ML15114A102

Requests for Additional Information on the LaSalle County Station Units 1 and 2 (LSCS) Analysis of Severe Accident Mitigation Alternatives

ML15114A102
Person / Time
Site:	LaSalle
Issue date:	04/30/2015
From:	License Renewal Projects Branch 2
To:
Shared Package
ML15114A390	List: ML15114A102 ML15114A320
References
TAC MF5567, TAC MF5568
Download: ML15114A102 (7)

Text

ENCLOSURE Requests for Additional Information on the LaSalle County Station Units 1 and 2 (LSCS) Analysis of Severe Accident Mitigation Alternatives

1. Provide the following information regarding the Probabilistic Risk Assessment (PRA) used for the Severe Accident Mitigation Alternative (SAMA) analysis

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant.

As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's treatment of internal events and calculation of core damage frequency in the Level 1 PRA model.

The requested information is needed for the NRC staff to reach a conclusion on the sufficiency of the applicant's Level 1 PRA model for supporting the SAMA analysis. a. Section F.2 of the Environmental Report (

ER) indicates that the LSCS PRA is a Unit 2 model and is considered applicable to Unit 1 since the two units are nearly identical, unless otherwise noted. Provide a brief description of the differences between the units, particularly those differences that might impact internal flooding and their impacts on the unit risk.

b. Relative to the contributions to the internal events Core Damage Frequency (CDF)

i. Provide the contributions to internal events CDF from Station Blackout (SBO)

. ii. Table F.2

-3 indicates the CDF contribution due to Anticipated Transient Without Scram (ATWS) is 4.87 x 10-7/year; discuss the reasons why the CDF due to ATWS is such a relatively high contribution to the total internal events CDF.

iii. Clarify what is meant by the statement "all ATWS events are modeled as a turbine trip" in Section F.2.3.1 of the ER

. c. Identify the plant systems shared or that could be cross

-tied between units and describe the modeling in the PRA, including the treatment of unavailability during outages or events involving the other unit.

d. Section F.2.1 of the ER indicates that no significant plant modifications affecting the risk profile were performed since the PRA model freeze date. Other than the hard pipe vent, identify any planned changes or modifications to the LSCS design, operation, reactor power level, fuel cycle, or fuel design that might impact the SAMA analysis and describe their expected impacts.

e. Table F.2

-1 of the ER includes a number of footnote citations, but the content of the footnotes are not included. Provide the footnotes.

f. Section F.2.2.4 of the ER states that the 2000A model included a seismic PRA model (removed when the 2006A model was developed). Confirm that the seismic risk is not included in the results given in Table F.2

-1 of the ER. If included, provide the results without the seismic risk.

g. Provide a summary of the more significant changes incorporated in the 2006A PRA model.

Page 2 of 7 h. Changes incorporated into the 2011A model included assignment of mitigated ATWS scenarios from Accident Class IV to Accident Class II in Section F.2.2.11 of the ER. However, Table F.2

-3 of the ER does not appear to include any Class II ATWS scenarios. Provide more detail on the assignment of ATWS scenarios, including any impacts from the 2013A Level 2 model upgrade.

2. Provide the following information relative to the Level 2 analysis

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant.

As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's treatment o f accident propagation and radionuclide release in the Level 2 PRA model.

The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant's Level 2 PRA model for supporting the SAMA evaluation.

a. Section F.2.2.11 of the ER is titled "2013A Upgrade." Describe the changes in the Level 2 model that are considered upgrades as opposed to updates. Also, describe the steps taken to assure technical adequacy of the 2013A Level 2 model.

b. Provide a description of the interface between the Level 1 CDF analysis and the Level 2 analysis, including the use of plant damage states or core damage accident classes and how the two models are linked.

c. Provide additional information on the containment event trees (CETs) utilized in the Level 2 analysis including: the number of CETs, the sequences handled by each CET, the events included in each CET

, and how the loss of offsite power (LOOP) and SBO sequences are addressed.

d. Table F.2

-6 of the ER includes Accident Classes II, IIE, IIV and IIVE. Classes IIE and IIVE are not defined in Table F.2

-3. i. Define these accident classes.

ii. Table F.2-6 also includes a significant intact frequency for the Class II accident classes which is normally considered and defined in Table F.2-3 to include sequences where core damage occurs after containment failure.

Discuss the reasons for these results, how the intact frequency was determined, and the potential for impacting all the release category frequencies.

e. It is noted in Table F.2

-6 of the ER that the majority of Class IV (ATWS) sequences result in a low/early (L/E) release and, as indicated in Table F.3

-17 for the majority of these sequences, core melt is arrested in the reactor vessel. Discuss the reasons and basis for these results.

f. In Table F.3

-17 of the ER, the reference Modular Accident Analysis Program (MAAP) case for Sequence IV

-04, which is the dominant contributor to the low/early (L/E) release category frequency, is stated not to have been used for the release fractions because this MAAP case does not model the core melt arresting in the reactor vessel. Another MAAP case, with a release fraction 1/65th of that for the ATWS case, was used to determine the release fractions for the L/E release category. Provide additional support for the use of these release fractions.

Page 3 of 7 g. The run time for several of the LSCS MAAP cases in Table F.3

-18 of the ER is less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> past the declaration of a general emergency when effective offsite resources might be available for mitigating the accident. While it is stated that the representative MAAP cases were run until plateaus of the cesium iodide (CsI) and cesium hydroxide (CsOH) release fractions were achieved, provide more information to justify the adequacy of the MAAP release fraction results for the SAMA analysis for those cases with run times less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the general emergency.

h. In Table F.3

-18 of the ER, the time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for declaration of a general emergency for the moderate/intermediate (M/I) release category is indicated to have been determined probabilistically. Describe how this time was determined, how it was utilized in the Level 2 and Level 3 analysis and the impact on the SAMA cost

-benefit analysis if the minimum steam coolant water level limit (MSCWLL) time of 27.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is used in the analysis.

3. Provide the following information with regard to the treatment and inclusion of external events in the SAMA analysis

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's treatment of external events in the Level 1 PRA model.

The requested information is needed in order for the NRC staff to reach a conclusion on the sufficiency of the applicant's Level 1 PRA model for supporting the SAMA evaluation.

a. Section F.5.1.6.2.1 indicates that the impact of using the LSCS 2013 seismic hazard curves on the Risk Methods Integration and Evaluation Program (RMIEP) analysis has been investigated. Provide more information on the source of these hazard curves including a characterization of how they compare with the LSCS hazard curves included in the industry's response to the NRC's Fukushima Daiichi Nuclear Power Plant Near Term Task Force (NTTF) Recommendation 2.1. Also, provide the LSCS seismic CDF using the updated hazard curves and the simplified methodology of Generic Issue (GI)-199. 4. Provide the following information relative to the Level 3 PRA analysis

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant.

As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's analysis of consequences in the Level 3 PRA model. The requested information is needed in order for the NRC staff to reach a conclusion on the acceptability of the applicant's Level 3 PRA model for supporting the SAMA evaluation

. a. Indicate the Sector Population and Economic Estimator (SECPOP2000) version used in the analysis. Confirm that manual entry of economic values was sufficient to prevent known SECPOP2000 errors from influencing the SAMA results. The known errors are described in NRC Request 4.b on page E1

-24 of the response letter to NRC from another licensee (ADAMS Accession No. ML102100588

).

Page 4 of 7 b. Compared to higher (total) estimates in Table F

.3-2 of the ER for an assumed uniformly distributed population, provide supporting justification that 2010 population estimates in Table F.3-5 are more representative of the actual 2010 population distribution.

c. Indicate the source of the county population growth rates in Table F.3

-1 of the ER

. d. Compare the more recent 2012 Census of Agriculture data to values used in the analysis from the 2007 Census of Agriculture data with any consumer price index updates. e. Section F.3

-7 of the ER mentions greater plume washout and radionuclide deposition during precipitation events. Estimate the sensitivity of the offsite population dose and

offsite economic cost risk to atmospheric conditions with and without precipitation events. f. Specify the software codes and versions used for calculating the core inventory.

g. NEI 05-01 lists several radionuclides to be considered in the core inventory for th e SAMA analysis. Describe how the core inventories for Co

-58 and Co

-60 were calculated and justify the appropriateness of inherent assumptions to LSCS Units 1 and 2.

h. Section F.2 of the ER indicates the PRA used in the SAMA analysis is a model for Unit 2, unless otherwise noted. Section F.3.5 indicates the plant

-specific calculation of core inventory represents bounding isotopic values. Clarify if the calculation of radionuclides shown in Table F.3-11 of the ER relates to LSCS Unit 1, Unit 2, or both units. Justify why the calculation is bounding for either unit.

5. Provide the following information with regard to the selection and screening of Phase I SAMA candidates

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's basis for the selection and screening Phase I SAMA candidates

. The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant's Phase I SAMA selection and screening process for the SAMA evaluation

. a. Relative to the identification of SAMAs from the Level 1 importance review documented in Table F.5

-1 of the ER

i. The turbine trip with bypass initiating event (%TT) is indicated to have a frequency of 0.8 per year, which implies that a number of turbine trips with bypass have occurred over the LSCS operating history. Do any of these occurrences suggest possible cost-beneficial SAMAs?

ii. The discussion of potential SAMAs for a significant number of events (RCVCL

-1BE, %TIA, RCVSEQ

-DLOP-041, BWTOPWTHXSTBYH

--, and others) involve water hammer induced loss of coolant accidents (LOCAs) related to the generation of a LOCA signal on high drywell pressures when an actual LOCA does not exist and are mitigated by SAMA 7. Provide more detail on the water hammer scenarios in general and more specifically those involving the above identified events including: how loss of instrument air initiated events involve water hammer induced LOCAs, Page 5 of 7 why alignment of the standby turbine building closed cooling water (TBCCW) heat exchanger train is needed, why water hammer LOCAs are not important for other non-LOCA transients, and the potential for preventing water hammer by preventing residual heat removal (RHR) pump start when water hammer conditions exist or preventing draining of the RHR line. Also, clarify the SAMA 7 statement in Table F.5-4 of the ER that the LOOP

-delayed LOCA scenario is not specifically modeled in the PRA.

iii. For event 2ADRX

-TRANS--H-- (p. F-244 of the ER), failure to manually depressurize the reactor pressure vessel, the action is indicated to be required for about 90 percent of the scenarios because failure to initiate suppression pool cooling (SPC) results in containment failure. Discuss these scenarios in more detail including the reason for this high percentage compared to other causes of failure of high-pressure injection.

iv. Regarding event 2VYFNNWVY01

--X-- (p. F-271 of the ER), VY NW corner room (RHR A) cooling fan 2VY01C fails to run, previous LSCS evaluations could not demonstrate that portable fans would provide adequate cooling for the reactor building corner rooms when the normal cooling system failed. Discuss this previous evaluation and if sufficient cooling can be achieved if the portable fans can provide air movement through the room coolers assuming cooling water remains available.

b. Relative to the identification of SAMA s from the Level 2 importance review documented in Table F.5

-2 of the ER

i. For event 2TDOP

-RECLPS2H-- (p. F-291 of the ER),

operator fails to recover low-pressure systems, the discussion indicates that low-pressure systems would not have the power to function. Consider the potential for utilizing a fire truck or other portable self

-powered pumps for injection into the containment.

ii. Event 2GVPHCMBSTGASF

-- (pp. F-286, F-300, and F

-316), Combustible ga s venting fails, appears in all three Level 2 importance lists. Discuss this event and if any SAMAs are possible to directly mitigate this failure.

c. The discussion in Section F.5.1.3.6 of the ER on Grand Gulf SAMA 59 states that "Rather than cycling large pumps in scenarios where the cooling system is lost, a more effective means of maintaining injection with the Emergency Core Cooling System (ECCS) pumps is considered to be through the use of portable/temporary cooling alignment, which is addressed in the LSCS importance list review by SAMA 16.

" Consider a SAMA similar to Grand Gulf's SAMA 59, Increase operator training for alternating operation of the low pressure emergency core cooling system pumps (low-pressure coolant injection and low pressure core spray) for loss of standby service water scenarios, for rooms where the use of portable fans may not be effective.

d. While the Commonwealth Edison Individual Plant Examination (IPE) submittal did not identify any vulnerability, as discussed in the NRC's IPE Safety Evaluation Report, the IPE did cite NUREG/CR4832, Vol. 3, Part 1

, for insights about potential improvements. These are: (

i) Eliminate the sneak circuit in the reactor core isolation cooling (RCIC) isolation logic that results in the RCIC steam line inboard isolation valve closing when offsite alternating current (AC) power is lost and the appropriate diesel generator starts

(ii) Change the RCIC room temperature isolation logic so that, in cases where AC power Page 6 of 7 from train A has failed but AC power from train B is available, this isolation logic does not isolate if no other emergency core cooling system is working

and (iii) Change the venting procedure so that venting does not result in severe environments in the reactor building. Discuss the current status of these potential improvements.

e. The RMIEP Summary (NUREG/CR

-4832, Volume 1) identifies the common cause failures of the core standby cooling system (CSCS) cooling water pumps as the dominant events in the seismic risk reduction importance assessment. Discuss these events and their importance in the seismic analysis used for SAMA identification.

f. Relative to the identification of SAMAs to mitigate internal fires:

i. As discussed in Section F.5.1.6.1 of the ER, the lower cutoff value of 5 percent of the total fire CDF is used for considering SAMAs. Using this cutoff value results in potentially not identifying cost

-beneficial SAMAs below a cost of $142,000.

Consider additional fire sequences below this threshold in order to identify potential low cost procedure enhancement SAMAs.

ii. As indicated in Section F.5.1.6.1.1 of the ER, fires in the Division 1 and Division 2 essential switchgear rooms for each unit make up 60 percent of the total fire CDF. The only SAMA proposed for mitigating these fires is SAMA 1, installing a reliable hard pipe vent. SAMA 1 only mitigates the adverse consequences of venting and does not mitigate the direct impact of the fire. Discuss these fire scenarios and the potential for other SAMAs to directly mitigate the fire at an earlier stage of the scenario or by mitigating events in the cutsets other than the adverse conditions due to venting

. iii. Section F.5.1.6.1.3 of the ER indicates that the largest contributing fire scenario to the Auxiliary Electrical Equipment Room fire risk is a bounding cable fire caused by hot work. Discuss the potential for a SAMA to mitigate this risk.

iv. The RMIEP Summary (NUREG/CR

-4832, Volume 1) identifies the three most important events in the fire analysis risk reduction importance assessment to be in the control room abandonment scenario. Discuss this scenario and its importance in the current fire analysis and the potential for a cost

-beneficial SAMA to mitigate this scenario.

6. Provide the following information with regard to the Phase II co st-benefit evaluations

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant.

As part of its review of the LSCS SAMA analysis, NRC staff evaluates the applicant's cost-benefit analysis of Phase II SAMAs

. The requested information is needed in order for the NRC staff to reach a conclusion on the acceptability of the applicant's cost estimations for individual SAMAs and cost-benefit evaluation

. a. Clarify if maintenance costs are included in the estimated implementation costs.

b. Discussions of SAMA 8 in Section F.6.8 of the ER and SAMA 14 in Section F.6.13 of the ER include the statement "Flow from the fire protection system, in its current configuration, is only adequate in cases where RCIC....

" Discuss what is meant by Page 7 of 7 "current configuration

" and the potential for a SAMA to address the fire protection system configuration and make its use possible without prior RCIC operation.

c. Discussion of SAMA 14 (

Provide a portable DC source to support RCIC and SRV operation) in Section F.6.13 of the ER indicates that the PRA model was changed to include a lumped event to represent the 480V AC power source that feeds the Division 1 battery chargers. The SAMA description and the basic events cited to be mitigated by this SAMA indicated that a direct current (DC) power source to directly supply an ESF DC distribution panel is needed. Discuss the inclusion of an AC power supply in the model. d. SAMA 7 (Water hammer prevention) and SAMA 25 (Periodic training on water hammer scenarios resulting from a false LOCA signal) both are intended to mitigate the water hammer scenarios involving SPC operation interrupted by a LOOP. The changes made to the model and the impact of the changes on person

-rem and offsite economic cost risk are significantly different for the two SAMAs. Discuss these SAMAs and their analyses in more detail to justify these differences.

e. A few SAMAs (e.g.

, 8, 14, and 27) involve use of equipment that may be available because of the B.5.b Program. Discuss this further and the impact on the cost

-benefit analysis.

f. SAMA 9 (Develop flood zone specific procedures) and SAMA 11 (Provide the capability to trip the FPS pumps) both address internal flooding whose principal contributor is a fire protection system (FPS) pipe rupture in the reactor building. The assumption for evaluating SAMA 9, that the risk from all internal floods will be eliminated, results in a 9 percent reduction in CDF and this SAMA being cost

-beneficial. The assessment of SAMA 14 results in less than a 2 percent reduction in CDF; while as stated in Section F.6.11, SAMA 11 is designed to eliminate the FPS's flow. Discuss the FPS design and the FPS pipe break scenarios and associated modeling to support the above results. 7. For certain SAMAs considered in the LSCS Environmental Report, there may be lower cost or more effective alternatives that could achieve much of the risk reduction.

In this regard, provide an evaluation of the following SAMAs

. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the LSCS SAMA analysis, NRC staff considers additional SAMAs that may be more effective or have lower implementation costs than the other SAMAs evaluated by the applicant

. The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant's determination of cost-beneficial SAMAs. a. Consider the potential for changes to the suppression pool cooling operating procedures/practices to reduce the chance of the 2 psi high drywell signal being reached for normal transients as an alternative to SAMA 7.