ML17309A582
| ML17309A582 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/19/1995 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9506280692 | |
| Download: ML17309A582 (98) | |
Text
.
PRIORITY 1 r<+, (ACCELERATED RZDS PROCESSING)
REGULATORY INFORMATION DXSTRIBUTION SYSTEM (RXDS)
F ACCESSION NBR: 9506280692 DOC. DATE: 95/06/19 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Roc ester 05000244 P AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp. R RECIP.NAME RECIPIENT AFFILXATION JOHNSON,A.R. Project Directorate I-1 (PD1-1) (Post 941001)
SUBJECT:
Informs that since 921105 submittal of ECCS evaluation including effects of upper plenum injection,ECCS evaluation 'I updated to include larger peaking factors. Rept summarizing 0 ECCS evaluation encl.
DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED:LTR Submittal: General Distribution I ENCL g SIZE:
NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 LA 1 1 PD1-1 PD 1 JOHNSON,A 1 1 XNTERNAL E CE 01 1 l. NRR/DE/EMCB 1 1 NRR DRCH/HICB 1 1 NRR/DS SA/S PLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 D
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AND ROCHESTER GAS AND ElECTRIC CORPORATION ~ 89 EASTAVENUE, ROCHESTER, M Y Id&9-0001 AREA CODE7M 546-2700 ROBERT C. MECREDY Vice President Nucleor Operotions June 19, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555
Subject:
ECCS Evaluation Including the Effects of Upper Plenum Injection R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson:
Reference (a) requested Rochester Gas and Electric Corp. (RG&E) to submit an ECCS evaluation that included the effects of Upper Plenum Injection (UPI). The evaluation was completed and submitted by reference (b). Since that submittal, the ECCS evaluation has been updated using reference (c) methodology to include larger peaking factors. The larger peaking factors are necessary to support conversion to an eighteen month fuel cycle scheduled to begin at the Spring 1996 Startup. This submittal supersedes Reference (b) in its entirety.
The attached report summarizes the ECCS evaluation and will be used to update section 15.6.4.2 of the Ginna UFSAR.
It is requested that this evaluation become the analysis of record with the startup of Cycle 26, currently scheduled for June 5, 1996.
In order to use the methodology of reference (c) exemptions from two specific sections of 10CFR50 Appendix K are necessary because of the hardware configuration of upper plenum injection. Those exemption requests were submitted by reference (d).
Very truly yours,
~~
Robert C. Mecredy g
Attachment RWEtl,380 260039
'st506280692 st50619 PDR ADOCK 05000244 pJ+ ~exp~i~&p
xc: Mr. Allen R. Johnson (Mail Stop 14B2)
Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 US NRC Ginna Senior Resident Inspector Ref. (a): NRC letter from S. A. Zwolinski to R. W. Kober (RG&E),
Subject:
"Development of an Acceptable ECCS Evaluation Model Which Includes the Effects of Upper Plenum Injection, " dated 2/12/85.
(b): RG&E letter from R. C. Mecredy to A. R. Johnson (NRC),
Subject:
ECCS Evaluation Including the Effects of Upper Plenum Injection, dated Nov. 5, 1992.
(c): NRC SER, Thadami (NRC) to W. Johnson (W), Feb. 8. 1991, Westin house Lar e Break LOCA Best Estimate Methodolo Vol. 1: Model Descri tion and Validation: Model Revisions, WCAP-10924-P, Rev. 1, Vol. 1, Addendum 4 (Proprietary Version) Aug. 1990.
(d): RG&E letter from R. C. Mecredy (RG&E) to A. R. Johnson (NRC),
Subject:
Request for Exemption to Selected 10 CFR Part 50, Appendix K Requirements, dated Nov. 5, 1992.
15.6.4 PRIMARY SYSTEM PIPE RUPTURES 15.6.4.2 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS-OF-COOLANT ACCIDENT)
The analysis specified by 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors,"<'> is presented in this section for a major rupture of the reactor coolant system (RCS) pressure boundary for the R. E. Ginna Nuclear Plant.
15.6.4.2.1 I, ssification and riteria A major pipe rupture (large break), as considered in this section, is defined as a breech in the reactor coolant pressure boundary with a total cross-sectional area greater than 1.0 ft'. This is considered a Condition IV event'">. Condition IV occurrences are faults which are not expected to occur during the lifetime of the R. E. Ginna Nuclear Plant, but are postulated because the consequences include the potential for the release of significant amounts of radioactive material.
The Condition IV major pipe rupture loss-of-coolant accident (LOCA) is the most drastic decrease in reactor coolant inventory event which must be designed against and thus represents the limiting design case for the Emergency Core Cooling System (ECCS). In Westinghouse nuclear steam supply system (NSSS) designs, Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values of 10 CFR 100 and a single Condition IV fault is not to cause a consequential loss of required functions of systems needed to cope with the fault, including those of the ECCS and the containment. WASH-1400"> presents the results of a study of the probability of occurrence of various accident sequences including pipe ruptures.
The R. E. Ginna Nuclear Plant reactor is designed to withstand the effects caused by a loss-of-coolant accident including the double ended severance of the largest pipe in the reactor coolant system. The reactor core and internals together with the Safety Injection System are designed so that the reactor can be safely shut-down, the essential heat transfer geometry of the core preserved following the accident, and the long-term coolability maintained. The ECCS is designed to meet Acceptance Criteria which preclude fission product release to the environment in excess of the guideline values of 10 CFR 100.
The Acceptance Criteria for the loss-of-coolant accident, as prescribed in 10 CFR 50.46, are:
- a. The calculated peak fuel element cladding temperature is below the limit of 2200'F.
- b. The cladding temperature transient is 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 after quenching.
- c. The amount of hydrogen generated by fuel element cladding that reacts chemically with water or steam does not exceed an amount corresponding to interaction of 1% of the total amount of Zircaloy in the reactor..
0 d. The core remains amenable to cooling during and after the break.
- e. The core temperature is reduced and decay heat is removed for an extended period of time, as 95062806g~
,F required by the long-lived radioactivity remaining in the core.
These criteria were established to provide significant margin in ECCS performance following a LOCA.
The ECCS is designed to meet Acceptance Criteria even when operating with the most severe single failure. During the injection mode, the loss of a safety injection train is the limiting single failure.
For large break LOCAs, the limiting single failure is one which minimizes the amount of ECCS flow delivered to the core without reducing the capability of the emergency safeguards systems to reduce the containment pressure. A lower containment backpressure reduces the reflooding rate due to the increased difficulty in venting steam from increased steam binding. The lowest containment pressure is obtained only if all containment spray pumps and fan coolers operate subsequent to the postulated LOCA.
Therefore, for the purposes of large break LOCA analyses, the most limiting single failure would be the loss of one high head SI and one low head safety injection pump with full operation of the spray pumps and fan coolers.
The large break LOCA analyses for the R. E. Ginna Nuclear Plant conservatively assumes both maximum containment spray and fan cooler operation (lowest containment pressure) and minimum ECCS safeguards (the loss of one high head SI and one low head SI pump), which results in the minimum delivered ECCS flow available to the RCS.
An additional conservative assumption is that the insertion of control rods to shut down the reactor is neglected in the large break LOCA analysis, although in reality some control rod insertion may occur.
15.6.4.2.3 e ienc of Event. and stem. erations 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 (high containment pressure or low pressurizer pressure) is reached. Continued RCS depressurization results in accumulator injection to the intact loop cold leg. These countermeasures will limit the consequences of the accident in two ways:
A. Although credit is not taken for control rod insertion in the LOCA analysis, in reality 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.
B. Injection of borated water provides for heat transfer from the core and prevents excessive clad temperatures.
I For the R. E. Ginna Nuclear Plant large break LOCA ECCS analysis using the WCOBRA/TRAC UPI methodology, one low head safety injection pump starts and delivers flow to the upper plenum.
Additionally, two high head safety injection pumps start with one assumed to inject directly to the intact loop and one spilling on the broken loop. The high head safety injection flows are modeled so that delivery to the RCS is minimized while maximizing spill to the containment. One low head and one high head safety injection pump are assumed to fail due to a bus failure, representing the worst single failure assumption for the analysis. In addition, both low head and high head safety injection pump performance curves are assumed to be degraded. Assuming power to both emergency system trains is consistent with modeling full operation of the active containment heat removal systems. Modeling the operation of all the containment heat removal systems is consistent with the US-NRC Branch Technical Position CSB 6-1P> and is conservative for the large break LOCA.
Following a reactor trip, offsite power may degrade to the point where vital loads are switched to emergency power. This would cause an interruption in safety injection flow. Interruption of the safety injection flow from the time of the reactor trip signal plus 40 seconds (minimum) and plus 150 seconds (maximum) is considered. From the time of interruption, high head safety injection is lost for 1.5 seconds then restored to full capacity and the low head safety injection flow is lost for 17 seconds then restored to full capacity. These times are associated with restarting the safety injection pumps on emergency power.
The time sequence of events following a large break LOCA is presented in Figure 15.6.4.2-1 and Table 15.6.4.2-3. The safety injection performance during the transient, as predicted in the R. E. Ginna Nuclear Plant large break LOCA WCOBRA/TRAC UPI analysis, are presented in Figures 15.6.4.2-8A and 15.6.4.2-9A, 15.6.4.2-8B and 15.6.4.2-9B, and 15.6.4.2-8C and 15.6.4.2-9C for the Appendix K calculations and in Figures 15.6.4.2-8D and 15.6.4.2-9D for the Superbounded calculation.
Figures 15.6.4.2-14 and 15.6.4.2-15 illustrate the safety injection pump performance modelled in the WCOBRA/TRAC input.
15.6.4.2.4 meri tion of I ar e Break A Tran. ient Before the break occurs, the R. E. Ginna Nuclear Plant is assumed to be in a full power equilibrium condition; i.e., the heat generated in the core is being removed through the steam generator secondary system. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. During blowdown, heat from fission product decay, hot internals and the vessel, continues to be transferred to the reactor coolant. After the break develops, the time to departure from nucleate boiling is calculated.
Thereafter, the core heat transfer is unstable, with both nucleate boiling and film boiling occurring. As the core becomes voided, both transition boiling and forced convection are considered as the dominant core heat transfer mechanisms. Heat transfer due to radiation is also considered.
The heat transfer between the RCS and the secondary system may be in either direction, depending on the relative temperatures. In the case of the large break LOCA, the primary pressure rapidly decreases below the secondary system pressure and the steam generators are an additional heat source. In the R.
E. Ginna Nuclear Plant large break LOCA analysis using the WCOBRA/TRAC UPI methodology, the secondary system is conservatively assumed to be isolated at the initiation of the event to maximize the secondary side heat load.
When the RCS depressurizes to approximately 715 psia, the accumulators begin to inject borated water into the reactor coolant loops. Borated water from the accumulator in the faulted loop is assumed to spill to containment and be unavailable for core cooling for breaks in the cold leg of the RCS. Flow from the accumulator in the intact loop may not reach the core during depressurization of the RCS due to the fluid dynamics present during the ECCS bypass period. ECCS bypass results from the momentum of the fluid flow up the downcomer due to a break in the cold leg which entrains ECCS flow out toward the break.
Bypass of the Safety Injection diminishes as mechanisms responsible for the bypassing are calculated to be no longer effective.
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 atmosphere. After the end of the blowdown, refill of the reactor vessel lower plenum begins. Refill is 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, BOC, recovery time).
The reflood phase of the transient is defined as the time period lasting from BOC recovery 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 on into the beginning of reflood, the intact loop safety injection accumulator tank rapidly discharges borated cooling water into the RCS. Although the portion injected prior to end of bypass is lost out the cold leg break, the accumulator eventually contributes to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for the reflooding of the reactor core. The safety injection pumps aid in the filling of the downcomer and core and subsequently supply water to help maintain a full downcomer and complete the reflooding process. The UPI also aids the reflooding process by providing water to the core through the vessel upper plenum.
The end of the refill phase and the beginning of the reflood phase, i.e., BOC time, is not as significant an event or as easily defined for the Two-Loop UPI Large Break LOCA WCOBRA/TRAC Evaluation Model when compared to previous Westinghouse Large Break Evaluation Models. The typical practice for WCOBRA/TRAC analyses, is to report the time when low void fraction mixture is seen at the bottom of the core for BOC time. However, since a significant portion of the upper plenum safety injection water can flow down the low power/periphery channel of the core, significant cooling of the hot rod can occur prior to this time due to transverse flows within the core. In some cases, this cooling can be sufficient to cause the peak cladding temperature to occur prior to BOC time.
Following safety injection interruption and full capacity being restored, continued operation of the ECCS pumps supplies water during long-term cooling. Core temperatures have been reduced to long-term 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 from the injection mode to the sump recirculation mode of ECCS operation. Spilled borated water is drawn from the containment sump by the low head SI pumps and returned to the upper plenum.
Long-term cooling includes long-term criticality control. Criticality control is achieved by maintaining subcriticality without credit for RCCA insertion due to the boron in the ECCS and sump. The necessary RWST and accumulator boron concentrations are a function of each core design and are verified each cycle.
15.6.4.2.5 Anal si. of Effects and Conse uences 15.6.4.2.5.1 Method of Analysis The analysis was performed using the Westinghouse Large Break LOCA WCOBRA/TRAC Best-Estimate Methodology for plants which incorporate Upper Plenum Injection (UPI) in the Safety Injection System design.'4'i The Westinghouse Best-Estimate Methodology was developed consistent with guidelines set forth in the SECY-83-472 document.<@ These guidelines provide for the use of realistic models and assumptions in the calculational framework. The technical basis for the use of this model is discussed in detail in References 4 and 5.
The SECY-83<72 document states that there are three areas of conservatism in the current licensing models: the required Appendix K conservatism, the conservatism added by both the NRC staff and industry to cover uncertainties, and the conservatism imposed by the industry in some cases to reduce the complexity of the analysis. Based on a review of the available experimental data and the best estimate computer code calculations, the NRC staff concluded that there is more than sufficient safety margin to assure adequate performance of the ECCS, and that this excess margin can be reduced without an adverse effect on plant safety. Therefore, in the SECY-83C72 approach, the NRC staff suggested that the
licensee could utilize a realistic model of the PWR to calculate the plant response to a LOCA at a conservative 95 percent probability level, i.e., the Superbounded Analysis. The calculation at the 95 percent probability level would account for uncertainties in such things as power level, fuel initial temperature, nuclear parameters, and computer code uncertainties. The parameters which imply uncertainty, and the methods by which the uncertainties would be combined (either statistically or as a one-sided bias) would have to be justified. The uncertainty analyses can be performed on a generic, realistic PWR model which is representative of a class of similar plants, that is, two-, three-, or four-loop PWRs so that generic uncertainties are applicable to the individual plants.
The WCOBRA/TRAC code uncertainty methodology calculation consists of two parts;
- 1) An assessment of the ability of WCOBRA/TRAC to model the PWR behavioit'i, and
- 2) A quantified assessment of WCOBRA/TRAC capability to predict the measured quantities from various separate effects and systems effects experiments which cover the range of PWR accident conditionsi'>.
The sources of uncertainty within the code, and the specific application of the code to the PWR calculation has been addressed in accordance with requirements of SECY-83-472ii. While performing this assessment it was determined that the uncertainty of several modeling effects could not be quantified by comparison to experimental data. Consequently, parametric sensitivity studies were performed which varied these modeling effects in the WCOBRA/TRAC computer code, and the uncertainty was determined based on the results of these sensitivity studies.
~ ~
The numerical value for the code uncertainty was derived by comparing WCOBRA/TRAC to a wide
~
range of experiments which covered the expected range of conditions for the PWR. The uncertainty
~ ~
analysis considered the following items:
t) Code bias - obtained by comparing the code calculated temperatures to the ~avera e of temperatures measured from various single effects and integral tests.
- 2) The uncertainty in the code bias - the standard deviation of the code bias is calculated as 6,.
- 3) The uncertainty in the data for each of the experiments. The individual test data uncertainties are sample size weighted and pooled together to obtain a data uncertainty for all the experiments analyzed as b,.
- 4) The initial test condition uncertainty used in the WCOBRA/TRAC code was assessed by examination of repeat experiments and is calculated as 8,.
- 5) The test modeling uncertainty was assessed by performing noding sensitivity analyses on different tests and averaging the differences between the different cases, and is calculated as The uncertainty analysis was undertaken for both a blowdown and reflood peak temperature. The code bias was a direct value added or subtracted from the calculated plant peak cladding temperature. The uncertainties from items 2 to 5 were statistically combined as the square-root-sum-of-squares and raised to the 95th percentile by multiplying by 1.645. The equation for the plant peak cladding temperature at the 95th percentile becomes:
PCT>>,= PCT,~~ g Code Bias + 1.645 6', + 6', + 63 + 64
~
The nominal calculation is performed to provide assurance that the most probable PCT is well below the
~
estimate of the 95 percent probability value. However, the nominal calculation is itself a conservative estimate since several conservative assumptions are retained.
To demonstrate compliance with the specific requirements of Appendix K to 10 CFR 50, a calculation is performed in which the plant-specific realistic best estimate calculation includes the required Appendix K features, such as 1971 ANS decay heat plus 20 percent, Moody break flow model, no return to nucleate boiling during blowdown, etc. The realistic calculation with the Appendix K required features could be used to demonstrate compliance with the Acceptance Criteria of 10 CFR 50.46, provided that the peak cladding temperature exceeded the peak cladding temperature calculated at the 95 percent probability level but was below the Acceptance Criteria limit of 2200'F.
15.6.4.2.5.2 ECCS Evaluation Model The Best Estimate UPI ECCS Evaluation Model is comprised of the WCOBRA/TRAC and COCO computer codes" i. The WCOBRA/TRAC code is used to generate the complete transient (blowdown through reflood) system hydraulics as well as the cladding thermal analysis.
+COBRA/TRAC is the Westinghouse version of the COBRA/TRACi'> code originally developed by Battelle Northwest Laboratory in the late 1970's. It is an advanced computer code used to simulate complex two-phase transient and steady-state phenomena in nuclear reactors or other large complex heat exchange equipment. WCOBRA/TRAC is a combination of two codes:
a) COBRA-TF, a 3-D, two-fluid, three-field model, capable of calculating complex flow fields in a wide variety of geometries.
b) TRAC-PD2, a 1-D, two-phase drift flux flow model used primarily to simulate piping systems.
The COBRA-TF computer code provides a transient or steady-state two-fluid, three-field representation of two-phase flow. Each field is treated in three dimensions and is compressible. Continuous vapor, continuous liquid and entrained liquid drops are the three fields. The conservation equations for each of the three fields and for heat transfer from the solid structures in contact with the fluid are solved using a semi-implicit, finite-difference numerical technique on an Eulerian mesh. The COBRA-TF vessel 1
model features extremely flexible noding for both the hydrodynamic mesh and the heat transfer solution.
The flexible noding allows representation of single rod bundle subchannel, or grouping of rod bundle subchannels into larger hydrodynamic channels.
Multiphase flows consisting of two or more fluids are separated by moving phase interfaces. In general, the phases can be present in any combination of liquid, solid, or gas. The flow pattern can assume any one of a wide variety of forms, such as bubbly flow, droplet flow, gas-particle flow, and stratified flow.
Since the quantities of interest are the average behavior of each phase within the control volume, most work in multiphase flow is done using average equations across the control volume.
The average conservation equations used in COBRA-TF are derived following the methods of Ishii n" The average used is a simple Eulerian time average over a time interval, ~t, assumed to be long enough to smooth out the random fluctuations present in a multiphase flow but short enough to preserve any gross unsteadiness in the flow. The resulting average equations can be formulated in either the mixture form or the two-fluid form. Due to its greater physical appeal and broader range of application, and the possibility of reduced uncertainty, the two-fluid approach is used as the foundation for COBRA/TF.
The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. ~ The effects of one phase on another are accounted for by interfacial friction, heat and mass transfer interaction terms appearing in the equations. The conservation equations have the same form for
~
each phase; only the constitutive relations and physical properties differ.
The three-field formulation used in COBRA-TF is a straight-forward extension of the two-fluid model.
The fields included are vapor, continuous liquid, and entrained liquid. Dividing the liquid phase into two fields is the most convenient and physically reasonable way of handling flows. For this representation of the liquid phase, the liquid can appear in both film and droplet form. This permits more accurate modeling of thermal-hydraulic phenomena such as entrainment, de-entrainment, fallback, liquid pooling and flooding.
One of the important features of the COBRA-TF vessel model is that the governing equations form a complete set. No terms are omitted particularly in the momentum equations where wall shear, momentum exchange due to turbulence and all the interfacial terms are represented. The COBRA-TF vessel model also has two energy equations to account for thermodynamic non-equilibrium between the two phases. This is particularly important for post CHF (dryout) conditions where the vapor phase can be superheated and the liquid phase remains at the saturation temperature.
A complete set of heat transfer and fiow regime models is incorporated into COBRA-TF. These models
are applicable over a wide range of fluid and heat transfer conditions, as required by the range of conditions found during light water reactor transients. The flow regime model covers the full range from low-void fraction, bubbly regimes to highly dispersed droplet regimes and corresponding heat transfer models exist for these flow regimes, for wall surface temperatures ranging from the fluid saturation temperature to approximately 3000'F.
I WCOBRA/TRAC has been successfully utilized to analyze Westinghouse two-loop PWRs with Upper Plenum Injection<i. The results of these calculations indicate that the WCOBRA/TRAC analysis method verified the safety performance of the upper plenum injection system for this class of plants. This successfully resolved a long standing US-NRC concern on the adequacy of this injection system design.
The system hydraulic transient is influenced by the containment pressure transient response to the mass and energy released from the reactor coolant system by the LOCA. In the Best Estimate UPI ECCS Evaluation Model, the containment pressure transient is provided as a boundary condition to the system hydraulic transient.. The containment pressure transient applied is to be conservatively low and include the effect of the operation of all pressure reducing systems and processes. The COCO computer codd" is used to generate the containment pressure response to the mass and energy release f'rom the break.
This containment pressure curve is then used as an input to the WCOBRA/TRAC code. It should be noted that safety injection actuation is based on the pressurizer low pressure SI signal, and not on containment pressure high pressure SI signal. Although the latter is computed to occur earlier, it is conservative to model a later time for SI injection. Additionally, since the WCOBRA/TRAC and COCO computer codes do not run interactively, it would be difficult to model SI actuation on high containment pressure.
15.6.4.2.5.3 Plant Input Parameters and Initial Conditions Important input parameters and initial conditions used in the analysis are listed in Tables 15.6.4.2-2A and 15.6.4.2-2B for the Appendix K and Superbounded cases respectively. The initial steady state fuel pellet temperature and fuel rod internal pressure used in the LOCA analysis was generated with the PAD 3.4 Fuel Rod Design Codei'" which has been approved by the US-NRC. The fuel parameters input to the code were at beginningwf-life (maximum densification) values for the hot assembly and hot rod, and the remainder of the assemblies were modeled at a conservatively low value representing average core burnup.
In determining the conservative direction for bounding values and assumptions for UPI plants, many sensitivity studies were performed'". These sensitivities were performed using a representative two-loop plant with Upper plenum Injection (UPI) in the ECCS design. Since the representative two-loop plant has a higher peak linear heat rate and a higher core power to pumped ECCS flow ratio than the R. E.
Ginna Nuclear Plant it will yield a greater change in peak cladding temperature for changes in plant 1
parameters.. These sensitivity studies were used to determine the direction of conservatism for choosing the bounding conditions for the 95th percentile calculation for the R. E. Ginna Nuclear Plant.
The parameters used in the COCO analysis to determine the containment pressure curve are presented in Table 15.6.4.2-5. The containment pressure transient used to calculate the system hydraulic transient is shown in Figure 15.6.4.2-2 for the Appendix K and Superbounded calculations.
Initial conditions for the R. E. Ginna Nuclear Plant large break LOCA analysis are delineated in Table 15.6.4.2-1. Most of these parameters were chosen at their limiting values in order to provide a conservative bound for evaluation of the calculated peak cladding temperature for the large break LOCA analysis. Note that this analysis incorporates the effect of accumulator water temperature presented in Reference 12 and a maximum expected value for accumulator water temperature as defined in Reference 13 has been used. This analysis incorporates BWI r'eplacement steam generators (SGs) and a T window of 15%. The replacement SGs results in a PCT benefit which compensates for the effect of accumulator water temperature and T,~ window. The hot assembly was located under a source plate location in the upper core plate. Past sensitivity studies showed the limiting location for peak cladding temperature to be an open hole' and the source plate location for the R. E. Ginna Nuclear Plant possesses the limiting flow area among the various upper core plate geometries.
15.6.4.2.5.4 Description of Appendix K Calculations Three Appendix K calculations were performed using the WCOBRA/TRAC Addendum 4 version of the model<. The cases consider minimum and maximum RCS average temperatures of 559'F and 573.5'F with minimum and maximum safety injection interruptions time of 40 seconds and 150 seconds (from time of reactor trip). The 3 cases are summarized below. Since PCTs generally occur prior to the maximum SI interruption time, only one case with the maximum interruption time will be performed to i
verify that it has no effect on the PCT.
- Appendix K Minimum Tavg with Minimum SI Interruption Time (Case A)
L 5
f>>"
- Appendix K Minimum Tavg with Maximum SI Interruption Time (Case B)
- Appendix K Maximum Tavg with Minimum SI Interruption Time (Case C) 15.6.4.2.5.4.1 Appendix K Minimum Tavg with Minimum SI Interruption Time Calculations A "Base Case" Appendix K calculation (Case A) was performed for the R. E. Ginna Nuclear Plant which conforms to the modeling requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50. The conservative assumptions used in the Appendix K Case A calculation are listed in Table 15.6.4.2-2A.
Shortly after the break is assumed to open, the vessel rapidly depressurizes (Figure 15.6.4.2-4A) and the core flow quickly reverses. During the flow reversal, the hot assembly fuel rods dry out and begin to heat up momentarily (Figure 15.6.4.2-3A).
At approximately 9 seconds into the transient, maximum downflow is reached in the high and low power regions of the core. Figure 15.6.4.2-6A shows the liquid, vapor and entrained liquid flow rates respectively for the average power interior assemblies. Figure 15.6.4.2-11A shows the same three quantities for the guide tube assemblies. Similarly, Figure 15.6.4.2-12A and Figure 15.6.4.2-13A show the same three quantities for the low power/pe riphery assemblies and for the hot assembly respectively. This flow is sufficient to cool the hot assembly and maintain the rest of the core near the fluid saturation temperature (Figure 15.6 4.2-3A).
As the vessel continues to depressurize, liquid inventory continues to be depleted, and core void fractions increase (Figure 15.6.4.2-7A). This results in reduced core flow and resulting cladding heatup, first for the hot assembly, and later for the other regions of the core.
At approximately 6 seconds into the transient, the accumulator begins to inject water into the intact cold leg (Figure 15.6.4.2-10A). This water fills the cold leg and upper downcomer region, and is bypassed to the break initially. At approximately 19 seconds, accumulator water begins to flow into the lower plenum.
At approximately 20 seconds, pumped injection into the cold leg and into the upper plenum begins (Figures 15.6.4.2-8A and -9A). This water begins to flow through the low power peripheral region of the core, and contributes to some core cooling, but primarily flows through the core into the lower plenum. Figures 15.6.4.2-8A and -9A also show the loss of safety injection flow at 44 seconds. High head and low head safety injection is restored to full capacity at 45.5 and 61 seconds respectively.
I' I 'I
% v)~
At approximately 26 seconds, the lower plenum has filled to the point that water begins to reflood the core from below. The void fraction in the upper plenum begins to decrease (Figure 15.6.4.2-5A), as well as the core void fraction (Figure 15.6.4.2-7A). At this time core cooling increases substantially and the peak cladding temperature begins to decrease.
Figure 15.6.4.2-5A shows the void fraction in the upper plenum interior global channel, below the hot leg elevation. The upper plenum interior global channel is that region above the upper core plate which has no flow path from the core (i.e., above the interior solid metal portions of the upper core plate).
The peak cladding temperature calculated for the R. E. Ginna Nuclear Plant Appendix K large break LOCA analysis is 2050.5'F, assuming a peak hot rod power of 14.34 kw/ft. This result is below the acceptance criteria limit of 2200'F. The maximum local metal-water reaction is well below the embrittlement Acceptance Criteria limitof 17 percent. The limiting total core metal-water reaction is also less than 1.0 percent, in accordance with the Acceptance Criteria. The clad temperature transient is 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.
15.6.4.2.5.4.2 Appendix K Minimum Tavg with Maximum SI Interruption Time Calculations A variation of the Case A Appendix K calculation shown in Section 15.6.4.2.5.1 was performed assuming the maximum delay time for safety injection interruption (time of reactor trip signal plus 150 seconds) for the R. E. Ginna Nuclear Plant. The transient is very similar and does not vary from the Case A until after the time core cooling has increased substantially and the peak cladding temperature begins to decrease. Figures 15.6.4.2-8B and 15.6.4.2-9B show the loss of safety injection flow at 155 seconds.
High head and low head safety injection is restored to full capacity at 156.5 and 172 seconds respectively.
Figures 15.6.4.2-3B through 15.6.4.2-13B describe the transient as discussed for the Base Case in Section 15.6.4.2.5.4.1.
The change in SI delay time has no effect on the calculated peak cladding temperature since the delay occurs after the peak cladding temperature is calculated to occur. The peak cladding temperature calculated for the R. E. Ginna Nuclear Plant Appendix K large break LOCA analysis is 2050.5'F, assuming a.peak hot rod power of 14.34 kw/ft. This result is below the acceptance criteria limit of
2200'F. The maximum local metal-water reaction is well below the embrittlement Acceptance Criteria limit of 17 percent: The limiting total core metal-water reaction is also less than 1.0 percent, in accordance with the Acceptance Criteria. The clad temperature transient is 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.
15.6.4.2.5.4.3 Appendix K Maximum Tavg with Minimum SI Interruption Time Calculations A variation of the Appendix K calculation was performed assuming the maximum average RCS temperature of 573.5 'F for the R. E. Ginna Nuclear Plant. The transient is very similar to Case A which assumes a minimum RCS average temperature. Figures 15.6.4.2-3C through 15.6.4.2-13C describe the transient as discussed for the Base Case in Section.15.6.4.2.5.4.1.
The peak cladding temperature calculated for the R. E. Ginna Nuclear Plant Appendix K large break LOCA analysis is 2006.1'F, assuming a peak hot rod power of 14.34 kw/ft. This result is below the acceptance criteria limit of 2200'F. The maximum local metal-water reaction is well below the embrittlement Acceptance Criteria limitof 17 percent. The limiting total core metal-water reaction is also less than 1.0 percent, in accordance with the Acceptance Criteria. The clad temperature transient is 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.
15.6 4.2.5.5 Description of Superbounded Calculation A calculation was performed for the R. E. Ginna Nuclear Plant which combined all the parameters at their conservative values. This calculation has been shown in previous studies to conservatively estimate the 95 percent probability PCT, and is called the Superbounded calculation. The conservative assumptions used in the Superbounded calculation are listed in Table 15.6.4.2-2B.
The figures described for the Appendix K calculation in Section 15.6.4.2.5.4.1 are provided for the Superbounded calculation.
The peak cladding temperature calculated for the R. E. Ginna Nuclear Plant bounded large break LOCA analysis, with uncertainty, is 1951.9'F, assuming a peak hot rod power of 14.04 kw/ft. This result is
below the acceptance criteria limit of 2200'F. The maximum local metal-water reaction is well below the embrittlement Acceptance Criteria limit of 17 percent. The limiting total core metal-water reaction is also less than 1.0 percent, in accordance with the Acceptance Criteria. The clad temperature transient is 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.
15.6.4.2.6 conclusions For breaks up to and including the double ended severance of a reactor coolant pipe, the emergency core cooling system will meet the acceptance criteria. These criteria are as follows:
- 1. The calculated peak fuel element cladding temperature is below the requirements 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 cladding 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 after 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.
In keeping with the SECY-83P72 approach, large break LOCA analyses were performed for the 95 percent probability level (known as a "Superbounded" calculation) and the Appendix K calculation.
Table 15.6.4.2-4 has a summary of the results for both the Appendix K cases and the Superbounded case.
The Appendix K calculations had a peak cladding temperature of 2050.5'F. The bounded calculation resulted in a peak cladding temperature of 1776.9'F. With a total bias and uncertainty value of 175'FPi, the 95th probability value is 1776.9'F + 183'F = 1959.9'F. These results clearly meet the Acceptance Criteria specified in 10 CFR 50.46.
REFERENCES
- 1. Acce tance ri eria for Emer enc ore lin s ems for Li ht Water poled Nucl r Power Reac rs 10 FRS 46 and A endix K f 1 FR 0.46, Federal Register, Vol. 39, No. 3, January 4, 1974.
1a. ANS-51.1/N18.2-1973, Nucl r fe ri ri fr he Desi n of S ati nar Press>rized W er Reac r Plan
- 2. U. S. Nuclear Regulatory Commission 1975, Re c r Safet Stud - An Ass <<men f Accident Ri k in A mmerci l N clear P w r Plant, WASH-1400, NUREG-75/014.
- 3. Branch Technical Position CSB 6-1, Minimum ontainmen Pressure Model F r PWR E S Perf rmance Ev lu i n, November 24, 1975.
r -Br k L A B -E im M h d I, Volume 1: Model Description and Validation, WCAP-10924-P-A, Vol. 1, Rev. 1, and Addenda, (Proprietary Version), December 1988.
- 5. Dederer, S. L, Hochreiter, L. E., Schwarz, W. R., Stucker, D. L., Tsai, C. K., an'd Young, M.
Y., Wes in h use Lar e-Br k L A B s-E. imate Methodol V lume 2' licati n o Two- PWRs E ui wi h er Plenum In'ection, WCAP-10924-P-A, Vol. 2, Rev. 2, and Addenda, December 1988.
- 6. NRC Staff Report, Emer enc r lin em Anal sis Me hods, USNRC-SECY-83-472, November 1983.
- 7. NRC SER, Thadani (NRC) to W. Johnson ~, February 8, 1991, W ~ in h i.e Lar e Break L A B ~t Estim e Meh l V 1>me 1' del Descri ion and Valida i n'odel
~Revi i n ", WCAP-10924-P Revision i, Volume i, Addendum 4 (Proprietary Version), August, 1990.
- 8. Bordelon, F. M., and Murphy, E. T.,~ ~ ~ n inmen Pressure Anal sis ode , WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June 1974. ~
REFERENC continued 9 ...,,Gi Therm l-H drauli de f r Tr nsien i..,,
Anal sis
...C of Nucl r Reac or Vessels and Prima oolan s ems'E uati nsan ns i iv M dels, NUREG/CR-3046, PNL-4385Vol. 1R4, March 1983.
- 10. Ishii, M., Thermo-Field D n mic Th r f Tw -Pha.e Flow, Eyrolles, 1975.
- 11. Weiner, R. A., etal., Im rov F el Perf rm nce Models for W s in h eF el R d Desi n nd C -999-- G 9 i ).* <<9
- 12. Letter ET-NRC-92-3699, Liparulo, N. J. g to NRC, "Results of Technical Evaluation of Containment Initial Temperature Assumptions for LBLOCA Analysis," June 1, 1992.
- 13. Letter 95RGE*-G-0017, Attachment A, from Hoskins, K. C. to Eliasz, R. W., March 8. 1995.
TABLE 15.6.4.2-1 (1 of 2)
INITIALCONDITIONS FOR THE R. E. GINNA NUCLEAR PLANT LARGE BREAK LOCA ANALYSIS APPENDIX K AND SUPERBOUNDED PARAMETER NALV I DESIRED VALUE Plant Internals Flat Upper Support Plate Barrel Baffle Design Downflow Core Bypass Flow 6.5%
NSSS Power, 102% of (MWT) 1520.
System Pressure (psia) 2280.
Primary System Fluid Temperatures Tavg = 559'F Tter ( F) 590.58 T ('F) 527.42 Turrsa>mn ( F) 585.0 e Tavg = 573.5'F Tier ( F) 604.56 T ('F) 542.44 Turrea>m,n ( F) 600.1 Fuel Type 14 x 14 OFA (Optimized Fuel Assembly)
Fuel Stored Energy Beginning of Life Fuel Data Source- Pad 3 4"u Fuel Rod Backfill Pressure (psig) 275.
FqT 2.45 FaH 1.75 Peak Linear Power, kw/ft 14.34 Relative Power in the Outer Core Channel 0.6 Loop (Thermal Design) Flow Rate (gpm) 85,100 Reactor Coolant Pumps Running TABLE 15.6.4.2-1 (2 of 2)
INITIALCONDITIONS FOR THE R. E. GINNA NUCLEAR PLANT LARGE BREAK LOCA ANALYSIS APPENDIX K AND SUPERBOUNDED PARAMETER ANALY IS DESIRED VAL E Steam Generator Tube Plugging (Symmetric) 15%
Steam Generator Secondary Pressure (psia)
Tavg = 559'F 707.77 Tavg = 573.5 F 811.63 Accumulators In Operation 2 (one injects into the Intact loop, one spills to containment)
Accumulator Conditions per Accumulator:
Water Volume (fP) 1,100.
Nitrogen Pressure (psia) 715.
Water Temperature ('F) 105.
Safety Injection Conditions-Pumps in Operation: 1 LHSI Into Upper Plenum and 2 HHSI Into Two Cold Legs Pump Flow Degraded Water Temperature ('F) 60.0 Delay Time (seconds) 12.0 (High Head Safety Injection) 19.0 (Upper Plenum Injection)
Containment Pressure FIGURE 15.6.4.2-2 TABLE 15.6.4.2-2A ASSUMPTIONS USED IN THE APPENDIX K CALCULATIONS
- 1. PLANT CONFIGURATION
- a. Pressurizer in Intact Loop
- b. Total Peaking Factor (F~T) at 2.45 C. Nuclear Enthalpy Rise Peaking Factor (F,) at 1.75
- d. 102% of 1520 MWt (1550.4 MWt) 15% steam generator tube plugging level Thermal design minimum loop flow rate Beginning of cycle fuel temperature
- h. Beginning of cycle fuel pressure Conservative power distribution
- 2. SAFETY INJECTION CONFIGURATION
- a. Worst single failure
- b. HHSI spilling to containment pressure
- c. Maximum safety injection delay time
- 3. MODEL ASSUMPTIONS
- a. Accumulator nitrogen modeled
- b. Conservative reactor coolant pump two-phase model C. Cross fiow de-entrainment
- d. No locked pump rotor during reflood
- e. Limiting break discharge coefficient (0.4)
Lower bound containment pressure g ANS 1971 Decay Heat Plus 20%
- h. Baker-Just Metal Water reaction Swelling and blockage model I
Can not return to nucleate boiling during blowdown
- k. Clad burst ECCS bypass TABLE 15.6.4.2-2B ASSUMPTIONS USED IN THE SUPERBOUNDED CALCULATION
- 1. PLANT CONFIGURATION
- a. Pressurizer in Intact Loop
- b. 95/95 Peaking Factor (F~T) at 2.40 (equivalent Appendix K F~ = 2.64)
Nuclear Enthalpy Rise Peaking Factor (F,) at 1.75
- d. 102% of 1520 MWt 15% steam generator tube plugging level Thermal design minimum loop flow rate g Beginning of cycle fuel temperature
- h. Beginning of cycle fuel pressure Conservative power distribution
- 2. SAFETY IN1ECTION CONFIGURATION
- a. Worst single failure
- b. HHSI spilling to containment pressure
- c. Maximum safety injection delay time
- 3. MODEL ASSUMPTIONS No accumulator nitrogen modeled
- b. Conservative reactor coolant pump two-phase model C. No cross flow de-entrainment Locked pump rotor during reflood Limiting break discharge coefficient (0.6)
Lower bound containment pressure g Decay heat at 95/95 upper bound for hot rod
- h. Metal water reaction at 95/95 upper bound for hot rod TABLE 15.6.4.2-3 LARGE BREAK LOCA ANALYSIS TIME SEQUENCE OF EVENTS FOR DECLG BREAK (All Results Are From The WCOBRA/TRAC Computer Code)
APPENDIX K SUPER Case A Case B Case C BOUNDED
~Tim sec T~ime sec T~ime sec ~Time eec Start 0.0 0.0 0.0 0.0 Reactor Trip Signal -4.0 -4.0 -4.0 -4.0 Accumulator Injection Begins 6.0 6.0 6.0 6.0 S.I. Signal 44 4.4 44 44 Blowdown Peak Cladding Temperature Occurs 8.0 8.0 7.0 7.0 End of ECCS Bypass 14.6 14.6 14.4 High Head Safety Injection Begins 16.4 16.4 16.4 16.4 End of Blowdown 19.8 19.8 19.8 15.4 Hot Rod Burst Time 19.9 19.9 19.1 Hot Assembly Burst Time 23.5 23.5 23.6 Low Head (RHR) Safety Injection Begins 23.4 23.4 23.4 23.4 Bottom of Core Recovery 26.0 26.0 25.3 Reflood Peak Cladding Temperature Occurs 36.0 36.0 37.0 37.9 Safety Injection Flow Cut-off 44.0 155.0 44.0 44.0 High Head Safety Injection Flow Restored 45.5 156.5 45.5 45.5 Accumulator Water Empty 52.3 52.3 52.4 50.3 Low Head (RHR) Safety Injection Flow Restored 61.0 172.0 61.0 61.0
TABLE 15.6.4.2-4 LARGE BREAK LOCA ANALYSIS RESULTS APPENDIX K SUPER-RI<~L~T Cnse A Cnse 8 ~a~e C ~BNDED Calculated Blowdown Peak Cladding Temp., 'F 1738.7 1738.7 1711.6 1623.9 Calculated Reflood Peak Cladding Temp., 'F 2050.5 2050.5 2006.1 1776.9 Peak Cladding Temp. (P ) 95%), 'F 2050.5 2050.5 2006.1 1951.9'.63 Peak Cladding Temp. Location, Ft. 7.63 7.63 7.25 I
Hot Rod Burst Time (sec.) 19.9 19.9 19.1 Burst Location (ft.) 7.63 7.63 7.63 Hot Assembly Burst Time (sec.) 23.5 23.5 23.6 Burst Location (ft.) 7.63 7.63 7.63
% Blockage 46.5 46.5 43.7 Local Zr/H,O Reaction (max), % < 17.0 Local Zr/H,O Location, Ft. 7.63 7.63 7.63 Total Zr/H,O Reaction, %
- Peak cladding temperature at the 95th percentile probability level with 95 percent confidence is obtained by adding the calculated peak cladding temperature to the code bias plus uncertainties as discussed in Section 15.6.4.2. The sum of the code bias plus uncertainties has been determined to be 183'F as specified in Reference 7.
TABLE 15.6.4.2-5 LARGE BREAK CONTAINMENTDATA (DRY CONTAINMENT)
NET FREE VOLUME 1.066 x 10't'NITIAL CONDITIONS Pressure 14.5 psia Temperature 90'F RWST Temperature 60'F Service Water Temperature 35'F Outside Temperature -10'F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate 1,800 gpm each Actuation Time 2 seconds'AFEGUARDS FAN COOLERS Number of Fan Coolers Operating Fastest Post Accident Initiation of Fan Coolers 0 seconds
'lthough a later spray initiation time can be justified, modelling an earlier spray start time is conservative for large break LOCA analyses.
TABLE 15.6.4.2-5 (Continued)
STRUCTURAL HEAT SINK DATA Descriptive Surface Thickness Material Area,(ft')
Insulated Portion of dome 1.25 in. Insulation 36285 and containment wall 0.375 in. Steel 2.5 ft. Concrete Uninsulated Portion of dome 0.375 in. Steel 12370 2.5 ft. Concrete Basement floor 2.0 ft. Concrete 7230 0.25 in. Steel 2.0 ft. Concrete Walls of sump in basement floor 2.0 ft. Concrete 2480 0.25 in. Steel 3.0 ft. Concrete Floor of sump 2.0 ft. Concrete 400 0.25 in. Steel 1.0 ft. Concrete Inside of refueling cavity 0.25 in. Steel 6170 2.5 ft. Concrete Bottom of refueling cavity 0.25 in. Steel 1260 2.5 ft. Concrete Area on outside of refueling 2.5 ft. Concrete 6750 cavity walls Area inside of loop and steam 4.0 ft. Concrete 10370 generator compartment Intermediate level floor area 6.0 ft. Concrete 5320 Operating floor 2.0 ft. Concrete 6500 1.48 in. thick I-beam 1.48 in. Steel 2000 0.94 in. thick I-beam 0.94 in. Steel 630 TABLE 15.6.4.2-5 (Continued)
STRUCTURAL HEAT SINK DATA (continued)
Descriptive Surface Thickness Material Area (ft')
0.52 in. thick I-beam 0.52 in. Steel 4220 0.61 in. thick I-beam 0.61 in. Steel 1190 Cylindrical Supports for steam generator and reactor coolant pumps
'.5 in. Steel 470 Plant crane support columns 0.75 in. Steel 4810 Beams used for crane structure 1.5 in. Steel 3390 Structure on operating floor 2.0 ft. Concrete 2060 Grating, stairs, miscellaneous steels 0.125 in. Steel 7000 FIGURE 15.6.4.2-1 LARGE BREAK LOCA SEQUENCE OF EVENTS BREAK OCCURS B PUMPED Sl SIGNAL L REACTOR TRIP (Sl SIGNAL) 0 ACCUMUIATORINJECTION BEGINS W
D 0 HIGH HEAD SAFElY INJECTION BEGINS (OFFSITE POWER AVAIIABLEI W END OF BYPASS N END OF BLOWDOWN LOW HEAD SAFElY INJECTION BEGINS (OFFSITE POWER AVAIIABLE)
CONTAINMENTSPRAY SYSTEM STARTS (OFFSITE POWER AVAILABLE)
BOTTOM OF CORE RECOVERY SAFETY INJECTION FLOW INTERRUPTION HIGH HEAD SAFElY INJECTION RESTORED ACCUMUIATORS EMP1Y LOW HEAD SAFElY INJECTION RESTORED CORE OUENCHED 0 SWITCH TO SUMP RECIRCUlATION ON RWST LOW LEVELALARM N
G 45.0 40.0
. 35.0 30.0 I 25.0 20.0 15.0 10.0 0 50 100 150 200 250 300 Time (seconds)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-2 LOCA - DECLG - ALL CASES Containment Pressure Transient Versus Time
Hot Rod Hot Aeeembly Guide Tub ~ Aeeembly Average Rod Aeeembly Low Power/Periphery Aeeembly 2500 2000
/I
/s 4
//
I ~/
1500 a
L
~ Iooo CL E
I r I Io
/
500
/
XI I ll I
<JL/i 50 0 I 0 2 2 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-3A LOCA - DECLG - APPENDIX K CASE A - Co = 0.4 Cladding Temperature Transient (7.63 ft from bottom of core)
Versus Time RGEFSAR.WPF
--- Rot Rod Hot Aeeembly Gula Tub ~ Aeeembly Average Rod Aeeembly Low Power/Periphery Aeeembly 2500 2000 ot 1 /)
/(U I (
/ )I
//
yl 1500 lD L
CI 1000 I)
I
)I 500 50 0 10 20 2 0 3 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-3B LOCA DECLG - APPENDIX K CASE B - Co = 0.4 Cladding Temperature Transient (7.63 ft from bottom of core)
Versus Time RGEfSAR.WP f <
Hot Rod Hot Aeeembly Gula Tub ~
Average Rod Aeeembly Aeeembly Low Power/Periphery Aeeembly 2500 2000 1
'L
/I /
f500 L
CI L
fooo I
/ l weal I // l!
i ri I
500 50 0 I 0 2 0 2 0 3 0 Time (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6,4.2-3C LOCA - DECLG - APPENDIX K CASE C Co = 0 4 Cladding Temperature Transient (7.63 ft from bottom of core)
Versus Time RGEFSAR.WPF c,
~
s Hot Rod Hot Assembly s Guide Tube Assembly Average Rod Assembly Low power/Periphery Assembly 2500 2000 1500 LtJ LLJ "1 000 CL bJ 500 0
0 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-3D LOCA DECLG SUPERBOUNDED CJJ 0 6 Cladding Temperature Transient (7.50 ft from bottom of core)
Versus Time REEFS AR.%PP,L-)
2500 2000 1500 cn 1000 CL 500 50 1010'2 T I Al8 S 2 0 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-4A LOCA - DECLG - APPENDIX K CASE A - Co = 0 4 Top of Core Pressure Transient (Channel 10, Node 7)
Versus Time RGEFShR.WPF; J5 .
/'
2500 2000 1500 cn 1000 CL 500 50 101020 Time (8) 2 0 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-4B LOCA - DECLG - APPENDIX K CASE B - C~ = 0.4 Top of Core Pressure Transient (Channel 10, Node 7)
Versus Time RGEFSAR.WPF; <~
2500 2000 1500 m 1000 Q
500 50 10102 Time s 2 0 ROCHESTER GAS AND EL'ECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-4C LOCA DECLG APPENDIX K CASE C Co = 0 4 Top of Core Pressure Transient (Channel 10, Node 7)
Versus Time RGEFSAR.WPF
2500 2000 1500 1000 LIJ 500 50 100 150 200 250 300 TIME (Sj ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-4D LOCA - DECLG - SUPERBOUNDED - C~ = 0.6 Top of Core Pressure Transient (Channel 10, Node 7)
Versus Time RGEFSAR.WPF
~ 8 O
~ 6
~ 2 50 I 0 0 2 0 2 0 3 0 Time s ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6,4.2-5A LOCA - DECLG - APPENDlX K CASE A - C~ = 0,4 Upper Plenum Void Fraction Transient (Channel 25, Node 2)
Versus Time RGEFSAR.WPP'. "=
~ 8
~ 6
~
4 o
~ 2 50 101020 Time (8) 3 0 ROCHESTER GAS AND EI ECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-5B LOCA DECLG APPENDIX K CASE B C() 0 4 Upper Plenum Void Fraction Transient (Channel 25, Node 2)
Versus Time RGEFSAR.WPF.
~ 8 O
~ 6 o
I 4
~ 2 50 10 Time I 020 (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-5B LOCA - DECLG - APPENDIX K CASE B - C() = 0.4 Upper Plenum Void Fraction Transient (Channel 25, Node 2)
Versus Time RGEFSAR.WPF
~ 2 0
0 50 100 150 200 250 300 TIME (Sj ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-5D LOCA - DECLG - SUPERBOUNDED - CG = 0.6 Plenum Void Fraction Transient (Channel 25, Node 2)
'pper Versus Time RGEFSAR.WPF - 42
Liquid Flow Vapor Flow Entrainment Flow 10000 co 5000 E
-5000
-10000 50 0 I 0 2 2 0 Time (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-6A LOCA - DECLG - APPENDIX K CASE A - Co = 0.4 8ottom of Core I iqvid, Vapor, Entrainment Flow Transient (Channel 10, Node 1) Versus Time RGEFSALWPF,
L1qu1d Flor Vapor Floe Entrainment Floe 10000 ca 5000 E
-5000
-10000 50 0 1 0 2 0 2 0 3 0 Time s ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-6D LOCA - DECLG - APPENDIX K CASE B - Co 0.4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 10, Node 1) Versus Time RGEFSAR.NtPF,
Llqold Flow Vapor Flow
-- Entrainment Flow 10000 co 5000 E
o -5000
-10000 50 10 10 20 20 30 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-6C LOCA - DECLG - APPENDIX K CASE C - Co = 0.4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 10, Node 1) Versus Time RGEFSAR.WPF,. d
~
o o Liquid Flow Vapor Flow Entrainment Flow 10000 5000
-5000
-10000 0 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-6D LOCA DECLG SUPERBOUNDED C() 0 6 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 10, Node 1) Versus Time RGEFSAR.WPF r
Top ot Core Choo' Bottom of Core Charms I 10 10 v
I I
I I
I I
C)
~ 8 I II II I II nI I I
II Lp
)
II )(
) ))I II I(
I I(
((I
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~ 2 I)
I)
)( ()I )
)I ((I )
)( ()I I
)I)( I IP( I If 50 10 10 2 2 0 3 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-7A LOCA DECLG - APPENDIX K CASE A - Co = 0.4 Top and Bottom Core Void Fraction Transient (Channel 10, Nodes 1, 7) Versus Time RGEBAR.WPF-
Bottomof Core Top Choo' of Core Channel 10 10 lg I
I I
I I
I r8 CO
~ 6 I
CL II I II I CD ] I( (
CO II ((I ( I I
I'I,) I,l I I I
(li PI I I(
I I Il I I(( I Il I
~ 2 I((
I)
)I I) I
)I( I II( I
) I) I IA) I 50 0 2 0 2 0 Time (e)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-7B LOCA - DECLG - APPENDIX K CASE B - Co = 0.4 Top and Bottom Core Void Fraction Transient (Channel 10, Nodes 1, 7) Versus Time RGEFSAR.WPF I
Top ol Core Channel TC Bottom of Core Channel 10 I
I II II II
- 8 II II II II II II II CO I
~ 6 I I I CD I I I lf I II 1 lf I I) I I C3 I) )
,4 CO I) )
lir
~ 2 r Il>
I 50 0 I 0 2 0 2 0 Time (I)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-7C LOCA - DECLG - APPENDIX K CASE C - Co = 0.4 Top and Bottom Core Void Fraction Transient (Channel 10, Nodes 1, 7) Versus Time RGEFSAR.WPP ll
~ Top of Core Channel l0 Bottom of Core Chonnot 10
~ 8 C) 6 I
~
Q
. 4
)
C)
~ 2 50 100 ,150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-7D LOCA DECLG SUPERBOUNDED Co= 06 Top and Bottom Core Void Fraction Transient (Channel 10, Nodes 1, 7) Versus Time RGEFSAR.WPF
200
~+v
< I I
I
~ 150 I
I I
E I I
I I
I I
o 100 I
I I
I I
I I
I I
I C7 50 I
I I
I I
I I
I I
I 50 10 10 2 0 2 0 3 0 Time (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-8A LOCA - DECLG - APPENDIX K CASE A - C() = 0.4 Upper Plenum Injection Transient (Component 22, Cell 1)
Versus Time RGEFSAR.WPF '
200
~ V~
r" 't I
I
~ 150 I
E a 100 I
I I
I I
I I
o I z 50 50 10 10 2 0 3 0 T Ime ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-8B LOCA DECLG - APPENDIX K CASE B - C~ = 0.4 Upper Plenum Injection Transient (Component 22, Cell 1)
Versus Time RGEFSAR.WPF
200 yV I
I I
I
~ 150 I
I I
I E I I
I I
I I
a 100 I I
I I
I I
I I
I 50 I
I I
I I
I I
I I
50 10102 Time S ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6,4.2-8C LOCA DECLG APPENDIX K CASE C " Co = 0 4 Upper Plenum Injection Transient (Component 22, Cell 1)
Versus Time RGEFSALWPF '
200 cn 150 100 50 100 150 200 250 300 TIME (Sj ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FiNAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-8D LOCA DECLG SUPERBOUNDED C() 0 6 Upper Plenum Injection Transient (Component 22, Cell 1)
Versus Time RGEFSAR.WPP
60
<<<<'1I I
50 E 40 a 30 O
Lx 20 D
10 50 10102 Time s 2 0 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-9A LOCA DECLG APPENDIX K CASE A Co 04 Intact Cold Leg Safety Injection Transient (Component 6, Cell 6) Versus Time RGEFSAR.WPF
60 50
~E 40 o 30 C!
20 10 50 0 1 0 2 3 0 Time ( s )
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-9B LOCA - DECLG - APPENDIX K CASE B - C~ = 0.4 Intact Cold Leg Safety Injection Transient (Component 6, Cell 6) Versus Time RCEFSAR.WPF,
$0 50 E 40 a 30 20 CD 10 50 1 0 0 2 2 0 Time s ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15,6.4.2-9C LOCA DECLG APPENDIX K CASE C Co = 0 4 Intact Cold Leg Safety Injection Transient (Component 6, Cell 6) Versus Time RGEFSAR.WPF ~ ~
60 50 40 LxJ 30 CO 20 10 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-9D LOCA DECLG SUPERBOUNDED Cp 0 6 Intact Cold Leg Safety Injection Transient (Component 6, Cell 6) Versus Time RGEFSAR.WPF ~ '~
- 2500 2000 E 1500
~ 1000 O
4 500 I
I I
I I
li I
-500 50 10102 Time s 2 0 3 0 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-10A LOCA - DECLG - APPENDIX K CASE A - C() = 0.4 Intact Cold Leg Accumulator Injection Transient (Component 10, Cell 2) Versus Time RGEFSAR WPF-
2500 2000 1
\
\
E 1500 a 1000 o
500 I
I I
I I
li (il'(
I I
-500 50 0 0 2 3 0 Time s ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-10B LOCA DECLG APPENDIX K CASE B Co = 04 Intact Cold Leg Accumulator Injection Transient (Component 10, Cell 2) Versus Time RGEFSAR.WPF ~
2500 2000 E 1500 a 1000 C) 500 I I
I I
I I ~ (
-500 50 0 Time 10,20 0 (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R.'. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-1 OC LOCA DECLG - APPENDIX K CASE C - Co = 0.4 Intact Cold Leg Accumulator Injection Transient (Component 10, Cell 2) Versus Time RGEFSAR.WPF,
2500 2000 1500 1000 C) 500
-500 50 100 150 200 250 300 TIME (Sj ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-1 OD LOCA - DECLG - SUPERBOUNDED - Co = 0.6 Intact Cold Leg Accumulator Injection Transient (Component 10, Cell 2) Versus Time RGEFSAR.WPF,
Lldufd Flow Vapor flow Entrainment Flow 6000 4000 2000 a
-2000
-4000 50 0 10 20 2 0 3 0 Time (e)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-11A LOCA DECLG - APPENDIX K CASE A - Co = 0.4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 11, Node 1) Versus Time RGEFSAR.WPF,
--- Ltquia Fiat Vapor Flow
-- -- Entrainment Flow OOOO 4000 2000
-2000
-4000 50 0 0 2 2 0 3 0 Time (s )
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-11B LOCA - DECLG - APPENDIX K CASE B - Co = 0.4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 11, Node 1) Versus Time
Llquld Flow Vapor Floe EntraInment Floe 6000 4000 E
2000 0
C7
-2000
-4000 50 0 1 0 2 2 0 Time (e)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-11C LOCA DECLG APPENDIX K CASE C Co = 0 4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 11, Node 1) Versus Time RGEFSAR.WPF
~
o a Liquid .Flow Vapor Flow Entrainment Flow 6000 4000 2000
-2000
-4000 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-11D LOCA DECLG SUPERBOUNDED CD 0 6 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 11, Node 1) Versus Time RGEFSAR.WPF i
Liquid FIoe Vapor Flow Entrainment Flow 4000 co 2000 E
o -2000
-4000 50 10 10 2 2 0 3 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT'igure 15.6.4.2-1 2A LOCA DECLG APPENDIX K CASE A Co 0 4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 13, Node 1) Versus Time RGEFSAR.WPF ~
~, ~
Liquid Flow Vapor Flow
-- -- Entrainment Flow 4000 co 2000 E
o -2000
-4000 50 0 10 20 2 0 Time (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-12B LOCA DECLG APPENDIX K CASE B Co 0 4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 13, Node 1) Versus Time RGEFSAR.WPF ~
~'
Llwold Flow Vapor Floe Entrainment Flow 4000 3000 co 2000 E
1000
-1000 a -2000
-3000
-4000 50 0 10 20 2 0 3 0 Time (8)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4,2-12C LOCA DECLG APPENDIX K CASE C Co = 04 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 13, Node 1) Versus Time RGEFSAR.WPF
~
o o Liquid Flow Vapor Flow Entrainment F low 4000 2000
-2000
-4000 0 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-12D LOCA - DECLG - SUPERBOUNDED- Cp =0.6 Bottom of Core Liquid, Vapor, Entrainment FlowiTransient (Channel 13, Node 1) Versus Time RGEFSAR.WPF-
Liquid Flow
Entrainment Vapor Flaw Flow 200 100 E
o -100 C7
-200
-300 50 0 10 20 2 0 3 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-1 3A LOCA DECLG APPENDIX K CASE A" Co = 04 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 12, Node 1) Versus Time RGEFSAR.WPF ~
--- Ltqutd Floe Vapor Flow
-- - - Entrainment Flow 150 100 50 E
0 a
-50 o
~ -100 C7
-150
-200 50 0 10 20 2 0 3 0 Time (s)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-1 3B LOCA DECLG APPENDIX K CASE B Co = 0 4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 12, Node 1) Versus Time RGEFSAR.WPF ~,
Ltdutd Ftdd Vapor Floe Entrainment Flow 200 150 co 100 E
50 o 0 Ck C)
-50 o -100
-150
-200 50 0 1 0 2 2 0 Time (e)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-13C LOCA - DECLG - APPENDIX K CASE C - Co = 0.4 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 12, Node 1) Versus Time RGEFSAR.WFF-
a a
a Liquid Flow Vapor Flow Eatraiamaai Flow 150 100 50 LQ I
CL
-50 C) cn -1 00 150
-200 50 100 150 200 250 300 TIME (S)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT 0 Figure 15.6.4.2-1 3D LOCA DECLG SUPERBOUNDED Co = 0 6 Bottom of Core Liquid, Vapor, Entrainment Flow Transient (Channel 12, Node 1) Versus Time RGEFSALWPP
4000 I
I I
I E I I
300.0 I I
cd I o I I
I K I oC I,0 L J.
I I
I I I, I I I I I I I I 100.0 0.0 0 200 400 600 800 1,000 1,200 1,400 Reactor Coolant System Pressure (psla)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6.4.2-1 4 LOCA - DECLG - ALL CASES Intact Cold Leg Safety Injection Versus Pressure RGEFSAR.WPP ',, )
1,400.0 1,200.0 6
1,000.0 K
G: 800.0 C
600.0 400.0 200.0 0.0 0 20 40 60 80 100 120 140 Reactor Coolant System Pressure (psia)
ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT I,
~ Figure 15.6,4.2-15 LOCA - DECLG - ALL CASES Upper Plenum (RHR) Injection Versus Pressure RGEFSAR.WPF i
~'
l l f I,
r l
,/