ML041270560
| ML041270560 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/03/2004 |
| From: | Sean Peters NRC/NRR/DLPM/LPD2 |
| To: | Stinson L Southern Nuclear Operating Co |
| Peters S, NRR/DLPM, 415-1842 | |
| References | |
| TAC MC0625, TAC MC0626 | |
| Download: ML041270560 (10) | |
Text
May 3, 2004 Mr. L. M. Stinson Vice President - Farley Project Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RE: REQUEST TO REVISE CONTAINMENT EQUIPMENT HATCH TECHNICAL SPECIFICATIONS (TAC NOS. MC0625 AND MC0626)
Dear Mr. Stinson:
By letter dated August 29, 2003, Southern Nuclear Operating Company requested amendments to the Technical Specifications for Farley Nuclear Plant, Units 1 and 2, that would allow the equipment hatch to be open during core alterations and during movement of irradiated fuel assemblies within containment. The U.S. Nuclear Regulatory Commission staff has reviewed the application and has determined that additional information is required, as identified in the Enclosure.
We discussed these issues with your staff on April 26, 2004. Your staff indicated that you would attempt to provide your response by May 10, 2004.
Please contact me at (301) 415-1842, if you have any other questions on these issues.
Sincerely,
/RA/
Sean E. Peters, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
Request for Additional Information cc w/encl: See next page
ML041270560 OFFICE PDII-1/PM PDII-1/LA (A)
PDII-1/SC(A)
NAME SPeters DClarke SCoffin DATE 05/3/04 05/03/04 05/3/04
REQUEST FOR ADDITIONAL INFORMATION SOUTHERN NUCLEAR OPERATING COMPANY, INC.
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364 The U. S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensees submittal dated August 29, 2003, regarding proposed changes to the Technical Specifications (TS) for the Containment Equipment Hatch. The NRC staff has identified the following information that is needed to enable the continuation of its review.
- 1.
Please provide the design-bases parameters, assumptions, and methodologies (other than those provided in the August 29, 2003, submittal) that were changed in the radiological design-basis accident analyses as a result of the proposed TS change.
Also, please provide justification for the changes. If there are many changes it would be helpful to compare and contrast them in a table.
2.
Based upon a preliminary review of the fuel handling accident for the proposed TS change, the reviewer is unable to match the calculated doses. Please provide the calculation for the design bases fuel handling accident.
3.
What types of hoses and cables will be allowed to pass through the open equipment hatch? What provisions will be made for the designated individual to separate these to close the air lock door while reducing hazards from these hoses and cables?
4.
Appendix B to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 establishes quality assurance requirements for the design, construction, and operation of those structures systems or components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public.
Appendix B, Criterion III, Design Control, requires that design control measures be provided for verifying or checking the adequacy of design. Appendix B, Criterion XVI, Corrective Action, requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, defective material and equipment, and nonconformances are promptly identified and corrected. Generic Letter (GL) 2003-01, Control Room Habitability, addresses current issues with respect to previously assumed values of unfiltered inleakage. Generally, these issues can only be resolved by inleakage testing. In light of Appendix B requirements and GL 2003-01, provide sufficient justification to explain why a value of zero for the control rooms unfiltered inleakage is appropriate for this proposed TS change request. Provide details regarding your control room, design, maintenance and assessments to justify the use of this number.
5.
The proposed TS specifies that a "designated trained hatch closure crew" is available to close the Containment Structure Equipment Hatch Shield Doors rather than a "dedicated" crew who would have no other duties. Specify what other duties the designated crew will have and where they will be stationed relative to the air lock doors.
6.
Provide a detailed account of the timing and flow rates, and filtration of the control room Heating Ventilation and Air Conditioning as it responds to the accident. Please justify the assumption that one of the two emergency control room filtration trains are operating within 10 minutes of the accident. Is this action automatic or manual?
7.
What criteria will be used to determine if closure of the containment is necessary in the event that environmental conditions could impact fuel handling? Has the impact of wind on fuel handling been evaluated (for example, reduced pool visibility due to pool surface disruption)? What steps would be taken in the event of severe weather to minimize the impact of flying debris?
8.
Criterion 64 of 10 CFR Part 50, Appendix A states that means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents. The proposed TS change should consider how Criterion 64 will be met in the event of a Fuel Handling Accident (FHA) with the equipment hatch open. Moreover, this information should be included as part of the Bases discussion.
Provide the bases for meeting Criterion 64 for the proposed TS change. Please confirm that your emergency planning dose assessment methodology includes the ability to assess this accident. For example, does your methodology include the capability to determine the source term, release rate out of containment, meteorology and consider feedback via field monitoring health physics survey teams? Have you evaluated the need for any special radiological monitoring or survey equipment (i.e., in-plant equipment or field team survey equipment) to evaluate the radiological conditions of this accident scenario? Will your emergency response personnel be trained to deal with this accident scenario?
9.
10 CFR 50.36 states that:
A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:
(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The proposed analysis utilizes an initial condition of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of fuel decay for the FHA.
The proposed TS does not provide a limiting condition of operation for this initial condition. Please justify why this decay time does not meet Criterion 2 of 10 CFR 50.36 or modify the TS to include the decay time.
Emergency Plan Considerations 10.
Will your Emergency Plan be updated to include an accident release through the equipment hatch? Will your Emergency Operating Procedures be updated to address the specific details needed to respond to this accident scenario?
11.
Will you inform the State Emergency Response personnel about this accident scenario?
Meteorological Monitoring Program:
12.
Confirm that the 2000 through 2002 meteorological data used in the atmospheric dispersion analyses are of high quality, representative of long term conditions, and suitable for use in the dispersion analyses. The primary intent of the following questions is to assess the overall quality of the meteorological data as collected and processed for use in the atmospheric dispersion calculations. These questions are also intended to ascertain whether there have been any changes made to the meteorological monitoring program from that currently described in the Farley Nuclear Plant, Final Safety Analysis Report Section 2.3.3 During the period of data collection, did the measurement program meet the guidelines of Regulatory Guide (RG) 1.23, Onsite Meteorological Programs? Was the tower base area on natural surface (e.g., short natural vegetation) and was the tower free from obstructions (e.g., trees, structures) and micro-scale influences to ensure that the data were representative of the overall site area? In the case of possible obstructions, were trees, structures, etc. at least 10 times their height away from the meteorological tower?
What types of surveillance activities were performed to ensure that data were of high quality? Were instrumentation problems and resulting questionable data identified and corrected in a timely manner? Were calibrations properly performed and systems found to be within guideline specifications? What additional data reviews were performed following data collection and prior to archival? If data abnormalities occurred, describe the abnormalities and why the data were still deemed to be adequate (a detailed response for each individual data point is not expected). Were the data compared with other site historical or regional data? If so, what were the findings?
What additional data reviews were performed prior to their input into the atmospheric dispersion calculations to ensure that conversion and reformatting to the ARCON96 format were properly performed?
Meteorological Data 13.
Staff review indicates apparent weaknesses in the submitted 2000 through 2002 meteorological data as discussed below. Please check and amend the data files as appropriate. Provide a copy of the amended files and the basis for any remaining departures from expected conditions and RG 1.23. Alternatively, provide a replacement set of high quality data.
a)
The data files for each year contain non-consecutive (missing) date/hour sequences. In addition, May 2001 contains several records containing duplicate hourly labels. ACRON96 requires that the meteorological data files contain one record per hour. Any hour with invalid data should be represented with each data field completely filled in with 9s (for example, 999" for a wind direction field, 9999" for a wind speed field, and 99" for the stability class field).
b)
The ASIIMETDATA files contain wind direction values of 0. If these values are intended to represent valid wind direction observations from the North, they are being misinterpreted by ARCON96 as invalid data (valid wind direction values provided as input to ARCON96 should range from 1 to 360). In addition, there are several occurrences of negative wind direction values. All valid wind direction values should range from 1 to 360; all invalid wind direction values should be identified by 999".
c)
Likewise, the ASIIMETDATA files contain wind speed values of -44700 mps. If these values are intended to represent invalid wind speed observations, they should be reset to 9999".
d)
There are a number of hours in the RAWMETDATA files with delta-temperature values of -9999 F that were classified as stability class A in the ASIIMETDATA files. There are also a number of hours with delta-temperature values of 100 F that were classified as stability class G. If these values represent invalid delta-temperature observations, the associated stability class value should be set to 99".
e)
A high frequency of stability class A was reported during 2002 (20.8%) as compared to an average of 2.4% for 2000 and 2001. In fact, nearly 65% of the hourly data reported from Julian day 262 (September 19, 2002) through the end of 2002 were classified as stability class A. One continuous period of stability class A data which began on hour 22 on Julian day 295 (October 22, 2002) lasted for 188 hours0.00218 days <br />0.0522 hours <br />3.108466e-4 weeks <br />7.1534e-5 months <br />. Please explain what might have caused this high occurrence of stability class A conditions during the last four months of 2002.
f)
A wind rose comparing wind direction frequency distributions between the Jan 2000 - Dec 2002 onsite meteorological database provided in the RAI response dated 11/11/2003 and the Apr 1971 - Mar 1975 onsite data provided in FSAR Table 2.3-8B is provided below:
0%
5%
10%
15%
N NNE NE ENE E
WNW NW NNW 1971-1975 2000-2002 This wind rose shows considerable discrepancies for the following wind direction sectors:
Period of Record Wind Direction Sector SSE S
SSW Apr 1971 - Mar 1975 5.6%
6.5%
6.5%
Jan 2000 - Dec 2002 1.4%
1.4%
2.3%
Please explain what might have caused these discrepancies in reported wind direction frequency distributions between the 1971-1975 data set and the 2000-2002 data set.
Atmospheric Dispersion Factors 14.
The dose calculations assume automatic control room isolation with manual actuation of the control room emergency filtration system occurs within 10 minutes of the accident.
During the isolation mode, unfiltered inleakage into the control room is assumed to be 10 cfm. This inleakage of unfiltered air, which can occur through doorways, envelope penetrations, and ventilation system components, was apparently modeled using the control room intake /Q values. Please verify that there are no other potential unfiltered inleakage pathways that could result in /Q values that are higher than the control room intake /Q values.
15.
Staff review indicates apparent deficiencies in the ARCON96 run inputs provided in your Request for Additional Information (RAI) response dated November 11, 2003, as discussed below. Please consider the impact of these modeling deficiencies, along with the meteorological database abnormalities discussed previously, on the ARCON96 results. If necessary, provide a set of revised /Q values, along with a copy of the associated ARCON96 printouts.
a)
The distance to receptor ARCON96 input parameter should represent the horizontal distance between the release point and the receptor. However, it appears from Table 1 in Enclosure 2 to the RAI response that the distance to receptor values provided as input to the ARCON96 runs also include the difference in height between the release point (e.g., the containment hatch door) and the receptor (e.g., CR air intake). Note that ARCON96 uses the difference in the release height and intake height input parameters (as well as the elevation difference input parameter) to determine the slant path distance between the release point and the receptor for ground level releases. Since the release height and intake height input parameters for the ARCON96 runs also account for the difference in elevation between the release point and the receptor, the difference in elevation between the release point and the receptor appears to have been double-counted.
b)
A value of 4.0 was used for the averaging sector width constant input parameter instead of a value of 4.3 as suggested in Table A-2 of Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants.
16.
In addition to describing the inputs, assumptions, and bases used to execute ARCON96 for containment hatch door releases to the control room air intakes, the RAI response dated 11/11/2003 also provides additional information regarding vent and containment leakage releases and Technical Support Center air intake /Q values. ARCON96 outputs for these additional source/receptor combinations are also provided. Since the resulting /Q values for these additional source/receptor combinations are neither presented or discussed in the original Containment Equipment Hatch License Amendment Request, what is the intent in providing this additional information in the RAI response submittal?
Health Physics
- 17.
In the event of an in-containment FHA, a designated crew of workers will be assigned to manually close the containment structure equipment hatch shield doors. A best estimate thyroid dose of about 46 rem was provided, based on a one-hour stay-time inside containment. Provide an estimate of whole body (deep dose) for a member of the crew to ensure that these doses are consistent with exposure guidelines of 10 CFR 50.47(b)(11). List all pertinent assumptions (e.g., airborne and external source terms, stay time, respiratory protection factors, etc.) taken to develop the dose estimates for these emergency workers. Crew member whole body doses should include external doses from all in-containment sources including worker immersion/shine dose from noble gas and iodine airborne cloud, and contained system/component sources (e.g. filters). Doses should include in-containment exposure received traversing to and from the hatch worksite.
18.
For the emergency response action of closing the equipment hatch, describe the radiation protection job planning and job-site coverage and the radiation surveys, personal protection, and dose monitoring equipment that will be provided to the crew members. Describe the initial (and continuing) radiological training that will be provided, including whether the crew workers will be qualified and trained to use respiratory protection devices or other means to limit intake of radioactive materials (e.g., use of KI to minimize radioiodine uptake of the thyroid). Describe any mockup (or actual) training or practice that will be provided to the crew members that would minimize in-containment stay-time during the accident.
Joseph M. Farley Nuclear Plant cc:
Mr. Don E. Grissette General Manager -
Southern Nuclear Operating Company Post Office Box 470 Ashford, Alabama 36312 Mr. B. D. McKinney, Licensing Manager Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201-1295 Mr. M. Stanford Blanton Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201 Mr. J. B. Beasley, Jr.
Executive Vice President Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701 Chairman Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Resident Inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 Columbia, Alabama 36319 William D. Oldfield SAER Supervisor Southern Nuclear Operating Company P. O. Box 470 Ashford, Alabama 36312