ML20099J150
| ML20099J150 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 11/30/1984 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19269B210 | List:
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| References | |
| NUDOCS 8503190607 | |
| Download: ML20099J150 (42) | |
Text
Q 15.0 ACCIDENT ANALYSES 15.0.1 Gener11_
fV 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 consnent, 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 f rom the reactor coolant pressure boundary in the NSSS PWR design penetrate the Containment.
(For the definition of the Reactor Coolant System boundary, refer to Section 5,
" Nuclear Safety Criteria for the Design of Stationary PWR Plants," 1973.)
15.0.2 Classification of Plant conditions Since 1970 the ANS classification of plant conditions has been used to divide plant conditions into f,our categories in accordance with anticipated f requency 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:
Limiting Faults.
O O
WAPWR-!&C/EP 15.0-1 NOVEMBER, 1984 2218e:1d 850319g6g 80 601 PDR A
PDR K
. _. - - _ ~ - -. _ -. _.
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.
15.0.2.1 Condition I - Normal Operation and Operational Transients Condition I occurrences are those which are expected f requently or regularly in the course of normal plant operation, ref ueling, 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 f rom the point of view of af fecting the consequences of f ault conditions (Conditions II, III and IV).
In this regard, analysis of each f ault 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.
~4.1e 1 - Power operation (> 5 to 100 percent of rated thermal power).
O b.
Mode 2 - Startup (K,ffl0.99, 5 5 percent of rated thermal power).
Mode 3 - Hot standby (K,ff < 0.99, T 1 350'F).
c.
avg O
d.
Mode 4 - Hot shutdown (K,ff < 0.99, 200*F 1 T 1 350*F).
AVG WAPER-!&C/EP 15.0-2 NOVEMBER, 1984 2211e:1d
d e.
Mode 5 - Cold Shutdown (K,ff < 0.99, T,yg < 200*F).
f.
Mode 6 - Refueling (K,9f 1 0.95, T,yg 1 140*F).
2.
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:
a.
Operation with components or systems out of service (such as power operation with a reactor coolant pump out of service).
b.
Radioactivity in the reactor coolant, due to leakage from fuel with cladding defects and other sources.
- 1) Fission products O'
- 2) Corrosion products
- 3) Tritium c.
Operation with steam generator primary-to-secondary leakage up to the maximum allowed by the Technical Specifications, d.
Testing as required by the Technical Specifications.
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 for the pressurizer during heatup).
O b.
Step load changes (up to 10 percent).
c.
Ramp load changes (up to 5 percent / minute).
O WAPWR-I&C/EP 15.0-3 NOVEMBER, 1984 2218e:1d
d.
Load rejection up to and including design full load rejection transient.
15.0.2.2 Cor!! tion II - Faults of Moderate Frequency At worst, a Condition 11 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 II events are not expected to result in fuel rod f ailure 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 Safeguards System / Stear and Power Conversion System").
2.
Feedwater system nalfunctions causing an increase in feedwater flow (Subsection 15.1.2 of RESAR-SP/90 PDA Module 6/8,
" Secondary Side Safeguards System / Steam and Power Conversion System").
l 3.
Excessive increase in secondary steam flow (Subsection 15.1.3 of RESAR-SP/90 PDA M6dule 6/0, " Secondary Side Saf eguards System / Steam and l
Power Conversion System").
l 4.
Inadvertent opening of a steam generator relief or safety valve causing a l
depressurization of the main steam system (Subsection 13.1.4 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System").
i l
l S.
Loss of external load (Subsection 15.2.2 of RESAR-SP/90 PDA Module 6/8,
" Secondary Side Safeguards System / Steam and Power Conversion System").
t 6.
Turbine trip (Subsection 15.2.3 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System").
HAPWR-I&C/EP 15.0-4 NOVEMBER, 1984 2218e:1d
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 Power Conversion System").
8.
Loss of condenser vacuum and other events esulting in turbine trip (Subsection 15.2.5 of RESAR-SP/90 PDA Module 6/8,
" Secondary Side Safeguards System / Steam and Power Conversion System").
9.
Loss of no'1 emergency A-C power to the statior, auxiliar ies (Subsection 15.2.6 of RESAR-SP/90 PDA Module 6/8,
" Secondary Side Safeguards System / Steam and Power Conversion System").
- 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 System").
- 11. Partial loss of forced reactor coolant flow (Subsection 15.3.1 of RESAR-SP/90 PDA Module 4. " Reactor Coolant System").
O
- 12. Uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power startup condition (Subsection 15.4.1 of RESAR-SP/90 PDA Module 5, " Reactor System").
- 13. Uncontrolled rod cluster control assembly bank withdrawal at power l
(Subsectica 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 O
(Subsection 15.4.3 of this module).
- 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").
O
- 16. Chemical and volume control system nelfunction that results in a decrease in the boron concentration in the reactor coolant (Subsection 15.4.6 of RESAR-SP/90 PDA Module 13. " Auxiliary Systems").
O WAPWR-I&C/EP 15.0-5 NOVEMBER, 1984 2218e:1d
r
- 17. Inadvertent operation of emergency core cooling system during power operation (Subsection 15.5.1 of RESAR-SP/90 PDA Module 1
" Primary Side Safeguards System").
- 18. Chemical and volume control system malfunction that increases reactor coolant inventory (Subsection 15.5.2 of RESAR-SP/90 PDA Module 13
" Auxiliary Systems").
- 19. Inadvertent opening of a pressurizer safety or relief valve (Subsection 15.6.1 of RESAR-SP/90 PCA Module 4, " Reactor Coolant System").
- 20. "ailure of small lines carrying primary coolant outside containment l
(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 f aults which may occur very infrequently during the life of the plant. They will be accommodated with the failure of only a small fraction of the fuel rods although suf ficient fuel damage might occur to preclude immediate resumption o' the operation.
The l
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 s 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 System").
2.
Complete loss of forced reactor coolant flow (Subsection 15.3.2 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System").
O WAPWR-I&C/EP 15.0-6 NOVEMBER, 1984 2218e:1d i
3.
Control rod misalignment - Single rod cluster control assembly withdrawal l
x at - full power (Subsection 15.4.3 of RESAR-SP/90 PDA Module 5
" Reactor System").
4.
Inadvertent loading and operation of a fuel essembly in an improper position (Subsection 15.4.7 of RESAR-SP/90 PDA Module 5, " Reactor System").
5.
Loss of reactor coolant f rom small ruptured pipes or f rom 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").
6.
Waste gas system failure (Subsection 15.7.1 of RESAR-SP/90 PDA Module 12
" Waste Management").
7.
Radioactive liquid waste system leak or failure (atmospheric release)
(Subsection 15.7.2 of RESAR-SP/90 PDA Module 12. " Waste Management").
8.
Liquid containing tank failure (Subsection 15.7.3 of RESAR-SP/90 PDA O
Module 12. " Waste Management").
v 15.0.2.4 Condition IV - Limiting Faults Condition IV occurrences are f aults 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 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, " Secondary Side Safeguards System / Steam and Power Conversior. System").
O WAPWR-I&C/EP 15.0-7 NOVEMBER, 1984 2218e:1d
. ~
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 System").
3.
Reactor coolant pump rotor seizure (locked rotor) (Subsection 15.3.3 of RESAR-SP/90 PDA Module 4. " Reactor Coolant System").
4.
Reactor coolant pump shaft break (Subsection 15.3.4 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System").
5.
Spectrum of rod cluster control assembly ejection accidents (Subsection 15.4.8 of RESAR-SP/90 PDA Module 5, " Reactor System").
6.
Steam generator tube failure (Subsection 15.6.3 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System").
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").
8.
Fuel handling accident (Subsection 15.7.4 of RESAR-SP/90 PDA Module 12,
' Waste Management").
15.0.3 Ootimization 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 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 dif ferent, compromises based on the optimum overall performance are made and verified.
A consistent set of control system parameters is derived satisfying plant operational requireme.ts 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.
WAPWR-!&C/EP 15.0-8 NOVEMBER, 1984 2218e:ld
O 15.0.4 Plant Characteristics and Initial Conditions Assumed in the Accident Analyses 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 O
is consistent with the license application rating described in Chapter 1.0.
Allowances for errors in the deterrination of the steady-state power level are made as described in Subsection 15.0.4.2.
The values of pertinent plant 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 15.0-3.
15.0.4.2 Initial conditions For most accidents which are DNB limited, nominal values of initial conditions are assumed..The allowances on power, temperature, and pressure noted above are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-8567 (Reference 1).
This procedure is known as the
" Improved 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 DNS 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 O
steady state errors were assumed in the analysis:
1.
Core power i 25 allowance for calorimetric error 2.
Average reactor coolant i 4"F allowance for controller deadband system temperature and measurement error 3.
Pressurizer pressure 1 30 psi allowance for steady-state fluctuations and measurement error.
O EAPWR-I&C/EP 15.0-9 NOVEMBER, 1984 2218e:1d u
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 Improved Thermal Design Procedure.
15.0.4.3 Power Distribution The limiting conditions occurring during reactor transients are dependent on the core powar 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 Chapter 7 of this module 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.
For transients which may be DNB limited both the radial and axial peaking f actors are of importance.
The core thermal limits illustrated in Figure 15.0-1 are based on a reference axial power shape. The low DNBR reactor trip 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 Chapter 7 of this module.
The radial peaking factor F increases with decreasing power and with AH increasing rod insertion.
The increase in F resulting from decreasing AH 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 f actor F is of importance.
F is continuously monitored through the high Kw/ft q
reactor trip as described in Chapter 7 of this module and tit sechnical 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 O
WAPWR-I&C/EP 15.0-10 NOVEMBER, 1984 2218e:1d I
l V
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
.the excessive load increase incident, which reaches equilibrium without 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 subcritical and rod cluster 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 condi-tions, fuel burnup, and rod power, a tyoical value at beginning-of-life for high power rods is approximately 5 seconds.
15.0.5 Reactivity Coefficients Assumed in the Accident Analyses The transient response of the reactor system is dependent on reactivity feedback ef fects, in particular the moderator temperature coef ficient and the Doppler power coefficient. These reactivity coefficients and their values are O.
discussed in detail i. this module.
In the analysis of certain events, conservatism ' requires the use of large reactiv;ty 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 I.
reactor coolant system, do not depend highly on reactivity feedback effects.
j The values used fcr each accident are given in Table 15.0-3.
Reference is nude in that table to Figure 15.0-2 which shows the upper and lower bound Doppler power coefficients as a function of power, used in the tre.sient analysis.
The justification for use of conservatively large vs. small reactivity coefficient values are treated on an event-by-event basis.
Conservative combinations of parameters are used for a ghen transient to bound the ef fects of core life, although these combinations may not represent possible realistic situations.
- a i
MAPWR-I&C/EP 15.0-11 NOVEMBER, 1984 2218e:1d
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 i
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 3.4 seconds.
The normalized rod cluster control assembly position vs. tinie assumed in accident analyses is shown in Figure 15.0-3.
Figure 15.0-4 shows the fraction of total negative reactivity insertion vs.
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.
This curve is used to compute the negative reactivity insertion vs. time 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 input into the point kinetics core model.
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 inf requently.
For cases other than those associated with unbalanced xenon distributions, significant negative r.eactivity 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 from Fqures 15.0-3 and 15.0-4.
A total negative reactivity insertion following a trip of 4% Ap is assumed in the transient analyses except 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".
O WAPWR-I&C/EP 15.0-12 NOVEMBER, 1984 2218e:ld
The r.armalized rod cluster control assembly negative reactivity insertion vs.
time :urve for en axial power distribution skewed to the bottom (Figure 15.0-5) is used for those transient analyses for which a point kinetics core A
model is used.
Where special analyses required use of three-dimensional or
(,,/
axial one-dimensional core models, the negative reactivity insertion resulting f rom 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 Trio 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 by the mechanism.
The total delay to trip is defined as the time delay f rom the time that trip conditions are reached to the time the rods are f ree 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.
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 locp 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".
s 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 all adverse instrumentation and setpoint errors so that under ncminal 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 limit value of 1.62.
All points below and to the lef t of a DNB line for a WAPWR-I&C/EP 15.0-13 NDVEMBER, 1984 2218e:1d
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 j
pressure (fixed setpoint); low pressure (fixed setpoint); low DNBR (variable setpoint); high kw/ft (fixed setpoin.).
The limit value, which was used as the DNBR limit for all accidents analyzed with the Improved Thermal Design Procedure (see Table 15.0-3), is conservative compared to the actual design DNBR value required to meet the DNB design basis is discussed in Section 4.4 of RESAR-SP/90 PDA Module 5, ' Reactor System".
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 l
Specifications.
15.0.8 Instrumentation Drift and Calorimetric Errors - Power Range Neutron Flux The instrumentation drif t and calorimetric errors used in establishing the power range high neutron flux setpoint a presented in Table 15.0-5.
The calorimetric error is the error assumed in the determination of core thermal power as obtained f rom secondary plant measurements.
The total ion chamber current (sum of the multiple sections) is calibrated (set equal) to this measured power on a periodic basis.
i O
HAPWR-I&C/EP 15.0-14 NOVEMBER, 1984 2218e:1d
b The secondary power is obtained f rom measurement of feedwater flow, feedwater inlet temperature to the steam generators and steam pressure.
High accuracy instrumentation is provided for these measurements with accuracy tolerances much tighter than those which would be required to control feedwater flow.
15.0.9 Plant Systems and ComDonents Available for Mitigation of Accident Effects n
v The Westinghouse nuclear steam supply system (NSSS) is designed to afford power protection againts the possible ef fects of natural phenomena, postulated environmental conditions, and the dynamic eff ects 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 nomally operating systems and components listed in Table 15.0-6 will be available for mitigation of the events discussed in Chapter 15.
In determining which systems are necessary to mitigate the effects of these postulated
- events, the classification system of ANSI-N18.2-1973 is utilized.
The design of " systems 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.
In the analysis of the Chapter 15 events, the operation of 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 severe consequences.
NC credit is taken for control system operation if that l
operation mitigates the results of an accident.
For some accidents, the analysis is perfomed 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.
WAPWR-I&C/EP 15.0-15 NOVEMBER, 1984 2218e:1d i
15.0.10 Fission Product Inventories 15.0.10.1 Inventory in ti,e Core The time dependent fission product inventories in the reactor core are
}
calculated by the ORIGEN code using a
data library based on
}
ENDF/8-IV.
Core inventories are shown in Table 15.0-7.
The fission product radiation sources considered to be released from the fuel to the containment following a maximum credible accident are based on the assumptions stated in TID-14844( ):
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 tods 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, I-127, 'and I-129, which are 30 percent) is accumulated in the fuel clad gap. The gap attivities are shown in Table 15.0-7.
15.0.10.3 Inventory in the Reactor Coolant Reactor coolant iodine concentrations for the Technict.1 Specification limit of 1 uti/gm of dose equivalent (0.E.) I-131 and for the assume <. pre-accident iodine spike concentration of 60 WCi/gm of 0.E. 1-131 are presented in Table 15.0-8.
Reactor coolant noble gas concentrations based on 1 percent fuel defacts are presented in Table 15.0-9.
Iodine appearance rates in the reactor coolant, for normal steady state operation at 1 pCi/gm of 0.E.
1-131, and for an assumed accident initiated iodine spike are presented in Table 15.0-10.
O WAPWR-I&C/EP 15.0-16 NOVEMBER, 1984 2218e:1d
_ ~ _
l
(
15.0.11 Residual Decay Heat 15.0.11.1 Total Residual Heat e
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 time before the core goes subcritical to deter 1nine fission product decay
()
energy. For all other accidents, the same models are used except that fission product decay energy is based on core average exposure at the end of the equilibrium cycle.
15.0.12 Computer Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below.
Other codes, in particular, very specialized codes in which i
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 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System"), are summarized in their respective accident
' analyses sections.
The codes used in the analyses of each transient are listed in Table 15.0-3.
15.0.12.1.
FACTRAN 4
FACTRAN calculates the transient temperature distribution in a cross section of a metal clad UO fuel rod and the transient heat flux at the surf ace of 2
the cladding using as input the nuclear power and the time-dependent.oolant parameters (pressure, flow, temperature, and density).
The codes uses a fuel l
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.
1 2.
Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation.
WAPWR-!&C/EP 15.0-17 NOVEMBER, 1984 2218e:1d
3.
The necessary calculations to handle post DNB transient:
film boiling teat transfer correlations, Zircaloy-water reaction and partial melting of the materials.
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 reacter vessel, hot and cold leg piping, steam generators (tube and shell sides) and the pressuri-zer.
The pressurizer heaters, spray, relief and safety valves are also con-sidered in the program.
Point model neutron kinetics, and reactivity effects of the moderator, f uel, boron and rods are included.
The secondary side of l
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, low DNBR, high linear power (kW/f t), high and low pressure, low flow,
(
and high pressurizer level.
Control systems are also simulated including rod I
contral, 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.
t l
LOFTRAN also has the capability of calculating the transient value of DNBR based on the input f rom the core limits illustrated in Figure 15.0-1.
The 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.
l O
WAPWR-I&C/EP 15.0-18 NOVEMBER, 1984 2218e:ld
15.0.12.3 TWINKLE The TWINKLE program is a multi-dimensional spatial neutron kinetics code, which was patterned af ter steady state codes presently used for reactor core design.
The code uses an implicit finite-difference method to solve the two-group transient neutron dif fusion equations in one, two or three dimen-sions.
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 ef fects.
The code handles up to 2000 spatial points, and performs its own steady state initialization.
Aside f rom 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 of f set, 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.
15.0.12.4 THINC The THINC Code is described in Section 4.4 of RESAR-SP/90 PDA Module 5,
" Reactor System".
i 15.0.13 REFERENCES 1.
H.
Chelemer, et al.,
" Improved Thermal Design Procedure", WCAP-8567-P (Proprietary), July 1975, and WCAP-8568 (Non-Proprietary) July,1975.
2.
J.
- Skaritka, ed.,
" Hybrid 8C Absorber Control Rod Evaluation Report",
4 i
WCAP-8846-A, October 1977.
i WAPWR-I&C/EP 15.0-19 NOVEMBER, 1984 2218e:1d
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 8.
R.
Rider, " Compilation of Fission Product Yields",
)
NEDO-12154-1, General Electric Corporation, January 1974.
5.
F.
M.
8ordelon et al.,
" SATAN-VI Program:
Comprehensive Space Time Dependent Analysis of Loss-of-Coolant", WCAP-8306, June 1974.
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 Thernal Transients in a UO 2
Fuel Rod". WCAP-7908, June 1972.
8.
T. W. T. 8urnett et al., "LOFTRAN Code Description". WCAP-7907-P-A, April, 1984.
9.
D. H. Risher, Jr., and R. F. Barry, " TWINKLE - A Multi-Dimensional [ Neutron Kinetics Computer Code":
WCAP-7979-P-A (Proprietary) January 1975, and WCAP-8028-A, (Non-Proprietary), January 1975.
- 10. 8 ell, M.
J.,
"0RIGEN - The ORNL Isotope Generation and Depletion Code,"
ORNL-4628, May 1973
- 11. "0RIGEN Yields and Cross Sections - Nuclear Transmutation and Decay Data From END F/8-IV",
Radiation Shielding Infornation Center, Oak Ridge National Laboratory, RSIC-DLC-38, September 1975.
WAPWR-I&C/EP 15.0-20 NOVEMBER, 1984 2218e:1d
TABLE 15.0-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS N-Loop Doeration Reactor core thermal power output (MWt)*
3800 Therwel 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.
MAPWR-I&C/EP 15.0-21 NOVEMBER, 1984 2218e:1d
TABLE 15.0-2 VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES
- I N-Loop Doeration Thermal output of nuclear steam supply system (MWt) 3816 Reactor core therwel power output (MWt) 3800 Core inlet temperature (*F) 560.8 Reactor coolant average temperature (*F) 592.6 Reactor coolant system pressure (psia) 2250 Reactor coolant flow per loop (gpm) 97900 Total reactor coolant flow (106 lb/hr) 145.0 Total steam flow from NSSS (106 lb/hr) 17.14 Steam pressure at steam generator outlet (psia) 1024 Maximum steam moisture content (%)
0.25 Feedwater temperature at steam generator inlet (*F) 450 2
Average core heat flux (Stu/hr-ft )
162960 I
- For accident analyses using the improved thermal design procedure.
{
i l
WAPWR-I&C/EP 15.0-22 NOVEMBER, 1984 2218e:1d l
t.
TABLE 15.0-2a VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES
- N-Loop Doeration Therwel output of nuclear steam supply system (MWt) 3816 Reactor core therwel power output (MWt) 3800 Core inlet temperature (*F) 560.7 Reactor coolant average temperature (*F) 592.9 Reactor coolant system pressure (psia) 2250 Reactor coolant flow per loop (gpm) 96900 Total reactor coolant flow (106 lb/hr) 143.5 Total steam flow from NSSS (106 lb/hr) 17.14 Steam pressure at steam generator outlet (psia) 1024 Maximum steam moisture content (%)
0.25 Feedwater temperature at steam generator inlet (*F) 450 2
Average core heat flux (Btu /hr-ft )
162960
- For accident analyses not using the improved thermal design procedure.
1
~
i WAPWR-I&C/EP 15.0-23 NOVEMBER, 1984 2218e:1d
1Atti 15.0-3 SUMMART of IN!!!AL COM0lllONS AND COMPU1tR CODES U$te R1nette Parameters Assumje Initial leproved
- 555 Reactor Vessel Press.
Computer Delayed Moder.
Theres)
Thermal Vessel Average Press.
Iseter feedwater Codes Neutron Density Dug Design Power Coolant Temp.
Pressure Volume Temp.
faults V1111 red fraction ( As/en/cc) Doppler Correlation Proc ed. Dettet (188t) flow femen
(*ft (esial (ft31
(*ft 15.1 tacrease in Heat Removal by the Secondary System feedwater Systee Mal-function Cassing an (See RESAR-SP/90 PSA Module 6/S, ' Secondary Side Safeguards System /5 team and Power Conversion System')
increase in feedwater flow facessive Iv rease in secondary Steam flow (See etSAR-SP/90 PSA Module 6/S, *5econdary Side Safeguards System /5 team and Power Conversion Systee')
i Accldental Depresser-j 1 ration of the Mais (See RtSAR-$P/90 PSA Module 6/s, *$econdary Side Safeguards System /5 team and Power Conversten Systen")
j 5 team Systes Steam Systen Piping (see RESAR-SP/90 PSA Module 6/t, ' Secondary Side Safeguards System / Steam and Power Conversion Systee')
i failure 1
15.2 Decrease in Meat Removal by the Setendary Systee toss of faternal (see RisAt-SP/90 Ptn Modele 6/t. *5econdary Side Safegeerds System /5 team and Power Conversion System *)
tiectrical load and/or Terhine trip toss on Non-tmergency r
A-C Power to the Sta-(See atSAR-SP/90 PDA Module 6/S, *5econdary Slee safegeerds System / Steam and Power Conversion Systee')
i tion Aust11 aries toss of Morms) f eed-(See etSAR-SP/90 PSA Module 6/S, '5econdary Side Safeguards System /5 team and Power Conversion System *)
water flow feedwater System Ptpe (See RESAR-5P/90 PSA Module 6/S, *5econdary Side Safeguards System /5 team and Power Conversion Systee')
Greek e
Isarwo-l&C/t p 7288e:Id IS.0-N HOVtMBit 1994 w
1Astt 15.0-3 (Con't)
Elmetic Parameters assumed Initial leproved N555 Reactor vessel
- Press, Computer Delayed neder.
Thersel Theres) wessel Average Press.
Water Feeduster Codes Neutron Density see Design Power Coolant leep.
Pressere volume
- Temp, Faults Utilfred fraction f eo/en/cci possier Correlation Proc ed. Outset (Insti Flow feoal
(*F1 fasial fft))
f*F1 15.3 Decrease in Seacter Coolant Systen Flow sate Partial and Complete (See etSAR-5P/90 PDA saadute 4.
toss of f orced Beactor Coolant Flow teactor Coolant Pump (See RC5AR-SP/90 PSA Itodule 4. ' Reactor Coolant Syston')
shalt 5elaure (locked rotor 15.4 Beactivity and power distributton anomelles uncontrolled God (See RtSAR-5P/90 PDA pendule 5. " Reactor Systee')
Cluster Control Assently Sant Withdrawal from a seIKritical or Low Power Startup Conditten.
Itacontrolled Red (See RC5at-5P/90 PSA Module 5.
- Reactor Systee')
C1siter Assembly Sant Withdrawal at Power Control Red Mis-(Later) alignment
, Startup of an (See RESAR-SP/90 PSA Module 4. ' Reactor Coolant System')
Inactive Reactor Coolant Loop at an lacorrect Temperature 9
tenrue.:LC/t P
" I'5.0-25 77tne:54 180VIMel R.1984
l i
-)
I laett 15.0-3 (Can't)
Etnetic Parameters Assume (
Initial leproved 10555 seacter vessel Press.
Computer pelayed feeder.
Thermal Theriaal vessel Average
- Press, tieter feeduster Codes leestren pensity Dus Design Power Coolant temp.
Pressere volume toep.
Feelts Utillied fractlen f ee/se/cc) W Correlation Proc ed. Outout (sett) flow (somj f*ft testa)
(ft3)
. (*ft Cheetcal and volume (See RESAa-SP/90 PSA leadele 13. 'Aeutilary Systems')
Contral Systee nel-functlen that Desults in a petrease in peron i
Concentratten in the Reacter Coolant Inadvertent toeding (See et1As-SP/90 PDA seedule 5. *Reacter Systee')
and Operation of a fuel Assembly in an leproper Positten Spectrum of Red (See etSAR-SP/10 PSA feedele 5. ' Reactor Systee')
Cisster Centrol Asseely (jectlen accidents 15.5 Increase in Coolant Inventary inadvertent Operatten 81 4 IBA IIA IIA lea IIA inn 31 4 sta Ig4 geA geA of (CC5 Suring Power Operatten 15.6 Decrease in Beector Coelant Inventory Inadvertent Opening (See atSAR-SP/90 PDA IIedule 4. 'Reacter Coolant System *)
of a Pressertrer Safety or tellef valve l
l Sef erence figure 35.0-2.
Itentmum refers to lower curve and ninteum refers to apper curve, mA - not applicante.
00C - Beginning of cycle (DC - End of cycle 1
inspws. I LC/E P 15.0-M slovtsista.1994 22 tee:Id w
TABLE 15.0-4 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Limiting Trip Point Assumed Time Delays Trio Function In Analysis (sec)
Power range high neutron flux, 118%
0.5 high setting Power Range high neutron flux, 35%
0.5 low setting Power range neutron flux, high (Later)
(Later) negative rate High neutron flux, P-8 85%
0.5 Low DNBR Variable, see 6.0**
Figure 15.0-1 High pressurizer pressure 2410 psig 2.0 Low pressurizer pressure 1836 psia 2.0 Low reactor coolant flow B7% loop flow 1.0 (f rom loop flow detectors)
RCP underspeed 92% or nominal 0.6 speed Turbine trip Not applicable 2.0 Safety injection reactor trip Not applicable 2.0 Low steam generator level High steam generator level -
produces feedwater isolation and turbine trip Total time delay (including RTD time response and trip circuit channel electronics delay) f rom the time the temperature in the coolant loops exceeds the trip setpoint until the rods are free to fall.
WAPWR-!&C/EP 15.0-27 NOVEMBER, 1984 2218e:1d
r-
+
TABLE 15.0-5 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT -
POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRUMENT ERRORS Effect on I
Accuracy of Thernal Power Measurement Determination of Variable (5 error)
Variable (5 error)
(Estimated)
(Assumed)
Calorimetric errors in the measurement of secondary system thermal power:
Feedwater temperature
! 0.5 Feedwater pressure (small 1 0.5 0.3 correction on enthalpy)
Steam pressure (small 2
correction on enthalpy Feedwater flow t 1.25 1.25 Assumed calorimetric error 1 2(a)
(% of rated power)
Axial power distribution effects on total ion chamber current Estimated error 3
i
(% rated power)
Assumed error 1 5(b) i (5 of rated power)
Instrumentation channel drift l
and setpoint reproducibility Estimated error 1
(% or rated power)
Assumed error i 2 (c)
(% of rated power)
Total assumed error in setpoint i9 (a) + (b) + (c)
WAPWR-I&C/EP 15.0-28 N09EM8ER, 1984 2218e:1d
TABLE 15.0-5 (Con't)
Percent Rated Power Nominal Setpoint 109 Maximum overpower trip point 118 assuming all individual errors are simultaneously in the most adverse direction.
I 4
1 l
WAPWR-!&C/EP 15.0-29 NOVEMBER, 1984 2218e:1d i
i
.- ~. -
TABLE 15.0-6 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT ANO ACCIDENT CONDITIONS Incident Reactor Trio Functions ESF Actuation Functions Other Equipment ESF Eautemen 15.1 Increase in Heat Removed by the Secondary System Feedwater System Power range high flux, High steam generator Feedwater isolation NA Malfunction Causing high steam generator level-produced feedwater valves ca 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 DN8R, high actuated safety valves; kw/ft steam generator safety valves Accidental Depres-Low pressurizer Low pressurizer pressure, Feedwater isolation Emergency feed strization of the pressure, manual, SIS low compensated steam valves, steamline stop system; Safety Main Steam System line pressure. Hi-1 valves Injection Syst containment pressure, manual, low 4 Tcold Steam System Piping SIS, low pressurizer low pressurizer pressure, Feedwater isolation Emergency feed Failure pressure, manual low compensated steam-valves, steamilne stop system; Safety line pressure Hi-1 valves Injection Syst containment pressure, manual, low 4 T old c
15.2 Otcrease in Heat Eemoval by the Secondary System Loss of External High pressurizer low steam generator Pressurizer safety Emergency feedwater Electrical Load /
pressure, low DN8R, level valves, steam generator system Turbine Trip low steam generator level, manual WAPWR
./EP J-30 N
SER, 1984
....a
f TABLE 15.0-6 (Con't)
Incident Reactor Trio Functions ESF Actuation Functions Other EquiDeent ESF Eauipment loss of Non-Emergency Steam generator low Steam generator low Steam generator safety Emergency feedwa A-C Power to the level, manual level valves system l
Station Auxiliaries I
Loss of Normal Steam generator low Steam generator low Steam generator safety Emergency feedwa Feedwater Flow level, manual valves system level Feedwater System Pipe Steam generator low Hi-1 containment Steamline isolation Emergency feedwa Break level, high pressurizer pressure, steam generator valves, feedline 1501-system Safety pressure SIS, penual low level, low compen-ation, pressurizer Injection Syste low DNBR sated steamline pressure safety valves steam generator safety valves 15.3 Decrease in i
Reactor Coolant System Flow Rate Partial and Complete Low flow, low RCP NA Steam generator NA Lcss of Forced Reactor speed, manaul safety valves Coolant Flow R actor Coolant Pump Low flow, manual NA Pressurizer safety NA Shaft Seizure (Locked valves, steam generator R: tor) safety valves.
15.4 Reactivity and Power Distribution Anomalies Uncontrolled Rod Power range high flux NA NA NA Cluster' Control (low s.p.), manual Assembly Bank Withdrawal from a Subtritical or low Power Startup Condition WAPWR-I&C/EP 15.0-31 NOVEMBER, 1984 2218e:1d
TABLE 15.0-6 (Con't)
Incident Reactor Trio Functions ESF Actuation Functions Other Eauipment ESF Eeuisment Uncontrolled Rod Power range high NA Pressurizer safety NA Cl; ster Control flux, Hi pressurizer valves, steam generator Assembly Bank pressure, manual, low safety. valves Withdrawal at Power DN8R Control Rod Misalign-Power range negative NA NA NA ment flux rate, manual Stcrtup of an Inactive Power range high flux, MA NA NA R; actor Coolant Loop P-8, manual et an Incorrect Temperature Chemical and volume Source range high flux.
NA Low insertion limit NA 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 Boron Concentration la the Reactor Coolant Spectrum.of Rod Cluster Power range high flux, NA NA NA Control Assembly high positive flux rate.
Ej:ction Accidents annual 15.5 Increase in Reactor Coolant Inventory Inadvertent NA NA NA NA Operation of ECCS During Power Operation WAPWR-I-
'P If 12 NOV!
1, 1984 2218e:lt
a
+
TABLE 15.0-6 (Con't)
Incident Reactor Trio Functions ESF Actuation Functions Other Eeuipment ESF Eauisment 15.6 Decrease in Reactor Coolant Inventory Inadvertent Opening of Pressurizer low Low pressurizer pressure NA Safety Injection l
c Pressurizer Safety pressure, manual, low System j
er Relief Valve DNSR I
St:am Generator Tube Reactor trip system Engineered Safety Service Water System, Emergency Core Rupture Features Actuation Component Cooling Water Cooling System, System System, steam generator Emergency Feedwe safety valves, PORVs.
System, Emergent SG overfill control Power Systems valves and steamline stop valves j
Less of Coolant Reactor trip system Engineered Safety Service Water System, Emergency Core j
Accident from Spectrum Features Actuation System Component Cooling Water Cooling System.
J cf Postulated Piping System, Steam Generator Emergency Feeds i
Breaks within the Safety Valves System, Contair l
System Heat Removal S) i Emergency Power 1
l I
i WAPWR-I&C/EP 15.0-33 NOVEMBER, 1984 l
2218e:ld
?
i h
_u.a.-----.J-n.S TABLE 15.0-7 FUEL AND ROD GAP INVENTORIES, CORE (Ci)(*
Isotone Fuel igg (b)
I-131 1.0E + 7 1.0E + 6 I-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 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 Xe-133m 3.1E + 7 3.1E + 6 Xe-133 2.0E + 8 2.dE+7 Xe-135m 4.3E + 7 4.3E + 6 Xe-135 4.5E + 7 4.5'E + 6 Xe-138 1.7E + 8 1.7E + 7 1-127 3.0 kg 0.90 kg I-129 12.2 kg 3.7 kg 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, 1-127, and 1-129, whose gap activities are assumed to be 30 percent of their core activities (Regulatory Guide 1.25 assumption).
WAPWR-I&C/EP 15.0-34 NOV5MBER,1984 2218e:1d i
--- l
TABLE 15.0-8 REACTOR COOLANT IODINE CONCENTRATIONS FOR 1 pC1/ GRAM AND 60 pCI/ GRAM OF DOSE EQUIVALENT I-131 Reactor Coolant Concentration (Ci/gm)
Nuclide.
1 uti /om D. E. 1-131 60 v Ci/am D.E. I-131 1-131 0.76 45.6 I-132 0.76 45.6 I-133 1.14 68.4 1-134 0.195 11.7 I-135 0.63 37.8 WAPWR-I&C/EP 15.0-35 NOVEMBER, 1984 2218e:1d
TABLE 15.0-9 REACTOR COOLANT NOBLE GAS SPECIFIC ACTIVITY BASED ON DNE PERCENT DEFECTIVE FUEL l
Nuclide Activity fuc/ cram)
Kr-85m 2.0 Kr-85 7.3 Kr-87 1.3 Kr-88 3.6 Xe-131m 2.2 I
Xe-133m 1.7 x 10 Xe-133 2.7 x 10 Xe-135m 4.8 x 10'I Xe-135 7.2
-I Xe-138 6.4 x 10 WAPWR-l&C/EP 15.0-36 NOVEMBER, 1984 2218e:1d
TABLE 15.0-10 I0 DINE APPEARANCE RATES IN THE REACTOR COOLANT (Curies /sec)
- Appearance Rates
- Equilibrium Appearance Due to an Accident Rates Due to Fuel Defects Initiated Iodine Soike I-131 3.4 x 10-3 3,7
-2 I-132 1.8 x 10 9.0 I-133 7.2 x 10-3 3.6
-2 I-134 1.1 x 10 5,$
I-135 6.8 x 10-3 3.4 Based on RCS concentration of 1 pCi/gm of dose equivalent I-131.
500 x equilibrium appearance rate.
1 l
I WAPWR-I&C/EP 15.0-37 NOVEMBER, 1984 2218e:1d L
680-Pressurizer Pressure (psia) 0 660 N
- Locus of Points Where Reactor Vessel 240 g
Exit Temperature Equals Saturation s
N Temperature s
N 640 --
s N
s, 1990 s
N N
N s
N N
g s
N N
N s
s i
620 -
\\
1765
\\
N N
N N
N N
N s
N N
s C
N N
N s
k 600 -
N N
s
\\
N-
's s
N.
N, N
s N
N h5 N,
\\ \\
N N
\\h 580 N
\\
\\
l
\\
\\
\\j Low DNBR N
s 6
Reactor Trip Lines N
N g
(IncludingInstrumentErrors)
N 5
t N
\\
T 560..
N 1
N N
N N
1 540 --
Locus of Points Where i
DNBR Equals Limit Value for Reference Axial Power Distribution 520 0
20 60 80 100 120 Core Power (% of Nominal)
Figure 15.0-1 Illustration of Core Thermal Limits and DNB Protection (N Loop Operation)
WAPWR-!&C/EP t!0VEf1 DER, 1984
6 1.0 1
0.9 0.4 0.7 g
I g 0.6 E
es.
W 0.5 W
S::: 0.4 s
0.3 I
l 0.2 l
l 0.1 0
0 0.2 0.4 0.6 0.8 1.0 R00 POSITION (FRACTIONINSERTED) i FIGURE 15.0 4 NORt'AI2ED RCCA REAC IVITY WORTH VS. FRACTION INSECICf.
NOVEMBER, 1984' WAPWR-I&C/EP
i 1.2.
I.I '
IFully 1.0.
Top of Deshoot 0
.9<
e 1 0.s<
i b5
}e0.1.
I e,
1 0.6 1
y2 l
=
0.5-i l
0.4<
l
~
l
- 0.3< -
8 E
I 0.2. -
t 1
0.1<
l 1
0 0
0.1 0.2 0.3 0.4 0.5 0.6 0.1 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.5 hermaltred Drop Time (Time After Drop Begins/Orop Time to Top of Deshoot)
FIGURE 15.0-3 ROCA POSITION VS. TIPI TO DASHPOT WAPWR-1&C/EP NOVEMBER, 1984
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.3 Rod Cluster Control Assembly Misalianment (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident Description Rod cluster control assembly (RCCA) misoperation accidents include:
o A dropped RCCA, discussed in this module (to be provided later in the integrated PDA document).
o A dropped RCCA bank, discussed in this module (to be provided later in the integrated PDA document),
o Statically misaligned RCCA (discussed in RESAR-SP/90 PDA Module 5,
" Reactor System").
o Withdrawal of a single RCCA (discussed in RESAR-SP/90 PDA Modul;e 5,
" Reactor System').
WAPWR-!&C/EP 15.4-1 NOVEMBER, 1984 l
2220e:ld
_