ML053560245
| ML053560245 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/23/2005 |
| From: | Milano P NRC/NRR/ADRO/DORL/LPLA |
| To: | |
| Milano P, NRR/DLPM 415-1457 | |
| References | |
| TAC MC7382 | |
| Download: ML053560245 (6) | |
Text
December 23, 2005 LICENSEE:
R.E. Ginna Nuclear Power Plant, LLC FACILITY:
R.E. Ginna Nuclear Power Plant
SUBJECT:
SUMMARY
OF NOVEMBER 17, 2005, MEETING WITH R.E. GINNA NUCLEAR POWER PLANT, LLC, REGARDING EXTENDED POWER UPRATE APPLICATION (TAC NO. MC7382)
On November 17, 2005, the Nuclear Regulatory Commission (NRC) staff met with representatives of R.E. Ginna Nuclear Power Plant, LLC (the licensee) and its contractor, Westinghouse Electric Company, in a Category 1 public meeting held at the NRC offices at 11555 Rockville Pike, Rockville, Maryland. The purpose of the meeting was to discuss information regarding the licensees application dated July 7, 2005, for a 16.8 percent increase in the maximum steady-state thermal power level at the R.E. Ginna Nuclear Power Plant (Ginna). This level of power increase is generally referred to as an extended power uprate (EPU). A list of attendees is provided as Enclosure 1.
The agenda for this meeting consisted of discussions regarding the loss-of-coolant accident (LOCA), long-term cooling, and boron precipitation analyses supporting the EPU application. In particular, the discussion centered on issues presented in the NRC staffs requests for additional information on October 25 and 28, 2005, and issues that the NRC staff identified during its November 1-3, 2005, audit of supporting calculations. The presentation handout slides are provided as Enclosure 2.
For the LOCA analyses, information was presented regarding the analyses and modeling that were conducted. Because Ginna utilizes upper plenum injection, Westinghouse stated that the large-break LOCA analysis is best addressed by modeling a hot-leg break. In this regard, recirculation flow from the containment sump may begin in about 24 minutes. Upon questioning from the NRC staff, the licensee indicated that no credit was taken for containment overpressure for pump net-positive suction head. The licensee agreed to provide COBRA/TRAC computer code results to better understand liquid mixing volumes and boil-off.
The licensee also stated that it would be conducting simulator and operator training within several months of the outage to validate operator performance times for such aspects as restoration of cold-leg injection.
In the small-break LOCA analysis, the concern is for a cold-leg break. Westinghouse indicated that the mixing volumes from the NOTRUMP analysis were used for the boric acid precipitation analysis and solubility was based at atmospheric conditions. Break sizes from 1.1 to 4 inches were evaluated to determine reactor coolant system (RCS) depressurization and the resultant boron precipitation times, the results of which were used to test operator actions. The licensee presented a draft plan for responding to the concerns about post-LOCA long-term cooling (Enclosure 2). The NRC staff stated that it would be looking at the guidelines and time limits for initiating depressurization as a job performance measure. The staff also questioned the need to include the operator action to depressurize the RCS in the emergency operating procedures (EOPs). The concern is that, without an explicit action in the EOPs, there may be no assurance that the 1-hour time to initiate cooldown in the analysis will always be achieved. In particular, the staff asked about the licensees plans for training on postulated bottom-mounted instrument tube breaks. Further, the NRC staff asked about the testing of the service water supply to auxiliary feedwater.
The NRC staff discussed its observations and issues identified during an audit of the calculations supporting the EPU application. The licensee agreed to answer the following questions:
1.
With respect to the rod withdrawal at power analysis (Section 2.8.5.4.2 of the Licensing Report):
a.
Figure 2.8.5.4.2-11, which depicts pressurizer pressure versus time for the RCS pressure case, indicates an uninterrupted rise in pressurizer pressure, to almost 2700 psia, until the reactor is tripped. It appears that the pressurizer safety valves are not modeled, and this seems to be an overly conservative assumption. Explain the modeling.
b.
Results of this event are presented as an assembly of the minimum departure from nucleate boiling ratio (DNBR) results of many transient analysis cases, mapping the minimum DNBRs as a function of initial power level, core reactivity feedback, and reactivity insertion rate. From this, it is concluded that the automatic reactor protection system prevents fuel clad damage for all analyzed cases. In a similar fashion, analyze or evaluate the case(s) of operation at low power level with only one loop in operation.
2.
With respect to the steam system piping failures analysis (Section 2.8.5.1.2), describe how the shutdown margin is determined and used in the single-loop in operation case.
3.
With respect to the spurious safety injection and chemical and volume control system (CVCS) malfunction event (Section 2.8.5.5), verify that the operator, following prescribed procedures, can terminate the CVCS malfunction event before the pressurizer fills (about 13 minutes).
4.
With respect to the boron dilution event (Section 2.8.5.4.5):
a.
Identify the Ginna licensing basis for the boron dilution event for all modes.
Indicate (1) whether the time allotted for operator action begins with operator notification or not, and (2) which operating modes must be analyzed.
b.
Explain why the parameters used in the boron dilution calculations are conservative.
c.
Since some sections of the boron dilution submittal are written in past tense, verify that the acceptance criteria are currently satisfied.
5.
With respect to the functional design of the control rod system (Section 2.8.4.1), the licensee stated that [c]hanges to the rod position indication systems... are currently being assessed to ensure that correct individual rod position indications are available to the operator. Provide the results of that assessment. Provide the rod drop time analysis.
- 6.
With respect to the fuel system design (Section 2.8.1) and nuclear design (Section 2.8.2):
a.
The Fuel Criteria Evaluation Process (FCEP) for the 9 grid 422V+ fuel does not provide the overall assembly flow loss coefficient, nor an explanation of how it is determined. Please provide that information, either in the FCEP or in the Ginna submittal.
b.
The analysis for the transition cycles indicates that the current fuel (Optimized Fuel Assembly) could possibly exceed 60 gigawatt-day per metric ton of uranium (GWD/MTU) during the first transition cycle. Since this does not represent an actual core loading, it is not necessary, at this time, to submit any information to support exceeding 60 GWD/MTU. The licensee agreed to provide this information in the future, in the event that an actual core loading is predicted to exceed 60 GWD/MTU.
c.
The issue and commitment in Item (b), above, also apply to the 422V+ fuel assemblies in the first equilibrium cycle.
During the meeting, the licensee also discussed the proposed power ascension and testing plans associated with the implementation of the EPU. The NRC staff discussed and clarified its prior concerns on large transient testing.
Members of the public were not in attendance. Public meeting feedback forms were not received.
Please direct any inquiries to me at 301-415-1457 or pdm@nrc.gov.
/RA/
Patrick Milano, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. List of Attendees
- 2. Licensee Handout cc w/encls: See next page the operator. Provide the results of that assessment. Provide the rod drop time analysis.
- 6.
With respect to the fuel system design (Section 2.8.1) and nuclear design (Section 2.8.2):
a.
The Fuel Criteria Evaluation Process (FCEP) for the 9 grid 422V+ fuel does not provide the overall assembly flow loss coefficient, nor an explanation of how it is determined. Please provide that information, either in the FCEP or in the Ginna submittal.
b.
The analysis for the transition cycles indicates that the current fuel (Optimized Fuel Assembly) could possibly exceed 60 gigawatt-day per metric ton of uranium (GWD/MTU) during the first transition cycle. Since this does not represent an actual core loading, it is not necessary, at this time, to submit any information to support exceeding 60 GWD/MTU. The licensee agreed to provide this information in the future, in the event that an actual core loading is predicted to exceed 60 GWD/MTU.
c.
The issue and commitment in Item (b), above, also apply to the 422V+ fuel assemblies in the first equilibrium cycle.
During the meeting, the licensee also discussed the proposed power ascension and testing plans associated with the implementation of the EPU. The NRC staff discussed and clarified its prior concerns on large transient testing.
Members of the public were not in attendance. Public meeting feedback forms were not received.
Please direct any inquiries to me at 301-415-1457 or pdm@nrc.gov.
/RA/
Patrick Milano, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. List of Attendees
- 2. Licensee Handout cc w/encls: See next page DISTRIBUTION PUBLIC PDI-1 Reading File RidsNrrDorl RidsNrrDorlLpla RidsNrrDssSpwb RidsNrrPMPMilano RidsNrrPMJStang RidsNrrLASLittle S. Miranda L. Ward K. Wood P. Prescott, NRR J. Trapp, R-I RidsOgcMailCenter RidsAcrsAcnwMailCenter Accession No.: ML053560245 Package No.:
OFFICE PDI-1/PM DLPM/LPM PDI-1/LA SRXB/SC PDI-1/SC NAME PMilano JStang SLittle YHsii for JNakoski RLaufer DATE 12/23/05 12/22/05 12/22/05 12/23/05 12/23/05 OFFICIAL RECORD COPY
R.E. Ginna Nuclear Power Plant cc:
Mr. Michael J. Wallace President R.E. Ginna Nuclear Power Plant, LLC c/o Constellation Energy 750 East Pratt Street Baltimore, MD 21202 Mr. John M. Heffley Senior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401 Mrs. Mary G. Korsnick Vice President R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519 Kenneth Kolaczyk, Sr. Resident Inspector R.E. Ginna Nuclear Power Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Carey W. Fleming, Esquire Senior Counsel - Nuclear Generation Constellation Generation Group, LLC 750 East Pratt Street, 17th Floor Baltimore, MD 21202 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Ms. Thelma Wideman, Director Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 1190 Scottsville Road, Suite 200 Rochester, NY 14624 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 MEETING ATTENDEES R.E. GINNA NUCLEAR POWER PLANT, LLC R.E. GINNA NUCLEAR POWER PLANT NOVEMBER 17, 2005 NRC Staff:
Name Samuel Miranda Leonard Ward Kent Wood Patrick Milano John Stang Title Reactor Engineer Sr. Reactor Engineer Reactor Engineer Sr. Project Manager Sr. Project Manager Organization NRR/Division of Safety Systems (DSS)/PWR Systems Branch (SPWB)
NRR/DSS/Nuclear Performance and Code Review Branch NRR/DSS/SPWB NRR/Division of Operating Reactor Licensing (DORL)/Plant Licensing Branch I-1 (LPLA)
NRR/DORL/LPLC Constellation Energy/R.E. Ginna Nuclear Power Plant, LLC:
Name Mark Flaherty Mark Finley George Wrobel Title Manager, Fleet Licensing Manager, Power Uprates Manager, Licensing Westinghouse Electric Company:
Name Dave Dominicis Dave Sklarsky Patricia E. Vaughn Brett Kellerman Dawn Crytzer David Fink J. Hartz Andy Gagnon Jayashree Iyengar Title Project Manager Project Manager Project Manager Senior Engineer Senior Engineer Principal Engineer Principal Engineer Principal Engineer Contractor for Westinghouse Other
Participants:
None