ML090090307
| ML090090307 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/26/2009 |
| From: | Thomas Wengert Plant Licensing Branch III |
| To: | Wadley M Northern States Power Co |
| Wengert, Thomas J, NRR/DORL, 415-4037 | |
| References | |
| TAC MD9140, TAC MD9141 | |
| Download: ML090090307 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 26, 2009 Mr. Michael D. Wadley Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST FOR TECHNICAL SPECIFICATION CHANGES RELATED TO LICENSING BASIS RADIOLOGICAL DOSE CONSEQUENCES (TAC NOS. MD9140AND MD9141)
Dear Mr. Wadley:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated June 26,2008 (Agencywide Documents Access and Management System Accession No. ML081790439), Nuclear Management Company, LLC, a predecessor license holder to Northern States Power Company, a Minnesota corporation, submitted a request for technical specification changes related to the licensing basis loss-of-coolant accident and main steamline break accident radiological dose consequences for Prairie Island Nuclear Generating Plant, Units 1 and 2.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on January 12, 2009, it was agreed that you would provide a response within 45 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely,
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I In.a\\..~o- ~_J ~ V Thomas J. Wengert, Senior Project Manager Plant Licensing Branch /11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-282 and 50-306
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ
REQUEST FOR ADDITIONAL INFORMATION PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306 In reviewing the Nuclear Management Company, LLC*, a predecessor license holder to the Northern States Power Company, a Minnesota corporation (the licensee), submittal dated June 26,2008 (Agencywide Documents Access and Management System Accession No. ML081790439), which requested technical specification (TS) changes related to the licensing basis loss-of-coolant accident (LOCA) and main steamline break (MSLB) accident radiological dose consequences for Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP), the U.S.
Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:
Meteorological Requests for Additional Information (RAts)
(1)
Regarding the June 26, 2008, license amendment request (LAR) to revise the dose consequences of the LOCA and MSLB accident, as well as affected TSs, for PINGP, modifications were made to the onsite atmospheric dispersion factors (Le., X/Q values) used in the subsequent dose analyses. Accordingly, Enclosure 2, "Offsite & Control Room Dose Consequences," states:
"... as part of NRC SER [Safety Evaluation Report] for License Amendment 166 and 156... the staff performed a quality review of the on-site hourly met data (1993 through 1997) and concluded that the data was an acceptable basis for making estimates of atmospheric dispersion for design basis accidents. The above NRC-approved on-site hourly met data was utilized to develop the ARCON96 on-site atmospheric dispersion factors used in the LOCA and MSLB dose consequence analyses."
- a.
For new or updated X/Q values which were not already specifically approved, provide the input files (electronic files for data input into the ARCON96 computer code) and a discussion of the assumptions used to generate the X/Q values, summary output files, and/or cite references where this information has been previously docketed. These input files should clearly indicate the release height, receptor height, distance, and direction (with respect to true north) for each release/receptor pair analyzed.
- b.
Provide figures which support the selection of the inputs and assumptions used to calculate all of the onslte X/Q values. Include a figure of the general arrangement of plant structures, drawn approximately to scale and showing true north, sufficient to enable the NRC staff to make confirmatory estimates of the selected inputs and assumptions and resultant X/Q values for both the LOCA and MSLB accident. For each accident, highlight the postulated release and receptor
- On September 22, 2008, NuclearManagement Company, LLC (NMC), transferred its operating authority to Northern StatesPower Company, a Minnesota Corporation (NSPM). By letterdated September 3, 2008, NSPM stated that it would assume responsibility for actions and commitments submitted by I\\JMC.
ENCLOSURE
-2 locations including control room locations that may experience unfiltered inleakage.
- c.
Were the distance inputs into the ARCON96 calculations directly estimated as horizontal straight line distances or was another methodology (e.g., a "taut string" methodology) used to estimate the distances? If the distances were not estimated directly as the straight line horizontal distance, how were they determined? Did the procedure used to estimate the distances properly factor in differences in heights between source and receptor?
(2)
Regarding the LOCA and MSLB accident reanalyzed in support of this proposed amendment, please confirm that the generated X/Q values model the limiting doses and all potential release scenarios were considered, including those due to loss of offsite power or other single failures.
Dose Consequence Analysis RAls (3)
On page 15 of Enclosure 2 to the June 26, 2008, LAR submittal, the licensee stated that PINGP control room inleakage tracer gas testing resulted in a measurement of 165 cubic feet per minute (cfm). A total control room unfiltered inleakage value of 175 cfm is thereby assumed in the dose consequence analysis, including an additional assumed 10 cfm of unfiltered inleakage to account for ingress/egress. Typically, there is a quantifiable uncertainty associated with tracer gas testing. The licensee's unfiltered inleakage assumption of 175 cfm does not appear to include such an allowance for measurement uncertainty.
Therefore, if 165 cfm was the inleakage measured by the PINGP tracer gas test, please either provide the uncertainty associated with this measurement and quantify its effect on control room dose consequence, or provide technical justification for not including the uncertainty in the analysis.
(4)
Please explain where "Containment Leakage that is collected in the annulus (Shield Building)" is released. Is this leakage at any time, prior to equilibrium exhaust flow being achieved and maintained, assumed to be released from the Shield Building Stack? If so, please explain why this assumption is acceptable.
(5)
On pages 17 and 18 of Enclosure 2 to the LAR submittal, the licensee explained that, at various times prior to 20 minutes after accident initiation, when equilibrium exhaust flow is achieved and maintained, both filtered and unfiltered containment leakage releases are assumed to be taking place.
Please explain how the dose from filtered releases to the control room intake is accounted for during this time period.
(6)
Please explain how unfiltered inleakage into the control room is modeled before the assumed time of control room isolation at 2 minutes. Is it assumed to be in addition to the normal unfiltered intake flow? If it is not assessed, please explain why ignoring unfiltered inleakage during this pre-isolation time period is acceptable.
-3 Containment and Ventilation Systems RAls (7)
The LAR proposes to remove TS 3.3.5, "Containment Ventilation Isolation Instrumentation" which provides requirements for instrumentation to close the Containment Inservice Purge System (CIPS) isolation valves, in its entirety.
- a.
Please confirm that pages 3.3.5-2, -3, and -4 will also be removed from the TS.
- b.
In Enclosure 1, page 11 of 21 of the LAR, the licensee states that, with the blind flanges installed, the CIPS isolation valves are no longer credited for isolation of CIPS following an accident in containment and therefore the isolation instrumentation is no longer required. What changes, if any, are being implemented in how these valves presently receive and operate in response to this instrumentation, e.g. physical modifications such as signal and wiring disconnect from the safety related portions of the isolation system?
- c.
In Enclosure 1, page 11 of 21 of the licensee states that CIPS may be operated in Modes 5 and 6. Will the manual initiation functions from the control room to operate the CIPS isolation valves be retained so they can be used during this mode of operation?
- d.
In Enclosure 1, page 11 of 21 of the LAR, the licensee states that the CIPS penetrations will continue to be isolated during movement of recently irradiated fuel in accordance with the provisions of TS 3.9.4, "Containment Penetrations."
Please clarify whether the blind flanges or the CIPS isolation valves will provide this isolation function.
(8)
Please confirm that the changes to the current design basis input used in the LOCA and MSLB dose analyses will not impact the post-accident containment pressures and temperatures.
(9)
Are the spool pieces and blind flanges provided for the CIPS penetrations located inside or outside containment?
(10)
In Enclosure 2, page 5 of 36 of the LAR, it is stated that the Auxiliary Building drawdown time prior to crediting the Auxiliary Building special ventilation system is 6 minutes.
Please confirm whether this value supported by periodic drawdown tests.
(11), page 5 of 36 of the LAR, identifies changes to the current design basis inputs for spray initiation time (42 seconds) and the control room isolation time (2 minutes). Please identify the basis for these changes.
Mr. Michael D. Wadley January 26, 2009 Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST FOR TECHNICAL SPECIFICATION CHANGES RELATED TO LICENSING BASIS RADIOLOGICAL DOSE CONSEQUENCES (TAC NOS. MD9140 AND MD9141)
Dear Mr. Wadley:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated June 26, 2008 (Agencywide Documents Access and Management System Accession No. ML081790439), Nuclear Management Company, LLC, a predecessor license holder to Northern States Power Company, a Minnesota corporation, submitted a request for technical specification changes related to the licensing basis loss-of-coolant accident and main steamline break accident radiological dose consequences for Prairie Island Nuclear Generating Plant, Units 1 and 2.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on January 12, 2009, it was agreed that you would provide a response within 45 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely, IRAJ Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-282 and 50-306
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ N. Karipineni, NRR A. Boatright, NRR DISTRIBUTION:
PUBLIC LPL3-1 RlF RidsNrrPMPrairielsland Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl3-1 Resource RidsNrrLATHarris Resource RidsOgcRp Resource RidsNrrDorlDpr Resource RidsNrrDraAadb Resource RidsNrrDssScvb Resource ADAMS Accession Number' ML090090307 OFFICE LPL3-1/PM LPL3-1/LA NRRlDRA/AADB/BC NRRlDSS/SCVB/BC LPL3-1/BC NAME TWengert THarris RTaylor*
RDennig**
LJames DATE 01/14/09 01/13/09 12/18/08 12/17/08 01/26/09
- via memo dated 12/18/2008 OFFICIAL RECORD COPY
- via memo dated 12/17/2008