ML20111B781
| ML20111B781 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/06/1985 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20111B769 | List: |
| References | |
| NUDOCS 8503130225 | |
| Download: ML20111B781 (79) | |
Text
{{#Wiki_filter:* TABLE 2.2-1 , E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS e Total Allowable Value
}
p Functional Unit
- 1. Manual Reactor Trip Allowance (TA)
Not Applicable Z NA S NA Trip Setpoint NA NA
- 2. Power Range, Neutron Flux 7.5 4.56 0 1109% of RTP 1111.2% of RTP High Setpoint Low Setpoint 8.3 4.56 0 125% of RTP 127.2% of HTP g.
- 3. Power Range, Neutron Flux 1.6 0.5 6 15% of RTP with <6.3% of RTP with a time constant h
High Positive Rate a time constant 12 seconds 12 seconds g
- 4. Power Range, Neutron Flux 1.6 0. 5' O 15% of RTP with 16.3% of RTP with f High Negative Rate a time constant a time constant k 7 a 12 seconds 12 seconds k
- 5. Intermediate Range, 17.0 8.4 0 125% of RTP $31% of RTP C
Neutron Flux 6 1
- 6. Source Range, Neutron Flux 17.0 10.0 0 110s cps $1.4 x 105 cps
- 7. Overtemperature AT 7.1 2.94 1.8 See note 1 See note 2
- 8. Overpower AT 4.5 1.4 1. 2 See note 3 See note 4
- 9. Pressurizer Pressure-Low 3.1 0.71 1.5 11870 psig 11859 psig
- 10. Pressurizer Pressure-High 3.1 0.71 1.5 12380 psig 12391 psig
- 11. Pressurizer Water Level-High 5.0 2.18 1.5 192% of instrument 193.8% of instrument span span
- 12. Loss of Flow 2.5 1.0 1.5 190% of loop 189.2% of loop design flow" design flow *
- 942DO '
'; Loop design flow = g e t gpm ATTACHMENT I PAGE 6 OF 15 RTP = PATFn THFRMAL POWER 8503130225 850306 PDR ADOCK 05000395 P PDR
ATTACHMENT I PAGE 7 OF 15 POWER DISTRIBUTION LIMfT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR NOT CHANNEL FACTOR (Continued)
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6. are maintained.
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F" will be maintained within its limits provided conditions a. through
- d. abo are maintained. As noted on Figur .2-3h0.2-- RCS flow rate and Fg may be " traded off" against one another (i.e. , a low measured RCS flow rateisacceptableifthemeasuredFhisalsolow)toensurethatthe The relaxation of f calculated DNBR will not be below the design DN8R value. >
F" as a function of THERMAL POWER allows changes in the radial power shape
~for all permissible rod insertion limits. i pascalculatedin3.2.3andusedinFigure3.2.3,accountsforFh less than or equal to 1.49. This value is used in the various accident N influences parameters other than DNBR, e.g., peak clad analyses where Fg 8 temperature and thus is the maximum "as measured" value allowed. 7.= ,
ce; :.,.. er. C;= v 1,. Th;;_ 4;f L.ed , .'. i n.; f:: +he Sehr hr ;f e .c.e';, f..
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measurement is taken, an allowance for both experimental error When an F g and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The radial peaking factor F,y(Z) is measured periodically to provide The assurance that the hot channel factor, F (Z), remains within its limit. RTP)0as provided in the Radial Peaking F xy limit for Rated Thermal Power (F xy . Factor Limit Report per specification 6.9.1.14 was determined from expected l power control maneuvers over the full range of burnup conditions in the core. l When RCS flow rate and F N are measured, no additional allowances are necessarypriortocomparisonwiththelimitsofFigure[3.2-3[2.0C N Measurement errors of 3.5% for RCS total flow rate and 4,% for F# have been l 9 ;;i;; C;C ei.;. q
~..;;i:n ;f 0.:
allowed for in e+ 7.1-3. delero iniq TG kis of Fywe. SUMMER - UNIT 1 8 3/4 2-4 l l l
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f POWER DISTRIBUT20N LIMIT I BASES . PY RISE NEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRAT HOT CHANNEL FACTOR (Continued) The 12 hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. creusi i; avu'Iowls kw valus wI CSC sobsu. I;;I . sd be iGy . ;d;;;; kris Inis GreQ1L Gome> Igvu. a ssne.-is u as g iu g LO d66uun6 tor Rud
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3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides A DNS and linear heat generation rate protection with x-y plane power tilts. limiting tilt of 1.025 can be tolerated before the margin for uncertainty The limit of 1.02 was selected to provide an allowance for F is depleted. q the uncertainty associated with the indicated power tilt. The- two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-In the ev rection of a dropped or misaligned centrol rod. is reinstated by , l not correct the tilt, the margin for uncertainty on Fg l reducing the maximum allowed power by 3 percent for each percent of ti) excess of 1.0. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm th the normalized symmetric power distribution is consistent These locations are C-8, E-5, POWER TILT RATIO. flux map or two sets of 4 syssetric thimbels. E-11 H-3, H-13. L-5, 2-11, N-8. 3/4.2.5 DNS PARAMETERS f The limits on the ONB related parameters assure that each of the l parameters are maintained within theThe normallimits steady state envelope are consistent I assumed in the transient and accident analyses. with the initial FSAR assumptions and have been The 12 hour periodic surveillance of these parameters through instrumen readout is sufficient to ensure that the parameters are restored within thei limits following load changes and other expected transient operation. ATTACHNENT I PAGE 8 OF 15 SUPMER - UNIT 1 8 3/4 2-5 l r _ _ - _ _ _ - __ ___m _ _ _
JE nrerib A Tech. Spec. Insert Fuel rod bowing reduces the value of DNB ratio.. Credit is available to offset this reduction in the generic margin. The generic, design. margins, totaling 9.1% DNBR, completely offset any rod bow penalties. This margin includes the following:
- 1) Design limit DNBR of 1.30 vs.1.28
- 2) Grid Spacing (K
- 3) Thermal Diffusi$n) of 0.046 Coefficient vs. 0.059 of 0.038 vs. 0.059
- 4) DNBR Multiplier of 0.86 vs. 0.88
- 5) Pitch reduction The applicable value of rod bow penalties are referenced in the FSAR.
t . . 6 d I l ATTACHMENT I PAGa 9' OF 15
, n ---e , .-- --n Q
i POWER DISTRIBUTION LIMITS l 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR I
' LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow
~
rate and shall be maintained within the region of allowable operation - shown on F gur 3.2-3 for 3 loop operation. ) Where: N -
- AH _ ,
- a. = 1.49 [.1.0 + 0.2 (1.0 - P)]
=-
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THERMAL POWER , 6 e. P = RATED THERMAL POWER c.d. F N
= Measured values of F" obtained by using the movable incore detectors to obtain a power distribution map. The measured values of Fh shall be used to calculate R since Figure 3.2-3 includes measurement uncertainties of 3.5% for flow and 4% for N
incore measurement of Fg , and
^
I0I (O'v') -- Eed Oww funel 6y na a 7.netien v7 asylwn sw.isys uvanup en Aa-.. $.. E ivvi s 3.2 4, unwis a region in usianeQ as T.n05r
----=h14mc with th. --- 19z n+; (j;baua) or enrir.nmuni.
/ f 4 v=ct en=4-APPLICABILITY: MODE 1.
ACTION: outside the region of With the combination of RCS total flow rate and R acceptable operation shown on Figure 3.2-3:
- a. Within 2 hours either:
- 1. Restore the combination of RCS total flow rate and to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POW and reduce the Power Range Neutron Flux - High trip setpoint to
' less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
s !
- SUPMER - UNIT 1 3/4 2-8 )
ATTACHMENT I PAGE 10 OF 15
-- -- -- - _ - . . . . . v . . -
ATTACHMENT I PAGE 11 OF 15 i POWER DISTRIBUTION LIMITS i ACTION: (Continued)
- b. Within 24 hours of initially being outside the above limits, verify ll through incore flux ma g and RCS total flow rate comparison that F the combination of and RCS total flow rate are restored to l
within the above li r reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.
- c. Identify and correct the cause of the out-of-limit condition prior e to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION items a.2. and/or b. above; subsequent POWERand OPERATION may proceed provided that the combination of flux indicated RCS total flow rate are demonstrated, through mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal 75% of RATED THERMAL POWER, and
- 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
i SURVEILLANCE REQUIREMENTS > 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. shall be 4.2.3.2 The combination of indicated RCS total flow rate and 3.2-3: determined to be within the region of acceptable operation of F gu
- a. Prior to operation above 75% of RATED THERMAL POWER after each f;e1 loading, and ,
- b. At least once per 31 Effective Full Power Days.
i 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the l region of acceptable operation of Figure 3.2-3 at obtainedleast once perper 12 hours when Specification ) themostrecentlyobtainedvalugofR .R 4.2.3.2, assumed to exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a' CHANNEL CALIBRATION at least once per 18 months. ' 4.2.3.5 The RCS total flow rate shall be detemined by measurement at least once per 18 months. SUPMER - UNIT 1 3/4 2-9 l 1 I
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- l EIEE i MEASUREMENT UNCERTAINTIES OF 3.5% FOR FLCW S' l AND 4.tife FO'R INCORE MEhSl.lREMENT OF'F"g' INCLUDE'D::W ,
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- f Figurs 2.11 Reactor Core Safety Lirst .Three Loops in Operation '
f.Avit M KTP (
- When operating in ".:;;i:- m of Techn cal Specificial 3.2.3 (Figure 3.2-3), the restricted power level n:ust be considered i 100% RTP for this figure.
SUMMER - UNIT 1 2-2 s ATTACHMENT I PAGE 14 OF 15
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g ATTACHMENT I PAGE 15 OF 15 Amendaant No. 19 5tMER - UNIT 1 3/4 2 11
----_____ .o
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- 2. Inlet Flow Ma1 distribution:
The consideration of inlet flow maldistribution in core thermal performances is discussed in Section 4.4.3.1.2. A design basis of 5 percent reduction in coolant flow to the hot assembly is used in the THINC-IV analysis.
- 3. Flow Redistribution:
l The flow redistribution accounts for the reduction in flow in the hot channel resulting from the high flow resistance in the channel due to the local or bulk boiling. The effect of the nonuniform power distribution is inherently considered in the THINC analysis for every operating condition which is evaluated.
- 4. Flow Mixing:
The subchannel in reactor mixing design modelonincorporated is based experimental in the(TH4NC Code and used data 51J discussed in l Section 4.4.3.4.1. The mixing vanes incorporated in the spacer grid design induce additional flow mixing between the various flow channels in a fuel assembly as well as between adjacent assemblies. This mi::ing reduces the enthalpy rise in the hot channel resulting j from local power peaking or unfavorable mechanical tolerances. (PLAct IA/sEPJ 4.4.2.4 6 6d Fluz Tilt Considerations i Significant quadrant power tilts are not anticipated during normal operation since this phenomenon is caused by some asymmetric perturbation. A dropped or misaligned rod cluster control assembly could cause changes in hot channel factors; however, these events are analyzed separately in Chapter 15. This discussion will be confined to flux tilts caused by x y xenon transients, inlet temperature mismatches, enrichment variations within tolerances and so forth. N The design value of the enthalpy rise hot channel factor F H' " " includes an eight percentuncertainty(asdiscussedinSechton4.3.2.2.7),
- is assumed to be sufficiently conservative that flux tilts up to and includingtgealarmpoint(seeTechnicalSpecifications)willnotresultin values of F greater-than that assumed in this submittal. The design l value of F oes not include a specific allowance for quadrant flux tilts.
i
)
4.4.2.5 Void Fraction Distribution The calculated core average and the hot subchannel maximum average void ! fractions are presented in Table 4.4-3 for operation at full power with l design hot channel factors. The void fraction distribution in the core at )
! various radial and axial locations is presented in' Reference (52]. The void models used in the THINC-IV computer code are described in Section 4.4.2.8.3.
I i . 4.4-15 Attachnent II Page 1 of 69 l
, _ . . _ _ . , _ . - - ~ . . , .
i FSAR INSERT B (NEW SECTION) 4.4.2.3.4.3 - Effects of Rod Bow on DNBR The phenomenon of fuel rod bowing, as described in Reference (95), must be accounted for in the DNBR safety analysis of Condition 1 and Condition II events for each plant application. Applicable generic credits for margin resulting f rom retained conservatism in the evaluation of the DNBR and/or margin obtained from the measured plant operating parameters (suchasF[N or core flow) -- which are less limiting that those required by the plant safety analysis -- can be used to offset the effect of rod bow. For the safety analysis of the Virgil C. Summer Unit, sufficient margin was maintained (9.1%)* to accommodate full and low flow DNBR penalties (<3% for the worst case, which is a burnup of 33,000 MWD /MTU) identified in Reference (96). The maximum rod bow penalties accounted for in the design safety analysis are based on an assembly average burnup of 33,000 MWD /MTU. At burnups greater than 33,000 MWD /MTU, credit istakenfortheeffectofF[g burndown, due to the decrease in fissionable isotopes and the buildup of fission product inventory, and no additional rod bow penalty is required.
- Design Limit DNBR of 1.30 vs. 1.28 Grid Spacing (Ks) of 0.046 vs. 0.059 Thermal Diffusion Coefficient of 0.038 vs. 0.059 DB Multiplier of 0.86 vs. 0.88 Pitch Reduction ATTACHMENT II PAGE 2 OF 69 l
l
FSAR MODIFICATION ' (_ Add following to FSAR Section 4.4 Reference List: l
- 95. Skaritka, J. ( Ed . ) , " Fuel Rod Bow Evaluation," WCAP-8691 Revision 1, July, 1979.
- 96. " Partial Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1" letter, E. P. Rahe, Jr. (Westinghouse to J. R. Miller (NRC), NS-EPR-2515, dated October 9,1931; " Remaining Response to Request Number 1 for Additional Information on WCAP-8 691, Revision 1" letter, E. P. Rahe, Jr. (Westinghouse) to J.
R. Miller (NRC), NS-EPR-2572, dated March 15, 1982. Attachment II Page 3 of 69
=
i 15.4.1 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT Identification of Causes and Frecuency Classification A loss-of-coolant accident (LOCA) is the result of a pipe rupture of the RCS pressure boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 f t . This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of The Virgil C. Sumer, but is postulated as a conservative design basis. results for the small break loss of coolant accident are presented in Section 15.3.1. The boundary considered for loss of coolant accidents are related to connecting pipe as defined in Section 3.6. l The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (10 CFR 50.46 and Appendix K of 10 CFR 50 1974)III as follows:
- 1. The calculated peak fuel element clad temparature is below the requirement of 2,200'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 to cooling. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
- 4. The core remains amenable to cooling during and af ter the break.
- 5. The core temperature is reduced and decay heat is removed for an extended period'of time, as required by the long-lived radioactivity remaining in the core.
1 ATTACHMENT II PAGE 4 0F 69 I 79890:10/103184
I i j These criteria were establishsd to provide significant margin in emergency core cooling system (ECCS) performance following a LOCA. WASH-1400 (USNRC 1975) presents a recent study in regards to the probability of occurrence of RCS pipe ruptures. Seouence of Events and Systems Operations Should a, major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. A safety injection signal is generated when the appropriate setpoint is reached. These countermeasures will limit the consequences of the accident in two ways:
- 1. Reactor trip and borated water injection supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat. However, no credit is taken in the LOCA analysis for the boron content of the injection water. In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis.
- 2. Injection of borated water provides for heat transfer from the core and prevent excessive clad temperatures.
Descriotion of large Break Loss-of-Coolant Accident Transient The sequence of events following a large break LOCA is presented in Table 15.4.1-1. Before the break occurs, the unit is in an equilibrium condition; that is, the heat generated in the core is being removed via the secondary system. During blowdown, heat from fission product decay, hot internals and the vessel, continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat i from the core by forced convection with some fully developed nucleate : boiling. After the break develops, the time to departure from nucleate l l l ATTACHMENT II PAGE 5 0F 69 , l l ! 79890:10/103184 l
boiling is calculated, consistent with Appendix K of 10 CFR 50.II) Thereafter the core heat transfer is unstable, with both nucleate boiling and film boiling occurring. As the core becomes uncovered, both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms. The heat transfer between the RCS and the secondary system may be in either direction, depending on the relative temperatures. In the case of continued heat addition to the secondary system, the secondary system pressure increases and the main steam safety valves may actuate to limit the pressure. Makeup water to the secondary side is automatically provided by the emergency feedwater system. The safety injection signal actuates a feedwater isolation signal which isolates normal feedwater flow by closing the main feedwater isolation valves, and also initiates emergency feedwater flow by starting the l ! emergency feedwater pumps. The secondary flow aids in the reduction of RCS pressure. When the RCS depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10CFR50. Since loss of offsite power (LOOP) is i assumed, the RCPs are assumed to trip at the inception of the accident. The effects of pump coastdown are included in the blowdown analysis. The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2280 psia) falls to a value approaching that of the containment i atmosphere. Prior to or at the end of the blowdown, the mechanisms that are responsible for the emergency core cooling water injected into the RCS bypassing the core are calculated not to be effective. At this time (called ; end-of-bypass) refill of the reactor vessel lower plenum begins. Refill is l completed when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time). ATTACHMENT II PAGE 6 0F 69 7989Q:10/103184
6 The reflood phase of the transient is dofined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated. From the latter stage of blowdown and then the beginning-of-reflood, the safety injection
; accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downtomer water elevation head provides the driving force required for the reflooding of the reactor core. The low head and high head safety injection pumps aid in the filling of the downccmer and subsequently supply water to maintain a full downtomer and complete the reflooding process.
Continued operation of the ECCS pumps supplies water during longterm cooling. Core temperatures have been reduced to longterm steady state levels associated with dissipation of residual heat generation. After the water level of the refueling water storage tank (RWST) reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching'to the cold recirculation phase of operation, in which spilled borated water is drawn from the engineered safety features (ESF) containment sumps by the low head safety injection (residual heat removal) pumps and returned to the RCS colo legs. Thecontainmentspraysys(emcontinuestooperatetofurtherreduce containment pressure. Approximately 24 hours after initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid concentration in the reactor vessel. Core and System Performance j Mathematical Model:
- s l The requirements of an acceptable ECCS evaluation model are presented in '
Appendix K of 10 CFR 50 (Federal Register 1974).O ATTACHMENT 11 PAGE 7 0F 69 19890:10/103184
Large Break LOCA Evaluation Model The analysire of.a large break LOCA transient is divided into three phases: (1) blowdown. (2) refill, and (3) reflood. There are three distinct
- transien*.s analyzed in each phase, including the thermal-hydraulic transient j in the RCS, the p essure and temperature transient within the containment, and !
the fuel and clad temperature transient of the hottesi. fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA. A description of the various aspects of the LOCA analysis methodology is given by Bordelon, Massie, and Zordon (1974),( ) This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, BART, LOCTA-IV, and COC0 codes, which are used in the LOCA analysis, are described in detail by Bordelon et a_t. (1974)(5); g,))y e_t a1. (1974)( I; Young et al. (1980)(8); Bordelon and Murphy (1974)(4); Bordelon and Murphy (1974)( I. Code modifications are specified in References 2, 7 and 14. These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyses the thermal-hydraulic transient in the RCS during blowdown ,and the WREFLOOD computer code calculates this transient during the refill and reflood phases to the accident. The COCO code is used for the complete containment pressure history for dry containments. The LOCTA-IV computer code calculates the thermal transient of the hottest fuel rod during the three phases. The Revised Pad Fuel Thermal Safety Model, described in Reference 14, generates the initial fuel rod conditions input to LOCTA-IC. ' SATAN-VI calculates the RCS pressure, enthalpy, density, and the mass and , l energy flow rates in the RCS, as well as steam generator energy t,ransfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator water mass and internal pressure and the pipe break mass and energy flow rates that ! ATTACHMENT II PAGE 8 OF 69 7989Q:10/103184
are assumed to be vented to the containment during blowdown. At the end of i the blowdown phase, these data are transferred to the WREFLOOD code. Also, at the end-of-blowdown, the mass and energy release rates during blowdown are input to the COCO code for use in the determination of the containment
- pressure response during blowdown. 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 f rom the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (that is, the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the reflood phase of the LOCA. WREFLOOD also calculates the mass and energy flow addition to the containment through the i break. WREFLOOO is also linked to the BART and LOCTA-IV codes. The heat i
- transfer calculation for average fuel channel in the hot assembly during the reflood phase of the LOCA is performed by the BART(I coinputer code using a i l
l mechanistic core heat transfer model. This information is used by LOCTA-IV throughout the analysis of the LOCA transient to calculate the fuel clad temperature and metal-water reaction of the hottest rod in the core. ] i The large break analysis was performed with the December 1981 version of the Evaluation Model modified to incorporate the BART( } computer code. l l Input Parameters and Initial Conditions: The analysis presented in this section was performed with a reactor vessel i upper head temperature equal t'o the RCS cold leg temperature. ! The bases used to select the numerical values that are input parameters to the ! analysis have been conservatively determined from extensive sensitivity studies (Westinghouse 1974( ); Salvatori 1974 I 'I). In addition, the requirements of Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions which were made pertain to the conditions of the reactor and associated safety system equipment at the time that tae LOCA t ATTACHMENT II' PAGE 9 OF 69 79890:10/103184 _____. . - . . - - .-. .-- .. T .,. - ,: - -
occurs, and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS. Oecay heat generated throughout the transient is also conservatively calculated. i Results: Based on the results of the LOCA sensitivity studies (Westinghouse 1974 I I; Salvatori 1974I ' }) the limiting large break was found to be the double f ended cold leg guillotine (DECLG). Therefore, only the DECLG break is ! considered in the large break ECCS performance analysis. Calculations were performed for a range of Moody break discharge coefficients. The results of I these calculations are summarized in Tables 15.4.1-1 and 15.4.1-2. The hot spot is defined to be the location of the maximum peak clad temperature. This location is given in Table 15.4.1-2 for each break size analyzed.
> Containment data used to calculated ECCS back pressure is presented in Table 3
15.4.1-3. Figures 15.4.1-la through 15.4.1-16 present the transients for the principal
- parameters for the break sizes analyzed. The following items are noted
Fiqures 15.4.1-la The following quantities are presented at the clad burst through 15.4.1 -3c location and at the hot spot (location of maximum clad temperature), both on the hottest fuel rod (hot rod):
- 1. fluid quality;
- 2. mass velocity; f
- 3. heat transfer coef ficient.
The heat transfer coefficient shown is calculated by the LOCTA-IV code. l 2 Fiqures 15.4.1 -4a The system pressure shown is the calculated pressure in through 15.4.1 -6c the core. The flow rate from the break is plotted as the j sum of both ends for the guillotine break cases. The core pressure drop shown is from the lower plenum, near the care, to the upper plenum at the core outlet. J ! l ATTACHMENT II PAGE 100F 69 l l i 79890:10/103184
l Fiqure 15.4.1 -7a These figures show the hot sp8t clad temper 3ture transient through 15.4.1 -9c and the clad temperature transient at the carst location. The fluid temperature shown is also for the hot spot and burst location. The core flow (top and bottom) is also shown. Fiqures- 15. 4.1 -10a These figures show the core reflood transient. through 15 4.1-10f Fiqures 15. 4.1 -11 a These figures show the Emergency Core Cooling System thrcuch 15 . 4 .1 -12 c flow for all of the cases analyzed. As described earlier, the accumulator delivery during blowdown is discarded until the end of bypass is calculated. Accumulator flow, however, is established in the refill and the reflood calculations. The accumulator flow assumed is the sum of that injected in the intact cold legs. Fiqures 15. 4.1 -13a The containment pressure transient is als'o provided. through 15. 4.1 -13c Fiqures 15.4.1-14 Thesa figures show the core power transient. a.b.c Fiqure 15.4.1-16 This figure provides the containment Gall condensing heat transfer coef ficient for the limiting case break. In addition to the above, Tables 15.4.1-4 and 15.4.1-5 present the reflood mass and energy release to the containment and the broken loop accumulator mass and energy flowrate to the containment, respectively. ; I The maximum clad temperature calculated for a large break is 2132.0*F which is less than the Acceptance Criteria limit of 2200*F. The maximum local metal-water reaction is 4.85 percent, which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46. The total core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10 CFR 50.46. The clad temperature transient is ATTACHMENT II PAGE 110F 69 1989Q: 10/103184 i
-a' ~
n terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided. 1 ATTACHMENT II PAGE 12 0F 69 l 79890:10/103184
REFERENCES FOR SECTION 15.4.1
- 1. " Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix X of 10 CFR 50, Federal Register 1974, Volume 39, Number 3.
Rahe, E. P. (Westinghouse), letter to J. R. Miller (USNRC) Letter No. 2. NS-EPRS-2679, November 1982.
- 3. Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary), WCAP-8326 (Non-Proprietary), June, 1974.
- 4. Bordelon, F. M. el a_l,. , "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary),1974.
- 5. Bordelon, F. M. et al . , " SATAN-VI Program: Comprehensive Space, Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), 1974.
- 6. Bordelon, F. M.; Massie, H. W.; and Zordon, T. A., " Westinghouse ECCS Evaluation Model - Summary," WCAP-8339, 1974.
- 7. Rahe, E. P., " Westinghouse ECCS Evaluation Model, 1981 Version,"
WCAP-9220-P-A (Proprietary Version), WCAP-9221-P-A (Non-Proprietary Version), Revision 1,1981.
- 8. Young, M. Y. et al., "8 ART-A1: A Computer Code for the 8est Estimate Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary) and WCAP-9695-A (Non-Proprietary) .
- 9. Kelly, R. D. et al., " Calculational Model for Core Reflooding Af ter a loss-of-Coolant Accident (WREFLOOD Code," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), 1974.
ATTACHMENT II PAGE 13 0F 69 79890:10/103184
w
- 10. U. S. Nuclear Regulatory Commission 1975, Reactor Safety Study - An Assessment of Accident 0,isks in U. S. Commercial Nuclear Power Plant," l l
WASH-1400, NUREG-75/014. l
- 11. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), 1974.
- 12. " Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP-8341 (Proprietary) and WCAP-8342 (Non-Proprietary), 1974.
- 13. "Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model -
Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-P rop ri eta ry) , 1975.
- 14. Letter from J. F. Stoltz (NRC) to T. M. Anderson (Westinghouse),
Subject:
Review of WCAP-8720, Improved Analytical Models used in Westinghouse Fuel Rod Design Computations. ATTACHMENT II PAGE I4 0F 69 7989Q:10/103184
TABLE 15.4.1-1 h LARGE BREAK TIME SEQUENCE OF EVENIS DECLG (C D = 0.8) DECLG (C D = 0.6) DECLG (C D = 0.4) (Sec) (Sec) (Sec) STAR 1 0.0 0.0 0.0 R2 actor Trip Signal 0.404 0.409 0.418 S. 1. Signal 0.63 0.71 0.87 Acc. Injection 9.24 11.5 15.5 End of Blowdown 22.22 25.14 30.996 Bottom of Core Recovery 34.45 37.11 43.45 Acc. Empty 41.309 44.197 49.576 Pump Injection 25.63 25.71 25.87 End of Bypass 22.22 25.06 30.923 l ATTACHMENT II PAGE 15 0F 69
TABLE 15.4.1-2 LARGE BREAK DECLG (CD = 0.8) DECLG (CD = 0.6) DECLG (CD = 0.4) Results Peak Clad Temp. *F 1866.0 2009.0 2132.0 Peak Clad location Ft. 7.25 7.0 7.0 Local Zr/H 2O Reaction 2.338 3.55 4.85 (max) % Local Zr/H2O Location Ft. 7.25 1.0 7.0 Total Zr/H,0 Reaction % <0.3 <0.3 <0.3 liot Rod Burst Time sec 63.8 46.0 41.0 Hot Rod Burst Location Ft. 6.75 6.0 6.0 Calculation NSSS Power Mwt 102% of 2775 Peak Linear Power kw/ft 102% of 12.632 Peaking Factor (At License Rating) 2.32 3 Accumulator Water volume (ft ) 1000 Fuel region + cycle analyzed :ycle Region UNil 1 ALL ALL 65 Steam Generator Tube Plugging in each steam generator is assumed. ATTACHMENT II PAGE 16 0F 69
TABLE 15.4.1-3 LARGE BREAK CONTAINMENT DATA (ORY CONTAINMENT) 6 3 1.9 x 10 ft NET FREE VOLUME INITIAL CONDITIONS 14.7 psia Pressure Temperature 90*F RWST Temperature 40*F Service Water Temperature NA Outside Temperature 19'F SPRAY SYSTEM Total Flow Rate 6000 gpm Actuation Time 52 secs SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating 2 Fastests Post Accicent Initiation of Fan Coolers 43 secs t ATTACHMENT II PAGE 17 0F 69 7989Q:10/103184
TABLE 15.4.1-3 (Continued) LARGE BREAK CONTAINMENT DATA (DRY CONTAINMENT) STRUCTURAL HEAT SINKS Thickness (In) Area (Ft ) 0.348 Carbon Steel 48.0 Concrete 57,397 0.264 Carbon Steel 36.0 Concrete 20,241 0.125 Carbon Steel 24.0 Concrete 11,694 18.0 Concrete 315 22.56 Concrete 43,537 12.0 Concrete 10,811 48.0 Concrete 19,020 1.52 Stainless Steel 409 1.13 Stainless Steel 551 0.6 Stainless Steel 1939 0.336 Stainless Steel 2194 0.06 Stainless Steel 88481 6.672 Carbon Steel 3300 3.504 Carbon Steel 130 2.376 Carbon Steel 2324 1.7568 Carbon Steel 4323 0.87 Carbon Steel 8787 0.744 Carbon Steel 17734 0.324 Carbon Steel 16929 0.06 Carbon Steel 654508 ATTACHMENT II PAGE18 0F.69 l l 79890:10/103184 _ _ _ _ _ _ x-
TABLE 15.4.1-4
= 0.4)
REFLOOD MASS / ENERGY RELEASES * (CD c TOTAL ENERGY FLOWRATE TOTAL MASS FLOWRATE 5 (L8M/SEC) 10 STU/SEC TIME (SEC) _ 0.0 0.0 44.023 48.177 0.617 50.045 8 66.77 0.841 64.970 127.19 1.121 84.170 292.42 1.515 104.170 ' 300.31 1.469 125.770 340.72 1.483 164.420
- Accumulator nitrogen was released between 53.0 and 93.0 seconds at a mass flow rate of 96.135 lbm/sec.
ATTACHMENT II PAGE 19 0F 69 l 7989Q:10/103184 1 e
~>
TABLE 15.4.1-5 BROKEN LOOP ACCUMULATOR FLOWRATE TO CONTAINMENT FOR LIMITING CASE - DECLG (C D
= 0.4)
TIME (SEC) MASS FLOWRATE* (LBM/SEC) 0.0 4598.0 3.01 3499.1 4.01 3281.7
~
5.01 3100.7 8.01 2690.5 11.01 2397.8 16.01 2052.5 21.01 1811.8 l 24.01 1700.5 24.87 0.0
*Enthalpy of accumulator water is 58 BTU /LBM i
1 ATTACHMENT II PAGE 20 0F 69 79890:10/103184
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