ML071410047

Request for Additional Information Regarding Nine Mile Point Nuclear Station, Unit No. 1, Implementation of Alternative Source Term

ML071410047
Person / Time
Site:	Nine Mile Point
Issue date:	05/21/2007
From:	David M NRC/NRR/ADRO/DORL/LPLI-1
To:	O'Connor T Nine Mile Point
david marshall NRR/DORL 415-1547
References
TAC MD3896
Download: ML071410047 (7)

Text

May 21, 2007 Mr. Timothy J. OConnor Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING NINE MILE POINT NUCLEAR STATION, UNIT NO. 1, IMPLEMENTATION OF ALTERNATIVE SOURCE TERM (TAC NO. MD3896)

Dear Mr. OConnor:

By letter dated December 14, 2006, Nine Mile Point Nuclear Station, LLC requested an amendment to the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) Renewed Facility Operating License. The proposed license amendment would revise the accident source term used in the NMP1 design basis radiological consequence analyses in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.67. The revised accident source term would replace the current methodology that is based on TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," with the alternative source term methodology described in Regulatory Guide (RG) 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided in that letter and has determined that additional information is needed to complete its review.

Enclosed is the NRC staffs request for additional information (RAI). The RAI was discussed with your staff on May 10, 15, and 18, 2007, and it was agreed that your response would be provided within 60 days from the date of this letter.

Sincerely,

/RA/

Marshall J. David, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosure:

RAI cc w/encl: See next page

May 21, 2007 Mr. Timothy J. OConnor Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING NINE MILE POINT NUCLEAR STATION, UNIT NO. 1, IMPLEMENTATION OF ALTERNATIVE SOURCE TERM (TAC NO. MD3896)

Dear Mr. OConnor:

By letter dated December 14, 2006, Nine Mile Point Nuclear Station, LLC requested an amendment to the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) Renewed Facility Operating License. The proposed license amendment would revise the accident source term used in the NMP1 design basis radiological consequence analyses in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.67. The revised accident source term would replace the current methodology that is based on TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," with the alternative source term methodology described in Regulatory Guide (RG) 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided in that letter and has determined that additional information is needed to complete its review.

Enclosed is the NRC staffs request for additional information (RAI). The RAI was discussed with your staff on May 10, 15, and 18, 2007, and it was agreed that your response would be provided within 60 days from the date of this letter.

Sincerely,

/RA/

Marshall J. David, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosure:

RAI cc w/encl: See next page DISTRIBUTION:

PUBLIC RidsNrrPMMDavid RidsNrrLASLittle RidsNrrDraAadb LPLI-1 PDI-1 Reading File RidsNrrDciCsgb RidsOgcRp RidsNrrDirsItsb RidsNrrAcrsAcnwMailCenter Accession Number: ML071410047 NRR-088 OFFICE LPL1-1/PM LPL1-1/LA ITSB/BC CSGB/BC AADB/BC LPL1-1/BC NAME MDavid SLittle TKobetz* AHiser* MKotzalas* MKowal DATE 5/21/07 5/21/07 04/11/07 04/02/07 04/05/07 5/21/07

• RAI provided by memo on date shown OFFICIAL RECORD COPY

Nine Mile Point Nuclear Station cc:

Mr. Michael J. Wallace Mark J. Wetterhahn, Esquire President Winston & Strawn Nine Mile Point Nuclear Station, LLC 1700 K Street, NW c/o Constellation Energy Group Washington, DC 20006 750 East Pratt Street Baltimore, MD 21202 Supervisor Town of Scriba Mr. Mike Heffley Route 8, Box 382 Senior Vice President and Chief Oswego, NY 13126 Nuclear Officer Constellation Generation Group Carey W. Fleming, Esquire 1997 Annapolis Exchange Parkway Senior Counsel Suite 500 Constellation Generation Group, LLC Annapolis, MD 21401 750 East Pratt Street, 17th Floor Baltimore, MD 21202 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 126 Lycoming, NY 13093 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy Electric Division NYS Department of Public Service Agency Building 3 Empire State Plaza Albany, NY 12223 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399

REQUEST FOR ADDITIONAL INFORMATION NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 IMPLEMENTATION OF ALTERNATIVE SOURCE TERM The Nuclear Regulatory Commission (NRC) staff has performed its initial review of your December 14, 2006, request to revise the accident source term used in the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) design basis radiological consequence analyses in accordance with 10 CFR 50.67. As a result of that review, we have determined that additional information is required to adequately evaluate the acceptability of the proposed revision.

ITSB-1 Technical Specification (TS) 3.4.5, Control Room Air Treatment System

[CRATS] is required in order to ensure that the CRATS is operable during a loss-of-coolant accident (LOCA). NUREG-1433, which contains Standard Technical Specifications (STS) for General Electric BWR/4 plants, has STS 3.7.4, Main Control Room Environmental Control (MCREC) System, which provides a similar function to the NMP1 CRATS. STS 3.7.4 is applicable in hot shutdown since plant pressure increases the likelihood of a LOCA and, therefore, a fission product release. Criterion 3 of 10 CFR 50.36(c)(2)(ii) states that a TS limiting condition for operation must be established for "a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." TS 3.4.5, Specification (a), does not require that the CRATS be operable when the plant is in the hot shutdown condition.

Please explain how NMP1 TS 3.4.5, Specification (a), satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

ITSB-2 TS 3.4.5, Specification (e), allows stated operations to continue to occur for 7 days with an inoperable CRATS. If the CRATS is inoperable after 7 days, Specification (f) requires that the reactor be shutdown to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and that refueling operations and operations with the potential to drain the reactor vessel (OPDRVs) be suspended within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

10 CFR 50.36(c)(2) states, "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met."

NUREG-1433 STS 3.7.4 for the MCREC System, which provides a similar function to the NMP1 CRATS, contains Condition E, which states to immediately shutdown and cooldown upon a complete loss of the MCREC System (as opposed to 7 days in NMP1 TS 3.4.5, Specification (e)). Also, STS 3.7.4, Condition F, states to suspend movement of recently irradiated fuel assemblies and to suspend OPDRVs immediately upon a complete loss of the MCREC System (as opposed to 7 days plus 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in NMP1 TS 3.4.5, Specifications (e) and (f)). The remedial actions provided in NMP1 TS 3.4.5, Specifications (e) and Enclosure

(f) do not appear conservative (i.e., safe remedial actions) when compared to the NUREG-1433 STS.

Please explain how NMP1 TS 3.4.5, Specifications (e) and (f), provide safe remedial actions in accordance with 10 CFR 50.36(c)(2).

AADB-1 From the spreadsheet provided in the submitted design analysis H21C090, it is difficult to determine which iodine decontamination factor (DF) is being used as an overall DF. In the submitted fuel-handling accident (FHA) design analysis, it is stated that an elemental iodine DF of 268 and an organic iodine DF of 1 is assumed. It is also stated in that same analysis that an overall DF of 200 is assumed to characterize the combined removal of elemental and organic iodine activity in the volume of coolant between the damaged assemblies and water surface. However, if 268 and 1 are respectfully assumed as elemental and organic DFs, the resulting overall iodine DF would likely be closer to 191.

Therefore, please clarify and justify what value is used as an overall iodine DF.

AADB-2 Please identify the amount of fuel assembly water coverage available during the postulated FHA, the reference for this value, and how this value is used to calculate the effective DF associated with each radioactive nuclide species.

AADB-3 With regard to the evaluation of reactor building (RB) cloud shine dose, there is a statement made in Appendix H of the submitted LOCA analysis that reads, dose varies as a function of the inverse of distance-squared It is agreed that this holds true for point sources; however, the RB to control room (CR) source-detector geometry described in that same attachment indicates a vast dissimilarity to the geometry that would make a 1/r2 relationship applicable. It is noted that this relationship is being used, not to calculate, but to validate the actual RB cloud shine dose model, whose code input is given in Appendix G of the submitted LOCA design analysis. Nevertheless, a comparison is made between the results of the Appendix H verification and the Appendix G model result.

Therefore, please provide a description of, and the plant general arrangements used to determine, the three-dimensional geometry implemented in the Appendix G model of RB cloud shine dose to the CR. The inclusion of an illustration may also be considered to help characterize material density and the specific geometry (source-shield-receiver) that is modeled.

AADB-4 Appendix A, Section 6.1, of Regulatory Guide (RG) 1.183 states that the activity available for release via MSIV [main steam isolation valve] leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage (see Regulatory Position 3). Regulatory Position 3 presents the source term, in the form of activity release fractions, released into containment. It is understood that this containment source term accounts for phenomena that would serve to inhibit activity release from the vessel, prior to transport through main steam and other bypass piping. The guidance of RG 1.183 further allows for the credit of other containment removal mechanisms

(i.e., natural deposition and drywell spray); however, applying these additional removal mechanisms, prior to and in addition to crediting pipe deposition, can substantially change the containment source term assumed to enter the main steam and other bypass piping, thus rendering the containment source term of Regulatory Position 3 inapplicable. Because the cumulative effect of these removal mechanisms was not explicitly addressed by the containment source term provided in Regulatory Position 3, consideration should be given to the interaction of each removal mechanism with the source term of RG 1.183 when modeling the transport of activity from the drywell through bypass pathways.

Therefore, please provide information to show that the cumulative effect of assuming the release of the containment source term, natural deposition, drywell spray removal, followed by pipe deposition, for the postulated LOCA at NMP1, does not compromise the conservative characteristics of the dose analysis.

AADB-5 The phenomena referred to as impaction is generally assumed to take place under conditions characterized by relatively high flow rates and turbulence, where the concentration of airborne particulate, or aerosols, is substantial. For the release model used for NMP1, impaction is credited; however, the assumed flow rate is low and potentially laminar, and particle settling in the pipe has been credited as well.

Therefore, please provide the effective DFs associated with the individual aerosol removal mechanisms that were credited for the NMP1 LOCA analysis, and explain how the credit taken for aerosol impaction accurately accounts for the relatively low flow rate assumed and the settling of large particulate in the pipe length.

AADB-6 Was leakage from a gland seal condenser considered as a potential post-control rod drop accident (CRDA) release path? If not, please provide justification for not considering this path. Also, either provide justification for excluding the steam jet air ejector as a potential post-CRDA release path, or indicate the dose consequence from this potential path as it applies to the dose analysis.

CSGB-1 There are two ways of calculating generation of hydrochloric acid (HCL) from cable insulation in the post-LOCA radiation environment. In the method used in your submitted analysis, production of HCL is based on the mass of cable and on radiation dose rate at the cable jacket surface, multiplied by a flux averaging factor. In the other methods, described in NUREG-1081, "Post-Accident Gas Generation from Radiolysis of Organic Materials," September 1984, generation of HCL depends on the cable jacket surface area and on the energy release per unit volume of containment. The main difference between these methods is that, in the first method, an average energy flux through the cable insulation material is used and, in the second method, the energy flux existing at the surface of the cable is used.

For -radiation the difference between these two methods is very small.

However, for -radiation the difference between these two methods is significant,

corresponding to the generation of significantly less HCL in your method. Please justify the use of your method which produces non-conservative results.