ML070990089
| ML070990089 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/17/2007 |
| From: | Tam P NRC/NRR/ADRO/DORL/LPLIII-1 |
| To: | Conway J Nuclear Management Co |
| References | |
| TAC MC8971 | |
| Download: ML070990089 (9) | |
Text
April 17, 2007 Mr. John T. Conway Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT (MNGP) - CORRECTION OF SAFETY EVALUATION ASSOCIATED WITH ALTERNATIVE SOURCE TERM AMENDMENT (TAC NO. MC8971)
Dear Mr. Conway:
On December 7, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 148 to Facility Operating License No. DPR-22 for MNGP. The amendment revised the MNGP licensing basis by implementing the full-scope alternative source term methodology, and revised portions of the Technical Specifications.
Subsequent to issuance, we noted that the safety evaluation (SE) associated with the amendment contains a number of typographical errors. Accordingly, we are re-issuing corrected Pages12, 15, 16, 19 and 20 (enclosed). The correction of these typographical errors does not change the NRC staffs conclusion regarding Amendment No. 148.
The NRC staff regrets any inconvenience these errors may have caused you.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
SE pages 12, 15, 16, 19, and 20 cc w/encl: See next page
April 17, 2007 Mr. John T. Conway Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT (MNGP) - CORRECTION OF SAFETY EVALUATION ASSOCIATED WITH ALTERNATIVE SOURCE TERM AMENDMENT (TAC NO. MC8971)
Dear Mr. Conway:
On December 7, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 148 to Facility Operating License No. DPR-22 for MNGP. The amendment revised the MNGP licensing basis by implementing the full-scope alternative source term methodology, and revised portions of the Technical Specifications.
Subsequent to issuance, we noted that the safety evaluation (SE) associated with the amendment contains a number of typographical errors. Accordingly, we are re-issuing corrected Pages12, 15, 16, 19 and 20 (enclosed). The correction of these typographical errors does not change the NRC staffs conclusion regarding Amendment No. 148.
The NRC staff regrets any inconvenience these errors may have caused you.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
SE pages 12, 15, 16, 19, and 20 cc w/encl: See next page DISTRIBUTION PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 RidsNrrPMPTam RidsNrrLATHarris RidsOGCRp RidsAcrsAcnwMailCenter RidsNrrDirsltsb G. Hill, OIS RidsRgn3MailCenter RidsNrrDorlDpr Amendment Accession No.: ML070990089 OFFICE LPL3-1/PM LPL3-1/LA LPL3-1/BC NAME PTam THarris LRaghavan DATE 4/12/07 4/12/07 4/17/07 OFFICIAL RECORD COPY
letter dated August 22, 2006, the licensee provided corrected control room elevated release /Q values and determined the impact on the CRDA dose analysis results due to this correction.
Because the fumigation control room /Q values were not affected by the error, and fumigation was assumed for the duration of the pre-isolation elevated release period for Case 2 (MVP isolation), the control room dose calculated for Case 2 was not impacted by the correction. The increase in the licensees calculated dose in the control room for the CRDA with SJAEs in operation (Case 1) was less than 0.04 rem TEDE, assuming a conservative 2.1 percent increase due to the corrected control room elevated /Q values. The corrected total CRDA Case 1 control room dose is 1.70 rem TEDE. The dose in the control room continues to meet l the requirements of 10 CFR 50.67 and GDC-19.
3.1.3.1 CRDA Summary The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed changes. The assumptions found acceptable to the NRC staff are presented in Table 1. The licensees calculated dose results are presented in Table 2. Based upon the information provided by the licensee, the NRC staff finds that the licensee used analysis methods and assumptions consistent with the guidance of RG 1.183 in its analysis of the CRDA. The NRC staff compared the radiation doses estimated by the licensee to the applicable acceptance criteria in SRP 15.0.1, and to the results estimated by the NRC staff in its confirmatory calculations. The NRC staff finds, with reasonable assurance, that the licensees estimates of the EAB, LPZ, and control room doses for the CRDA will continue to comply with the dose guidelines in 10 CFR 50.67, GDC-19, and the accident-specific dose acceptance criteria in SRP 15.0.1 and RG 1.183.
3.1.4 Fuel Handling Accident (FHA)
The FHA analysis postulates that a spent fuel assembly is dropped during refueling operations.
For the licensing basis case, the postulated FHA involves the drop of a fuel assembly in the reactor vessel cavity over the reactor core during refueling operations. The licensee also considered the consequences of dropping a fuel assembly during fuel movement at other locations.
The drop of a fuel assembly in the reactor vessel cavity over the reactor core was found to be the limiting design basis case. A fuel assembly is postulated to drop from the maximum height allowed by the refueling platform and to fall onto the fuel in the reactor. The drywell head and reactor vessel head are assumed to be removed. At this location, the maximum drop (free fall distance) is approximately 27 feet for the fuel assembly, and fuel pin damage is postulated to occur to both the dropped assembly and to a portion of those assemblies impacted in the reactor core. The extent of damage is calculated based on the free fall distance and the resulting kinetic energy of the dropped assembly. In accordance with the current licensing basis, this drop is conservatively assumed to damage 125 fuel pins.
The gap activity from the damaged pins is the radioactive source term for this event. A radial peaking factor of 1.7 is assumed in the analysis. A post-shutdown 24-hour decay period was used to determine the release activity inventory. An effective total decontamination factor (DF) of 200 for the released radioiodine species is assumed based on a minimum water depth of 23 feet. The nominal water depth (i.e., the distance from the top of the water in the vessel to the Corrected by letter of 4/17/07 l
The NRC staff generated revised data files to include the above criteria for zero values of wind direction and wind speed, for both the 10-meter and 43-meter hourly measurements previously provided by the licensee and the 10-meter and 100-meter hourly data provided for the current amendment application. The NRC staff performed a quality review of these corrected data files using the methodology described in NUREG-0917, Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data, and also computer spreadsheets. With respect to atmospheric stability measurements, the time of occurrence and duration of stable and unstable conditions were found to be consistent with expected meteorological conditions.
Stable and neutral conditions were reported to occur at night and unstable and neutral conditions during the day, with neutral or near-neutral conditions predominating during each year. During 1998-2002, wind direction frequency occurrences were similar among the three measurement heights and from year to year at each height. Winds were predominately from the north northwest and south at all levels. Wind speed at each height was also consistent for each of the 5 years. A comparison of the joint frequency distributions derived by the NRC staff from these ARCON96-formatted hourly data with joint frequency distributions developed by the licensee showed some slight differences. These differences appear to be the result of use of two slightly different data sets in generation of the joint frequency distributions. The licensee calculated joint frequency distributions based upon a meteorological data set of measurements from the single measurement train having the highest data recovery. The ARCON96 files used a meteorological data set based upon measurements from a second train, when data were not available from the first train. The NRC staff made an overall comparison of wind rose calculations from joint frequency distributions determined from the above two data sets and found the two sets of joint frequency distributions to be similar.
3.1.5.2 Control Room Atmospheric Dispersion Factors The licensee used guidance provided in RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, to generate the new control room atmospheric dispersion factors for assumed releases from the turbine building vent, the reactor building vent, and the reactor building wall nearest the two assumed intake locations, i.e., the control room and administrative building intakes. All releases were assumed to be point source releases. The licensee calculated these new control room /Q values using the ARCON96 computer code. Because the heights of these postulated release locations are less than 21/2 times the height of adjacent buildings, they were modeled using the ARCON96 ground level release option in accordance with RG 1.194. The licensee executed ARCON96 using the 1998-2002 hourly data from the MNGP onsite meteorological tower from wind measurements made at a height of 10 meters and 43 meters and atmospheric stability measurements between the 43-meter and 10-meter levels. The licensee used the taut string methodology described in RG 1.194 to estimate the shortest distance around or over intervening buildings which was input as the source-to-receptor distance. RG 1.194 states that ARCON96 is an acceptable methodology for assessing control room /Q values for use in DBA radiological analyses. The NRC staff evaluated the applicability of the ARCON96 model and concluded that there are no unusual siting, building arrangements, release characterization, source-receptor configuration, meteorological regimes, or terrain conditions that preclude use of the ARCON96 model for the MNGP site.
The licensee also generated new /Q values for releases from the off-gas stack using PAVAN for the actual distance to the intake locations. Fumigation was assumed to occur for one-half hour during the worst 2-hour EAB exposure period. Guidance in RG 1.194 states that l Corrected by letter of 4/17/07 l
comparative calculations should be made using both the PAVAN and ARCON96 computer codes. PAVAN estimates should be made for several distances in each wind direction sector with the objective of identifying the maximum /Q value, but fumigation need not be assumed.
The NRC staff made comparison estimates using the RG 1.194 methodology. The licensees
/Q values for the 0-4 day time periods were larger than the NRCs /Q values. The NRC-generated 4-30 day /Q value was somewhat higher than the 4-30 day /Q value generated by the licensee using PAVAN only. The only DBA dose analysis assuming elevated releases after 4 days is the LOCA, for the containment release pathway. Because the majority of the containment release from the LOCA occurs within the first 4 days, and since the NRC staff found that the licensees /Q for the periods up to 4 days are conservative compared to the NRC staffs calculated values, the impact of the higher /Q value for the 4-30 day time period on the total control room dose for the LOCA is minimal.
For each release point, dispersion factors were calculated for both the control room intake and the administrative building intake. The limiting dispersion factor of these two was chosen by the licensee to model all pathways to the control room. Unfiltered inleakage conservatively assumed the limiting X/Q value, due to the proximity of the local intake to the control room. The licensee modeled unfiltered inleakage as though it entered the control room at the same locations as the more limiting of the two /Q values.
The NRC staff qualitatively reviewed selected inputs to the licensees computer runs and found them generally consistent with site configuration drawings and NRC staff practice, made confirmatory calculations using the corrected data files discussed above. Therefore, the NRC staff finds the licensees estimates to be acceptable, with the stipulation on use of the 4-30-day offgas stack X/Q values as discussed above. These control room /Q values are listed in Tables 3 through 5.
3.1.5.3 Offsite Atmospheric Dispersion Factors The licensee used RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, and the PAVAN computer code to calculate /Q values for the EAB and LPZ.
For postulated releases from the reactor building wall, reactor building vent and turbine building vent, the licensee used the PAVAN ground level release mode, inputting a reactor building height of 43.6 meters and reactor building cross-sectional area of 1829 square meters. The licensees meteorological input to PAVAN consisted of a joint frequency distribution of wind speed, wind direction, and atmospheric stability data for the 1998-2002 period. Wind speed and direction data from the meteorological towers 10-meter level were used. Stability class was based on the temperature difference data between the 43-meter and 10-meter levels on the onsite meteorological tower.
The licensee used the elevated release mode option of the PAVAN computer code to model the release from the 100-meter free-standing off-gas stack at MNGP. Consistent with RG 1.183, Section 5.3, fumigation was assumed to occur for one-half hour during the worst 2-hour EAB l exposure period. The licensee modeled the reduction in effective stack height due to topography by inputting the highest terrain height of 9 meters in all directions.
Corrected by letter of 4/17/07 l
The NRC staff considered the transport of the sodium pentaborate from the reactor vessel to the suppression pool. The SLC system injects the sodium pentaborate to the reactor vessel through a sparger that consists of a vertical pipe in the core region. SLC system injection occurs approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the onset of the LOCA. At this time, low-pressure coolant injection (LPCI) pumps are transferred from the injection mode to recirculating the suppression pool water for mixing. Core spray from at least one core spray pump (approximately 2000 gallons per minute) is injected at the top of the core for cooling and liquid level control. The core spray flows down, mixes with SLC injection and transports it to the break and, ultimately, to the suppression pool. Although transport of SLC contents by core spray alone would be slow, sufficient sodium pentaborate is present in the SLC, which, when transferred to the suppression pool, will ensure long-term pH control.
Using the LPCI in the recirculation mode for suppression pool cooling provides mixing to assure l that the sodium pentaborate is evenly distributed. The NRC staff concluded that there would be l mixing and transport at some rate and that it was reasonable to assume the concentration of l sodium pentaborate in the core would equalize with the concentration in the suppression pool l within an acceptable time after SLC injection. As a consequence, there would be sufficient pH l control to deter and prevent iodine re-evolution. l The specific changes being made to TS Section 3.1.7, to require the SLC system to be l operable at all times during Run, Startup, and Hot Shutdown time, is appropriate for this action.
On the basis of the above discussion, the staff finds this change to be acceptable.
3.3 Control of Post-LOCA Suppression Pool Water pH The NRC staff reviewed the licensees analysis regarding maintaining suppression pool pH$7 for 30 days following a LOCA. According to RG 1.183, maintaining pH basic will prevent re-evolution of iodine from the suppression pool water.
After a LOCA, a variety of different chemical species are released from the damaged core One of them is radioactive iodine, which, when released to the outside environment, will significantly contribute to radiation doses. It is, therefore, essential to keep the iodine confined within the plants containment. According to NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plant, the iodine is release from the core in three different chemical forms; at least 95 percent is released in ionic form as cesium iodide (CsI) and the remaining 5 percent as elemental iodine (I ) and as hydriodic acid (HI), with not less than 1 percent of each as I and HI. CsI and HI are ionized in water environment and are, therefore, soluble. However, elemental iodine is scarcely soluble. It is of interest, therefore, to maintain as much as possible of the released iodine in ionic form. However, in the post-LOCA radiation environment in the containment, some of the ionic iodine dissolved in water is converted into elemental form.
The degree of conversion varies significantly with the pH of water. At higher pH, conversion to elemental form is low and at pH> 7, conversion is negligibly small. The relationship between the degree of conversion and pH is specified in Figure 3.1 of NUREG/CR-5950, Iodine Evolution and pH Control.
In MNGP, most of the iodine released from the core is assumed to end up in the suppression pool. Therefore, in order to keep it dissolved, the suppression pool water should be kept at pH$7 throughout the 30 day post-LOCA period. The licensee has demonstrated that because Corrected by letter of 4/17/07 l
of strong acid formation in the suppression pool, this is not achievable without adding buffering chemicals to control the water pH.
After a LOCA, the pH of the suppression water would be continuously decreasing due to formation of hydrochloric and nitric acids in the containment. Hydrochloric acid is formed from decomposition of the Hypalon cable insulation by radiation. About 3.08E-5 mols/liter of hydrochloric acid would be formed during the 30 day period. Nitric acid is formed by irradiation of air and water and about 2.28E-5 mols/liter of nitric acid would be formed during the same period. Both acids are strong acids and will significantly contribute to lowering suppression pool pH. In order to neutralize their effect, the licensee chooses to buffer suppression pool water by using sodium pentaborate from the SLC system. The main purpose of the SLC system is to control reactivity in the case of control rod failure. Since sodium pentaborate is derived from a strong base and a weak acid, it can also act as a buffer. Such buffering action could maintain pH in the basic range in the suppression pool despite the presence of strong acids. The licensee has calculated that in order to maintain pH in the suppression water basic, 1404 gallons of 10.7 percent solution of sodium pentaborate should be added into the suppression l pools water. The licensee stated that the addition is accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after a LOCA.
In order to evaluate beneficial effect of sodium pentaborate, the licensee calculated suppression pool pH for unbuffered and buffered cases. Without addition of sodium pentabarate but taking only credit for the presence of Cs(OH), the value of pH during the 30-day period was below 7, reaching a minimum value of 4.26. With the addition of sodium pentaborate, but without taking credit for Cs(OH), the pH will increase rapidly above 7 and can reach a value as high as 8.6, depending on the core thermal power and the volume of water in the suppression pool.
The NRC staff reviewed the licensees analysis and concludes that, based on the analysis, the suppression pool pH will stay basic for a period of 30 days after a LOCA.
3.4 Changes to the Technical Specifications The initial application, dated September 15, 2005, arrived when MNGP was still operating under the former custom TS. On June 5, 2006, the NRC staff issued Amendment No. 146, fully converting the former custom TS to the Improved Technical Specifications (ITS) format. The most obvious impact of this conversion is that the locations of various requirements are no longer the same, and that many requirements have been changed. The licensee's August 21, 2006, letter, provided reprinted ITS pages reflecting implementation of the AST methodology. Accordingly, the following subsections of this safety evaluation identify the changes both in the former custom format and in the current ITS format.
ITS 1.1, "Definitions" (Custom TS 1.0, "Definitions")
The definition of Dose Equivalent I-131 is changed in two ways: (1) "thyroid dose" is changed to "dose, and (2) the source of dose conversion factors is changed from TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," dated 1962, and RG 1.109, Rev. 1 to Federal Guidance Report (FGR)-11, "Limiting Values of Radionuclide Intake and Air Concentration Factors for Inhalation, Submersion, and Ingestion," September 1988, and Corrected by letter of 4/17/07 l
Monticello Nuclear Generating Plant cc:
Jonathan Rogoff, Esquire Commissioner Vice President, Counsel & Secretary Minnesota Department of Commerce Nuclear Management Company, LLC 85 7th Place East, Suite 500 700 First Street St. Paul, MN 55101-2198 Hudson, WI 54016 Manager - Environmental Protection Division U.S. Nuclear Regulatory Commission Minnesota Attorney Generals Office Resident Inspector's Office 445 Minnesota St., Suite 900 2807 W. County Road 75 St. Paul, MN 55101-2127 Monticello, MN 55362 Michael B. Sellman Manager, Nuclear Safety Assessment President and Chief Executive Officer Monticello Nuclear Generating Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 2807 West County Road 75 Hudson, MI 54016 Monticello, MN 55362-9637 Nuclear Asset Manager Commissioner Xcel Energy, Inc.
Minnesota Pollution Control Agency 414 Nicollet Mall, R.S. 8 520 Lafayette Road Minneapolis, MN 55401 St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 November 2005