ML080850183

Draft Request for Additional Information

ML080850183
Person / Time
Site:	Oyster Creek
Issue date:	03/25/2008
From:	Geoffrey Miller NRC/NRR/ADRO/DORL/LPLI-2
To:	Chernoff H NRC/NRR/ADRO/DORL/LPLI-2
Miller G, NRR/DORL, 415-2481
References
TAC MD7261
Download: ML080850183 (6)

Text

March 25, 2008 MEMORANDUM TO: Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM: G. Edward Miller, Project Manager /ra/

Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MD7261)

The attached draft request for information (RAI) was transmitted on March 25, 2008, to Mr. David Robillard of AmerGen Energy Company, LLC (AmerGen). This information was transmitted to facilitate an upcoming conference call in order to clarify the licensee=s amendment request for the Oyster Creek Nuclear Generating Station (Oyster Creek), dated November 2, 2007.

The proposed amendment would revise the Oyster Creek Technical Specifications regarding secondary containment operability requirements during refueling.

The draft questions were sent to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed.

Additionally, review of the draft RAI would allow AmerGen to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensees request.

Docket No. 50-219

Enclosure:

Draft RAI

March 25, 2008 MEMORANDUM TO: Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM: G. Edward Miller, Project Manager /ra/

Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MD7261)

The attached draft request for information (RAI) was transmitted on March 25, 2008, to Mr. David Robillard of AmerGen Energy Company, LLC (AmerGen). This information was transmitted to facilitate an upcoming conference call in order to clarify the licensee=s amendment request for the Oyster Creek Nuclear Generating Station (Oyster Creek), dated November 2, 2007.

The proposed amendment would revise the Oyster Creek Technical Specifications regarding secondary containment operability requirements during refueling.

The draft questions were sent to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed.

Additionally, review of the draft RAI would allow AmerGen to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensees request.

Docket No. 50-219

Attachment:

Draft RAI DISTRIBUTION PUBLIC RidsNrrDorlLpl1-2 PDI-2 Reading RidsNrrPMGMiller RidsNrrDorlDpr

• Memorandum dated 3/12/2008, no significant changes ACCESSION NO.: ML080850183 OFFICE LPI-2/PM DRA/AADB/BC NAME G. E. Miller R. Taylor*

DATE 3/25/2008 3/12/2008 OFFICIAL RECORD COPY

DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT SECONDARY CONTAINMENT OPERABILITY REQUIREMENTS DURING REFUELING OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 By letter dated November 2, 2007, AmerGen Energy Company, LLC (Amergen) submitted an amendment request for the Oyster Creek Nuclear Generating Station (Oyster Creek). The proposed amendment would revise the Oyster Creek Technical Specifications regarding secondary containment operability requirements during refueling.

The Nuclear Regulatory Commission staff has reviewed the information provided in support of the proposed amendment and finds that the following information is required to complete its review:

INTEGRITY OF FACILITY DESIGN BASIS

1. For the Oyster Creek revised Fuel Handling Design Basis Accident (DBA) source term analysis done in accordance with 10 CFR 50.67, Alternative Source Term (AST) methodology, provide the current licensing basis (CLB) parameters along with the revised values where applicable, as listed in Table 4-3 of the November 2, 2007 submittal. In addition, provide the basis for any changes to the CLB parameters as a result of the proposed AST application for the fuel handling accident (FHA). The NRC staff requests that the licensee include CLB information and the revised parameters in a table whether or not the individual parameter changed for this amendment request (See regulatory positions 1.3.2 and 1.3.4 of Regulatory Guide (RG) 1.183 and NRC Regulatory Issues Summary (RIS) 2006-04).

ACCIDENT SOURCE TERM

2. In the FHA AST assumptions Amergen stated that the General Electric (GE)-11 fuel burnup is 27.6 giga-watt-days (GWD)/ Metric Ton Uranium (MTU) so that the 6.3 kilo-watt (kW) / foot (ft) restriction outlined in footnote 11 of RG 1.183 does not apply.

Please clarify that the GE 11 9X9 fuel, in the course of its projected power history for any specific fuel load, will not exceed the value of 54 GWD/MTU (See Regulatory Position 3.2 and footnote 11 of RG 1.183).

3. Given the information above to be correct, does Oyster Creek have other fuel types that would exceed 54 GWD/MTU and be more limiting for a FHA at Oyster Creek as far as the maximum fuel inventory at the end of fuel life (See Regulatory Position 3.2 and footnote 11 of RG 1.183).

ENCLOSURE

DOSE CALCULATIONAL METHODOLOGY

4. Please clarify the term Reactor Well discussed in your submittal in Section 4.3 under the heading Decontamination Factor.

5. In Section 4.3 of the submittal, Amergen determined that, A drop over the reactor well is more limiting than accidents in the spent fuel pool (SFP). Provide the reference or detail that shows the drop in the well or as described in the Oyster Creek final safety analysis report (FSAR), a drop onto the reactor core from the maximum height allowed by the refueling equipment, as the limiting accident (See Regulatory Position 2.4 of RG 1.183).

6. Appendix B of RG 1.183 allows an overall decontamination factor of 200 if the depth of water above damaged fuel is 23 feet or greater. If the depth of the water is less than 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1) of RG 1.183. Provide either a detailed analysis that proves the statement in Section 4.3 that a shorter drop in the fuel pool would result in less radiological release to the containment or conservatively adjust the decontamination factor used for the FHA to account for water depth less than 23 feet (See Regulatory Position 2.4 of RG 1.183 and Regulatory Position 2 of Appendix B of RG 1.183).

ANALYSIS ASSUMPTIONS AND METHODOLOGY

7. In Table 4-3 of the submittal, the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) limiting dispersion factors are listed as Normal RB exhaust stack with limiting atmospheric dispersion factor (/Q values) of 1.41E-03 second per cubic meter (sec/m3 ) for the 414 m EAB and 1.35E-4 sec/m3 for the 3218 m LPZ. The reference to stack release and the dispersion values are in conflict with a worst-case ground level release as well as the values listed in Attachment 1 (Calculation C-1302-822-E310-082) of the submittal Table 6.1.3 which shows ground level release EAB

/Q value of 1.10E-3 sec/m3 and LPZ /Q value of 5.60E-5 sec/m3. Please clarify the above apparent discrepancies in your submittal METEOROLOGY ASSUMPTIONS

8. In Table 4-3, on page 15 of Enclosure 1 to the submittal dated November 02, 2007, the licensee states:

The most limiting 0-2 hr potential release point was determined to be the MAC facility entrance.

However, the table later lists the most limiting 0-2 hour (i.e., highest /Q value) potential release point for the FHA was as the Monitor and Control (MAC) Facility Personnel Airlock source location with a value of 6.75E-03 sec/m3. The 0-2 hour /Q value for the MAC Facility Entrance source location is 6.62E-03 sec/m3. Please clarify this statement and confirm which 0-2 hour value was used in calculating the control room dose estimate for the Oyster Creek Nuclear Generating Station.

9. Two control room Heating, Ventilating, and Air-Conditioning System intakes (Intake A and Intake B) were used as the respective receptor points for each source location for the onsite /Q analyses. ARCON96 used to perform these calculations of estimated atmospheric dispersion values. The guidance for ARCON96, RG 1.194, notes that certain considerations should be evaluated in identifying the control room outside air intakes for which /Q values should be considered.

Please clarify if the two intakes to the control room used to model the onsite /Q values are: (1) in the same wind direction window or not and (2) share an imbalance of the intake flow rate to the control room. Based on these clarifications, please identify which ARCON96 /Q equation was used for each source/receptor pair analyzed in this submittal and confirm that this was the appropriate methodology for the relative calculation. For example, if the Oyster Creek control room has dual intakes (e.g.,

Intake A and Intake B) with equal intake flow rates and share a wind direction window, then equation (5a) is deemed appropriate for this particular onsite /Q analysis.

10. ARCON96 was uesd to compute the onsite /Q values, which is supported by the guidance of RG 1.194. On page 14 of this regulatory guide (i.e., 1.194-14), it states that:

If the distance to the receptor is less than about 10 meters, the ARCON96 code and the procedures in Regulatory Position 4 should not be used to assess /Q values.

Considering this statement, please justify the use of the ARCON96 computer code to estimate the /Q value for the Reactor Building Wall source location to the Control Room Intake A receptor point. The horizontal distance for this particular source/receptor pair is noted as 7.9 meters (25.8 feet), which is considerably less than 10 meters (as specified appropriate for use of the ARCON96 computer code in RG 1.194). Please indicate what impact this may have on the resulting /Q values if used inappropriately.

11. Numerous assumptions were made relative to the building wake area and distance calculations for source/receptor pairs used in the assessment of control room /Q values at the Oyster Creek Nuclear Generating Station. In order to justify these assumptions, a more graphically detailed schematic is needed than the plant layout provided in Attachment A to the amendment request.

Please provide a legible and suitably scaled schematic of the Oyster Creek Nuclear Generating Station site area showing true north. The drawing should indicate site boundaries and the location and orientation of principal plant structures within the site area representative of those used in the FHA most limiting /Q analyses as presented in Table 4-3 of the submittal. Highlight postulated release and receptor points or locations and include relative straight-line or taut string distances, as deemed appropriate, on the drawing

ADDITIONAL INFORMATION

12. Provide the Oyster Creek proposed FSAR Markup, as suggested by Regulatory Position 1.5 of RG 1.183, or revised updated final safety analysis report pages outlining the Oyster Creek revised AST licensing basis for the FHA. Regulatory Position 1.6 of RG 1.183 outlines the FSAR update requirements including a reference to 10 CFR 50.71.