RS-14-128, LaSalle, Units 1 & 2, Updated Final Safety Analysis Report, Revision 20, Chapter 15.0, Accident Analyses

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LaSalle, Units 1 & 2, Updated Final Safety Analysis Report, Revision 20, Chapter 15.0, Accident Analyses
ML14113A116
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Issue date: 04/11/2014
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LSCS-UFSAR 15.0-1 REV. 20, APRIL 2014 CHAPTER 15.0 - ACCIDENT ANALYSES 15.0.0 Introduction The following UFSAR sections evaluate th e capability of the plant to control or accommodate postulated failures and events. Prior to the initial startup on both units, Chapter 15 was written to provide a description of the analyses performed that showed the plant was capable of withstanding all credible events. The remainder of Section 15.0 describes the approach and methodology that was used in analyzing the events described in sectio ns 15.1 through 15.10. Sections 15.1 through 15.10 describe various events and their causes, initial conditions, sequence of events, probable consequences and pl ant performance for a number of nuclear system transients/accidents which were supposed to pose potential challenges to the Nuclear Steam Supply System. These analyses were performed for the initial cores on each unit. However, each analysis is not required to be performed on subsequent cycles as only a few of the events could cause a decrease in the margin to the Minimum Critical Power Ratio (MCPR) Safety Limit. Events analyzed in these sections are either bounded by other events or the analysis is good for all cycles provided no modifications have been done th at would affect the validity of the initial analysis. Consequently, for subsequent cycles on each unit only the limiting

transients are analyzed for their effects on MCPR. The results of the limiting transients are provided by General Electric in the Supplemental Reload Licensing Submittal for each cycle. A description of each of the limiting transients analyzed each cycle can be found in Section 15.A. "Typical" descriptions of the event can be found in the appropriate subsections of Sections 15.1 through 15.10. The information contained in Sections 15.1 throug h 15.10 is being retained for historical reference.

Analytical methods used by AREVA in support of reload fuel are outlined in Reference 7.

The AREVA evaluation of accident conditions begins with a Disposition of Events. AREVA documents their disposition of events for MUR PU conditions in Reference

18. At each follow-on reload cycle, AREVA confirms the validity of Reference 18.

Reference 18 is applicable to both Units 1 and 2. The objective of the disposition is to identify potentially events which require further evaluation to support an operating fuel cycle. As a result of the disposition, events may be categorized as:

1. The consequences of the event are potentially limiting and need to be evaluated on a cycle-specific basis.
2. The consequences of the event are bound by those of another event, i.e., the event is non-limiting.

LSCS-UFSAR 15.0-2 REV. 15, APRIL 2004 3. The event is not fuel related and the current analysis of record remains applicable.

The 1999 LaSalle County Station Power Uprate Project included re-evaluating a broad set of most limiting transient events at the power uprate conditions. The basis for the selection of the transient events for re-analysis is documented in "NEDC-31897P-A, LTR, General Guidelines for GE BWR Power Uprate, May 1992, Appendix E" where it stipulates, "Analy sis will be performed for the limiting transient events. This includes all events that establish the core thermal operating limits and the events that show bounding conformance to the other transient protection criteria (e.g., ASME overpressure limits)." The transient events which are re-analyzed with power uprate condit ions from 3323 MWth to 3489 MWth core thermal power are documented in Section 15.B and summarized in Table 15.B-1.

The initial conditions for the transient analysis are listed in Table 15.B-2.

15.0.1 Approach to Safety Analysis The safety analysis described in Sections 15.0.0 through 15.10 evaluates the ability of the plant with the initial core to operate within the regulatory guidelines without undue risk to the public health and safety.

The analysis investigates the categories of events by type and expected frequency to delineate the limiting cases where the radiological consequences are significant.

This approach ensures that a broad spectrum of initiating events is considered. It also enables the focusing of more detailed treatment of the radiologically important cases, while subordinating trivial and nondominant cases to lesser relative importance. A hypothetical ATWS event is also included at the request of the NRC. It has an extremely low probability of occurrence at LSCS.

In the treatment of specific safety case s initiated by typical plant events, the concept of expected frequency was mutually considered with the mechanisms of radiological release, to scope the safety risk associated with that particular event.

The safety analysis presents two categories of events: transien ts and accidents. Transients are subdivided into two subsets: moderate frequency events and

infrequent incidents. For the purpose of simplifying a summary of results, however, all transient events are tabulated, independ ently of frequency, with regard to their Critical Power Ratio (CPR) operating limit. The accident results are also tabulated in the same summary (Table 15.0-2). The init iating events were assigned one of the following expected frequencies based upon practical Exelon operating experiences with six nuclear power stations:

a. Transients Moderate frequency - Events which may oc cur during a calendar year to once per 20 years for a particular plant. Anticipated operational transients are in this frequency class.

LSCS-UFSAR 15.0-2a REV. 14, APRIL 2002 Infrequent incidents - Events which are expected to occur once during the lifetime of the plant (including those that may occur once every 20 to 100 years). Unexpected or abnormal operational transients are includ ed in this frequency class.

b. Accidents Limiting fault frequency - refers to those incidents that are never expected to happen but for which safety analyses are arbitrarily made to represent upper bounds on the radiological LSCS-UFSAR 15.0-3 REV. 13 consequences. The design-basis accident is included in this frequency class.

It should be noted, for example, that th e frequency of an initiating event may be described by the term limiting fault frequency and not be characterized as the limiting fault per se. For the LaSalle County Station (LSCS), no limiting fault was found, hence this UFSAR does not attempt to describe a limiting fault, but treats it only as a concept.

In the treatment of particular events, the product of the expected frequency and the scale of the radiological consequence was us ed to categorize the safety significance of the event. For example, an initiating incident of moderate frequency that has no radiological consequence (because no radioactivity was released beyond the primary pressure boundary) was subordinated in safety importance to an infrequent incident which allowed radioactivity to be released beyond the primary pressure boundary.

The plant design takes into account the fact that the integrity of this pressure boundary, the primary containment and the secondary containment constitute significant safety barriers. Indirectly then, the conceptual probability for breaches to all of these barriers aids not only in the analytical treatment for the UFSAR appraisals but also in the delineation of physical processes important to knowledgeable safety design.

15.0.2 Categories of Safety Events Transient and accident events are categorized by their initiating cause via the process variable whose change may have a deleterious effect on the nuclear reactor fuel. Each postulated initiating incident is assigned to one of the following categories:

a. Nuclear system pressure increase threatens to overstress the reactor coolant pressure boundary from internal pressure. A pressure increase collapses the voids in the moderator thereby increasing reactivity and power which threaten fuel cladding due to overheating.
b. Reactor vessel water (moderator) temperature reduction results in an increase in core reactivity as density increases. Positive reactivity increases have the effect described in item a.
c. Reactivity and power distribution anomalies may reduce the void content of the moderator thus resulting in increased reactivity and power levels. Such transient anomalies may affect the fuel cladding.

LSCS-UFSAR 15.0-4 REV. 13

d. Reactor vessel coolant inventory decrease could threaten the fuel as the coolant becomes less able to remove the heat generated in the core.
e. Reactor core coolant flow decrease could result in overheating of the cladding as the coolant becomes unable to adequately remove the heat generated in the fuel.
f. Excess of coolant inventory co uld result in damage resulting from excessive moisture carryov er to the main turbine.
g. Postulated radioactive releas e from a subsystem or component due to loss of integrity.
h. Postulated (anticipated) transients without scram which results from multisystem maloperation plus active component failures. This is a hypothetical situation with an extremely low probability of occurrence.
i. Postulated thermal-hydraulic instabilities in certain portions of the core and flow operating domain, which could threaten MCPR limits. These nine categories include all of the effects on the nuclear system caused by abnormal operational transients which might lead to degradation of the reactor fuel barrier or reactor coolant pressure boundar
y. The variation of any one of these parameters may affect another. For purposes of analysis in the UFSAR, events are analyzed in groups, according to the initiating incident or event. For example, positive reactivity insertions resulting from sudden pressure increases are evaluated in the increase in reactor pressure classification.

The input parameters and initial conditions used for the initial core transient and accident analyses are listed in Table 15.0-1.

15.0.3 Judgment of Nonacceptable Safety Results For all transients of moderate and infr equent frequencies, the following are considered to be unacceptable safety results:

a. Release of radioactive material to environs in excess of 10 CFR 20 limits.
b. Reactor operation induced fuel clad failures.

LSCS-UFSAR 15.0-5 REV. 13 c. Nuclear system stresses in excess of those allowed for in the transient classification by applicable industry codes.

d. Containment stresses in excess of those allowed for in the transient classification by applicable industry codes.

For the design-basis accidents (limiting faul ts), the following are considered to be unacceptable safety results:

a. Release of radioactivity resultin g in dose consequences in excess of 10 CFR 100 values.
b. Failure of fuel cladding sufficient to cause changes in core geometry such that core cooling would be inhibited.
c. Nuclear system stresses in excess of those allowed for the accident (faulted) classification by applicable industry codes.
d. Containment stresses in excess of those allowed for the accident classification by applicable industry codes when containment is required as a barrier.
e. Radiation exposure to plant operations personnel in the main control room in excess of 5 rem whole body, 30 rem inhalation dose and 75 rem skin dose.

15.0.4 Method of Analysis Each transient or accident analyzed for the initial core is discussed and evaluated in terms of a sequence of events from the initiating condition to final stabilized state. The normal operation of unfailed equipment and controls is assumed. Credit is taken for plant systems and reactor protection systems in their normal functioning mode. The operation of unfaile d engineered safety features (ESF) is also included. The effect of a single operator error or a single failure of active equipment is also included in certain analyses; however, this is done on an or basis for transient evaluations. In the evaluati on of these postulated events, the plant damage allowances or limits are the same as those for normal operation. The evaluation presented herein interprets the accidents and transients in consonance with the historical frequency classification for the initiating events, i.e., the margin or limit was reported on the initiating ev ent frequency rather than upon contingent or conditional frequencies of multiple ev ents involved in certain sequences.

It is important to recognize that certain arbitrary accident scenarios require the application of single failures and operator errors. Others, such as ATWS, require multiple failures for the postulated end condition. In these accidents the event LSCS-UFSAR 15.0-6 REV. 19, APRIL 2012 frequency has a much lower probability.

Credible frequency classes for these multifailure, multierror scenarios are not currently recognized, hence recourse was made to the former more simple classifica tion for the convenience of cataloging these very low probability transients.

Most events postulated for consideration ar e already the results of single equipment failures or single operator errors hypothesized during normal or planned plant operations. Typical operational equipment failures or operator errors that can initiate events important to safety are as follows:

a. Undesired opening or closing of a single valve (a check valve is not assumed to close against normal flow).
b. Undesired starting or stopping of a single component (a change of state is not assumed without an assignable cause).
c. Malfunction or maloperation of any single control device.
d. Single failure of any electrical component.
e. Single operator error event by one person.

In general, the analyzed events have numerical input parameters and initial (state) conditions as specified in Table 15.0-1. Note that these are analytical values.

Analyses that assume data inputs different from these values are designated accordingly in their discussion and the specific parameters are defined therein for such cases.

Initial Power/Flow Operating Constraints The analytical basis for most of the initial core transient safety analysis is the thermal power at rated co re flow (100%) correspond ing to approximately 105% Nuclear Boiler Rated steam flow. This operating point is the apex of a bounded operating power/flow map which, in response to any abnormal operational transients, will yield the minimum pressure and thermal margins of any operating point within the bounded map. Referring to Figure 15.0-1, the apex of the bounded power/flow map is point A, the upper bound is the design flow control line (105% rod line A-D'), the lower bound is the zero power line H-J', the right bound is the rated valve position line A-H', and the left bound is either the low pump speed, minimum valve position line D-J or the natural circulation line D'-J'.

The power/flow map, A-D'-J'-H'-A, represen ts the acceptable operational constraints based on abnormal operational transient evaluations.

LSCS-UFSAR 15.0-7 REV. 19, APRIL 2012 Any other constraint which may truncate the bounded power/flow map must be observed, such as the recirculation valve and pump cavitation regions, the licensed power limit and other restrictions based on pressure and therma l margin criteria. For LaSalle, the initial cycle analyzed power was 3454 MWt(point A), the power/flow map is not truncated by the line B-C. Reactor operation must be confined within the boundary A-D'-J'-L-K-A. For a derated operating power level, such as point F which is applicable to satisfy a pressure margin criteria (for example), the upper constraint on power/flow is correspondingly reduced to the rod line, such as line F-G', which intersects the power/flow coordinate of the new operating basis. For this ex ample, the operating bounds wo uld be F-G'-J'-J-L-K-F.

Operation would not be allowed at any point along line F-M, left of point F, at the derated power but at reduced flow. On the other hand, if derated operation is restricted to point F by some MCPR limitation, operation at point M (or right of it) would be allowed provided the MCPR safety limit is not violated. Consequently, the upper operating power/flow limit of the reactor is predicated on the operating constraint of the analysis (i.e., flow or MCPR) and the corresponding constant rod pattern line. Consequently, the upper operating power/flow limit of a reactor is predicated on the operating basis of the analysis and the corresponding constant, rod pattern line.

Certain localized events are evaluated at other than the above mentioned conditions. When applicable, such conditions are discussed specifically for that appropriate event.

The power and flow used in the reload analyses are given in the Technical Requirements Manual. A typical power to flow map is shown in Figure 15.0-1(a).

The GE models used are given in Referenc e 1 and Reference 2. The AREVA models are discussed in References 7 and 9 thro ugh 11. For LOCA see section 6.3.3 for a detailed description.

Core and System Performance Section 4.2 describes the various fuel failure mechanisms. An acceptable criterion was determined to be that 99.9% of the fuel rods in the core would not be expected to experience boiling transition (Refer ence 1). This criterion is met by demonstrating that transients and accidents do not result in a minimal critical power ratio (MCPR) less than 1.06 for the initial core, or the value given in Technical Specification for reload cores which is defined as the safety limit MCPR for LaSalle 1&2.

The steady-state reactor operating limit is determined as follows:

a. The change in the critical power ratio (CPR) which would result in the safety limit CPR being reached, is calculated for each event. These CPR values are shown in Table 15.0-2 for LSCS-UFSAR 15.0-8 REV. 15, APRIL 2004 the initial cores and in the Technical Requirements Manual for reload cores.
b. For GE the CPR value for each event is then added to the Safety Limit CPR value to yiel d the event-based MCPR, FANP develops an operating limited based on a delta CPR and safety limit that bounds the limiting events.
c. The exception to b are those events whose CPRs are calculated using ODYN. For events whose CPR is determined by ODYN (GE only) (all rapid pressurizati on events), the event-based MCPR is determined in conjunction with NRC-additive correction factors, the CPR, and the Safety Limit CPR. These correction factors are listed in Table 15.0-1 for the initial core and in the Reload Licensing Onsite Review Package for reload cores. FANP documents their approach to account for the transient analysis code (CONTRANSA) uncertainties in the statistical determination of the MCPR in References 10 and 11.

These results are given graphically in Figure 15.0-2 for limiting initial core transients and accidents. The operating limit MCPR is the maximum locus of values from these event MCPRs calculated with the above method. The maximum calculated MCPR for the initial core is de picted by the solid line in Figure 15.0-2. Maintaining the CPR operating limit at or above the operating limit assures that the LaSalle Safety Limit CPR is never violated. The MCPR operating limit for reload cores can be found in the LaSalle Administrative Technical Requirements.

In addition to MCPR, MAPLHGR and LHGR are also fuel design limits.

For situations in which fuel damage is sustained, the extent of damage is determined by correlating fuel energy content, cladding temperature, fuel rod internal pressure, and cladding mechanical characteristics. The bases for these correlations are the fuel rod failure tests discussed in Section 4.4, and in Section 6.3.

Barrier Performance If there is no cladding damage, fission products are constrained to the fuel and only activation products are present in the reactor coolant. The performance of the Reactor Coolant Pressure Boundary (RCPB) and the containment system during transients and accidents is the prim ary evaluation of this section.

During transients that occur with no release of coolant to the containment only RCPB performance is considered. If releas e to the containment occurs as in the case of limiting faults, then challenges to the containment are evaluated as well.

LSCS-UFSAR 15.0-9 REV. 13 Releases from containment are evaluated for those specific cases which involve source terms outside the primary containm ent boundary. The normal operation of the SGTS and the single point stack is covered in such cases, as applicable.

Reactor Coolant Pressure Boundary Damage

The only significant areas of interest for internal pressure damage are the high-pressure portions of the reactor coolant pressure boundary (the reactor vessel and the high-pressure pipelines attached to the reactor vessel). The overpressure below which no damage can occur is define d as the pressure increase over design pressure allowed by the applicable ASME Boiler and Pressure Vessel Code,Section III, Class 1, for the reactor vessel and the high-pressure nuclear system piping.

Because the ASME Boiler and Pressure Ve ssel Code,Section III, Class 1, permits pressure transients up to 10% over design pressure, the design pressure portion of the reactor coolant pressure boundary meets the design requirement if peak nuclear system pressure remains belo w 1375 psig (110%

x 1250 psig).

Peak fuel enthalpy (discussed in Subsection 4.3) is used to evaluate whether reactor coolant pressure boundary damage occurs as a result of reactivity accidents. If peak fuel enthalpy remains below 280 cal/g, no reactor coolant pressure boundary (clad) damage results from nuclear excursion accidents and therefore no other barriers are challenged to retain concentrated fission products.

Radiological Consequences In this section, the consequences of radioactivity release during both types of events: (1) operational transients, and (2) limiting faults or design-basis accidents are considered. For all events whose consequences are limiting, a detailed quantitative evaluation is presented.

For non-limiting events, a qualitative evaluation is presented or results are referenced from a more limiting or enveloping case or event.

For limiting faults or design-basis accidents, two quantitative analyses are considered:

a. The first is based on conservative assumptions considered to be acceptable to the NRC for the purposes of the worst case bounding event which determines the adequacy of the plant design to meet 10 CFR Part 100 guidelines. This analysis is referred to as the "desi gn-basis analysis".
b. The second is based on realistic assumptions considered to reflect expected radiological co nsequences, i.e., what could be LSCS-UFSAR 15.0-10 REV. 19, APRIL 2012 measured as an average value. This analysis is referred to as the "realistic analysis." Results for both are shown to be within NRC guidelines.

Results The results of analytical evaluations are pr ovided for each initial core event. In addition the result summary is shown in Table 15.0-2. From that table, a comparison can be made of the limiting event for any particular safety event category. Radiological analyses are not performed for reloads as the UFSAR Section 15.6.5 analyses are bounding.

15.0.5 Meteorological Parameters (Other than Alternative Source Terms)

Atmospheric dilution factors (X/Q's sec/m 3 ) are summarized here for use in this chapter. The atmospheric dilution factor s for the conservative analyses are based on the diffusion models presented in U.S. NRC Regulatory Guide 1.3, Revision 2 (June 1974) and U.S. NRC Regulatory Gu ide 1.145, Revision 1 (November 1982).

X/Q values determined per RG 1.145 methodology are determined for an elevated release and for a ground level release. Th e X/Q values for an elevated release are based on 1978 through 1987 historical met eorology at 375 feet above grade and represent a release via the plant exhaust stack. The X/Q values for a ground level release are based on 1982 through 1987 hist orical meteorology at 33 feet above grade and represent a release via the turbine building.

The atmospheric dilution factors at the 50th percentile for the realistic analyses have been derived from 2 years of onsite meteorological data.

Estimates of atmospheric dispersion for effluents released through the standby gas treatment system (SGTS) vent are based on values given in Table 2.3-48, 2.3-50, and 2.3-58. Estimates of atmospheric disper sion for effluents released through the plant common stack are based on values gi ven in Tables 2.3-33, 2.3-35, and 2.3-37. These data reflect a "realistic" average value estimate of expected consequences.

The atmospheric dilution factors are:

a. Conservative NRC Regulatory Guide 1.3 Values Time Periods-hrs X/Q sec/m 3 1. Exclusive Area Boundary (509 meters) 0-0.5 1.8x10-4 0.5-2.0 1.5x10-5 LSCS-UFSAR 15.0-11 REV. 13
2. Low Population Zone (6400 meters) 0-0.5 1.8x10-5 0.5-8 4.4x10-6 8-24 1.7x10-6 24-96 5.0x10-7 96-720 1.7x10-7 b. Conservative NRC Regulatory Guide 1.145 Values b.1 Elevated Release Out the Plant Exhaust Stack Time Periods-hrs X/Q sec/m 3
1. Exclusion Area Boundary (2800 meters)***

0-0.5 8.4x10-5 (fumigation) 0.5-2.0 2.6x10-6 2. Low Population Zone (6400 meters) 0-0.5 8.9x10

-6 0.5-2.0 1.6x10

-6 2.0-8.0 9.2x10

-7 8.0-24 5.5x10

-7 24-96 2.5x10

-7 96-720 8.2x10

-8 b.2 Ground Level Release via the Turbine Building Time Periods-hrs X/Q sec/m 3 1. Exclusion Area Boundary 0-2 5.1x10

-4 (423 meters) **** 2. Low Population Zone (6400 meters) 0-8 1.0x10

-5 8-24 6.7x10

-6 24-96 2.6x10

-6 LSCS-UFSAR 15.0-12 REV. 14, APRIL 2002 96-720 6.5x10

-7 c. Realistic Values - Standby Gas Treatment System Vent Time Periods-hrs X/Q sec/m 3 1. Exclusion Area Boundary (509 meters) 0-2* 4.0x10-7** 2. Low Population Zone (6400 meters) 0-8 1.47x10

-7 8-24 3.29x10

-7 24-96 1.37x10

-8 96-720 1.14x10

-8 d. Realistic Values - Plant Common Stack Time Periods hrs X/Q (sec/m

3) 1. Exclusion Area Boundary (509 meters) 0-2* 1.71x10-7** 2. Low Population Zone (6400 meters) 0-8 5.75x10-8 8-24 1.33x10

-8 24-96 5.62x10

-9 96-720 4.96x10

-9

  • A predicated fumigation condition at the onset of an accident is not considered realistic for observed meteorology at LSCS.
    • The "realistic boundary" for maximum dose is not the EAB; the releases are elevated, and therefore not monotonic with distance. The "realistic boundaries" for the SGTS vent and the plant common st ack are 4500 and 6400 meters, respectively. The EAB 50th percentile X/Q's for the SGTS vent and the plant common stack are 5.03x10-13 and 6.13x10

-25 sec/m 3 , respectively.

      • The distance of 2800 meters is in the SW downwind direction. This distance is greater than the Exclusion Area Boundary di stance in the SW downwind direction.

For elevated releases, the maximum sector X/Q value can occur at a distance LSCS-UFSAR 15.0-13 REV. 19, APRIL 2012 greater than the EAB boundary. In accordance with Regulatory Guide 1.145, the maximum sector X/Q value is used.

        • The distance of 423 meters is the sh ortest distance between the turbine building and the Exclusion Area Boundary (EAB) within a 45

° sector centered on the WNW downwind direction. This is the methodology for determining sector distances to the EAB stipulated in Regulatory Guide 1.145.

Comparison between the realistic values with those obtained from Regulatory Guide 1.3 and Regulatory Guide 1.145 shows that the realistic values are consistently lower. 15.0.5a Meteorological Parameters (Alternative Source Terms)

Alternative Source Terms (AST) Atmospheric dilution factors (X/Q's sec/m

3) for the EAB and LPZ were calculated with th e model PAVAN which implements the guidance provided in Regulatory Guide 1.145.

X/Q values were calculated for an elevated release via the stack and ground-level release via the turbine building utilizing the 1999-2003 meteorological tower data.

This tower data consists of wind speed and direction measurements at 33 ft, 200 ft and 375 ft and delta temperature measuremen ts at 375 - 33 ft and 200 - 33 ft. The meteorological data and the X/Q values are contained in Chapter 2.3.4a.

15.0.6 Nuclear Safety Operational Analysis (NSOA) Relationship The main objectives of the operational analyses are to identify all essential protection sequences and to identify the detailed hardware conditions essential to satisfying the nuclear safety operational criteria. The main objective of the analyses of Chapter 15.0 is to provide detailed analyses of the "worst cases."

15.0.7 MSIV Closure Change from Reac tor Water Level 2 to Water Level 1 The sequence of events for the cases an alyzed in Chapter 15 indicate the MSIV isolation occurs at reactor water Level 2. The MSIV Isolation was changed from Level 2 to Level 1 and as stated Reference 2 in this design change has been included in the current analyzed licensing accident events described in Chapters 15 and 6.3.3.

15.0.8 MSIV Closure Position Limit Switch Change from 90% Open to 85%

Open Reference 7 provides documentation of the re-analysis of the MSIV Inadvertent Closure Event allowing a Technical Specific ation Change to the MSIV limit switch. The previous limit switch set point were based upon a 90% open analytical limit, a 94% open trip and a 93% open allowable. The revised setpoints are based upon an 85% open analytical limit, a 88% open Technical Specification Limit, and a 92% open LSCS-UFSAR 15.0-14 REV. 19, APRIL 2012 nominal setpoint. This rean alysis resulted in a peak vessel pressure of 1203 psig, 4 psig higher than the original co re load analysis value of 1199.

15.0.9 Impact of Increased Initial Suppression Pool Temperature The initial conditions for the cases evaluated in Chapter 15 indicate an initial suppression pool temperature of 100

°F. The maximum suppression pool temperature limit (for normal operation) was changed to 105

°F as stated in Section 6.2.1.8. This temperature limit change was verified to have an insignificant impact on the accident events described in Chapter 15.

15.0.10 Reduction in the Total Number of SRVs

An evaluation and analysis has been performed based on the removal of five (5)

SRVs, for a total of 13 installed SRVs (Ref erences 12 and 13). See Table 5.2-9 for a summary of the remaining valves. None of the ADS or Low-Low Setpoint Valves are affected, thus postulated accidents involving ADS valves are not impacted. For the applicable transients which utilize the SRVs, a conservative analysis based on the MSIV closure event has been performed. Events involving the MSIV isolation of the NSSS at high power levels, are the most demanding on the SRVs. The analysis was done as part of an ASME overpressure analysis to show feasibility. This analysis is required to be reverified for each operating cycle through the cycle specific safety analysis process (See Section 15.A).

Details of the analysis are discussed in Section 5.2.2.2.3. The analysis conservatively assumed a reactor scram based on a high neutron flux signal (in lieu of the MSIV position trip logic). The analysis assumed only 10 SRVs (out of the total 13 SRVs) are available. The calculated peak vessel pressure increases slightly but is still significantly below the ASME Code limiting pressure of 1375 psig. Reference 14 documents an evaluation of the effect on MCPR for the reduction in number of SRVs.

This effect has been determined to be neg ligible. Based on this analysis it can be concluded that there is no significant effect on the consequences associated with any of the accidents or transients associated with the SRVs. The effect on ATWS was evaluated separately and is discussed in Section 15.8.

15.0.11 References

1. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A, (latest approved revision).
2. Letter Dated March 6, 1987, from C.M. Allen (CECO NLA) to H.R.

Denton (NRC) concerning MSIV level setpoint change from Level 2 to Level 1.

LSCS-UFSAR 15.0-15 REV. 19, APRIL 2012

3. GE document NEDC-31455, "Extended Operating Domain and Equipment Out of Service for LaSalle County Station Units 1 and 2," dated March 1990, with addenda.
4. Deleted
5. GE Document GE-NE-187-62-1191, "Equipment Out-of Service in the Increased Core Flow Domain For LaSalle County Station Units 1 and 2," (latest approved version).
6. GE Letter LS-2209, lt. R. Peffer (GE) to G. R. Crane (ComEd),

Subject:

LaSalle County Station Unit 1 & 2 Technical Specifications on MSIV Closure Scram, dated March 15, 1982.

7. Boiling Water Reactor Licensing Methodology Summary, EMF-94-217(P), Revision 1, Siemens Power Corporation, Nuclear Division, October 1995.
8. Intentionally Deleted.
9. Exxon Nuclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis , XN-NF-80-19(P)(A), Volume 1 and Supplement 3, Exxon Nuclear Company, Inc., March 1983.
10. Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description , XN-NF080-19(P)(A), Volume 3, Exxon Nuclear Company, Inc.
11. EXXON Nuclear Methodology for Boiling Water Reactors, Application of the ENC Methodology for BWR Reloads, XN-NF-80-19(P)(A) Volume 4 Revision 1, Siemens Power Corporation, June 1986.
12. "Safety Review for LaSalle County Station Units 1 and 2 Safety Relief Valves Reduction and Setpoint Toleranc e Relaxation Analyses", Rev. 2, GE Report, GE-NE-B13-01760, by H.

X. Hoang, dated February 1996. (On Site Review No.96-020)

13. Safety Evaluation Report (SER) by NRC, dated 06-03-99, for Amendment Nos. 133 and 118 for LaSalle County Station Units 1 & 2, respectively.
14. Letter No. NFS:BSA:96-084, dated 8 96, from R. W. Tsai to P. Antonopoulos, "MCPR and LOCA Impact for Safety Relief Valve Setpoint Tolerance Relaxation and Removal LSCS-UFSAR 15.0-16 REV. 19, APRIL 2012
15. Design Analysis L-003560, Revision 0, "T200 Series - Operating Power/Flow Map," July 2010.
16. Power Uprate Project Task 900, "Transient Analysis," GE-NE-A1300384-08 Revision 1, September 1999.
17. Intentionally Deleted.
18. Design Analysis L-003505, Revision 0, "Disposition of Events Summary for the LaSalle MUR Power Uprate," July 2010.

LSCS-UFSAR

TABLE 15.0-1 (SHEET 1 OF 3) TABLE 15.0-1 REV. 13 INPUT PARAMETERS AND INITIAL CONDITIONS FOR ANALYSIS OF INITIAL CORE TRANSIENTS AND ACCIDENT

1. Thermal power level, MWt Analysis value 3454 (104.8% NBR) 2. Steam flow, lb per hr 14.81 x 10 6 (104% NBR) 3. Core flow, lb per hr 108.36 x 10 6 4. Feedwater flow rate, lb per sec 4115 5. Feedwater temperature, °F 420
6. Vessel dome pressure, psig 1020 7. Vessel core pressure, psig 1031 8. Turbine bypass capacity, %NBR 25
9. Core coolant inlet enthalpy, Btu per lb 529 10. Turbine inlet pressure, psig 962 11. Fuel lattice 8 x 8
12. Core average gap conductance, Btu/sec-ft 2 - °F 0.1662 13. Core leakage flow, % 12 14. Required MCPR operating limit See Figure 15.0-2
15. MCPR Safety Limit 1.06 16. Doppler coefficient (-)¢/°F Nominal EOC-1 0.221 Analysis data 0.221

LSCS-UFSAR

TABLE 15.0-1 (SHEET 2 OF 3) TABLE 15.0-1 REV. 13

17. Void coefficient (-)¢/% rated voids Nominal EOC-1 7.479 Analysis data for power Increase events 12.63 Analysis data for power Decrease events 7.01 18. Core average rated void fraction,% 41.03 19. Scram reactivity, Analysis data FSAR Figure 15.0-2
20. Control rod drive speed, position versus time FSAR Figure 15.0-2 21. Jet pump ratio, M 2.28
22. Safety/Relief valve capacity, % NBR at 1165 psig 111.5 Manufacturer Crosby Quantity Installed 18 23. Relief function delay, seconds 0.1 24. Relief function response, seconds 0.1
25. Analytical setpoints for safety/relief valves Safety function, psig 1150, 1175, 1185, 1195, 1205 Relief function, psig 1076, 1086, 1096, 1106, 1116 26. Number of valve groupings simulated Safety function, No. 5 Relief function, No. 5 27. High flux trip, % NBR Analysis setpoint (120 x 1.038), % NBR 124.6

LSCS-UFSAR

TABLE 15.0-1 (SHEET 3 OF 3) TABLE 15.0-1 REV. 0 - APRIL 1984

28. High-pressure scram setpoint, psig 1071 29. Vessel level, inches above steam dryer skirt bottom (instrument zero 527.5)

Level 8 - (L8) Analytical Value 57.1* Level 3 - (L3) Actual Setpoint 12.5 Level 2 - (L2) Actual Setpoint -50 30. APRM thermal trip Setpoint, % NBR 121.54 31. RPT delay, seconds 0.190

32. RPT inertia time constant, sec** 6 TIME LR/TT w/o BP FW CONTROL FAILURE BOC -0.004 +0.029
33. CPR/ICPR Adjustment Factor to be applied to LaSalle ODYN deterministic results to establish Option 95/95 pressure transient CPR operating limits. MOC -0.021 +0.016
  • NRC reference level is 60.0 inches. GE sensitivity analyses for the 2.9 inch difference showed that a CPR of 0.003 would exist; however, this small increment is dropped in rounding off the MCPR's in the ODYN solution. Actual setpoints are not used at L3 or L2 in the transient/ accident analyses. where t = inertia time constant; J 0 = pump motor inertia, n = rated pump speed, g = gravitational constant; T 0 = pump electrical torque 0 0 gT n J 2 t**=

LSCS-UFSAR

TABLE 15.0-2 (SHEET 1 OF 4) TABLE 15.0-2 REV. 13

SUMMARY

OF EVENTS RESULTS FOR INITIAL CORES DURATION OF BLOWDOWN

PARA-GRAPH+

FIGURE DESCRIPTION MAXIMUM NEUTRON FLUX  % NBR MAXIMUM DOME PRESSURE psig MAXIMUM VESSEL PRESSURE psig MAXIMUM STEAM LINE PRESSURE psig MAXIMUM CORE AVERAGE SURFACE HEAT FLUX % OF INITIAL

CPR FREQUENCY CATEGORY ff NO. OF VALVES 1st BLOW-DOWN DURATION OF BLOW-DOWN SEC fff 15.1 DECREASE IN CORE COOLANT TEMPERATURE 15.1.1 15.1.1-1 Loss of Feedwater Heater, AFC 111.4 1020 1058 994 106.1 0.06 a 0 0 15.1.1 15.1.1-2 Loss of Feedwater Heater, MFC 123 1030 1067 1017 117.2 0.16 a 0 0 15.1.2A 15.1.2-1 Feedwater Cntl Failure, MAX DEMAND HI PWR w/bypass 215 1140 1168 1135 114.4 0.11* a 18 6 15.1.3 15.1.3-1 Pressure Regulator Fail Open 115%

Flow 103.9 1068 1083 1066 100.0 0.05* a 0 0 15.2 INCREASE IN REACTOR PRESSURE 15.2.2A 15.2.2-1 Generator Load Rejection, Bypass-On, RPT-On 214 1140 1166 1131 106.4 0.07* a 18 15.2.2A 15.2.2-2 Generator Load Rejection, Bypass-Off, RPT-On 350 1166 1192 1163 113.6 0.15* b 19 LSCS-UFSAR

TABLE 15.0-2 (SHEET 2 OF 4) TABLE 15.0-2 REV. 0 - APRIL 1984 DURATION OF BLOWDOWN

PARA-GRAPH+

FIGURE

DESCRIPTION MAXIMUM NEUTRON FLUX  % NBR MAXIMUM DOME PRESSURE psig MAXIMUM VESSEL PRESSURE psig MAXIMUM STEAM LINE PRESSURE psig MAXIMUM CORE AVERAGE SURFACE HEAT FLUX % OF INITIAL

CPR

FREQUENCY CATEGORY ff NO. OF VALVES 1st BLOW-DOWN DURATION OF BLOW-DOWN SEC fff 15.2.3 15.2.3-1 Turbine Trip, Bypass-On, RPT-On 165 1138 1164 1123 103.1 0.08 a 18 5.4 15.2.3A 15.2.3-2 Turbine Trip, Bypass-Off, RPT-On 352 1165 1190 1162 112.1 0.13 b 18 15.2.4 15.2.4-1 Main Steam Line Isolation, Position Scram 269 1163 1199 1152 108.6 <0.04** a 18 6.4 15.2.5 15.2.5-1 Loss of Condenser Vacuum at 2 inches per sec 151 1134 1159 1120 104 <0.08** a 14 6.0 LSCS-UFSAR

TABLE 15.0-2 (SHEET 3 OF 4) TABLE 15.0-2 REV. 0 - APRIL 1984 DURATION OF BLOWDOWN

PARA-GRAPH+

FIGURE

DESCRIPTION MAXIMUM NEUTRON FLUX  % NBR MAXIMUM DOME PRESSURE psig MAXIMUM VESSEL PRESSURE psig MAXIMUM STEAM LINE PRESSURE psig MAXIMUM CORE AVERAGE SURFACE HEAT FLUX % OF INITIAL

CPR FREQUENCY CATEGORY ff NO. OF VALVES 1st BLOW-DOWN DURATION OF BLOW-DOWN SEC fff 15.2.6 15.2.6-1 Loss of Auxiliary Power Transformer 103.9 1092 1103 1092 100.0 ~0.0 a 2 5.6 15.2.6 15.2.6-2 Loss of All Grid Connections 150.4 1135 1161 1121 101.8 <0.08** a 18 6.4 15.2.7 15.2.7-1 Loss of All Feedwater Flow 103.9 1094 1105 1094 100.0 ~0.0 a 2 5.5 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOWRATE 15.3.1 15.3.1-1 Trip of One Recirculation Pump Motor 104.0 1020 1058 994 100.0 ~0.0 a 0 0 15.3.1 15.3.1-2 Trip of Both Recirculation Pump Motors 103.9 1094 1107 1092 100.0 ~0.0 a 2 5.3 15.3.2 15.3.2-1 Fast Closure of One Main Recirc Valve - 30% /sec 103.9 1095 1108 1093 100.0 ~0.0 a 2 5.4 15.3.2 15.3.2-2 Fast Closure of Two Main Recirc Valves - 11% /sec 103.9 1095 1108 1099 100.0 ~0.0 a 6 5.3 15.3.3 15.3.3-1 Seizure of One Recirculation Pump 103.9 1107 1119 1101 100.2 ~0.0 c 6 5.6 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.2 15.4.2-1 Rod Withdrawal Error at Power 110.4 1020 1050 985 103.4 0.18 a 0 0 15.4.4 15.4.4-1 Startup of Idle Recriculation Loop 110.2 100.2 982 982 995 985 971 79.2 <0.18*** a 0 0

LSCS-UFSAR

TABLE 15.0-2 (SHEET 4 OF 4) TABLE 15.0-2 REV. 0 - APRIL 1984 DURATION OF BLOWDOWN

PARA-GRAPH+

FIGURE

DESCRIPTION MAXIMUM NEUTRON FLUX  % NBR MAXIMUM DOME PRESSURE psig MAXIMUM VESSEL PRESSURE psig MAXIMUM STEAM LINE PRESSURE psig MAXIMUM CORE AVERAGE SURFACE HEAT FLUX % OF INITIAL CPR FREQUENCY CATEGORY ff NO. OF VALVES 1st BLOW-DOWN DURATION OF BLOW-DOWN SEC fff 15.4.5 15.4.5-1 Fast Opening of One Main Recirc Valve - 30% /sec 281.8 980 1000 971 76.2 <0.18*** a 0 0 15.4.5 15.4.5-2 Fast Opening of Both Main Recirc Valves - 11% /sec 193.8 973 980 961 72.0 <0.18*** a 0 0 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 15.5.1-1 Inadvertent HPCS Pump Start 103.9 1020 1058 994 100.0 ~0.0 a 0 0

ff = Incidents of moderate freq; b = infrequent incidents; c = limiting faults. fff = Estimated value. + = Paragraphs denoted with suffix A indicate Reanalyses with ODYN Code (see Reference 2 of Subsection 15.1.2.6). * = ODYN results without adders. ** = Events obviously bounded by more severe transients, hence unique MCPR were not calculated, but a limiting calculation was used because there is no threat to the MCPR safety limit when the event is initiated from less than 100% power. *** = Events initiated from low power and therefore resulting MCPR is well above the MCPR safety limit.

LSCS-UFSAR TABLE 15.4-1 TABLE 15.4-1 REV. 10 - APRIL 1994 SEQUENCE OF EVENTS FOR CONTINUOUS ROD WITHDRAWAL DURING REACTOR STARTUP TIME (sec). EVENT 0 1. The reactor is critical and operating in the startup range. >0 2. The operator selects and withdraws an out-of-sequence control rod at the maximum normal drive speed of 3.6 ips.

~4 sec 3. The RWM fails to block the selection (selection error) and Continuous withdrawal (withdraw error) of the

out-of-sequence rod.

4-8 sec 4. The reactor scram is initiated by the intermediate range monitor (IRM) system or the average power range monitor system (APRM).

5-9 sec 5. The prompt power burst is terminated by a combination of Doppler and/or scram feedback.

10 sec 6. The transient is finally terminated by the scram of all rods, including the control rod being withdrawn. (Technical Specification scram insertion times are assumed, 5 seconds to 90% insertion.)

LSCS-UFSAR TABLE 15.4-2 TABLE 15.4-2 REV. 0 - APRIL 1984

SUMMARY

OF RESULTS FOR DETAILED AND POINT KINETICS EVALUATIONS OF CONTINUOUS ROD WITHDRAWAL IN THE STARTUP RANGE CASE CONTROL ROD WORTH (%k) _ hf(cal/gm)

P/A (c)^ h(cal/gm) 1 1.6 17.3 24.2 42.7 2 2.0 17.3 30.9 50.0 3 2.5 17.2 46.0 58.5 4 1.6 (a) 18.3 19.7 (b)56.2 5 1.6 (d) 18.3 19.7 59.6

_____________________________

(a) Detailed transient calculation. All other data reported are for point kinetics calculations.

(b) The P/A = 19.7 is the initial value. For the detailed analysis this value will decrease during the course of the transient since the power shape will flatten du e to Doppler feedback.

(c) P/A = global peaking factor (Radial x Axial).

(d) Point kinetics calculation with IRM initiated scram and 3-D simulator global peaking.

LSCS-UFSAR TABLE 15.4-3 TABLE 15.4-3 REV. 13 INPUT PARAMETERS AND INITIAL CONDITIONS FOR ROD WITHDRAWAL TRANSIENT AT POWER FOR INITIAL CORE Reactor Power, MWt 3323 Average Core Exposure, MWd/t 6000 Xenon State None Average Linear Heat Generation Rate, kw/ft 5.39 Maximum Linear Heat Generation Rate, kw/ft 13.48 Location of Maximum LHGR Bundle (21-38) Minimum CPR 1.298 Location of Minimum CPR Bundle (21-38) Maximum Worth Control Rod (26-35) Rod Withdrawal Speed, in/sec 3.6 Core Coolant Flow Rate #hrx10 6 108.5 Core Coolant Inlet Enthalpy, Btu/# 527.6 Core Average Steam Volume Fraction 0.371 Reactor Coolant Pressure, Average, Psia 1035 Control Rod Pattern Figure 15.4.2-1 RBM Trip Setpoint 106%

LSCS-UFSAR TABLE 15.4-4 TABLE 15.4-4 REV. 12 - MARCH 1998

(INTENTIONALLY LEFT BLANK)

LSCS-UFSAR TABLE 15.4-5 TABLE 15.4-5 REV. 14 - APRIL 2002 INCREMENTAL ROD WORTHS USING BANK-POSITION WITHDRAWAL SEQUENCE WORST CASE FOR EACH OF THE GIVEN ROD GROUPS (INITIAL CYCLE ANALYSES)

CORE CONDITION CONTROL ROD GROUP

  • BANKED AT NOTCH CONTROL ROD (X,Y) DROPS FROM-TO k BOC-1 Sequence A G1 through G4 W/D all others at 0 7 12 26-35 0 - 48 .004658 BOC-1 Sequence A G1 through G4 W/D all others at 0 8 12 26-43 0 - 48 .002518 BOC-1 Sequence A G1 through G4 W/D G5 through G8 at 12 G10 at 0 9 4 30-31 0 - 8 .002154 BOC-1 Sequence A G1 through G4 W/D G5 through G8 at 12 G9 at 0 10 4 22-31 0 - 8 .002141 NOTE: The following assumptions were made to ensure that the rod worths were conservatively high for the banke d-position withdrawal sequence:
a. BOC,
b. hot startup, c. no xenon.
  • For definition of rod groups, see Figures 4.3-27 and 4.3-28.

LSCS-UFSAR TABLE 15.4-6 TABLE 15.4-6 REV. 15, APRIL 2004 (Sheet 1 of 2) INPUT PARAMETERS AND INITIAL CONDITIONS: ROD DROP ACCIDENT DESIGN-BASIS ASSUMPTIONS I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level corresponding to 102% of rated core thermal power 3559 Mwt (Reference 24) B. Radial peaking factor 1.5 C. Fuel Damaged 770 rods (GE), 850 rods (FANP) D. Fraction of damaged fuel assumed to melt 0.0077 E. Fractions of fission products in damaged fuel assumed release to coolant Subsection 15.4.9.5 F. Iodine fractions (1) Organic (2) Elemental (3) Particulate 0

1 0 G. Release of activity to environment by nuclide (1) Case 1 (2) Case 2 (3) Case 3 Table 15.4-8 Table 15.4-10 Table 15.4-12 II. Data and assumptions used to estimate activity released A. Condenser leak rate (%/day) 1.0 B. Mechanical vacuum pump operating period (1) Case 1 (2) Case 2 (3) Case 3 NA 15 min continuous C. Condenser discharge rate via MVP (%/day)

(1) Case 1 (2) Case 2 (3) Case 3 0 3157 3157 LSCS-UFSAR TABLE 15.4-6 TABLE 15.4-6 REV. 12 (Sheet 2 of 2) INPUT PARAMETERS AND INITIAL CONDITIONS: ROD DROP ACCIDENT III. Dispersion data A. Exclusion Area Boundary and LPZ distances (m) Section 15.0.5, Item b B. Ground Level Release /Q's for time intervals of (1) 0 - 2 hr - EAB/LPZ (2) 2 - 8 hr - LPZ (3) 8 - 24 hr - LPZ (4) 1 - 4 day - LPZ (5) 4 - 30 day - LPZ 5.1E-4/1.0E-5 1.0E-5 6.7E-6 2.6E-6 6.5E-7 C. Elevated Release /Q's for time intervals of (1) 0 - 1/2 hr - EAB/LPZ (2) 1/2 - 2 hr - EAB/LPZ (3) 2 - 8 hr - LPZ (4) 8 - 24 hr - LPZ (5) 1 - 4 day - LPZ (6) 4 - 30 day LPZ 8.4E-5/8.9E-6 2.6E-6/1.6E-6 9.2E-7 5.5E-7 2.5E-7 8.2E-8 IV. Dose Data A. Method of dose calculation Reference 8 B. Dose conversion assumptions Reference 8 C. Peak activity concentrations in condenser (1) Case 1 (2) Case 2 (3) Case 3 Table 15.4-7 Table 15.4-9 Table 15.4-11 D. Doses Table 15.4-13

LSCS-UFSAR TABLE 15.4-7 TABLE 15.4-7 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 1 - NO MVP OPERATION)

AIRBORNE ACTIVITY IN CONDENSER (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 2.29E 03 2.28E 03 2.28E 03 2.27E 03 2.25E 03 2.21E 03 2.18E 03 2.08E 03 1.56E 03 1.28E 02 I-132 3.33E 03 3.10E 03 2.87E 03 1.83E 03 1.00E 03 2.99E 02 8.94E 01 2.39E 00 8.76E-10 0.0 I-133 4.78E 03 4.74E 03 4.70E 03 4.47E 03 4.18E 03 3.65E 03 3.19E 03 2.13E 03 1.87E 02 1.35E-07 I-134 5.20E 03 4.32E 03 3.54E 03 1.08E 03 2.22E 02 9.40E 00 3.97E-01 3.00E-05 0.0 0.0 I-135 4.51E 03 4.40E 03 4.28E 03 3.66E 03 2.96E 03 1.95E 03 1.28E 03 3.61E 02 1.84E-01 0.0 TOTAL I 2.01E 04 1.88E 04 1.77E 04 1.33E 04 1.06E 04 8.12E 03 6.74E 03 4.57E 03 1.74E 03 1.28E 02 KR-83M 2.81E 04 2.57E 04 2.34E 04 1.32E 04 6.21E 03 1.36E 03 2.99E 02 3.16E 00 4.41E-12 0.0 KR-85M 6.05E 04 5.84E 04 5.62E 04 4.45E 04 3.26E 04 1.75E 04 9.43E 03 1.47E 03 2.06E-02 0.0 KR-85 2.72E 03 2.72E 03 2.72E 03 2.72E 03 2.71E 03 2.71E 03 2.70E 03 2.69E 03 2.61E 03 2.00E 03 KR-87 1.15E 05 1.02E 05 8.87E 04 3.91E 04 1.31E 04 1.48E 03 1.67E 02 2.40E-01 0.0 0.0 KR-88 1.64E 05 1.55E 05 1.46E 05 1.01E 05 6.20E 04 2.33E 04 8.77E 03 4.67E 02 1.06E-05 0.0 KR-89 1.65E 05 7.72E 03 2.91E 02 8.26E-07 0.0 0.0 0.0 0.0 0.0 0.0 XE-131M 1.43E 03 1.42E 03 1.42E 03 1.42E 03 1.41E 03 1.39E 03 1.38E 03 1.33E 03 1.09E 03 1.84E 02 XE-133M 2.08E 04 2.07E 04 2.06E 04 2.02E 04 1.97E 04 1.86E 04 1.76E 04 1.50E 04 5.62E 03 1.15E 00 XE-133 4.98E 05 4.98E 05 4.97E 05 4.92E 05 4.87E 05 4.75E 05 4.64E 05 4.32E 05 2.82E 05 7.00E 03 XE-135M 8.99E 04 4.83E 04 2.49E 04 4.61E 02 2.27E 00 5.47E-05 1.32E-09 0.0 0.0 0.0 XE-135 6.43E 04 6.32E 04 6.20E 04 5.53E 04 4.74E 04 3.49E 04 2.57E 04 1.02E 04 4.07E 01 0.0 XE-137 3.65E 05 2.90E 04 1.92E 03 1.62E-04 6.00E-14 0.0 0.0 0.0 0.0 0.0 XE-138 3.96E 05 1.99E 05 9.58E 04 1.17E 03 3.30E 00 2.63E-05 2.09E-10 0.0 0.0 0.0 TOTAL NG 1.97E 06 1.21E 06 1.02E 06 7.72E 05 6.72E 05 5.77E 05 5.30E 05 4.63E 05 2.91E 05 9.18E 03 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these va lues should be increased by a factor 1.029 for all isotopes except I-131. I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-8 TABLE 15.4-8 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 1 - NO MVP OPERATION)

ACTIVITY RELEASED TO ENVIRONMENT (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 1.59E-02 2.38E-01 4.76E-01 1.90E 00 3.78E 00 7.50E 00 1.12E 01 2.18E 01 7.59E 01 2.24E 02 I-132 2.32E-02 3.35E-01 6.46E-01 2.09E 00 3.23E 00 4.20E 00 4.49E 00 4.61E 00 4.62E 00 4.62E 00 I-133 3.32E-02 4.96E-01 9.88E-01 3.85E 00 7.46E 00 1.40E 01 1.97E 01 3.28E 01 5.67E 01 5.91E 01 I-134 3.63E-02 4.98E-01 9.06E-01 2.20E 00 2.66E 00 2.77E 00 2.77E 00 2.77E 00 2.77E 00 2.77E 00 I-135 3.13E-02 4.64E-01 9.17E-01 3.39E 00 6.14E 00 1.02E 01 1.28E 01 1.64E 01 1.79E 01 1.79E 01 TOTAL I 1.40E-01 2.03E 00 3.93E 00 1.34E 01 2.33E 01 3.86E 01 5.09E 01 7.84E 01 1.58E 02 3.09E 02 KR-83M 1.96E-01 2.81E 00 5.37E 00 1.65E 01 2.43E 01 2.96E 01 3.07E 01 3.11E 01 3.11E 01 3.11E 01 KR-85M 4.21E-01 6.20E 00 1.22E 01 4.35E 01 7.54E 01 1.16E 02 1.38E 02 1.59E 02 1.63E 02 1.63E 02 KR-85 1.89E-02 2.83E-01 5.66E-01 2.26E 00 4.53E 00 9.04E 00 1.36E 01 2.70E 01 1.07E 02 7.03E 02 KR-87 8.05E-01 1.13E 01 2.12E 01 5.91E 01 7.89E 01 8.78E 01 8.88E 01 8.90E 01 8.90E 01 8.90E 01 KR-88 1.14E 00 1.67E 01 3.23E 01 1.09E 02 1.75E 02 2.41E 02 2.66E 02 2.80E 02 2.81E 02 2.81E 02 KR-89 1.28E 00 6.27E 00 6.51E 00 6.52E 00 6.52E 00 6.52E 00 6.52E 00 6.52E 00 6.52E 00 6.52E 00 XE-131M 9.90E-03 1.48E-01 2.97E-01 1.18E 00 2.36E 00 4.70E 00 7.01E 00 1.38E 01 4.99E 01 1.82E 02 XE-133M 1.44E-01 2.16E 00 4.31E 00 1.71E 01 3.37E 01 6.56E 01 9.58E 01 1.77E 02 4.64E 02 6.36E 02 XE-133 3.46E 00 5.19E 01 1.04E 02 4.13E 02 8.21E 02 1.62E 03 2.40E 03 4.64E 03 1.52E 04 3.46E 04 XE-135M 6.38E-01 7.15E 00 1.08E 01 1.47E 01 1.47E 01 1.47E 01 1.47E 01 1.47E 01 1.47E 01 1.47E 01 XE-135 4.47E-01 6.65E 00 1.32E 01 4.98E 01 9.25E 01 1.60E 02 2.11E 02 2.94E 02 3.50E 02 3.50E 02 XE-137 2.78E 00 1.57E 01 1.67E 01 1.68E 01 1.68E 01 1.68E 01 1.68E 01 1.68E 01 1.68E 01 1.68E 01 XE-138 2.82E 00 3.07E 01 4.54E 01 5.88E 01 5.90E 01 5.90E 01 5.90E 01 5.90E 01 5.90E 01 5.90E 01 TOTAL NG 1.42E 01 1.58E 02 2.73E 02 8.08E 02 1.40E 03 2.43E 03 3.35E 03 5.81E 03 1.68E 04 3.71E 04 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these va lues should be increased by a factor 1.029 for all isotopes except I-131. I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-9 TABLE 15.4-9 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 2 - MVP TRIPPED AT 15 MINUTES)

AIRBORNE ACTIVITY IN CONDENSER (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 2.24E 03 1.64E 03 1.64E 03 1.63E 03 1.62E 03 1.59E 03 1.57E 03 1.50E 03 1.12E 03 9.18E 01 I-132 3.25E 03 2.23E 03 2.07E 03 1.32E 03 7.20E 02 2.15E 02 6.44E 01 1.72E 00 6.31E-10 0.0 I-133 4.68E 03 3.41E 03 3.39E 03 3.22E 03 3.01E 03 2.63E 03 2.30E 03 1.53E 03 1.35E 02 9.69E-08 I-134 5.08E 03 3.11E 03 2.55E 03 7.79E 02 1.60E 02 6.77E 00 2.86E-01 2.16E-05 0.0 0.0 I-135 4.41E 03 3.17E 03 3.08E 03 2.63E 03 2.13E 03 1.40E 03 9.19E 02 2.60E 02 1.33E-01 0.0 TOTAL I 1.97E 04 1.36E 04 1.27E 04 9.58E 03 7.64E 03 5.85E 03 4.85E 03 3.29E 03 1.26E 03 9.18E 01 KR-83M 2.75E 04 1.85E 04 1.68E 04 9.54E 03 4.47E 03 9.80E 02 2.15E 02 2.27E 00 3.17E-12 0.0 KR-85M 5.92E 04 4.20E 04 4.04E 04 3.20E 04 2.35E 04 1.26E 04 6.79E 03 1.06E 03 1.49E-02 0.0 KR-85 2.66E 03 1.96E 03 1.96E 03 1.95E 03 1.95E 03 1.95E 03 1.95E 03 1.94E 03 1.88E 03 1.44E 03 KR-87 1.13E 05 7.31E 04 6.38E 04 2.82E 04 9.46E 03 1.07E 03 1.20E 02 1.73E-01 0.0 0.0 KR-88 1.61E 05 1.12E 05 1.05E 05 7.28E 04 4.47E 04 1.68E 04 6.32E 03 3.36E 02 7.60E-06 0.0 KR-89 1.61E 05 5.56E 03 2.09E 02 5.95E-07 0.0 0.0 0.0 0.0 0.0 0.0 XE-131M 1.39E 03 1.03E 03 1.02E 03 1.02E 03 1.01E 03 1.00E 03 9.92E 02 9.59E 02 7.81E 02 1.32E 02 XE-133M 2.03E 04 1.49E 04 1.49E 04 1.46E 04 1.42E 04 1.34E 04 1.27E 04 1.08E 04 4.05E 03 8.25E-01 XE-133 4.87E 05 3.58E 05 3.58E 05 3.54E 05 3.50E 05 3.42E 05 3.34E 05 3.11E 05 2.03E 05 5.04E 03 XE-135M 8.79E 04 3.48E 04 1.79E 04 3.32E 02 1.63E 00 3.94E-05 9.51E-10 0.0 0.0 0.0 XE-135 6.30E 04 4.55E 04 4.46E 04 3.98E 04 3.41E 04 2.51E 04 1.85E 04 7.35E 03 2.93E 01 0.0 XE-137 3.57E 05 2.09E 04 1.38E 03 1.17E-04 4.32E-14 0.0 0.0 0.0 0.0 0.0 XE-138 3.87E 05 1.44E 05 6.89E 04 8.44E 02 2.38E 00 1.89E-05 1.50E-10 0.0 0.0 0.0 TOTAL NG 1.93E 06 8.72E 05 7.35E 05 5.55E 05 4.84E 05 4.15E 05 3.82E 05 3.34E 05 2.10E 05 6.61E 03 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these valu es should be increased by a factor 1.029 for all isotopes except I-131. I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-10 TABLE 15.4-10 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 2 - MVP TRIPPED AT 15 MINUTES)

ACTIVITY RELEASED TO ENVIRONMENT (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 4.96E 01 6.41E 02 6.41E 02 6.42E 02 6.43E 02 6.46E 02 6.49E 02 6.56E 02 6.95E 02 8.02E 02 I-132 7.23E 01 9.04E 02 9.04E 02 9.05E 02 9.06E 02 9.07E 02 9.07E 02 9.07E 02 9.07E 02 9.07E 02 I-133 1.04E 02 1.34E 03 1.34E 03 1.34E 03 1.34E 03 1.34E 03 1.35E 03 1.36E 03 1.38E 03 1.38E 03 I-134 1.13E 02 1.35E 03 1.35E 03 1.35E 03 1.35E 03 1.35E 03 1.35E 03 1.35E 03 1.35E 03 1.35E 03 I-135 9.78E 01 1.25E 03 1.25E 03 1.25E 03 1.25E 03 1.26E 03 1.26E 03 1.26E 03 1.26E 03 1.26E 03 TOTAL I 4.37E 02 5.48E 03 5.48E 03 5.48E 03 5.49E 03 5.50E 03 5.51E 03 5.53E 03 5.59E 03 5.70E 03 KR-83M 6.11E 02 7.58E 03 7.58E 03 7.59E 03 7.60E 03 7.60E 03 7.60E 03 7.60E 03 7.60E 03 7.60E 03 KR-85M 1.31E 03 1.67E 04 1.67E 04 1.67E 04 1.68E 04 1.68E 04 1.68E 04 1.68E 04 1.68E 04 1.68E 04 KR-85 5.89E 01 7.62E 02 7.62E 02 7.63E 02 7.65E 02 7.68E 02 7.71E 02 7.81E 02 8.38E 02 1.27E 03 KR-87 2.51E 03 3.06E 04 3.06E 04 3.07E 04 3.07E 04 3.07E 04 3.07E 04 3.07E 04 3.07E 04 3.07E 04 KR-88 3.57E 03 4.49E 04 4.49E 04 4.50E 04 4.50E 04 4.51E 04 4.51E 04 4.51E 04 4.51E 04 4.51E 04 KR-89 4.00E 03 1.82E 04 1.82E 04 1.82E 04 1.82E 04 1.82E 04 1.82E 04 1.82E 04 1.82E 04 1.82E 04 XE-131M 3.09E 01 4.00E 02 4.00E 02 4.00E 02 4.01E 02 4.03E 02 4.04E 02 4.09E 02 4.35E 02 5.30E 02 XE-133M 4.50E 02 5.81E 03 5.81E 03 5.82E 03 5.84E 03 5.86E 03 5.88E 03 5.94E 03 6.15E 03 6.27E 03 XE-133 1.08E 04 1.40E 05 1.40E 05 1.40E 05 1.40E 05 1.41E 05 1.41E 05 1.43E 05 1.50E 05 1.64E 05 XE-135M 1.99E 03 1.96E 04 1.96E 04 1.96E 04 1.96E 04 1.96E 04 1.96E 04 1.96E 04 1.96E 04 1.96E 04 XE-135 1.40E 03 1.79E 04 1.79E 04 1.79E 04 1.80E 04 1.80E 04 1.80E 04 1.81E 04 1.81E 04 1.81E 04 XE-137 8.68E 03 4.50E 04 4.50E 04 4.50E 04 4.50E 04 4.50E 04 4.50E 04 4.50E 04 4.50E 04 4.50E 04 XE-138 8.79E 03 8.42E 04 8.42E 04 8.42E 04 8.42E 04 8.42E 04 8.42E 04 8.42E 04 8.42E 04 8.42E 04 TOTAL NG 4.42E 04 4.31E 05 4.31E 05 4.32E 05 4.32E 05 4.33E 05 4.33E 05 4.35E 05 4.43E 05 4.58E 05 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these valu es should be increased by a factor 1.029 for all isotopes except I-131.

I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-11 TABLE 15.4-11 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 3 - MVP OPERATING CONTINUOUSLY)

AIRBORNE ACTIVITY IN CONDENSER (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 2.24E 03 1.64E 03 1.18E 03 1.64E 02 1.17E 01 5.98E-02 3.06E-04 4.09E-11 0.0 0.0 I-132 3.25E 03 2.23E 03 1.49E 03 1.32E 02 5.19E 00 8.07E-03 1.25E-05 4.71E-14 0.0 0.0 I-133 4.68E 03 3.41E 03 2.44E 03 3.22E 02 2.17E 01 9.86E-02 4.48E-04 4.19E-11 0.0 0.0 I-134 5.08E 03 3.11E 03 1.84E 03 7.80E 01 1.16E 00 2.54E-04 5.57E-08 0.0 0.0 0.0 I-135 4.41E 03 3.17E 03 2.22E 03 2.64E 02 1.54E 02 5.25E-02 1.79E-04 7.10E-12 0.0 0.0 TOTAL I 1.97E 04 1.36E 04 9.17E 03 9.59E 02 5.52E 01 2.19E-01 9.45E-04 8.99E-11 0.0 0.0 KR-83M 2.75E 04 1.85E 04 1.21E 04 9.55E 02 3.22E 01 3.68E-02 4.19E-05 6.22E-14 0.0 0.0 KR-85M 5.92E 04 4.20E 04 2.91E 04 3.21E 03 1.70E 02 4.74E-01 1.32E-03 2.89E-11 0.0 0.0 KR-85 2.66E 03 1.96E 03 1.41E 03 1.96E 02 1.41E 01 7.31E-02 3.79E-04 5.30E-11 0.0 0.0 KR-87 1.13E 05 7.31E 04 4.59E 04 2.82E 03 6.83E 01 4.00E-02 2.35E-05 4.73E-15 0.0 0.0 KR-88 1.61E 05 1.12E 05 7.56E 04 7.29E 03 3.22E 02 6.30E-01 1.23E-03 9.18E-12 0.0 0.0 KR-89 1.61E 05 5.56E 03 1.51E 02 5.96E-08 0.0 0.0 0.0 0.0 0.0 0.0 XE-131M 1.39E 03 1.03E 03 7.38E 02 1.02E 02 7.33E 00 3.76E-02 1.93E-04 2.62E-11 0.0 0.0 XE-133M 2.03E 04 1.49E 04 1.07E 04 1.46E 03 1.02E 02 5.03E-01 2.48E-03 2.95E-10 0.0 0.0 XE-133 4.87E 05 3.58E 05 2.57E 05 3.55E 04 2.53E 03 1.28E 01 6.51E-02 8.51E-09 0.0 0.0 XE-135M 8.79E 04 3.48E 04 1.29E 04 3.33E 01 1.18E-02 1.48E-09 0.0 0.0 0.0 0.0 XE-135 6.30E 04 4.55E 04 3.21E 04 3.98E 03 2.46E 02 9.42E-01 3.60E-03 2.01E-10 0.0 0.0 XE-137 3.57E 05 2.09E 04 9.94E 02 1.17E-05 0.0 0.0 0.0 0.0 0.0 0.0 XE-138 3.87E 05 1.44E 05 4.96E 04 8.45E 01 1.72E-02 7.10E-10 0.0 0.0 0.0 0.0 TOTAL NG 1.93E 06 8.72E 05 5.29E 05 5.56E 04 3.49E 03 1.56E 01 7.44E-02 9.12E-09 0.0 0.0 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these valu es should be increased by a factor 1.029 for all isotopes except I-131.

I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-12 TABLE 15.4-12 REV. 15, APRIL 2004 CONTROL ROD DROP ACCIDENT - DESIGN BASIS (CASE 3 - MVP OPERATING CONTINUOUSLY)

ACTIVITY RELEASED TO ENVIRONMENT (CURIES)

ISOTOPE 1-MIN 15-MIN 30-MIN 2-HOUR 4-HOUR 8-HOUR 12-HOUR 1-DAY 4-DAY 30-DAY I-131 4.96E 01 6.41E 02 1.10E 03 2.12E 03 2.27E 03 2.28E 03 2.28E 03 2.28E 03 2.28E 03 2.28E 03 I-132 7.23E 01 9.04E 02 1.51E 03 2.61E 03 2.72E 03 2.72E 03 2.72E 03 2.72E 03 2.72E 03 2.72E 03 I-133 1.04E 02 1.34E 03 2.29E 03 4.35E 03 4.64E 03 4.66E 03 4.66E 03 4.66E 03 4.66E 03 4.66E 03 I-134 1.13E 02 1.35E 03 2.14E 03 3.24E 03 3.29E 03 3.29E 03 3.29E 03 3.29E 03 3.29E 03 3.29E 03 I-135 9.78E 01 1.25E 03 2.13E 03 3.94E 03 4.17E 03 4.18E 03 4.18E 03 4.18E 03 4.18E 03 4.18E 03 TOTAL I 4.37E 02 5.48E 03 9.16E 03 1.63E 04 1.71E 04 1.71E 04 1.71E 04 1.71E 04 1.71E 04 1.71E 04 KR-83M 6.11E 02 7.58E 03 1.25E 04 2.12E 04 2.19E 04 2.20E 04 2.20E 04 2.20E 04 2.20E 04 2.20E 04 KR-85M 1.31E 03 1.67E 04 2.83E 04 5.14E 04 5.42E 04 5.43E 04 5.43E 04 5.43E 04 5.43E 04 5.43E 04 KR-85 5.89E 01 7.62E 02 1.31E 03 2.52E 03 2.70E 03 2.72E 03 2.72E 03 2.72E 03 2.72E 03 2.72E 03 KR-87 2.51E 03 3.06E 04 4.99E 04 8.03E 04 8.23E 04 8.23E 04 8.23E 04 8.23E 04 8.23E 04 8.23E 04 KR-88 3.57E 03 4.49E 04 7.53E 04 1.33E 05 1.39E 05 1.39E 05 1.39E 05 1.39E 05 1.39E 05 1.39E 05 KR-89 4.00E 03 1.82E 04 1.87E 04 1.87E 04 1.87E 04 1.87E 04 1.87E 04 1.87E 04 1.87E 04 1.87E 04 XE-131M 3.09E 01 4.00E 02 6.87E 02 1.32E 03 1.42E 03 1.42E 03 1.42E 03 1.42E 03 1.42E 03 1.42E 03 XE-133M 4.50E 02 5.81E 03 9.98E 03 1.91E 04 2.05E 04 2.06E 04 2.06E 04 2.06E 04 2.06E 04 2.06E 04 XE-133 1.08E 04 1.40E 05 2.40E 05 4.61E 05 4.94E 05 4.96E 05 4.96E 05 4.96E 05 4.96E 05 4.96E 05 XE-135M 1.99E 03 1.96E 04 2.68E 04 3.11E 04 3.11E 04 3.11E 04 3.11E 04 3.11E 04 3.11E 04 3.11E 04 XE-135 1.40E 03 1.79E 04 3.05E 04 5.71E 04 6.07E 04 6.09E 04 6.09E 04 6.09E 04 6.09E 04 6.09E 04 XE-137 8.68E 03 4.50E 04 4.72E 04 4.73E 04 4.73E 04 4.73E 04 4.73E 04 4.73E 04 4.73E 04 4.73E 04 XE-138 8.79E 03 8.42E 04 1.13E 05 1.29E 05 1.29E 05 1.29E 05 1.29E 05 1.29E 05 1.29E 05 1.29E 05 TOTAL NG 4.42E 04 4.31E 05 6.54E 05 1.05E 06 1.10E 06 1.11E 06 1.11E 06 1.11E 06 1.11E 06 1.11E 06 NOTE: For power uprate to 3559 MWt (102% of 3489 MWt), these valu es should be increased by a factor 1.029 for all isotopes except I-131.

I-131 should be increased by a factor of 1.129 (Reference 24).

LSCS-UFSAR TABLE 15.4-13 TABLE 15.4-13 REV. 14 - APRIL 2002 CONTROL ROD DROP ACCIDENT RADIOLOGICAL EFFECTS DOSE (REM) (Note 1) EAB (2HR) LPZ (30 DAY)

Case Mechanical Vacuum Pump Operation Thyroid Whole Body Thyroid Whole Body 1 None 0.72 0.052 0.17 0.0030 2 15 Minutes 41.3 5.9 4.5 0.63 3 Continuous 72.1 8.7 8.8 0.99 10 CFR 100 Limits 300 25 300 25

Note 1: For 105% licensed core thermal power uprate to 3489 MWt, the most limiting dose consequence (Case 2) was evaluated at 3559 MWt (i.e., 102% of 3489 MWt). It was determined that th e whole body dose increases by 1.029 and the thyroid dose by 1.129 (See Section 15.4.9.5). This results in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB thyroid dose of 46.7 rem and a whole body dose of 6.07 rem for Case 2. (Reference 24)

LSCS-UFSAR 15.7-1 REV. 14, APRIL 2002 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 15.7.1 Radioactive Gas Waste System Leak or Failure Radioactive Gas Waste System Leaks or Failu res are not analyzed for reload cores.

The information presented in this Section 15.7.1 is retained for historical documentation only, since it is part of the original design basis of the Station. This information is based on pre-power uprate normal operating design basis source terms. The existing normal operating de sign basis source terms are bounding for power uprate. Therefore, 105% power up rate (i.e., 105% of 3323 MWt) has no impact on the Section 15.7.

1 analyses (Reference 12).

The NRC Standard Review Plan (SRP)

NUREG-0800, Revision 1 dated July 1981 (formerly NUREG 075/087) deleted this section from the SRP.

The following radioactive gas waste system components were examined under severe failure mode conditions for effects on the plant safety profile:

a. main condenser off-gas treatment system failure,
b. failure of air ejector lines, and
c. malfunction of main turbine gland sealing system.

15.7.1.1 Main Condenser Gas Treatment System Failure

15.7.1.1.1 Identification of Causes and Frequency Classification Those events which could conceivably cause a gross failure of a charcoal adsorber tank, a prefilter vessel, or a holdup pipeline in the off-gas treatment system are:

a. a seismic occurrence - greater than design basis;
b. a hydrogen detonation which ruptures the system pressure boundary;
c. a fire in the filter assemblies; and
d. failure of adjacent equipme nt, which could subsequently compromise off-gas equipment.

A seismic occurrence is consid ered to be the most probable and severe event which the system is designed to accommodate. Nevertheless, the seismic failure is the only conceivable event which could cause significant hardware damage that might LSCS-UFSAR 15.7-1a REV. 14, APRIL 2002 result in a radiological release. The eq uipment and piping are designed to contain any hydrogen-oxygen detonation which has a reasonable probability of occurring. A detonation is not considered as a possible failure mode. The decay heat on the filters is insignificant and cannot serve as an ignition source for the filters. It can be easily accommodated inherently by the system and certainly by the available air flows. The system is reasonably isolated from other systems or components which could cause any serious interaction or failure.

LSCS-UFSAR 15.7-2 REV. 13 Thus the only credible event which could result in the release of significant activity to the environment is an earthquake that causes building damage and subsequent rupture of off-gas components from falling building debris. Even though the off-gas system is designed to uniform building code seismic requirements, an event more severe than the design requirements is arbitrarily assumed to occur, resulting in the failure of the off-gas system. Failure of the off-gas treatment system is postulated to occur with the frequency of a limiting fault. The design basis, description, and performance evaluation of the subject system is given in Section 11.3. 15.7.1.1.2 Sequence of Events and Systems Operation Gross failure of the off-gas system results in the activation of area radiation alarms which alert the plant personnel. When the station vent stack radiation monitor reaches its limit, the off- gas system is manually isolated from the condenser, resulting in a high backpressure on the condenser and a reactor scram. The postulated failures which can release radioactivity into their respective building atmospheres include: a. Adsorber tanks and prefilter vessels release to the off-gas building. b. Delay line pipe failure is assumed to release radioactivity from the inlet end into the turbine building. The sequence of events following equipment failure is as follows: Approximate Elapsed Time Events 0 sec Event begins - system fails 0 sec Noble gases are released < 1 min Area radiation alarms alert plant personnel LSCS-UFSAR 15.7-3 REV. 13

< 1 min Operator actions begin with: (a) initiation of appropriate system isolations (b) manual scram actuation (c) assurance of reactor shutdown cooling.

Normal operator actions a ssure evacuation of the affected radiation areas and assure isolation of the turbine building via its HVAC dampers. The turbine building HVAC flow is then exhausted thro ugh the station vent stack. Only 2 minutes is needed to accomplish these operator actions.

In analyzing the postulated off-gas system failure, no credit is taken for the operation of plant and reactor protection syst ems, or of engineered safety features. Credit is taken for functioning of normally operating plant instruments and controls and other systems only to the following extent:

a. capability to detect the failure itself - indicated by an alarmed increase in radioactivity levels seen by area radiation monitoring system, by an alarmed loss of flow in the off-gas system, or by an alarmed increase in activity at the station vent stack; b. capability to isolate the off-gas system and shut down the reactor; and
c. operational indicators and annunciators in the main control room. The seismic event, which is assumed to occur beyond the present plant design basis for non-safety equipment, will cause the tripping of turbine or will lead to a load rejection. This initiates a scram and negates the need to postulate operator actions for a reactor shutdown.

15.7.1.1.3 Core and System Performance The postulated failure results in a system isolation necessitating reactor shutdown because of loss of vacuum in the main cond enser. This transient has been analyzed in Subsection 15.2.5. Otherw ise, this auxiliary system d oes not directly affect the reactor core nor the power cycle systems but is coupled only through operator alarms in the control room.

The iodine activity leaving the off-gas recombiner has been assumed to be entirely retained in the first charcoal tank. Thus failure of this tank results in the highest potential iodine release. Iodine absorbs strongly to charcoal. A conservative LSCS-UFSAR 15.7-4 REV. 13 evaluation leads to an assumption of the re lease of 1% of the iodine in the first charcoal tank.

Of primary importance in determining the potential release is the inventory of radioactive products in the equipment pieces before failure. The fractional releases from the equipment, after normal operation, are indicated in Table 15.7-2. These release fractions were used for analysis pu rposes for both the design- basis analysis and the realistic analysis. Table 15.7-3 lists the isotope inventories contained in the various components of the off-gas system.

The transport pathway to the environment co nsists of a release of fission products from the failed equipment through the build ing ventilation system to the station vent stack.

15.7.1.1.4 Barrier Performance The postulated failure is the rupture of the off-gas system pressure boundary. The events described here occur outside cont ainment, hence do not involve barrier integrity. No credit is taken for performa nce of secondary barriers, except to the extent inherent in the assumed equipment release fractions discussed herein.

15.7.1.1.5 Radiological Consequences Separate radiological analyses are provided for the "design basis" case and for the "realistic" case as follows:

a. The "design basis analysis" uses conservative assumptions considered acceptable to the NRC for determining adequacy of plant design to meet 10 CFR 100 guidelines.
b. The "realistic analysis" uses engineering inputs and field experience (actual meteorological data for instance), to determine the radiological consequences.

Both analyses assume the following equipment characteristics with respect to the retention or passage of radioactive solid daughter products prior to initiation of the seismic event that causes failure of the off-gas equipment:

a. Off-gas condenser and water separator - 100% retained and continuously washed out with condensate.
b. Delay line pipe - 60% retained and continuously washed out with condensate.

LSCS-UFSAR 15.7-5 REV. 14, APRIL 2002

c. Cooler condenser and moisture separator - 100% retained and continuously washed out with condensate.
d. Prefilter - 100% retained, element changed annually.
e. Charcoal adsorbers - 100% retained.
f. After filter - 100% retained, element changed annually.

Components not listed are assumed to have zero retention of solid daughter products.

The differences in system performance in the postulated accidents are due to differences in the magnitude of the source term for the equipment and the assumptions relative to release fractions. The same analytical techniques were used for these separate analyses.

Table 15.7-1 lists the paramete rs used in the analyses.

Design-Basis Analysis There are no specific NRC guidelines for this analysis. Nevertheless, an evaluation has been performed to provide comparative results for this event.

The activity in the off-gas system is based on the following initial conditions: 2 scfm air inleakage, and 100,000 µCi/sec noble gas release after 30 minute delay for a period of 11 months, followed by 1 month of 350,000 µCi/sec at 30 minutes.

Additionally, the following assumptions were used with respect to equipment failures:

a. Charcoal adsorber tanks - these tanks are assumed to fail circumferentially from falling concrete that tears 50% of the circumference, or that creates a shearing load resulting in the same size break in the tank. These failures would result in no more than 10% to 15% of the ch arcoal (carbon) being displaced from the vessel. Iodine is stro ngly bonded to the charcoal and would not be expected to be remo ved by exposure to the air. Nevertheless, to be conservative, the assumption is made that 1% of the iodine activity contained in the adsorber tank is

released to the vault where the tanks are located. Additionally, the conservative assumption is made that 1% of the solid daughter products retained in the charcoal is also released.

LSCS-UFSAR 15.7-6 REV. 14, APRIL 2002 Measurements made at KRB indica te that off-gas is about 30%

richer in Kr than air. Therefore, if this carbon is exposed to air, it will eventually reach equilibri um with the noble gases in the air. However, the first few inches of carbon will blanket the underlying carbon from the air. A 10% loss of noble gas activity from a failed vessel is conservati ve because of the small fraction of carbon exposed to the air.

b. Prefilters - Because of the design features of the prefilter vessel (approximately 24-inch diameter, 4-foot height, 350 psig design pressure, 1/2-inch wall thickness and collapsible filter media), a failure mechanism cannot be postulated that will result in emission of filter media or daught er products from this vessel. However, to illustrate the consequences of a radioactivity loss from this vessel, 1% release of particulate activity is assumed.
c. Delay line pipe - Pipe rupture and depressurization of the pipe is considered. For the design-bas is analysis, 100% of the noble gases and all of the remaining solid daughters after the 60% washout are assumed to be released.
d. Piping - It is assumed that the seismic event causing the pipe failure is accompanied by a reactor isolation that stops steam flow to the steam jet air ejecto rs. Therefore, the resulting release from failed piping is not significant compared to those failures previously considered.

Results The calculated exposure for the design-basis analysis is tabulated in Table 15.7-4.

The doses are well within the 10 CFR 100 guidelines.

A 105% power uprate to 3489 MWt has no i mpact on either the gaseous radwaste system or the design basis radioactive gas source term. Therefore, the results presented in this analysis are not impacted by power uprate (Reference 12).

Realistic Analysis The realistic analysis is based on an engineered but still conservative assessment of this event. The specific models, assumptions, and the program used for computer evaluation are described in Reference 1. Sp ecific values of parameters used in the evaluation are presented in Table 15.7-1.

The activity in the off-gas system is based on the following initial conditions:

a. 18.5 scfm air inleakage, and
b. 100,000 µCi/sec noble gas after 30-minute delay.

LSCS-UFSAR 15.7-7 REV. 13 Additionally, the following assumptions were used with respect to equipment failures:

a. Charcoal adsorber tanks - The same tank failure mechanisms are assumed as those used in th e design-basis analysis. The release fractions for radionuclides are the same, except for the solid daughters. No logical mechanism exists for any of the solid daughter products formed and retained within the micropore structure of the carbon to be rele ased. Hence, no such release is assumed for the realistic analysis.
b. Prefilters - Because of the design features of the prefilter vessel, approximately 24-inch diameter, 4-foot height, 350-psig design pressure, 1/2-inch wall thickness, and collapsible filter media, a failure mechanism cannot be postulated that will result in emission of filter media or daught er products from this vessel. However, to illustrate the consequences of a radioactivity loss from this vessel, 1% release of particulate activity is assumed.
c. Delay line pipe - Pipe rupture and depressurization of the pipe is considered. Normally, the pipe will operate at less than 16 psia and depressurize to 14.7 psia. The possible loss of solid daughters and noble gases is conservatively taken as 20%. The model used assumes retention and washout of 60% of the particulate daughters for the calculation of the holdup pipe inventory.
d. Piping - It is assumed that the seismic event causing the pipe failure is accompanied by a reactor isolation, stopping steam flow to the steam jet air ejecto rs. Therefore, the resulting release from failed piping is not significant compared to those failures previously considered.

The release of activity to the environment is determined by applying the release fractions in Table 15.7-2 to the inventorie s shown in Table 15.7-3. The calculated exposures are presented in Table 15.7-4.

The only credible failure that could result in loss of carbon from the vessels is the failure of the concrete structure surrounding the vessel. A circumferential failure of the vessel could result from concrete falling on the vessel in either of two ways:

a. Bending Load - The vessel supported in the center and loaded on each end. This could result in a tear around 50% of the circumference.

LSCS-UFSAR 15.7-8 REV. 14, APRIL 2002 b. Shearing Load - The vessel being supported and loaded near the same point from above.

In either case, no more than 10-15% of the carbon would be displaced from the vessel. Iodine is strongly bonded to the charcoal and would not be expected to be removed by exposure to the air. However, a conservative assumption is made, that 1% of the iodine activity contained in the absorber tanks is released to the vault containing the off-gas equipment.

Measurements made at KRB indicate that off-gas is about 30% richer in Kr than air. Therefore, if this carb on is exposed to air, it w ill eventually reach equilibrium with the noble gases in the air. However, the first few inches of carbon will blanket the underlying carbon from the air. A 10% loss of noble gas activity from a failed vessel is conservative because of the small fraction of carbon exposed to the air.

There is no reason to believe that any of the solid daughter products formed and retained within the micropore structure of the carbon will be released. Hence no such release is assumed for the realistic analysis.

Results The calculated exposures for the realistic analysis are presented in Table 15.7-4. Note that station 2-year observations were used for the /Q's used for the realistic analysis.

A 105% power uprate to 3489 MWt has no i mpact on either the gaseous radwaste system or the design basis radioactive gas source term. Therefore, the results presented in this analysis are not impacted by power uprate (Reference 12).

15.7.1.2 Failure of Air Ejector Lines

15.7.1.2.1 Identification of Causes and Frequency Classification Failure of the air ejector inlet line upstream of its isolation valve to the condenser is arbitrarily assumed to occur with the frequency of a limiting fault. This line is designed to contain an explosion, theref ore an explosion is not considered as a possible cause of failure.

15.7.1.2.2 Sequence of Ev ents and Systems Operation It is assumed that the line leading to the steam jet air ejector fails near the condenser. This results in activity no rmally processed by the off-gas treatment system being discharged directly to the turbine building and subsequently through the ventilation system to the environment. This failure results in a loss-of-flow signal to the off-gas system, thus closing the condenser isolation valve downstream of the assumed break.

LSCS-UFSAR 15.7-9 REV. 14, APRIL 2002 The operator initiates a normal shutdown of the reactor within 30 minutes to reduce the gaseous activity being discharged. A loss of condenser vacuum will result in a turbine trip and reactor shutdown.

15.7.1.2.3 Core and System Performance

This auxiliary system does not directly affect the reactor core nor the power cycle systems but is coupled only through operator alarms in the contro l room. Failure of the air ejector lines has no applicabl e effect on NSSS safety performance.

15.7.1.2.4 Barrier Analysis This failure occurs outside the containment, hence it does not involve barrier integrity.

15.7.1.2.5 Radiological Consequences Specific regulatory guide calculation me thod and assumptions are not available, therefore only the realistic basis analys is is provided. The specific models, assumptions, and computer program are discussed in Refe rence 1. Table 15.7-5 itemizes those parametric values applicable to the realistic analysis.

The radiological exposures have been eval uated for the meteorological conditions defined in Subsection 15.0.5. There is no fuel damage as a consequence of this event, therefore the only activity released to the environment is that associated with the fluid processed by the off-gas syst em. This activity is discharged to the turbine building ventilation system and subsequently to the environment via the station vent stack. Because operator action is assumed to occur within 30 minutes, an uncontrolled release to the environment for this time period has been assumed. No credit is taken for plateout of iodine in the turbine building.

The activity released is shown in Table 15.7-6, and the calculated exposures are given in Table 15.7-7.

A 105% power uprate to 3489 MWt has no i mpact on either the gaseous radwaste system or the design basis radioactive gas source term. Therefore, the results presented in this analysis are not impacted by power uprate (Reference 12).

15.7.1.3 Malfunction of Tu rbine Gland Sealing System 15.7.1.3.1 Identification of Causes and Frequency Classification Failure of various components of the turbine gland sealing system can lead to system malfunction.

LSCS-UFSAR 15.7-9a REV. 14, APRIL 2002 The source for sealing steam for the turbine gland seals is a separate low-radioactivity steam drum on an evaporator unit that is heated by main steam. The leakage from the turbine gland sealing syst em is therefore low-radioactivity steam for normally anticipated leaks. Those le aks resulting from component malfunctions which invalidate or LSCS-UFSAR 15.7-10 REV. 17, APRIL 2008 partially compromise this source of low-radioactivity sealing steam are discussed below. This event is categorized as a limiting fault.

15.7.1.3.2 Sequence of Ev ents and Systems Operation Loss of Sealing Steam The instrumentation on the steam seal header detects low sealing steam pressure. The steam seal evaporator can be bypasse d manually, and the main steam supply to the steam seal header can be connected.

Assuming all sources of sealing steam are lost during the postulated malfunction, the following events will take place. In the high-pressure turbine glands, high-pressure steam will filter through the glands into the steam seal exhaust header. In the low-pressure turbine glands, air will be drawn into the sealing glands. The action of cool air quenching the hot turbine shaft will cause excessive shaft vibration.

Water Induction to Turbine Shaft Seal Glands During normal operation, the level contro l system on the steam seal evaporator guards against a high water level by alarm. Automatic control then closes off the condensate quality water feedline, which causes the water level in the steam seal evaporators to lower. In case the high water level situation is not alleviated, the motor-operated valve on the SSE condensate header side of the evaporators will be closed by the high level control. This prohibits water intrusion into the turbine glands and the turbine casing.

Loss of Vacuum in the Gland Steam Condenser During normal operation, noncondensibles are removed from the gland steam condenser by one of two gland steam conden ser blowers. In the event this blower malfunctions, the backup blower is manually started to take over the gas removal requirements. Assuming loss of both blowers, vacuum will be lost in the gland steam condenser. The pressure in the glan d steam exhaust header will increase to greater than atmospheric, thus causing sealing steam to leak to the turbine building through the turbine glands.

Loss of Cooling Medium for Gland Steam Condenser During normal operation, main condensa te provides the cooling medium for the gland steam condenser. In the event th is cooling water capacity is lost, the exhausted sealing steam will not be condense

d. This results in the same situation as loss of vacuum described above and causes a pressure buildup in the exhaust header of the sealing steam system with consequent leakage of sealing steam.

LSCS-UFSAR 15.7-11 REV. 17, APRIL 2008 Loss of Sealing Steam to Control, Intercept, and Stop Valves During normal operation, the stems of the intercept valves, stop valves, and control valves are sealed by steam from the turbin e gland sealing system. In the event of a sealing steam malfunction, the valves will leak main steam and reheated steam.

Description of Operator Action No operator action except to manually start the redundant gland steam condenser blower or an orderly shutdown is required for any of the previous malfunctions in the turbine gland sealing system.

15.7.1.3.3 Core and System Performance The plant is operating at full power prior to failure of some component of the normal operation gland sealing system. There are three means of providing heated steam to the steam seal evaporators. Main steam is used by the evaporators during startup and low load operation. An extraction steamline provides the evaporators with heating steam during normal and high-load operation. There is also a supply line from an auxiliary boiler which can provide steam to the evaporators during testing and startup.

In the event of the occurrence of loss of sealing steam, the action of cool air being drawn into the sealing glands and the resultant shaft vibration will cause tripping of the turbine-generator by the excessive shaft vibration trip. This tripping mechanism is independent of operator ac tion and, consequently, will produce a rapid and safe shutdown of the turbine-generator.

The input parameters and initial conditions for this event are listed in Subsection 15.0.2.

There are no cladding perforations as a re sult of this event an d no fission product release occurs.

15.7.1.3.4 Barrier Performance The event does not involve the barrier since the release occurs outside of containment.

15.7.1.3.5 Radiological Consequences

The amount of steam released to the turb ine building due to the above hypothesized events would be small due to the close clearances in the turbine shaft sealing glands. Any release would be well below th e release used for th e failure of the air ejector lines (Subsection 15.7.1.2).

LSCS-UFSAR 15.7-12 REV. 14, APRIL 2002 The radiological effects would be inconseq uential since the releases would be well below the releases used for the failure of the air ejector lines.

A 105% power uprate to 3489 MWt has no i mpact on either the gaseous radwaste system or the design basis radioactive gas source term. Therefore, the results presented in this analysis are not impacted by power uprate (Reference 12).

15.7.2 Radioactive Liquid Waste System Leak Radioactive Liquid Waste System Leaks are not analyzed for reload cores.

15.7.2.1 Miscellaneous Small Releases Outside Containment

Releases which could occur from piping fa ilures outside the containment include the feedwater system piping break (Subsectio n 15.6.6) and the main steamline break (Subsection 15.6.4) accidents. The analysis of these events provides doses which might occur for such a classificati on of piping fa ilure events.

Other releases which could occur outside containment include small spills and leaks of radioactive materials inside structures housing process equipment. Conservative values for leakage have been assumed and evaluated in Chapter 11.0 under routine plant releases. The offsite dose that results from any small spill which could occur outside containment would be negligible in comparison to the dose resulting from the previous postulated leakages of 15.6.

15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure

Liquid Radwaste Tank Failures are not analyzed for reload area.

15.7.3.1 Identification of Causes and Frequency Classification An unspecified event causes the complete release of the average radioactivity inventory in the tank containing the largest quantities of significant radionuclides in the liquid radwaste system. This is one of the concentrator waste tanks in the radwaste building. The airborne radioactivity released during the accident passes directly to the environment via the station vent stack.

The radioactive liquid releases seep through the radwaste building foundation and move through the soil toward the cooling lake.

Postulated events that could cause rele ase of the radioactive inventory of the concentrator waste tank are cracks in the vessels and operator error. The possibility of small cracks and consequent low-level release rates receives primary consideration in system and component design. The concentrates waste tank is designed to operate at atmospheric pressure and 200 F maximum temperature so LSCS-UFSAR 15.7-12a REV. 14, APRIL 2002 the possibility of failure is considered small. A liquid radwaste release caused by operator error is also considered a remote possibility. Operating techniques and administrative procedures emphasize de tailed system and equipment operating instruction. A positive action interloc k system is also provided to prevent LSCS-UFSAR 15.7-13 REV. 16, APRIL 2006 inadvertent opening of a drain valve. Should a release of liquid radioactive wastes occur, floor drain sump pumps in the floor of the radwaste building will receive a high water level alarm, activate automa tically, and remove the spilled liquid.

Much of the exposition concerning the remote likelihood of a leakage or malfunction accident of the concentrates waste tank applies equally to a complete release accident. The probability of a complete rupture or complete malfunction accident is, however, considered even lower.

Although not analyzed for the requirements of Seismic Category I equipment, the liquid radwaste tanks are constructed in accordance with sound engineering principles. Therefore, simultaneous failu re of all the tanks is not considered credible. This accident is expected to oc cur with the frequency of a limiting fault.

Note: Although the concentrator waste tanks have been abandoned -in- place (UFSAR Section 11.2.2.5.5), the use of the assumed tank contents for this analysis remains bounding for all other liquid radwaste tanks.

15.7.3.2 Sequence of Even ts and Systems Operation

The sequence of events expected to occur is as follows:

Sequence of Events Elapsed Time

1. Event begins -- failure occurs 0 2. Area radiation alarms alert plant personnel 1 minute 3. Operator actions begin 5 minutes The rupture of a concentrates waste tank would leave little recourse to the operator. No method of recontaining the gaseous phase discharge is available, however, isolation of the radwaste area would minimize the results. High radiation alarms both in the radwaste ventilation exhaust and in the radwaste area would alert the operator to the failure.

Normal isolation of the radwaste area ventilation is actuated upon initiation of the above alarms. However, no credit for any operator action or for ventilation isolation has been taken in evaluating this event.

15.7.3.3 Core and System Performance The failure of this liquid radwaste system component does not directly affect either the core or the nuclear steam supply system (NSSS). It will, of course, lead to decoupling of NSSS with the subject system.

LSCS-UFSAR 15.7-14 REV. 13 The analytical methods and associated assumptions which are used to evaluate the consequences of the accident are considered to provide a conservative assessment of the consequences.

The liquid radwaste tank failure is evaluated in accordance with the following assumptions and conditions:

a. One hundred percent of the shielding design-basis inventory of a concentrates waste tank is released into a concentrated waste tank cubicle.
b. The partition factor for radioiod ine from water to air is 0.001.
c. The airborne radioactivity released into a concentrates waste tank cubicle passes to the environment via the station vent stack. d. The (/Q) values are from Subsection 15.0.5.
e. The concentrated waste tank cubicle is designed to retain the entire contents of the tank.
f. The cubicle floor has no floor drains.
g. The cubicle has a steel liner which is capable of holding the entire contents of one tank.
h. The liquid waste is unable to seep out of the cubicle.
i. Eighty percent of the liquid inventory of one concentrated waste tank is released to the cubicle.
j. The hydraulic gradient is toward the lake.
k. The shielding design-basis inventory given in Table 11.2-5 is dissolved in a sufficient quan tity of liquid to produce 5000 gallons of concentrated waste.
l. No credit has been taken for dilution in groundwater or for dispersion.
m. Additional parameters are given in Subsection 2.4.12.

LSCS-UFSAR 15.7-15 REV. 13 Results The conservative assessment of the liquid radwaste tank failure leads to consideration of iodine partitioning from a spill which would be drained in a rapid manner, thereby minimizing the release of iodine to the air.

The radwaste building exhaust is filtered through HEPA filters for which no reduction in released iodine has been assumed. Table 15.7-8 presents the parametric values used in the conserva tive analysis and Table 15.7-10 lists the radionuclide activity release to the environment.

Because of the design features incorporated in LSCS, i.e., as the liquid radwaste tanks are each located in vented and drai ned cells, the failure of a liquid radwaste tank will produce virtually no additive ef fects above doses from normal operation.

The most conservative liquid effluence model would involve no groundwater dilution or dispersion. The model relies only on the soil permeability, the soil porosity, and the hydraulic gradient to determine the liquid concentrations that will reach the lake. These parameters are used to derive the travel time from the radwaste building to the lake.

If the concentrated waste volume remains constant, 600 years would be required for the radioisotope concentration to become less than the limits specified in 10 CFR 20, Appendix B, Table II, Column 2. Table 5.7-11 gives maximum liquid effluent concentrations versus time for the assumptions given above and in Subsection 2.4.12.

The liner in the concentrate tank cubicle will prevent any liquid waste from reaching the groundwater or the lake. Water sampling (see LSCS-ER (OLS), Chapter 6.0) has measured background beta radiation levels between 4 pCi/liter and 35 pCi/liter. Since no concentrated wa ste liquid can reach the lake, the activity in the lake water will fall between these values. All other radwaste tanks are located below the level of the lake surface. No liquid radwaste from these is released to the environment (see Subsection 2.4.12).

15.7.3.4 Barrier Performance This event does not involve the barrier since the release occurs outside of containment.

LSCS-UFSAR 15.7-16 REV. 14, APRIL 2002 15.7.3.5 Radiological Consequences Realistic Analysis The radiological effects are based on a short-term release to the atmosphere using the meteorological parameters presented in Subsection 15.0.5. The whole-body dose results from the gamma radiation emitted by the iodine. The doses from airborne releases are presented in Table 15.7-9.

The liquid releases due to the rupture of on e concentrated waste tank will not have any adverse effects on the cooling lake or the Illinois River. No radioactivity is added to these bodies of water.

The radwaste equipment's radioactive invent ory is a combination of the collection volume, the source stream(s) supply rates, and the activity in the source streams.

For liquid radwaste following the 5% power uprate, all these parameters remain unchanged (Reference 12). Therefore, the re sults of this analys is are not impacted by power uprate.

15.7.4 Fuel Handling Accident 15.7.4.1 Identification of Causes and Frequency Classification The fuel handling accident is assumed to o ccur as a consequence of a failure of the fuel assembly lifting mechanism resulting in the dropping of a raised fuel assembly onto other fuel bundles in the core. A variety of events which qualify for the class of accidents termed "fuel handling accidents" has been investigated. The accident

which produces the largest num ber of failed spent fuel rods (including consideration of the drop of a fuel bundle onto the Unit 2 consolidated fuel storage pool) is the drop of a spent fuel bundle onto the reactor core when the reactor vessel head is off. This accident is expected to occur wi th the frequency of a limiting fault.

15.7.4.2 Sequence of Even ts and Systems Operation

The most severe fuel handling accident from the radiological viewpoint is the dropping of a fuel assembly onto the top of the core. The sequence of events is as follows: Event Approximate Elapsed Time

1. Fuel assembly is being handled by refueling equipment. The assembly drops onto the top of the core.

0 LSCS-UFSAR 15.7-16a REV. 14, APRIL 2002

2. Some of the fuel rods in both the dropped assembly and reactor core are damaged, resulting in the release of gaseous fission products to the reactor coolant and eventually to the reactor building

atmosphere.

0 LSCS-UFSAR 15.7-17 REV. 14, APRIL 2002

3. The reactor building ventilation radiation monitoring system alarms to alert plant personnel, isolates the ventilation system, and starts operation of the SGTS.

<1 minute

4. Operator actions begin.

<5 minutes

Normally, operating plant instrumentation and controls are assumed to function, although credit is taken only for the isolation of the normal ventilation system and the operation of the standby gas treatment system. Operation of other plant or reactor protection systems or ESF systems is not expected.

The automatic ventilation isolation system, which includes: a) the radiation monitoring detectors, b) the isolation valves, and c) the SGTS is designed to single-failure criteria and safety requirements.

Refer to Sections 7.6 and 9.4.

15.7.4.3 Core and System Performance The analytical methods and associated assumptions used to evaluate the

consequences of this accident are considered to provide a realistic yet conservative assessment of the consequences.

The kinetic energy acquired by a falling fuel assembly may be dissipated in one or more impacts. To estimate the expected nu mber of failed fuel rods in each impact, an energy approach is used.

The fuel assembly is expected to impact on the reactor core at a small angle from the vertical, possibly inducing a bending mo de of failure on the fuel rods of the dropped assembly. It is assumed that each fuel rod resists the imposed bending load by a couple consisting of two equal, opposite concentrated forces.

Therefore, fuel rods are expected to absorb little energy prior to failure as a result of bending. Actual bending tests with concentrated point-loads show that each fuel rod absorbs approximately l ft-lb prior to cl adding failure. Each rod that fails as a result of gross compression distortion is expected to absorb approximately 250 ft-lb before cladding failure (based on 1% uniform plastic deformation of the rods). The energy of the dropped assembly is conservatively assumed to be absorbed by only the cladding and other core structures. Because a fuel assembly consists of 72%

fuel, 11% cladding, and 17% other structural material by weight, the assumption that no energy is absorbed by the fuel material results in considerable conservatism in the mass-energy calculations that follow.

LSCS-UFSAR 15.7-18 REV. 13 The energy absorption on successive impacts is estimated by considering a plastic impact. Conservation of momentum under a plastic impact shows that the fractional kinetic energy absorbed during impact is: where M1 is the impacting mass and M2 is the struck mass. The assumptions used in the analysis of this accident are listed below: a. The fuel assembly is dropped from the maximum height allowed by the fuel handling equipment, which is less than 30 feet. b. The entire amount of potential energy, referenced to the top of the reactor core, is available for application to the fuel assemblies involved in the accident. This assumption neglects the dissipation of some of the mechanical energy of the falling fuel assembly in the water above the core and requires the complete detachment of the assembly from the fuel hoisting equipment. This is only possible if the fuel assembly handle, the fuel grapple, or both grapple cables break. c. None of the energy associated with the dropped fuel assembly is absorbed by the fuel material (uranium dioxide). Energy Available Dropping a fuel assembly onto the reactor core from the maximum height allowed by the refueling equipment, less than 30 feet, results in a maximum impact velocity of 40 ft/sec. The kinetic energy acquired by the falling fuel assembly is less than 17,300 ft-lb and is dissipated in one or more impacts. Energy Loss Per Impact Based on the fuel geometry in the reactor core, four fuel assemblies are struck by the impacting assembly. The fractional energy loss on the first impact is approximately 80%. The second impact is expected to be less direct. The broad side of the dropped assembly impacts approximately 24 more fuel assemblies, so that after the second impact only 136 ft-lb (approximately 1% of the original kinetic energy) is available for a third impact. Because a single fuel rod is capable of absorbing 250 ft-lb in 17.15 121 LSCS-UFSAR 15.7-19 REV. 13 compression before cladding failure, it is unlikely that any fuel rod will fail on a third impact. If the dropped fuel assembly strikes only one or two fuel assemblies on the first impact, the energy absorption by the core support structure results in approximately the same energy dissipation on the first impact as in the case where four fuel assemblies are struck. The energy relations on the second and third impacts remain approximately the same as in the original case. Thus, the calculated energy dissipation is as follows: first impact 80% second impact 19% third impact 1% (no cladding failures) Fuel Rod Failures For the initial core, the first impact dissipates 0.80 x 17,300 or 13,800 ft-lb of energy. It is assumed that 50% of this energy is absorbed by the dropped fuel assembly and that the remaining 50% is absorbed by the struck fuel assemblies in the core. Because the fuel rods of the dropped fuel assembly are susceptible to the bending mode of failure and because 1 ft-lb of energy is sufficient to cause cladding failure as a result of bending, all 62 rods of the dropped fuel assembly are assumed to fail. Because the 8 tie-rods of each struck fuel assembly are more susceptible to bending failure than the other 54 fuel rods, it is assumed that they fail on the first impact. Thus 4 x 8 = 32 tie-rods (total in 4 assemblies) are assumed to fail. Because the remaining fuel rods of the struck assemblies are held rigidly in place in the core, they are susceptible only to the compression mode of failure. To cause cladding failure of one fuel rod as a result of compression, 250 ft-lb of energy is required. To cause failure of all the remaining rods of the 4 struck assemblies, 250 x 54 x 4 or 54,000 ft-lb of energy would have to be absorbed in cladding alone. Thus, it is clear that not all the remaining fuel rods of the struck assemblies can fail on the first impact. The number of fuel rod failures caused by compression is computed as follows: Thus, during the first impact, fuel rod failures are as follows: dropped assembly 62 rods (bending)2)-(15.7 11250171111800,135.0 LSCS-UFSAR 15.7-20 REV. 20, APRIL 2014 struck assemblies 32 tie-rods (bending) struck assemblies 11 rods (compression) 105 failed rods Because of the less severe nature of the second impact and the distorted shape of the dropped fuel assembly, it is assumed that in only 2 of the 24 struck assemblies are the tie-rods subjected to bending failure. Thus 2 x 8 = 16 tie-rods are assumed to fail. The number of fuel rod failures caused by compression on the second impact is computed as follows: Thus, during the second impact the fuel rod failures are as follows: struck assemblies 16 tie-rods (bending) struck assemblies 3 rods (compression) 19 failed rods The total number of failed rods resulting from the accident is as follows: first impact 105 rods second impact 19 rods third impact 0 rods 124 total failed rods The above evaluation, the FSAR Base Case, was based on the GE 8x8 fuel design and considered a drop distance of 30 feet but did not consider the impact of the grapple mast and head. The current grapple mast and head weigh 619 pounds. Alternative fuel designs have been loaded in the core since the GE 8x8 design:GE9 (8x8): ATRIUM-9B (9x9): ATRIUM-10 (10x10); GE14 (10x10); and GNF2 (10x10). The number of fuel pin failures resulting from the Refuel Accident for these fuel designs are summarized below: 3)-(15.7 3250171111300,17219.0 LSCS-UFSAR 15.7-21 REV. 20, APRIL 2014 GE 8x8 (FSAR Base Case) ATRIUM-9B ATRIUM-10 GE14 GNF2 Fuel Assembly Drop Distance 30 ft. 34 ft. 34 ft. 34 ft. 34 ft. Grapple Mast and Head Weight Not considered 620 lbs 620 lbs 619 lbs 619 lbs Pins Failed (First impact) 105 116 140 - - Pins Failed (Second impact) 19 15 16 - - Total Pins Failed 124 131 156 172 172 Reference 6 15 2 23 Reference 22 confirms that Reference 15 remains applicable for ATRIYM-10 fuel at MUR conditions and that Reference 21 remains applicable fo r ATRIUM 10XM LTA at MUR conditions.

15.7.4.4 Barrier Performance The reactor coolant pressure boundary and the containment are assumed to be open. All radioactivity is released thro ugh the SGTS to the environment. The transport of fission products from the reactor building is discussed below.

15.7.4.5 Radiological Consequences

LSCS-UFSAR 15.7-22 REV. 19, APRIL 2012 AST Analysis The AST analysis (Reference 3) is based on engineering inputs from RG 1.183, which yield a conservative assessment of this accident. Specific input parameters used in this evaluation are shown in Table 15.7-21. The leakage path is the same as that used in the previous non-AST design-basis evaluation, Figure 15.7-1.

RG 1.183 is the basis for the AST evaluations. Concerning the FHA, this AST guidance has smaller gap fractions and a larger pool decontamination factor (DF).

Fission product release estimates for the evaluation of this fuel handling accident use the following assumptions:

  • The reactor fuel has an average irra diation time of 1000 days at 3559 MWt up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the accident. This assumption results in an equilibrium fission product concentration at the time the reactor is shut down. Longer operating histories do not increase the concentration of fission products of concern. The 24-hour decay time allows time to shut down the reactor, depressurize it, remove the reactor vessel head, and remove the reactor internals above the core. It is not expected that these operations could be accomplished in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and they would normally require approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
  • This analysis is applicable to fuel whose burnup and power limits are bounded by those specified in RG 1.183, footnote 11. This allows application of the gap activity fractions for Non-Lo ss of Coolant Accident (LOCA) events per Table 3 of RG 1.183, which are as follows:

5% of the noble gases (excluding Kr-85) 10% of the Kr-85 5% of the iodine inventory (excluding I-131) 8% of the I-131 12% of the Alkali metal inventory

  • Because of the negligible particulate activity available for release in the fuel rod plena, none of these solid fission products is assumed to be released.
  • It is assumed that GE14 fuel is used an d that 172 fuel rods fail as can be seen in the following table.

LSCS-UFSAR 15.7-23 REV. 20, APRIL 2014 Bundle Type Fuel Type Assumed Pins-in Bundle Failed Pins Damaged Core Fraction Radial- Peaking Factor (PF) Damaged Core Fraction with PF GE-Various 8x8 62 124 0.002618 1.5 0.003927 FANP Atrium-9B 9x9 72 131 0.002381 1.5 0.003572 Atrium-10 10x10 91 156 0.002244 1.7 0.003815 GE11&GE13 9x9 74 140 0.002476 1.5 0.003714 GE12&GE14* 10x10 87.33 172 0.002578 1.7 0.004382 GNF2 10x10 85.6 172 0.002630 1.6 0.004208

  • Bounding Assembly type, with Radial Peaking Factor commensurate with full core application
  • The SGTS filter efficiency for HEPA and Charcoal is credited as 99% (with a 0.5% bypass leakage).
  • Control Room Emergency Makeup Filter efficiency credited for HEPA is 99% and for Charcoal 90%.
  • Recirculation Filtration unit has an effective credit of 70% (not including bypass leakage). The following assumptions and conditions are used in calculating the release of activity to the environment:
  • All of the noble gases released to the fuel pool become airborne in the secondary containment (reactor building).
  • The decontamination factors for iodine elemental and organic species are 500 and 1, respectively, when the depth of water above the fuel is 23 feet or greater, therefore giving an effective decontamination factor of 200.
  • The ventilation rate from the secondary containment to the environment through the SGTS is 0.1 volume change per hour. This assumption results in 77.678% of the available activity to be released during the drawdown period, 98.889% by the end of the fumigation period, and 99.9994% (effectively all) to be released with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The SGTS filter efficiency of 99.0% with a 0.5% bypass leakage allowance is credited only after a 15 minute drawdown period. The results for this analysis are shown in Table 15.7-22.

LSCS-UFSAR 15.7-23a REV. 19, APRIL 2012 15.7.5 Spent Fuel Cask Drop Accident The Spent Fuel Cask Drop accident is not considered a credible "design basis accident," because the reactor building overhead crane meets the single-failure proof criteria of ASME NOG-1-2004, NUREG-0554 and NUREG-0612, Appendix C. The main hoist is classified as a Type I main hoist per ASME NOG single-failure proof for all identified critical loads. The overhead crane is designed to ensure that a single credible failure of the crane system will not re sult in the loss of the capability of the system to safely retain a load. A spent fuel cask will follow the restricted critical "L" path shown in Figure 9.1-6.

[Historical Information]

The following discussion reflects the results of the original 100-Ton cask drop evaluation. This evaluation was completed prior to the reactor building overhead crane being upgraded to single-failure-proof.

The Spent Fuel Cask Drop Accident is not analyzed for reload cores.

For analysis of spent fuel cask drops, the 100-ton spent fuel cask as described in Subsection 9.1.4.2 is used.

LSCS-UFSAR 15.7-24 REV. 15, APRIL 2004 The cask drop analysis (32 irradiated fuel bundles) makes the assumption that the fuel is normally cooled for a minimum of 360 days prior to cask loading and has assumed maximum values for irradiation time for a given fuel type. For shipments involving a small quantity of irradiated fuel rods or one or two fuel assemblies (generally for Research & Development or root cause investigation into performance issues), a shorter cooling period and/or longer period of irradiation than assumed for the large cask drop analysis are permissible provided the following are verified: a. The fuel rods meet the shipping cask requirements for minimum cooling time, for maximum decay heat limitations, and maximum fuel rod irradiation time in addition to other cask specific requirements. b. The source term inventory in the contained fuel rods (those isotopes which are pertinent to the offsite dose consequences of a cask drop as specified in Section 15.7.5) are bounded by the cask drop accident source term and, hence, the offsite dose consequences of the large scale cask drop analysis remains bounding. Credible cask drop accidents have been broken down into two categories: those in which the cask is assumed to fall less than 30 feet, where there are no radiological consequences because the cask is able to withstand a free drop through a distance of 30 feet without rupturing; and those cask drops greater than 30 feet, where conservative assumptions are made to model postaccident conditions because there is no accurate way of predicting the cask conditions after such a drop. 15.7.5.1 Identification of Causes and Frequency Classification Postulation of a cask drop accident resulting from a failure of the reactor building crane would require the failure of a component in the main hoist drive train, the upper or lower load block, the cable, or the crane hook. Due to the safety margins involved in the design of these components, the initial load tests on the crane hook and the crane and the inservice inspection programs, it is considered extremely unlikely that such an accident could happen. Therefore, this accident is categorized as an event with the frequency of a limiting fault. The only other cask failure mechanism possible is an in-transit event. Although accidents with railroad and truck type vehicles are not uncommon, transportation restrictions and transport vehicle designs also justify the expected frequency of a limiting fault for this event. Recent input tests at Sandia National Laboratories indicate that integrity of shipping casks designed to 10 CFR 71 Appendix B criteria is not compromised during in-transit collisions or upsets. For spent fuel cask drops of greater than 30 feet, only two drop cases are of any consequence. One results from a reactor building crane failure when the cask is over the cask storage well; this sets up a maximum drop of approximately 43 feet.

LSCS-UFSAR 15.7-24a REV. 15, APRIL 2004 The other results when the cask is over the equipment hatch, where the maximum drop is approximately 133 feet. An analysis of the structural ef fects on the building from such a drop are included in Refere nce 5. The LaSalle Station design was specifically modified to accept the more severe fuel cask drop accident in the reactor building.

Spent fuel cask drops of less than 30 feet can be postulated for the refueling period when the cask is supported by the reactor building crane or for a cask transport vehicle accident. During refueling, perm issive operations are constrained by the critical L-path control on the crane system as described in Subsection 9.1.4.1.2. The maximum cask drop height under this path control is about 10 inches. No drop will result in the spent fuel cask tipping over or structur ally damaging the building while under critical path control. Although contingent upon both cask and vehicle design, cask drops from a transport vehicl e are less than 30 feet and thus yield trivial results even though this event is categorized as a limiting fault.

This event is categorized as a limiting fault.

LSCS-UFSAR 15.7-25 REV. 13 15.7.5.2 Sequence of Even ts and Systems Operation The sequence of events and approximate elapsed time before the event occurs for each accident previously described are as follows:

Event Description - The spent fuel cask is dr opped 43 feet into the cask storage well. The cask head is assumed to be latched to the cask.

Sequence of Events Approximate Elapsed Time

1. Event begins - cask drops to

bottom of cask well 0

2. Area is evacuated until radiological consequences are assessed < 5 minutes
3. Spent fuel pool exhaust radiation monitor alarm and initiation of standby gas treatment system

< 5 minutes

4. Assessment of radiological consequences, structural

damage to building, and cask cooling requirements 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

5. Cleanup and repair operations begin Dictated by damage assessment Event Description - The spent fuel cask is dropped 133 feet down through the equipment hatches. A detailed analysis of this drop is presented in Reference 5.

Sequence of Events Approximate Elapsed Time

1. Event begins - cask drops to elevation 673 of the reactor building 0 LSCS-UFSAR 15.7-26 REV. 13
2. Area is evacuated until radiological consequences are assessed < 5 minutes
3. Reactor building exhaust radiation monitor alarm and initiation of standby gas treatment system.

< 5 minutes

4. Assessment of radiological consequences, structural damage to building and cask, and cask cooling requirements 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
5. Cleanup and repair operations begin dictated by damage assessment Event Description - The spent fuel cask is dropped either 10 inches into the decontamination pit or 6 inch es onto the refueling floor.

Sequence of Events Approximate Elapsed Time

1. Event begins - cask drops onto refueling floor 0 2. Cask damage and crane

malfunction are assessed 5 minutes

3. Auxiliary cooling is provided to the cask if required

< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

4. Cask is unloaded and inspected 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Event Description - The spent fuel cask is dropped from its transport vehicle.

Sequence of Events Approximate Elapsed Time

1. Event begins - cask falls from transport 0

LSCS-UFSAR 15.7-27 REV. 13 2. Transport vehicle operator stops vehicle 1 minute 3. Cask is loaded back on transport vehicle for continuation of journey 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 15.7.5.3 Core and Systems Performance This event does not have a direct effect on the core and does not affect the continued normal operation of the reactor system. Damage to facility safety-related equipment, which is possible only in the case of the 133-foot drop described previously, has been outlined in Reference 5. The reactor building ventilation and the standby gas treatment systems are expected to function normally. This allows automatic isolation of the reactor building and gives assurance that any release of radioactive gases associated with a cask drop will be filtered and treated through the standby gas treatment system. 15.7.5.4 Barrier Performance This event occurs outside the primary containment, therefore this section is not directly applicable. The only fission products expected to be released from the cask would be those contained in the external shield coolant annulus in the event at loss-of-cask coolant. For the purpose of analysis, the assumption is made that 1000 Ci of noble gas is released from the cask. The cask handling area is served by the SGTS. All releases within this structure are capable of being filter-decayed and released through the elevated station vent stack. The integrity of the secondary containment boundary (reactor building) is maintained for the spent fuel cask drop accident. Therefore only an elevated release of radioactive gases is postulated. 15.7.5.5 Radiological Consequences A fully loaded spent fuel cask is arbitrarily assumed to drop 133 feet while being lowered through the facility hatches to a waiting flatcar inside the reactor building. The cask impacts a yielding surface thus resulting in compromised cask integrity with a postulated release of a fraction of the radioactivity content of all rods to the secondary containment atmosphere. This gap activity is assumed to be immediately LSCS-UFSAR 15.7-28 REV. 16, APRIL 2006 exhausted out the station vent stack. Credit taken for charcoal adsorption of radioactivity released through the SGTS is assumed to be 95% for radioiodine.

The input parameters and initial conditions are as follows:

a. The cask is loaded to a maximum of 32 fuel assemblies; this is equivalent to 4.2% of the total core inventory. An operating period of 1300 days was assumed for design-burnup condition (38.3 MWD/MTU core average exposure) for the GE analysis, 1953 days was assumed for the FANP ATRIUM-9 analysis, and 2078 days was assumed for the FANP ATRIUM-10 analysis.
b. The cask drop event occurs at 360 days after shutdown, which is the earliest normal time interval before spent fuel elements are loaded into the cask.
c. It is conservatively assumed that all of the gap activity from the 32 fuel assemblies is released to the atmosphere of the secondary containment.
d. The released gap activity consists of 10% of the noble gases other than Kr-85, 30% of the Kr

-85, and 10% (12% was used for ATRIUM-10) of the total radioiodine in the damaged fuel rods at the time of the event.

Table 15.7-19 shows the fission product invent ory in the core, in the cask for 32 fuel assemblies, and in the fuel rod gaps which is assumed to escape the fuel casks.

Peak core activity was assumed with a radi al peaking factor of 1.5 applied for the GE and FANP ATRIUM-9 fuel assemblies in the cask when it was dropped. A radial peaking factor of 1.7 is applied to ATRIUM-10 fuel assemblies in the cask when it was dropped.

The doses shown in Table 15.7-20 for the entire cloud passage are well below the limits applicable to an individual receptor at the exclusion area boundary following this postulated event. Doses at th e LPZ boundary are also well below the appropriate limits. The negligible thyroid dose estimates result from the extremely small release of radioiodines from the fuel that has decayed for approximately 1 year. The fission product inventory increases in proportion to the power level. Therefore, a 105% core licensed power uprate to 3489 MWt increases the dose consequences from a refueling cask drop accident by 5% (Reference 12). As seen in Table 15.7-17, this is still a small fraction of the 10 CFR 100 requirements.

LSCS-UFSAR 15.7-28a REV. 19, APRIL 2012 As outlined in Reference 9, the reload analyses performed for L1C09 and L2C08 bound the extended burnup to support 24-month fuel cycles. These analyses included, in part, the fuel handling accident, cask drop accident, and the control rod drop accident. The dose consequences for the fuel handling and control rod drop accidents remain bounded by the corresponding Updated Final Safety Analysis Report (UFSAR) Chapter 15 analyses previously performed for General Electric Company (GE) supplied fuel. The dose co nsequences for the cask drop accident increased by an insignificant amount.

Below is the listing of the parameters, from References 9, 13 and 14, that are the basis for the high exposure ATRIUM-9B Source Term used in the L1C09 and L2C08 analyses for ATRIUM-9B fuel. The units given below are Megawatt-days (MWd) per Metric Ton of Uranium (MTU) and Megawatts Thermal (MWt).

Spent Fuel Cask Drop Accident:

50,000 MWd/MTU; 3323 MWt core power; 1953 core residence days Below is the listing of the parameters, from Reference 19, that is the basis for the high exposure ATRIUM-10 Source Term used in the analyses for ATRIUM-10 fuel.

Reference 22 confirms that the Reference 19 evaluation remains applicable for ATRIUM-10 fuel at MUR conditions and that the Reference 21 evaluation remains applicable for ATRIUM 10XM LTA at MUR conditions.

Spent Fuel Cask Drop Accident: 60, 000 MWd/Mtu; 3910 MWt Core power; 2078 Core residence days.

[End of Historical Information]

LSCS-UFSAR 15.7-29 REV. 19, APRIL 2012 15.7.6 References 1. D. Nguyen, "Realistic Accident Analysis - the RELAC Code and User's Guide," NEDO-21142, October 1977. 2. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A, (latest approved revision). 3. LSCS Design Analysis L-003067, Rev. 2, "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms". 4. Deleted. 5. "Supplemental Crane Design Features and Fuel Cask Drop Analysis," Special Report 2, LaSalle County Station PSAR, April 1974. 6. FANP document, "LaSalle Fuel Handling Accident for ATRIUM-9B Fuel," EMF-96-171(P), Revision 2, Framatome - ANP, March 2001. 7. SPC letter, JHR : 96 : 399, "Re-Transmittal of Response to comment 25 LaSalle Cask Drop Accident with Table," J.H. Riddle to R.J. Chin, October 10, 1996. 8. Letter from R. M. Krich, Commonwealth Edison (ComEd) Company, to U.S. NRC, "Request for License Amendment for Power Uprate Operation," dated July 14, 1999, with Attachment E: General Electric Nuclear Energy, Licensing Topical Report NEDC-32701P, Revision 2, "Power Uprate Safety Analysis Report for LaSalle County Station, Units 1 and 2," dated July 1999 (Proprietary). 9. Letter from C. G. Pardee, Commonwealth Edison (ComEd) Company, to U.S. NRC, "Supplement to the License Amendment Request for Power Uprate Operation," dated 04/14/2000. 10. Letter from D. M. Skay, U.S. NRC, to O. D. Kingsley, Commonwealth Edison (ComEd) Company, "LaSalle - Issuance Of Amendments Regarding Power Uprate (TAC NOS. MA6070 AND MA6071)," dated May 9, 2000 (OL Amendments 140/125). 11. "Safety Evaluation By The Office Of Nuclear Reactor Regulation Related To Amendment No. 140 To Facility Operating License No. NPF-11 and Amendment No. 125 To Facility Operating License No. NPF-18; Commonwealth Edison Company Lasalle County Station, Units 1 And 2; Docket Nos. 50-373 And 50-374" dated May 9, 2000.

LSCS-UFSAR 15.7-30 REV. 20, APRIL 2014 12. ENDIT H041, S&L Task 21b, "Radiological - Chapter 15 Accidents." 13. Memorandum NFM:BSA 00-028 from R. W. Tsai, Nuclear Fuel Management, to F. A. Spangenberg III, Regulatory Assurance, "Summary of Key Input Parameters for the High Exposure Siemens ATRIUM-9B Source Term," dated April 13, 2000 DG00-000413. 14. Letter JH:96:188 from J. H. Riddle (Siemens) to Dr. R. J. Chin (ComEd), "Radioactive Release Analysis Source Term Values," dated May 20, 1996 DG96-001013. 15. F-ANP Document EMF-2679(P), Revision 0, "LaSalle Fuel Handling Accident for ATRIUM-10 Fuel," dated January 2002. 16. FANP document, "Dropped Cask Accident Analysis Results for LaSalle with ATRIUM-10 Fuel", AWW:02:021, dated December 9, 2002. 17. GNF Report, "GE14 Fuel Design Cycle" Independent Analysis for LaSalle Unit 1 and Unit 2", GE-NE-0000-0026-4769-00, Rev.0 dated January 2005. 18. Letter, D.E. Garber (FANP) to F.W. Trikur (Exelon), "Radioactive Source Term Values for ATRIUM-10 Fuel at LaSalle, " DEG:01:170 dated October 22, 2001. 19. Letter, A.W. Will (FANP) to F.W. Trikur (Exelon), "Dropped Cask Accident Analysis Results for LaSalle with ATRIUM-10 Fuel," AWW:02:021 dated December 9, 2002. 20. GNF Report, "GE14 Fuel Design Cycle - Independent Analyses For LaSalle Unit 1 and Unit 2", GE-NE-0000-0026-4769-00, Rev. 0, dated January 2005. 21. AREVA document 51-9093566-000, "Radioactive Source Term Values, Fuel Handling Accident and Dropped Cask Accident Analyses at LaSalle with ATRIUM 10XM Fuel," transmitted to Exelon under letter FAB08-2573, October 30, 2008. 22. Design Analysis L-003509, Revision 0, "Evaluation of Appendix R, Station Blackout, Containment and Source Terms for LaSalle MUR Power Uprate," July 2010. 23. Design Analysis L-003696, Revision 1, "NEDC-33647P, GNF2 Fuel Design Cycle-Independent Analyses for Exelon LaSalle County Station Units 1 and 2" February 2012.

LSCS-UFSAR 15.8-1 REV. 13 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS)

The ATWS event is not analyzed for reload cores.

Prior to the issuance of the ATWS Rule (Reference 7) for the mitigation of the consequences of postulated anticipated transients without scram (ATWS), a potential concern about the possibility of common mode failure of the scram system was noted in ACRS reports for various reac tors. General Electric examined the BWR reactor protection system and concluded that no credible common mode failure could prevent a scram signal from reaching the scram actuators when scram is called for. This study is documented in Reference 1. The same type of equipment is used in the actuation and drive mechanism for each control rod. However, each of the 185 control rods has its own separate scram actuator and rod drive equipment.

Therefore, the probability that the control rod drives or scram actuators for every rod would all fail at exactly the same time is extremely low. This coupled with the technical specification requirements for peri odic testing provides a very high degree of assurance that any failure mode which might affect the scram function for all control rods would be detected long before the failure had affe cted enough of the rods to prevent reactor shutdown in the event scram was called for. Thus, the occurrence of a common mode failure which would cause complete failure to scram, or even failure to insert enough rods to shut down the reactor to the hot standby condition, is an extremely unlikely event.

General Electric believed that the complete failure of the BWR scram system due to common mode failure is of such extremely low probability that no change in BWR design to account for the event is warrante

d. However, at the request of the AEC in 1971, a study was performed to evaluate the consequences of an undefined failure of the scram protection system. GE reported on those features which could mitigate the effects of a reactor shutdown from ATWS without a control rod drive scram. Based on its evaluation of the nature of ATWS, General Electric established appropriate criteria and provided an analysis of the event, including a suggested design change which enabled the then current BWR product line to meet the criteria. This analysis was reported in Reference 2.

As was indicated in Reference 2, tripping the recirculation pumps at the start of an ATWS event prevents the violation of any of the criteria when the reactor is shut down shortly after the beginning of the ATWS event. This shutdown can be achieved by manual insertion of control ro ds. The following criteria established in NEDO 10349 (Reference 2) are satisfied: ra diological releases are less than 10 CFR 100 limitations; there is no failure of the reactor coolant pressure boundary; the core is in a long-term coolable geometry; and containment integrity is maintained.

LSCS-UFSAR 15.8-2 REV. 14, APRIL 2002 References 3, 4, and 5 addressed the regulatory staff position published via WASH-1270. This NEDO-20626 document di scusses design mo difications which would mitigate ATWS consequences in Class 1.B plants.

In September 1976, GE submitted the reliab ility evaluation for RPT and an alternate rod insertion system (ARI), which identified the simple modifications to BWR scram systems that would increase their reliability by several orders of magnitude. Edison committed to install the RPT plus ARI via lette r of September 29, 1976, Edison to the NRC (Lee to Hanauer).

NUREG-0460 was issued by NRC in late 1978 with recommendations for modifications to specific categories of BWR plants. GE's generic response was solicited by a February 15, 1979, request fr om Dr. Mattson to Dr. Sherwood. That proprietary response, as provided in May 1979, covered Alternate 3 specifically (RPT

+ ARI + Autoboron) as requested, but also concluded that Alternate 2 (RPT + ARI) adequately met the NRC's criteria and did it in a cost-effectiv e way, whereas the NRC's Alternate 3 recommended solution was not justified by value assessment. "Implementation of Alternate 2 of NUREG-0460 as defined by the Staff is all that would be necessary to make the ATWS ri sk acceptable based on the staff's own statement that the present likelihood of severe consequences arising from an ATWS event is acceptably small and presently there is no undue risk to the public from ATWS.". The value of implementing any modifications beyond Alternate 2 is diminishingly small and non-economic.

A plant-unique evaluation was performed for the LaSalle plant to demonstrate the adequacy of the RPT and ARI design for preventing and mitigating ATWS events (reference 6). The limiting transient events analyzed thus confirm the prior conclusion that RPT and ARI are adequate design modifications for the ATWS hypothesis. Nevertheless, in response to 10CFR50.62 (C)(4), LSCS has incorporated design changes in each Unit, to mitigate the consequences of certain ATWS events. The rule requires that each BWR have a Standby Liquid Control System which a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate soluti on. To meet this requirement the SBLC system will use a boron 10 enriched sodium pentaborate. The concentration and temperature requirements will remain exactly as before. The LaSalle County Station plant design currently contains protection against failure to scram via recirculation pump trip on high reactor dome pressure or on low-low reactor water level using one-out-of-two taken twice logic.

The Main Steam Isolation Valve Closure (MSIVC) event has been reanalyzed under ATWS condition to support removal of five (5) SRVs (total of 13 SRVs remaining). This was performed by General Electric using the ODYN computer code which has been approved for use by the NRC. The analysis credits proper op eration of 12 SRVs.

This allows for one SRV to be out-of-service. The results show that the vessel pressure reaches a maximum pressure of 1 378 psig which is within the ASME vessel overpressure criterion of 1500 psig fo r ATWS events (References 9 and 10).

LSCS-UFSAR 15.8-2a REV. 20, APRIL 2014 For the transition to the FANP ATRIUM-9B fuel, GE evaluated the impact on the ATWS analysis for LaSalle with respect to the alternate core characteristics (such as void coefficient) associated with the FANP ATRIUM-9B fuel. In Reference 8 GE documented that the changed core characteristics do not result in a significant impact on the ATWS analysis and thus the analyses of record are still applicable.

For the 1999 Power Uprate Project, GE eval uated the impact of the uprated power level on the limiting ATWS events. The analysis results, documented in Reference 12, demonstrate that all ATWS acceptance criteria are met for the most severe ATWS events when operating at uprated conditions.

For the transition to FANP ATRIUM-10 fuel, Exelon evaluated the impact of ATWS analysis for LaSalle with respect to alternative core characteristics associated with FANP ATRIUM-10 fuel. Reference 12 provid ed the FANP documentation that the changed core characteristics do not result in a significant impact on the ATWS analysis and thus the analysis of record are still applicable.

An ATWS analysis was performed by GE to support the 2004 transition to the GE14 fuel design. The results of the plant specific ATWS analysis (Reference 13) for LaSalle based on a mixed core of GE14, FANP ATRIUM-9B and FANP ATRIUM-10

fuels indicated that the deployment of GE14 does not adversely affect the plant's ability to meet the ATWS acceptance criteria. Therefore, the analysis of record are still applicable. Additionally, an increase in the Level 2 trip delay of up to four (4) seconds has no effect on the limiting ATWS events and key peak ATWS parameters.

An additional steam flow induced process measurement error in the Level 3 scram was evaluated by GE in Reference 14 for AT WS event and it was concluded that it is not affected by change in L3 analytical limit (AL) as there is no L3 function directly credited by the ATWS events. Ho wever since there is no scram there is bypass steam flow in the annular region ou tside the dryer, which causes an induced error in the L2 trip.

An ATWS analysis was performed by GE H to support the 2012 transition to the GNF2 fuel design. The results of the plant-specific ATWS analysis (Reference 16) for LaSalle based on an all GNF2 core indicate that the deployment of GNF2 does not adversely affect the plant's ability to meet the ATWS acceptance criteria.

Additionally, this conclusion is applicable for future LaSalle cycles with a mixed core of GNF2 (GNF) and ATRIUM-10 (AREVA). Therefore, the analyses of record are still applicable.

For the limiting ATWS events, the scenario involves pressurization due to the MSIV closure. The reactor isolation leads to a recirculation pump trip (ATWS RPT) very early in the transient and the trip is usually reached at about the same time the MSIVs are full closed. The ATWS RPT rapidly reduces power and steaming rate and is the key feature that reduces the steaming rate to be within the capacity of the Safety / Relief Valves. The post RPT po wer level is on the order of 50 to 55%

LSCS-UFSAR 15.8-2b REV. 20, APRIL 2014 power and by the time the level is near the Level 2 AL, the power and steaming rate is below 50%. With the reactor steaming rate reduced to 50%, the error will be significantly reduced, and its effect with be approximately 1/4 of the effect at rated conditions. This would reduce the error of 6 inches, for example, at rated power to about 1.5 inches at these conditions. Since the LaSalle RCIC/HPCS initiation at this water level for ATWS is not critical to the event mitigation, this error and delay to L2 is considered insignificant. A small delay for the RCIC / HPCS initiation would be slightly beneficial as the water level would be lower during a portion of the transient and would result in a reduced reactor power and reduced steaming rate to the suppression pool. The long-term mitigation of these events involves controlling water level to low levels in the vessel. Again the small error at these conditions (< 2 inches) is insignificant for water level control and power generation compared to the analysis. Based on above discussion, the analysis of record in Reference 13 for a mixed core of GE and AREVA fuel types re mains applicable with respect to steam flow induced process measurement error.

Non-limiting ATWS events that may initiate the Level 2 ATWS-RPT or other L2 functions for ATWS would also be affected by L3 analytical limit error. An example would be the LOFW event. This event would result in recirculation runback associated with the loss of flow and low level (e.g., level 4). This would reduce the power and steaming rate. The power would also reduce due to the reduced subcooling associated with the loss of feedwater flow. The combined effect would reduce the error to approximately half of the condition at rated power (based on an estimated power and steaming rate reduced to 70% prior to level 2). As events that trip ATWS-RPT on low level are power and pressure reduction events, they do not challenge the ATWS acceptance criteria and therefore a low level ATWS RPT delay due to L3 scram error is not significant for compliance to the ATWS acceptance

criteria. Therefore, the expected steam flow induced error (approximately half of the error at rated conditions) will have no significant affect on the power and pressure events, and these events will remain far from limiting.

For MUR power uprate, GE evaluated the impact of the MUR power level on the limiting ATWS events. The analysis results, documented in Reference 15, demonstrate that all ATWS acceptance criteria are met for the most severe ATWS events when operating at MUR power level.

LSCS-UFSAR 15.8-3 REV. 16, APRIL 2006 15.8.1 References

1. "An Analysis of Functional Common Mode Failures in GE BWR Protection and Control Instrumentation," NEDO-10l89, July 1979.
2. L. A. Michelotti, "Analysis of Anticipated Transients Without Scram," NEDO-10349, March 1971.
3. L. B. Claassen and E.C. Eckert, "Stu dies of BWR Designs for Mitigation of Anticipated Transients Without Scrams," NEDO-20626, October 1974.
4. L. B. Claassen and E. C. Eckert, "Studies of BWR Designs for Mitigation of Anticipated Transients Without Scrams, Amendment 1," NEDO-20626-1, June 1975.
5. L. W. Baysinger, E. C. Eckert, and D. G. Weis, "Studies of BWR Designs for Mitigation of Anticipated Transients Without Scrams, Amendment 2," NEDO-20626-2, July 1975.
6. "Evaluation of Anticipated Transients without Scram for LaSalle County Station Unit 2" (1983), QUAD-1-83-007.
7. ATWS Rule - 10CF50.62 Reduction of Risk from Anticipated Transient Without Scram (ATWS) events for Light-Water-Cooled Nuclear Power Plants.
8. GE Letter A096-004, "LaSalle Station, Al ternate Core Characteristics Impact on ATWS Analysis Results", J. Casillas (GE) to B. Karas (ComEd), dated March 1, 1996.
9. "Safety Review for LaSalle County Station Units 1 and 2 Safety Relief Valves Reduction and Setpoint Tolerance Relaxation Analyses", Rev. 2, GE Report, GE-NE-B13-01760, by H. X. Hoang, da ted February 1996. (On Site Review No.96-020)
10. Safety Evaluation Report (SER) by NRC, dated 06-03-99, for Amendment Nos. 133 and 118 for LaSalle County St ation Units 1 & 2, respectively.
11. Power Uprate Project Task 902, "Antic ipated Transient Without Scram," GE-NE-A1300384-25-01, Revision 0, October 1999.
12. FANP letter DEG:01:174, "ATWS Kinetic Pa rameters for LaSalle," D. Gerber to F. Trikur, dated October 25, 2001.
13. GE-Document NE-0000-0026-4769-00, Revision 0, "GE14 Fuel Design Cycle -

Independent Analyses for LaSalle Unit 1 and Unit 2", " January 2005.

LSCS-UFSAR 15.8-4 REV. 20, APRIL 2014

14. "BWR Owners Group Evaluation of Steam Flow Induced Error (SFIE) Impact on the L3 Setpoint Analytic Limit" GEH-NE-0000-0077-4603-R1.
15. Design Analysis L-003565, Revision 0, "T900 Series - ATWOS," July 2010.
16. Design Analysis L-003696, Revision 1, "NEDC-33647P, GNF2 Fuel Design Cycle- Independent Analyses for Exel on LaSalle County Station Units 1 and 2" February 2012.

LSCS-UFSAR 15.9-1 REV. 14, APRIL 2002 15.9 Loss of All Alternating Current Power (Station Blackout)

The Loss of all Alternating Current Power (Station Blackout) (SBO) is not analyzed for reload cores.

15.9.1 Identification of Caus es and Frequency Classification The Station Blackout Rule (Reference 1), requires that each light- water-cooled nuclear power plant licensed to operate must be able to withstand for a specified duration and recover from a station blackout. This event is not given a frequency classification based on event frequency catego ries, but is required to be analyzed per the Station Blackout Rule, 10 CFR 50.63.

Station Blackout occurs as a result of a Loss Of Off-site Power (LOOP) in conjunction with a loss of on-site AC power, failure of Diesel Generators 0 and 1A or 2A. Diesel Generators 1B and 2B are assumed to be available to support the operation of the HPCS system during th e Blackout, but are not classified as "Alternate AC" power sources, because Division 3 does not supply power to safe shutdown loads. Therefore, even though Diesel Generators 1B and 2B are available, LaSalle coping analysis uses the AC-independent approach.

15.9.2 Sequence of Events and System Operation

a. The first 50 seconds of a Station Blackout are the same as for the Loss of A-C Power event in section 15.2.6. The initial conditions for the analysis of a Station Blackout are included in Table 15.9-1. Station Blackout is assumed for analysis to occur on both units. Immediately prior to the postulated Station Blackout event, the reactor and supporting systems are within normal operating ranges for pressure, temperature, and water level. All plant equipment is either normally operating or available from the standby state.

The Division 3 Diesel Generators are assumed to operate

normally during the Station Blackout, allowing the HPCS systems to supply make-up water to the reactor vessel from the suppression pool. Also, RCIC systems are assumed to operate normally during the Station Blackout.

The Station Blackout coping duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on:

a. Plant offsite AC power design characteristic Group "P1", b. Emergency AC (EAC) power configuration Group "D",

LSCS-UFSAR 15.9-2 REV. 19, APRIL 2012 c. Target Emergency Diesel Generator (EDG) reliability of 0.975.

15.9.3 Station Blackout Coping Capability 15.9.3.1 Condensate Inventory for Decay Heat Removal

LaSalle's station blackout coping method uses RCIC or HPCS to provide makeup water for core cooling. The HPCS system normally takes suction from the suppression pool, and RCIC suction automatically transfers to the suppression pool on low condensate storage tank level. De cay heat is removed by discharge of steam through the Safety/Relief Valves into the suppression pool, where the steam is condensed. As a result, gradual heatup of the suppression pool is expected.

Per analysis the suppression pool water inventory is sufficient to make up for decay heat removal requirements and expected leakage during a four hour station blackout. The suppression pool temperature will remain below the heat capacity temperature limits while providing this water, if cooldown is limited to 20°F/hr. (Reference 13).

The condensate inventory analysis was performed to determine if the suppression pool contains sufficient water for a four-hour SBO with the reactor at cooldown (Using either RCIC or HPCS cooling system).

The analysis (Reference 22) assumes:

a. Water is not available from the cycled condensate tanks.
b. Losses in suppression pool inventory are associated with changes (an increase) to reactor vessel inventory, T.S. leakage to the drywell, RR pump seal leakage, and RCIC pump turbine gland seal leakage. These inventories do not return to the suppression pool.
c. The Reactor Coolant System leakage is 61 GPM.

The total amount of water needed for deca y heat removal, cooldown, and leakage is 142,703 gallons, HPCS mode, and 140,349 ga llons, RCIC mode, of which 101,921 gallons (HPCS) and 99,447 gallons (RCIC) is returned to the pool.

With applied margin, the calculated suppression pool water level at the end of a four hour SBO (plus 15-min recovery period) is 698.04 feet (a drop of 1.5 feet), for either HPCS or RCIC operation. NPSH calculations for the SBO coping period use an analytical suppression pool water le vel of 696.88 ft. This value includes additional applied margin beyond the minimum computed level above (Reference 22).

LSCS-UFSAR 15.9-2a REV. 19, APRIL 2012 The Reactor Coolant System inventory anal ysis demonstrates that the RCIC system is capable of maintaining the water level abo ve the top of active fuel (Reference 22) assumes that:

a. The reactor is operating at full power of 3559 MWt, dome pressure of 1040 psia, and normal water level at time of initiation. (The power level and dome pressure are analytical values that include uncertainty above the plant operating limits to allow for instrument uncertainty).
b. The reactor scrams at event initiation.
c. The MSIVs are fully closed in 5 seconds.
d. The ANSI/ANS 5.1-1979 decay heat correlation is used.
e. RCIC initiates automatically when water level drops to Level 2.
f. Operator action to control the maximum depressurization cooldown to a rate of 20° F/hr is assumed fo llowing RCIC startup.

LSCS-UFSAR 15.9-3 REV. 20, APRIL 2014 Evaluations performed by GE for LaSalle indicate that the introduction of GE14 and GNF2 fuel designs (References 15 and 23, re spectively) does not affect the existing decay analysis design basis and, as a result, is not expected to affect this aspect of the station blackout coping capability.

AREVA documents their disposition for the ATRIUM-10 reload fuel and the ATRIUM 10XM LTA at MUR conditions in Reference 21 which confirms that this event is bounded by the events that are analyzed each reload by AREVA.

15.9.3.2 Class 1E Battery Capacity The 125 Vdc (Divisions 1 and 2) and 250 Vdc Class 1E batteries are sized to provide

SBO loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. A calculation was performed to ensure that these batteries have sufficient capacity (i.e.) a minimum remaining margin of 5% to meet the station blackout loads for four hours assuming that loads not needed to cope with a station blackout are shed. The required loads include power restoration from either the emergency ac power supplies or the preferred power source. The loads that need to be shed are listed in Table 15.

9-2. The shedding of these loads is proceduralized (Reference 5 and 6).

The methodology used to determine battery capacity is as follows (Reference 6):

The existing 125 Vdc and 250 Vdc computer models were used for this SBO calculation because they include the battery manufacturer's characteristics required in determining battery sizing, all dc loads (with their various characteristics such as inrush and continuous current, when the load is energized, and the duration of the load), and they calculate the required number of positive plates and the battery capacity remaining. The loads for SBO were verified to be energized for the entire four-hour SBO plus recovery, unless the load was shed. The exceptions to this are switchgear breaker cubicles, excitation cu bicles, fire protection sirens, the TIP panel, and the diesel generator field flashing loads which are verified to be energized for fifteen minutes or less after the inception of SBO.

LSCS-UFSAR 15.9-4 REV. 19, APRIL 2012 The design margin for this study was assumed as 1.0. The maximum temperature was set at the Technical Specification limits.

The aging factor may be adjusted (less than 1.25) to maintain a minimum 5% re maining margin. Appropriate battery performance procedure(s) require verification that the batteries have a minimum capacity consistent with the aging factor us ed in the station blackout battery sizing calculation. The temperature factor, de sign margin and aging factor are all incorporated into the computer models. Note , battery sizing is calculated using the methodology of IEEE-485.

15.9.3.3 Compressed Air There is no AC power to station air compressors during SBO, however, instrument nitrogen is required for th e relief mode operation of the main steamline SRVs. The Automatic Depressurization System (ADS) va lves (7 of the SRV's) have existing backup nitrogen bottle banks that have been analyzed to ensure they are sufficient to support SRV actuations for the four-hour coping duration. Manual opening and closing of individual ADS valves, to depressurize the reactor in a controlled manner, requires sending an operator to the Auxiliary Electric Equipment Room (AEER).

Emergency lighting and communications alre ady exist to gain access to the AEER and to utilize the required controls. Co ntrol of the ADS valves from the AEER can be established within 20 minutes. During this time, the Low Low Set function of five of the ADS valves will automatically control reactor pressure in the normal operating range. The evaluation of the nitrogen bottle bank is based on the opening of the individual SRVs as necessary to ma intain an average 20°F/hr. cooldown over the four-hour coping period (Reference 13). Additionally, the ADS valves (all 7 at once) can be manually initiated by either of two divisions of DC logic from the Control Room. As a backup, the mechanic al safety mode of SRV operation is available independent of nitrogen bottles to control pressure (Reference 5).

15.9.3.4 Effects of the Loss of HVAC The areas of concern at LaSalle Station due to the loss of ventilation were chosen from rooms that based on documented engineering judgement (1) contained SBO response equipment, (2) have substantial heat generation terms and (3) lack normal heat removal systems due to the blackout. The Main Control Room, Auxiliary Electric Equipment Rooms (AEER), and the RCIC room satisfy these criteria.

Areas immediately adjacent to these areas of concern were considered, as well as the floors immediately above and below, in determining heat flow and heat contributions. Additionally, a temperature transient analysis was performed on both the Drywell and Suppression Pool to determine the maximum expected temperature and equipment operability.

The Main Steam Tunnel was considered for the temperature heat up analysis but a review revealed that it did not cont ain SSD equipment credited for SBO, nor RCIC isolation temperature instrumentation.

LSCS-UFSAR 15.9-5 REV. 17, APRIL 2008 The RCIC pipe tunnel contains ambient and differential temperature instrumentation for steam leak detection. However, the RCIC turbine isolation valves which are affected by this logic (1(2)E51-F008, 1(2)E51-F063, and 1(2)E51-F076) are AC powered and AC controlle

d. These valves are maintained in the position required for proper operatio n of the RCIC system when the system is lined up in the standby condition, that is, 1(2)E51-F008 and 1(2)E51-F063 are open and 1(2)E51-F076 is closed. Thus, during an SBO event the loss of HVAC is not an isolation concern for RCIC operation as these valves remain in their required positions on a loss of AC power.

The HPCS diesel is available to power the HPCS pump and its associated systems during a station blackout. This source powers ventilation in the HPCS rooms. Since ventilation will be provided if the HPCS is used during a station blackout, equipment operability is established and no he atup analysis is required in this area (Reference 6).

The Main Control Room, AEER, and RCIC temperature transient calculations are based on evaluation which included a surveillance review of summertime room temperatures. The RCIC room temperature calculations also included the maximum calculated steam leakage from the RCIC turbine gland seals at a conservative turbine backpressure of 50.0 psig (Reference 13). The initial room temperatures and heat loads used in the calculations necessitated procedural requirements to open panel doors in both the Main Control Room and AEER during this event. The access doors to the AEER and Main Control Room are not required to be open during a SBO for cooling purposes. NUMARC 87-00 (Reference 3) provides guidelines for determining "Reaso nable Assurance of Operability," (RAO).

The initial temperatures assumed, final temperatures at the end of a four hour SBO, and the RAO justification for the Main Control Room, AEER areas, and RCIC rooms are provided in Table 15.9.3.

The control room, AEER, and RCIC room temperatures are monitored daily. If the initial temperatures assumed in the SBO analyses are exceeded, then appropriate action should be taken to investigate the problem and resolve it in a timely manner.

15.9.3.5 Primary Containment Calculations The initial conditions for the primar y containment calculations are:

Suppression Chamber water temperature:

105 °F Drywell Air temperature:

135 °F Reactor Recirculation pump seal leakage per pump:

18 gpm Technical Specification Reactor coolant system leakage rate (excluding RR pump seal leakage):

25 gpm LSCS-UFSAR 15.9-5a REV. 14, APRIL 2002 The calculations for Suppression Chamber water temperature assume an average reactor pressure vessel depressurization rate of 20

°F/hr (Reference 13). The slower the depressurization rate, the slower the heatup of the suppression chamber water. The water temperature after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 15 minutes with no cooling is 200

°F LSCS-UFSAR 15.9-6 REV. 17, APRIL 2008 with HPCS supplying RPV inventory makeup during SBO and 196

°F with RCIC supplying RPV inventory makeup during SBO. The cooldown rate is administratively controlled.

There were calculations performed for two different major assumptions for determination of drywell air temperature at the end of a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO. The analysis assumed that the reactor coolant system remains at saturated conditions of 1020 psia and the corresponding saturation temperature. Also, all the primary system leakage (61 gpm) is assumed to stay in the drywell atmosphere, with no atmosphere transfer to the suppression chamber. This results in a final drywell air temperature of 315

°F, assuming the reactor is not depressurized du ring a 4-hour station blackout. Should the reactor be depressurized and normal venting of the drywell to suppression pool occur, this temperature will be lower (Reference 6 and 8).

Therefore, the Drywell maximum air temperature decreases as reactor depressurization rate increases. Thus, there is a trade-off between having lower drywell air temperature by increasing reactor depressurization rate versus a lower final suppression pool temperature by a slower reactor depressurization rate. However, the analyses bound depressurization rates from 0

°F/hr up to and including an average of 20

°F/hr (Reference 13).

The equipment qualification curve for the drywell is a step function with the following temperature limits:

340 °F from 0 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 320 °F from 3 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 250 °F from 6 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Thus, the EQ equipment inside the drywell is designed to operate at 320

°F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and envelopes the SBO drywe ll temperatures (Reference 8).

15.9.3.6 Containment Isolation The list of containment isolation valves in UFSAR Table 6.2-21 were reviewed to ensure that the isolation functions and positi on indication can be provided during an SBO event. Position indication is considered acceptable if it includes local mechanical indication, DC powered indication (including AC indicators powered from inverters) and HPCS DG powered indication. The acceptable means of valve closure include manual operation, air operation (including air operated valves that are mechanically closed on loss of air), DC powered operation and HPCS DG LSCS-UFSAR 15.9-7 REV. 17, APRIL 2008 powered operation. As recommended in NUMARC 87-00 the following criteria were used to exclude valves from consideration:

1. valves normally locked closed during operation; 2. valves that fail closed on loss of AC power or air; 3. check valves;
4. valves in non-radioactive clos ed-loop systems not expected to be breached in a station blackout (with the exception of lines that communicate directly with the containment atmosphere); and, 5. all valves less than 3-inch nominal diameter.

Since independent valve failures are not assumed to occur during a station blackout, a valve in line with an excl uded valve was also excluded from consideration. In addition, valves which continue to be powered and operable during a station blackout do not require manual operation capability. Table 5-1 of Reference 6 lists the valves reviewed and their exclusion justification. When multiple valves are in line with one penetr ation, all the valves are listed but only one valve would need to be closed. Table 15.9-4 lists the valves that would require operator action and verification of closure.

Full containment isolation is not expected to be necessary as a result of a station blackout. However, Regulatory Guide 1.155 requires reactors to have the ability to maintain "containment integrity" in station blackout conditions should this be necessary for other reasons. Such other reasons could include requirements to close certain valves following a loss of offsite power, loss of decay heat removal capability, or other casualties affecting the reactor coolant system (Reference 6).

15.9.3.7 Recovery from a Station Blackout The SBO analyses for suppression pool temperature assume that suppression pool cooling is established 15 minutes following a SB O. The suppression pool temperature 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 15 minutes after a SBO occurs is 200

°F when using the HPCS system and 196

°F when using the RCIC system for decay heat removal and reactor coolant inventory (Reference 13). These temperatures are within the environmental qualification temperature rating for HPCS materials, the RCIC pump materials, and RHR materials (Reference 8).

Per Reference 16, the Net Positive Suction Head (NPSH) requirements versus the Net Positive Suction Head Available at the end of a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO plus a 15 minute allowance to start an RHR pump in the Suppression Pool Cooling mode are as follows: 1. RCIC Pump NPSH Available (at pump inlet centerline) is 27.25 feet.

LSCS-UFSAR 15.9-8 REV. 17, APRIL 2008 2. RCIC Pump NPSH Required (at pump inlet centerline) is 20.5 feet.

3. HPCS Pump NPSH Available (at pump inlet centerline) is 16.7 feet.
4. HPCS Pump NPSH Required (at pump inlet centerline) is 6.0 feet.
5. RHR Pump NPSH Available (at pump inlet centerline) is 16.45 feet.
6. RHR Pump NPSH Required (at pump inlet centerline) is 15.0 feet.

The RCIC turbine backpressure was dete rmined based on the worst case suppression pool water levels, suppression chamber pressure and RCIC turbine exhaust flow following the SBO. The calculated maximum RCIC turbine

backpressure is 36.4 psig (Reference 13).

This pressure is below the RCIC turbine backpressure trip setpoint of 43.7 psig (Reference 8 and 11).

15.9.4 Quality Assurance A QA program meeting the requirements of Regulatory Guide 1.155 Appendices A and B has been applied to cover non-safety related equipment needed for coping

with a station blackout that were not alre ady covered by existing QA requirements in Appendices B or R of 10 CFR 50 (References 5 and 8).

15.9.5 References

1. 10 CFR 50.63, Loss of All Alternating Current Power.
2. Regulatory Guide 1.155, Rev. 0, Dated June 1988; Station Blackout.
3. NUMARC 87-00, Rev. 1, Date d August, 1991; Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors.
4. Letter dated April 17, 1989 from M.H. Richter to Dr. T.E. Murley, Director of the Office of Nuclear Reactor Regulation; Response to Station Blackout Rule, including NRC Docket Nos. 50-373 and 50-374.
5. Letter dated June 22, 1990 from M.H. Richter to Dr. T.E. Murley, Director of the of Nuclear Reactor Regulation; LaSalle County Station Units 1 and 2, Revised Office Response to Station Blackout Rule, NRC Docket Nos. 50-373 and 50-374.
6. Letter dated September 23, 1991 from P.L. Piet to the Office of Nuclear Reactor Regulation; LaSalle County Station Units 1 and 2, LSCS-UFSAR 15.9-9 REV. 20, APRIL 2014 Supplemental Response to Station Blackout Rule, NRC Docket Nos. 50-373 and 50-374.
7. Letter dated March 6, 1992 from B.L. Siegel, NRR Project Manager; Safety Evaluation of the LaSalle County Station Response to the Station Blackout Rule (TAC Nos. M68559 and M68560).
8. Letter dated May 15, 1992 from J.N. Shields to Dr. T.E. Murley, Director of the Office of Nuclear Reactor Regulation; LaSalle County Station Units 1 and 2, Response to Safety Evaluation on the Station Blackout Rule, NRC Docket Nos. 50-373 and 50-374.
9. Letter dated July 17, 1992 from B.L. Siegel, NRR Project Manager; Safety Evaluation Related to Station Blackout Analysis, LaSalle County Station, Units 1 and 2 (TAC Nos. M68559 and M68560.
10. Chron# 300100 dated March 28, 1994, LaSalle County Station Units 1 and 2 Station Blackout Analysis Review of S&L Calculations ATD-0117 and 3C7-0390-011.
11. Chron #302940 dated September 30, 1994, LaSalle County Station, Units 1 and 2, Approval of RCIC Turbine Exhaust High Pressure Trip Setpoint Change.
12. Calculation L-001260, ECCS & RCIC Suppression Pool Suction Strainer Head Loss for a 50% Plugged Strainer.
13. NDIT LAS-ENDIT-1255, Upgrade 3, "Station Blackout".
14. NEDO-32701P, Rev. 2, Power Uprate Safety Analysis Report for LaSalle County Station Units 1 and 2, July 1999.
15. Section 3.0, GE Document GE-NE-0000-0026-4769-00, Revision 0, "GE14 Fuel Design Cycle-Independent Analyses for LaSalle Unit 1 and Unit 2", dated January 2005.
16. Calculation L-002540, "NPSH Margin for HPCS, RHR, and RCIC Pumps, Backpressure for RCIC Turbine." 17. Calculation ATD-0351, RCIC Pump Room Temperature Transient Following Station Blackout With Gland Seal Leakage.
18. Intentionally Deleted.

LSCS-UFSAR 15.9-9a REV. 20, APRIL 2014

19. Calculation (Design Analysis) 3C7-0290-001, Main Control Room Temperature Transient Following Station Blackout.
20. Calculation (Design Analysis) 3C7-0289-001, Aux Electric Equipment Room Temperature Transient Following Station Blackout.
21. Design Analysis L-003509, Revision 0, "Evaluation of Appendix R, Station Blackout, Containment and Source Terms for LaSalle MUR Power Uprate," July 2010.
22. Calculation (Design Analysis) 3C7-0189-001, Station Blackout Condensate Inventory Coping Assessment.
23. Section 3.0, GNF Document NEDC-33647P, Revision 2, "GNF2 Fuel Design Cycle-Independent Analyses for Exelon LaSalle County Station Units 1 and 2," dated February 2012.

LSCS-UFSAR 15.10-1 REV. 17, APRIL 2008 15.10 Thermal Hydraulic Instability Transient

This section covers events that result in a thermal hydraulic instability event.

Additional information regarding the transient and the system designed to respond to

it, namely the Oscillation Power Range Mo nitor (OPRM) system, is contained in chapters 4 and 7.

15.10.1 Identification of Causes and Frequency Classification

Events such as Reactor Recirculation (RR) pump trips and runbacks, turbine/generator runbacks, loss of feedwater heating, and RR flow controller

failures can result in unplanned entry into the high power and low flow region of

the power to flow map. Under these conditions, axially varying moderator density

in the fuel channels can cause flux oscillations that increase in amplitude. Without

manual or automatic suppression, such os cillations can cause the MCPR Safety Limit to be exceeded (Reference 15.10.6.1).

This event is controlled by a system designed for detection and suppression of

oscillations in accordance with GDC 10 and

12. The system is the Oscillation Power Range Monitor (OPRM) system. It provides automatic protection for this event, when it is installed and fully functional. For operation prior to the installation of

OPRM, or when OPRM is not fully functional, the operator controls the oscillations

by scramming the reactor upon entry into th e region of power to recirculation flow map when such oscillations are possible.

Anticipated stability-related neutron flux oscillations are those instabilities that result

from normal operating conditions, including conditions resulting from anticipated

operational occurrences. This category of events is equivalent to the standard

terminology for the analysis of events of moderate frequency (Reference 15.10.6.2).

15.10.2 Sequence of Events and System Operation

For this event, the plant must be operating in mode 1.

A. As a result of some manual actions or equipment problems (e.g., RR pump runback, loss of feedwater heating), the core power and flow

combination may be such that oscillations of neutron flux may be

possible.

B. Due to forced flow being inadequate to control density wave transit time up the fuel channels, flux oscillations start and begin to increase

in amplitude.

C.1 If the OPRM is not operable, the operator manually scrams the reactor upon recognition of the instability.

LSCS-UFSAR 15.10-2 REV. 17, APRIL 2008 C.2 With the OPRM operable, the operator may be able to take action based on pre-trip alarms to insert control rods or increase flow. If not able to because of the rate of increasing oscillations, the OPRM

automatically scrams the reactor be fore the Safety Limit MCPR is violated.

15.10.3 Core and System Performance

The OPRM system contains 4 LPRMs per OPRM cell (using the Bockstanz-

Lehmann LPRM assignment methodology described in Reference 15.10.6.3) and

requires 1 LPRM input for the cell to be operable. The amplitude setpoint for

oscillation magnitude and the number of confirmation counts are specified for the

analysis. Since core thermal hydraulic instability is characterized by a consistent

period for the oscillations, the OPRM logic includes a check for a set number of

consecutive counts as well as a magnitude.

The specified system setpoints are used to determine the hot bundle oscillation

magnitude. This information is used, along with empirical data applicable to the

fuel in the core, to determine the fractional change of CPR (delta CPR/IMCPR, where IMCPR is initial MCPR).

The Initial (pre-oscillation) MCPR (IMCPR) is determined as the lower of the following:

1. The MCPR following a dual RR pump trip from rated power on the highest allowed flow control line, after the coastdown to natural

circulation and after feedwater temperature reaches equilibrium. The

assumption is that the core was operating at the Operating Limit MCPR

prior to the dual pump trip.

2. The MCPR Operating Limit with the reactor at steady state conditions at 45% core flow on the highest allowed flow control line.

The Final MCPR (FMCPR) is determined using the IMCPR and CPR/IMCPR data (Reference 15.10.6.3).

The FMCPR is then verified to be greater than the Safety Limit MCPR.

Alternatively, a minimum IMCPR can be determined for a given Safety Limit and

checked against the cycle specific Operating Limit (Reference 15.10.6.4).

If the minimum IMCPR is greater than the Operating Limit determined from other

cycle analyses, or the FMCPR is less th an the Safety Limit MCPR, the system setpoint may be changed and the reload confirmation performed again.

Alternatively, the Operating Limit MCPR may be changed, or the LPRM

assignment scheme may be modified.

The above is confirmed for each cycle as part of the reload analysis.

LSCS-UFSAR 15.10-3 REV. 17, APRIL 2008 15.10.4 Barrier Performance

Since the successful completion of this analysis demonstrates that the MCPR Safety

Limit is not exceeded, fuel-cladding integrity is not challenged.

15.10.5 Radiological Consequences

Since fuel-cladding integrity is not challenged, there are no radiological

consequences warranting evaluation for this event.

15.10.6 References

15.10.6.1 USNRC Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors," dated July 11, 1994. 15.10.6.2 GE Document NEDO-31960-A, Supplement 1, "BWR Owner's Group Long-Term Stability Solutions Licensing Methodology (Supplement 1)," dated November 1995. 15.10.6.3 GE Document NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload applications," dated August 1996. 15.10.6.4 BNDG 96-011

LSCS UFSAR TABLE 15.B-1 REV. 14, APRIL 2002 TABLE 15.B-1 TRANSIENT ANALYSIS POWER / FLOW STATE POINTS AT POWER UPRATE Transient Event (100P =3489 MWt) 100P/105F Nom FWT 100P/105 FFWTR =100F 1 100P/81F Nom FWT 102P/105F Nom FWT 2 LRNBP X X TTNBP X X FWCF X X X LRNBP w/o RPT X TTNBP w/o RPT X FWCF w/o TBP X X MSIVF X LRNBP w/o RPT 3 X Loss of Feedwater

Heater 4 X Source of information: Reference 59.

Notes:

1. Initial feedwater temperature of 326.5

°F assumes Feedwater Temperature Reduction of 100

°F from the normal feedwater temperature at 3489 MWt.

2. Overpressurization event, analyzed at a dome pressure of 1035 psia and a 2%

overpower for initial power. The nominal feedwater temperature for 102% power is 428.5 °F.

3. Turbine Control Valve slow closure, only TCV #1 closes at 50% per second.
4. 145 °F worst loss of feedwater temperature.

LSCS UFSAR TABLE 15.B-2

REV. 14, APRIL 2002 TABLE 15.B-2 TRANSIENT ANALYSIS INIT IAL PLANT CONDITIONS Parameter FSAR Basis

  • Cycle 8 Basis Power Uprate Basis Rated Thermal Power (MWt) 3296 3323 3489 Analysis Power (% Rated) 104.8 100 100 Rated Core Flow (Mlb/hr) 108.4 108.5 108.5 Rated Power Core Flow Range

(% of rated) 87 - 105 81 - 105 Rated Vessel Steam Flow and FW flow (Mlb/hr) 14.24 14.30 15.14 Analysis Steam Flow (% rated steam flow) 104 100 100 Analysis Dome Pressure (psig) 1020 986.3 ** 986.3 **

Analysis Turbine Pressure (psig) 962 933.1 926.3 Normal Feedwater Temperature ( °F) 420 420 426.5 Feedwater Temperature Reduction, ( °F) 100 100 Steam Bypass Capacity (% rated steamflow) 25 25 23.6 Number of SRVs assumed in Analysis 18 17 12 SV Setpoint

(# of valves @ psig)

Analytical Limit 2 @ 1150 4 @ 1175 4 @ 1185 4 @ 1195 4 @ 1205 2 @ 1185 ***

4 @ 1210 4 @ 1221 4 @ 1231 4 @ 1241 1 @ 1210 ***

4 @ 1221 4 @ 1231 4 @ 1241 RV Setpoint

(# of valves @ psig)

Analytical Limit 2 @1076 4 @1086 4 @1096 4 @1106 4 @1116 2 @ 1091 ***

4 @ 1101 4 @ 1111 4 @ 1121 4 @ 1131 1 @ 1101 ***

4 @ 1111 4 @ 1121 4 @ 1131 MCPR Safety Limit 1.06 1.07 1.07 CRD Speed Figure 15.0-2 67B 67B Source of information: Reference 59.

  • These values are for the initial core analysis.
    • For MCPR calculations a conservatively lower operating pressure is assumed in the analysis. For the overpressure analysis the technical specification maximum dome pressure of 1020 psig is used in the analysis. *** The lowest setpoint valve is assumed Out-of-Service.

LSCS UFSAR TABLE 15.B-3 REV. 14, APRIL 2002 TABLE 15.B-3 LASALLE POWER UPRATE TRANSIENT ANALYSIS RESULTS Transient (a) Initial Power /Flow (b)

Peak Neutron Flux (%NBR) Peak Heat Flux (%NBR) Peak Steamline Pressure (psig) Peak Vessel Pressure(psig) CPR GE8x8NB (c) Notes** Equipment

In Service { } = Cycle 8 Results LRNBP 100P/105F 487 {491} 119 {119} 1171 {1152} 1206 {1190} 0.21 {0.20} d d (Ref 4) TTNBP 100P/105F 474 118 1170 1205 0.19 d FWCF 100P/105F 375 118 1137 1172 0.17 d FWCF 100P/105F 350 {345} 121 {121} 1129 {1116} 1156 {1145} 0.18 {0.18} d,g d,g (Ref 4) MSIVF 102P/105F 493 {467} 130 {130} 1303 {1260} 1332 {1296} d d (Ref 4) LRNBP 100P/81F 338 116 1174 1202 0.15 d TTNBP 100P/81F 307 114 1173 1201 0.14 d FWCF 100P/81F 253 113 1139 1167 0.12 d LFWH 100P/81F 0.17 Equipment out of Service { } = Cycle 8 Results LRNBP 100P/105F 592 {595} 124 {124} 1174 {1154} 1216 {1200} 0.24 {0.24} d,e d,e (Ref 4) TTNBP 100P/105F 602 123 1172 1215 0.23 d,e FWCF 100P/105F 474 126 1164 1200 0.23 d,f,g LRNBP 100P/105F 426 125 1192 1225 0.26 d,e,h FWCF 100P/105F 523 124 1170 1204 0.23 d,f Source of information: Reference 59.

FOOTNOTES for Table 15.B-3 (a) LRNBP = Generator Load Rejection with Bypass Failure FWCF = Feedwater Controller Failure (to maximum demand) TTNBP = Turbine Trip with Bypass Failure MSIVF = Main Steam Isolation Valve Closure, Flux Scram HPCS = Inadvertent Actuation of High Pressure Core Spray LFWH = Loss of Feedwater Heaters (145 deg F FWTR) (b) 100P = uprate power of 3489 MWt 100F = rated core flow of 108.5Mlb/hr 105F = ICF flow point at uprated power 81F = MELLLA flow point at uprated power (87.9 Mlb/hr) (c) CPR based on initial CPR which yields MCPR = 1.07

    • Notes: d = 1SRVOOS e = RPTOOS f = TBSOOS g = FFWTR h = TCV slow closure

LSCS UFSAR TABLE 15.B-4

REV. 14, APRIL 2002 TABLE 15.B-4 LASALLE POWER UPRATE TOP/MOP RESULTS Transient (a) Initial Power/Flow (b) TOP (%) TOP Design limit (%) MOP (%) MOP design limit (%) Notes** Equipment In Service { } = Cycle 8 Results LRNBP 100P/105F 24.9 {24.9} 38.0 {38.0} 25.3 {25.2} 38.0 {38.0} d d (Ref 4) TTNBP 100P/105F 23.8 38.0 24.2 38.0 d FWCF 100P/105F 20.2 37.0 20.9 39.0 d FWCF 100P/105F 21.8 {21.9} 37.0 {37.0} 23.7 {23.6} 39.0 {39.0} d,g d,g (Ref 4) LRNBP 100P/81F 23.0 38.0 23.4 38.0 d TTNBP 100P/81F 22.6 38.0 22.9 38.0 d FWCF 100P/81F 17.8 37.0 18.2 39.0 d LFWH 100P/81F 37.3 25.0 37.3 45.0 Equipment Out of Service { } = Cycle 8 Results LRNBP 100P/105F 30.1 {30.3} 38.0 {38.0} 30.6 {30.6} 38.0 {38.0} d,e d,e (Ref 4) TTNBP 100P/105F 29.9 38.0 30.4 38.0 d,e FWCF 100P/105F 28.6 37.0 30.0 39.0 d,f,g LRNBP 100P/105F 30.7 38.0 30.9 38.0 d,e,h FWCF 100P/105F 27.5 37.0 28.4 39.0 d,f Source of information: Reference 59.

FOOTNOTES for Table 15.B-4 (a) LRNBP = Generator Load Rejection with Bypass Failure FWCF = Feedwater Controller Failure (to maximum demand) TTNBP = Turbine Trip with Bypass Failure MSIVF = Main Steam Isolation Valve Closure, Flux Scram HPCS = Inadvertent Actuation of High Pressure Core Spray LFWH = Loss of Feedwater Heaters (145 deg F FWTR) (b) 100P = uprate power of 3489 MWt 100F = rated core flow of 108.5Mlb/hr 105F = ICF flow point at uprated power 81F = MELLLA flow point at uprated power (87.9 Mlb/hr)

    • Notes: d = 1SRVOOS e = RPTOOS f = TBSOOS g = FFWTR h = TCV slow closure

LSCS UFSAR TABLE 15.B-5

REV. 14, APRIL 2002 TABLE 15.B-5 ARTS VERIFICATION FOR MAPFAC(P) RESULTS Power Dependent MAPLHGR Factor without TPS Credit 25% P < 85% MAPFAC(p) = 0.

910 + 0.00503(P-85) 85% P < 100%

MAPFAC(p) = 0.910 Power Dependent MAPLHGR Factor with TPS Credit 25% P < 95% MAPFAC(p) = 0.

960 + 0.00503(P-95) 95% P < 100%

MAPFAC(p) = 0.960 Source of information: Reference 59.

LSes*UFSAR UIO 140 120*'" 3: g-Q in.i.......a;>C;;('"=>e S'"'" c: H'00 120 o__.L....L o eo 100§l-e'"!'!a: '".0 oJ e::I'"!eo COJI'EFLOW C1I.AATEOt LASALLE COUNTY STATION UPDATED FINAL SAFETY Al"lALYSIS REPORT FIGURE 15.0*1 TYPICAL POWERJFLOW MAP REV.13 GE-NE-0000-0099-8344-R1 TASK REPORT T0201 GEH PROPRIETARY INFORMATION 3-7 Figure 3-2: Revised LaSalle Power Flow Map (Revised TPO - ~101.65% CLTP) 0 10 20 30 40 50 60 70 80 901001101200102030405060708090100110120Core Flow (%)

Thermal Power (% Revised TPO) 0 400 800120016002000240028003200360040000102030405060708090100110120130Core Flow (Mlbm/hr)

Thermal Power (MWt) 100.0% Revised TPO = 3546 MWt 100.0% CLTP = 3489 MWt100.0% Core Flow = 108.5 Mlbm/hr A: 60.0% Power / 36.1% Flow B: 100.0% Power / 82.8% Flow B': 98.4% Power / 80.8% FlowC: 100.0% Power / 100.0% Flow C': 98.4% Power / 100.0% FlowD: 100.0% Power / 105.0% Flow D': 98.4% Power / 105.0% Flow E: 56.2% Power / 105.0% Flow F: 22.5% Power / 48.0% Flow G: 22.5% Power / 32.6% Flow H: 49.3% Power / 31.5% FlowMELLLA BoundaryP = (22.191+0.89714W T-0.0011905W T 2)(1.132)3546 MWt3489 MWt ICFNatural CirculationTwo PumpMinimum Flow LineFlow Control ValveCavitation ProtectionJet Pump and Recirc PumpCavitation Protection A B B'C C'D D'E F G HTPO 100% Load LineTPO 66.7% Load LineTPO 71.0% Load Line LSCS-UFSA R REV. 19, APRIL 2012LASALLECOUNTYSTATIONUPDATEDFINALSAFETYANALYSISREPORTFIGURE15.0-1(a)LASALLEPOWER/FLOWMAP

-1.30 Haxlluum allowdble (ore averdye serLllll lIme to noldl pos ilion 39 wilen operat ioy under lhe GLMCPR U 1 j",1t of 1.24 Mepl!.lectlfllcd I Spec lficaLi on I illl1 l on core s(ram lillie to notch pas i tlon19.whtot'L" Salle i>0.lJ60>econos.measured core S((dfll L iUt!to Holch pos it iOIl 39.OLMCPR A OperaIi ny li.1l t bd sed on OHm A.OLMCPR a c limit bJ,ud on OI!YN Opllull B.Tlte Operating Iilllit HeI'R at rated core flow I, the do",i/lating lacu'af event HCPR value, fnJil.Figore.The La Sa 11 e"eador, dre I illlited by the rGd withMawd 1 error evenl (Rwf)at an OlMCPRU'1.24 up to thaI"oHlt wllere the generator load reject event wi UIOU[llyi(tR WID BPI intersects

,,"d dominates the IIW[event.Then the OLMCPRU beCOlIlCS 1.26 at the vdloe of['1\'Applicaole definitions

<Ire as fall""" 1.10 1.15 I.20 I.25-----MINIMUM OI'EllATiNG U'R LIMIT-----CVCNT HCPR...---_....------------------------------LR W UP_----------Rill l.fWIi_------_----.....

FIIC"

LII 11/0 01'_--..._---""",,--------;;W/O up_----.--

---_...I.ltJ 1.25 1.3D1.2U*0.5 T

-T[lJT A-T a-LA SALLE COUNTY STATION FINAL SI\F£l¥ANALYSIS RLPOHI flliUHf 15.0-2 INI'rIAL CORE t1HHMtJlt OreRA!I N6 U'H I Ii'll T REV.4-APRIL 1988 LSCS-UFSAR!liB-03 i§Ii!2-14.c:i!g i=aD.LlllI..,.lIG:lacI1

!-0!!f!Ii!il.....il"::'i--I:-.

..............§-LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-1 (INITIAL CORE)LOSS OF FEEDW ATER HEATER AUTO FLOW CONTROL MODE REV.13 LSCS*UFSAR,i!j p',...............

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CJ l!..8-.lS

..........,.;..---:.."..---.:,.--

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COIlPOEHTSlfl CI.f-/...,............- j;,CJ......::jIlc c.._--......c......-......LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1,2 (INITIAL CORE)LOSS OF FEEDW ATER HEATER MANUAL FLOW CONTROL REV, 13 LSCS*UFSAR

.."'"--.,;J..leif§IJoJ....r;....w_c:..=.

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..." ll:!i g Q r03ltlll:lO lKDlI3i/I--!i----t---t.-H--*

-!':--!:+----f-tlt-----f-----l

...--';'I.!_c LASALLE COUNTY STATTON UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1*.'1 FEEDWATER CONTROLLER

FAILURE, DEMAND, WITH HIGH WATER LEVEL TRIPS.104%POWER-ODYN REANALYSIS (INITlAL CORE RESCLTS)REV.1.'1 LSCS*UFSAR Iii-I03JtftI Jll JIIDIf30II LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1*4 FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIPS, 104%POWER-ODYN REANALYSIS WITHOUT BYPASS (INITIAL CORE RESULTS)REV.13 LSCS-UFSAR 1--2":-'---S":-*---jj:""........1.I.I.I...t;{*:5-(lB.UII.:m.J.Ja:IBd1

....LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-5 PRESSURE REGULATOR FAILURE TO 115%DEMAND (INITIAL CORE RESULTS)REV.13 LSCS-UFSAR dgg c:i a:i..."'!!!Vl-...:::>11:1...."""'....D_c.>!r=!....':1'1I111 t:Jl:j:10 1N 3Jl:13d I.0 8.11.1 o e.J II.a:i..........oJ...:x........""...l!5 ,.u:i_--u:i_u u Int!t!....:::r LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-1 GENERATOR LOAD REJECTION WITH BYPASS ON 104 LOAD-OD1'N REAi"iALYSIS (INITIAL CORE RESULTS)REV.13 LSCS*UFSAR

'I:-.

..1...........-..1........,..).0 i0 1031 tllI:10 lN3JlGJ1.;.;II>.....J"-IX...II>S!" c'i_:i...........__iilli.o LASALLE COUNTY STATION UPDATED FINAL SAFETY REPORT FIGURE 15.2-2 GENERATOR LOAD REJECTION WITHOUT BYPASS 104%POWER-OD'{N REANALYSIS (INITIAL CORE RESULTS)REV.13 LSCS-UFSAR

--4-IJ---I----+---_.§

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........oa i Q-

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,.I"::II J-,I.i5 Ii if.., iii VI i.., ,..,,;-!...:J'LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-3 GENERATOR LOAD REJECTION WITH BYPASS ON 104%LOAD-ODYN REA.'1ALYSIS (INITIAL CORE RESULTS)REV.13 LSCS-UFSAR!It!-4---+.,..-f--1---___ifa:.

--+----:l:If.,'--......,I-----iS

..&.U........ B ii" (GilIlll IiJ.l.NitlWiWl

.--

LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-4 TURBINE TRIP, TRIP SCRAM, BYPASS-ON, RV-ON, RPT-ON (INITIAL CORE RESULSj REV.13 LSCSUFSAR o o Ii lil I01UI:I itJ JHDBlI*,l£[-I'-..i....... l!t'""I""..:i--""'...-.

lil 8"l'.,..:.:.-,--!iii ileac_N'"..'".c ,-":;,*r..*-,..;.all..0.-..-WI'" rt-"'"'"--..!!t." , LASALLE COUNlY STATION UPDATED FINAL SAFElY ANALYSIS REPORT FIGURE 15.2-5 TURBINE TRIP WITHOUT BYPASS,TRIP SCRAM, 104%POWER-ODYN REANALYSIS (INITIAL CORE RESUL TSI REV.13 LSCS-UFSAR l 1'.'I*s..r ii_e..J...!! s (QlLIAI:IlJ.1HD&Il 1.

I I.--+--......I,-'--_

..................

IoJ.'Q Ii;LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15,2-6 3-SECOND CLOSURE OF ALL MSIV'S 105%POWER (INITIAL CORE RESULTS)REV.13

!i1-4---++-+7-".....----ifi.=y!!!f-iii....fi....!!lEi

  • .w.l..I.W.

....

__--LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT LOSS OF CONDENSER VACUUM AT 2 INCHES PER SECOND 105%POWER, 100%FLOW (INITIAL CORE RESULTS)REV.13 r (J)(")(J)C'"T1 (J);J>;;0 60.30.115.T1HE (SECI 15.120*1 if 0&(.*I*fIJ ttl I[I 1..O.25.50.75.100.COIlE flOW 1%1...zflO.I i:l I" 1 ll.r/1;>I e 80.1 I:/" I Itt'I!I...If 15 ISO.1II Ix DE'" rL 120.1II"i

...120J"!lI",,!

I I ,..O.15.30.ijS.60.TlHE ISH)l'0 c (J)" (J)0'"T1 rd;J>C ul'><r z;J>:;f:: r:;;0.-<o (J)tTl" C;J>(")0;;0'"T10tTl tTl C_:!z;;0 1::;;J>:!-J'Z;;0 00(J);;0;J>t::;;tTl Z.-<-J<(J)(J)-'"T1-0 0 (J)Z....;J>"" tTl 0;;0 OJ;;0 r-J N 0 0 N I'I'I.----

LSCS-UFSAR 1.

l--4-_--:..---+---j'g

-....L--_,1...---I;-&

  • ................

""""-!.c:ii 0 0_1c:J:iUW 4J>>Gl:ll&ll--+---fJiM:tY---+-----is!

&i-LASALLE COUNTY STATION UPDATED FINAL SAFE1Y ANALYSIS REPORT LOSS OF ALL GRID CONNECTIONS 105%POWER, 100%FWW (lNlTIAL CORE RESUL TSj REV.13

si!-8 iii...LSCS-UFSAR l.

1.-

S...JiJ!t1

...--+-+-+-::il-il--+-----:j.s LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-10 LOSS OF ALL FEEDWATER FLOW 105%POWER.IOO%FLOW (INITIAL CORE RESULTS)REV.13 COLD SHUTDOWN ACHIEVED (VESSEL HEAD REMOVED)D NORMAL SHUTDOWN COOLING (ONSIlE POWER)ADS/RELIEF VALVE ALTERNATE PATH ADS/RELIEF VALVE ALTERNATE PATH 100 psi, 330 0 F TO--..J 14.7 ps i, 125 0F-III NORMAL SHUTDOWN z:...COOLING (OFFSITE POWER)DEPRESSURIZE VESSEL VIA MANUAL RELIEF p" I DIESEL...IADS/RELIEF VALVE VALVE ACTUATION t RHR ALTERNATE PATH SUPP.POOL COOLING DEPRESSURIZE VESSEL VIA MANUAL RELIEF VALVE ACTUATION+RHR SUPP.POOL COOLING AUTOMATIC RELIEF VALVE ACTUATION CONDENSER NOT AVAILABLE r---1045 psia, 560 0 F TO 100 psi.330 0 F*DEPRESSURIZE VESSEL TO MAIN CONDENSER w-'co ct-':;;::>ct ct: w 3: 0 0-W I-.....VI Lt..Lt..0 NORMAL SHUTDOWN INITIATED P=1045 psia Vl cII T=560 0 F c:-0 3:§;r 3:......I-):::0;:;1>1-2:::0.....w VI.....-<OUJ Ll..Vl Lt..Z 0..,.,>oct..,.,......r ct: u......-0..,., 0 ct: C):::o.....IJTJ VlW o-t en Vl3: I:J: c: U1()00 o Vl::0-'a..IT1;u Vl):::o trJ:I:<:......<: c:):::0 U1-<Z-t........01 N):::0-1I0:E:co......::Zl......

IT!-t Vl-ll-1>0 Vl-l"0;u):::0;;0-H C)mO t"i:I: 2Sz.....IT!::0 I-'<:-l 1.0 rrl ill ,j:>.

LSCS-UFSAR

...__-1040 psta.549 of II III ,.a, ," r 111"'Ie 14.J lB'r........,-.-.......1040....1-549 of*.-....--..........

,_..':::::........\1 REV.14-.....,.....a.........e--....._.-....LASALLE COUNTY STATION UPDATED FINAL SAFETY ANAL YSIS REPORT FIGURE 15.2-12 ADSlRHR COOLING LOOPS (Sheet 1 of 2)APRll..2002 LSCS-UFSAR REV. 14 APRIL 2002 Activity A Initial pressure - 1040 psia Initial temperature - 549° F For purposes of this analysis, the following worst-case conditions are assumed to exist: a. The reactor is assumed to be operating at 102% nuclear boiler rated steam flow: b. a loss of power transient occurs (see Subsection 15.2.6); and c. a simultaneous loss of onsite power (Division 1 or 2), which eventually results in the operator not being able to open one of the RHR shutdown cooling line suction valves. Activity B Initial pressure = 1040 psia Initial system temperature = 549° F Operator Actions During approximately the first 30 minutes, reactor decay heat is passed to the suppression pool by the automatic operation of the reactor relief valves. Reactor water level will be returned to normal by the HPCS and RCIC system automatic operation. After approximately 10 minutes, it is assumed one RHR heat exchanger will be placed in the suppression pool cooling mode to remove decay heat. The operator initiates depressurization of the reactor vessel to control vessel pressure. Controlled depressurization procedures consist of controlling vessel pressure and water level by using the ADS, and RCIC and HPCS systems. When the reactor pressure approaches 100 psig, the operator would normally prepare for operation of the RHR system in the shutdown cooling mode. Activity C1 (Division 1 fails, Division 2 available) System pressure ~100 psi System temperature ~330°F Operator Actions The operator establishes a closed cooling path as follows: a. Three to five ADS valves (DC Division 2) are powered open; b. Either of the following cooling paths is established: 1. Utilizing RHR loop B, water from the suppression pool is pumped through the RHR heat exchanger (where a portion of the decay heat is removed) into the reactor vessel. The cooled suppression pool water flows through the vessel (picking up a portion of the decay heat) out of the ADS valves and back to the suppression pool. This alternate cooling path is shown in Figure 15.2-13. 2. Utilizing RHR loops B and C together, water is taken from the suppression pool and pumped directly into the reactor vessel. The water passed through the vessel (picking up decay heat) and out the ADS valves returning to the suppression pool as shown in Figure 15.2-4. Suppression pool water is then cooled by operation of RHR loop B in the cooling mode (see Figure 15.2-15). In this alternate cooling path RHR loop C is used for injection and RHR loop B for cooling. Activity C2 (Division 2 fails, Division 1 available) (Figure 15.2-16) System pressure ~ 100 psi System temperature ~ 330° F Operator Actions The operator establishes a closed cooling path as follows: a. Three to five ADS valves (DC Division 1) are powered open; b. Utilizing RHR loop A instead of loop B, and alternate cooling path is established as in Activity C1 Item 2 (a) above. LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-12 ADS/RHR COOLING LOOPS (Sheet 2 of 2)

SIR VA.....E MSL.REACTOR VESSEL SUPPRESSION

"'REA\SERVICE WATER.....--1 DISCHARGE

.1-04-------

RHA SE AVICE WATEA SYSTEM I L HEAT LA SALLE COUNTY STAT/ON UPDATED FINAL SAFETY ANALYSIS REPOKT FIGURE 15.2-13 ACTIVITY Cl ALTERNATE SHUTDOWN COOLING PATH UTILIZING RHR LOOP B REV.a-APRIL 1984 MSL SIR VALVE REACTOR VESSEL SUPPR E SSION AREA J LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-14 RHR lOOP C REV.0-APRIL 1984 SERVice WATER.....-"1 DISCHARGE H-L...-J-....---......-AHA SE RVICE WATER SYSTEM LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-15 RHR LOOP B (SUPPRESSION POOL COOLING MODE)REV.0-APRIL 1984 S'R MSL VALVE DRYWELL REACTOR VESSEL SUPPRESSION AREA LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT 1-_....SERVICE WATER DISCHARGE r-H-;;:;-I EXCHANGER I I I I I RHR SERVICE WATER SYSTEM--+---..':1--1:.....""'1 I I L J FIGURE 15.2-16 ACTIVITY C2 ALTERNATE SHUTDOWN COOLING PATH UTILIZING RHR LOOP A REV.0-APRIL 1984 LSCS-UFSAR i§lii_¥isI!Iii::: S iii.0.0 iij.2 i C>llIi.tW:II JJG:JI3c/l j j lii_Iii_!l Ri'"¥li-S S--';-........C>LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-1 TRIP OF ONE RECIRCULATION PUMP MOTOR-101)%POWER (Typical)REV.13 LSCS-UFSAR

--l---.--t---+----4§

--+.---..I:-.----l,...........L..........:I.

c!'j i fi c LASALLE COUNTY STATION UPDATED FINAL SAFETY Ai'lALYSIS REPORT FIGURE 15.3-2 TRIP OF BOTH RECIRCULATION PUMP MOTORS-105%PO\VER REV.13 LSCS-UFSAR

.§.cl clS cl.cl., Q llIlUll Jg J.H]:)lG"II

=:r::::'--A-I--I+---l---.../i'L!!-+---l7H-+1--I----1ra';:.

LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-3 FAST CLOSURE OF ONE RECIRCULATION FLOW CONTROL VALVE AT 30°I SEC*105%POWER (Typical)REV.13 LSCS*UFSAR

--1----:J::"-+---+-----l§

--++--+-t---.illr--t----tJlig

  • M

...LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-4 FAST CLOSURE OF BOTH RECIRCUL.,\TION FLOW CONTROL VALVES AT 11*/SEC-105%Po\VER (Typical)REV.13 r,sCS-UFSAR

---i...i:*

.....

LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-5 SEIZURE OF ONE RECIRCULATION PUMP 105%POWER, 100%FLOW (Typical)REY.13 LSCS*UFSAR

.026--------------------.,\CONTROL ROD BEING PULLED,,I tI,I.,\-4-.SCRAM INSERTS,I" CONTROL ROD,,I,,I II, II , , o J<<:--.L---1..i..L._..L----JI..-..--1--.....1--J----l--:1-----J o 4-81216 20 24 40 TIME{SECONDS}.010.002 LASALLE COlJNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-1 (INITIAL CORE)POINT KINETICS CONTROL ROD REACTlv1TY INSERTION REV.13 60,..-----------------------.

50 40:J<<x<<)(30...J<<i3<<a::-PIA FROM ANALYSIS 10 3.0 2.0 CONTROL ROD WORTH 1.0 LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-2 PIA VS.ROD WORTH NEDO-10527 SUPPLEMENT 1(2)AND DETAILED ANALYSIS REV.0-APRIL 1984 LSCS*UFSAR

2.5%ROO WORTH-------2.001.ROO WORTH-----1.6%ROD WORTH" 1\I\/\/\I I V I'\I/\I/\I,//I I//I'I ,II//'/,////'/,////"," ,,'""-", ,..8.0 4.0 6.0 TIME (SECONDS)/(5 2'---1.."---"-__.......2.0 LASALLE COUNTY STATION UPDATED FINAL SAF'ETY ANALYSIS REPORT FIGURE 15.4-3 (INITIAL CORE)CONTINUOUS RWE IN THE STARTUP RANGE CORE AVERAGE POWER VS.TIME FOR 1.6%, 2.0%and 2.5%ROD WORTH'S (POINT MODEL KINETICS)REV.13 LSCS*UFSAR I02L__-.l.____+=-____---;: 2 3 4 9 ASSUMPTIONS:

1.1.6%cl.ROD 2_0-3 fps VELOCITY 3.IRM SCRAM FOR WORST BYPASS CONDITION 4.Po=10.2 OF RATED 5.1967 PROUUe"!'ULJ.\lE ECH SPEC SCRAM RATE 6.EXPOSURE=0_0 GWDn'LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4*4 (INITIAL CORE)CONTINUOUS CONTROL ROD WITHDRAWAL FROM HOT STARTUP REV.13

LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-5 INITIAL CYCLE ANALYSIS ON LIMITS ROD PATTERN FOR ROD (26,35)REV, 14, APRIL 2002 I LSCS*UFSAR 1.8 1.7 1.8 1.5 0;::<<a: a: w1.*1.115......<<...;:: it...!1.3'OWER:Ii:Ii a:.....1.2 1.0..a......<<a:.................

121-381...0!.-----...._.--2 1.1 1.02 I:;

BUNDLE 123-3$1<<a:..........""""...................

1.0"""'-..........---1.01..............

-ROO BLOCK LINE.'"""'"""-0..1.00 0:I***10 12 DIST ANeE CONTROL ROO WITHDRAWN 11.11 LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4*6 (INITIAL CORE)FRACTION OF RATED PO\VER AND MCPR VS.MCPR VS.DISTANCE ROD (26*35)WITHDRAWN REV.13 LSCS-UFSAR It.",..-(I/BUNDLE 121.381//.//'---..--...

18 13 17 i:!!.....0(" cr;z 2..0(II:...Z...'"..15 0(...:c a:<<IU Z:;:I;;).1.:!:I-.D 8UNOLE 12S-341'2 ROO 8LOCK LINE" 10**,,'-----""----""'---_I-I-L...-L.J o OI$TAHCe CONTROL ROD WITHDRAWN""" LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-7 (INITIAL CORE)MLHGR VS.DISTANCE ROD (26,35)WITHDRAWN REV.1.3 LSCS-UFSAR

.,0.....,: 1II'"..0:(z..<<1II-.....,:........., N......Z........<., U 0...5:...at Z 0;: 0:I 0 z 0 J: II:<<..IU 0:: O.QCZ..I r i 0".,z ,.;oz.."-Ii::u ,;.....'"'"?i l a::;i.....-'...!!!!a:: D I l M..,..,:!:!!!...LL__-l...L I:-__

o OJ oJ;(..""__nZM o roi":it M N...:li 8 z It;r;oJO<<o.It0::t....u Z i 0%.,;:;5.....u%....;:;It..I...It!!::l D......*.'..L..L L__--L..L 0 1YIJ.INI.-

M" LASALLE COUNTY STATION UPDATED FINALSAFE'l'YANALYSIS REPORT FIGURE 1,5.4-9 (INITIAL CORE)RBM RESPONSE TO CONTROL ROD MOTION, CHANNELB&D REV.13 17S.0 150.0 125.0 Cl...,-100.0<a:...Cl-Z..., 75.0 u*..., a.50.0 LSCS*UFSAR IDLE LOOP STARTUP POWER: 35.0%RATED FLOW: 47.0%RATED SNUME: 0012A1N UTRON FLUX2A VE SUR ACE HE AT FLU TEAHFL OW l)5C ORE IN ET FLC W (:0J.I\f/....--r""-I, V\"-r""'--25.0 0.0 G.O$.0 10.0 15.0 20.0 25.0 50.0 Plots from REDYV04 Output for the Simulation of an Abnormal Idle Recirculation Loop Start-Up Event: (35%Power, 47%Core Flow).LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE lOA-lOa ABNORMAL IDLE RECIRCULATION LOOP START*UP EVENT-35%POWER, 47%CORE FLOW REV.13 LSCS-UFSAR IDLE LOOP STARTUP PO\\'ER: 35.0%RATED FLOW: 47.0%RATED SNUMB: 0012A 10.0 25.0 20.0 15.0 10.0 S.D 1 LE VEL(1 CH-REF-SEP-S RT)21P bSITION(%)3 SP ED..1(%)4

'r:'nuI, 1'1'\5 DRI VE F OW2.(I I"--,,,,J......I r""..J.I w'" w...w , Tf....4-v';0.0 50.0 15.0 25**lZS.O 150.0 100.0-as.o 1.0 TIME (SECONDS)PLOTS from REDYV04 Output for the Simulation of an Abnormal Idle Recirculation Loop Start-up Event: (35%Power, 47%Core Flow).Rated drive flow=9917 Ibm/sec)LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-lOb ABNORl\-IAL IDLE RECIRCULATION LOOP ST ART.UP EVENT-35%POWER, 47%CORE FLOW REV.13 LSCS-liFSAR IDLE LOOP STARTUP POWER 35.0%RATED FLOW'47.0%RATED SNUB: 0012A 0 1 DPHE PRISS RI 9£(PSI I 2 CPRE IN T SUB (BTU/8M)3 DIFFUSE F"LOW 1 (;l),..U r U::it

(;.:)r5T RBINE STEAM ,LOW (7-'"'/.----,,, V¥.2-r...""...............

...17/*I 1"I"" V*j I I 0.0 75.0 25.0$0.0 I1S.150.0 125.0 100.0-2'5.0 0.0 5.0 10.0 15.0 20.0 2$.0 30.0 Plots from REDYV04 Output for the Simulation of an Abnormal Idle Recirculation Loop Start-up Event: (35%Power, 47%Core Flow).L<\SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-10c ABNOIThl-\L IDLE RECIRCULATION LOOP START-UP EVENT*35%POWER, 47%CORE FLOW REV.13 LSCS-UFSAR IDLE LOOP STARTUP POWER: 35,0%RATED FLOW: 47,0%RATED SNUB: 0012.'1.5 1 vbro RE ClIVI y20 VITYACTIV TYn.AI , 0 V\-, j5 r f\rA--0-")<oJ I"'\-,./--........

.."***1.L o.---.S ,..--OJ C...ae******0.0 5.0 U.O 15.0 20.0 25.0 50.0 Plots from REHYV04 Output for the Simulation of an Abnormal Idle Recirculation Loop Start-up Event: (35%Power, 47%Core Flow).LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE l5A-lOd ABNORM.-\L IDLE RECIRCULATION LOOP START-UP EVENT.35%POWER, 47%CORE FLOW REV 13 s§s_l!-i-2 Id.c:l.0ii i 8 0'7 IlWW JQ JIG:ll£UI iie--H+-+---r+--+---1.1ii LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-11 FAST OPENING OF ONE RECIRCULATION VALVE 30%/SEC AT 58%POWER, 35%FLOW (Typical)REV.13 LSCS-UFSAR Iii-viai!i G:1ii 5 tj....Iii I'JI!--tH---ir-i--t---....,rj.=

--+---="":--.1..--.":"....................'1 LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15-4-12 FAST OPENING OF BOTH RECIRCULATION FLOW CONTROL VALVES l1%/SEC AT 68%POWER 50%FLOW (Typical)

LSCS-UFSAR FIGURE 15.4-13 REV. 12 - MARCH 1998 (THIS FIGURE INTENTIONALLY LEFT BLANK) LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-13 CRITICAL ROD PATTERN AND FUEL BUNDLE EXCHANGE LOCATIONS FOR MISPLACED BUNDLE ACCIDENT 0.0 Gwd/t LSCS*UFSAR§ll!8 uei lQ-.!llllW Zi lH3:IGIl2-.;-:ltZ_N'"::!...:!l..-:!l...I..,..0 o Jl,., J j--........,v.1 I.,v.(CB1.I:Il Zi.l.tGJIG,j)

L<\SALLE COUNTY STATION CPDATED FINAL SAFETY A.;."iALYSIS REPORT FIGURE 155-1 INADVERTENT PUMP START OF HPCS PUMP (CE, Typical)REV.13 LSCSUFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-1 INITIAL CYCLE ANALYSIS SUPPRESSION POOL TEMPERATURE RESPONSE STUCK OPEN RELIEF VALVE FROM POWER OPERATION Rev.14, APRIL 2002 LSCS-UFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-2 INITIAL CYCLE ANALYSIS SUPPRESSION POOL TEMPERATURE RESPONSE STUCK OPEN RELIEF VALVE FROM HOT STANDBY Rev.14.APRIL 2002 ENVIRONS SECONDARY I I CONTAINMENT CONTAINMENT*

I SOTS LIKE STRUCTURE I INSTRUMENT INSTRUMENT LINE BREAK LINE BREAK POTENTIAL DESIGN CASE LA SALLE COU NT Y ST ATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3 LEAKAGE PATH FOR INSTRUMENT LINE BREAK REV.0-APRIL 1984

l...a: 11.1...e(w w....cr 11.1 C/)ZZ-11.1 o (,)11.1 Z:Jc(...>III 11.1>..J c(>:.:: (J 11.1:r: (J FIGURE 15.6-4 STEAM FLOW SCHEMATIC FOR STEAM BREAK OUTSIDE CONTAINMENT LA SALLECOU NTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT<<...z o (J..J 11.1 Z Z::::l...11.1(:1 ZZ;0__-+-1...........J II 14 11.1>a: o...11.1 cr REV.0-APRIL 1984
  • SECONDARY CONTAINMENT ENVIRONMENT
  • DRvWELL I SGTS I---1 t*RPV n j,"-BREAK PRIMARY{CONTAINMENT SUPPRESSION POOL-LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-5 LEAKAGE FLOW FOR LOCA REV.0-APRIL 1':Hi4 TURBINE BYPASS CHECK VALve FLOW_____PUMPSIHEATEAS/Ioo4!-p404 PUMPS IDEMINI CONTROLS CONTROLS CHECK VALVES LA SALLE COUNTY STATION UPDATED FINAL-SAFETY ANALYSIS REPORT FIGURE 15.6-6 LEAKAGE PATH FOR FEEDWATER LINE BREAK OUTSIDE CONTAINMENT REV.0-APRIL 1984 ENVIRONMENT SECONDARY CONT AINMENT--SGTS.t.I POOL RELEASE AREA LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.7-1 LEAKAGE PATH FOR FUEL HANDLING ACCIDENT REV.0-APRIL 1984 LSCS-UFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.8 I RESPONSE TO MSIV CLOSURE WITH FLUX SCRAM (102%uprated power, 105%core flow and 1035 psi a initial dome pressure)Source of information:

Reference 59 Rev.14, APRIL 2002 LSCS-UFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.82 TURBINE TRIP WITH BYPASS FAILURE (@100%Uprated Power,&105%Core Flow)Source of information:

Reference 59 Rev.14, APRIL 2002 LSCS UFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE I5.B-3 GENERATO R LOAD REJECTION WITH BYPASS FAILURE (@100%Uprated Power.&105'){, Core Flow)Source of information:

Reference 59 Rev.14, APRIL 2002 LSCS-UFSAR LASALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSTS REPORT FIGURE 15.B-4 FEEDWATER CONTROLER FAILURE-MAXIMUM DEMAND(@100%Uprated Power.&105%Core Flow&326.5 of Feedwater Temperature)

Source of information:

Reference 59 Rev.14, APRIL 2002

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