ML092330691

Ria Related to Alternative Source Term License Amendment Request

ML092330691
Person / Time
Site:	LaSalle
Issue date:	08/27/2009
From:	Goodwin C Plant Licensing Branch III
To:	Pardee C Exelon Nuclear
Goodwin, Cameron S, NRR/DORL, 415-3719
References
TAC ME0068, TAC ME0069
Download: ML092330691 (6)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 27, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville,IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO ALTERNATIVE SOURCE TERM LICENSE AMENDMENT REQUEST (TAC NOS. ME0068 - ME0069)

Dear Mr. Pardee:

By letter to the Nuclear Regulatory Commission (NRC) dated October 23,2008 (Agencywide Documents Access and Management System Accession No. ML083100149), Exelon Generation Company, LLC, submitted a request to replace the current accident source term used in design-basis radiological analysis with an alternative source term pursuant to Title 10 of the Code of Federal Regulations Part 50.67, "Accident Source Term," for the LaSalle County Station, Units 1 and 2.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on August 17, 2009, it was agreed that you would provide a response 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3719.

Sincerely,

~~{0(J ~

carleron S. Goodwin, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosure:

Request for Additional Information cc w/encl: Distribution via ListServ

REQUEST FOR ADDITIONAL INFORMATION LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374 Containment and Ventilation Branch Question

Background:

The LaSalle County Station (LSCS) standby liquid control (SLC) system activation steps are in a safety-related plant procedure (Emergency Operating Procedure (EOP) LGA-001, "RPV Control"). LSCS states that they will revise LGA-001 to ensure that the SLC system starts injection from the boron solution storage tank during a design-basis accident (DBA) loss-of coolant accident (LOCA). In addition, LSCS intends to revise LGA-001 to ensure no steps would terminate the injection during a DBA LOCA prior to emptying the SLC storage tank (i.e., injection of the full content into the reactor pressure vessel (RPV)). The technical specification Bases state that the borated solution [sodium pentaborate] is discharged near the bottom of the core shroud.

Describe the emergency core cooling system (ECCS) flow paths that will be in use in LGA-001 when the SLC system is activated.

Describe the SLC flow path from injection into the RPV, through the RPV, and from the RPV to the suppression pool in conjunction with the ECCS flow path(s) in use per LGA-001.

Please describe how the EOPs will be changed to assure that adequate sodium pentaborate reaches the suppression pool to maintain a pH of at least 7.

Describe how LSCS will verify the changes to the EOPs assure completion of the injection of sodium pentaborate to the suppression pool within the required time limit.

The proposed license revision indicates that the LSCS SLC system cannot be considered redundant with respect to its active components. The submittal indentified the inboard and outboard check valves as non-redundant components. Exelon Generation Company has Enclosure

-2 determined that the 1(2) C41-F006 and 1(2) C41 -FOO? check valves have an acceptable performance history at LSCS.

Please describe how the inboard and outboard check valves are protected from foreign material when maintenance is performed on the SLC flow path.

Mechanical and Civil Engineering Branch

RAI-1

Page 1.183-8 of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," (Section 1.3.2, Re-Analysis Guidance) denotes the responsibility of licensees to evaluate the radiological and nonradiological impacts of an alternative source term (AST) implementation and the impacts of any associated plant modifications. Section 3.6 of your October 23, 2008, submittal, indicates that credit will be taken for portions of the Control Room (CR) and Auxiliary Electric Equipment Room Heating Ventilation and Air Conditioning (HVAC) systems in support of the proposed AST implementation at LSCS. Please discuss the methodologies used to demonstrate the seismic ruggedness and/or the seismic qualification of the aforementioned HVAC systems and their components. Additionally, please provide the references which provide the regulatory acceptance bases of these methodologies.

RAI-2

Please verify whether the secondary containment bypass piping is credited for the proposed AST. If affirmative, please confirm whether these piping runs have been seismically evaluated and discuss the methodologies used in these evaluations.

Please provide a list of any additional mechanical equipment or systems (i.e. air handling units, motor control centers), fans and I&C cabinets, etc.) not mentioned above which will be credited as part of the implementation of the AST at LSCS. For these items:

a) Indicate whether the equipment is new or existing.

b) Describe the location of the equipment and the seismic qualification method employed to demonstrate the seismic ruggedness of this equipment, such as the plant licensing basis or a Nuclear Regulatory Commission (NRC)-endorsed industry standard.

c) Summarize the results of the seismic qualification of the equipment, indicating whether any modifications or re-design will be necessary in support of the AST.

Accident Dose Branch Please explain how the LaSalle 1998 through 2003, hourly meteorological data provided in support of the October 23, 2008, AST license amendment request (LAR), in general, was

-3 processed from the raw measurements and discuss the LaSalle site meteorological characteristics. The following are some NRC staff observations and concerns with respect to the reported atmospheric stability and wind speed measurements.

• There appears to be a relatively low frequency of reported unstable conditions. This is particularly noticeable when compared with historic data in the LaSalle Updated Final Safety Analysis Report (UFSAR). In addition, extremely stable conditions (G) were reported almost 12 percent of the time between 1998 and 2003 for measurements between the 61.0 and 10.1 meter levels.

Information Atmospheric Stability Frequency Occurrence (%)

Source Data Period A B C D E F G Total Delta-T Interval LaSalle UFSAR 10/1976 3.5 3.6 4.7 45.8 24.3 14.1 4.2 100 09/1978 114.3 -10.1 m NRC estimate 1998 - 2003 0.1 0.4 1.6 48.4 29.2 14.8 5.5 100 114.3-10.1m NRC estimate 1998 - 2003 1.9 3.4 6.4 37.2 26.2 13.3 11.7 100 61.0-10.1m

• LaSalle appears to be a relatively high wind speed site for all stability categories. In particular, staff estimated average wind speed during G stability at about 7.5 miles per hour (mph) and maximum yearly values ranging from about 16 mph to almost 30 mph at the 10.1 meter level. Such wind speeds seem too high to support G stability conditions.

Reported wind speeds also seemed generally high for other stable categories.

• Wind speed and atmospheric stability data for 1995 through 1998, from the Dresden and Quad Cities sites in eastern and western northern Illinois, respectively, seem more similar to each other than the 1998 through 2003, meteorological data from LaSalle which is located between these two sites in central northern Illinois.

RAI-2

With regard to page H1 of Calculation No. L-003063, "Alternative Source Term Onsite and Offsite X/Q Values," Rev. No.1, Attachment H, which direction is true north? What is the scale of the figure? Where are the reactor building hatch access to auxiliary building roof and reactor building truck bay door release locations? How was the release point shown for the integrated leak rate test location determined to be appropriate for use in calculating the atmospheric dispersion factors (X/Q values)? Where is the assumed location for releases from the reactor building wall? Where is the control room located and why is it appropriate to assume that use of the control room air intake X/Q values as the input to the dose assessments is appropriate for unfiltered inleakage into the control room?

-4 Section 3.3.2.2 of Attachment 1, "Evaluation of Proposed Change," to the LAR states that the stack was modeled as a ground-level release to obtain X/a values to be utilized for the fuel handling accident (FHA) only. Why was this calculation made? Why was a diffuse release from secondary containment assumed to be appropriate for the FHA? Section 3.2.4.1 of RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," states that diffuse source modeling should be used only for those situations in which the activity being released is homogeneously distributed throughout the building, and when the assumed release rate from the building surface would be reasonably constant over the surface of the building. Please confirm that no other source/receptor pair was more limiting.

RAI-4

For each DBA... which sourcelreceptor pairs were compared to determine the limiting control room X/a values? Which X/a values were input into each of the limiting dose assessments?

RAI-5

Regarding the DBAs analyzed in support of this LAR, please confirm that the generated X/a values model the limiting doses and all potential release scenarios were considered, including those due to loss of offsite power or other single failures.

Why does the joint frequency distribution for ground level releases to the exclusion area boundary and outer boundary of the low population zone use stability classifications based upon measurements between the 114.3 to 10.1-meter intervals rather than between the 61.0 and 10.1 meter levels? RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Section C.1.1, states that atmospheric stability should be determined by temperature difference between the release height and the 1O-meter level.

Section 4.3 of Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant Accident," to RG 1.183 states that ambient temperatures used in assessments on the ability of the secondary containment to maintain a negative pressure should be the 1-hour average value that is exceeded only 5 percent or 95 percent of the total numbers of hours in the data set, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded only 5 percent). What is the limiting temperature for the LaSalle site?

Mr. Charles G. Pardee August 27,2009 President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUB..IECT: LASALLE COUNTY STATION, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO ALTERNATIVE SOURCE TERM LICENSE AMENDMENT REQUEST (TAC NOS. ME0068 - ME0069)

Dear Mr. Pardee:

By letter to the Nuclear Regulatory Commission (NRC) dated October 23, 2008 (Agencywide Documents Access and Management System Accession No. ML083100149), Exelon Generation Company, LLC, submitted a request to replace the current accident source term used in design-basis radiological analysis with an alternative source term pursuant to Title 10 of the Code of Federal Regulations Part 50.67, "Accident Source Term," for the LaSalle County Station, Units 1 and 2..

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on August 17, 2009, it was agreed that you would provide a response 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3719.

Sincerely,

/RA by M. David for C. Goodwin/

Cameron S. Goodwin, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosure:

Request for Additional Information cc w/encl: Distribution via ListServ DISTRIBUTION:

PUBLIC LPL3-2 R1F RidsNrrDorlLpl3-2 Resource RidsNrrPMLaSalle Resource RidsNrrLATHarris Resource RidsNrrDoriDpr Resource RidsOgcRp Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource ADAMS Accession No ML092330691 NRR-106 OFFICE LPL3-2/PM LPL3-2/LA LPL3-2/BC NAME MDavid for THarris SCampbell CGoodwin DATE 8/27/09 8/25/09 8/27/09 OFFICIAL RECORD COpy