ML19323H019
| ML19323H019 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/30/1980 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19323H016 | List: |
| References | |
| TASK-03-12, TASK-15-02, TASK-15-2, TASK-RR NUDOCS 8006110187 | |
| Download: ML19323H019 (42) | |
Text
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YANKEE ROWE CONTAINMENT ANALYSIS FOR A MAIN STEAM LINE RUPTURE INSIDE CONTAINMENT s
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Yankee Atomic Electric Company i
Rowe, Massachusetts June 1980 l
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ABSTRACT This report provides a complete analysis of the transient resulting from a Steamline Rupture Inside Containment for the Yankee Rowe Nuclear Power Station. Containment integrity is demonstrqted for this postulated event by showing that the limiting pressure transient resulting from the most severe steamline rupture yields a maximum pressure less than the containment design pressure. This report also provides the limiting containment temperature transient following the most severe steamline rupture for use in containment equipment qualification analyses.
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TABLE OF CONTENTS Page ABSTRACT..................................
11 1.0 I NTROD UCT I ON AND SU MMARY................................ 1 2.0 DISCUSSION OF MAIN STEAM LINE RUPIURE AT YANKEE ROWE.... 3 3.0 BLOWDOWN ANALYSIS.......................................
4 3.1 Anal yt ic al M e t ho d................................... 4 3.2 Initial Conditions and Transient Assumptions........ 5 3.2.1 Genera 1......................................
5 3.2.2 Co re Powe r Le ve 1............................. 5 3.2.3 Initial S te cm Generator Inventory............ 6 3.2.4 Feedwate r Sys tem Ope rational Mode............ 6 3.2.5 B r e ak S i z e................................... 7 3.2.6 Ancilla ry Initial Condi tions................. 7 3.3 RELAP4-EM Blowdown Results..........................
9 3.4 Summa ry of Tot al Blowdown F1uid..................... 10 4.0 CONTAINMENT ANALYSIS....................................
19 4.1 General.............................................
19 l
4.2 Analytical Mode 1....................................
19 4.2.1 Cont ainme nt Vo1ume........................... 19 4.2.2 Heat conducting Structure Modeling........... 19 4.2.3 Heat Trans fe r Coef fici ent s................... 20 4.2.4 Time Step Size...............................
21
- 4. 3 In i t i al Cond i t io ns.................................. 21 4.4 Cont ainme nt Transient Result s....................... 22 4.5 Sensitivity Studies on Analytical and Initial Condition Assumptions...............................
22 5.0 00 L o u JS I O NS............................................. 35
)
6.0 REF E RENC E S....................................,......... 36 APPENDIX A: MODIFIED H FOR CONCRETE EXPOSED TO STEAM..yygg 39
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1.0 INTRODUCTION
AND
SUMMARY
Reference 1) requested information required by NRC with respect to containment environmental conditions and systems required for accident mitigation. The purpose of this report is to provide the containment response following a main steam line rupture inside containment.
A licensing analysis of the containment transient resulting from a main steam line rupture inside containment has not been performed.
However, scoping analyses, as part of the YR Systematic Evaluation Program effort, to determine containment response to a steamline rupture (SLR) have been performed by Yankee. These analyses, based on RELAP4 blowdown anlysis and 00NTEMPT-LT026 containment analysis, indicate that the design containment pressure of 34.5 psig is not exceeded for the most severe steam line rupture.
Additionally, this analysis provides the limiting containment temperature l
transient following the rupture for use in containment equipment qualification analyses.
The results of this analysis were previously reported to the NRC in the following correspondence:
- 1) In support of SEP Equipment Qualification via Reference 2,
- 2) In response to NRC letter concerning non-safety grade equipment qualification via Reference 3,
- 3) In response to IE Bulletin 80-04 (Reference 4) via Reference 5, and
- 4) In numerous conversations with your staff.
The most recent reporting of the scoping analysis (Reference 5) was discussed with your staff (Mr. C. Tinkler) on May 29, 1980. During this conversation, Mr. Tinkler and YAEC (J.R. Chapman) resolved the following:
- 1) YAEC should submit the data pertinent to the scoping analysis discussed in Reference 5, in lieu of the specific information requested in Enclosure (3) to Reference 1,
- 2) Since YAEC is not using a currently approved model.for this scoping analysis (see NUREG-0588) as much detail as possible should be submitted to describe the analysis,
- 3) Any additional information required for NRC to perform independent analysis for either LOCA or main steam line rupture may be submitted on a schedule consistent with NRC need.
The remainder of this report provides the information required by Items (1) and (2). Information discussed in Item (3) above, will be provided as requested from NRC. Additionally, YAEC is currently requesting proposals from a number of vendor / consultants to perform containment analysis for a main steam line rupture. -
4 2.0 DISCUSSION OF MAIN STEAM LINE RUPTURE AT YANKEE R0WE Each of the four main steam lines at Yankee Rowe (YR) has a Non-Return Valve (NRV) in the line outside containment. These valves act as a check valves to preclude reverse flow and also can be manually closed to I
preclude forward or reverse flow.
A steam line rupture inside containment results in the blowdown of secondary fluid from the steam generator connected to the ruptured steam line into the containment atmosphere. Any backflow from the unaffected steam generators would be rapidly terminated by closure of the NRV located outside containment in the intact portion of the ruptured steam line.
Reactor protection would be assured by a number of trips including high containment pressure, low pressurizer pressure and high neutron flux levels.
R -, - _
3.0 BLOWDOWN ANALYSIS 3.1 Analytical Method The blowdown analysis was performed with the Yankee approved version of RELAP4-EM (Reference 6).
The model employed s based on the Small Break LOCA Blowdown model used to license Yankee Rowe Core XIII (Reference 6).
Figure 3-1 provides the nodalization used in the SLR analysis and Figure 1
3-2 provides the Small Break LOCA model on which the SLR model is based.
The following discussion outlines the development from the Small Break LOCA model to the SLR model:
1.
Volumes 18-25 were eliminated since the Emergency Core Cooling System is not modeled in the SLR analysis.
2.
Junctions 18-27 and 30-33 were eliminated for the same reason.
3.
Volume 26 and Junctions 28 and 29 were replaced by fill Junctions 21 and 22.
4.
Junction 35, the negative fill junction for the SG Safety Valves in the SG connected to the ruptured steamline was eliminated.
5.
Junction 34 was renumbered Junction 23.
6.
Junctions 36 and 37 were renumbered Junctions 19 and 20, respectively.
7.
A containment volume, number 18, and Junction number 18 were added to model the break flow from SG Volume 17.
8.
The three heat slabs used to model the core in the Small Break.
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LOCA analysis were reduced to one slab for the SLR analysis.
9.
Two heat slabs were added - one for each of the two SG metal masses (excluding the tube masses which were already modeled explicitly).
The result is an eighteen (18) volume, twenty-three (23) junction, twenty-five (25) heat slab model. Basic analytical inputs necessary to model the transient effectively with RELAP are detailed in Section 3.2.
3.2 Initial Conditions and Transient Assumptions 3.2.1 General To determine the most limiting combination of consistent initial conditions and system operation modes requires a detailed parametric s tudy.
To circumvent an extensive parametric study, the most limiting set of conditions covering the full range of cperating conditions from zero to full power was assumed coincident for the transient. The following discussion provides the basic assumptions and the reasoning behind utilizing the coincident philosophy approach.
3.2.2 Core Power Level A power level of 112 percent of 618 MWt (the trip setpoint for high neutron flux including uncertainty) was used and held constant until the occurrence of a trip on high containment pressure. This assumption encompasses all possible break sizes that could yield an increase in core power level (due to moderator cooldown in the presence of a negative moderator tempe ?ature coef ficient of reactivity) without tripping the plant.
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This rational is not overly conservative since the overal.1 results (i.e.,
containment maximum pressure) are not strongly dependent on core power level since the higher core power level serves only to maintain the RCS fluid passing through the SG tubes warmer than would be expected, and hence by creating a larger potential for heat transfer from the primary to secondary yield a slightly faster SG blowdown.
3.2.3 Initial Steam Generator Inventory Yankee Rowe's SG inventory varies from 20,000 ihm at 100 percent power to 31,500 lbm at zero power. Using the maximum value of 31,500 lbm is the most conservative assumption especially in combination with assuming the core power level remains at the high neutron flux trip setpoint until the containment trip signal.
3.2.4 Feedwater System Operational Mode Detailed procedures are followed which define the number of feedwater pumps and flowpath to the SGs (i.e., through the main fe2dsater piping or through bypass piping) used as a function of plant power level. However, being a bounding analysis, full system operation was assumed with feedwater assumed to be delivered only to the SG connected to the ruptured cteamline.
By making this assumption, the simulation is a low power level case in which a low rate of feedwater flow is being delivered to the four SGs initially and when the break occurs the Feedwater Control Valve in the feedline connected to the ruptured SG goes full open. This results in the maximum possible feedwater flow to the SG connected to the ruptured steamline.
A calculation to determica the maximum feedwater flow to one SG in coincidence with zero flow to the remaining three SGs was performed assuming 1.
the feedwater system was in the maximum flow operational mode (i.e., all three Boiler Feed Pumps operational) and the results used in the RELAP blowdown analysis.
This feedwater system operational assumption was reviewed in response to IE Bulletin No. 80-04 (Reference 4) concerns and the results of the review reported to NRC via Reference 5.
In Reference 5, YAEC committed to an equipment and a procedurc change that ensures the applicability of the feedwater system assumptions made in this report.
3.2.5 Break Size The largest break size (a double-ended guillotine of the main steamline) yields the fastest blowdown rate and subsequently the most severe containment transient. The break size explicitly modeled with RELAP is the double ended severence of a main steamline (14 inch Schedule 80 piping 1
2 with an internal cross-sectional flow area of.8522 f t ),
3.2.6 Ancillary Initial Conditions In addition to the major assumptions discussed in the preceding sections, numerous other initial conditions and assumptions were required.
1hese remaining boundary conditions are discussed below:
1.
An infinite bubble rise velocity was used in the SGs to preclude moisture carryover and thus maximize the energy content of the blowdown fluid.
2.
RCS pumps were not tripped thus allowing a maximum heat transfer rate from the RCS to the ruptured SG. -.
3.
RCS temperatures were based on the maximum Technical Specification allowable core inlet temperature of 515 F + 4 F 0
uncertainty (i.e., 5190F) to yield the maximum possible stored RCS energy.
t 3
4.
SG secondary initial temperature was set at 519 F, the maximum 0
possible value, to yield the maximum possible stored energy and initial pressure.
5.
Feedwater temperature was maintained at the maximum expected temperature of 3540F.
6.
Nucleate boiling heat transfer from the SG tubes to the secondary was assumed to occur for the entire blowdown period. This assumption maximizes the heat transfer rate to the secondary and consequently the blowdown rate. The Thom nucleate boiling correlation was used to conservatively model this heat transfer mode.
7.
No credit was taken for the reduced heat transfer rate with uncovering of the SG tubes as the SG inventory decreased. The steam generator tubes were at all times assumed covered by a l
two phase fluid thea assuring Thom nucleate boiling heat transfer for the entire transient. This was assured in the RELAP model by inputting the SG tube height equal to a small value (0.02 ft).
. 8.
The full heat transfer area of the tubes was assumed in the analysis with no credit taken for any SG tube fouling. 1
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9.
Feedwater termination was conservatively modeled as occurring 10 seconds af ter reactor trip.
(This feedwater system operational assumption was reviewed in response to IE Bulletin No. 80-04 (Reference 4) concerns and the results of the review reported to NRC via Reference 5.
In Reference 5, YAEC committed to an equipment and a procedure change that ensures the applicability of the feedwater system assumptions made in this report.)
- 10. Throughout the transient, the steam flow from the intact SGs to the turbine was conservatively assumed to be zero.
Table 3-1 provides a summary of the initial conditions and Table 3-2 provides a summary of the assumptions made in the analysis.
3.3 RELAP4-EM Blowdown Results Table 3-3 provides a summary of the pertinent results. The time of high containment pressure trip signal actuation at 5.0 psig was conservatively determined in an iterative process based on an initial containment pressure of 0.0 psig by first running the blowdown analysis without trip to determine the blowdown rate. Using this blowdown rate, the time to reach 5.0 psig was calculated to be less than 5 seconds. This value was then conservatively doubled to yield a signal occurrence of 10.0 seconds. The blowdown analysis was then rerun with a reactor trip at 11 seconds (includes I second signal delay time). Table 3-4 provides the blowdown rate and enthalpy as a function of time for this analysis.
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3.4 Summary of Blowdown Fluid To determine the entire blowdown fluid for input into the CONTEMPT-LT analysis, it is necessary to account for each of the components of the total blowdown shown in Table 3-5.
The first four components were modeled by initializing the SG inventory at a value in excess of the sum of these four components. The feedwater was explicitly modeled by a fill junction as described in Section 3-1.
The final two components of the total blowdown were accounted for by conservatively including the additional blowdown mass in the blowdown rate in the firs t 2.1 seconds of blowdown. Table 3-6 provides the final blowdown rates and enthalpy for use in the CONTEMPT-LT containment analysis. '
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TABLE 3-1 1
I Blowdown Analysis Initial Conditions i
Parameter Value i
Core Power Level (percent)
,e 112 SG Inventory (Ibm) 31,500-Core Inlet Temperature (OF) 519 SG Secondary Temperature (OF) 519 Feedwater Temperature (OF) 354 1
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TABLE 3-2 Blowdown Analysis Assumptions 1.
Feedwater System Operational Mode - three boiler feed pumps operational 2
2.
Break Size - double-ended guillotine (.8522 f t )
6
-1 3.
SG Bubble Rise Velocity - 10 ft-sec 4.
RCS Main Coolant Pumps - remain active 5.
SG Tube -to Secondary Side Heat Transfer Mode - nucleate boiling via Thom correlatiot, throughout transient -
6.
SG Tube Area - the entire SG tube area was assumed active 7.
Reactor Trip Signal - occurs 1.0 second af ter. containment signal 8.
Feedwater Termination - occurs 10.0 seconds af ter reactor trip signal 9.
Flow to Turbine - zero at all times J
1 4 -
TABLE 3-3 Blowdown Analysis Results Event Time (sec)
Rupture occurs 0.0 Feedwater flow to o
SG connected to ruptured steamline increases to runout value 0.0 Flow to turbine goes to zero 0.0 Containment high pressure signal occurs at 5.0 psig 10.0 Reactor trip signal occurs 11.0 l
Feedwater flow to SG connected 2
to ruptured steamline goes to zero 21.0 SG connected to ruptured steamline empties 112.0.-
TABLE 3-4 Blowdown Rate and Enthalpy Resuits for the Guillotine SLR Analysis Time (sec)
Blowdown Rate (Ibm-sec )
Enthalpy (Btu-lbm-1) l O.00 1404.0 1199.1 0.04 1395.6 9
1199.3 0.10 1381.8 1199.5 0.20 1355.8 1199.9 0.30 1331.0 1200.4 0.40 1307.2 1200.7 0.50 1284.4 1201.0 1.10 1171.6 1202.4 2.10 1018.0 1203.9 3.10 902.6 1204.5 4.10 812.6 1204.7 5.10 741.4 1204.7 6.10 684.7 1204.5
/.10 639.5 1204.2 8.10 603.0 1203.9 9.10 573.3 1203.6 10.1 549.3 1203.3 11.1 529.7 1203.0 12.1 513.7 1202.7 14.1 489.6 3202.3 16.1 471.6 120>.9 18.1 455.1 1201.6
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19.1 447.0 1201.4 24.0 415.8 1200.5 30.0 38 8.7 1199.6 36.0 368.3 1199.1 40.0 357.4 1198.6 46.0 343.2 1198.0 50.0 334.4 1197.6 56.0 321.5 1197.1 60.0 312.7 1196.7
/0.0 29 5.0 1195.8
/8.0 282.2 1195.0 88.0 266.5 1194.1 96.0 255.0 1193.4 102.
246.6 1192.8 110.
236.0 1192.0 112.
0.0 1192.0 7
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TABLE 3-5 Total Blowdown Fluid (Ibm)
Total zero power liquid inventory 31500 Total zero power vapor inventory 760 I
Feedwater that enters SG following feedwater isolation due to elevation 860 differences in piping height to nozzle location Feedwater that enters SG following feedwater termination due to flashing i
of feedwater when the SG pressure decreases below the feedwater saturation pressure 1060 21 seconds of feedwater flow at 350 l
lbm-sec-1 7350 Mass of steam between SG nozzle and check valve 425 Flow through check valve prior to closure 790 Total fluid available to enter containment 42745 Minus vapor remaining in SG
-40 Resulting tota) fluid entering containment 42705 TABLE 3-6 Total Blowdown Rates and Enthalpy Time (sec)
Flow Rate (Ibm-sec-1)
Enthalpy (Btu-lbm-1) 0.00 2193.0 119y.1 0.04 2184.6 1199.3 0.10 2170.8 1199.5 0.20 2144.8 1199.9 0.30 2120.0 1200.4 0.40 2096.2 1200.7 0.50 2073.4 1201.0 1.10 1960.6 1202.4 2.10 1018.0 1203.9 3.10 902.6 1204.5 4.10 812.6 1204.7 5.10 741.4 1204.7 6.10 684.1 1204.5
/.10 639.5 1204.2 8.10 603.0 1203.9 9.10 513.3 1203.6 10.0 549.3 1203.3 11.1 529.7 1203.0 12.1 513.7 1202.7 14.1 489.6 1202.3 16.1 471.6 1201.9 18.1 455.1 1201.6 19.1 447.0 1201.4 24.0 415.8 1200.5 30.0 388.7 1199.6 36.0 368.3 1199.1 40.0 357.4 1198.6 46.3 343.2 1198.0 50.0 334.4 1197.6 56.0 321.5 1197.1 60.0 312.7 1196.7 70.0 29 5.0 1195.8 78.0 282.2 1195.0 88.0 266.5 1194.1 96.0 255.0 1193.4 102.
246.6 1192.8 110.
236.0 1192.0 112.
0.0 1192.0 I
Figura 3-1 StGamlina Ruptura RELAP4-EM Blowdown Model J
18 CONTAIN-FEED SAFETY TO MENT 5
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FIGURE 3-2 Small Break LOCA RELAP4-EM Blowdown Model t@
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4.0 C0ffrAINMEffI ANALYSIS 4.1 General This section describes the analysis of the containment transient
- resulting from the blowdown transient defined in ethe previous sections.
The analysis was performed utilizing the YAEC version of the CONTEMPT-LT computer code as documented in References 7 and 8.
4.2 Analytical Model 4.2.1 Containment Volume The best estimate net free containment volume was assumed in the
- model, i.e.,
860,000 cubic feet (Reference 10).
4.2.2 Heat Conducting Structures Modeling The structural heat sinks (Reference 10) used in the model are provided in Table 4-1.
The values of exposed surface area used in Reference 14 included a +5% conservatism since the analysin was used as input to
)
Emergency Core Cooling System performance analyses. This 5% conservatism was removed and the nominal "Best-Estimate" values were used in the model as shown in Table 4-1.
Excepting the containment sphere (heat conducting structure number 1), all heat conducting structures were modeled as symmetrical slabs exposed on both sides to the containment vapor space.
The exposed surface area provided in Table 4-1 is the total for both sides, and the thickness is the half-thickness for these slabs. The containment sphere, which is exposed to the outside atmosphere on its outer surface and containment vapor space conditions on the inner surface, was modeled full thickness and the values shown in Table 4-1 are the nominal full thickness of the shell and area of one exposed surface only.
All painted surfaces are modeled as having a.024 inch thick paint layer which is conservatively thick based on actual measurements taken at the plant. The thermophysical properties used in the analysis are listed in Table 4-2.
Table 4-3 provides the mesh structure used for each heat sink.
Heat conducting Structures 4, 5 and 6 from Table 4-1 were lumped into a single heat sink, Structure No. 4 in Table 4-3.
This is possible since the structures are similar painted concrete symmetrical slabs with uniform initial temperatures exposed to the vapor space and with half-thicknesses greater than the expected 24-hour thermal penetration distance (<4.5 feet in concrete). The exposed surface area of the lumped slab is equal to the total for both sides of all the component structures. The mesh spacing was chosen to comply with recommendations given in References 7 and 9.
4.2.3 Heat Transfer Coefficients For the outer surface of the containment exposed to the natural environment, the CONTEMPT-LT calculated Macadams (Reference 13) turbulent natural convection correlation was used. As a boundary condition for all internal heat conducting surfaces exposed to the vapor space, the CONTEMPT-LT calculated Uchida condensing heat transfer coefficient (Reference 12) vas used. To conservatively apply the Uchida correlation to the conceste structures, the exposed surface area for the conceete structuess was reduced by 40%.
To maintain the same thermal diffusivity relative to the true mass of the concrete, the thermal conductivity and heat capacity of the concrete !
structures were adjusted as discussed in Appendix A.
The condensate " Dropout" dial was set at 100% since this results in a more limiting transient than assuming the same quantity of energy is removed from the vapor space without " Dropout" of condensed liquid until the entire steam mass is saturated.
4.2.4 Time 5tep Eize Time step values were chosen sufficiently small to assure solution convergence throughout the analysis. As recommended in Reference 7, the convergence parameter felta E/E edited in the CONTEMPT-LT output format was kept below 0.1%.
The time steps used are provided in Table 4-4.
4.3 Initial Conditions The initial conditions used were selected to yield a conservative calculation of the maximum resulting temperature and pressure transient.
Table 4-5 provides a listing of the initial conditions used. Note that two base case analyses were performed - the first case to determine the maximum pressure transient - the second to determine the most limiting temperature transient. The maximum pressure transient is attained for the maximum initial containment pressure ard temperature situation while the most limiting temperature transient is attained for the maximum initial containment temperature and minimum initial pressure. The limiting values assumed are based on Technical Specification limits as stated in Table 4-5.
Table 4-6 provides typical values for the containment parameters of pressure, temperature and relative humidity to indicate the conservative approach used.
4.4 Containment Transient Results Figures 4-1 through 4-4 provide the containment pressure and temperature transients resulting from the blowdown data contained in Table 3-6.
Figures 4-1 and 4-2 provide the results of,the analysis performed to determine the limiting pressure transient which has a peak of 31.7 psig.
Figures 4-3 and 4-4 provide the results of the analysis performed to determine the limiting temperature transient which has a peak of 3690F.
4.5 Sensitivity Studies on Analytical and Initial Ccndition Assumptions To determine the sensitivity of the containment transient to changes in analytical and initial condition assumptions a parametric study was pe rf o rmed. Table 4-7 provides a list of the parametric a =1yses performed, while Table 4-8 provides a detailed description of each of the parameters examined.
The results of the parametric studies show that the analytical assumptions made in the two base case analyses are conservative with respect to maximizing the pressure and temperature transient.
1 TABLE 4-1 Containment Structural Heat Sinks Heat Sink Material Exposed Surface Material Thickness Identification Composition Area (f t2)
(ft) 1.
Containment Painted Carbon Steel 49087.4 Sphere Paint
.002 Carbon Steel
.0833 Paint
.002 2.
Painted Concrete 505.7 Paint.
.002 Concrete 1.50 3.
Painted Concrete 8820.9 Paint
.002 Cone ete 2.0 4.
Painted Concrete 7196.2 Faint
.002 Concrete 4.50 5.
Painted Concrete 17278.0 Paint
.002 Concrete 5.0 6.
Painted Concrete 1730.5 Paint
.002 Concrete 6.0 7.
Unpainted Stainless 980.95 0.10 Steel
TABLE 4-2 Thermophysical Materials Properties Conductivityl Volume Heat Capacity 2 Paint 3 0.125 43.0 Carbon Steel 4 26.0 53.6 Stainless Steel 4 10.0 58.6 Concrete 4 0.8 19.04 Concrete 5 1.33 31.73 1.
BTU /HR-FT OF 2
2.
BTU /FT _op 3.
Reference No.11 4.
Reference No.10 5.
V:Alfied to reduce HFILM, See Section 4.2.2 and Appendix A 4
1 4
9,- -.
4 TABLE 4-3 HEAT CONDUCTING STRUCTURE NODALIZATION Structure No.1 Containment Shell LB/.002 ( 1 ) /.002 ( 1 ) /.00264 (30 ) /. 002 ( 1 ) /.002 ( 1 )/RB Structure No. 2 Painted Concrete LB/.002 ( 1 )/.002 ( 1 )/.006 (2 )/.016 (3 ),.04814 )/.072(2 )/.108 (4 )/.111 (6 )/C Structure No. 3 Painted Concrete LB/.002 ( 1 )/.002 ( 1 ) /.006 (2 )/.016 (3 ) /.048 (4 ) /.072 (2 ) /.108 ( 4 )/.117 ( 10 ) /C Structure No. 4 Painted Concrete LB/.002 (1 )/.002 ( 1 )/.006 (2 ) /.016 (3 ) /.048 (4 ) /.072 (2 ) /.108 (4 ) /.130 (10 )/
.118(20)/C Structure No. 5 Stainless Steel LB/0.00833(12)/C xxx(y) xxx = Mesh size, ft, y = # of mesh spaces LB = Left Boundary RB = Right Boundary C = Centerline L
TABLE 4-4 1
i i
Time Step Size Time Interval (sec)
Time Step Size (sec)
I j
0 - 200 0.1 200 - 600 1.0 4
600 - 1200 5.0 4
1200 - 3600 10.0 9
1 3600 - 86400 25.0 i
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f 4
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f TABLE 4-5 Containment Data for Main Steanline Rupture Analysis Parameter Value Containment Net Free Volume 860,000 ft3 Initial Containment Pressure 0.0 psig for Limiting Temperature Transient Initial Containment Pressure 3.0 psig(3) for Limiting Pressure Transient Initial Containment Temperature 1200F Relative Humidity 1%(2)
Outside Att. spheric Temperature 100 F Initial Containment Pool Water 3(4)
Volume 10.0 ft 1 Maximum allowable Technical Specification Value 2 Less than any observed value 3 Greater than expected environmental value i
4 Necessary in order to force CONTEMPT-LT to use the correct tempe rature for the fluid condensate along the heat slabs. This requirement results from the use of the Uchida condensing-steam film heat transfer coef ficient as a boundary condition for the exposed surfaces. Omission of this volume yields erroneous results.
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TABLE 4-6 Typical Yankee Rowe. Containment Operating Conditions Parameter Value Containment Pressure 0.8-2.2 psig e,
0 600 -90 F F
Containment Temperature Relative Humdity 20%-60%
1 i
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PARAMETERS
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- "g CASE PCONT TCONT REL CONCRETE CONCRETE DAW 2 FAC CHTC CMTC PMAX TMAX INITIAL INITIAL HUMID SURF AREA K+ Cp (PSIG)
( F) mn (psig)
(LF)
(%)
yk
$E rr 1
3.0 120 1
- l.0
- l.0 1.0 1.0 1.0 1.0 31.5 350
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2 3.0 120 1
- .6
.6 1.0 1.0 1.0 1.0 31.7 351 0 0.
"a a
b 3
3.0 120 1
- .6 t.6 0.0 1.0 1.0 1.0 31.7 351 E."
D i
OE N
4 3.0 120 100
- .6
.6 0.0 1.0 1.0 1.0 31.1 347
" U, c
0h b
5 3.0 120 100
- .6 f.6 1.0 1.0 1.0 1.0 31.1 347 N"
6 3.0 120 1
- .6 f.6 1.0 0.0 1.0 1.0 26.1 260
$8 En 7
0.0 120 1
- 6 f.6 1.0 1.0 1.0 1.0 28.2 3'69 8
0.0 120 1
- .6 f.6 1.0 1.0 0.0 0.0 28.2 368
>r EE 9
0.0 120 1
- l.0
- 1.0 1.0 1.0 1.0 1.0 27.9 366 6
7, 10 0.0 120
)
- l.0
- l.0 0.0 1.0 0.0 0.0 27.9 366 0
\\
TABLE 4-8 Description of Parameters Studied in the Analyses Provided in Table 4-7 1.
"P
" is the initial containment pressure.
OONT INITIAL 2.
"TOONT INITIAL" is the initial containment te,mperature.
3.
"REL HUMID" is the initial containment relative humidity.
4.
" Concrete Surf Area" - This parameter refers to the fraction of actual concrete material surface area assumed in the analysis (refer to Section 4.2.2).
5.
" Concrete K and C " - This parameter refers to the fraction of actual concrete material Ehermal conductivity (K) and heat capacity ( C )
p used in the analyses (refer to Section 4.2.2).
6.
" DAW 2" - This parameter is a CONTEMPT-LT dial to select use of evaporation - condensation model in drywell prior to end of blowdown:
0.0 indicates use of model,1.0 bypasses model.
7.
"FAC" - Fraction of Uchida Condensate transferred to the pool.
J 8.
"CHTC" - Film heat transfer coef ficient multiplier for sensible heat transfer between the liquid pool and vapor region.
9.
"CMTC" - Mass transfer multiplier for evaporation model.
10.
"PMAX" is the maximum containment transient pressure attained.
11.
'TMAX" is the maximum containment transient temperature attained.
=
9
-3 0-
UUNIH1NMENI RESl>0NSE 10 S TE AM L I NE RUP TURE aua ao. ncox l
oo O
C)-
v PNRX = 31.7 PSIG TMAX = 351 DEG F oo 9
cs Ra-mW a
Vapor Space L
ap x1 "
o o
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t i
1E-01 IE+00 IE+01 IE+02 IE+03 IE+04
'lE+05
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CONTAINMENT
RESPONSE
IU S I l-H N L 1 N t; K U t' I U K t.
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1E-01 1E+00 1E+01 1E+02 IE+03 IE+04 IE+05 1
SECONDS
Figura 4-3 Contcinment Pracsura Respen03 for Limiting Tcmperctura Analysis u
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~u
5.0 CONCLUSION
Containment integrity following the most severe postulated steamline rupture is assured since the maximum resulting containment pressure of 31.7 Psig is less than the containment design pressure,of 34.5 psig. Figure 4-4 provides the limiting containment temperature response following the most severe SLR for use in containment equipment qualification analyses.
I l
1 t
I '
- ,.o
6.0 REFERENCES
1)
USNRC letter to YAEC, dated March 28,1980, " Environmental Qualification of Electrical Equipment".
2)
YAEL letter to USNRC, dated November 27, 1978, WYR 78-103, " Systematic l
Evaluation Program (SEP)".
3)
YAEC letter to USNRC, dated October 9, 1979, WYR 79-145, " Potential Unreviewed Safety Question Regarding Safety Function Interactions with Non-Safety Grade Systems".
4)
USNRC letter to YAEC, dated February 8,1980.
5)
YAEC letter to USNRC, dated May 8,1980, WYR 80-50 " Response to IE Bulletin No. 80-04".
6)
Proposed Change No.125, Supplement No. 5, " Yankee Rowe Core XIII ECCS Performance Analysis", dated August 1,1977, Appendix A.
7)
L. L. Wheat, R. J. Wagner, C. F. Niederauer, and C. F. Obenchain,
" CONTEMPT-LT: A Computer Program for Predicting Containment Pressure -
Temperature Response to a Loss-of-Coolant Accident", ANCR-1219 (June, 1975).
8)
" Version 26 Modifications to the CONTEMPT-LT Program", Report No. SRD-83-76, April 1976, W. J. Mines, System Research Division, INEL, Aerojet Nuclear Company.
9)
"R. N. Gupta, Containment Pressure Analysis Model Using CONTEMPT-LT, TAEC-1086, September,1975.
e, o
- 10) Stone and Webster Engineering Corporation. Letter from S. Frank to A. E. Ladieu, dated April 8,1974, " Containment Data Required for ECCS Evalua tion".
- 11) Stone and Webster Engineering Corporation Letter from S. Frank to John W. Heard, MYAPCo., dated May 14, 1976 Attachment 1, " Containment Data Maine Yankee Nuclear Plant".
12)
H. Uchida, A. Oyama, and Y. Fogo, " Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors", ex. Proceeding of the Third International Conference on the " Peaceful Uses of Atomic Energy", held in Geneva 31 August - 9 September 1964, Vol.13, New York, United Nations, 1965, pp.93-104, (A/ CONF 28/p/436).
13)
W. H. Macadam, Heat Transmission, Chapter 7, pp. 172-174.
14)
R. Byron Bird et. al, Transport Phenomena, John Wiley and Sons, Inc.,
pp. 353-354, 1960.
t APPENDIX A Modified HpIng For Concrete Exposed to Steam From Reference 7, the following heat conduction equation solved by CONTEMPT-LT was obtained.
hA (T,- T ) " 3 (T1 - T ) + pCpA 2
1 where:
h
=HFILM = heat transfer coefficient between vapor and surface A
= surface area T,
= vapor temperature T1
= surface temperature k
= material thermal conductivity AX
= material mesh size T2
= temperature of 2nd node from surface node p
= material density Cp
= material heat capacity t
time The amount of heat conducted through the film can be effectively reduced by 40 percent by decreasing the left-hand side of the above equation by h'
.6*h.
Since h cannot be conveniently modified, the input area (A) is modified by A' =.6*A.
However, A also occurs on the right-hand side of the equation, which should not be modified. Therefore, in order to restore the right-hand side to its original form, K and pCp must be modified by: kl = k/.6 and (p Cp)1 = pCp/.b.
Therefore, the resulting equation is the following.
ll 1 - T ) + (pCp)1 Al AX dT1 l (T, - T ) = k A (T hA 1
AX 2
7