B" 2
30 2
l" E
v, 20 -
NOTE: S.G. "A" BOILS.ORY WITHIN 20.0 SECONOS 10 -
1514 321 r r- - -
0 100 200 30(, 400 Time, SeC Revision 1 ASB-65 12/79
RESPONSE TO 10 CFR 50.54(f)
O FIGURE A&B-33 SLB, CASE 2, 177 FA 250 _
,s LOOP "A"
. I s g \
.l I CANDY CANE g
' \
200 -
I g
= \
\
i i I
= \
I ,
= 150 es -l 1 5 i g 5
'I \
I \
100 I \
t \
I I UPPER S.G. TUBCS 50
! 1 l HOT LEG i 4 . s I \
,/ _ . r- k - -b - i 0 25 50 75 100 Time, see 1514 322 Revision 1 12/79 ASB-66
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-34 SLB, CASE 2, 177 FA 40 ,
LOOP "3"
/
30 -
CAN0Y CANE e-s 20 -
U a
10 -
\5\4 23 g ---------~T~~~
~~~ '
~T'---
r 0 5 .
10 15 Time, see Revision 1 ASB-67 12/79
RESPONSE 'IO 10 CFR 50. 5 4 ( f )
FIGURE ASB-35 SLB, CASE 3,177 FA 1.2 1.0 -
t h
- 0. 8 5.
0.6 ' -
B 3
0.4 - -
0.2 -
0 0 100 200 300 400 500 600 Time, sec 1514 324 Revision 1 A&B-68 12/79
RESPONSE F 10 CFR 50.54(f)
FIGURE ASB-36 SLB, CASE 3, 177 FA 2000 .J 1800 -
n a
f 1600
=
o 5 1400 3
1200 1000
,,, 7._
0 100 200 300 400 500 600 Time, see 1514 325 Revision 1 12/79 ASB-69 F
RESPONSE TO 10 CFR 50.54(f)
FIGUR2 ASB-37 SLB CASE 3, 177 FA 600 550 - -
1 I
I I
= k HOT LEG i \
s .
1 5 sa 2 g/ n \
1/ \
\
\
\
\
\
COLE LEG \
\
\
N N
N N
N N
N N
400 q g <
N -
\
,- s 0 100 200 300 400 500 600 15.14 e sm 1 326 ASB-70 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-38 SLB, CASE 3, 177 FA 33.2 t
27.67 -
i I ,
I 22.14 J
= .
9 3
g 16.80 -
5 e
- 11. 07 I
5.53 -
0 :_
0 100 200 300 400 500 600 Time, see 1514 327 Revision 1 12/79 AEB-71
RESPONSE 'IO 10 CFR 50. 54 ( f )
FIGURE ASB-39 SLB, CASE 3, 177 FA 1200 -
1000 -
.2 8.
=
3 800 '-
e 8
0 600 -
E o
5 E 400 -
200 .
' ~ ^ ' ' ~ - - - -
0 0 100 200 300 400 500 600 Time, see 1514 328 Revision 1 "A&B-72
RESPONSE 'Io 10 CFR 50.54(f)
FIGURE ASB-40 St.B, CASE 3, 177 FA 1200 1000 -
I C3.
. 800 5
E 600 -
a r.
C 3
a 400 -
200 -
0 0 100 200 300 W W
- Time, see 1514 329 Revision 1 12/79 A&B-73
RESPONSE 'IO 10 CFR 50.54(f)
FZe"n : A&B-41 SLB, CASE 3, 177 FA FULL 50 .
i 40 i
E 2
m ,
a 30 e
o 3
m 20 t
4 NOTE: SG "A" BOILS ORY WITHIN 10 20.0 SECONOS l
0 t-- t r r --- - r 50 100 150 200 250 300 50 0
Time, see 1 4 330 0 Revision 1 ASB-74 12/79
RESPONSE TO 10 CPR 50.54(f)
FIGURE A&B-42 SLB CASE 3, 177 FA 300 -
f it l ,q\
250 - l \
g I s I \
" l 200 I [ CANDY CANE "A" i i i 3 I I ,
!u 150 -I ,
j l I
\
100 -l \
SG UPPER TUBE REGION, l \
SG "A" HOT LEG "A" g Il . s l
iI 50 .s l
s_, -- s i
N
, , /
/
yx rs _ _
g 50 100 150 Time, sac 1514 33I Revision 1 12/79 ASB-75
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-43 SLB, CASE 3, 177 FA 50 A 40 -
CANDY CANE E "B" j 30
=
20 10
, , , 1514 332 0 5 to 15 Timegg_rguoture.Sec 7 , Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-44 SLB, CASE 4, 377 pg 1.0 1
0.8 - -
o.
- 0. 6 - a
- i
,e 0.4 - -
0.2 .
0 100 200 300 400 500 Time, sec 1514 333 Revision 1 ,
ASB-77 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-45 SLB, CASE 4, 177 FA 1800 -
1600 - -
.5 E.
5 1400 5
e j 1200
=
3 1000 800 600 r- -
T i 100 200 300 400 500 600 0
Time, see 1514 334 Revir, ion 1 A&B-78' 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-46 SLB, CASE 4, 177 FA 600 --
t l
l 525'y
\
I HOT LEG
." t/N \
g V
\
3 \
a .\
450 - \
N COLO LEG \
\
N l %
N N
N N
375 -; \
- s%
l 's N
~
300 '
100 200 300 400 500 600 Time, see 1514 335 Revision 1 ASB-79 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB- 47 SLB, CASE 4, 177 FA 33.20 27.67 22.14 -
=
=
B 16.60 --
g.
e a.
11.07 5.53 .
I -
-~iA - .m. ._
0 109 200 300 400 500 600 Time, sec 1514 336 Revision 1 12/19 ASB-80
RESPONSE TO 10 CFR 50.54(f)
FIGUPE ASB-48 SLB, CASE 4, 177 FA 1000 -
- 800 ' -
=
I (
5 '
f 2
600 E
o
.?.
400 --
]
E w
b E 200 O , .
G 100 200 300 400 500 600 Time, Sec Revision 1 12/79
^'"-*'
1514 337
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-49 SLB, CASE 4, 177 FA 1000
. 800 ..
E 600
} -
8 y !00 -
C a
l-
= 200 _
0 I I I I I O 100 200 300 400 500 600 Time, Sec Revision 1 12/79
^5"-82 1514 338
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-50 SLB, CASE 4, 177 FA FULL 50 -
40 O
~
ll; 3 30 -
E
- a g STEAM GENERATOR "B" 2
w 20 -
NOTE: SG "A" BOILS ORY WITHIN 20.0 SECONOS 10 1514 339 O_ l i I I ! l 1 0 50 100 150 200 250 300 350 400 Time, Sec Revision 1 A&B-83 12/79
RESPONSE 'IO 10 CFR 50. 54 ( f )
FIGURE A&B-51 SLB, CASE 4, 177 FA
\
300 -
\1 l
270 l 1 i i 240 1
l CANDY CANE 210 I ,
m O 180 l
, LOOP "A" E 1 5 i S 150 , 1 l 1 i
120 - -
1
! l
' T\ /
/ N
\
\\ '
/ \
60
- \, UPPER
/ SGTU2E\ HOT LLL I
- REGION 30 . -
\
0 1 I \I ' '
O 50 100- 150 200 250 300 350 Time, Sec 1514 340 ,,,1,10, 1 ASB-8 t; 12/79
RESPONSE TO 10 CFR 50.54(f)
FIGU.tE A&B-52 SLB, CASE 4, 177 FA 63 - - - -
/~T
/ \
54 f CANDY CANE "B"
/ \
l I \
45 -
j l
" \
= I r \
p 36 I E i (
i \
- I 27
\
i l \m 18 l
l i
2 _ i I
I \
t i I I I .l I l I I I I I I . . _ _ _ _ _
Pj 0 1 2 3 4 5 6 7 8 9 10 11 12 Time, Sec
~
Revision 1 ASB-85 '" 1514 341
RESPONSE TO 10 CFR 50.54(f)
FIGURE A&B-53 SLB, CASE 5, 177 FA
~
1.2 1.0 -
- 0. 8 0.6 - -
2 0.4 --
0.2 ..
0.0 t i .
0 100 200 300 400 500 600
!! rwa, Sec Revision 1 12/79 ASB-86 1514 342
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-54 SLB, CASE 5, 177 FA 2200 p 2000 .
1800
.5 E.
s' a
y 1600 ..
E 8
e 1400 0
1200 _
1000 I I I I I 800 . _ . _
0 100 200 300 400 500 600 Time, Sec Revision 1 12/79 A&B-87 1514 343
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-55 SLB, CASE 5, 177 FA 609 550 1
( HOT LEG B
5 E
A g 500 i
/
/%
'l N
} COLD LEG N
N N
450 -
\
N N
N N
N N N 400 I I I I I I I I I I ~
0 100 200 300 400 500 600 Time, Sec Revision 1 A&B-88 1514 344
RESPONSE TO 10 CFR 50.54(f)
~
FIGURE ASB-56 SLB CASE 5, 177 FA 27.67 - --
22.44
- 16.60 ..
=
2 E
3 11.07 E
5.53 g _
_7 y _i .
0 100 200 300 400 500 600 Time, Sec Revision 1 12/79 ASB-89 , -
1514 343
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-57 SLB CASE 5,177 FA 1200 1000 -
2 E
E 3 800 E
a r 600 -
E a
m 400 .. ,
a 200 -
0 .
0 100 ~ 200 300 400 500 600 Time, Sec Revision 1 12/79 ASB-90 1R1/! Tan
s RESPONSE TO 10 CFR 50.54(f)
! i FIGURE ASB-58 SLB CASE 5, 177 FA 1400 . . _ _ . ..
1200 -
=
M !
5 1000 - .
8 1
E a.
as W
800 O ,
5 i
- i h 600 ;
3 400 200 I I f I i ,
0 100 290 300 400 500 600 Time, Sec - - -
J Revision 1 12/79 ASB-91 lbl4 347 [
,s, .-
? .
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-59 SLB CASE 5,177 FA FULL 50 -
STEAM GENERATOR B 40
}* 30 a
5 b 20 - NOTE: SG A B0llS DRY WITHIN 22.0 SECONOS 10 0 I I I I I l 0 50 100 150 200 250 300 Time, Sec Revision 1 ASB-92 12 M 1514 348
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-60 SL,8 CASE 5, 177 FA 264 252 240 - 1 I
228
-l L I
216 L
204 192 -
180 168 "O 156
. CANDY CANE "A"
$ 140 o 1
> 132 l
=
5 120 t
100 i
96 - i i
84 -
-
- UPPER TUBE REGION, SG A 72 g i
60 -
l 48 . . ,'
I' \
- HOT LEG "A" 36 i
.h .i ,
24 - .
!,1 12 f i O
t I I I [I O 100 230 300 400
. Time, see Revision 1 ASB-93 12/79 1514 349
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-61 NO V010 LNG IN LOOP B Revision 1 12/79 ASB-94 1514 3IP30
RESPONSE TO 10 CFR 50.54(f)
FIGUPE ASB-62 SLB CASE 6, 177 FA 1.0 m 0.8
- 0. 6 --
E -
2 30.4 - .
0.2 .
- 0. 0 0 100 200 300 400 500 600 Time, Sec Revision 1 12/79 A&B-95 1514 351
. RESPONSE TO 10 CFR 50.54(f) mm sB-63 SLB CASE 6,177 FA
% 2190 2000 1800 3 1600 - -
5 5
I 1400 -
5
=
1200 1000 -
800 0 100 200 300 400 500 600 Time, Sec Revision 1 12/79 ASB-96 1514 352
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-64 SLB CASE 6, 177 FA 600 550 ' -
k HOT LEG I
e .
I
~
5~ l N
'00 V j/ .
\
g y \
\
U
" \ COLO LEG N
\
N 450 N
N N
N N
N N
N w*%
400 l I I I I O 100 200 300 400 500 600 Time, Sec Revision 1 ASB-97 1514 353
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-65 SLB CASE 6, 177 FA 22.14
= 16.60 . -
i 5 I 2 11.07 _ _
Z 5
2 5.53 .
_. e I -
O 100 200 300 400 500 600 Time, ;ec
\b\4 ~b54 Revision 1 12/79 ASB-98
RESPONSE TO 10 CFR 50.54lf)
FIGURE ASB-66 SLB CASE 6, 177 FA 1000
.5 E 800 -
a E
5 a
, 600 -
2 o
z 3
8 400 - -
3 m
=
M 200 f
t i i t i i O , , , ,
0 100 200 300 400 500 600 Time, Sec Revision 1 12/79 AEB-99 1514 355
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-67 SLB CASE 6,177 FA
. 1000 -
O a
a a
j 800 -
5 m
3 600 --
5 vs h'
ui 400 -
vi 200 - I I l l l 0 100 200 300 400 500 600
. Time, Sec 1514 356 Revision 1 12/79 ASB-100
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-68 SLB CASE 6, 177 FA FULL 50 -
45 40 0 35 --
30 -
.a STEAN GENERATOR "B" 25 -
w 20 -
NOTE: SG "A" BOILS ORY WITHIN 20.0 SECONOS 15 -
10 - -
5 -
0 I I I I I I _
1514 357 0 50 100 150 200 250 300 350 Time, Sec Revision 1 ASB-101 12/79
RESPONSE TO 10 CFR 50.54(f)
FIF,URE ASB-69 SLB CASE 6, 177 FA 300 O
11
!D LOOP "A" I \
250
-l g I \
l \
I
, _l CANDY CANE R l n I \l i i I
~
g 150 I \
- 1 \
I I 100 l $
l I
\
SG UPPER l HOT LEG
- TUBE REGION
\
I I l \' e s%
0 k ! Ew 0 50 100 Time, Sec Revision 1 12/79 ASB-102
RESPONSE 'IO 10 CFR 50.54 ( f)
FIGURE AF,B-70 SLB CASE 6, 177 FA 50 L
1 40 -
CANDY CANE "B" l
3G E
3 3
2 3
20 -
10 a l I 5 1 15 Time, Sec Revision 1 ] } j f } } g ASB-103 12/79
RESPONSE 'IO 10 CFR 50.54 ( f)
FIGUDI ASB-71 SLB CASE 7, 177 FA 1.2
- 1. 0 .
0.8 - -
E 0.6 - -
0.4 .
0.2 - -
i !
- 0. 0 I I O 100 200 300 400 500 Time, Sec Revision J.(:, 4 ,.g 12/79 s ASB-104
4 4 s
1'
- +e+% , eE Ev <e 1,.
$A %4 4
TEST TARGET (MT-3)
I'O '" EA D22 EE4 625 m
l-l
! ,,, 5 IlllE e.=
1.25 IA 1.6 4% ' '#
+ 'b t
- S$e#N/
ffff b<;%g I
I'
4 9'?b't>, //h('4
%4
- ++ ,_E Ev <e -
TEST TARGET (MT-3) 4 1.0
[l9EMEM
@ NE I.I [ 52 Ol!M l.8 1.25 1.4 l 1.6
- 4 + 4%
- ?f$?
5, 7 .
'hQ b
+y:sl 1
1 7
A+f>+ n'*
,. ee . <e.1,.
TEST TARGET (MT-3)
Ao 1.0 lj IE BB ll9E4 a L" EB l.8 1.25 1.4 1.6
=
< 6" -
4% + ss
I8h I 3(ff!;
p .~ Y
4 4
- +e)<%> 4 _e . .._1,.
$<s&
T.ST TARG.T (MT-3) 10 '"Du A 6$
El n2 gm 1,l
['u OM
.8
'j'l.25 l.6 l1.4 j h %7 $ /4%
$8 4 fax)/
// Q
% I
. 4<v
RESPONSE TO 10 CFR 50.54(f)
PIGURE A&B-72 St.B CASE 7,177 FA 3000 3m. 2500 -
d 15
- I E 2000 . _
8
=
O 1500 -
1000 x
500 I !
0 100 200 300 400 500 600 Time, Sec Revision 1 12/79
^'"-' S 1515 001
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-73 SLB CASE 7, 177 FA 600 570 1
520 .
\'
e i
l/s l
V
\
\
3 \ -
h
~
470 \
\
y g HOT LEG
\
\
\
\
\
420 \
coto L G \
\
\
\
\
\
N N
37, i i i i i i i i i I i 1 1515 002 0 100 200 300 400 500 600 Time, See Revision 1 ASB-106 1!'
RESPONSE M 10 CFR 50.54(f)
FIGURE ASB-74 SLB CASE 7,177 FA .
33.20 27.67 _
22.14
=
5 16.60
=
0-a 11.07 5.53 -
0 I ' ' I O 100 200 300 400 500 600 Tirza, Sec
, Revj.sion 1 12/79
^'"-' '
1515 003
RESPONSE TO 10 CFR 50.54(f)
FIGURE A&B-75 SLB CASE 7, 177 FA 1200 1000 E
. 800 -
E a
r
- 600 E
o O
m b
j 400 -
E
=
S 200
' ' ' I I I 0
0 100 200 300 400 500 600 Time, Sec Revision 1 ASB-108 12/79
RESPONSE TO 10 CFR 50.54(f)
~~
FIGURE ASB-76 SLB CASE 7, 177 FA 1200 -
2 1000 E
1 i
5 I 800 U
3=
8 3
= 600 - -
? '
N 400 200 -
0 I I I I O 100 200 300 400 500 600 Time, Sec Revision 1 12/79 ASB-109 -
1515 003
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-77 SLB CASE 7, 177 FA FULL 50 40 -
O STEAM GENERATOR "B"
~
~
~
g 30 -
%=
!J w
20 -
NOTE: SG "A" BOILS ORY WITHIN 20.0 SECONOS 10 -
0 I I O 100 200 300 Time, Sec 1515 006 Revision 1 ASB-110 !
RESPONSE TO 10 CFR 50.54(f)
FIGURE ASB-78 SLB CASE 7, 177 FA 287 l
1 1
11 250 -l{
11 11 11 Il li 200 -
_}i , ,\
il / \
ll /
\
11 / n ll / \/\
s
- i / V \
I \
- I / \
j 150 I g j
3 3
QlI I
f \
g / C AN0Y CANE "A" iI lI I
/
100 _L I
/
I I l! 1 /
/ HOT LEG "Aa l '
l /
\ ;/
i l
50 _L
'i[i
\
l .i UPPER SG TUBE REGION l'
a l
l ,
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RESPONSE 'IO 10 CFR 50. 54 ( f )
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RESPONSE TO 10 CFR 50.54(f)
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RESPONSE TO 10 CFR 50.54(f)
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RESPONSE 'IO 10 CFR 50. 54 ( f )
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RESPONSE TO 10 CFR 50.54(f) e FICURE ASB-83 SLB CASE 8, 177 FA 41.50 -
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RESPONSE 'IO 10 CFR 50. 54 ( f )
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RESPONSE TO 10 CFR ~'.34(f)
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RESPONSE TO 10 CFR 50.54(f)
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RESPONSE 'IO 10 CFR 50. 54 ( f )
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RESPONSE TO 10 CFR 50.54(f)
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RES PONSE TO 10 CFR 50. 54 ( f)
APPENDIX C Question c Provide a schedule of completion of installation of the
~
identified systems and components.
Response ,
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 t ase items and the issue of OTSG sensitivity. Therefore, hardware and procedural changes for these systems tre not addressed.
The table consists of a matrix for both units showing, for each 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 exchar.gers, and tanks),
electrical equipment (cable, conduit, and trays,), and instrumentation ( transmitters , controllerse, tubing, and indicators) is basea on actual quantities installed as of 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 components within the scope of the system or component and when 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 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, increased electrical quantities for raceway, small pipe, wire, and cable, plus constraints due to space limitations, are 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 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 1515 0l8
RESPONSE TO 10 CFR 50.54(f)
APPENDIX C panel fabricator. The remainder of the devices are currently being installed in the field. HVAC installation work is complete, and electrical connectioins and terminations work is being perfarmed on the installed panels.
C-2 11/79 1515 019
TABLE C-1 CONSTRUCTION STATUS HIDLAND PLANT UNITS 1 AND 2 Unit 2 % Complete (Installed) Unit 1 % Complete ( I ns tal '. ed )
Smallin Largem Completion Sma l l"' La rg e"' Completion System / Component Pipe s2 Pipa Equip Elec3' t I n s t r Date Pipe s2h Pipe Equip Elec"3 I n s t r Date l1 HPI NA 75 100 68 6 01/10/80 NA 65 100 68 6 04/24/80 AIM <10 80 100 31 0 0/28/80 <10 75 100 26 0 08/08/80 DHR 50 65 100 57 4 12/21/79 20 55 100 55 1 04/15/80 CFT NA 95 100, 43 17 01/25/80 NA 90 100 43 0 0F/06/80 RCS pressure control 10 50 20 37 1 05/05/80 0 40 10 28 2 07/23/80 Makeup and letdown 5 65 100 39 3 12/14/79 5 55 100 27 3 04/11/80 h Steam generator 75"' h pressure control 60 75 50 36 03/06/80 60 75 7 5 "' 33 36 08/21/80 g Steam generator NA NA 90 NA NA 05/02/80 NA NA 90 MA N4 08/01/80 3 Pressurizer NA 50 90 NA - NA 05/05/80 NA 0 90 NA NA 07/23/80 5 Quench tank NA 50 90 NA HA 05/05/80 NA 10 90 V' NA 7/23/80 h Control room '
layout 75 20 f
NA NA 75 Uy system NA NA 75 20 75 By system -
RCS piping NA 90 NA NA 9 06/12/80 NA 75 NA NA 0 08/16/80 ICS'" NA NA 100 S i a' NA 02/06/81 NA NA 100 3:s' NA 03/18/81 MFW 70 95 100 47 17 02/25/80 55 92 100 36 2 06/04/80 84 Pipe percent represents the summation of pipe spools, hangers, restraints, and field welds.
g 2'Small pipe is primarily vents and drains.
t' Pulling cable was weighted as 75% of ef fort and terminating cable was weighted a' 25% of effort.
a t-
'3', Tubing, mounting, and electrical connections were weighted equally .
"'Remai ni ng 25% !s installation of MAD valves, which are onsite.
@ '6' Includes surge line.
W H
" As scoped , the ICS includes only the control cabinet. Also, field transmitters are included in each respective system.
---. ie> Abo ut 50% of ICS system cables are pulled. Few are terminated. Cables in this system are mostly multi-(2) conductor.
m 'Ihis n'.mber of connection greatly influences total percent complete.
CJ N
C3
RESPONSE TO 10 CFR 50.54(f)
APPENDIX D Ouestion d Identify the feasibility of halting installation of these systems and components as compared to the feasibility of completing installation and then effecting 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 discussed in Appendix C.
- a. Civil - Civil construction work is 82% complete; concrete work is 94% complete. Primary and secondary shield wall construction is complete in both containment bu:' dings. The civil construction opening for the Unit 1 contai ment 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 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%
complete, and both large and small pipe hanger installation, 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; conduit is 72%
D-1 11/79 1515 021
RESPONSE TO 10CFR 50.54(f)
APPENDIX D complete; and electrical cable is 41% complete for both 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 ef fecting 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 effecting 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 associated with a halt in construction, key personnel would be lost to the project.
Consumers Power Company has evaluated the Three Mile Island accident for impact on the Midland design and has closely 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.
D-2 11/79 1515 022
RESPONSE TO 10 CFR 50.54(f)
APPENDIX E Question e Comme 7t on the OTSG sensitivity to feedwater transients.
Response
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. The overall primary-to-secondary heat transfer is controlled by the rate of feedwater introduction to the generator. This, in turn, 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 eu icie cy than a recirculating steam -generator, 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 disturbances has demonstrated the ability to maintain safe core cooling without radioactive release under the applicable licensing assumptions. However, the frequency o.' feedwater transients leading to disturbances of pressure and/or pressurizer level in the primary system of B&W plants has been higher than desir ed . Thi- 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 defense-in-depth; and existing plant features accomplish this defense-in-depth as indicated on the attached Figure E-1, which uses an overcooling event as an example.
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 identified for further investigation as a result of this examination are presented in Appendix F.
s E-1 Revision i 12/79
~ ~
1515
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 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 of tne 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 maintenance of reactor coolant natural circulation. It is recognized that upper head injection 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 t' tis concern.
- b. Concern: "Further system modifications provide control grade reactor trips based on secondary system malfunctions such as turbine or feedwater pump trip. While these reactor trips do serve,to reduce undercooling feedwater transients 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 syctems.
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 reckamination 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 1515 024 E-2 Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f)
APPENDIX E feedwater these (sic) actions may amplify undesireable l1 primary system responses."
Comment: The addition of an automatic reactor coolant pump trip, based upon the coincidence or' 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 challenges to the power-operated relief valve (PORV) by raising the PORV 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 the high-pressure reactor trip have increased the number of 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 system 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 attention 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 featcre of the design. Some B&W operating plants do place reliance upon the operator to limit feedwater excursions which may result from control system failure. However, Midland already includes a number of features *.o reduce this reliance upon the operator. Several additional modifications are being E-3 Revision 1 12/79 1515 025
RES PONSE TO 10 CFR 50. 54 ( f)
APPENDIX E investigated which will further reduce the requirements for the operator to act in response to a control system failure and will thus improve our defense-in-depth against primary system parameter excursions resulting from moderate secondary system upsets. These changes are discussed in Appendix F.
- f. Concern: "It appears that in many cases the main feedwater control system does not react quickly enough or is not sufficiently stable to meet feedwater requirements. Rather, the system will of ten oscillate from underfeed to overfeed conditions, causing a reactor trip and sometimes a high-pressure injection initiation. One undesirable element of this lack of stability is that overcooling transients on the primary side proceed very much like a small break LOCA (decrease in pressurizer level and pressure). Thus, for a certain period of time, the operators may not know whether they are having a LOCA or an overcooling event."
Comment: Overcooling transients in all PNR systems proceed initially like a small break LOCA, and thus are not a unique problem of the OTSG. For example, on a recent reactor trip in a PWR with a recirculating steam generator, a stuck-open turbine bypass valve with approximately 5% capacity caused an 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 coolant sufficient 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 events to the greatest degree practicable, which, when combined with satisfactory design mitigative capability, adequate operator indications, and detailed training, will help the plant operators to act confidently and safely when l1 these abnormal events occur. Consumers Power Co.tpany efforts to improve identification and response to overcooling events are addressed in Appendix F.
- g. Concern: " A major area of concern arising from the B&W OTSG sensitivity is the response of pressuri?er level indication.
Several B&W feedwater transients have led to loss of pressurizer level indiation. Most notable was a November 1977 incident a t Davis-Besse where level indication was lost for several minutes."
Comment: The loss of pressurizer level indication following a reactor trip is an operational concern and should be minimized for expected abnormal occurences. However, it should be noted that the loss of indicated pressurizer level on B&W operating plants is not synonymous with a loss of liquid in the pressurizer. Certain B&W plants, such as the Davis-Besse Unit 1 reactor, have pressurizer level indicators which do not cover the full span of the pressurizer volume.
E-4 Revision 1 12/79 .. ,
1515 Utu
/
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 indicati a 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 pressurizer level range will more closely relate to the full fluid volume of the pressurizer and, therefore, the loss of indicated pressurizer level will be minimized. With this expanded 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 trips the heaters off when pressurizer level is lou. If these nonsafety grade cutoffs should fail, the heaters weuld be kept on while uncovered."
Comment: B&W operating plants include a control grade circuit 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.
- i. Concern: " Overfeed transient (MFW) (not uncommon to B&W) causes overcooling; pressurizer level shrinks, pressure reaches 1,600 psi, RS actuates, RCP is tripped; AFW on (possible RCP seal failure)."
Comment: For the Midland plant, automatic equipment will be installed to eliminate the reactor 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 investigated as a means to terminate main feedwater flow l1 before excessive overcooling occurs. This investigation is further discussed in Appendix F.
- j. Concern: " Operator manually controls AFW (possible MFW instead or in addition if MFW is not isolated such that OTSG 1evel 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 circulntion even if pressurizer does not empty)."
l5l5 C27 E-5 Revision 1 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 found 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 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 operatin'g 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 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 maintain the reactor plant online following various secondary l1 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 meaJures which would improve the reliability of all control functions related to the B&W operating plant ICS design. Consumers Power Company's response to the results of this FMEA is addressed in Appendix F. Control of auxiliary feedwater is provided by a safety grade system independent of the ICS. As 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 Midland design and those under consideration should provide E-6 Revision 1 12/79 n7^
151Jt VL
RESPONSE TO 10 CFR 50.54(f)
APPENDIX E substantial defense-in-depth against sequences of the sort discussed in this section.
Cor. 3rn: "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 issues raised in this paper by improving system reliability and reducing the consequences e,f secondary system upsets. In addition to this, a carefully considered group of invastigations 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.
E-7 11/79 nen 151-3 u/-
1 Defense in depth approach to maintain adequate core cooling IND DEFENSES SEQlJENCE OF EVENTS PRESENT DESIGN FEATURES INFROVE OVERALL PLANI REll438till Ako i
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RUSPONSE TO 10 CFR 50.54(f)
APPENDIX F Question f Provide recommendations on hardware and procedural changes related to the need for and methods for damping the primary system sensitivity to perturbations in the OTSG. Include details 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 OTSG 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 Midland as a result of new regulatory refuirements, 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:
- a. Upgrade of required pressurizer heaters and controls to l1 safety classification to ensure RCS subcooling
- 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, Consumero 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 t concern of overcooling in B&W type plants. The following material discusses changes that have been identified for 1515 031 F-1 Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f)
APPENDIX F incorporation in the Midland design pertinent to this issue and also addresses additional evaluations pertaining to overcooling that are being conducted by Consumers Power Company. In general, 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 logic actuated by either turbine trip or loss of main feedwater. These changes addressed ,l I 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 through prompt reactor trip during transients that resulted in primary system pressure increases.
Although these changeo 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 of trip frequency of certain plants. Because overcooling events 11 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.
Consumers Power Company intends to adopt an alternative solution l1 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) 0,3<
1515 F-2 Revision 1 12/79
RES PONSE TO 10 CFR 50. 54 ( f)
APPENDIX F B. Safety grade anticipatory reactor trip on total loss of 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 assume? 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 events 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 primsry system.
In summary, the design features discussed above address the issue 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 frcm 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.
This potential change is being examined to gage its potential impact on system reliability in an in-depth
, F-3 11/79 1515 033
"ESPONSE TO 10 CFR 50.54(f)
APPENDIX F reliability assessment of the Midland AFW system. Tb 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 corponent 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 scenario will be calculated via fault and event tree techniques. No 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 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 en 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 2110w time for the operator to take actions necessary to presen* 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 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 1515 034 F-4 11/79
RES PONSE TO 10 CFR 50. 54 ( f )
APPENDIX F 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 3725 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 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 following two options for correctit.g the identified 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 vnloc.
- 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.}} }}} F-5 11/79 O
RESPONSE 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 upgrade provides the ensured capability of maintaining adequate subcooling after all anticipated transients through proper RCS pressure control. Addi tionally, redundant pressurizer level and reactor coolant pressure indications have been upgraded to safety grade on the aain control boards and auxiliary shutdown panel. To specifically address the concern of loss of pressurizer 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 320 inches creates a greater potential for an actual off-scale 1cw level as demonstrated by operating plant history. The 40-inch increase at both ends of the present operating range, in conjunction with anticipated improvements in the AFW 1evel control scheme discussed in Section II. A,, will provide additional assurance that the pressurizer level will remain on-scale for all anticipated operational occurrences. Thin design modification is currently being implemented. In the everJ - 'oss of liquid inventory in the pressurlzer, 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 inpact on construction schedule is expected. C. Transient Identification As discussed in Appendix E, overcooling events in all PNR systems proceed initially like a small break LUCE. Therefore, it is important that any auotmatic plant response required for either of these events be able to differentiate between similar appearing transients in order that system actuations only occur when desired. Additionally, sufficient 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 areas is discussed below. In general, specifically identified design changes can be accommodated within the F-6 Revision 1
'Idh5036
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, however, 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 sersing 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 of void formation will not reach a point where degraded pump current will actuate RCP trip.
- 2. Plant Indications
- 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 m Consumeis 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 assessed for possible use in providing core map temperature trending 3argin to saturation of the average incore thermocouple temperature, 33 and margin to saturation of the hottest valid I thermocouple temperature.
1515 037 F-7 Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f) APPENDIX F
- c. Primary Coolant Level Indication - Consumers Power Company is participating in a B&W engineering study to determine the most appropriate meth'i for operator recognition of inadequate coolant evel. The " . od presently being conridered to ...o 'i the information is hot leg vY r leve' i. lieu of reactor vessel water lev?). Differential pressure on une hot les fe m the top of the candy cane tc the b;ttom the hot leg piping) is ti.e S thnique ceing evaluated for measuring the level.
- d. Natural Circulation Flow Indication -
Consumers Power Company is reviewing the 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 wire display of the above-mentioned plant l1 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 (BAW-1564, August 1979). As a result of this 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. 1515 038 F-8 Revision 1 12/79
RES PONSE 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 addit 'anl 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 faults 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 system diagrams are being used as a vehicle for this review. The methodology for 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 F-9 11/79 1515 03?
RESPONSE TO 10 CFR 50.54(f) APPENDIX F review. The analysis has presently identified some 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), che 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 Nuclear for Consumers PoWar Company. This analysis 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 program. Expected changen resulting from ATOG will be procedural in nature and therefore do not impact construction. Finally, B&W has performed an analysis of the dynamic post-trip response of the NSSS to overcooling 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, qg, iS15 040 F-10 Revision 1 12/79
RESPONSE TO 10 CFR 50.54(f) APPENDIX F 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 ensura proper actions are taken and to identify additional steps to counteract failures or degraded conditions as identi fied. 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 mitigation for those transients which do occur. Design changes desc*ibed in this appendix, along with additional modifications and 1 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. In conclusion, Consumers Power Company is actively pursuing the issue of overcooling and has committed to modifications which, coupled with the outcome of various ongoing plant studies, will result in a final design which provides adequate l1 de'fenses to these events. We believe that most changes which ere significantly impacted by construction status have already been initiated and that any remaining modifications resulting from completion of design studics can be accommodated during system construction.
\5\5 041 F-ll ,evision 1 12/79
RES PONSE TO 10 CFR 50. 5 A ( f) APPENDIX F TABLE F-1 DL_'ENSE-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. AFW 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 Cutient Coincidence Logic l1
- 13. Instrumentation to Detect Inadequate Core Cooling
- 14. ICS FMEA
- 15. MFW System Review
- 16. ATOG Program
- 17. Restoration of Original PORV and High RCS Pressure Trip Setpoints 1515 042 F-12 Revision 1 12/79
. RESPOWSE TO 10 CHI 50.54(f) Figure F-1
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