ML19263E213
| ML19263E213 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/31/1979 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| 388, NUDOCS 7906050318 | |
| Download: ML19263E213 (46) | |
Text
1 t-VIHUINIA Ex.ucrHIc ann Pownu COMPANY Ricun own.Va nox x A 20201 May 31, 1979 Mr. Harold R. Denton, Director Serial No. 388 Office of Nuclear Reactor Regulation LQA/DWSjr: jab Attn: Mr. Albert Schwencer, Chief Operating Reactors Branch flo.1 Docket No.
50-281 Division of Reactor Licensing U. S. Nuclear Regulatory Commission License No. DPR-37 Washington, DC 20555
Dear Mr. Denton:
AMENDMENT TO OPERATING LICENSE SURRY POWER STATION UNIT NO. 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE N0. 78 Pursuant to 10 rFR 50.90, the Virginia Electric and Power Com-pany hereby requests an amendment, in the form of changes to the Technical Specifications, to Operating License No. DPR-37 for the Surry Nuclear Power Station, Unit No. 2.
The proposed changes are attached and have been designated as Change No. 78 The proposed amendment is in response to the requirement for a LOCA-ECCS analysis which was identified in the Safety Evaluation Report transmitted by the A. Schwencer (NRC) to W. L. Proffitt (Vepco) letter dated December 15, 1978 provides the required LOCA-ECCS analysis results which will support the continued full rated power operation of both Surry Unit Nos.1 and 2 after replacement of their respective steam generators. Attachment 2 provides the associated changes to the Technical Specifications for Surry Unit 2.
Prior to completion of the Surry Unit 1 Steam Generator Replacement Program, the additional Technical Specifications changes required to support operation of Surry Unit 1 will be provided.
The LOCA-ECCS analysis was performed with a calculational model which fully conforms to the provisions of Appendix K,10 CFR 50 and which specifically represented the replacement steam generators.
In addition, the analysis accounted for plant modifications required during this outage and sat.isfies our commitment contained in the letter from Mr. C. M. Stallings (Vepco) to Mr. E. G. Case (NRC), Serial No.
382A/092477, dated November 22, 1977 The analysis results demonstrate compliance to the requirements of 10 CFR 50.46, 2299 270 790605 0 3/cf
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VaucixA ELzcTunc Awo Powra COMPAVY To Mr. Harold R. Denton 2
Since the proposed changes complement our response initiated in November,1977, to the Net Positive Suction Head concerns, an amendment fee payment is not enclosed.
This proposed amendment 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 this request does not involve an unreviewed safety question as defined in 10 CFR 50.59.
Your review of the attached Technical Specifications change is requested by August 1,1979 Should you have questions, we would be happy to meet with you at your earliest convenience.
Very truly yours, b.Nl.%:dN0nif C. M. Stallings Vice President-Power Station and Production Operation Attachments:
(1) Safety Analysis (2) Proposed Technical Specifications cc: Mr. James P. O'Reilly, Director Office Inspections and Enforcement Region II 2299 271
e t
COMMONWEALTH OF V IRGINI A
)
)
S. S.
CITY OF RICHMOND
)
Before me, a Notary Public, in and ror the City and Common-weal th aforesaid, today personally ap;. eared 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 duly authorized to execute and file the fore-going Amendment in behalf of that Company, and (3) that tne statements in the Amendment are true to the best of his knowledge and belief.
Given under my hand and notarial seal this 3/ <1 day of uf., v
,em.
My Commission expires
.L,,,c v yr /9e#
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Nota'ry Public 2299 272 (SEAL)
i I
ATTACHMENT 1 2299 273 k
i t
t l.0 INTRODUCTION A reanalysis of the ECCS cooling 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 presented herein* and are in compliance with 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors. This reanalysis was performed with the NRC approved (1) February, 1978 version of the Westinghouse LOCA-ECCS evaluation model. The analytical tecniques used are in full compliance with 10 CFR 50, Appendir K and satisfy the requirements of Reference 2.
As required by Appendix K of 10 CFR 50, certain conservative as-sumptions 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 performance of the emergency core cooling system (ECCS). All assumptions and initial operating condition input data used in this reanalysis were the same as was used in the previously applicable LOCA-ECCS anlysis(3) except for 1) the assumed steam generator plugging level was reduced to 3%**,
- 2) the limiting value of the heat flux hot channel factor was increased to 2.19, 3) the core inlet temperature value was increased to 536.0*F, 4) the value of the limiting enthalpy rise hot channel factor was increased to 1.55, 5) the change of several system parameters was made to conservatively reflect the current system configuration,
- 6) the change of several steam generator model parameters was made to reflect the replacement steam generators, and 7) the appropriate value of the Thermal Design RCS Flowrate was increased from 79,650 gpm to 88,500 gpm.
- As noted in Reference 2, the reanalysis of the small break LOCA is not necessary, and therefore, the analysis of this accident submitted by Reference 4 remains applicable.
- With the completion of the Steam Generator Replacement Programs at the respective Surry units, the percentage of steam generator tube plugging will be zero. The more conservative level of 3% was assumed in the analysis.
2299 274
r
2.0 DESCRIPTION
OF POSTULATED MAJOR REACTOR COOLANT PIPE RUPTURE (LOSS OF COOLANT ACCIDENT - LOCA)
A LOCA is the result of a rupture of the Reactor Ceolant System (RCS) piping or of any line connected to the system. The system boundaries considered in the LOCA analysis are defined in the FSAR.
Sensitivity studies (5) have indicated that a double-ended cold leg guillotine (DECLG) pipe break is limiting.
Should a DECLG break occur, rapid depressurization of the RCS occurs. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. A Safety Injection System (SIS) signal is actuated when 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, assumes the most limiting single failure event. These countermeasures will limit the consequences of the accident in two ways:
1.
Reactor trip and borated water injection complement void for-mation in causing rapid reduction of power to a residual level corresponding to fission product decay 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 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 blowdown, heat from decay, hot internals and the vessel continues to be transferred to the reactor coolant system. At the beginning of the blow-down 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 CFR50.
Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection to steam os the major heat transfer mechanisms. During the~ refill 2299 275
e a
period, it is assumed that rod-to-rod radiation is the only core heat transfer mechanism.
The heat transfer between the reactor coolant system and the secondary system may be in either direction depending on the relative temperatures.
For the case of continued heat addition to the secondary side, secondary side pressure increases and the main safety valves may actuate to reduce the pressure.
Make-up to the secondary side is automatically provided by the auxiliary feedwater system. Coincident with the Safety Injection Signal, normal feed-water flow is stopped by closing the main feedwater control valves and tripoing the main feedwater pumps.
Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps. The secondary side flow aids in the reduction of reactor coolant system pressure. When the reactor coolant system depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. The conservative assumption is then made that 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 10CFR50.
In addition, the reactor coolant pumps are assumed to be tripped at the initiation of the accident and ef fects of pump coastdown are included in the blowdown analysis.
The water injected by the accumulators cools the core and subsequent operation of the low head safety injection punps 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 water 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. Thet e are three-distinct 2299 276
e 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.
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.(0) 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 described in detail in WCAP-8306(7), WCAP-8326(0), WCAP-8171(9) and WCAP-8305 (10), respectively.
These codes are able to assess whether sufficient heat transfer geometry and core amenability to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown and the WREFLOOD computer code is used to calculate the transient during the refill and reflood phases of the accident. The COC0 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 C0CO 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 2299 277
e 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 model to determine the core flooding rate (i.e., the rate at which coolaat 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 flooding rate and the local 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 from WREFLOOD are used by LOCTA-IV in its calcu-lation 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 to be the shape that produces the naximun peak clad temperature (11} )
The COCO code, which is also used throughout the LOCA analysis, calculates'the containment pressure.
Input to COC0 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 containment cooling system are input to COCO. These initial containment conditions and assumed modes of operation are provided in Table 2.
2299 278
3.0 DISCUSSION OF SIGNIFICANT INPUT Significant changes in the input used in this reanalysis from those used in the currently applicable analysis are delineated in Section 1.0 and are discussed in more detail below.
The changes made in this analysis reflect the operational conditionc and limits necessary to allow full power operation at steam generator tube plugging levels of up to 3%.*
The most significant input change for this analysis is the reduction in assumed steam generator plugging. The currently applicable analysis had incorporated a plugging level of 28%. The plugging level assumed for this analysis was reduced to 3%.
The reduction in steam generator plugging level assumed in the analysis allowed a change in several other input values. The reduction in plugging level is an analysis benefit which allowed the heat flux hot channel factor to be increased from the current value of 2.05 to a value of 2.19 for this analysis. The plugging level benefit also allowed an increase in the limiting enthalpy rise hot channel factor from a value of 1.45 to a value of 1.55.
The corresponding limiting assembly enthalpy rise factor was also increased consistent with Westinghouse methodology. The assumed plugging level also affects the core inlet temperature used in the analysis.
The reduction in plugging level required an increase in core inlet temperature from 534.5'F to a value of 536*F for this analysis. This value is the best estimate inlet temperature as determined from past operational data and is adequate to encompass the assumed steam generator plugging level of 3%.
No uncertainty has been added to the inlet temperature input for conservatism.
RCS flow rate has been increased from 79.650 gpm to 88,500 gpm which is the appropriate Thermal Design RCS Flowrate for use in this analysis.
- With the completion of the Steam Generator Replacement P.ograms at the respective Surry units, the percentage of steam generator tube plugging will be zero. The more conservative level of 3% was assumed in the analysis.
2299 279
In conjunction with the Steam Generator Replacement Program, several additional system modifications were made primarily to satisfy Net Positive Suction Head (NPSH) concerns and were conservatively accounted for in the analysis. Specifically, changes to the Containment Spray System, the Outside Recirculation Spray System, the Low Head Safety Injec-tion System and the minimum allowable containment temperature were incor-porated in the model.
The Containment Spray System was modified by adding additional spray header capacity. The addirianal capacity requires a longer fill time which results in a greater time to actuation of the spray system.
Consequently, the actuation time for the Containment Spray System was conservatively increased from 52 to 59 seconds.
In addition, the Outside Recirculation Spray System was modified by removing the flow reducing devices which had been installed as part of the interim NPSH solution. The runout pump flowrate assumed in the analysis was conservatively increased from 2250 gpm to 3500 gpm with a corresponding conservative reduction in actuation time from 410 seconds to 365 seconds.
Since the peak clad temperature occurs at approximately 150 seconds (See Figure 7), the changed Outside Recirculation Spray System parameters did not affect the limiting analysis results. NPSH concerns also required the installation of flow limiting devices in the Low Head Safety Injection system.
The minimum Safety Injection flow based on the modified system was determined and is conservative with respect to that provided in Reference 16.
The conserva-tive Safety Injection flow values were used in this analysis. Another system change assumed in the analysis was a reduction in the containment initial temperature from 90*F to 80*F.
This change is in a conservative direction for LOCA-ECCS ar.alyses and was made to be consistent with the minimum value of the allowable containment temperature range. Current containment tempera-ture range restrictions relating to NPSH concerns will be maintained and are consistent with the minimum temperature reduction used in this analysis.
2299 280
a Several changes to the steam generator model parameters assumed in this analysis were made to accurately reflect the replacement steam generators. As discussed in the Vepco Steam Generator Repair Program submittal (Reference 17) and delineated in the NRC Safety Evaluation Report (Reference 2),
the only thermal-hydraulic parameters that changed were the reactor coolant side pressure drop and flow area and the tube length.
The magnitude of these parameter changes was small, and the operational characteristics of the old and the replacement steam generator are essentially the iame.
Therefore, the specific use of the new steam generator design in this LOCA-ECCS analysis did not significantly influence the analyses results.
Finally, this analysis wac conducted with the February, 1978 version of the Westinghouse LOCA-ECCS Evaluation Model(12, 13, 14). This model version includes a modification to the SATAN VI and LOCTA IV codes to properly account for the volumetric heat generatio.n due to the metal-water reaction 2299 28i
e 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 precents 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 DECLG has been determined to be the limiting break size and location based on the sensitivity studies reported in Reference 5.
Further, all previous LOCA-ECCS submittals for the Surry units have resulted in the Cp = 0.4 dis-charge coefficient being the limiting break size. The applicability of this conclusion (i.e. Cp = 0.4 is the limiting break size) for this analysis was explicitly verified.
(See Table 4).
Consequently, only the results of the most limiting break size are presented in the figures and remaining tables in this submittal. The current analysis resulted in a limiting peak clad temperature of 2190*F, a maximum local cladding oxidation level of 7.99 pe rcent, and a total core metal-water reaction of less than 0.3 percent. The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2 through 18.
2299 282
5.0 CONCLUSION
S For breaks up to and including the double-ended severance of a reacter coolant pipe and for the operating conditions specified in Tables 1 and 2, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46. That is:
1.
The calculate > 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 ir terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17% are not exceeded during or af ter quenching.
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.
2299 283
6.0 REFERENCES
1.
Letter from NRC (J. F. Stolz) to Westinghouse (T. M. Anderson) dated August 29, 1978.
2.
Letter from NRC (A. Schwencer) to Vepco (W. L. Proffitt) dated December 15, 1978.
3.
Letter from Vepco (C. M. Stallings) to NRC (H. R. Denton) dated December 26, 1978, Serial No. 736.
4.
Letter from Vepco (C. M. Stallings) to NRC (K. R. Goller), Serial No.
500-S, dated June 6, 1975.
5.
Buterbaugh, T.
L., Johnson, W. J. and Kopelic, S.
D., " Westinghouse ECCS-Plant Sensitivity Studies," WCAP-8356, July 1974.
6.
Bordelon, F.
M., Massie, H. W., and Zordan, T.
A., " 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 (C0CO)," WCAP-8326, June 1974.
9.
Kelly, R. D.,
et al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974.
10.
Rordelon, F.
M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, June, 1974 11.
Letter from Vepco (C. M. Stallings) to NRC (E. G. Case) dated February 17, 1978, Serial No. 092.
12.
" Westinghouse ECCS Evaluation Model - February,1978 Version", WCAP-9220:
P-A (Proprietary) and WCAP-9221-P-A (Non Proprietary), February 1978.
13.
Letter from Westinghouse (T. M. Anderson) to NRC (J. F. Stolz), dated November 1, 1978, Serial No. NS-TMA-1981.
14.
Letter from Westinghouse (T. M. Anderson) to NRC (R. Tedesco), dated December 11, 1978, Servial No. NS-TMA -2014.
15.
Letter from Westinghouse (C. Eicheldinger) to NRC (J. F. Stolz), dated
~
April 7, 1978, Serial No. NS-CE-1751.
16.
19tter from Vepco (C. M. Stallings) to NRC (H. R. Denton), dated November 22, 1978, Serial No. 069C/013178.
17.
Letter from Vepco (C. M. Sta? lings) to NRC (E. G. Case) dated August 17, 1977, Serial No. 351 as revised.
18.
Letter from Vepco (C. M. Stallings) to NRC (E. G. Case) dated December 28, 1977, Serial No. 587.
2299 284
TABLE 1 INITIAL RCS CONDITIONS Core Power, Mut, 102% of 2441 Peak Linear Power, Kw/ft, 102% of 13.59 Peaking Factor (Fq) 2.19 Accumulator Water Volume, ft 975 (per accumulator)
Reactor Coolant System Flow, gpm 88,500(per loop)
(Thermal Design)
Steam Generator Tube Plugging Level, %
3 Inlet Temperature,
'F, 536.0 Temperature of the Fluid in the Upper Head 100% of THOT Fuel Temperatures Generic Westinghouse 15 x 15 Enthalpy Rise Hot Channel Factor 1.55 Most Limiting Fuel Region Cycle Region Unit 1 All All Unit 2 All All 2299 285
TABLE 2 CONTAINMENT DATA (DRY CONTAINMENT)
NET FREE VOLUME 1.863x10 Ft INITIAL CONDITIONS Pressure (total) 9.35 psia Temperature 80*F RWST Temperature 40*F Service Water Temperature 32.5'F**
Outside Temperature 9*F SPRAY SYSTEM I CONTAINMENT SPRAY SYSTEM Number of Pumps Operating 2
Runout Flowrate 3200 gpa Actuation Time 59 secs SPRAY SYSTEM II - INSIDE RECIRCULATION SPRAY SUBSYSTEM Number Pumps Operating 2
Runout Flowrate (each) 3500 gpm Actuation Time 190 secs 6
Heat Exchanger (UA (per pump))
5.18x10 BTU / H2 *F Service Water Flow (per exchanger) 6900 gpm SPRAY SYSTEM II - OUTSIDE RECIRCULATION SPRAY SUBSYSTEM Number Pumps Operating 2
Runout Flowrate (each) 3500 gpm Actuation Time 365 secs b
Heat Exchanger (UA (per pump))
5.18x10 BTU /HR *F Service Water Flow (per exchanger) 6900 gpm 2299 286
TABLE 2 (Continued)
STRUCTURAL HEAT SINKS Type / Thickness (in.)
Area (ft ), w/ uncertainty Concrete /6 8,393 Concrete /12 62,271 concrete /18 55,365 Concrete /24 11,591 Concrete /27 9,404 Concrete /36 3,636 Carbon Steel /0.375 46,489*
Concrete /54 Carbon Steel /0.50 25,652*
Concrete /30 Concrete /26 (Floor) 12,110 Carbon Steel /0.239 158,059*
Stainless Steel /0.306 17,519 Aluminum /0.0091 3,911
- Credit for painted surfaces was taken only for the nominal surface area
- Sensitivity analyses provided in Reference 18 demonstrate that service water temperature levels as low as 25'F will have a negligible impact on the limiting results of the LOCA-ECCS analyses.
2299 287
TABLE 3 TIME SEQUENCE OF EVENTS DECLG CD = 0.4 (Sec)
START 0.0 Reactor Trip 0.520 S. I. Signal 2.33 Acc. Injection 17.3 End of Bypass 26.23 Pump Injection 27.33 End of Blowdown 29.72 Bottom of Core Recovery 37.82 Acc. Empty 46.50 2299 288
TABLE 4 RESULTS FOR DECLG CD = 0.4 Cp = 0.6 Cp = 1.0 Peak Clad Temp, *F 2190 2045 1868 Peak Clad Location, Ft.
7.75 7.75 7.5 Local Zr/H O RXN (max), %
7.99 5.08 2.83 2
Local Zr/H,0 RXN (max),
7.75 7.75 7.5 Location, Ft.
Total Zr/H2O RXN,
<0.3
<0.3
<0.3 Hot Rod Burst Time, sec.
27.90 28.4 57.2 Hot Rod Burst Location, Ft.
5.75 5.75 6.0 2299 289 I"
TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLG (C 0.4)
=
D TIME (SEC)
TOTAL MASS TOTAL ENERGY 5
FLOWRATE (LB/SEC)
FLOWRATE (10 BTU /SEC) 42.1 44.56 0.5775 49.9 246.03 1.5419 63.3 279.42 1.5716 80.2 291.62 1.5401 99.1 297.73 1.4964 119.9 302.30 1.4429 1 66.6 309.76 1.3239 221.8 317.80 1.1179 294.0 328.78 1.0028 2299 290
TABLE 6 BROKEN LOOP ACCUMULATOR ROW TO CONTAINMENT DECLG, C "
- D TIME (SEC)
MASS FLOWRATE* (LBm/SEC) 0.0 0.0 1.0 4212 3.0 3598 5.0 3194 7.0 2901 10.0 2573 16.0 2129 20.0 1924 22.0 1844 22.84 0.0 2299 291
- For energy flowrate multiply mass flowrate by a constant of 49.67 BTU /LBM.
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ATTACIDfENT 2 2299 310
TS 3.12-4 Unit 1 Unit 2 Fq(Z) 12.05/P x K(Z) for P > 0.5 Fq(Z) 1 2.19/P x K(Z) for P > 0.5 F (Z) 1 4.38 x K(Z) for P 1 0.5 F (Z) 1 4.10 x K(Z) for P 1 5
9 q
ggi1.0 Q+0.2@$ x M (H 1 1.55 (1+0.2(1-P)) x T(BU)
F A
A Assm.
1.476/P Assm.
1.38/P F
F AH LH Y
1 1.45/P F
1 1.55/P H
H where P is the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8a for Unit 1 and Figure 3.12-8b for Unit 2, Z is the core height location of Fq, and T(BU) is theinterinthimblecellrodbowpenaltyonF{H given in TS Figure 3.12-9.
2.
Prior to exceeding 75% power following each core loading, and during each effective full power month of operation thereafter, power distribu-tion maps using the movable detector system, shall be made to confirm that the hot channel factor limits of this specification are satisfied.
For the purpose of this confirmation:
Themeasurementoftotalpeakingfactor,Fheas,shallbeincreased a.
by eight percent to account for manufacturing tolerances, measure-ment error, and the effects of rod bow.
The measurement of enthalpy rise hot channel factor, the hot assembly enthalpy rise factor,
^
FfH and the hot rod enthalpy rise factor, FII shall be H
Ro increased by four percent to account for measurement error.
If any measured hot channel factor exceeds its limit specified under 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under 3.12.B.1 are met.
If the hot channel factors cannot be brought to within the limitg listed below within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower LT and Overtemperatu e AT trip setpvints shall be similarly reduced.
2299
- 11
TS 3.12-4a Unit 1 Unit 2 q 1 2.05 x K(Z)
Fq 12.19 x K(Z)
F FfH 1 1.55 x T(BU)
FfH 1 1.55 x T(BU)
LOCA FN LOCA AH Assm. < 1,3g pN Assm.
l.476 AH FN LOCA yN-LOCA < l.55 AH Rod 1 1.45 AH Rod -
b.
Fq(Z) shall be evaluated for normal (Condition I) operation of Unit 2 by combining the measured values of Fxy(Z) with the design Condition I axial peaking factor values, Fg(Z), as listed in TS Table 3.12-1B.
For the purpose of this specification Fxy(Z) shall be determined between 1.5 feet and 10.5 feet elevations of the core exclusive of grid plane regions located at 25.9 13.2 inches, 52.1 13.2 inches, 78.3 13.2 incher., and 104.5 13.2 inches. The measured values of Fxy(Z) shall be increased by nine percent to account for manufactur-ing tolerances, measurement error, rod bow, xenon redistribution, and any burnup dependent peaking factor increases. If the results of this evaluation predict that Fq(Z) could potentially violate its limiting values as established in Specification 3.12.B.1, either:
(1) the thermal power and high neutron flux trip setpoint shall be reduced at least 1% for each 1% of the potential violation (for the purpose of this specification, this power level shall be called P I
THRESHOLD '
(2) movable detector surveillance shall be required for operation when the reactor thermal power exceeds P s sur-THRESHOLD.
veillance shall be performed in accordance with the following:
j (a) The normalized power discritution, Fq(Z)-
, from APDM thimble j at core elevation Z shall be measured utilizing at least two thimbles of the movable incore flux system for 2299.12
TS 3.12-14 F (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the maximus q
local heat flux on the surface of a fuel rod at core elevation Z d!.vided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
Ff, Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
FN, Nuclear Enthalpy Rise Hot Channel Frctor, is defined as the ratio of the g
integral of linear power along the rod with the highest integrated power to the average rod power for both LOCA and non-LOCA considerations.
A F
Assm., Hot Assembly Nuclear Enthalpy Rise Factor, is defined as the ratio AH of the integral of linear power along the assembly with the highest integrated power to the average assembly power.
It should be noted that the enthalpy rise factors are based on integrals and are used as such in the DNB and LOCA calculations. Local heat fluxes are obtained by using hot channel and udjacent channel explicit power shapes which take into account variations in re. dial (x-y) power shapes throughout the core.
Thus, the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factors. The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using an upper bound envelope of 2.05 (Unit 1) or 2.19 (Unit 2) times the hot channel factor nor=alized operating envelope given by TS Figures 3.12-8a and 3.12-8b.
2299
.:13
l TS FIGURE 3.12-8a HOT CHANNEL FACTOR NORMALIZED OPERATING ENVF. LOPE SURRY POWER STATION UNIT NO. 1 ei m i m
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