ML19260A762
| ML19260A762 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/29/1979 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19260A763 | List: |
| References | |
| NUDOCS 7912030209 | |
| Download: ML19260A762 (39) | |
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- s VIRGINI A ELECTRIC AND POWER COMP ANY, RICHMOND, VIRGINI A 23261 November 29,1979 Mr. Harold R. Denton, Director Serial No. 986 Office of Nuclear Reactor Regulation FR/MLB: mye Attn: Mr. A. Schwencer, Chief Operating Reactors Branch No.1 Docket No.: 50-338 Division of Operating Reactors 50-339 U. S. Nuclear Regulatory Commission Washington, DC 20555 License No.: NPF-4
Dear Mr. Denton:
AMENDMENT TO OPERATING LICENSE NPF-4 NORTH ANNA POWER STATIONS UNITS NO.1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE NO. 27 We have reviewed Mr. D. G. Eisenhut's letter of November 9,1979 concerning the applicability of fuel cladding swelling and incidence of rupture modeling in the ECCS evaluation model usea in the analysis of North Anna Units 1 and 2. In addition, we have initiated discussions with representatives of the Westinghouse Electric Corporation concerning their responses to the NRC Staff on this issue contained in the letters from T. M. Anderson to D. G. Eisenhut, Serial Nos. NS-TMA-2147 and 2163, dated November 2,1979 and November 16, 1979, respectively. As a result of this review, we have concluded that the burst temperature curve used in our currently applicable analysis (reference the O. D. Parr (NRC) to W. L. Proffitt (Vepco) letter dated May 19,1978) was based on a nonconservative heat-up rate. In order to explicitly represent the use of a conservative heat-up rate, we have now completed a new LOCA-ECCS analysis as discussed in my letter of November 21, 1979 to Mr. J. P. O'Reilly, Serial No. 953. This analysis uses the appropriate heat-up rate calculation procedure indicated in Mr. Anderson's letter, NS-TMA-2163, in conjunction with the NRC approved February 1978 Version of the Westinghouse LOCA-ECCS Evaluation Model. The implications of using revised heat-up rates on the clad strain and flow blockage concerns expressed in Mr. Eisenhut's letter of November 9, 1979 were addressed in the above Westinghouse letter of November 16,1979 (Serial No. NS-TMA-2163), and it is our understanding that this issue is currently being reviewed by the NRC. As a result, this anlaysis is in compliance with Appendix K to 10 CFR 50 and meets the criteria of 10 CFR 50.46. To implement the results of the new analysis, it is necessary, pursuant to 10 CFR 50.90, to request an amendment in the form of a change to the Technical Specifications, to Operating License No. NPF-4 for the North Anna Power Station, Unit No. 1. The proposed change (designated Change No. 27) and supporting analysis are attached, and the results require a reduction in the limiting heat flux hot channel factor (Fq) from 2.21 to 2.10. This change is also applicable to the Technical Specifications for North Anna Unit No. 2. since the current Technical Specifications limit on Fnis nonconservative, we are immediately implementing an administrative reduction of tffenF limit from 2.21 to 2.10 in order to support continued operation. This limit will be adjusted, if necessary, consistent with your final review of the Westinghouse ECCS Evaluation Model and our attached safety analysis. However, it should be noted that the 2.10 F q limit being 1460 .503 79120302 o9
VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton SHEP1T NO. 2 administratively imposed which is based on the attached analysis, assumes a more conservative input condition than will actually be experienced during this time frame. Specifically, the attached analysis assumes that approximately 500 tubes in the steam generators are plugged. This is conservative, since we currently only plan to plug a few (approximately four) tubes for North Anna Unit 1 during the Cycle 2 refueling as part of our routine steam generator inspection and maintenance program. The supporting safety analysis for this request, as well as the specific changes to the Technical Specifications, are provided in Attachmerits 1 and 2, respectively. This information has been reviewed and approved by both the Station Nuclear Safety and Operating Committee and the System Nuclear Safety and Operating Committee. It has been determined that these changes to the Technical Specifications do not involve an unreviewed safety question as defined in 10 CFR 50.59. We have evaluated the prcposed Unit 1 Technical Specifications changes in accordance with the criteria in 10 CFR 170.22. It has been determined that this request requires a Class III amendment fee. Accordingly, you will find enclosed a voucher check in the amount of $4,000.00 in payment of the amendment fee. Should you have questions, please contact us at your earliest convenience. Very truly yours, C. M. Stallings Vice President-Power Supply and Production Operations Attachments (1) LOCA-ECS Safety Evaluation for North Anna Unit 1 (2) Proposed Technical Specifications Change No. 27 (3) Voucher Check for $4,000.000 cc: Mr. James P. O'Reilly, Director Office of Inspection and Enforcement, Region II Mr. O. D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management Mr. Jack Rosenthall, Reactor Safety Branch Division of Operating Reactors 1460 304
COMMONWEALTH OF V IRGINI A )
) s. S.
CITY OF RICHMOND ) Before me, a Natary Pubilc, in and for the City and Common-wealth aforesaid, today personally appeared C. M. Stallings, who being duly sworn, made oath and said (1) that he is Vice President-Power Supply and Production Operations, of the Virginia Electric and Power Company, (2) that he is du!y authorized to execute and file the fore-going Amendment in behalf of that Company, and (3) that the statements in the Amendment are true to the best of his knowledge and belief. Given under my hand and notarial seal this 29/Aday of Ah> gem A. , - ,mo . My Commission expires .)er > 1;/g v =C 1941 . o
}& >l}W l$s Notary Public (SEAL) 1460 305
e 8 ATTACHMENT 1 1460 306
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 Appendix K to 10 CFR 50. The results of this reanalysis are pre-sented herein and are in comrliance with 10 CFR 50.46, Acceptance Cri-teria for Emergency Core Cooling Systems for Light Water Reactors. This analysis was performed with the NRC approved (Ref. 2) February 1978 version of the Westinghouse LOCA-ECCS evaluation model in conjunction with an appropriate fuel rod burst curve calculational procedure (Ref. 3). The analytical techniques used are in full compliance with 10 CFR 50, Appendix K. As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the LOCA-ECCS analysis. The assumptions pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur and include such items as the core peaking factors, the containment pressure, and the perfo.imance of the emergency core cooling system (ECCS). All assumptions and initial operating conditions used in this reanalysis were the same as those used in the previous LOCA-ECCS analysis (Ref. 4) with the following exceptions: 1) The limiting value of the heat flux hot channel f actor was decreased from
- The reanalysis of the small break LOCA is not necessary and therefore the analysis of this accident submitted by Reference 1 remains appli-cable.
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PACE 2 2.21 to 2.10; 2) !!ominal design power rating was used versus engineered safeguards rating; 3) RCS cold leg temperature of 5500F based on operational data was used versus 555 F; 4) More accurate data for several containment parameters were used;
- 5) 5% of the steam generator tubes were assumed to be plugged; 6) 17x17 generic fuel parameters were used instead of plant specific fuel parameters.
2.0 DESCRIPTION
OF POSTULATED MAJOR REACTOR COOLANT pipe RUPTURE (LOSS OF COOLAMT ACCIDENT - LOCA) A LOCA is the result of a rupture of the reactor coolant system (RCS) piping or of any line connected to the system. The system boundaries considered in the LOCA analysis are defined in the TSAR. Sensitivity studies (Reference 5) have indicated that a double-ended cold leg guillotine (DECLG) pipe break is limiting. Should a DECLG break occur, rapid depressuriration of the RCS occurs. The reactor trip signal subsequently occurs when the pressurirer low pressure trip setpoint is reached. A safety injection system (SIS) signal is actuated uhen the appropriate setpoint is reached and the high head safety injection pumps are activated. The actuation and subsequent activation of the ECCS, which occurs with the SIS signal, as s u:a e s the most limiting single failure event. These countermeasures will limit the consequences of the accident in tue uays.
- 1. Reactor trip and borated unter injection complpment void formatisn in causing rapid reduction of power to a residual level corresponding to dission product decay hett.
(It shculd be noted, however, that no credit in taken in the analysis for the insertion of control rods to shut doun the reactor)
- 2. In ]e c t:.o n of borated unter provides heat tr an:, rer from t h .-
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PAGE 3 core and prevents excessive clad temperatures. Before the break occurs, the unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. During bloudown, heat from decay, hot internals and the vessel continues to be transferred to the reactor coolant system. At the beginning of the bloudoun phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10 CFR 50. Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms. During the refill period, it is assumed that rod-to-rod radiation is the only core heat transfer mechanism. The Feat transler between the reactor coolant system and the secondary syste m may be in either direction depending on the Islative temperatures. For the case of continued heat addition to the secondary side, secondar/ side pressure increases and the main safety valves may actuate to reduce the pressure. Makeup to the secondary side is automatically provided ly the auxiliary feeduater system. Coincident with the safety injection signal, ncrmal feeduater flou is stopped by tlosing the main feedwater coittrol valves and tripping the main feedwater pumps. Emergency feeduater flou is initiated by starting the auxiliary feeduater pumps. The secondtry side flow aids in the reduction of reactor coolant system pressure. When the reactor coolant system depressuri=es to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. The conservative assumption is then made that injected accumulator unter bypasses the core and goes out through the 1460 309
PAGE 4 break until the termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50. In addition, the reactor coolant Pumps are assumed to be tripped at the initiation of the accident and effects of pump coastdown are included in the blowdown analysis. The water injected by the accumulators cools the core and subsequent operation of the lea head safety insection pumps supplies water for long term cooling. When the RWST is nearly empty, long term cooling of the core is accomplished by switching to the recirculation mode of core cooling, in which the spilled borated unter is drawn from the containment sump by the low head safety injection pumps and returned to the reactor vessel. The containment spray system and the recirculation spray system operate to return the containment environment to a subatmospheric pressure. The 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
- 7. c d in the core. Based on these considerations, a system of inter-related computer codes has been developed for the analysis of the LOCA.
The description of the various aspects of the LOCA analysis methodology is given in WCAp-8339(Ref. G). This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the 1460 310
PAGE 5 codes which ensure compliance with 10 CTR 50, Appendix K. The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in WCAp-8306(Ref. 7),WCAp-8326,(Ref. 8),WCAp-8171(Ref. 9), and WCAp-8305(Ref. 10), respectively. These codes are able to assess whether sufficient heat transfer geometry and core amenablity to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA. The SATAM-VI computer code analy=es the thermal-hydraulic transient in the RCS during blowdown and the reflood computer code is used to calculate the containment pressure transient during all three phases of the LOCA analysis. Similarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases. SATAN-VI is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the RCS and steam generator secondary, 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 mass 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 COCO code for use in the determination of the containment pressure response during this first phase of the 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.
- 1460 311
PAGE 6 With input from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the refill and reflood phases of the LOCA. WREFLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment. Since the mass flow rate to the containment depends upon the core pressure, which is a function of the containment backpressure, the WREFLOOD and COCO codes are interactively linked. WREFLOOD is also linked to the LOCTA-IV code in that thermal-hydraulic parameters fr'm WREFLOOD are used by LOCTA-IV in its calculation of the fuel temperature. LOCTA-IV is used throughout the analysis of the LOCA transient to calculate 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 WREFLOOD and conservatively selected initial RCS operating conditions. These initial conditions are summarized in Table 1 and Figure 1. (The axial power shape of Figure 1 assumed for LOCTA-IV is a cosine curve which has been previously verified (Ref. 11) to be the shape that produces the maximum peak clad temperature). The COCO code, which is also used throughout 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 SATAN-VI and WREFLOOD codes. In addition, conservatively chosen initial containment conditions and an assumed. mode of operation for the 1460 312
PAGE 7 containment ~ cooling system are input to CCCO. These initial containment con'ditions and assumed modes of operation are provided in Table 2. 3.0 DISC 11SSION 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 detail below. The changes made in the anaysis reflect the operational conditions and limits necessary to allow full power operation at steam generator tube plugging levels of up to 5%. The most notable input change for this analysis is the increase in assumed steam generator tube plugging. The currently applicable analysis made no allowance for tube plugging. The plugging level assumed for this analysis is 5%. The assumption of a small amount of steam generator tube plugging in the analysis also affects the assumed core inlet temperature by requiring a decrease in this parameter. Consequently, a core inlet temperature of 550 F was assumed. This value is the best estimate core inlet temperature as determined from operational data and is adequate to encompass the applicable steam generator tube plugging range. In order to ensure compliance with the 10 CFR 50.46 acceptance criteria, a change to the assumed value of the heat flux hot channel factor was made. S p e cific alli' , the assumed heat flux hot channel factor was decreased from 2.21 to 2.10. 1460 313
PAGE 8 previous analyses had assumed ar Engineered Safeguards power level in the blowdown portion of the analysis for future operational flexibility. The current analysis assumed a nominal design power level as the basic power input throughout the analysis. Several changes were made to the containment parameters. The amounts of the various structural heat sinks provided in Table 2 were reviewed in detail. Based on the as-built plant containment, the heat sinks were conservatively revised to include the additiion of a 3% uncertainty to all surface areas. As allowed by the NRC, credit has been taken for paint on carbon steel surfaces. The calculation was performed assuming conservative generic 17x17 fuel parameters. The previous analysis had assumed cycle specific 17x17 parameters. Finally, the analysis was conducted with the February, 1978 version of the Westinghouse LOCA-ECCS Evaluation Model (Ref 12,13,14). In particular, recent concerns , regarding the fuel performance calculaton part of the analysis, have been conservatively accounted for by the use of new fuel rod burst calculational methodology. 4.0 RESULTS Tables 1 and 2 and Figure 1 present the initial conditions and modes of operation that were assumed in the analysis. Table 3 presents the time sequence of events and Table 4 presents the results for the double-ended - cold leg guillotine break (DECLG) for the CD=0.4 discharge coefficient. The 1460 314
PAGE 9 DECLG has been determined to be the limiting break si=e and location based on the sensitivity studies reported in Reference 5. Further, all previous LOCA-ECCS submittals for the North Anna units have resulted in the CD=0.4 discharge coefficient being the limiting break si=e. The applicability of this conclusion (i.e. CD=0.4 is tha limiting break si=e) for this analysis was explicitly verified. (See Table 4). Consequently, only the results of the most limiting break si=e are presented in the figures and remaining tables in this submittal. The current analysis resulted in a limiting peak clad temperature of 2195 T, a maximum local cladding oxidation level of 7.7%, and a total core metal-water reaction of less than 0.3%. The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2 through 18.
5.0 CONCLUSION
S For breaks up to and including the double-ended severance of a reactor pipe and for the operating conditions specified in Tables 1 and 2, coolant the . Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46. That is:
- 1. The calculated peak fuel rod clad temperature is below the requirement of 2200 F.
- 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.
- 3. The clad temperature transient is terminated at a time when the core geometry is still amenable te cooling. The locali=ed cladding oxidation limits of 17% are not exceeded during or after quenching.
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PAGE 10
- 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.
6.0 REFERENCES
- 1. Final Safety Analysis Report, North Anna Power Station, Units
, 1 and 2, Virginia Electric and Power Company.
- 2. Letter irom J.F. Stol=(NRC) to T.M. Anderson (Westinghcuse), dated August 29, 1978.
- 3. Letter from T.M. Anderson (Westinghouse) to D.G. Eisenhut(NRC),
Serial No. MS-TMA-2163, November 16,1979.
- 4. Letter from C.M. Stallings(Vepco) to E.G. Case (NRC), Serial No.
258, May 5, 1978.
- 5. Buterbaugh, T.L., Johnson, W.J., and Kopelic, S.D., " Westinghouse ECCS Plant Sensitivity Studies," WCAP-8356, July 1974.
- 6. Bordelon, F.M., et. al., " Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, July, 1974.
- 7. Bordelon, F.M., et. al., " SATAN-VI Program Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306, June 1974.
- 8. Bordelon, F.M., and Murphy, E.T., " Containment Pressure Analysis Code (COCO)," WCAP-8326, June 1974.
- 9. Kelly, R. et. al., " Calculational Model for Core Reflooding after a Lo %-af-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974.
- 10. Bordelon, F.M., et. al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, June 1974.
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PAGE 11
- 11. Letter from C.M. Stallings(Vepco) to E.G. Case (HRC), Serial No. 092, February 17, 1978.
- 12. " Westinghouse ECCS Evaluation Model- February 1978 Version,"
WCAP-9220.
- 13. Letter from T.M. Anderson (Westinghouse) to J.F. Stol=(NRC),
Serial No. MS-TMA-1981, November 1, 1978.
- 14. Letter from T.M. Anderson (Westinghouse) to R. Tedesco(HRC), Serial No. MS-TMA-2014, December 11, 1978.
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TABLE 1 INITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE-ENDED COLD LEG GUILLOTINE BREAK (DECLG) Calculational Input Core Power (MWt , 102% of) 2775 i Peak Linear Power (kw/f t, 102% of) 11.43 Heat Flux Hot Channel Factor (Fq) 2.10 Enthalpy Rise Hot Channel Factor (FN} H 1.55 Accumulator Water Volume (ft , each) 1025 Reactor Vessel Upper Head Temperature Equal hot Limiting Fuel Region and Cycle Cycle Region Unit 1 ALL ALL Regions Unit 2 ALL ALL degions 1460 318
TABLE 2 CONTAINMENT DATA NET FREE VOLME 1.916 x 10 ft INITIAL CONDITIONS Pressure 9.6 psia Temperature 90 F RWST Temperature 35 Outside Temperature -10,FF SPRAY SYSTEM Number of Pumps Operating 2 Runout Flow Rate (per pump) 2000 gpm 63 Time in which spray is effective 59 secs STRUCTURAL HEAT SINKS Thickness (In) Area (Ft ), w/ uncertainty 6 Concrete 8,393 12 Concrete 62,271 18 Concrete 55,365 24 Concrete 11,591 37 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
.458 Steel 160,328
.882 Steel 9,894
.059 Steel 60,875 (1) See the response to Comment S6.106 of the FSAP for a detailed breakdown of the containment heat sinks and for justification of the other input parameters used to calcul.1te containment pressure.
1460 al9
TABLE 3 TIME SEQUENCE OF EVENTS DECLG CD=0.4 (Sec) START 0.0 Reactor Trip , 0.72 S. I. Signal 2.42 Acc. Injection 16.3 End of Bypass 26.33 Pump Injection 27.42 End of Blowdown 29.52 Bottom of Core Recovery 39.74 Ac c. Emp ty 52.3 1460 320
TABLE 4 RESULTS FOR DECLG CD=0.4 CD =0.6 Peak Clad Temp, F 2195 2125 Peak Clad Location Ft. 7.5 7.5 Local Zr/H O RXN (max), % 7.7 6.34 2 Local Zr/H 0 Location, Ft. 7.5 7.5 Total Zr/H 0 RXN, ', <0.3 <0.3 2 Hot Rod Burst Time, sec 27.5 25.8 Hot Rod Burst Location, Ft. 6.0 6.0 1460 32;
TABLE 5 REFLOOD MASS AND ENERGY RELEASES - DECLC (C ~
- D TOTAL MASS TOTAL ENERGY 5
TIME (SEC) FLOWRATE (LB/SEC) FLOWRATE (10 BTU /SEC) 39.7 0.0 0.0 40.8 0.748 0.0098 46.0 35.45 0.4623 55.0 210.96 1.395 68.4 245.57 1.412 85.0 258.49 1.380 103.7 266.58 1.338 124.2 273.18 1.290 170.4 284.97 1.194 225.6 297.34 1.091 299.5 317.77 1.003 457.3 347.37 0.8037 0 .'22
TABLE 6 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT DECLG, C "
- D TIME (SEC) MASS FLOWRATE* (LBM/SEC) 0.0 4010 1.0 3622 3.0 3104 5.0 2761 7.0 2506 10.0 2225 15.0 1894 20.0 1673 25.0 1523 30.0 1420
*For energy flowrate multiply mass flowrate by a constant of 59.60 BTU /LBM.
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