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#REDIRECT [[LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 15, Accident Analysis - Classification of Plant Conditions]]
| number = ML17046A563
| issue date = 01/30/2017
| title = Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 15, Accident Analysis - Classification of Plant Conditions
| author name =
| author affiliation = PSEG Nuclear, LLC
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000272, 05000311
| license number = DPR-070, DPR-075
| contact person =
| case reference number = LR-N17-0034
| package number = ML17046A230
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 28
}}
 
=Text=
{{#Wiki_filter:SECTION 15 ACCIDENT ANALYSIS Classification of Plant conditions Since 1970 the American Nuclear Society (ANS) classification of plant conditions has been used which divides plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows: 1. Condition I: Normal Operation and Operational Transients 2. Condition II: Faults of Moderate Frequency 3. Condition III: Infrequent Faults 4. Condition IV: Faults The basic principle applied in relating design requirements to each of the conditions is that the most probable occurrences should yield the least radiological risk to the public and those extreme situations having the potential for the greatest risk to the public shall be those least likely to occur. Where applicable, reactor trip system and engineered safeguards functioning is assumed to the extent allowed by considerations, such as the single failure criterion, in fulfilling this principle. 15.1 CONDITION I -NORMAL OPERATION AND OPERATIONAL TRANSIENTS Condition I occurrences are those which are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, Condition I occurrences are accommodated with margin between any plant 15.1-1 SGS-UFSAR Revision 6 February 15, 1987 parameter and the value of that parameter which would require either automatic or manual protective action. Inasmuch as Condition I occurrences occur frequently or regularly, they must be considered from the point of view of affecting the consequences of fault conditions (Conditions II, III, and IV). In this regard, analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to the most adverse set of conditions which can occur during Condition I operation. The following is a typical list of condition I events. Steady State and Shutdown Operations REACTIVITY MODE CONDITION, K.rr THERMAL POWER* 1. Power Operation 2:0.99 >5% 2. Startup 2:0.99 3. Hot Standby <0.99 0 4. Hot Shutdown <0.99 0 s. cold Shutdown <0.99 0 6. Refueling** 0
* Excluding decay heat AVERAGE COOLANT TEMPERATURE 2:350&deg;F 2:350&deg;F 2:350&deg;F 200&deg;F < Tavq < 350&deg;F ** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. Operation with Permissible Deviations various deviations which may occur during continued operation as permitted by the plant Technical specifications must be considered in conjunction with other operational modes. These include: 1. Operation with components or systems out of service (such as power operation with a reactor coolant pump out of service) 15.1-2 SGS-UFSAR Revision 16 January 31, 1998 
: 2. Leakage from fuel with cladding defects 3. Activity in the reactor coolant a. Fission products b. Corrosion products c. Tritium 4 . Operation with steam generator leaks up to the maximum allowed by Technical Specifications Operational Transients 1. Plant heatup and cooldown {up to 100&deg;F/hour for the Reactor Coolant System; 200&deg;F/hour for the pressurizer). 2. Step load changes (up to +/-10 percent) 3. Ramp load changes (up to 5 percent/minute) 4. Load rejection up to and including design load rejection transient 15.1.1 Optimization of Control Systems A setpoint study has been performed to simulate performance of the Reactor Control and Protection Systems. Emphasis is placed on the development of a control system which will automatically maintain prescribed conditions in the plant even under the most conservative set of reactivity parameters with respect to both system stability and transient performance. For each mode of plant operation, a group of optimum controller setpoints is determined. In areas where the resultant setpoints are different, compromises based on the optimum overall performance are made and verified. A consistent set of control system parameters is derived satisfying plant operational 15.1-3 SGS-UFSAR Revision 6 February 15, 1987 requirements throughout the core life and for power levels between 15 and 100 percent. The study comprises an analysis of the following control systems: rod cluster assembly control, steam dump1 steam generator level, pressurizer pressure and pressurizer level. 15.1.2 Initial Power Conditions Assumed in Accident Analyses 15.1.2.1 Power Rating Table 15.1-1 lists the principal power rating values which are the basis for analyses performed in this section. The guaranteed Nuclear Steam Supply System {NSSS) thermal power output includes the thermal power generated by the reactor coolant pumps. The post-accident dose analyses performed in this section generally add conservatism by assuming some higher thermal power level ( 3632 MWt, 105% of 3459 MWt) than shown in Table 15.1-1. Where initial power operating conditions are assumed in accident analyses, the "guaranteed NSSS thermal power output" plus allowance for errors in steady state power determination is assumed. The thermal power values for each transient analyzed are given in Table 15.1-2. 15.1-4 SGS-UFSAR Revision 23 October 17, 2007 * 
-15.1.2.2 Initial Conditions For accident evaluation, the initial conditions are obtained by adding maximum steady state errors to rated values. The following steady state errors are considered for events not analyzed with Revised Thermal Design Procedure (RTDP):
* 1. 2. 3. Core power Average Reactor Coolant System (RCS) temperature Pressurizer pressure +/- 2 percent allowance calorimetric error * +/- 5&deg;F allowance for deadband and measurement error +/- 50 psi allowance for steady state fluctuations and measurement error The non-RTDP analyses were explicitly analyzed assuming a power calorimetric uncertainty of 2% together with an NSSS power of 3423 MWt (core power = 3411 MWt) . An evaluation has been performed that concludes that this is equivalent to analyzing with a calorimetric uncertainty of + 0. 6% and an NSSS power of 3471 MWt (core power = 3459 MWt). Initial values for core power, average RCS temperature and pressurizer pressure are selected to minimize the initial departure from nucleate boiling ratio (DNBR) unless otherwise stated in the sections describing specific accidents. The outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660&deg;F for steady state operation at rated power throughout core life due to the onset of nucleate boiling. Initially (beginning of life), this temperature is that of the cladding metal outer surface. During operation over the life of the core, the buildup of oxides and crud on the fuel rod surface causes the cladding surface temperature to increase. Allowance is made in the fuel center melt evaluation of this temperature rise. Since the thermal-hydraulic design basis limits departure from nucleate boiling (DNB), adequate heat transfer is provided between the fuel cladding and the reactor coolant so that the core thermal output is not limited by considerations of the cladding temperature. Figure 4.4-4 shows the axial variation of average cladding temperature for a typical rod (17 x 17 fuel assembly) both at beginning of life (BOLl and end of life {EOL) . 15.1-5 SGS-UFSAR Revision 19 November 19, 2001 I End of life is after three typical cycles of operation {approximately 20,000 effective full-power hours). These temperatures are calculated using the Westinghouse fuel rod model ( 1) which has been reviewed and approved by the Nuclear Regulatory Commission (NRC) . 15.1.2.3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core minimizes adverse_ power distribution through the placement of control rods and operation instructions. The power distribution may be characterized by the radial factor F AH and the total peaking factor F q The peaking factor limits are given in the Technical Specifications. For transients which may be DNB limited, the radial peaking factor is of importance. The radial peaking factor increases with decreasing power level due to rod insertion. This increase in F AH is included in the core limits illustrated on Figure 15.1-1. All transients that may be DNB limited are assumed to begin with a FAH consistent with the initial power level defined in the Technical Specifications. The axial power shape used in the DNB calculation is the chopped cosine as dlscussed in Section 4.4.3.2, For transients which may be overpower limited, the total peaking factor F is q of importance. The value of F may increase with decreasing power level such q that. full power hot spot heat flux is not exceeded, i.e., F Power = design q hot spot heat flux. All transients that may be overpower limited are assumed to begin with a value of F consistent with the initial power level as defined q in the Technical Specifications. The value of peak kW/ft can be directly related to fuel temperature as illustrated on Figures 4.4-1 and 4.4-2. For 15.1-6 SGS-UFSAR Revision 18 April 26, 2000 transients which are slow with respect to the fuel rod thermal time the fuel temperatures are illustrated on Figures 4.4-l and 4.4-2. For transients which are fast with respect to the fuel rod thermal time for example, rod ejection, a detailed heat transfer calculation is made. 15.1.3 Trip Points and Time Delays to Trip Assumed in Accident Analyses A reactor trip signal acts to open two trip breakers connected ir: series feeding power to the control rod drive mechanisms (CRDM). The loss of power to the mechanism coils causes the mechanisms to release the rod cluster control assemblies (RCCA) which then fall by gravity into the core. There are various instrumentation delays associated with each trip function, including delays in signal actuation, in opening the trip breakers, and in the coil release of the rods by the mechanisms. The coil release of the rods is* conservatively assumed to be 0.15 second. The total delay to trip is defined as the time from when the monitored parameter exceeds its trip setpoint at the channel sensor to the time when the rods begin to drop. Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in Table 15.1-3. Reference is made in that table to overtemperature and overpower 6.T trip shown on Figure 15.1-1. The overtemperature dT setpoints shown on Figure 15. 1-l along with all other evaluated DNBRS were calculated assuming approximately 15 percent margin in the crltical heat flux calculation, as discussed in Section 4.4.2.1. The difference between the limiting trip point assumed for the analysis and the nominal trip point represents an allowance for instrumentation channel error and setpoint error. During preliminary startup tests, it will be demonstrated that actual instrument errors and time delays are equal to or less than the assumed values. 15.1-7 SGS-UFSAR Revision 18 April 26, 2000 Public Service Electric & Gas, in its letter dated February 25, 1985, addressed NRC concerns regarding the replacement of the existing RCS resistance temperature detectors (RTD) with environmentally qualified RTDs. The new RTDs have a slower response time than the originally installed RTDs, and, therefore, a review of the accidents in which these RTDs are relied upon was performed. The review determined that reanalysis was only required for the uncontrolled RCCA bank withdrawal at power accident described in Section 15. 2. 2. The reanalysis was performed using the same methodology and inputs as the original J analysis except that a 7-second delay was assumed for the overtemperature trlp. It was concluded that a lower DNBR than originally calculated would be reached; however, in no case would the minimum DNBR fall below the limit value. 15. 1. 4 Instrumentation Drift and Calorimetric Errors Power Range Neutron Flux The instrumentation drift and calorimetric errors used in establishing the maximum overpower setpoint are presented in Table 15.1-4. The calorimetric error is the error assumed in the determination of core thermal power as obtained from secondary plant measurements. The total ion chamber current (sum of the top and bottom sections) is calibrated (set equal} to this measured power on a periodic basis. The secondary power is obtained from measurement of feedwater flow, feedwater inlet temperature to the steam generators and steam pressure. High accuracy instrumentation is provided for measurements with accuracy tolerances much higher than those which would be required to control feedwater flow. 15.1-8 SGS-UFSAR Revision 18 April 26, 2000 15.1.5 Rod Cluster Control Assembly Insertion Characteristics The negative reactivity insertion following a reactor trip is a function of the acceleration of the RCCAs and the variation in rod worth as a of rod position. With respect to accident analyses, the critical parameter is the time of insertion up to the dashpot entry or approximately 85 percent of the rod cluster travel. For accident analyses it is conservatively assumed that, after the total delay to trip {defined in Section 15.1.3), the insertion time from beginning of rod motion to dashpot entry is 2. 7 seconds. The RCCA position versus time assumed in accident analyses is shown on Figure 15.1-2. Figure 15.1-3 shows the fraction of total negative reactivity insertion for a core where the axial distribution is skewed to the lower region of the core. An axial distribution which is skewed to the lower region of the core can arise from a xenon oscillation or can be considered as representing a transient axial distribution which would exist after the RCCA bank had already traveled some distance after trip. This lower curve is used as input to all point kinetics core models used in transient analyses. There is inherent conservatism in the use of this curve in that it is based on o. skewed distribution which would exist relatively infrequently. For cases than those associated with xenon oscillations significant negative would have been inserted due to the more aistribution existing prior to trip. favorable axial The normalized RCCA negative reactivity insertion versus time is shown on Figure The curve shown on this figure was obtained from Figures 15.1-2 and 15.1-3. A total negative reactivity insertion following trip of 4 percent 6k is assumed in the transient analyses except where specifically noted This assumption is conservative with respect to the calculated trip reac:1vity worth available as shown in Table 4.3-3. 15.1-9 SGS-UFSAR Revision 18 April 26, 2000 I The normalized RCCA negative reactivity insertion versus time curve for an axial power distribution skewed to the bottom (Figure 15.1-4) is used in transient analyses. Where special analyses require use of three-dimensional or axial one-dimensional core models, the negative reactivity insertion resulting from reactor trip is calculated directly by the reactor kinetic code and is not separable from other reactivity feedback effects. In this case, the RCCA position versus time on Figure 15.1-2 is used as code input. 15.1.6 Reactivity Coefficients The transient response of the Reactor System is dependent on reactivity feedback effects, in particular the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are discussed in detail in Section 4. In the analysis of certain events, conservatism requires the use of large reactivity coefficient values whereas, in the analysis of other events, conservatism requires the use of small reactivity coefficient values. Some analyses such as loss of reactor coolant from cracks or ruptures in the RCS do not depend on reactivity feedback effects. The values used are given in Table 15.1-2; reference is made in that table to Figure 15.1-5 which shows the current lower and upper Doppler only power coefficient, as a function of power used in the transient analysis respectively. The basis for the revised most negative Doppler curve is the safety analysis performed for the Salem Unit 1 Cycle 6 reload design. (22) Those incidents found to be sensitive to the revised Doppler coefficient were reanalyzed. Table 15.1-2 gives a list of accidents presented in this FSAR and denotes those events reanalyzed for a new coefficient. The results of the analysis showed that the revised most negative Doppler curve can be accommodated with ample margin to the applicable FSAR safety limits. 15.1-10 SGS-UFSAR Revision 28 May 22, 2015 The justification for use of conservatively large versus small reactivity coefficient values are treated on an event-by-event basis. To facilitate comparison, individual sections in which justification for the use of large or small reactivity coefficient values is to be found are referenced below: Condition II Events l. Uncontrolled RCCA Bank Withdrawal From A Subcritical Condition 15.1-lOa SGS-UFSAR Section 15.2.1 Revision 7 July 22, 1987 THIS PAGE INTENTIONALLY BLANK 15 .1-lOb SGS-UFSAR Revision 7 July 22, 1987 
') .:.. 3. 4. 5. 6. 7. 8. 9. Uncontrolled RCCA Bank Withdrawal At Power Rod Cluster Control Assembly Misalignment Uncontrolled Boron Dilution Partial Loss of Forced Reactor Coolant Flow Loss of External Electrical Load and/or Turbine Trip Loss of Normal Feedwater Loss of Offsite Power to The Station Auxiliaries Excessive Heat Removal due to Feedwater System Malfunctions 10. Excessive Load Increase Incident 1:. Accidental Depressurization of The RCS 12. Accidental Depressurization of Main Steam Systems 13. Spurious Operation of the Safety Injection System (SIS) at Power Condition III Events 2. Complete Loss of Forced Reactor Coolant Flow Single RCCA Withdrawal at Full power 15.1-11 SGS-UFSAR 15.2.2 15.2.3 15.2.4 15.2.5 15.2.7 15.2.8 15.2.9 15.2.10 15.2.11 15.2.12 15.2.13 15.2.14 15.3.4 15.3.5 Revision 18 April 26, 2000 Condition IV Events 1. Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) 2. Major Secondary System Pipe Rupture 3. Major Rupture of a Main Feedwater Line 4. Stearn generator Tube Rupture 5. Slngle Reactor Coolant Pump Locked Rotor and Reactor Coolant Pump Shaft Break 6. Fuel Handling Accident 7. Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) B. Containment Pressure Analysis 15.1.7 Fission Product Inventories Activities in the Core 15.4.1 15.4. 2 15.4.3 15. 4
* 4 15.4.5 15.4.6 15.4.7 15.4.8 The fission product inventories which are important from a health hazards point of view consider inhalation dose and external dose due to immersion. The bases for the total core iodine (inhalation dose) and noble gas (external dose) inventories are described in Section 11. 1.1. These inventories are given in Table 11.1-1. 15.1.7.2 Activities in the Fuel Pellet Cladding Gap The fraction of core activity assumed to be in the gap can vary depending on the specific application. Gap activity is the primary source term for the locked rotor, rod ejection and fuel handling accidents. The gap activity basis is discussed as part of the assumptions described in the specific accident section of Chapter 15. 15.1-12 SGS-UFSAR Revision 18 April 26, 2000 ---.......--
THIS PAGE INTENTIONALLY BLANK 15.1-13 SGS-UFSAR Revision 16 January 31, 1998 15.1.8 Residual Decay Heat (ANS-1979) Residual heat in a subcritical core consists of: 1. Fission product decay energy, 2. Decay of neutron capture products, and 3. Residual fissions due to the effect of delayed neutrons. These constituents are discussed separately in the following paragraphs. 15.1-14 SGS-UFSAR Revision 18 April 26, 2000 *-
15.1.8.1 Fission Product Decay For short times (< 103 seconds) after shutdown, data on yields of short half-life isotopes is sparse. Very little exprrimental data is available for the y-ray contributions and even less for the contribution. Several authors have compiled the available data into a conservative estimate of fission product decay energy for short times after shutdown, notably Shure (7), Dudziak (8), and Teage (9). Of these three select ions, Shure 1 s curve is the highest and it is based on the data of Stehn and Clancy (10) and Obenshain and Foderaro (11). The fission product contribution to decay heat which has been assumed in the accident analyses is the curve of Shure increased by 20 percent for conservatism. This curve with the 20-percent factor included is shown on Figure 15.1-6. 15.1.8.2 Q5:cay of U-238 Capture Products Betas and gammas from the decay of U-239 (23.5 minute half-life) and Np-239 (2.35 day half-life) contribute significantly to the heat generation after shutdown. The cross-section for production of these isotopes and their decay schemes are relatively well known. For long irradiation times their contribution can be written as: P1/Po SGS-UFSAR E)'1 + E(?l 200 Mev c(l+a) e -A. watts/watt t 1 watts/watt -A t (e 2 15.1-15 -A t -e I ) -A t + e 2 ) Revision 6 February 15, 1987 where: Pl/Po P2/Po t c(l+a) -. = the energy from U-239 decay = the energy from Np-239 decay = the time after shutdown (seconds) = the ratio of U-238 captures to total fissions = 0.6(1 + 0.2) =the decay constant of U-239 = 4.91 x 10-4 seconds-! = the decay constant of Np-239 decay = 3.41 x 10-6 seconds-l = total y-ray energy from U-239 decay = .06 Mev = total y-ray energy from Np-239 decay = . 30 Mev = total (3-ray energy from U-239 decay = 1/3 X 1.18 Mev = total (3-ray energy from Np-239 decay = 1/3 X .43 Mev (Two-thirds of the potential (3-energy is assumed to escape by the accompanying neutrons.) This expression with a margin of 10 percent is shown on Figure 15.1-6. The 10-percent margin, compared to 20 percent for fission product decay, is justified by the availability of the basic data required for this analysis. The decay of other isotopes, produced by neutron reactions other than fission, is neglected. 15.1.8.3 Residual Fissions The time dependence of residual fission power after shutdown depends on core properties throughout a transient under consideration. Core average conditions are more conservative for the calculation of reactivity and power level than actual local conditions as they would exist in hot areas of the core. Thus, unless otherwise stated in the text, static power shapes have been assumed in the analyses and these are factored by the time behavior of core average fission power calculated by a point model kinetics calculation with six delayed neutron groups. 15.1-16 SGS-UFSAR Revision 6 February 15, 1987 For the purpose of illustration only a one delayed neutron group calculation, with a constant shutdown reactivity of -4 percent ak, is shown on Figure 15.1-6. 15. 1. 8. 4 Distribution of_ Decay Heat Followin_g -Coolant Accident During a loss-of-coolant accident (LOCA) the core is rapidly shut down by void formation or RCCA insertion, or both, and a large fraction of the heat generation to be considered comes from fission product decay gamma rays. This heat is not distributed in the same manner as steady state fission power. Local peaking effects which are important for the neutron dependent part of the heat generation do not apply to the gamma ray contribution. The steady state factor of 97.4 percent which represents the fraction of heat generated within the clad and pellet drops to 95 percent for the hot rod in a LOCA. For example, consider the transient resulting from the postulated double-ended break of the largest RCS pipe; one-half second after the rupture about 30 percent of the heat generated in the fuel rods is from gamma-ray absorption. The gamma power shape is less peaked than the steady state fission power shape, reducing the energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect is a reduction of 10 percent of Lhe gamma-ray contribution or 3 percent of the total. Since the water density is considerably reduced at this time, an average of 98 percent of the available heat is deposited in the fue] rods, the remaining 2 perC'ent being absorbed by water, thimbles, sleeves and grids. The net effect is a factor of 0.95 rather than 0.974, to be applied to the heat production in the hot rod. 15.1.9 Computer Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular, 15.1-17 SGS-UFSAR Revision 6 February 15, 1987 very specialized codes in which the modeling has been developed to stimulate one given accident, such as the SATAN-V Code as in the analysis of the RCS pipe rupture (Section 15.4. 1), and which consequently have a direct bearing on the analysis of the accident itself, are sununarized in their respective accident analyses sections. The codes used in the analyses of each transient have been listed in Table 15.1-2. 15.1.9.1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section of a metal clad uo2 fuel rod and the transient heat flux at the surface of the clad using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temperature, density). The code uses a fuel model which exhibits the following -features simultaneously: 1. A sufficiently large number of radial space increments to handle fast transients such as rod ejection accidents. 2. Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation. 3. The necessary calculations to handle post-DNB transients: film boiling heat transfer correlations, Zircaloy-water reaction and partial melting of the materials. The gap heat transfer coefficient is calculated a\cording to an elastic pellet model (refer to Figure 15. l-7). The thermal expansion of the pellet is calculat<*d as the sum of the radial (one-dimensional) expansions of the rings. Each ring is assumed to expand freely. The cladding diameter is calculated based on thermal expansion and internal and external pressures. 15.1-18 SGS-UFSAR Revision 6 February 15, 1987 If the outside radius of the expanded pellet is smaller than the inside rad1us of the expanded clad, there is no fuel-clad contact and the gap conductance calculated on the basis of the thermal conductivity of the gas contained in the gap. If the pellet outside radius so calculated is larger than the clad inside radius (negative gap), the pellet and the clad are pictured as exerting upon each other a pressure sufficiently important to reduce the gap to zero by elastic deformation of both. This contact pressure determines the gap heat transfey coefficient. FACTRAN 1s further discussed in Reference 12. 15.1-19 SGS-UFSAR Revision 18 April 26, 2000 THIS PAGE INTENTIONALLY BLANK 15.1-20 SGS-UFSAR Revision 18 April 26, 2000 *-*
5.1.9.2 LOFTRAN The LOFTRAN program is used for studies of transient response of a pressurized water reactor system to specified perturbations in process parameters. LOFTRAN simulates a multi-loop system by a lumped parameter single loop model containing reactor vessel, hot and cold leg piping, steam generator (tube and shell sides) and the pressurizer. The pressurizer heaters, spray, relief and safety valves are also considered in the program. Point model neutron kinetics, and reactivity effects of the moderator, fuel, boron and rods are lncluded. The secondary side of the steam generator utilizes a homogeneous, saturated mixture for the thermal transients and a water level correlation for indication and control. The Reactor Protection System is simulated to include reactor trips on neutron flux, overpower and overtemperature reactor coolant hlgh and low pressure, low flow, and high pressurizer level. Control systems are also simulated including rod control, steam dump, feedwater control, and pressurizer pressure control. The Safety Injection System, including the accumulators, is also modeled. LOFTRAN is a versatile program which is suited to both accident evaluation and control studies as well as parameter sizing. LOFTRAN also has the capability of calculating the transient value of DNB ratio basec on the input from the core limits illustrated on Figure 15.1-1. The core .'..lmits represent the minimum value of DNBR as calculated for typical or thimble is further discussed in Reference 15. :5.1.9.3 PHOENIX-P P!-lOENIX-F 1s a two-dimensional, multi-group transport theory computer code. The cross-section library used by PHOENIX-P contains cross-section data based on a 7C energy group structure derived from ENDF/B-VI files. PHOENIX-P per:orms o. 20 70 group nodal flux calculation which couples the individual subcel.'.. reg1ons (pellet, cladding, and moderator) as well as surrounding rods via a col!lsion probability technique. This 70 group solution is normalized by a 2oarse energy group flux solution derived from a discrete ordinates calculation. PHOENIX-P is capable of modeling all cell types needed for PWR core design application. SGS-UFSAR 15.1-21 Revision 18 April 26, 2000 PHOENIX-P calculates macroscopic cross-sections as a function of burnup, fuel type, and temperature for ANC (Section 15.1.9.4}. PHOENIX-P is further discussed in Reference 16. 15.1.9.4 ANC ANC is an advanced nodal code capable of two-dimensional and three-dimensional neutronics calculations. ANC is the reference model for certain safety analysis calculations, power distributions, peaking factors, critical boron concentrations, control rod worths, reactivity coefficients, etc. In addition, three-dimensional ANC validates one-dimensional and two-dimensional results and provides information about radial {x-y) peaking factors as a function of axial position. It can calculate discrete pin powers from nodal information as well. ANC is further discussed in Reference 17. 15.1.9.5 TWINKLE The TWINKLE program is multi-dimensional spatial neutron kinetics code, which was patterned after steady-state codes presently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transie!l.t neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed rnutli-region fuel-heat transfer model for calculating pointwise Doppler and moderator feedback effects. The code handles up to 2000 spatial points, and performs its own steady state initialization. Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits provide channelwise power, axial of:set, enthalpy, volumetric surge, pointwise power, fuel temperatures, and so on. SGS-UFSAR 15.1-22 Revision 18 April 26, 2000 The TWINKLE code is used to predict the kinetic behavior of a reactor for transients which cause a major perturbation in the spatial neutron flux distribution. TWINKLE is further described in Reference 18. 15.1-23 SGS-UFSAR Revision 11 July 22, 1991 15.1.9.6 THINC The THINC code is described in Section 4.4.3.1. 15.1.10 References for Section 15.1 1. Supplemental information on fuel design transmitted from R. Salvatori, Westinghouse NES, to D. Knuth, AEC, as attachments to letters NS-SL-518 (12/22/72), NS-SL-521 (1/4/73), NS-SL-524 (1/4/73) and NS-SL-543 ( l/12/73), (Westinghouse NES Proprietary}; and supplemental information on fuel design transmitted from R. Salvatori, Westinghouse NES, to D. Knuth, AEC, as attachments to letters NS-SL-527 (1/2/73) and NS-SL-544 (1/12/73). 2. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000. 3. DELETED 4. DELETED 5. DELETED 6. DELETED 7. Shure, K., "Fission Product Decay Energy in Bettis Technical Review," WAPD-BT-24, p. 1-17, December 1961. 15.1-24 SGS-UFSAR Revision 23 October 17, 2007 * * * 
: 8. Shure, K. and Dudziak, D. J., "Calculating Energy Released by Fission Products," Trans. Am. Nucl. Soc. 4 (1} 30 (1961). 9. U.K.A.E.A. Decay Heat Standard. 10. Stehn, J. R. and Clancy, E. F., "Fission-Product Radioactivity and Heat Generation" in "Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy," Volume 13, pp. 49-54, United Nations, Geneva, 1958. 11. Obenshain, F. E: and Foderaro, A. H., "Energy from Fission ?-reduct Decay," WAPD-P-652, 1955. 12. Hargrove, H. G., .,FACTRAN, a Fortran IV Code for Thermal Transients in a uo2 Fuel Rod," WCAP-7908, December 1989. 13. DELETED 14. DELETED 15. Burnett, T. W. T. , et al, "LOFT RAN Code Oeser iption, " WCAP-7 907, April 1974. 16. Nguyen, T. Q., et al, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores", WCAP-11596/ June 1988. 17 . Li u, Y. S. , et a l, "ANC: A Westinghouse Advanced Nodal Computer Code" 1 WCAP-10965, September 1986. 18. Risher, D. H., Jr. and Barry, R. F., "TWINKLE -A Multi-Dimensional Neut:ron Kinetics Computer Code," WCAP-7979, November 1972. 15.1-25 SGS-UFSAR Revision 18 April 26, 2000 
: 19. Deleted 20. Deleted 21. Liden, E. A., PSE&G to Varga, S. A., USNRC, "Supplemental Information Request for Amendment, Salem Generating Station Unit Nos. 1 and 2, Docket Nos. 50-272 and 50-311," February 25, 1985. 22. Letter from T. R. Croasdaile (Westinghouse) to J. T. Boettger (PSE&G),
 
==Subject:==
Safety Analysis for PSE&G Proposed Doppler Curve (Proprietary_ Document}, June 28, 1984; 84PS*-G-058, NFUI 84-366. 15.1-26 SGS-UFSAR Revision 11 July 22, 1991}}

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