ML20062L469
| ML20062L469 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/16/1982 |
| From: | Leasburg R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Clark R, Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20062L470 | List: |
| References | |
| 490, NUDOCS 8208190120 | |
| Download: ML20062L469 (37) | |
Text
VIItOINIA l$LECTHIC AND Pown:It CO)11%NY H ICIIMOND, VIItOINI A 211261 It. II. LE Asen 8:N0 m2.Y"".'.","!.. Auguet 16, 1982 Mr. liarold R. Denton, Director Office of Nuclear Reactor Regulation Serial No.: 490 Attn: Mr. Robert A. Clark, Chief FR/NPW: bjc Operating Reactors Branch No. 3 Docket Nos.: 50-338 Division of Licensing 50-339 U. S. Nuclear Regulatory Commission Washington, D.C. 20555 License Nos.: NPF-4 NPF-7 Centlemen:
AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTil ANNA POWER STATION UNIT NOS.1 AND 2 PROPOSED TECICi1 CAL SPECIFICATION CHANCE Pursuant to 10CFR50.90, the Virginia Electric and Power Company requests an amendment, in the form of changes to the Technical Specifications, to Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station Unit Nos. 1 and 2.
A LOCA-ECCS reanalysis for North Anna Unit Nos. 1 and 2 has been performed using the NRC approved 1981 version of the Westinghouse LOCA-ECCS Evaluation Model. The analysis has been conducted in compliance with Appendix K to 10CFR50 and meets the acceptance criteria delineated in 10CFR50.46.
This reanalysis was performed by Vepco under supervision of Westinghouse, and the results will support continued full power operation for both North Anna Units 1 and 2 at steam generator tube plugging levels of up to 7 percent.
The results of this reanalysis also support a new Fq limit of 2.20. 'Ihe se results are provided in Attachment 1. Proposed changes to the Technical Specifications, consistent with the reanalysis, are provided in Attachment 2.
In conjunction with the higher Fq limit, we are providing, as Attachment 3, new Core Surveillance Reports for North Anna 2 Cycl 2 2 and North Anna 1 Cycle 4. khile these reports increase the Fxy limits to provide increased operational flexibility, the Fq flyspecks remain within the new Fq limit of 2.20 as shown in the figures.
b\ cW go k 8208190120 820816 PDR ADOCK 05000338 p PDH }
viuoix 4 EucTuic amo Powtw Comvr to Mr. Harold R. Denton 2 This request has been reviewed by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control staff. It has been determined that this request does not involve any unreviewed safety questions as defined in 10CFRSO.59.
We have evaluated this reques t in accordance with the criteria in 10CFR170.22. Since this request involves a safety issue which the staff should be able to determine does not involve a significant hazards consideration for Unit 1 and a duplicate safety issue for Unit 2, a Class III license amend-ment fee and a Class I license amendment fee are required for Unit 1 and Unit 2, respectively. A voucher check in the amount of $4,400.00 is enclosed in payment of the required fees.
Very trily yours, I ll R.[hl II. Leasburg Attachments (1) LOCA-ECCS Safety Evaluation for North Anna Unit Nos.1 and 2 (2) Proposed Technical Specification Changes (3) Core Surveillance Reports (4) Voucher Check for $4,400.00 cc: Mr. James P. O'Reilly _
Regional Administrator Region Il
Attcchment 1 LOCA-ECCS Safety Evaluation for North Anna Unit Nos. 1 and 2 8
1 i
PAGE 1
1.0 INTRODUCTION
A reanalysis of the ECCS performance for the postulated large break Loss of Coolant Accident (LOCA)* has been performed which is in compliance with ,
!j Appendix K to 10 CTR 50. The results of this reanalysis are presented herein :
i cnd are in compliance with 10 CTR 50.46, Acceptance Criteria for Emergency l Core Cooling Systems for Light Water Reactors. This analysis was performed with the NRC approved (Ref. 2 , 11) 1981 version of the Wesinghouse LOCA- .
ECCS evaluation model. The analytical techniques used are in full compliance with 10 CFR 50, Appendix K. l Ao required by Appendix K of 10 CFR 50, certain conservative assumptions were ;
onde for the LOCA-ECCS analysis. The assumptions pertain to the conditions ;
of the reactor and associated safety system equipment at the time that the i
LOCA is assumed to occur and include such items as the core peaking factors, tho containment pressure, and the performance of the emergency core cooling system (ECCS). All assumptions and initial operating conditions used in this recnalysis were the same as those used in the previous LOCA-ECCS analysis l (Raf. 3) with the following exceptions: 1) the 17 x 17 generic fuel j parameters were updated to reflect the current valuess 2) the 1981 model, l l
Which incorporates the impact of the fuel rod burst and blockage models :
i required by NUREG-0630, was used to perform this analysis.
l c The reanalysis of the small break LOCA is not necessary and therefore the analysis of this accident submitted by Reference 1 remains applicable.
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- PAGE 2
2.0 DESCRIPTION
OF pC57ULATED MAJOR REACTOR COOLANT pipe RUPTURE (LOSS OF COOLANT ACCIDENT - LOCA) !
A LOCA is the result of a rupture of the reactor coolant system (RCS) piping l or of any line connected to the system. The system boundaries considered in ths LOCA analysis are defined in the F5AR. Sensitivity studies (Reference 4) hevo indicated that a double-and cold leg guillotine (DECLG) pipe break is limiting. In the unlikely event of a DECLG break, a rapid depressurization '
i of the RCS will result. The reactor trip signal subsequently occurs when the !
L pressurizar low pressure trip setpoint is reached. A safety injection system !
(SIS) signal is actuated when the appropriate setpoint is reached and the i high head safety injection pumps are activated. The actuation and subsequent f I
activation of the ECCS, which occurs with the SIS signal, assumes the most '
i liniting single failure event. These countermeasures will limit the f i
consequences of the accident in two ways: ,
, 1. Reactor trip and borated water injection complement void .
l formation in causing rapid reduction of power to a [
residual level corresponding to fission product decay
[
i heat. (It should be noted, however, that no credit is ;
taken in the analysis for the insertion of control rods to shut down the reactor).
- 2. Injection of borated water provides heat transfer from the l core and prevents excessive clad temperatures. i i
Bofore the break occurs, the unit is in an equilibrium condition, i.e., the l I
hont generated in the core is being removed via the secondary system. During r blowdown, heat from decay, hot internals and the vessel continues to be trcnsferred to the reactor coolant system. At the beginning of the blowdown f
- phase, the entire RCS contains subcooled liquid which transfers heat from the coro by forced convection with some fully developed nucleate boiling. After th9 break develops, the time to departure from nucleate boiling is calcuated, b
PAGE 3 consistent with Appendix K of 10 CFR 50. Thereafter, the core heat transfer in based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms. During the refill period, it is cocumed that rod-to-rod radiation is the only core heat transfer mechanism.
Tho heat transfer between the reactor coolant system and the secondary side coy be in either direction depending on the relative temperatures. For the coce of continued heat addition to the secondary side, secondary side pressure increases and the main safety valves may actuate to reduce the pressure. Makeup to the secondary side is automatically provided by the cuxiliary feedwater system. Coincident with the safety injection signal, normal feeduater flow is stopped by closing the main feedwater control valves cnd tripping the main feedwater pumps. Emergency feedwater flow is initiated ,
by starting the auxiliary feedwater pumps. The secondary side flow aids in tha reduction of reactor coolant system pressure. When the reactor coolant cyntem depressurizes to 600 psia, the accumulators begin to inject borated unter into the reactor coolant loops. A conservative assumption is then made that the injected accumulator water bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50. In addition, the reactor coolant l
pumps are assumed to be tripped at the initiation of the accident and effects t of pump coastdown are included in the blowdown analysis.
Tho water injected by the accumulators cools the core and subsequent oporation of the low head safety injection pumps supplies water for long term ;
o cooling. When the RWST is nearly empty, long term cooling of the core is I
accomplished by switching to the recirculation mode of core cooling, in which the spilled borated water is drawn from the containment sump by the low head i
l pAGE 4 ociety injection pumps and returned to the reactor vessel.
l Tho containment spray system and the recirculation spray system operates to roturn the containment environment to a subatmospheric pressure.
l Tho large break LOCA transient is divided, for analytical purposes, into three phases blowdown, refill, and reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment, and the fuel clad temperature transient of the hottest fuel rod in the core.
Boced on these considerations, a system of inter-related computer codes has l
baan developed for the analysis of the LOCA.
Tho description of the various aspects of the LOCA analysis methodology is given in WCAp-8339(Ref. 5). This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with 10 CFR 50, Appendix K. The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, are l
doacribed in detail in WCAp-8306 (Ref, 6), WCAp-8326(Ref. 7), WCAp-8171(Ref.
l 8), and WCAp-8305(Ref. 9), respectively. These codes are able to assess -
whether sufficient heat transfer geometry and core amenability to cooling are proserved during the time spans applicable to the blowdown, refill, and roflood phases of the LOCA. The SATAN-VI computer code analyzes the tharmal-hydraulic transient in the RCS during blowdown and the COCO computer code is used to calculate the containment pressure transient during all three l pheces of the LOCA analysis. Similarly, the LOCTA-IV computer code is used i
! to compute the thermal transient of the hottest fuel rod during the three phoces.
l:
PAGE 5 SATAN-VI is used to determine the RCS pressure, enthalpy, and density, as wall as the mass and energy flow rates in the RCS and steam generator accondary, as a function of time during the blowdown phase of the LOCA.
SATAN-VI also calculates the accumulator mass and pressure and the pipe break oces and energy flow rates that are assumed to be vented to the containment during blowdown. At the end of the blowdown, the mass and energy release rates during blowdown are transferred to the COC0 code for use in the dotermination of the containment pressure response during this first phase of ths LOCA. Additional SATAN-VI output data from the end of blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.
With input from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic codel to determine the core flooding rate (i.e., the rate at which coolant onters the bottom of the core), the coolant pressure and temperature, and the qu2nch front height during the refill and reflood phases of the LOCA. ,
6 WREFLOCD also calculates the mass and energy flow rates that are assumed to ha vented to the containment. Since the mass flow rates to the containment ,
depends upon the core pressure, which is a function of the containment backpressure, the WREFLOOD and COC0 codes are interactively linked. WREFLOOD 10 also linked to the LOCTA-IV code in that thermal-hydraulic parameters from WREFLOOD are used by LOCTA-IV in its calculation of the fuel temperature.
l LOCTA-IV is used throughout the analysis of the LOCA transient to calculate l
the fuel and clad temperature of the hottest rod in the core. The input to LOCTA-IV consists of appropriate thermal-hydraulic output from SATAN-VI and I
WREFLOOD and conservatively selected initial RCS operating conditions. These l
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PAGE 6 ;
initial conditions are summarized in Table 1 and Figure 1. (The axial power l chcpe of Tigure 1 assumed for LOCTA-IV is a cosine curve which has been i proviously verified (Ref. 10) to be the shape that produces the maximum peak c1cd temperature).
Tho COCO code, which is also used in the LOCA analysis, calculates the containment pressure. Input to COCO is obtained from the mass and energy flow rates assumed to be vented to the containment as calculated by the SATAM-VI and WREFLOOD codes. In addition, conservatively chosen initial containment conditions and an assumed mode of operation for the containment cooling system are input to COCO. These initial containment conditions and casumed modes of operation are provided in Table 2.
1 1
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PAGE 7 3.0 DISCUSSION OF SIGNIFICANT INPUT Significant differences in input between this analysis and the currently applicable analysis are delineated in Section 1.0 and discussed in more dotail below. The changes made in the analysis are those that are required to reperform the Reference 3 analysis with the 1981 model for the current 17 x 17 fuel. The steam generator tube plugging level was maintained at 7 percent.
Ths notable change for this analysis is the use of the 1981 version of the Wautinghouse Appendix K Evaluation Model. The changes in the 1981 model which had the most impact on this analysis were: 1) the impact of the fuel I
rod burst and blockage models required by NUREG-0630; 2) the use of "UHI 1
Software Technology" models.
Tha calculation was performed assuming conservative generic 17 x 17 fuel l l
parameters consistent with the current methodology.
Whon the above changes were incorporated into the analysis, it was found that ths assumed heat flux hot channel factor could be kept at the 2.20 value used i
in the Reference 3 analysis and still ensure compliance with the 10 CFR 50.46 acceptance criteria.
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r 4.0 RESULTS j r
Tchles 1 and 2 and Figure 1 present the initial conditions and modes of oporation that were assumed in the. analysis. Table 3 presents the time coquence of events and Table 4 presents the results for the double-ended cold log guillotine break (DECLG) for the CD=0.4 discharge coefficient. All provious LOCA-ECCS submittals for the North Anna units have resulted in the ,
CD=0.4 discharge coefficient being the limiting break size. The applicability of this conclusion (i.e. CD=0.4 is the limiting break size) for ,
this analysis was explicitly verified. Consequently, only the results of the j nost limiting break size are presented in the figures and remaining tables in
- i
- this submittal. The current analysis resulted in a limiting peak clad
^
tsaperature of 2194.7'F, a maximum local cladding oxidation level of 7.88X, and a total core metal-water reaction of less than 0.3X. The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures i 2 through 18.
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PAGE 9
5.0 CONCLUSION
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l For breaks up to and including the double-ended severance of a ;
reactor coolant pipe and for the operating conditions specified in Table 1 and 2, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46. That is: I i
- 1. The calculated peak fuel rod clad temperature is below the requirement of 2200*F. !
l t
- 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
i
- 3. The clad temperature transient is terminated at a time when I the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17X are not exceeded during or after quenching. i.
- 4. The core remains amenable to cooling during and after the ;
break. .
- 5. The core temperature is reduced and the long-term decay heat is removed for an extended period of time.
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PAGE 10
6.0 REFERENCES
- 1. Final Safety Analysis Report, North Anna Power Station, Units 1 and 2, Virginia Electric and Power Company.
- 2. Letter from J. R. Miller (NRC) to E. P. Rhae(Westinghouse), dated December 1, 1981.
- 3. Letter from R. H. Leasburg(Vepco) to H. R. Denton(NRC), Serial No. 627,
- November 12, 1981.
- 4. Buterbaugh. T. L., Johnson, W. J., and Hopelic, S. D., " Westinghouse ECCS Plant Sensitivity Studies," WCAP-8356, July 1974.
- 5. Bordelon, F. M., et. al., " Westinghouse ECCS Evaluation Model- Summary," ,
WCAP-8339, July, 1974.
- 6. Bordelon, F. M., et.al., " SATAN-VI Programs Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306, June 1974.
- 7. Bordelon, T. M., and Murphy, E. T., " Containment Pressure Analysis Code (COCO)," WCAP-8326, June 1974.
- 8. Kelly, R. D., et. al., " Calculation Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974.
- 9. Bordelon, T. M., et. al., "LOCTA-IV Program Loss-of-Coolant Transient Analysis," WCAP-8305, June 1974.
- 10. Letter from C. M. Stallings(Vepco) to E. G. Case (NRC), Serial No. 092, February 17, 1978. '
- 11. Eicheldinger C., " Westinghouse ECCS Evaluation Model - 1981 Version",
WCAP-9220-P-A Revision I (Proprietary Version), WCAP-9221-A Revision I (Non-Proprietary Version), February, 1982.
PAGE 11 TABLE 1 INITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE-ENDED COLD LEG GUILLOTINE BREAK (DECLG)
CALCULATIONAL INPUT Core Power (MWt, 102X of) 2775 '
I Pack Linear Power (kW/ft, 102X of) 11.98 Hant Flux Hot channel Factor ( Fe ) 2.20 N
Enthalpy Rise Hot Channel Factor (F6H ) 1.55 Accumulator Water Volume (ft3 , each) 1025 R3 actor Vessel Upper Head Temperature Equal to Thot LIMITING FUEL REGION AND CYCLE CYCLE REGION Unit 1 ALL ALL Regions ,
Unit 2 ALL ALL Regions 1
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PAGE 12 TABLE 2 CONTAINMENT DATA NET FREE VOLUME 1.916 x 105 ft3 INITIAL CONDITIONS 1 pressure 9.5 psia Temperature 90'F RWST Temperature 35'T Outside Temperature -10*F SPRAY SYSTEMS Number of Pumps Operating 2 Runout Flow Rate (per pump) 2000 gym Time in which spray is effective 59 secs STRUCTURAL HEAT SINKS 1 Thickness (In) Area (Ftt), w/ uncertainty 6 Concrete 8,3?3 12 concrete 62,271 18 Concrete 55,365 24 Concrete 11,591 27 Concrete 9,404 36 Concrete 3,636
.375 Steel, 54 Concrete 22,039
.375 Steel, 54 Concrete 28,933
.500 Steel, 30 concrete 25,673 26.4 Concrete, .25 Steel, 120 Concrete 12,110
.407 Stainless Steel 10,527
.371 Steel 160,328
.882 Steel 9,894
.059 Steel 60,875 1 5ee the response to Comment 56.106 of the FSAR for a detailed breakdown of the containment heat sinks and for justification of the other input parameters used to calculate containment pressure.
PAGE 13 TABLE 3
_ TIME SEQUENCE OF EVENTS DECLG l CD=0.4 (Sec)
Stort 0.0 Rocctor Trip 0.72
- 5. I. Signal 2.19 Acc. Injection 15.80 :
r Pump Injection 27.19 End of Bypass 30.00 End of Blowdown 30.02 {
Bottom of Core Recovery 42.93 !
Acc. Empty 54.47 I i
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PAGE 14 TABLE 4 r
RESULTS FOR DECLG l
CD=0.4 Pock Clad Tamp, 'T 2194.7 i
Pock Clad Location. Ft. 7.25 i Local Zr/H20 RXN (max), X 7.88 Local Zr/H2O Location, Ft. 7.5 -
Total Zr/H2O RXN, X <0.3 i
Hot Rod Burst Time, sec. 35.70 Hot Rod Burst Location, Ft. 6.0 L
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PAGE 15 TABLE 5 '
REFLOOD MASS AND ENERGY RELEASES i DECLG (CD= 0.4)
TIME (SEC) TOTAL MASS TOTAL ENERGY FLOWRATE (LB/SEC) FLOWRATE (105 BTU /SEC) ,
42.925 0.0 0.0 43.500 0.761 0.0099 44.050 1.533 0.1990 49.360 35.920 0.4665 58.922 212.90 1.4013 74.022 253.54 1.4377 93.222 267.14 1.4095 115.122 275.48 1.3642 139.422 282.28 1.3152
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195.922 295.24 1.2116 1 ;
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i 271.022 313.46 1.1063
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L PAGE 16 TABLE 6 9 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT DECLG, CD=0.4 TIME (SEC) MASS FLOWRATE8 (LBM/SEC) 0.00 4010.1 1.01 3622.8 3.01 3106.1 (
5.01 2763.8 7.01 2512.0 10.01 2228.8 !
15.01 1902.0 20.01 1678.5 25.01 1522.5 :
29.01 1583.2 **
- For energy flourate multiply mass flourate by a constant of 59.60 BTU /LBM. ;
i naFor energy flourate at this time multiply mass flourate by 54.08 BTU /LBM. 1 t
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