ML042430504
| ML042430504 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/10/2004 |
| From: | Mahesh Chawla NRC/NRR/DLPM/LPD3 |
| To: | Solymossy J Nuclear Management Co |
| Chawla M., NRR/DLPM, 415-8371 | |
| Shared Package | |
| ML042580009 | List: |
| References | |
| TAC MC1843, TAC MC1844 | |
| Download: ML042430504 (20) | |
Text
September 10, 2004 Mr. Joseph M. Solymossy Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -
ISSUANCE OF AMENDMENTS RE: SELECTIVE IMPLEMENTATION OF ALTERNATE SOURCE TERM FOR FUEL HANDLING ACCIDENTS (TAC NOS. MC1843 AND MC1844)
Dear Mr. Solymossy:
The Commission has issued the enclosed Amendment No. 166 to Facility Operating License No. DPR-42 and Amendment No. 156 to Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated January 20, 2004, as supplemented by letters dated May 19, July 13, and August 16, 2004.
The amendments change the Prairie Island TSs on containment to implement a portion of TS Task Force Traveler 5, Revise containment requirements during handling irradiated fuel and core alterations. The amendments also selectively implement an alternate source term per Title 10 of the Code of Federal Regulations, Part 50, Section 50.67, to perform the radiological consequences analysis of the design-basis fuel handling accident which supports the TS changes.
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosures:
- 1. Amendment No. 166 to DPR-42
- 2. Amendment No. 156 to DPR-60
- 3. Safety Evaluation cc w/encls: See next page
September 10, 2004 Mr. Joseph M. Solymossy Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -
ISSUANCE OF AMENDMENTS RE: SELECTIVE IMPLEMENTATION OF ALTERNATE SOURCE TERM FOR FUEL HANDLING ACCIDENTS (TAC NOS. MC1843 AND MC1844)
Dear Mr. Solymossy:
The Commission has issued the enclosed Amendment No. 166 to Facility Operating License No. DPR-42 and Amendment No. 156 to Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated January 20, 2004, as supplemented by letters dated May 19, July 13, and August 16, 2004.
The amendments change the Prairie Island TSs on containment to implement a portion of TS Task Force Traveler 5, Revise containment requirements during handling irradiated fuel and core alterations. The amendments also selectively implement an alternate source term per Title 10 of the Code of Federal Regulations, Part 50, Section 50.67, to perform the radiological consequences analysis of the design-basis fuel handling accident which supports the TS changes.
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosures:
- 1. Amendment No. 166 to DPR-42
- 2. Amendment No. 156 to DPR-60
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC OGC WRuland DLPMDPR PDIII-1 Reading ACRS CHarbuck LRaghavan PLouden, RGN-III HWalker MChawla GHill(4) LBrown THarris RDennig MHart ADAMS Accession No.: ML042430504(Memo) ADAMS Accession No.: ML042570194(TS)
ADAMS Accession No.: ML042580009(Package)
OFFICE PDIII-1/PM PDIII-1/LA SC:SPSB SC:IROB OGC PDIII-1/SC HChernoff for NAME MChawla THarris RDennig TBoyce CMarco LRaghavan DATE 09/01/04 09/01/04 08/27/04 9/10/04 9/08/04 9/10/04 OFFICIAL RECORD COPY
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 166 License No. DPR-42
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 20, 2004, as supplemented by letters dated May 19, July 13, and August 16, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 166, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA HChernoff for/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 10, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 166 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.3.5-2 3.3.5-2 3.3.5-3 3.3.5-3 3.3.5-4 3.3.5-4 3.3.5-5 ------
3.9.4-1 3.9.4-1 3.9.4-2 3.9.4-2
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 156 License No. DPR-60
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 20, 2004, as supplemented by letters dated May 19, July 13, and August 16, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 156, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA HChernoff for/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 10, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 156 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.3.5-2 3.3.5-2 3.3.5-3 3.3.5-3 3.3.5-4 3.3.5-4 3.3.5-5 ------
3.9.4-1 3.9.4-1 3.9.4-2 3.9.4-2
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 166 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 156 TO FACILITY OPERATION LICENSE NO. DPR-60 NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By application dated January 20, 2004, as supplemented by letters dated May 19, July 13, and August 16, 2004, Nuclear Management Company (NMC, the licensee) requested a license amendment for Prairie Island Nuclear Generating Plant, Units 1 and 2. The proposed change will revise the Prairie Island Technical Specifications (TS) on containment to implement a portion of Technical Specifications Task Force Traveler 51 (TSTF-51), Revise containment requirements during handling irradiated fuel and core alterations. The licensee also proposed to selectively implement an alternative source term (AST) per Title 10 of the Code of Federal Regulations (10 CFR) Section, 50.67 to perform the radiological consequences analysis of the design-basis fuel handling accident (FHA) which supports the proposed TS changes.
The proposed changes would allow the movement of fuel that has not been recently irradiated within the containment without requiring containment integrity or in the spent fuel pool (SFP) enclosure without SFP enclosure integrity. Recently irradiated fuel is defined in TSTF-51 as fuel that has not been part of a critical core within a time period that is shown through analyses to allow for enough radiological decay so that the radiological consequences of a design-basis FHA will remain acceptable.
For partial adoption of TSTF-51, the licensee requested changes to the following TS:
TS 3.3.5, "Containment Ventilation Isolation Instrumentation" TS 3.9.4, "Containment Penetrations" The supplemental letters of May 19, July 13, and August 16, 2004, contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.
2.0 REGULATORY EVALUATION
This safety evaluation addresses the impact of the proposed changes on previously analyzed design-basis accident radiological consequences and the acceptability of the revised analysis results. The regulatory requirements for which the Nuclear Regulatory Commission (NRC) staff
based its acceptance are the accident dose criteria in 10 CFR 50.67. The NRC staff also based its acceptance on the guidance provided in Regulatory Position 4.4 of Regulatory Guide 1.183 (RG 1.183), Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors, and Section 6.4 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP). Except where the licensee proposed a suitable alternative, the NRC staff utilized the regulatory guidance provided in SRP Section 15.0.1, Radiological Consequence Analysis Using Alternative Source Terms, in performing this review. The NRC staff also considered relevant information in the Prairie Island Updated Safety Analysis Report (USAR), TSs, and TSTF-51.
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the regulatory and technical analyses, as related to the radiological consequences of design-basis accidents, performed by NMC in support of its proposed license amendment. Information regarding these analyses was provided in Exhibit B Section 4.0 and Exhibit E of the submittal and in supplementary letters dated July 13, and August 16, 2004.
The staff reviewed the assumptions, inputs, and methods used by NMC to assess these impacts. The NRC staff performed independent calculations to confirm the conservatism of the licensees analyses. However, the findings of this safety evaluation are based on the descriptions of the licensees analyses and other supporting information docketed by NMC.
TSTF-51, Revision 2, allows for removal of the TS operability requirements for engineered safeguards features, such as secondary containment and standby gas treatment during movement of fuel, provided that sufficient radioactive decay has taken place to ensure that offsite doses remain well within (i.e., 25 percent of) 10 CFR Part 100 limits. Because the licensee has applied to implement an AST selectively for the FHA, the regulatory dose criteria of 10 CFR 50.67 are used in lieu of the 10 CFR Part 100 dose limits in determining acceptability of the proposed TS changes.
3.1 Changes to Containment Technical Specifications The licensee is requesting changes to TS 3.9.4, Containment Penetrations, and TS 3.3.5, Containment Ventilation Isolation Instrumentation, and the applicable Bases.
With respect to TS 3.9.4, Containment Penetrations, the licensee is requesting that the Applicability of Limiting Condition for Operation (LCO) 3.9.4 be revised to require this TS to apply during the movement of recently irradiated fuel assemblies within containment. The licensee stated that this Applicability change will also be accompanied by a change to the Required Actions, and an accompanying change in the applicable Bases. The licensee requested that this specification be revised to require all penetrations to be closed during movement of recently irradiated fuel assemblies within containment, including one door in each air lock and the Containment Purge and Inservice Purge System penetrations. In addition, the licensee stated that the Bases of TS 3.9.4 will be changed to define recently irradiated fuel as fuel that has occupied the critical reactor core within the past 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
The NRC staff have reviewed this request and considered the licensees proposed actions.
The NRC staff notes that in the licensees submittal dated January 20, 2004, the licensee stated that the requested change implements a portion of TSTF-51, Revise containment requirements during handling irradiated fuel and core alterations as it applies to TS 3.9.4. In
this submittal, the licensee commits to the guidelines of TSTF-51 Reviewers Note for the assessment of systems removed from service during movement of irradiated fuel at Prairie Island Nuclear Generating Plant. Specifically, the guidelines of NUMARC 93-01, Revision 3, Section 11.3.6, Assessment Methods for Shutdown Conditions, Subsection 11.3.6.5, that will be adopted as a commitment:
During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the reactor coolant system decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in dose due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay, and to avoid unmonitored releases.
A single normal contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.
In addition, Exhibit I of the licensees submittal provides draft containment controls that the licensee states will implement the commitment to assess systems removed from service during movement of irradiated fuel. The licensee also states that changes to these draft containment controls will be made under plant procedures.
The NRC staffs review found the requested change acceptable because the licensees analysis of the design basis FHA shows compliance with dose criteria in 10 CFR 50.67 and the regulatory dose acceptance criteria of RG 1.183, Alternative Radiological Source Terms for Evaluation Design-Basis Accidents at Nuclear Power Reactors. This analysis was reviewed by the NRC staff and was found to have taken no credit for containment closure. The licensees FHA dose analysis results remain well within the dose criteria of 10 CFR 50.67, assuming the fuel has not been in the critical reactor core within the last 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, and therefore, not recently irradiated. Further discussion of the NRC staffs review of the licensees dose analysis is set forth in Section 3.2 below.
Note that TS 3.9.4 does require all containment penetrations to be closed during handling of recently irradiated fuel assemblies within containment, including one door in each air lock and the Containment Purge and Inservice Purge System penetrations. It is the NRC staffs understanding that the requested change does not impact TS requirements for systems needed to prevent or mitigate core alteration events other than the FHA. The NRC staff also understands that the proposed change does not change the requirements for systems needed for decay heat removal or requirements to maintain the specified water levels over irradiated fuel, and does not result in changes to the design basis. The NRC staff concludes that the requested change is acceptable.
With respect to LCO 3.3.5, Containment Ventilation Isolation Instrumentation (CVI), the licensee is requesting that the Applicability Note (b) of Table 3.3.5-1 and Condition C be removed from this TS. Proposed TS 3.9.4 will be revised to require the containment purge and
inservice purge systems to be isolated at all times when recently irradiated fuel is handled, which means isolation of these systems by the CVI will not be required.
The licensee stated that the FHA dose analysis demonstrates that releases from an FHA meet the dose acceptance criteria and are well within the dose criteria of 10 CFR 50.67 without credit for containment ventilation system isolation, assuming the fuel has not been in a critical reactor within the last 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. The licensee also stated that CVI is not required when handling recently irradiated fuel since the containment purge and inservice purge systems are required to be isolated as specified in proposed TS 3.9.4, and that the Containment Purge and Inservice Purge Systems are not credited for filtration during handling of fuel which is not recently irradiated. Therefore, CVI requirements during reactor shutdown are proposed to be removed from TS 3.3.5. The NRC staff finds this acceptable because the proposed changes to TS 3.9.4, approved by this amendment, eliminate the need for Note (b) and Condition C in LCO 3.3.5.
3.2 Radiological Consequences of the FHA In the postulated FHA, a fuel assembly is assumed to be dropped and damaged during fuel handling. This accident may take place either in the containment or the SFP. The licensee does not take credit for containment isolation or filtration by the SFP special ventilation system in its analysis, and has chosen the analysis inputs and assumptions so that the results of the single FHA analysis are bounding for the accident occurring in either location. In order to do that, the limiting control room atmospheric dispersion factor (i/Q) was chosen from the various possible release points. Section 3.2.1 below discusses the NRC staffs review of the licensees assumed atmospheric dispersion factors.
The entire gap activity from the damaged assembly is assumed to be released directly to the outside atmosphere at a constant rate over a 2-hour period. The licensee calculated the activity in the gap of the fuel rods assuming the assembly has been operated at 102 percent of the maximum core thermal power times a radial peaking factor of 1.65, and the fuel has undergone radioactive decay for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. The analysis assumed the RG 1.183 Table 3 non-loss-of-coolant-accident gap fractions. Although the licensee could not ensure that the fuel would meet the footnote 11 maximum linear heat generation rate limitation, the licensee chose to follow the footnotes alternative to calculate fission gas release fractions with an NRC-approved methodology. The licensee additionally performed an analysis to show that the RG 1.183 gap fractions are conservative compared to the Prairie Island specific analysis, and therefore, are acceptable.
To calculate the gap release fraction for alternate source term, the NRC staff considers acceptable the use of approved methodologies and bounding power histories. The NRC staff endorses the American National Standards Institute/American Nuclear Society-5.4-1982 model, entitled American National Standard Method for Calculating the Fractional Release of Volatile Fission Product from Oxide Fuel, as an acceptable gap fractional release model. Based on this model, the licensee developed a computer code, GAP, to perform the calculation.
The temperature and power input data for the GAP code are acquired using the approved Westinghouse PAD fuel performance code based on bounding power histories. The GAP code analyses include short-lived and long-lived radionuclides, and low-temperature and high-temperature releases. A bounding value is then chosen for the final release result. The NRC staff reviewed the GAP code and determined that the analytical approach is consistent
with the ANSI/ANS-5.4-1982 model. Therefore, the NRC staff concludes that the GAP code is acceptable for analyzing the gap release fraction.
In accordance with RG 1.183, the licensee assumed the iodine species released from the fuel gap to the water was 95 percent cesium iodide, 4.85 percent elemental, and 0.15 percent organic, and the effective iodine decontamination factor for the water pool was 200.
The licensee assumed that the control room heating, ventilation and air conditioning system was initially in the normal operation mode. Within a few seconds, the activity level in the control room would cause a high radiation signal, which in turn activates the isolation of the control room envelope and filtration of recirculated air. The licensees analysis assumed this isolation and recirculation occurred 2 minutes after the FHA. The licensee assumed 410 cfm of unfiltered inleakage into the control room envelope during the emergency mode of operation.
This assumption bounds the results of tracer gas testing of the control room envelope inleakage that was performed in January 1998.
On June 12, 2003, the NRC staff issued Generic Letter (GL) 2003-01, "Control Room Habitability." This GL identifies NRC staff concerns regarding the reliability of current surveillance testing to identify and quantify control room inleakage, and requests licensees to confirm the most limiting unfiltered inleakage into their control room envelope. NMC was requested by the GL to respond to the information request within 180 days of its issue. The Prairie Island response was submitted to the NRC by letter dated December 9, 2003, but the NRC staff has not completed review of the response. The NRC staff has determined that there is reasonable assurance that the Prairie Island control room will be habitable during an FHA with the proposed changes to containment closure TS, and this amendment may be approved prior to the NRC staff's review of the NMC response to the GL. The NRC staff bases this determination on the dose analyses provided and the verification of the control room unfiltered inleakage assumption through tracer gas testing. The NRC staff's approval of this amendment does not relieve NMC of addressing the information requests in GL 2003-01 and does not imply that the NRC staff would necessarily find the analysis in this amendment acceptable as a response to information request 1(a) in GL 2003-01.
Although the licensees analysis does not take credit for filtration by the SFP special ventilation system (SFPSVS) or the shield building ventilation system, the systems are not prevented from operating after an FHA. The licensees January 20, 2004, submittal did not evaluate a release from the SFPSVS or shield building ventilation system, but instead assumed a release from the common area of the auxiliary building. The NRC staff asked the licensee to verify that the submitted FHA dose analysis was bounding for a potential release from the SFPSVS or the shield building ventilation system.
By letter dated August 16, 2004, the licensee provided a discussion in which NMC concluded that the analysis assuming an unfiltered release from the common area of the auxiliary building would bound the dose results for a filtered released from the SFPSVS or shield building ventilation system. Both these systems exhaust to the atmosphere through the shield building vent stack. The licensee estimated the atmospheric dispersion factor (i/Q) to be less than 10 percent greater for the release from the shield building vent stack than was assumed in the FHA dose analysis, which would make the resulting dose greater by the same degree, assuming the same activity release in both cases. However, the release from the SFPSVS or shield building ventilation system would be filtered, and the lowest credited iodine filtration
efficiency for the filters (therefore allowing more iodine activity to be released) is 70 percent.
This corresponds to a reduction in the iodine release by a factor of 3.3, which also would reduce the resulting dose by the same factor. The reduction in the dose due to crediting the filtration system is greater than the increase in the dose due to the change in release location and control room i/Q.
The NRC staff finds the licensees discussion technically sound. The NRC staff additionally performed an independent confirmatory calculation of the shield building vent stack control room i/Q and found the licensees assumed control room i/Q for the shield building vent stack release was lower than the NRC staffs calculated value by approximately 9 percent. Using the NRC staffs calculated control room i/Q for the shield building vent stack release, the NRC staff also performed an independent confirmatory calculation of the control room dose for an FHA with credit for filtered release through the SFPSVS or shield building ventilation system, and found the doses were bounded by the case assuming an unfiltered release. Therefore, the NRC staff agrees that the licensees analysis as submitted is bounding for the FHA in either the SFP or containment. The licensees conclusion and the NRC staffs finding that the submitted FHA dose analysis is bounding for any release pathway is predicated on the SFPSVS and shield building ventilation system filters remaining operable for any release through those systems.
The licensee should evaluate the effect on the FHA dose analysis of any future proposed change to the credited filtration efficiencies for the SFPSVS and shield building ventilation system filters or any change in assumed operation of these systems. That evaluation should include an analysis of the dose in the control room from a release from the SFPSVS or shield building ventilation system.
3.2.1 Atmospheric Relative Concentration Estimates 3.2.1.1 Meteorological Data The licensee used 5 years of hourly onsite meteorological data collected during calendar years 1993 through 1997 to generate a new Control Room (CR) atmospheric dispersion factor (i/Q value) for analyzing the in-containment and fuel pool related design-basis fuel handling accident for this License Amendment Request (LAR). The resulting CR atmospheric dispersion factor represents a change from that value used in the current USAR analysis. The only new atmospheric dispersion factor generated for this LAR was for the CR; existing USAR i/Q values were used to evaluate doses for the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ).
The 1993 through 1997 onsite hourly meteorological data were provided for staff review in the form of hourly meteorological data files. All releases were considered to be ground level and, as such, were modeled using wind data measured at 10 meters. Wind measurements were also provided for the 60 meter level, and the atmospheric stability categorization was based on temperature difference measurements between these two levels. The licensee stated that the monitoring system location and instrument accuracies satisfied RG 1.23, Onsite Meteorological Programs, guidance. Redundant measurements were taken at both levels on the onsite meteorological tower which helped facilitate a data recovery rate of approximately 99 percent.
These data were then supplemented by measurements from a back-up tower when data were unavailable from the primary tower. Data were reviewed daily to identify potential irregularities
and weekly to check trends, consistency and that the data were reasonable. Site quality assurance measures include monthly tests and annual calibrations of instruments to check for problems and ensure data quality. If results indicate that the instruments are not working properly, corrective actions such as calibrations, repairs or replacement are made.
The NRC staff performed a quality review of the 1993 through 1997 onsite hourly meteorological data using the methodology described in NUREG-0917, Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data. Further review was performed using computer spreadsheets. With respect to atmospheric stability measurements, stable and neutral conditions were consistently reported to occur at night and unstable and neutral conditions during the day. The frequency, length, and time of occurrence of stable and unstable atmospheric conditions were very congruent with expected meteorological conditions.
Wind speed and direction frequency distributions for each measurement channel were also very consistent from year to year and when comparing measurements between the two heights.
In summary, NRC staff has reviewed the available information relative to the onsite meteorological measurements program provided by the licensee. On the basis of this review, the NRC staff concludes that these data provide an acceptable basis for making estimates of atmospheric dispersion estimates for design-basis accident assessments.
3.2.1.2 CR Atmospheric Dispersion Factors The licensee calculated CR air intake i/Q values using 1993 through 1997 onsite meteorological data and the ARCON96 atmospheric computer code (NUREG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wakes). Control room i/Q values were calculated for releases from the Common Area (CA) of the Auxiliary Building, the Spent Fuel Pool Normal Exhaust Stack, and the Units 1 and 2 Equipment Hatches to the 121 and 122 Control Room Intakes. The resulting limiting (highest) i/Q value, associated with a release from the CA, was utilized in the subsequent dose analysis and is presented in Table 1.
The NRC staff qualitatively reviewed the inputs to the ARCON96 computer runs and found them generally consistent with site configuration drawings and NRC staff practice. In addition, the licensee utilized calculations made for other release/receptor pairs as discussed below.
Specific areas of note are as follows:
C Postulated releases were modeled as point sources, other than those from the CA which were assumed to be diffuse. 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. The CA, which provides access to the containment maintenance airlocks for both units, encloses the spent fuel pools and transfer canal. The licensee stated that the CA is a sheet metal sided building that is not leak tight and is not serviced by a ventilation system nor does it have penetrations to the environment that could result in a more limiting release than that modeled as a diffuse release by the licensee. The licensees calculations were conservatively modeled from the closest point of the CA to the CR intakes although the fuel pools and transfer
canals are at a more distant location within the CA. Given these conditions, the NRC staff finds it acceptable to model the postulated release as a diffuse source.
C The licensee stated that effluent from the CA could also leak into the spent fuel enclosure to be exhausted by the SFPSVS or from the containment building to the shield building to be released through the shield building ventilation system, but concluded that the dose from such releases would be less limiting than that calculated as a diffuse release from the CA. The SFPSVS and shield building ventilation system contain charcoal absorbers that are both tested as part of the ventilation testing program. Based on the maximum allowed penetration, the calculated decontamination factor reduction was estimated to be a minimum of 3.3.
The licensee did not initially provide an estimate of the i/Q value for the SFPSVS release. After discussions with the NRC staff, the licensee provided an estimate based upon an interpolation from a plot of i/Q values calculated using ARCON96 for other postulated release locations. This estimate was less than 10 percent higher than the i/Q value for the CA diffuse release calculation. The NRC staff then calculated a i/Q value for comparison using ARCON96 with scenario-specific inputs that showed that the licensees value was about 9 percent lower, but still supported the licensees conclusion that the postulated release from the CA bounds the filtered release through the SFPSVS and shield building ventilation system.
In summary, the NRC staff has reviewed the licensees assessment of CR post-accident dispersion conditions generated from the licensees meteorological data and atmospheric dispersion modeling for the FHA. On the basis of this review, the NRC staff concludes that the CR i/Q value for a postulated release from the CA as presented in Table 1 is acceptable for use in the fuel handling accident dose assessment.
The NRC staff notes that if changes are made in facility design or operation that could impact potential releases from the SFPSVS and shield building ventilation system, the licensee should re-examine the FHA assessment. As noted above, the licensee did not calculate a scenario-specific i/Q value and dose for a release through the SFPSVS or shield building ventilation system in analyzing this design-basis accident scenario. The licensee would be expected to provide this calculation or justify its omission with respect to other design-basis accident assessments.
3.2.1.3 EAB/LPZ Atmospheric Dispersion Factors The licensee stated that the dose assessment for this LAR is based upon EAB and LPZ i/Q values from initial facility licensing which are based on guidance in RG 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, and RG 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors. As presented in these guides, the shortest time interval for which i/Q values are calculated is a 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period. Since the same standard Gaussian short-term centerline atmospheric dispersion equation applies and the licensee chose inputs and assumptions that are suitable for a 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period, the
resultant i/Q values can be applied to both a 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and a 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period.
Therefore, in this case, it is acceptable to use the values listed for the 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period in the dose estimate for the 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period. However, the NRC staff notes that if a different methodology or different assumptions were used, it may not be appropriate to apply i/Q values calculated for a 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period to a 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period. The EAB and LPZ i/Q values are presented in Table 1.
3.3 Radiological Consequences Analysis Conclusion The NRC staff finds that the licensees analysis of the FHA follows the guidance in RG 1.183.
The FHA assumptions and inputs the licensee used and The NRC staff evaluated are presented below in Table 1. The licensees analysis results show the radiological consequences of the FHA remain within the regulatory dose acceptance criteria within RG 1.183, both for persons offsite and operators in the control room. Table 2 presents the licensees analysis results.
The NRC staff evaluated the licensees analyses against RG 1.183 and found the methodology, inputs and assumptions to be in accordance with the guidance. The NRC staff independently calculated the activity source terms based on the licensees information and found the licensees source terms acceptable. The NRC staff performed independent dose analyses using the licensees assumptions, and confirmed the licensees results. Therefore, the NRC staff finds acceptable the licensees analysis of the impact of the proposed changes on the radiological consequences of the postulated FHA.
3.4 Summary As described above, the NRC staff reviewed the assumptions, inputs, and methods used by NMC to assess the radiological impacts of implementing a portion of TSTF-51 and selectively implementing an AST for the FHA at Prairie Island Units 1 and 2. The NRC staff finds that NMC used analysis methods and assumptions consistent with the regulatory requirements and guidance identified in Section 2.0 above. The NRC staff compared the doses estimated by NMC to the applicable criteria identified in Section 2.0. The staff finds, with reasonable assurance, that the licensees estimates of the EAB, LPZ, and control room doses will continue to comply with these criteria. Therefore, the proposed changes to TS 3.3.5 and 3.9.4 are acceptable with regard to the radiological consequences of postulated design-basis accidents.
This licensing action is considered a selective implementation of the AST. With this approval, the selected characteristics of the AST and the total effective dose equivalent (TEDE) criteria, become the design basis for the analysis of the FHA at Prairie Island Units 1 and 2. This approval is limited to this specific implementation. Subsequent modifications based on the selected characteristics of the AST incorporated into the Prairie Island Units 1 and 2 design basis by this action may be possible under the provisions of 10 CFR 50.59. However, the selected characteristics of the AST as described above and the TEDE criteria may not be extended to other aspects of the plant design or operation without prior NRC review under 10 CFR 50.67. All future radiological analyses performed to demonstrate compliance with regulatory requirements, and which are within the scope of the selective implementation, shall address the selected characteristics of the AST and the TEDE criteria as described in the Prairie Island design basis.
Table 1 Assumptions for Fuel Handling Accident Analysis Reactor power 1683 MWt Radial peaking factor 1.65 Fission product decay period 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> Number of fuel assemblies damaged 1 Fuel gap fission product inventory I-131 8%
Kr-85 10%
Other iodines and noble gases 5%
Pool iodine effective decontamination factor 200 Chemical form of iodine in pool CsI 95%
Elemental 4.85%
Organic 0.15%
Duration of release 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Control room volume 61,315 ft3 Normal ventilation unfiltered makeup 2000 cfm Emergency ventilation flow rates Filtered recirculation 3600 cfm Unfiltered inleakage 410 cfm Control room filter efficiencies Elemental 95%
Organic 95%
Particulate 99%
Atmospheric dispersion factors (i/Q), duration of accident EAB, 0 - 8 hr 6.49E-04 sec/m3 LPZ, 0 - 8 hr 1.77E-04 sec/m3 Control room,* 0 - 2 hr 4.19E-03 sec/m3
- Release from the common area of the auxiliary building
Table 2 FHA Licensee Calculated Radiological Consequences TEDE (rem)
Design-Basis Accident EAB LPZ Control Room Fuel Handling Accident 1.95 0.53 1.3 Dose acceptance criteria 6.3 6.3 5
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (69 FR 29769). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: M. Hart L. Brown H. Walker Date: September 10, 2004
Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esquire Tribal Council Vice President, Counsel & Secretary Prairie Island Indian Community Nuclear Management Company, LLC ATTN: Environmental Department 700 First Street 5636 Sturgeon Lake Road Hudson, WI 54016 Welch, MN 55089 Manager, Regulatory Affairs Nuclear Asset Manager Prairie Island Nuclear Generating Plant Xcel Energy, Inc.
Nuclear Management Company, LLC 414 Nicollet Mall, R.S. 8 1717 Wakonade Drive East Minneapolis, MN 55401 Welch, MN 55089 John Paul Cowan Manager - Environmental Protection Division Executive Vice President & Chief Nuclear Minnesota Attorney Generals Office Officer 445 Minnesota St., Suite 900 Nuclear Management Company, LLC St. Paul, MN 55101-2127 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office Craig G. Anderson 1719 Wakonade Drive East Senior Vice President, Group Operations Welch, MN 55089-9642 Nuclear Management Company, LLC 700 First Street Regional Administrator, Region III Hudson, WI 54016 U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 121 Seventh Place East Suite 200 St. Paul, MN 55101-2145 November 2003