ML19263D735
| ML19263D735 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/30/1979 |
| From: | Hyman R, Krajicek J COMMONWEALTH EDISON CO., SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19263D731 | List: |
| References | |
| XN-NF-79-24, NUDOCS 7904130202 | |
| Download: ML19263D735 (28) | |
Text
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XN NR9 24 i
ECCS EVALUATION OF DRESDEN-1 USING THE EXXON NUCLEAR COMPANY WREM NON-JET PUMP EVALUATION MODEL LARGE BREAK EXAMPLE PROBLEM I
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ECCS EVALUATION OF DRESDEN-1 USING THE EXXON NUCLEAR COMPANY WREM NON-JET PUMP EVALUATION MODEL LARGE BREAK EXAMPLE PROBLEM Prepared by R. D. Hyman J. E. Krajicek Concur: / hh/d a /4 *-79 Nucle @ar Safety Engineering K. P lbraith, Manager
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G. A. Sofer, Managgfr' Nuclear Fuels Engiheering I/ "
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G. J. Busselman, Manager Contra PerformJnce Ym3 Approved
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ERON NUCLEAR COMPANY,Inc.
NUCLEAR REGUL ATOR Y COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THl$ DOCUMENT PLE ASE READ CAREFULLY This technical report was denved through research and development programs sponsored by E x xon Nuclear Company, Inc.
It is treing sub-mittel by Exxon Nuclear to the USNRC as part of a technical contri bution to facilitate safety analyses by licensees of the USNRC which 3
utibie Exxon Nuclear fabricattuj r elos t fuel or other technical services providest by Exxon Nuclear for liaht water power reactors and it is true arx1 correct to the best of Exxon Nuclear's knowlair, information, aruj behef. The information contained herrn may be usnt by the USNRC in its review of this report, and by liwnsees or apphrants before the USNRC which are customers of Ex xon Nuclear in their demonstration of comphance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf.
A.
Makes any warranty, express or emphewf, with respect to the acc ar zu:y, completeness, or usefulness of the infor mation containeu
- this docummt, or that the use of any information, apparatus, method, or process disclosal in this document will not ininnge privately owned rights, or B.
Assumes any liabihties with respect to the u.e of, or for darrages resulting from the use of, any information, ao para tu s, me t hod, or process disclosed in th's document.
I XN-NF F 00, 760 I
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I
-i-XN-NF-79-24 TABLE OF CONTENTS Page l.0 INTRODUCTION AND
SUMMARY
1 2.0 SYSTEM MODEL FOR THE DRESUEN-1 ECCS ANALYSIS.
4 2.1 MODEL DE$CRIPTION.
4 2.2 BLOWDOWN RESULTS 5
3.0 HEATUP MODEL FOR THE DRESDEN-1 ECCS ANALYSIS.
20
=
3.1 HEATUP MODEL DESCRIPTION 20 3.2 RF90 TS OF THE HEATUP ANALYSIS 20
4.0 CONCLUSION
S 22
5.0 REFERENCES
23 S
e e
5 E
-ii-XN-NF-79-24
[
LIST OF TABLES Table No.
Page l.1 ANALYSIS RESULTS OF EXAMPLE PROBLEM 9.0 DEG RECIRC INLET) ENC 6x6 RELOAD FUEL AT BOL...
2 1.2 LARGE BREAK RESULTS - TIME SEQUENCE OF EVENTS......
3 2.1 CORE DATA AT FULL POWER, WITH ENC FUEL 6
_e 2.2 THERMAL AND HYDRAULIC SYSTEM DATA..
7 m
N
i
-iii-XN-NF-79-24 LIST OF FIGURES Figure No.
Page 2.1 BLOWDOWN SYSTEM N0DALIZATION FOR DRESDEN-1 8
2.2 H0T CHANNEL N0DALIZATION...
9 2.3 PRESSURE IN REACTOR UPPER PLENUM...
10 2.4 TOTAL MASS FLOW OUT OF THE BREAK..
11 2.5 MIXTURE LEVEL IN STEAM DRUM 12 2.6 AVERAGE CORE INLET MASS FLOW.
13 2.7 AVERAGE CORE EXIT MASS FLOW 14 2.8 HOT CHANNEL INLET MASS FLOW 15 2.9 HOT CHANNEL EXIT MASS FLOW.
16 2.10 HEAT TRANSFER COEFFILIENT IN HOT NODE OF HOT CHANNEL 17 2.11 HIGH PRESSURE COOLANT INJECTION FLOW.
18 2.12 LOW PRESSURE CORE SPRAY FLOW 19 3.1 ROD AND CANNISTER TEMPERATURES VERSUS TIME 21
__. XN-NF-79-24
1.0 INTRODUCTION
AND
SUMMARY
This document presents an application of the ENC Non-Jet-Pump BWR (NJP-BWR) WREM-3ased(1,2) ECCS model for the Dresden-1 nuclear plant.
The hypothesized Loss-of-Coolant Accident (LOCA) was a guillotine con-figuration with a discharge coefficient of 1.0 and each break area equal to the pipe cross-sectional area. The break was assumed to occur in one recirculation inlet pipe near the reactor vessel.
The analyses involved calculations using the ENC NJP-BWR ECCS evaluation model(I)
Specifically, the following codes were used: RELAP4-EM/ ENC 28E for the blowdown analysis and HUXY/MAY77 for the heatup calculation. The criteria and assumptions used in the evaluation of the LOCA are those listed under Section 50.46 and Appendix K of Title 10 of the Code of Federal Regulations Part 50.
Principal results of these analyses for the recirculation inlet break are contained in Tables 1.1 and 1.2.
A Peak Cladding Temperature (PCT) of 1992 F and a maximum local Zr/H 0 reaction of less than 3.9% were cal-2 culated. These analyses were performed at 714 MWt which is 102 percent of rated power as indicated in Table 1.1.
Removing the Appendix K re-quired two percent core power uncertainty factor, the analysis supports operation of the plant with a Maximum Average Planar Linear Heat Genera-tion (MAPLHGR) of 11.58 kW/ft for ENC 6x6 reload fuel.
Table 1.2 presents summaries of the transient times calculated for the major events. Additional results of this analysis are presented in subsequent sections of the report.
These results indicate that full power operation with a MAPLHGR of 11.58 kW/ft will be achievable white satisfying 10CFR50.46 and Appendix K criteria when the ENC NJP-BWR ECCS model is applied to the Dresden-1 nuclear plant.
E E
- XN-NF-79-24 E
TABLE 1.1 Analysis Results of Example Problem (1.0 DEG Recirc Inlet)
ENC 6x6 Reload Fuel at BOL Analysis Results 1.0 DEG RI Peak Clad Temperature, F 1992 m
f Local Zirconium - Water Reaction, %
3.9%
Core Wide Zirconium - Water Reaction, %
<l%
E Calculation License Core Power, MWt 700 Power Used for Analysis, MWt 714 g
F xF.
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MAPLHGR, KW/FT 11.58**
- Axial (F ) times radial (F ) nuclear peaking factor E
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-- XN-NF-79-24 TABLE 1.2 Large Break Results - Time Sequence of Events Event Time, sec Start 0.0 Initiate Break 0.05 Safety Initiation Signal 1.9 Primary Steam and Feedwater Flows Stop 10.1 Steam Drum Empties 20.0 High Pressure Coolant Injection Valve Starts 18.9 to Open Rated High Pressure Coolant Injection Spray 27.9 Reached Low Pressure Core Spray Valve Starts to Open 51.5 Low Pressure Rated Core Spray Reached 59.0
-. XN-NF-79-24 2.0 SYSTEM MODEL FOR THE DRESDEN-1 ECCS ANALYSIS 2.1 SYSTEM MODEL DESCRIPTION The system blowdown was modeled using the RELAP4-EM/ ENC 28E*
computer code. The Dresden-1 plant was modeled with a total of 54 control volumes, 66 junctions and 50 heat conducting slabs.
Figure 2.1 shows the system nodalization diagram and identifies key components.
This model includes the steam drum, all recirculation piping, reactor vessel, reactor core and four steam generators.
The high pressure coolant injection and low pressure core spray systems were modeled as fill junctions into the reactor vessel upper plenum. Primary and secondary steam and feedwater flows were also modeled as fill junctions. The reactor core power is cal-culated by the RELAP4-EM solution of the space independent core kinetics equations with radioactive fission product decay energy (ANS +20%) and actinide contributions. The pump performance curves used were the homologous pump curves which are integral to the RELAP4 code.
The RELAP4 homologous pump data was used based on comparisons with Dresden-1
~
plant recirculation pump data. The reactor was divided into a single hot assembly and an average assembly.
This is the same type of core representa-tion that was used for previous NJP-BWR model applications.
The hot channel response was calculated initially as a part of the system blowdown analysis.
Further, hot channel analyses will use the model shown in Figure 2.2.
Table 2.1 summarizes the core data. The Dresden-1 thermal-hydraulic system data are shown on Table 2.2.
An 18" recirculation inlet pipe with a double-ended guillotine break was chosen for the example problem.
The total pipe break
- RELAP4-EM/ ENC 28E differs from version RELAP4-EM/ ENC 28B, which has been re-viewed by NRC, only due to minor revisions in the plotting, editing, and input routines. Analytical models and calculated results remain unchanged from the RELAP4-EM/ ENC 288 version.
l XN-NF-79-24 2
area is 2.836 ft.
Loss of offsite power is assumed to occur simultaneously with the recirculation pipe break.
No credit was taken for the emergency condenser.
2.2 BLOWDOWN RESULTS The results from the application of the ENC NJP-BWR ECCS s ' stem blowdown model to the Dresden-1 nuclear plant are shown in Figures E 3 through 2.12, and key events are summarized in Table 1.2 After the break, the recirculation pumps coast down, feedwater pumps coast down and the main turbine condenser loses vacuum.
Primary and secondary feedwater flow is shut off by check valve closure when the feedwater pumps coast down.
Primary and secondary steam flows are shut off by turbine bypass valve closure coincident with loss of vacuum in the main condenser. High con-tainment pressure (2 psig) gives a signal to start the diesel generator to supply power to the core spray pumps.
There is a 10 sec delay for the diesel generator to come up to speed after which the HPCI and LPCS pumps l
are allowed to start.
1
E
~ XN-NF-79-24 Table 2.1 2
Core Data At Full Power, With ENC Fuel Parameter Value Licensed Core Power, MW 700 Power Used for Analysis. MW (102%)
714 Axial Power Peaking Factor (Chopped Cosine Profile) 1.55 Radial Power Peaking Factor 1.62
]
Fuel Length, In 108.25 Fuel Pellet Diameter, In
.4610 Cladding ID, In
.5054 Cladding OD, In
.5645 Cross-Sectional Flow Area in Core, Ft 29.90 V'
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l 1 XN-NF-79-24 Table 2.2 Thennal and Hydraulic System Data I
Parameter Value Reactor Flow, Ib/sec 7,083.
Flow Through Average Core, lb/sec 6,502.
Flow Through Hot Channel, lb/sec 14.2 Flow Through Core Bypass, lb/sec 566.7 Reactor Inlet Enthalpy, Btu /lb 481.9 Reactor Outlet Enthalpy, Btu /lb 577.5 Reactor Outlet Pressure, psia 1,015.0 Steamdrum Pressure, psia 990.0 Steamdrum Temperature, F 543.4 Primary Steamflow, lb/sec 444.4 Primary Feedwater Flow, Ib/sec 444.4 Primary Feedwater Temperature, F 429.2 Secondary Steamflow, lb/sec 396.7 Secondary Feedwater Flow, lb/sec 396.7 Secondary Feedwater Temperature, F 403.4 Steam Generator Shellside Pressure, psia 510.0 Steam Generator Shellside Temperature, F 469.
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XN-NF-79-24 3.0 HEATUP MODEL FOR THE DRESDEN-1 ECCS ANALYSIS 3.1 HEATUP MODEL DESCRIPTION The results of the RELAP4-EM/ BLOWDOWN calculation for power, heat transfer coefficient and fluid conditions versus time, and time of rated spray flow are used as input data to the ENC HUXY multirod heatup code.( ) The fluid condition input includes temperatures and quality.
The HUXY calculations cover the entire LOCA transient from break initiation until the Peak Cladding Temperature (PCT) rod quenches at the plane of interest. The example calculations are for ENC 6x5 fuel at BOL.
The calculations were made for an axial-times-radial combined peaking factor of 2.50, which corresponds to a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) of 11.58 kW/ft.
3.2 RESULTS OF THE HEATUP AfMLYSIS
_I Figure 3.1 presents the HUXY calculated temperature results.
Temperatures are shown for the hottest fuel rod (rod #12 as identified in Figure 3.1), the inert rod (rod #19), and the cannister.
For the bundle, a PCT of 1992 F was calculated to occur on rod #12 at 437 seconds after break initiation. The maximum local zirconium water reaction was 3.9% on rod #12.
The core wide zirconium water reaction was below 1%.
The above results apply for a MAPLHGR of 11.58 kW/ft in the Dresden-1 plant.
I I
L 3
- s t DRESDEN-1 FXAMPLE PROBLEM.
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.1 HERT UP RNRL, ENC 6XG RELORD
- 2 BOL 2.82 FT2 DEG 4 9 13 16 17 18 9 10 14 17 19 20 C
- 1. 15 19 20 21 l R0D 12 2FC0 INERT ROD 19 l
c CANISTER 2000 1992 i
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4.0 CONCLUSION
S The results of the LOCA ECCS analysis for ENC 6x6 fuel at BOL, with a MAPLHGR of 11.58 kW/ft, give a PCT of 1992 F on rod #12 at 437 seconds after break initiation, a local metal-water reaction of 3.9%
and a core wide metal-water reaction of less than one percent.
For the guillotine break of the reactor vessel recirculation inlet, this LOCA ECCS example problem analysis shows that the Dresden-1 plant with XN-2 reload fuel at BOL meets the Acceptance Criteria as legislated in 10CFR50.46 and Appendix K, that is:
1.
The calculated peak fuel element clad temperature does not exceed the 2200 F limit.
2.
The total anount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.
3.
The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The hot fuel rod local cladding oxidation limits of 17% are not exceeded during or af ter quenching.
I I
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I XN-NF-79-24 I
5.0 REFERENCES
I 1.
Exxon Nuclear Company, "The Exxon Nuclear Company WREM-Based NJP-BWR ECCS Evaluation Model and Application to the Oyster Creek Plant" Xti-75-55, Revision 2, and Supplement 1 and 2, April 1977.
2.
Safety Evaluation Report by the Office of Nuclear Regulatory Regula-tion Regarding Review of the Exxon Nuclear Company Non-Jet Pump Boiling Water Reactor ECCS Evaluation Model Described in Exxon I
Topical Reports, XN-75-55, Revision 2, dated August 1976, XN-75-55, Revision 2, Supplement 1, dated September 1976, XN-75-55, Revision 2 Supplement 2, dated December 1976, for Conformance to Appendix K to 10CFR50, USNRC, February 1977.
3.
Exxon Nuclear Company, "HUXY: A Generalized Multirod Heatup Code With 10CFR50 Appendix K Heatup Option User's Manual," XN-CC-33(A),
Revision 1, November 1975.
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