ML20009E461
| ML20009E461 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/17/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-81-07-063, LSO5-81-7-63, NUDOCS 8107280227 | |
| Download: ML20009E461 (68) | |
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July 17,1981 o
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Mr. David P. Iloffman
'h.( u.s, %%=D Q Nuclear Licensing Adninistrator Consumers Power Compcny 6
1945 W Parnall Road N
D Jacksca, Michigan 49201 m
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Dear Mr. Hoffman:
SUBJECT:
SYSTEMATIC EVALUATION PROGRAM (SEP) FOR PALISADES m
NUCLEAR POWER PLANT, UNIT 1 - EVALUATION REPORT ON TOPICS VI-2.0 AND VI-3 (DOCKET NO. 50-255)
Enclosed is an updated copy of our contractor's (Lawrence Livermore National Laboratory) draf t technical evaluation report for SEP Topics VI-2.D " Mass and Energy Release for Possible Pipe Break inside Con-tainment," and VI-3, " Containment Pressure and Heat Removal Capability."
The NRC comments which appeared as pen and ink changes in Appendix A of our June 17, 1981 letter have been incorporated into this corrected
. version.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
Draf t SEP Topics VI-2.0 ar.d VI-3 cc w/ enclosure:
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e Mr. David P. Hoffman cc M. I. Miller, Esquire U. S. Environmental Protection Isham, Lincoln & Beale Agency Suite 4200 Federal Activities Branch One First National Plaza Region V Office Chicago, Illinois 60670 ATTN: EIS COORDINATOR 230 South Dearborn Street Mr. Paul A. Perry, Secretary Chicago, Illinois 60604 Consumers Power Cocpany 212 West Michigan Avenue Charles Bechhoefer, Esq., Chairman Jackson, Michigan 49201 Atomic Safety' and Licensing Board Panel Judd L. Bacon, Esquire U. S. Nuclear Regulatory Commission Consumers Power Coapany Washington, D. C.
20555 212_ West Michigan Avenue Jackson, Michigan 49201 Dr. George C. Anderson Department of Oceanography Myron M. Cherry, Esquire University of Washington Suite 4501 Seattle, Washington 98195 One IBM Plaza Chicago, Illinois 60611 Dr. M. Stanley Livingston 1005 Calle Largo Ms. Mary P. Sinclair Santa Fe, New Mexico 87501 Great Lakes Energy Alliance 5711 Summerset Drive Resident Inspector Midland, Michigan 48640 c/o U. S. NRC P. O. Box 87 Kalamazoo Public Library South Haven, Michigan 49090
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315 South Rose Street Kalamazoo, Michigan 49006 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan 49043 Office of the Governor (2)
Room 1 - Capitol Building Lansing, Michigan 48913 William J. Scanlon, Esquire 2034 Pauline Boulevard Ann Arbor, Michigan 48103 Palisades Plant ATTN: Mr. Robert Montross Plant Manager Covert, Michigan 49043 e
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Appendix A SEP Containment Analysis and Evaluation for the Palisades Nuclear Power Plant Contents Page,
1 1.0 Introduction and Background 2.0 Containment Functional Design 1
2.1 Review of Palisades Containment Design Analysis 2
2.2 Primary System Pipe Break 3
2.3 Secondary System Pipe Break 3
2.4 Reanalysis of Palisades Containment Design 4
3.0 o imary System Pipe Break 4
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Initial and Boundary Conditions 4
3.2 Blowdown Phase 6
3.3 Reflood Phase 7
3.4 Post-Reflood and Containment Response Calculation 9
3.5 Containment Response Results 14 4.0 Secondary System Pipe Break 16 4.1 Assumptions 16 4.2 Containment Response Results 17 e
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LIST OF TABLES Page Table 3.1 Palisades Double-Ended Suction Leg Break Blowdown Energy 23-Balance 3.2 Paiisades Double-Ended Suction leg Break Reflood Energy 24 Balance 3.3 Heat Structures in Palisades Containment Model 25 3.4 Palisades Containment--Initial Conditions 26 4.1 Main Steam Line Break Mass / Energy Release Data 27 4
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LIST OF FIGURES 1
Figure Page 3.1 Break Flow, Steam Generator Side of Break 29 3.2 Break Enthalpy, Steam Generator Side of Break 30 l
3.3 Break Flow. Pump Side of Break 31 3.4 Break Enthalpy, Pump Side of Break 32 3.5 Break Flow - Steam Generator Side 33 3.6 Break Flow Enthalpy - Steam Generator Side 34 3.7 Break Flow - Pump Side 35 3.8 Break Flow Enthalpy - Pump Side 36 3.9P Containment Pressure - Superheated Steam Only (Method A) 37 3.9T Containment Temperature - Superheated Steam Only (Method A) 38 3.10T Containment Pressure - Superheated Steam Only (Method B) 39 3.10T Containment Temperature - Superheated Steam Only (Method B) 40 3.11P Containment Pressure - Water and Superheated Steam (Method A) 41 3.11T Containment Temperature - Water and Superheated Steam (Method A) 42 3.12P Containment Pressure - Water and Superheated Steam (Method S) 43 3.12T Containment Temperature Water and Superheated Steam (Method B) 44 3.13P Containment Pressure - Water and Saturated Steam,(Method A) 45 3.13T Containment Temperature - Water and Saturated Steam (Method A) 46 3.14P Containment Pressure - Water and Saturated Steam (Method B) 47 3.14T Containment Temperature - Water and Saturated Steam (Method B) 48 4.1 Main Steam Line Break Flow 36" Diameter 49 4.2 Main Steam Line Break Flow 24" Diameter 50 1
4.3P Comparison of MSLB Case 1 and Reference 1, Containment Pressure 51 4.3T-A Containment Temperatures - MSLB Case 1 with Spray at 84 seconds 52 4.3T-B Containment Temperatures - MSLB Case I with Spray at 30 seconds 53 4.4P Containment Pressure - MSLB Case 2 54 4.4T Containment Temperature - MSLB Case 2 55 4.5P Contaiament Pressure - MSLB Case 3 56 57 4.5T Contairment Temperature - MSLB Case 3 4.6?
Centainment Pressure - MSLS Case 4 58 4.6T Containment Temperature - MSLB Case 4 59 4.7P Containment Pressure - MSLS Case 5 60 61 4.*7T Contairment Temperature - MSLB Case 5 4.8P Containment Pressure - MSLB Case 6 62 63 4.9T Containment Temp'erature - MSLB Case 6 l
o 1.0 Introduction and Background On January 1,1980 the Office of Nuclear Reactor Regulation (NRR) initiated a two-year program with Lawrence Livermore National Laboratory (LLNL) titled Containment Analysis Support for the Systematic Evaluation Program (SEP). This program is directed toward resolution of SEP Safety Topic VI-2.0, Mass and Epergy Release for Possible Pipe Break Inside Containment, and Safety Topic VI-3, Containment Pressure and Heat Removal Capability. The containment structure encloses the reactor system and is the final barrier against the release of radioactive fission products in the event of an accident. The containment structure must, therefore, be capable of withstanding, without loss of function, the pressure and temperature conditions resulting frem postulated LOCA and steam line break accidents.
Furthermore, equipment having a post-accident safety function must be environmentally qualified for the resulting adverse pressure and temperature conditions. To accomplish the objectives of this program, first, the existing docket information was reviewed and evaluated and then additional analyses were performed as required. The purpose of this report is to document original analyses performed by the LLNL on the containment functional design capability of the Palisades Nuclear Power Plant and evaluate existing analyses for conformance with current NRC criteria.
2.0 Contair. ment Functional Design Palisades is a Combustion Engineering PWR licensed to operate at 2200 MWt. The primary coolant system is a two loop system consisting of two steam generators with two cold leg loops per steam generator. The containment systems include the contairnent structure and associated systems. These systems include containment heat removal systems, containment isolation systems and a-combustible gas control system..
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e The containment is a steel-lined, pre-stressed, post-tension concrete structure with a net free volume of 1,640,000 cubic feet. The containment structure houses the nuclear steam supply system, including the reactor, steam generators, reactor coolant pumps and pressurizer, as well as certain components of the engineered safety features systems. The containment is e
designed for an internal pressure of 55 psig and a temperature of 283'F. '
2.1 Review of Palisades' Containment Design Analysis There are two separate calculations which make up the containment design analysis. Fir st is the mass and energy release analysis for postulated L OCA's. This consists of a blowdown, reflood and post-reflood phases. The results are mass and energy release rates into the containment. For PWR's there are two possible break types which must be analyzed, a primary system pipe break and a secondary system pipe break. A break on the primary side generally results in the most severe pressure response in the containment while a break on the secondary side results in the most severe temperature conditions in the containment. The second calculation which is performed in the containment design analysis is the containment response calculation. This results in the containment temperature and pressure response to the mass and energy release from the postulated breaks.
The acceptance criteria used to evaluate Palisades' Containment Design Analysis was based on the Standard Review Plan (SRP).
In order for the containment design analysis to be found acceptable both the mass and energy release and containment response calculation must meet the acceptance criteria specified in the SRP.
2.2 Primary System Pipe Break The SRP specifies several acceptance criteria applied to the mass and energy release analysis for primary system pipe breaks. Among these are break location.
In the Palisades FSAR the most severe mass and energy celease rate, calculated for containment design was done assuming a double-ended cold leg discha'rge break with no accounting for the reflood phase or energy in the secondary system. Since this does not meet the acceptance criteria specified in the SRP or previously accepted methods by the NRC staff, this analysis is unsui*.able for containment casign calculation. Since the mass and energy release rate analysis is found unacceptable, so is the containment response calculation based on the mass and energy release rates.
2.3 Secondary System Pipe Break The most recent secondary system pipe break analysis that was reviewed was submitted by Consumers Power Co. to the U.S. NRC on January 21, 1980.I In this analysis a main steam line break (MSLB) analysis was performed.
In this analysis the blowdown of one steam generator with feedwater isolation and loss-of-offsite power was considered. However, the analysis did not address the possibility of a single failure of one of the main steam isolation valves which could lead to the blowdown of both steam generators. Therefore, the analysis was considered inccmplete and unacceptable. A more thorough discussion of the MSLB analysis is given in Section 4.0, Secondary System Pipe Breaks.
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Palisades Plant - Autmoatic Initiation of Auxiliary feedwater Ssytem at Palisades Plant, Occket 50-255 - License OPR-20, January 21, 1980 letter from Reger W. Huston of Consumers Pover Co. to Dennis L. Zieman of NRR, NRC....
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2.4 Reanalysis of Palisades' Containment Design As mentioned earlier in Section 2.1, Review of Pali <ades' Containment Analysis, there are two separate calculations which make up the containment design analysis, the mass and energy release rate and the containment response. The mass and energy release rate calculation can be the result of either a primary or secondary pipe break. The primary pipe break generally results in the limiting condition for calcalating the peak pressure 'inside the containment. The secondary pipe break analysis generally is the most. limiting case for temperature conditions inside the containment. Both of these analyscs were performed and are discussed below.
3.0 Primary System Pipe Break For a primary system pipe break there are three phases in calculating mass and energy release rates. These are the blowdown, reflood, and post-reflood phases.
In each of these phases the calculation was done in accordance with the Section 6.2.1.3 of the Standard Review Plan under tite limitations of the computer codes used. The primary limitation was the carry-over rate fraction which was discussed in some detail in the Methodology Report for the Palisades Nuclear Power Plant.
In general, the analysis was done in a manner that conservatively establishes the containment design pressure; i.e., maximizes the post-acciden*
'tainment pressure. The worst break location was determined to be at the cold-leg pump suction sice because of the consideration of energy input during the reflood phase and flow resistance.
3.1 Initial and Boundary Conditions The initial and boundary conditions for this analysis were defined to satisfy the requirements of the Standard Review Plan. The single failure assumption for these analyses was a loss of one diesel generator. The initial s
power was specified to be 102% of safeguards design rating or 2690.76 MWt. A steady-state mass and energy distribution was provided in the primary and secondary coolant systems consistent with the conservative core power. The break flow'. were calculated using a discharge coefficient of 1.0, with the Henry-Fauske correlation for subcooled and the Moody correlation for saturated fluid. The safety injection flows were minimum, cor responding to the diesel generator failure. The mass and energy release analysis was performed with RELAP4 M006. Steam quenching by the safety injection water occurred due to
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the homogeneous equilibrium (HEM) assumptions of the RELAP4 M006 code. All of 0
the safety injection water temperatures were defined to be 90 F.
Scram was assumed to occur with a low pressurizer pressure of 1750 psia.
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A 1.0-second delay tine was used in the model for conservatism; however, tne moderator reactivity feedback caused core shutdown before the control rods ware effective. The main coolant pump power was tripped off at the time of the break. Steam generator isolation was initiated one second after the break and the valves were assumed to completely close in five seconds. A 15-psia constant containment backpressure was assumed to maximize mass and energy release throughout the blowdown. The end of blowdown was defined as the time the primary system pressure reached the containment design pressure of 55 psig.
The RELAP4 input deck was obtained from NRC and was carefully reviewed for code options, initial and boundary conditions. The plant physical description was generally assumed to be correct. Additional information required for the analysis was obtained from the Palisades FSAR, and telephone conversations with C. Tinkler of NRC and D. Vandewalle of Consumers Power Company. A thorough discussion of the model can be found in the Methodology Report for the Palisades Nuclear Power Plant.
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3.2 Blowcown Phase The blowdown analysis,results are summarized in Table 3.1 and Figures 3.1 through 3.4.
Table 3.1 itemizes the energy sources for the duration of the blowdown which ended at 20.4 seconds after the break. The total energy released during blowdown was approximately 253.4 million Btu. Figures 3.1 through 3.4 provide break flow and enthalpy out the break.
The accumulator flows start after 16 seconds and do not reach maximum flow rates by the end-of-blowdown. The pumps coast down at different rates. The pump nearest the break reaches zero rpm before two seconds because of reverse flow through the pump. The pumps were not allowed to reverse, providing a conservati.vely high resisttece which allows more flow through the steam generator side of the break. The other pump in the brn'.an loop coasts down to zero rpm at about 11 seconds. The pumps in the unbroken loop continue to have a positive rotation throughout the blowdown, although it decreases to 500 rpm in about 10 seconds. Although~the scram occurred at about eight seconds, moderator reactivity feedback had already reduced the power to less than 7-1/2% of the initial power.
The mass and energy release rates and energy sources were qualitatively compared to the CESSAR results for a double-ended suction leg slot break with the same area. The similarity of the results suggests the RELAP4 calculated blowdown results are reasonable.
3.3 Reflood Phase The reflood analysis for the double-ended pump-suction break was assumed to immediately follow the LOCA blowdown analysis. The analysis was performed using RELAP4 M007. Within the limitations of RELAP4 M007, the analysis was performed in accordance with the requirements of Section 6.2.1.3 of the Standard Review Plan (SRP)..
Initial conditions for the start of the reflood analysis were based on the end-of-blowdown (E08) results. EOS was defined to occur when the primary system pressure fell below the Palisades containment design pressure of 55 psig which occurred at 20.4 seconds after the start of blowdown. At that time, the core power level had dropped to 159.41 MWt or approximately 6% of the initial The accumulator flows had been initiated on low cold leg pressure po wer.
trips of 262.5 psia which occurred at about 16 seconds into the blowdown and had reached a total of 5900 lbm/sec at the start of reflood. The reactor coolant pumps had coasted dcwn and the rotors were locked.
For the reflood analysis, the primary system was initialized at the containment design pressure, 69.7 psia. The primary system junction flows were zero except for the accumulator and lower plenum inlet and outlet junctions. Heat conductor temperature and primary system state conditions were established based on the EOS conditions. Core power continued to decrease according to the ANS decay heat curve.
A natural circulation heat transfer model was used in the steam generator secondary to maximize the energy transfer rates to the break. The primary coolant pump rotors were assumed locked to conservatively provide resistance to flow. A closed valve was modeled in the intact cold leg of the broken loop to conservatively increase the flow through the steam generator.
For numerical stability of the RELAP4 computer code, the Emergency Core l
Cool'ng System (ECCS) flow was modeled as being injected directly into the i
0 downcomer at a temperature of 300 F.
Plant specific information was predominantly derived from a RELAP4 Feflood i
input listing for the Palisades power plant which was obtained from the Nuclear Regulatory Commission (NRC), and from the Palisades Final Safety analysis Report (FSAR).
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Several sensitivity calculations were performed to evaluate various input model and code options. The result's of the sensitivity studies are documented in the methodology report. The Palisades reflood transient results are presented in Table 3.2 and ' Figures 3.5 through 3.8.
Table 3.2 i,s a summary of the energy balance at the beginning and end of reflood. Figures 3.5 through 3.8 provide break flow and enthalpy out the break.
The accumulator flow is initiat'ed at 5900 lbm/sec and quickly rises to 6400 lbm/sec. The flow remains constant until 40 seconds, and then is ramped down to 0 lbm/sec at 50 seconds when the accumulator is empty. The HPI flow comes on at 0.6 seconds and remains at about 650 gal / min for the duration of the transient. The LPI flow comes on at 7.6 seconds and varies in magnitude between 400 and 600 lbm/sec for the duration of the transient, depending on the primary system pressure.
The primary system pressure starts at 69.7 psia, increases to.160 psia at 20 seconds, and then slowly decreases to 100 psia. The pressure increase can be attributed to steam binding in the crimary system. As the ECCS water enters the core, it boils away faster than the generated steam can escape through the break. After 20 seconds, the core is quenched and the steam generation rate reaches a new pseudo-steady-state with the break flow.
Normally, the end of reflood is defined as the time when the core recovers to within two feet from the top of the core.
In the case of Palisades, the maximum mixture level is less than seven feet at 50+ seconds into the transient wnich is still four feet below toe top of the 11-foot core.
However, the core-stored energy was essentially removed at 30 seconds into the l
The reflood calculation was extended to 100 seconds to determine when and if the steam generator side break flow would begin a rapid decay expected after the accumulators emptied at 50 seconds. Since the rapid flow decay did I.
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e not occur, the reflood calculation was continued beyond the time when the containment calculation predicts the peak pressure and temperature at 84 seconds after break or 64 seconds after start of reflood.
Because the safety injection water was assumed to be at 300 F, the extended duration of the reflood analysis is considered to provide a conservatively high energy transfer rate to the secondary.
3.4 Post-Reflood and Containment Response Calculation The containment model used was based on a CONTEMPT deck received frcm the NRC. The mass and energy flows to the containment were replaced and the remaining data carefully citecked against the FSAR and other sources. The analysis was performed using CONTEMPT-LT/028.
The heat structures used are listed in Table 3.3.
All the structures are represented in rectangular geometry. The thermal conductivity and the
. volumetric heat capacity were checked for the four materials used: steel, concrete, insulation,andair(gap). The heat capacity was found to be about
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two orders of magnitude low for insulation and was changed. Tagami/Uchida boundary conditions were used for all heat structure surf aces except the base slab, which was assumed to be covered with water. The Tagami peak time used was 20 seconds, 0.5 second before the end of blowdown.
The basic assumption was that off-site power was lost and that one diesel generator failed to start. The cooler and spray pump start times are based on the generator loading sequence.
It was assumed that one fan cooler was operating and that it started at 23 l
seconds after the break. The heat removal rate was variable, ranging from 97.5 M8tu/hr at a containment temperature at 350 F to 3.0 MBtu/hr at 104 F.
The one operating diesel generator was also assumed capable of 0
powering two spray pumps. Both pumps together were capable of 1.34 Mlbm/hr (2700 gpm) with spray efficiency of 90%. The containment spray started at 84
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seconds and used water f-om the Refueling Water Storage Tank until 30 minutes when the tank was empty, at which time the water source switched to the containment sump. The heat removal rate for the shell and tube heat e.schanger in the containment spray system is computed in CONTEMPT. The entered parameters were: 'the product of the heat exchange surface area and the overall heat transfer coefficient was 2.28 MBtu/hr/F; the coolant inlet temperature was 114 F; and the coolant flow rate was 2.0 Mlbm/hr.
General initial conditions are given in Table 3.4.
Initial conditions for the primary system refer to the end of blowdown. No water was introduced to the containment as an initial step input. The evaporation-condensation model in the containment was bypassed until the end of blowdown. The fraction of wall or cooling coil condensate transferred from the superheated containment atmosphere to the pool was set at 0.52.
The heat and mass transfer multipliers were set at 1.0, and the temperature flash option was used.
Two methods of trea, ting the post-reflood period were used for this
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analysis and three dif.ferent assumptions,made about the mass and energy release during reflood, resulting in six cases. The blowdown mass and energy release was the same for all cases. Peak flow was about 77,000 lbm/sec at 525 Btu /lbm, and the blowdown ended at 20.4 seconds. The reflood data lasted 100 seconds, which corresponds to 120.4 seconds 'after the break. The core was not covered, or even two-thirds covered at this time, but it was substantially cooled. Therefore, the end of the RELAP4 reflood run was defined to be the end of reflood even though the accumulator flow had been ramped down to zero between 60 and 70 seconds after the break.
During the post-reflood period, decay heat, heat from the sacondary system, and heat from the heat structures in the primary system are released to the containment. The decay heat is released over the duration of the run based upon the ANS standard decay heat curvi plus 20% and an ultimate reactor
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power of 2638 MWt plus 2% for instrument error (excluding pump heat). The heat from the secondary system (61 MBtu)'and the primary heat structures (53 MBtu) was all released by one hour af ter the break. A linear ramp to zero was used.
The amount of heat released to the containment by the secondary was determined by obtaining the sto-ed energy in the water in steam generators and in the SG tubes at the end of the RELAP4 reflood calculation. Assuming that this was based on 32 F and that the entire steam generator would be at 212 F af ter one hour, the amount of heat available to be released was computed to be 61 MBtu. This is conservative because the containment pressure will not decrease to atmospheric pressure in one hour, and so the secondary 0
system will be hotter than 212 F.
For the primary heat structures, the energy stored in all of the heat structures used in the reflodd model, except the core (fuel rods) and the steam generator tubes, was used in the model.
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.was conservatively assumed that all this metal would be at 212 F after one hour, with the difference (53 MBtu) being released to the containment.
Juring the post-reflood period, two different methods were used which differ only in the manner in which the energy from the secondary system and 4
primary heat structures is released after reflood.
In Method A, the energy is released directly to the containment without considering the primary coolant system. The amount of mass accompanying this energy release is required, and I
it is obtained by assuming that the heat is used in converting water at saturation to steam. A typical value for the heat of vaporization at the pressures experienced in the containment for the first hour is 925 Btu /lbm, and this has been used to calculate ".he mass release rate. This method is conservative, since af ter some time the water in the primary system will cool to belo'v the boiling point and most of the decay heat will go into heating the water to saturation and leaving very little to generate steam. The systems.
l (HPIS and LPIS) that inject water into the primary are not modeled in Method A since the mass and energy flow from the primary is already calculated as described above.
In Method B, the decay heat and heat from the secondary system and from,
the primary heat structures is passed to the compartment through the reactor coolant system. The model in CONTEMPT then determines how much steam is produced, how much heat goes into increasing water temperature, and so on.
This is much more realistic than Method A since it allows the steam production to decrease with time.
It is still conservative since the heat input has been calculated to be conservatively high.
In Method B, injection into the primary is explicitly modeled. The LPIS (5000 gpm) and the HPIS (450 gpm) both take water from the refueling water storage tank until it empties at 30 minutes.
After that, only the HPIS continues, taking water from the containment sump.
There is no heat exchanger on either of these systems. The containment sump is gradually cooled, since the water that is' recirculated through the containment spray system does pass through a heat exchanger.
Three assumptions were considered for the mass and energy release to the containment during reflood.
In the first assumption, only the steam flow frcim the SG side of the break was used frcm the RELAP4 results. To be l
conservative, it was then assumed that all of this dry steam w?s superheated to 1300 Btu /lba (about 500 F), and the energy release rate was obtained by multiplying the steam flow rate by 1300 Stu/lbm. The actual effluent enthalpy is about 1200 Stu/lbm for the first 20 seconds of reflood and gradually decreases to about 600 Stu/lbm after that. Thus, assuming that only dry superhe ed steam is released is conservative.
In the second assumption, the mass and energy flow rates from the SG side of the break were used, but the energy flow was augmented to account for superheating. At 70 psia, saturation is about 310 F, and the specific l l
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500 F is only 1282 Btu /lbm. While the SG tubes are a little below 500 F at the start of reflood, their temperature.is on the order of 350 F at the end of reflood. Therefore, adding 100 Btu /lbm for each pound of steam flow is conservative. The steam flow rate used to calculate this added er.ergy was the same as that used in the first assumption. The additional energy was about 8%
of that computed by RELAP4 at the beginning of reflood and about 6% at the end i of reflood.
In the third assumption, the mass and energy released from the SG side of the break by RELAP4 were used directly. The liquid phase falls'to the pool as released. Naturally, this case results in lower. peak temperatures and pressures than the superheated-steam-only case, but it is more realistic and it is conservative.
For all three assumptions, the release of water from the pump side of the break is ignored. This release is all liquid phase and goes directly to,the containment sump. The amount of water in the sump has a negligible effect on the temperature and pressure history of the containment vapor region. The RELAP4 model had to use ECCS water at 300 F in order to avoid instabilities, whereas the ECCS water would actually be about 100 F.
Since the water coming out the pump side of the break will have had no contact with the core and little with any of the metal enclosing the primary system, it should not be significantly warmer than when it left the accumulators.
In view of the large difference between th actual and the model ECCS water temperatures, neglecting the liquid flow from the pump side of the break is more realistic than including it.
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l 3.5 Containment Results The results of the CONTEMPT runs are shown in figure 3.9 through 3.14.
Figures 3.9 and 3.10 show the results for the case where only dry, superheated steam flow from the SG side of the break was considered, and the energy release rate during reflood was obtained by multiplying the steam flow rate by 0
1300 Stu/lbm (which is approximately the specific enthalpy at 500 F and 70 psia). Figure 3.9 shows the results for Method A and Figure 3.10 shows the results for Method B.
The tw cases are identical to 120.4 seconds since the difference is in how the mass and energy releases are handled after reflood.
The figures show that the containment atmosphere reached almost 70 psia and 325 F at 84 seconds just before the containment spray began. Since only' dry steam was released to the containment during reflood, the containment spray has an immediate and dramatic effect on the containment vapor region temperature and pressure. The peak pressure equals the containment design pressure of 69.7 psia.
In view of the conservative assumption of releasing only dry, superheated steam during reflood, this is not considered significant.
It is inconceivable that the superheated steam could flow frcm the steam generator to the break with the saturated water and not mix to form a homogeneous, two-p?ase flow.
Since the Method A assumptions are not suitable for a long-term model, the run shown in Figure 3.9 was terminated at two hours, while the Method B run in Figure 3.10 was continued to ten days. The results of the two methods are quite close at two hours. The dip in the atmosphere pressure and temperature at 20 minutes (1800 sec) in Method 8 is due to the shutdown of the LPIS at the time when the RWST runs dry. This does not show up in Method A since the primary system is not modeled. The change in slope at 30 minutes in the Method A result is due to the fact that the source of water for the containment spray changes from the RWST to the warmer containment sump. The.
containment vapor region temperature reaches 135 F at about 8.08 days.
(698,400 sec) in Method 8 (see Figure 3.10).
Figures 3.11 and 3.12 show the results for the release of a two-phase mixture with the energy flow augmented to account for superheating the steam.
fraction. Peak pressure is about two psi below the design pressure, and the peak containment atmosphere temperature is about 283%. Behavior after a few hundred seconds is nearly identical for all the A cases and all the B cases.
This is to be expected since events are dominated by the absorptive capacity of the heat structures and the effect of the sprays. The containment 0
atmospheric temperature reaches 135 F at about 8 days.
In view of all the conservative assumptions made elsewhere, the results sht,wn in Figures 3.11 and 3.12 are sufficiently conservative and meet the Standard Review Plan requirement for superheated steam. Since the effluent to the containment will certainly not be dry steam in view of the carryover rate fraction in the core, this as:umption of wet steam with the steam fraction arbitrarily superheated appe'ars to be the maximum which can be justified as realistic.
Figures 3.13 and 3.14 show the results when the RELAP4 calculated releases from the SG side of the break without adding any energy to account for superheating steam release during the reflood period. Peak pressure was over.
I 4 psi below the design pressure, and the peak containment atmospheric temperature was more less than 273 F.
The containment atmospheric temperature dropped to 135 F after about 8 days. So many conservatisms, including the _use of 300 F ECCS water, were made in arriving at the releases i
l to the containment during reflood and post-reflood that these results are definitely conservative. However, they do not meet the SRP mandate for superheated steam.,
4.0' Secondary System Pipe Break Analyses of the containment response to a secondary system pipe break were also made. For PWR's the most limiting break location is a main steam line break with pure steam blowdown.
In the case of Palisades the results show that a single failure assumption which allows both steam generators to blowdown will produce peak pressures and temperaturs which exceed design values. The model and assumptions that were used in analyzing the main steam line break are given in the following discussion.
4.1 Assumptions A main steam line break (MSLB) analysis was performed by Consumers Power Company.I Results given in this reference are used for ccmparison purposes.
In particular, mass and energy release data for the full power MSLB case discussed in the Palisades FSAR is provided in Table' 1 of reference 1.
The Palisades FSAR full power a'nalysis assumed:
(1) A double-ended guillotine rupture of a main steam line inside the containment.
(2) A reduction in feedwater flow from full flow to zero over the 60 seconds immediately following scram at less than 2 seconds on high containment pressure.
(3) Both main steam isolation valves would close on low steam generator pressure (500 psia) causing the unruptured steam generator to isolate in eight seconds.
(4)
Off-site power was available.
i I
Palisades Plant - Automatic Initiation of Auxiliary Feedwater System at Palisades Plant, Occket 50-255-License DPR-20, Janury~ 21, 1980 letter from Roger W. Huston of Consumers Power Co to Dennis L. Ziemann of NRR, Nuclear Regulatory Commission.
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(5) Only two containment spray pumps were available; no a'r coolers were available.
(6) Pure steam blowdown (no moisture carryover).
(7) A h'ighly conservative containment heat transfer.model.
In addition to the results reported in reference 1, a number of analyses for Palisades containment response to the MSLB were made. The analyses employed RELAP4 to obtain mass and energy release rates and CONTEMPT-LT/028 to obtain containment response.
The RELAP4 mass and energy release rates were obtained using a simplified model based on one volume, one heat conductor, one break junction, and one feedwater fill junction. Two break sizes and three feedwater flow assumptions were analyzed. The resulting break flow rates are summarized and compared to reference 1 results in Figures 4.1 and 4.2.
For the RELAP4 analyses, th?
steam generators were assumed to be at 770 psia, with an averaga water enthalpy of 552.2 Stu/lbm and contain.128,456 lbm each. The primary system was assumed to' be held constant during the blowdown with 513.83 F local temperatu'e and a 952 Stu/hr/ft / F heat *.ransfer coefficient in the steam r
generator. Figures 4.1 and 4.2 show that the main effect of the feedwater is to prolong the time of blowdown period and increase the total mass and energy to the containment.
l The containment responses for a number of MSLB cases have been compared.
The results are given in the following discussion.
4.2 Containment Response Results Case 1 The first case selected for analysis was intended to determine if the f
CONTEMPT-LT/028 model used would give results similar to those in reference 1 if similar assumptions were employed. Consequently, two CONTEMPT runs were
- I :. : 2.^ T ~ L
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made, using the reference 1 mass and energy release rates. These CONTEMPT runs assumed two spray pumps and one. fan cooler were available and a TPEAK of eight seconds for the CONTEMPT Tagami/Uchida heat transfer correlation.
The one run assumed that off-site power was not available so the two spray pumps were started at 84 seconds. The other CONTEMPT run assumed off-site power was available so the spray pumps were started at 30 seconds.
The containment histories for the two Case 1 runs are compared to the reference 1 pressure history in Figure 4.3P.
The comparison between the results with spray after 30 seconds and the reference i results is close, indicating an acceptable CONTEMPT model. The containment temperature histories for the Case 1 model are shown in Figures 4.3T-A and 4.3T-8.
Case 2 The purpose of the Case 2 analysis was to determine if the one volume RELAP4 model was adequate for obtaining mass and energy release rates to,the containment. The blowdown mass and energy release rates for various feedwater and break area combinations have been noted in Figures 4.1 and 4.2 For Case 2, the ruptured steam generator blowdown was si.nulated by the 36-inch break 2
(area = 6.12 ft ) with main feedwater only. This feedwater flow was initially 1650 lb/sec and ramped down to zero flow at 60 seconds. The unruptured steam generator was assumed to isolate (MSIV closure, not failure) so the mass and energy release rates were obtained from refrence 1.
The mass and energy release rates for the two steam generators were added for input to the CONTEMPT model.
The CONTEMPT assumptions for Case 2 were similar to the assumptions for the Case 1 run with spray after a 30 second delay. The containment pressure and temperature response from Case 2 is shown in Figures 4.4P and 4.4T, respectively. The peak pressure is about 65 psia, which is slightly less than,
for the comparable Case 1 run and the reference 1 value. Therefore, the one volume RELAP4 model was judged to be adequate for obtaining mass and energy release rates.
I-t is noted thct complete phase separation is modeled in the RELAP4 anclyses so that pure steam blowdown occurs.
Case 3 Cases 1 and 2 established that the CONTEMPT and RELAP4 models were adequate for obtaining containment response to a MSLB. Cases 3 and 4 were designed to determine the response of the Palisades containment to the MSLB for blowdown of both steam generators, with off-site power available. Case 3 assumed that each steam generator would blow down through a 24-inch break 2
(3.06 ft ).
Case a assumed that the ruptured steam generator would blow 2
down through the 36-inch (6.12 f t ) break and the unruptured steam generator would blow down through a 24-inch break.
Th,e assumptions.used for Case 3 include:
(1)
If off-site power is available, the spray pumps will be activated at 30 seconds after high containment pressure (5 psig). High containment pressure occurs in about 1.7 seconds. A conservative value of 33 seconds was used in the analysis.
(2) Single failure is the Main Steam Isolation Valve (MSIV) failure causing both stea;n generators to blow.down.
(3) Ruptured steam generator blows down through one-half the maTimum 2
areas, or 3.06 ft, as areas larger than this would not give a pure l
i steam blowdown.
(4)
Isolated steam generator blows down through one-half the steam-line 2
area becuase of MSIV flow area restrictions, thus through 3.06 ft.
(5) All three spray pumps and all four fan coolers will be available.
l The single failure is assumed in the MSIV.
4 l
7
(6) The CONTEMPT time TPEAX for the Tagami/Uchida heat transfer correlation was changed to 99 seconds to correspond with the end of blowdown.
(7) Main feedwater is available to each steam generat'or at 1650 lb/sec initially and ramps down to zero flow at 60 sec.
The resulting containment pressure and tempeCature history is shown in Figures 4.5P and 4.5T, respectively. The peak pressure is about 107 psia, which is substantially greater than the 69.7 psia design pressure.
Case 4 The assumptions used for Case 4 were identical to those used in Case 3 except that the ruptured steam generator was allowed to blowdown through the 2
maximum area of 6.12 f t. The resulting containment pressure and temperature predictions are shown in Figures 4.6P and 4.67, respectively. The peak pressur'e is about 106 psia.
Case 5 Cases 3 and 4 assumed that off-site power was available. Case 5 was run to investigate the containment response for the loss of off-site power assumption. Case 5 is similar to Cases 3 and 4 except for two assumptions.
First, because off-site power is lost, the spray pumps are not available until 84 seconds.
In addition, the loss of off-site power results in a complete and immediate loss of feedwater.
Case 5 was based on each steam generator blowing down through a 24-inch diameter break. The pressure and temperature response is shown in Figures 4.7P and 4.7T, respectively. The peak pressure is about 98 psia and a peak containment ate: spheric temperature of 465 F at 70 seconds.
i Case 6 Analyses have also been performed assuming a fix which would prevent the blowdown of both steam generators.
In this case the single failure assumption is loss-of-offsite power with a failure of one diesel generator. The mass and energy release data used in the analysis is for the full power MSLB with one steam generator blowdown. This is discussed by Consumer Power Company in Ref. 1 and provided in Table 4.1.
The assumptions made in the mass and energy release analysis are the following:
1.
A double-ended guillotine rupture of a main steam line inside the containment.
2.
A reduction in feedwater flow from full flow to zero over the 60 seconds immediately following scram at less than 2 seconds on high containment pressure.
3.
Both mainsteam isolation valves would close on low steam generator pressure (500 psia) causing the unrupture steam generator to isolate in eight seconds.
4.
Off-site power was available to 'aximize the rate of energy transfer from the primary to secondary.
5.
Pure steam blowdown (no moisture carryover).
For the containment response calculation the following assumptions were made:
1.
Loss-of-offsite power and failure of one diesel generator.
2.
2 of 3 containment spray pumps available.
3.
Containment spray initiation at 36.7 seconds (200 gpm) and full flow at 52.5 seconds (2680 gpm).
(Reference 2) 2.
Palisades Plant--Proposed Technical Specifications C.. ge Related to Containment Spray Initiation Times. Dockt 50-255 License DPR-20, Nov. 24, 1980. Letter from D. P. Hoffman of Consumer Power Co. to D. Crutchfield of NRC..
4.
1 of 4 air coolers available at 23 seconds.
5.
Tagami/Uchida heat transfer correlation with Tagami peak time at end of blowdown (68 seconds).
The results cf this analysis are the pressure and temperature responses shown in Figures 4.8P and 4.8T.
The calculated peak pressure is 68.5 psia reached at 67 seconds. This is 1.2 psi below design. The calculated peak temperature is 413 F reached at 37 seconds. Therefore, based on this analysis, a fix which would prevent the blowdown of both steam generators would limit the calculated peak pressure to 1.2 psia below design.
Conclusion Based on the results of C. a 1 and 2, the CONTEMPT and RELAp 4 models are adequate for obtaining concainment response to a MSLB. The results of Cases 3, 4, 'and 5 showed that the blowdown of both steam generators will result in containment pressure which exceeds design values. This is regardless of whether off-site power is available. The results of Case 6 showed that a design change which would prevent the blowdown of both steam generators would keep the containment peak pressure within the design limit.
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Table 3.1 e
Palisades Double-Ended Suction Leg Break Blowdown Energy Balance (Million Stu)
J O Seconds Decrease 20.4 Seconds l
D:imary System Coolant Inventory 253.7 246.6 7.1 Steam Gen'erator Coolant Inventory 140.0 2.0 138.0 Secondary Flow to Turbine (I)
-9.2 AccumulatorSystemInventory(2) 19.5 0.7 18.8 Core Stored Heat 18.9 4.6 14.3 Conductor Stored Heat (3) 111.0 3.7 107.3 5.0 Decay and Fission Heat s
543.1 253.4 285.5 Note:
(1) Flow continues until valve is fully closed at six seconds after the break. Energy value is net loss for steam and feedwater flows.
(2) Accumulators and lines at 900F.
(3) Conductors include ali metal transferring heat to the primary coolant system except for the fuel rods..
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Taole 3.2 Palisades Double-Ended Suction Lag Break Reflood Energy Balance (MillionStu) 20 Seconds Decrease 120 Seconds Reactor Coolant System Inventory 22.4
-3.1 25.5 Safety Injection Tank Water (I) 77.6 Saf ety Injection Pump Flow (I) 13.5 CoreStoredHeat(2) 16.2 11.0 5.2 12.3 Decay and Fission Heat 59.3 3.3 56.0 Primary Vessel Walls Primary Vessel Internals 12.4 3.5 8.9 Primary Loop Metal 28.4 1.0 27.4 Steam Generator Inventory 137.9 33.7 104.2 (I.L. 69.8 - 59.1 = 10.7)(3)
Steam Generator Tube Metal 14.3 3.6 10.7 (I.L. 7.6 - 6.3 = 1.1)(3)
TOTALS 290.9 156.4 237.9 Approximate Break Flow Energy 6
S.G. Side 72 (10 ) Stu 6
Pump Side 81 (10 ) Stu 6
Total 153 (10 ) Stu Reference Temperature is 320F Notes:
1 (1) The S.I. water temperature was 3000F to prevent numerical instabilties.
0 Actual value should be 110 F.
(2) Based on ANS + 20% decay heat curve.
(3) Energy from intact loop steam generator.
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Table 3.3 Heat Structures in Palisades Containment Model 2
Structure Area ft Thickness ft 1.
Tanks and piping (.453 inch) 19,332
.0378 2.
Ducts (.10 inch) 20,072'
.0083 3.
Reactor crane (2.35 inch) 6,973
.1958 4.
Internal concrete (33 inch) 9,401 2.75*
5.
Gratings and trusses 20,996
.0144 6.
Containment dome 7,270 3.0217 7.
Containment dome base 11,000 7.75 8.
Containment side wall 50,600 3.5217 j
9.
Storage pool floor and shielded walls 4,456 4.35 l
- 10. Containment base slab 8,229 12.44
- 11. Biological shield wall 2,340 7.8672
- 12. Structura1' support steel 26,320
.45 Deck received had 2.25, which was in error.
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Table 3.4 Palisades Contaimnent - Initial Conditions 0
Outside air temperature 95 F Outside air pressure 14.7 psia' Relative humidity of outside air 0.60 3
Volume of primary capable of holding liquid 3050.3 ft Temperature of primary system vapor region 250 F 0
Temperature of primary system liquid region 250 F 3
Volume of containment 1.6E6 ft 3
Volume of liquid pool in containment sump 10 ft 0
Temperature of containment vapor region 120 F Temperature of containment liquid region 120 F Pressure in containment 14.7 psia Relative humidity in containment 1.0 2
Horizontal cross-sec'.ional area of containment 8,229 ft t
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TABLE 4.1 Main Steam Line Break Mass / Energy Release Data Ruptured Steam Generator Time (Sec)
Lbm/hr Btu /Lbm 0.0 3.266E07 1200.7 0.1 3.266E07 1200.8 0.2 3.186E07 1201.2 0.3 3.106E07 1201.5 0.4 3.037E07 1202.0 0.5 2.957E07 1202.3 0.7 2.826E07 1202.8 1.0 2.637E07 1203.5 1.3 2.501E07 1204.0 1.5 2.409E07 1204.3 1.8 2.283E07 1204.5 2.0 2.215E07 1204.6 2.5 2.067E07 1204.8
~
3.' O 1.942E07 1204.8 4.0 1.759E07 1204.5 5.0 1.621E07 1204.1 7.2 1.370E07 1203.2 8.0 1.336E07 1202.9 10.0 1.233E07 1202.3 15.0 1.062E07 1200.9 20.0 9.592E07 1199.9 30.0 8.221E06 1197.6 40.0 7.193E06 1196.0 45.0 6.852E06 1195.1 50.0 6.566E06 1194.1 54.0 6.280E06 1193.5 60.0 5.938E06 1191.1 63.0 5.481E06 1191.1 68.0 0.0 l _2 l-
TABLE 4.1 (cont'd)
Main Steam Line Break Mass / Energy Release Data Isolated Steam Generator Time (Sec)
Lbm/hr Stu/Lbm 0.0 1.656E07 1200.4 0.1 1.656E07 1200.4 0.2 1.656E07 1200.6 0.6 1.587E07 1201.3 1.0 1.530E07 1201.8 1.3 1.496E07 1202.2 1.5 1.473E07 1202.5 2.0 1.416E07 1202.9 3.0 1.279E07 1203.6 3.5 1.233E07 1203.8 4.0 1.153E07 1204.0 5.0 9.930E06 1204.2 5.4 9.135E06 1204.2 6.0 7.650E06 1204.3 6.4 6.622E06 1204.2 6.8 5.709E06 1204.2 7.2 4.339E06 1204.1 7.8 1.484E06 1203.9 8.1 0.0 1203.8 8.1 0.0 l
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