ML20207S369

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Chapter 15, Accident Analyses, to Nonproprietary RESAR-SP/90 Westinghouse Advanced Pwr,Module 10, Containment Sys
ML20207S369
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Issue date: 11/30/1986
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0 15.0 ACCIDENT ANALYSES 15.0.1 General This chapter addresses the representative initiating events listed on Table 15-1 of Regulatory Guide 1.70, Revision 3, the " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", as they apply to a Westinghouse pressurized water reactor.

Certain items of Table 15-1 in the guide warrant comment, as follows:

1. Items 1.3 and 2.1 - There are no pressure regulators in the Nuclear Steam Supply System (NSSS) pressurized water reactor (PWR) design whose malfunction or failure could cause a steam flow transient.
2. Item 6.2 - No instrument lines from the reactor coolant pressure boundary f

j in the NSSS PWR design penetrate the Containment. (Forthedefinitionof the Reactor Coolant System boundary, refer to Section 5, ANSI-N18.2,

" Nuclear Safety Criteria for the Design of Stationary PWR Plants," 1973.)

l 15.0.2 Classification of Plant Conditions Since 1970 the ANS classification of plant conditions has been used to divide 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.

l 4. Condition IV: Limiting Faults.

O The basic principle applied in relating design requirements to each of the conditions is that the most probable occurrences should yield the least O WAPWR-CS 15.0-1 NOVEMBER, 1986 5723e:1d 8703190321 870309 PDR ADOCK 05000601 K PDR L

t 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.

15.0.2.1 Condition I - Normal Operation and Operational Transients O Condition I occurrences are those which are expected frequently or regularly in the course of normal plant operation, refueling, and maintenance. As such, Condition I occurrences are accommodated with margin between any plant 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 adverse conditions which can occur during Condition I operation.

Typical Condition I events are as follows:

1. Steady state and shutdown operations
a. Mode 1 - Power operation (> 5 to 100 percent of rated thermal power).

! b. Mode 2 - Startup (K,ffl 0.99, 5 5 percent of rated thermal power).

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c. Mode 3 - Hot standby (K,ff < 0.99, T,yg 1 350*F).
d. Mode 4 - Hot shutdown (K,ff < 0.99, 200*F 5 TAVG s 350*F).
O e. Mode 5 - Cold Shutdown (K,ff < 0.99, T,yg < 200*F).
f. Mode 6 - Refueling (K,ff 5 0.95, T,yg 1 140'F).

O WAPWR-CS 15.0-2 NOVEMBER, 1986 5723e:1d i

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O 2. Operation with permissible deviations

Various deviations which may occur during continued operation as permitted O by the plant Technical Specifications must be considered in conjunction O with other operational modes. These include
a. Operation with components or systems out ofservice(suchaspower operation with a reactor coolant pump out of service).

O b. Radioactivity in the reactor coolant, due to leakage from fuel with cladding defects and other sources.

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! 1) Fission products

! 2) Corrosion products

3) Tritium b
c. Operation with steam generator primary-to-secondary leakage up to the maximum allowed by the Technical Specifications.

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d. Testing as required by the Technical Specifications, t
3. Operational transients
a. Plant heatup and cooldown (up to 100*F/ hour for the reactor coolant system; 200'F/ hour for the pressurizer during cooldown and 100*F/ hour l l

for the pressurizer during heatup).  ;

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b. Step load changes (up to + 10 percent).  !

! c. Ramp load changes (up to 5 percent / minute).  ;

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d. Load rejection up to and including design full- load rejection j transient.

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15.0-3 NOVEMBER, 1986  !

4 WAPWR-CS j 5723e:1d i.

r~S b 15.0.2.2 Condition II - Faults of Moderate Frequency At worst, a Condition II fault results in a reactor trip with the plant being capable of returning to operation. By definition, these faults (or events) do not propagate to cause a more serious fault, i.e., Condition III or IV events. In addition, Condition 11 events are not expected to result in fuel rod failure or reactor coolant system or secondary system overpressurization.

The following faults are included in this category:

1. Feedwater system malfunctions causing a reduction in feedwater temperature (Subsection 15.1.1 of RESAR-SP/90 PDA Module 6/8, " Secondary Side SafeguardsSystem/SteamandPowerConversion").
2. Feedwater system malfunctions causing an increase in feedwater flow (Subsection 15.1.2 of RESAR-SP/90 PDA Module 6/8, " Secondary Side SafeguardsSystem/SteamandPowerConversion").

O 3. Excessive increase in secondary steam flow (Subsection 15.1.3 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and PowerConversion").

4. Inadver tent opening of a steam generator relief or safety valve causing a depressurization of the main steam system (Subsection 15.1.4 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and PowerConversion").
5. Loss of external electrical lead (Subsection 15.2.2 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").

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6. Turbinetrip(Subsection 15.2.3ofRESAR-SP/90PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion"). (

i 7. Inadvertent closure of main steam isolation valves (Subsection 15.2.4 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and l

i PowerConversion"),

i 15.0-4 NOVEMBER, 1986 WAPWR-CS 5723e:1d i

8. Loss of condenser vacuum and other events resulting in turbine trip 6/8, " Secondary Side (Subsection 15.2.5 of RESAR-SP/90 PDA Module Safeguards System / Steam and Power Conversion").

O' 9. Loss of nonemergency A-C power to the station auxiliaries (Subsection 15.2.6 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System /SteamandPowerConversion").

10. Loss of normal feedwater flow (Subsection 15.2.7 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").
11. Partial loss of forced reactor coolant flow (Subsection 15.3.1 of RESAR-SP/90 PDA Module 4 " Reactor Coolant System").
12. Uncontrolled rod cluster control assembly bank withdrawal from a suberitical or low power startup condition (Subsection 15.4.1 of RESAR-SP/90 PDA Module 5, " Reactor System").

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13. Uncontrolled rod cluster control assembly bank withdrawal at power (Subsection 15.4.2 of RESAR-SP/90 PDA Module 5, " Reactor System").
14. Control rod misalignment - Dropped full length assembly, dropped full length assembly bank, or statically misaligned full length assembly)

(Subsection 15.4.3ofRESAR-SP/90PDAModule5,"ReactorSystem").

15. Startup of an inactive reactor coolant loop at an incorrect temperature (Subsection 15.4.4 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System").
16. Inadvertent operation of emergency core cooling system during power operation (Subsection 15.5.1 of RESAR-SP/90 PDA Module 1 " Primary Side SafeguardsSystem").

O 17. Inadvertent opening of a pressurizer safety or relief valve (Subsection 15.6.1 of RESAR-SP/90 PDA Module 4. " Reactor Coolant Systems").

O WAPWR-CS 15.0-5 NOVEMBER, 1986 5723e:1d

18. Failure of small lines carrying primary coolant outside containment (Subsection 15.6.2 of RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System").

15.0.2.3 Condition III - Infrequent Faults By definition, Condition III occurrences are faults which may occur very infrequently during the life of the plant. They will be accommodated with the O- failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude immediate resumption of the operation. The release of radioactivity will not be sufficient to interrupt or restrict public use of those areas beyond the exclusion area boundary. A Condition III fault will not, by itself, generate a Condition IV fault or result in a consequential loss of function of the reactor coolant system or containment barriers. The following faults are included in this category:

1. Minor steam system piping failures (Subsection 15.1.5 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").
2. Complete loss of forced reactor coolant flow (Subsection 15.3.2 of RESAR-SP/90 PDA Module 4, " Reactor Coolant Systems").
3. Control rod misalignment - Single rod cluster control assembly withdrawal at full power (Subsection 15.4.3 of RESAR-SP/90 PDA Module 5, " Reactor Systems").
4. Inadvertent loading and operation of a fuel assembly in an improper position (Subsection 15.4.7 of RESAR-SP/90 PDA Module 5, " Reactor System").
5. Loss of reactor coolant from small ruptured pipes or from cracks in large pipes, which actuate the emergency core cooling system (Subsection 15.6.4 of RESAR-SP/90 PDA Module 1. " Primary Side Safeguards System").

O WAPWR-CS 15.0-6 NOVEMBER, 1986 5723e:1d

,O G 15.0.2.4 Condition IV - Limiting Faults Condition IV occurrences are faults which are not expected to occur, but are

[) postulated because their consequences would include the potential for release

' of significant amounts of radioactive material. They are the most drastic which must be designed against and represent limiting design cases. Plant design must be such as to preclude a fission product release to the environment resulting in an undue risk to public health and safety in excess V of guideline values of 10 CFR 100. A single Condition IV fault must not cause a consequential loss of required functions of systems needed to mitigate the consequences of the fault including those of the emergency core cooling system and containment. The following faults have been classified in this category:

1. Steam system piping failure (Subsection 15.1.5 of RESAR-SP/90 PDA Module 6/8,"SecondarySideSafeguardsSystem/SteamandPowerConversion").
2. Feedwater system pipe break (Subsection 15.2.8 of RESAR-SP/90 PDA

/ Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").

3. Reactor coolant pump rotor seizure (locked rotor) (Subsection 15.3.3 of RESAR-SP/90 PDA Module 4, " Reactor Coolant Systems").
4. Reactor coolant pump shaft break (Subsection 15.3.4 of RESAR-SP/90 PDA Module 4,"ReactorCoolantSystem").
5. Spectrum of rod cluster control assemblyejectionaccidents(Subsection 15.4.8 of RESAR-SP/90 PDA Module 5, " Reactor System").
6. Steam generator tube rupture (Subsection 15.6.3 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").
7. Loss-of-coolant accidents resulting from the spectrum of postulated piping breaks within the reactor coolant pressure boundary (Subsection 15.6.4 of RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System").

i O WAPWR CS 15.0-7 NOVEMBER, 1986 I

5723e:1d

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b 15.0.3 Optimization of Control Systems A control system automatically maintains prescribed conditions in the plant even under a conservative set of reactivity parameters with respect to both

O' system stability and transient performance. For each mode of plant operation, f 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 requirements throughout the core life and for various levels of power operation.

The system setpoints are derived by an analysis of the following control systems: rod control, steam dump, steam generator level, pressurizer pressure and pressurizer level.

15.0.4 Plant Characteristics and Initial Conditions Assumed in the Accident i

Analyses

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! 15.0.4.1 Design Plant Conditions Table 15.0-1 gives the guaranteed nuclear steam supply system thermal power output which is assumed in analyses performed in this report. This power

! output includes the thermal power generated by the reactor coolant pumps and l is consistent with the license application rating described in Chapter 1.0.

l Allowances for errors in the determ Sation of the steady-state power level are made as described in Subsection 15.0.4.2. The values of pertinent plant l'

parameters utilized in the accident analyses are given in Table 15.0-2. The thermal power values used for each transient analyzed are given in Table f

15.0-3 in RESAR-SP/90 PDA Modules 4, 5 and 6/8. ,

15.0.4.2 Initial Conditions iO For most accidents which are DNB limited, nominal values of initial conditions are assumed. The allowances on power, temperature, and pressure noted above i l

! are determined on a statistical basis and are included in the limit DNBR, as i WAPWR-CS 15.0-8 NOVEMBER, 1986 !

i 5723e:1d l

O described in WCAP-8567 (Reference 1). This procedure is known as the "laproved Thermal Design Procedure," and is discussed more fully in Section 4.4 of RESAR-SP/90 PDA Module 5, " Reactor System".

For accidents which are not DNB limited, or for which the Improved Thermal Design Procedure is not employed, initial conditions are obtained by adding the maximum steady state errors to rated values. The following conservative steady state errors were assumed in the analysis:

1. Core power 12% allowance for calorimetric error
2. Average reactor coolant 14'F allowance for controller deadband system temperature and measurement error h.
3. Pressurizar pressure 1 30 psi allowance for steady-state fluctuations and measurement error.

Table 15.0-3 summarizes initial conditions and computer codes used in the accident analysis, and shows which accidents employed a DNB analysis using the O Improved Thermal Design Procedure.

15.0.4.3 Power Distribution The limiting conditions occurring during reactor transients are dependent on the core power distribution. The design of the core and the control system minimizes adverse power distribution through the placement of control rods and operating methods. In addition, the core power distribution is continuously monitored by the integrated protection system as described in RESAR-SP/90 PDA Module 9, "l&C and Electric Power" and the Technical Specifications. Audible alarms will be activated in the control room whenever the power distribution exceeds the limits assumed as initial conditions for the transients presented in this chapter.

O For transients which may be DNB limited both the radial and axial peaking factors are of importance. The core thermal limits illustrated in Figure 15.0-1 are based on a reference axial power shape. The low DN8R reactor trip O WAPWR-CS 15.0-9 NOVEMBER, 1986 5723e:1d

O setpoint is automatically adjusted for axial shapes differing from the reference shape by the method described in Section 4.4 of RESAR-SP/90 PDA Module 5, " Reactor System" and also described in RESAR-SP/90 PDA Module 9, "I&C and Electric Power". The radial peaking factor F AH increases with decreasing power and with increasing rod insertion. The increase in FAH resulting from decreasing reactor power and increased rod insertion is accounted for in the low DNBR reactor trip through measurement of power and control rod position.

For transients which may be overpower limited, the total peaking factor F q is of importance. F q

is continuously monitored through the high Kw/ft reactor trip as described in RESAR-SP/90 PDA Module 9, "I&C and Electric Power" and the Technical Specifications to assure that the limiting overpower conditions are not exceeded.

For overpower transients which are slow with respect to the fuel rod thermal time constant, fuel rod thermal evaluations are determined as discussed in Section 4.4 of RESAR-SP/90 PDA Module 5, " Reactor System". Examples of this are the uncontrolled boron dilution incident, which lasts many minutes, and f

the excessive load increase incident, which reaches equilibrium witnout causing a reactor trip. For overpower transients which are fast with respect to the fuel rod thermal time constant (for example, the uncontrolled rod cluster control assembly bank withdrawal from suberitical and rod cluster I

control assembly ejection incidents, which result in a large power rise over a few seconds), a detailed fuel heat transfer calculation is performed.

Although the fuel rod thermal time constant is a function of system conditions, fuel burnup, and rod power, a typical value at beginning-of-life for high power rods is approximately 5 seconds.

I 15.0.5 Reactivity Coefficients Assumed in the Accident Analyses

The transient response of the reactor system is dependent on reactivity i' feodback effects, in particular the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are

! discussed in detail in RESAR-SP/90 PDA Module 5, " Reactor System".

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! WAPWR CS 5723e:1d 15.0 10 NOVEMBER, 1986 I

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O v 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 reactor coolant system, do not depend highly on reactivity feedback effects.

The values used for each accident are given in Table 15.0-3 of RESAR-SP/90 PDA Modules 4, 5, and 6/8. Reference is made in that table to Figure 15.0-2 which shows the upper and lower bound Doppler power coefficients as a function of q

V power, used in the transient analysis. The justification for use of con.crvatively large vs. small reactivity coefficient values are treated on an event-by-event basis. Conservative combinations of parameters are used for a given transient to bound the effects of core life, although these combinations i may not represent possible realistic situations.

I 15.0.6 Rod Cluster Control Assembly Insertion Characteristics '

The negative reactivity insertion following a reactor trip is a function of the position vs. time of the rod cluster control assemblies and the variation

- in rod worth as a function of rod position. With respect to accident analyses, the critical parameter is the time of insertion up to the dashpot entry or approximately 85% of the rod cluster travel. For all accidents the insertion time to dashpot entry is conservatively taken as [ ] seconds. The H a,c) normalized rod cluster control assembly position vs. time assumed in accident

analyses is shown in Figure 15.0-3.

Figure 15.0-4 shows the fraction of total negative reactivity insertion v's.

normalized rod position 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 an unbalanced xenon distribution.

j This curve is used to compute the negative reactivity insertion vs. time i following a reactor trip which is input to all point kinetics core models used in transient analyses. The bottom skewed power distribution itself is not an i

1 input into the point kinetics core model.

WAPWR-CS 15.0-11 NOVEMBER, 1986 5723e:1d l

A U There is inherent conservatism in the use of Figure 15.0-4 in that it is based on a skewed flux distribution which would exist relatively infrequently. For cases other than those associated with unbalanced xenon distributions, significant negative reactivity would have been inserted due to the more favorable axial distribution existing prior to trip.

The normalized rod cluster control assembly negative reactivity insertion vs.

time is shown in Figure 15.0-5. The curve shown in this figure was obtained k from Figures 15.0-3 and 15.0-4. A total negative reactivity insertion following a trip of ( ) Ap is assumed in the transient analyses except M**C) where specifically noted otherwise. This assumption is conservative with respect to the calculated trip reactivity worth available as shown in Section 4.3 of RESAR-SP/90 PDA Module 5, " Reactor System".

The normalized rod cluster control assembly negative reactivity insertion vs.

time curve for an axial power distribution skewed to the bottom (Figure 15.0-5) is used for those transient analyses for which a point kinetics core model is used. Where special analyses required use of three-dimensional or axial one-dimensional core models, the negative reactivity insertion resulting from the reactor trip is calculated directly by the reactor kinetics code and is not separable from the other reactivity feedback effects. In this case, the rod cluster control assembly position vs. time (Figure 15.0-3) is used as code input.

15.0.7 Trip Points and Time Delays to Trip Assumed in Accident Analyses A reactor trip signal acts to open eight trip breakers, two per channel set, feeding power to the control rod drive mechanisms. The loss of power to the mechanism coils causes the mechanisms to release the rod cluster control' assemblies 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 release of the rods O by the mechanism. The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the roos are free and begin to fall. Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in Table 15.0-4.

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WAPWR-CS 15.0 12 NOVEMBER, 1986 5723e:1d

A Reference is made in Table 15.0-4 to the low DNBR trips shown in Figure 15.0-1. These figures present the allowable reactor power as a function of the coolant loop inlet temperature and primary coolant pressure for N loop

/9 operation (4-loop operation), for the design flow and power distribution, as described in Section 4.4 of RESAR-SP/90 PDA Module 5, " Reactor System".

The boundaries of operation defined by the low DNBR trip are represented as

" protection lines" on this diagram. The protection lines are drawn to include

(]

V all adverse instrumentation and setpoint errors so that under nominal conditions trip would occur well within the area bounded by these lines. The DNB lines represent the locus of conditions for which the DNBR equals the

( ). All points below and to the left of a DNB line for a H a,c) given pressure have DNBR greater than the limit value with the assumed axial and radial power distributions. The diagram shows that the DNB design basis is not violated for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable DNBR line at any point.

The area of permissible operation (power, pressure and temperature) is bounded by the combination of reactor trips: high neutron flux (fixed setpoint); high pressure (fixedsetpoint);lowpressure(fixed setpcint); low DNBR (variable setpoint);highkw/ft(fixedsetpoint).

The limit value, which was used as the DNBR limit for all accidents analyzed withtheimprovedThermalDesignProcedure(seeTable15.0-3), is conservative compared to the actual design DNBR value required to meet the [sNB design basis is discussed in Section 4.4 of RESAR-SP/90 PDA Module 5, " Reactor System".

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( 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. Nominal trip setpoints are specified in the plant Technical Specifications. During plant startup tests, it is demonstrated that actual instrument time delays are equal to or less than the assumed values.

Additionally, protection system channels are calibrated and instrument response times determined periodically in accordance with the plant Technical Specifications.

O WAPWR CS 15.0-13 NOVEMBER, 1986 5723e:1d

O 15.0.8 Instrumentation Drift and Calorimetric Errors -

Power Range Neutron Flux O The instrumentation drift and calorimetric errors used in establishing the power range high neutron flux setpoint are presented in Table 15.0-5. The l calorimetric error is the error assumed in the determination of core thermal power as obtained from secondary plant measurements. The total ion chamber p current (sum of the multiple sections) is calibrated (set equal) to this V measured power on a periodic basis.

The secondary power is obtained from measurement of feedwater flow, feedwater i inlet temperature to the steam generators and steam pressure. High accuracy 1 instrumentation is provided for these measurements with accuracy tolerances much tighter than those which would be required to control feedwater flow.

t 15.0'.9 Plant Systems and Components Available for Mitigation of Accident Effects 4

j O The Westinghouse nuclear steam supply system (NSSS) is designed to afford power protection againts the possible effects of natural phenomena, postulated environmental conditions, and the dynamic effects of the postulated accident.

In addition, the design incorporates features which minimize the probability and effects of fires and explosions. Chapter 17.0 of the RESAR-SP/90 integrated PDA document will discuss the quality assurance program which is implemented to ensure that the plant will be designed, constructed, and operated without undue risk to the health and safety of the general public.

The incorporation of these features, coupled with the reliability of the design, ensures that the normally operating systems and components listed in Table 15.0-6 will be available for nitigation of the events discussed in Chapter 15. In determining which systems are necessary to mitigate the i

effects of these postulated events, the classification system of l ANSI-N18.2-1973 is utilized. The design of "systers important to safety" (including protection systems) is consistent with IEEE 379-1972 and Regulatory

Guide 1.53 in the application of the single failure criterion.

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I O WAPWR-CS 15.0-14 NOVEMBER, 1986 j 5723e:1d I

O In the analysis of the Chapter 15 events, the operation oi the non-safety-related rod control system, other than the reactor trip portion of the Control rod drive system (CRDS), is considered only if that action results in more O severe consequences. No credit is taken for control system operation if that b operation mitigates the results of an accident. For some accidents, the analysis is performed both with and without control system operation to determine the worst case. The pressurizer heaters are not assumed to be energized during any of the Chapter 15 events.

15.0.10 Fission Product Inventories 15.0.10.1 Inventory in the Core The time dependent fission product inventories in the reactor core are calculated by the ORIGEN code (10) using a data library based on ENDF/B-IV.(11) Core inventories are shown in Table 15.0-7.

The fission product radiation sources considered to be released from the fuel O to the containment following a maximum credible accident are based on the assumptions stated in TID-14844(3): 100 percent of the noble gases and 50 percent of the halogens.

15.0.10.2 Inventory in the Fuel Pellet Clad Gap The radiation sources associated with a gap activity release accident are based on the assumption that the fission products in the space between the fuel pellets and the cladding of all fuel rods in the core are released as a result of cladding failure.

The gap activities were determined using the model suggested in Regulatory Guide 1.25. Specifically,10 percent of the iodine and noble gas activity (except Kr-85, 1-127, and 1-129,whichare30 percent)isaccumulatedinthe d fuel clad gap. The gap activities are shown in Table 15.0-7.

O WAPWR-CS 15.0-15 NOVEMBER, 1986 5723e:1d

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O 15.0.10.3 - Inventory in the Reactor Coolant

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' Reactor coolant iodine concentrations for the Technical Specification limit of 1 pCi/gm of dose equivalent (D.E.) I-131 and for the assumed pre-accident i iodine spike concentration of 60 vCi/gm of D.E. 1-131 are presented in Table 15.0-8. Reactor coolant noble gas concentrations based- on 1 percent fuel j defects are presented in Table 15.0-9. Iodine appearance rates in the reactor i coolant, for normal steady state operation at 1 pCi/gm of D.E. I-131, and for an assumed accident initiated iodine spike are presented in Table 15.0-10.

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! 15.0.11 Residual Decay Heat 15.0.11.1 Total Residual Heat i

i i Residual heat in a suberitical core is calculated for the loss-of-coolant accident per the requirements of Appendix K, 10 CFR 50.46, as described in References 5 and 6. These requirements include assuming infinite irradiation f

time before the core goes suberitical to determine fission prtfuct decay O energy. For all other accidents, the same models are used except that fission f product decay energy is based on core average exposure at the end of the I equilibrium cycle.

15.0.12 Computer Codes Utilized l

Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular, very specialized codes in which  !

the modeling has been developed to simulate one given accident, such as those used in the analysis of the reactor coolant system pipe rupture (Section 15.6 of RESAR-SP/90 PDA Module 4 " Reactor Coolant System"), are summarized in

[ their respective accident analyses sections. The codes used in the analyses ,

l of each transient are listed in Table 15.0-3.

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O WAPWR-CS 15.0-16 NOVEMBER, 1986 5723e:1d i

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O 15.0.12.1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section h of a metal clad 00 2 fuel rod and the transient heat flux at the surface of the cladding using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temperature,anddensity). The codes uses a fuel model which exhibits the following features simultaneously:

1. A sufficiently large number of radial space increments to handle fast l

transients such as rod ejection accidents.

2. Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation.

l 4

3. The necessary calculations to handle post DNB transient: film boiling heat transfer correlations, Zircaloy-water reaction and partial melting of I

the materials.

O- FACTRAN is further discussed in Reference 7.

15.0.12.2 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 multiloop system by a model containing reactor vessel, hot and cold leg piping, steam generators (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 l

effects of the moderator, fuel, boron and rods are included. 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 l flux, low DNBR, high linear power (kW/ft), high and 104 pressure, low flow, i and high pressurizer level. Control systems are also simulated including rod i

O WAPWR-CS 15.0-17 NOVEMBER,1986 5723e:1d I

O control, steam dump, feedwater control and pressurizer pressure control.

ECCS, 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.

I LOFTRAN also has the capability of calculating the transient value of DNBR p based on the input from the core limits illustrated in Figure 15.0-1. The V core limits represent the minimum value of DNBR as calculated for typical, small thimble, large thimble, corner or side cell.

LOFTRAN is further discussed in Reference 8.

15.0.12.3 TWINKLE The TWINKLE program is a 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 transient neutron diffusion equations in one, two or three dimensions. The code uses six delayed neutron groups and contains a detailed multi-region fuel-clad-coolant 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 functions such as inlet temperature, pressure, flow, boron concentration, control rod motion. Various edits are provided, e.g.,

channelwise power, axial offset, enthalpy, volumetric surge, pointwise power,

( and fuel temperatures.

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 9.

O WAPWR-CS 15.0-18 NOVEMBER, 1986 l 5723e:1d L

O 15.0.12.4 THINC The THINC Code is described in Section 4.4 of RESAR-SP/90 PDA Nodule 5, -

" Reactor System".

15.0.13 REFERENCES

1. H. Chelemer, et al., " Improved Thermal Design- Procedure", WCAP-8567-P O, (Proprietary), July 1975,andWCAP-8568(Non-Proprietary) July,1975.
2. J. Skaritka, ed., " Hybrid BC 4 Absorber Control Rod Evaluation Report".

WCAP-8846-A, October 1977.

3. J. J. DiNunno et al., " Calculation of Distance Factors for Power and Test Reactor Sites", TID-14844, March 1962.
4. M. E. Meek and B. R. Rider, " Compilation of Fission Product Yields",

NED0-12154-1, General Electric Corporation, January 1974.

1

5. F. M. Bordelon et al., " SATAN-VI Program: Comprehensive Space Time l

! Dependent Analysis of Loss-of-Coolant", WCAP-8306, June 19744

6. F. M. Bordelon et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis", WCAP-8305, June 1974.
7. C. Hunin, "FACTRAN, A FORTRAN IV Code for Thermal Transients in a UO2 Fuel Red", WCAP-7908, June 1972.

O 8. T. W. T. Burnett et al., "LOFTRAN Code Description", WCAP-7907-P-A, April, i 1984.

D. H. Risher, Jr., and R. F. Barry, " TWINKLE - A Multi-Dimensional Neutron O 9.

Kinetics Computer Code": WCAP-7979-P-A (Proprietary) January 1975, and WCAP-8028-A, (Non-Proprietary), January 1975.

O WAPWR-CS 15.0-19 NOVEMBER, 1986

5723e:1d

i l'

l l

l l 10. Bell, M. J., "0RIGEN - The ORNL Isotope Generation and Depletten Code,"

i ORNL-4628, May 1973.  !

i j 11. "0RIGEN Yields and Cross Sections - Nuclear Transmutation and Decay Data From END F/B-IV", Radiation Shielding Information Center. Oak Ridge j j' National Laboratory, RSIC-DLC-38, September 1975. l i

i

. O i 4

k I

i i

i t

i O  ;

l I

i r i

t E

f O  :

O O WAPWR-CS 15.0-20 NOVEP.BER, 1986 5723e:1d I

_ _ _ _ _ _ _ _ , _ . . _ . . . _ _ _ _ . , _ , ~ , , . . . . _ _

O TABLE 15.0-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS O

N-Loop Operation Reactor core thermal power output (MWt)* 3800 Thermal power generated by the reactor coolant 16 pumps (MWt)

Guaranteed nuclear steam supply system thermal 3816 power output (MWt)

  • Radiological consequences based on 3565 (MWt) power level.

O O

O WAPWR-CS 15.0-21 NOVEMBER, 1986 5723e:1d I i

TABLE 15.0-2 VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES

  • O N-Loop Operation Thermal output of nuclear steam supply system (MWt) 3816

- - + (a , c)

Reactor core thermal power output (MWt) ,

Core inlet temperature (*F)

Reactor coolant average temperature (*F)

Reactor coolant system pressure (psia)

Reactor coolant flow per loop (gpm) 0 Total reactor coolant flow (10 lb/hr) 0 Total steam flow from NSSS (10 lb/hr)

Steam pressure at steam generator outlet (psia)

O Maximum steam moisture content (%)

Feedwater temperature at steam generator inlet (*F)

Average core heat flux (Btu /hr-ft2) 1

  • For accident analyses using the improved thermal design procedure.

O 1

0 -

O WAPWR-CS 15.0-22 NOVEMBER, 1986 5723e:1d

O TABLE 15.0-2a VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES

  • O N-Loop Operation Thermal output of nuclear steam supply system (MWt) 3816

- - +(a,c)

Reactor core thermal power output (MWt)

Core inlet temperature (*F)

Reactor coolant average temperature ('F)

Reactor coolant system pressure (psia)

Reactor coolant flow per loop (gpm) 6 Total reactor coolant flow (10 lb/hr) 0 Total steam flow from NSSS (10 lb/hr)

Steam pressure at steam generator outlet (psia)

Maximum steam moisture content (%)

Feedwater temperature at steam generator inlet (*F) 2 Averagecoreheatflux(Btu /hr-ft)

  • For accident analyses not using the improved thermal design procedure.

0 lO WAPWR-CS 15.0-23 NOVEMBER, 1986 5723e:1d l

i

g s  %  %  %

) \ < \ \

J N Y Y TABLE 15.0-3 Supe 8ARY OF INITIAL CCNDITIONS AND COMPUTER CODES USED Kinetic Parameters Assumed Initial Improved NS$$ Reactor Vessel Press, i Computer Delayed Moder. Thermal Thermal Vessel Average Press. Water Feedwater

Codes Neutron Density DNB Design Power Coolant Temp. Pressure volume Temp.

Faults Utilized Fraction ( Ap/qu/cc) Doppier Correistion Proced. output (Mwt) Fiov (ope) (*F) (pois) (ft3) ('F) 15.1 Increase in Heat Removal by the Secondary System Feedwater System Mai- (See RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System")

function Causing an Increase in Feedwater Flow Excessive increase in (See RESAR-SP/90 PDA Module 6/8,

  • Secondary Side Safeguards System / Steam and Power Conversion System *)

Secondary Steam Flow Accidental Depressur- (See RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System")

tration of the Main Steam System Steam System Piping (See RESAR-SP/90 PDA Module 6/8, " Secondary Stoe Safeguards System / Steam and Power conversion System *)

Failure 15.2 Decrease in Heat Removal by the Secondary System Loss of External (See RESAR-SP/90 PDA Module 6/8, *$econdary Side Safeguards System / Steam and Power conversion System")

Electrical Load i

and/or Turbine Trip i

Loss on Non-Emergency (See RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System")

A-C Power to the Sta-tion Auxiliaries Loss of Normal Feed- (See RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System *)

water Flow Feedwater System Pipe (See RESAR-SP/90 PDA Modulo 6/8,

  • Secondary Side Safeguards System / Steam and Power Conversion System")

Break WAPWR-CS 15.0-24 MIVEM8ER, 1986 5723e:1d

1 A /

TABLE 15.0-3 (Con't)

, Kinetic Parameters Assumed

! Initial Improved NSSS Reactor vessel Press.

I Computer Delayed moder. Thermal T hermal Vessel Average Press. Water Feedwater Codes Neutron Density DN8 Design Power Coolant Temp. Pressure Volume Temp.

Faults Utilized Fraction (Ap/qu/cc) Doppler Correlation Proced. Dutput (MWt) Flow (Opm) ('F) (psia) (ft3) , ('F)

, 15.3 Decrease in i Reactor Coolant System Flow Rate l Partial and Complete (See RESAR-SP/90 PDA Module 4 ' Reactor Coolant System")

Loss of Forced Reactor j Coolant Flow I Reactor Coolant Pump (See RESAR-SP/90 PDA Module 4 ' Reactor Coolant System')

I Shaf t Seizure (locked rotor

]

4 1 15.4 Reactivity and j power distribution s anomalies I

Uncontrolled Rod (See RESAR-SP/90 PDA Module 5. " Reactor System *)
Cluster Control

! Assembly Bank j Wi t hdrawal f rom a j Subcritical or Low l

Power Startup Condition.

Uncontrolled Rod (See RESAR-SP/90 PDA Module 5. " Reactor Systema )

! Cluster Assembly

. Bank Withdrawal l at Power i

Control Rod Mis- (See RESAR-SP/90 PDA Modulo 5

  • Reactor System")

alignment Startup of an (See RESAR-SP/90 PDA Module 4, ' Reactor Coolant System *)

Inactive Reactor

{ Coolant Loop at an j Incorrect Temperature

  • Inadvertent Loading (See RESAR-SP/90 PDA Module 4 " Reactor Coolant System) and Operation of a Fuel Assembly in an j Improper Position A

i I NOVEM8ER, 1996 WAPWR-CS 15.O-25 f 5723e:1d i

1

1 a

  • av ,  %

l i

! TABLE 15.0-3 (Con't,)

Kinette Parar.ctars AamuWid 4

Initial Improved NSSS Reactor Vessel Press.

Computer Delsyed Moder. T hermal Thermal Vessel Average Press. Water Feedwater Codes Neutron Density DNB Design Power Coolant Temp. Pressure Volume Temp.

Faults Ut11tred Fraction (Ap/qu/ce) Doppler Correlation Proced. Output (MWt) Flow (gpe) ('F) (psia) (ft3) (*F)

Spectr u of Rod (See RESAR-SP/90 PDA Module 4 " Reactor Coolant System) 1 Cluster Control

! Assembly Ejection j Accidents j 15.5 Increase in i Coo 1 Ant Inventory I

f Inadve tT-t Operetton NA NA NA NA NA NA NA NA NA NA NA of ECCS During Power l j Dperation J

I 15.6 Decrease in Reactor Coolant Inventory s

i Inadvertene AWT.-4 e +(See RESAR-SP/90 PDA Module 4 ' Reactor Coolant System)

! of a Pressurizer Safety or Relief Valve d

}

l l

i l

l l

t i

! i NA - Not applicable.

1 1

l WAPWR-CS 15.0-26 NDVEMBER. 1996 j 5723e:1d l

I l

i I . . _ _ _ _ _ _ _

TABLE 15.0-4 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Limiting Trip Point Assumed Time Delays Trip Function In Analysis (sec)

Power range high neutron flux, f high setting 4

Power Range high neutron flux, low setting Power range neutron flux, high negative rate High neutron flux, P-8 Low DNBR High pressurizer pressure Low pressurizer pressure Low reactor coolant flow (fromloopflowdetectors)

RCP underspeed Turbine trip Safety injection reactor trip Low steam generator level High steam generator level - -

i produces feedwater isolation and turbine trip

  • See RESAR-SP/90 PDA Module 5, " Reactor System"
    • Total time delay (including RTD time response and trip circuit channel electronics delay) from the time the temperature in the coolant loops exceeds the trip setpoint until the rods are free to fall.
      • See RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion").

WAPWR-CS 15.0-27 NOVEMBER, 1986 5723e:1d

( TABLE 15.0-5 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT -

POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRUMENT ERRORS O Effect on Accuracy of Thermal Power Measurement Determination of Variable (% error)

Variable (% error) (Estimated) (Assumed)

Calorimetric errors in the measurement of secondary system thermal power: .

Feedwater temperature Feedwater pressure (small ccrrection on enthalpy)

Steam pressure (small correction on enthalpy Feedwater flow Assumed calorimetric error

(% of rated power)

Axial power distribution effects on total ion chamber current Estimated error

(% rated power)

Assumed error

(% of rated power)

Instrumentation channel drif t and setpoint reproducibility Estimated error

(% or rated power)

Assumed error

(% of rated power)

Total assumed error in setpoint (a) + (b) + (c)

WAPWR-CS 15.0-28 NOVEMBER, 1986 5723e:1d

I-4 Y

TABLE 15.0-5(Con't) .

2 Percent Rated Power

+(a, c)

Nominal Setpoint

' Maximum overpower-trip point assuming all individus1 errors are l

simultaneously in the most adverse -

i direction.

1 i

i 4

?

i .

I i

1 l

I l

O l

l WAPWR-CS 15.0-29 NOVEMBER, 1986 5723e:1d ,

r

O O O O O O O i

TABLE 15.0-6 (Sheet 1 of 4)

PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS

. Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 15.1 Increase in Heat Removed by the Secondary System Feedwater System Power range high flux, High steam generator Feedwater isolation NA Mm1' unction Causing high steam generator level produced feedwater valves

, an increase in Feed- level, manual, low DNBR, isolation and turbine

! water Flow high kw/ft trip Excessive Increase Power range high flux, NA Pressurizer self- NA Secondary Steam Flow manual, low DNBR, high actuated safety valves; i kw/ft steam generator safety valves

Accidental Depres- Low pressurizer low pressurizer pressure, Feedwater isolation Auxiliary feed i surization of the pressure, manual, SIS low compensated steam valves, steamline stop System Safety i Main Steam System line pressure, Hi-1 valves Injection System containment pressure, j manual, low 4 T cold l Steam System Piping SIS, low pressurizer low pressurizer pressure, Feedwater isolation Auxiliary feed Failure pressure, manual low compensated steam- valves, steamline stop system; Safety

! line pressure, Hi-1 valves Injection System

! containment pressure, j manual, low 4 T cold i

1 l WAPWR-CS 15.0-30 NOVEMBER, 1986 5723e:Id 1

I i

O O O O O O O l

l TABLE 15.0-6 (Sheet 2 of 4)

Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 15.2 Decrease in Heat Removal by the Secondary System Loss of External High pressurizer Low steam generator Pressurizer safety Auxiliary feed j Electrical Load / pressure, low DNBR, level valves, steam generator system

Turbine Trip low steam generator I level, manual l' Loss of Non-Emergency Steam generator low Steam generator low Steam generator safety Auxiliary feed A-C Power to the level, manual level valves system Station Auxiliaries Loss of Normal Steam generator low Steam generator low Steam generator safety Auxiliary feed l Feedwater Flow level, manual level valves system 4

l Feedwater System Pipe Steam generator low Hi-1 containment Steamline isolation Auxiliary feed i

Break level, high pressurizer pressure, steam generator valves, feedline isol- system, Safety i pressure, SIS, manual low level, low compen- ation, pressurizer Injection System low DNBR sated steamline pressure safety valves steam generator safety valves

! 15.3 Decrease in Reactor Coolant System Flow Rate

! Partial and Complete Low flow, low RCP NA Steam generator NA

! Loss of Forced Reactor speed, manaul safety valves I Coolant Flow 1

l R: actor Coolant Pump Low flow, manual NA Pressurizer safety NA l Shaft Seizure (Locked valves, steam generator l Rotor) safety valves.

'l 3 WAPWR-CS 15.0-31 NOVEISER, 1986 l 5723c:1d

O O O O O O O TABLE 15.0-6 (Sheet 3 of 4) j Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 15.4 Reactivity and Power Distribution '

Anomalies Uncontrolled Rod Power range high flux NA NA NA Clutter Control (low s.p.), manual Assembly Bank Withdrawal from a

, Subcritical or low

! Power Startup Condition l Uncontrolled Rod Power range high NA Pressurizer safety NA Cluster Control flux, Hi pressurizer valves, steam generator i Assembly Bank pressure, manual, low safety valves Withdrawal at Power DNBR i

Control Rod Nisalign- Power range negative NA NA NA

) ment flux rate, manual

] Startup of an Inactive Power range high flux, NA NA NA

.i R: actor Coolant Loop P-8, manual l at an Incorrect ,

Temperature i

]

i j Chemical and Volume Source range high flux, NA low insertion limit NA J

Control System Mal- power range high flux, annunciators for bora-

! function that Results manual, low DNBR, high tion, VCT outlet isol-

! in a Decrease in kw/ft ation valves B:r<.n Concentration in the Reactor Coolant Spectrum of Rod Cluster Power range high flux, NA NA NA Control Assembly high positive flux rate, '

Ejection Accidents manual i

WAPWR-CS 15.0-32 NOVEMER, 1986

.i 5723c:1d

f O O O O O O O i

i j TABLE 15.0-6 (Sheet 4 of 4)

Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 15.5 Increase in Reacter Coolant i Inventory I

Inadvertent NA NA NA NA

! Operation of ECCS During Power Operation

15.6 De'rease c in Reactor j Coolant Inventory l' Inadvertent Opening of Pressurizer low Low pressurizer pressure NA Safety Injection -

a Pressurizer Safety pressure, manual, low System i or Relief Valve DNBR Steam Generator Tube Reactor trip system Engineered Safety Service Water System, Emergency Core ,

l Rupture Features Actuation. Component Cooling Water Cooling System, l System System, steam generator Auxiliary Feedwater

) safety valves, steam- System, Emergency ,

! line stop valves Power Systems Loss of Coolant Reactor trip system Engineered Safety Service Water System, Emergency Core

Accident from Spectrum Features Actuation System Component Cooling Water Cooling System, of Postulated Piping System, Steam Generator Auxiliary Feedwater Breaks within the Safety Valves System Containment System Heat Removal System, j Emergency Power System i

l WAPWR-CS 15.0-33 NOVEMBER, 1986-5723e:1d 4

TABLE 15.0-7 FUEL AND ROD GAP INVENTORIES, CORE (Ci)(a)

Core Isotope Fuel G_ag(b)

I-131 1.0E + 7 1.0E + 6

1-132 1.5E + 8 1.5E + 7 I-133 2.1E + 8 2.1E + 7

' I-134 2.3E + 8 2.3E + 7 I-135 2.0E + 8 2.0E + 7 i Kr-83m 1.3E + 7 1.3E + 6 Kr-85m 2.9E + 7 2.9E + 6 Kr-85 7.0E + 5 2.1E + 5 Kr-87 5.2E + 7 5.2E + 6 Kr-88 7.5E + 7 7.5E + 6 Kr-89 9.3E + 7 9.3E + 6 Xe-131m 7.5E + 5 7.5E + 4 O Xe-133m 3.1E + 7 2.0E + 8 3.1E + 6 2.0E + 7 Xe-133

, Xe-135m 4.3E + 7 4.3E + 6 l

Xe-135 4.5E + 7 4.5E + 6 l Xe-138 1.7E + 8 '1.7E + 7 l I-127 3.0 kg 0.90 kg I-129 12.2 kg 3.7 kg i

O -

a. Three-region equilibrium cycle core at end of life. The three regions have operated at a specific power of 40.03 MWt per metric ton of uranium for 300, 600, and 900 effective full power days, respectively.

! b. Gap activity is assumed to be 10 percent of core activity for all isotopes except Kr-85, I-127, and 1-129, whose gap activities are assumed to be 30 percent of their core activities (Regulatory Guide 1.25 assumption).

l l

l O WAPWR-CS 5723e:1d 15.0-34 NOVEMBER, 1986 l

l- -- - _ _ _

j l

f 1

TABLE 15.0-8 REACTOR COOLANT IODINE CONCENTRATIONS FOR i

1 vCi/ GRAM AND 60 vCI/ GRAM OF DOSE EQUIVALENT I-131 lO i,

Reactor Coolant Concentration (Ci/gm)

\

Nuclide 1 uti/gm D.E. 1-131 60 v Ci/gm D.E. 1-131 l l

lO

,, I-131 0.76 45.6 i 1-132 0.76 45.6 I-133 1.14 68.4- ,

j I-134 0.195 11.7 j I-135 0.63 37.8 i

l O.

i (O

i t

I

!O I

1 WAPWR-CS 15.0-35 NOVEMBER, 1986 5723e:1d I

i

-- - - - . , - - _ _ - - . ~ . _

O TABLE 15.0-9 REACTOR COOLANT NOBLE GAS SPECIFIC ACTIVITY BASED ON ONE PERCENT DEFECTIVE FUEL O

Nuclide Activity (uc/aram)

Kr-85m 2.0 7.3 Kr-85 Kr-87 1.3 Kr-88 3.6 Xe-131m 2.2 Xe-133m 1.7 x 10 1 2

Xe-133 2.7 x 10 Xe-135m 4.8 x 10 -1 Xe-135 7.2 O Xe-138 6.4 x 10 -1 O

O O WAPWR-CS 15.0-36 NOVEMBER, 1986 5723e:1d ,

1 TABLE 15.0-10 IDDINE APPEARANCE RATES IN THE REACTOR COOLANT (Curies /sec)

    • Appearance Rates l ,

'

  • Equilibrium Appearance Due to an Accident Rates Due to Fuel Defects Initiated Iodine Spike i

1-131 3.4 x 10'3 1.7 ,

9.0 i I-132 1.8 x 10 1-133 7.2 x 10 3.6

-2 5.5 1-134 1.1 x 10

! I-135 6.8 x 10 -3 3.4 I

'l

  • Based on RCS concentration of 1 vCi/gm of dose equivalent I-131.
    • 500 x equilibrium appearance rate.

i i <

i O

l I

lO 4

l l

!O WAPWR-CS 15.0-37 NOVEMBER, 1986  !

! 5723e:1d I

i i E . . _ _ _ _ _ _ _ _ _ . - _ ___ _ _

_ _ . . - . .m . _____ _ _ . __ _ _ _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ - _ . _ . . _ . _ _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ _

4 1

i

, O I

O sac -

7 1

Pressuriser Pressure (pata)

, 440 < s 24o

} g

N see - 's N 's N 1990 4

N N 1

N s N N g j \

s N i g N g l sto - uss N N

N N

's s I w

i j s 's N 's s N j y '

s N s '%

! , soo -. s s, s 's i

- s 's % s,

% N  %

8  %, N '

N N \ \

{ N N \ 'g I N N \ s l

$40 i \ g \ \ l i N N \ h

\ \ \

g h s s i

N '

ses .. \  !

s \t t' s N

5 540- ',  !

I j

-=-

J sze c ro ao so a: ioc 12:

i l 4

core e e a er =e- m u l l0 i

! Figure 15.0-1 Illustration of Core Thermal Limits and DNB Protection (N Loop Operation)  ;

O NOVEMBER, 1985 WAPWR-CS 5723e:1d [

1  ;

1 I

__-_~e, . , < - -

l l l i i O ..

1 I

.I E NOTE 1 i

l .6 i

3 .6 .

l  !

W l

3 .to . .

i .u .

i 1

-18 '

NOTE 2 1

4 i

l 16 I

i 60 80 100 0 20 A0 i

m ta Ltytt (s or n'tt e tt)

Note 1. Upper Curve, Least Negative Doppler Only Power i

Defect . 6.95% sp (0 to 100: Power)

Note 2 . Lower Curve. Most Negative Doppler Only Power j i

i Defect . 1.49: :p (0 to 100 Power) l ,.

Figure 15.0-2 Doppler Power Coefficient Used in Accident Analysis 1 O WAPWR-CS NOVEMBER, 1956 5723e:1d  !

--re-w--.-. -ew-+-y-u.%. - _ -

l 4

1 2

f .

l I

^

1 i t I

!Oi i

f 1.2 .

I 1.1, qFully j Inserted j

! 1.0 Top of Desh ct j -

0.9 a

4 0.s- , e i 2 -

l ge0.7< I ,

y.*50.6 .

, . l l

TE E 0.5 a 0,4, i

= I l l

l I L  ; 0.3<. I I

E.

l j O.2' l

l 0.1- l l

l l t 1 0

! 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5 homeli2ed 0700 Time (Time Af ter Orop legins/ Deep ttae to Top of Desnoct) i O .

Figure 15.0-3 RCCA P0sition Vs. Time to Dashp0t t i

NOVEMBER, 1985 WAPWR-CS i 5723e:1d I

i I

l l

1 O

i i

O 1.0 0.9 -

i

! 0.8 -

)

i

, 0.7 5 -

! 3: 0.6 E

3::

W 0.5 -

W ,

O, g t 0.4 0.3 -

0.2 -

0.1

'

  • I I 0

0.4 0.6 0.8 1.0 0 0.2 I

l O 200PCSITION(FRACil0MINSERTD)

O Figure 15.0-4 Normalized RCCA Reactivity Worth Vs. Fraction Insertion NOVEMBER, 1985 i WAPWR-CS 5723e:1d

- - - ,--- - . - . . , . , , _ . - , , - , . - - - - __w, - . . - - - , .- -.

O O

1.0-O .

.9 "

.8 "

S6

  • 7 b

T .6 "

3 g .5' O s W .,4-

}

I*3<

E, a

.2-

.1 O O .1 .5 .3 4 .5 .'6 .7 .8 .9 1.0 Normalized Drop Tim Tire After Drop Begins 1.1 1.2 Tirne to Top of Dashpot O

Figure 15.0-5 Normalized RCCA Bank Reactivity Worth Vs. Normalized Drop Time

, NOVEMBER,1985 5723e:1d

..

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