ML052100254
| ML052100254 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/27/2005 |
| From: | Gregory Suber NRC/NRR/DRIP/RLEP |
| To: | Nuclear Management Co |
| Suber G, NRR/DRIP/RLEP 301-415-1124 | |
| References | |
| Download: ML052100254 (11) | |
Text
July 27, 2005 LICENSEE:
Nuclear Management Company, LLC FACILITY:
Point Beach Nuclear Plant, Units 1 and 2
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCES HELD ON JUNE 30, 2005 AND JULY 8, 2005, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND NUCLEAR MANAGEMENT COMPANY, LLC CONCERNING CONFIRMATORY ITEMS PERTAINING TO THE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2, SAFETY EVALUATION REPORT The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Nuclear Management Company, LLC (NMC) held telephone conferences on June 30, 2005 and July 8, 2005, to discuss and clarify the staffs confirmatory items (CIs) concerning the Point Beach Nuclear Plant, Units 1 and 2, safety evaluation report. The conference call was useful in clarifying the intent of the staffs concerns. provides a listing of the meeting participants. Enclosure 2 contains a listing of the CIs discussed with the applicant, including a brief description on the status of the items.
The applicant had an opportunity to comment on this summary.
/RA/ (V. Rodriguez for)
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket Nos.: 50-266 and 50-301
Enclosures:
As stated cc w/encls: See next page
ML052100254 DOCUMENT NAME: E:\\Filenet\\ML052100254.wpd OFFICE PM:RLEP LA:RLEP PM:RLEP SC:RLEP NAME GSuber (VMR for)
MJenkins VRodriguez SLee DATE 7/25/05 7/18/05 7/25/05 7/27/05
Point Beach Nuclear Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esq.
Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Mr. Frederick D. Kuester President and Chief Executive Officer We Generation 231 West Michigan Street Milwaukee, WI 53201 Aldo Capristo Manager, Regulatory Affairs Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, WI 54228 Chairman Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Resident Inspectors Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241 Mr. Jeffrey Kitsembel Electric Division Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854 David Weaver Nuclear Asset Manager Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, WI 53201 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Douglas E. Cooper Senior Vice President - Group Operations Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043 James Ross Nuclear Energy Institute 1776 I Street, NW., Suite 400 Washington, DC 20006-3708 Roger A. Newton 3623 Nagawicka Shores Drive Hartland, WI 53029 James E. Knorr License Renewal Project Manager Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Point Beach Nuclear Plant Two Rivers, WI 54241
DISTRIBUTION: Summary of Telecon with NMC concerning Pt. Beach, Dated: July 29, 2005 Adams accession no.: Ml052100254 HARD COPY RLEP RF G. Suber E-MAIL:
RidsNrrDrip RidsNrrDe G. Bagchi K. Manoly W. Bateman J. Calvo R. Jenkins J. Fair RidsNrrDssa RidsNrrDipm D. Thatcher R. Pettis G. Galletti C. Li M. Itzkowitz (RidsOgcMailCenter)
R. Weisman M. Mayfield A. Murphy S. Smith (srs3)
S. Duraiswamy Y. L. (Renee) Li RLEP Staff P. Lougheed, RIII J. Strasma, RIII A. Stone, RIII H. Chernoff C. Marco L. Raghavan T. Mensah OPA LIST OF PARTICIPANTS FOR BOTH TELEPHONE CONFERENCES TO DISCUSS THE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT JUNE 30, 2005 AND JULY 8, 2005 Participants Affiliations Devender Reddy U.S. Nuclear Regulatory Commission (NRC)
Chang-Yang Li NRC Gregory Suber NRC Jim Knorr Nuclear Management Company, LLC (NMC)
John Thorgersen NMC Todd Mielke NMC CONFIRMATORY ITEMS DISCUSSED DURING TELEPHONE CONFERENCES BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND NUCLEAR MANAGEMENT COMPANY, LLC FOR POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT JUNE 30, 2005 AND JULY 8, 2005 The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Nuclear Management Company, LLC (NMC) held telephone conference calls on June 30, 2005 and July 8, 2005, to discuss and clarify the applicants responses to the staffs confirmatory items (CI) concerning the Point Beach Nuclear Plant, Units 1 and 2, Safety Evaluation Report (SER).
The following CI was discussed during the telephone conference calls.
CI 2.1-1 Section 2.1.2.1.2 - Application of the Scoping Criteria in 10 CFR 54.4(a)
In RAI 2.1-1, dated November 16, 2004, the staff requested additional information regarding the scoping methodology associated with the 10 CFR 54.4(a)(2) evaluation. The staff requested the applicant to adequately define short term exposure duration as it relates to the evaluation of low and moderate energy piping failures that could affect safety-related electrical equipment.
Since this equipment may not be environmentally qualified, it could fail due to 10 CFR 54.4(a)(2) piping failures.
In its response, by letter dated January 31, 2005, the applicant stated that for the purpose of license renewal, the term exposure duration will be removed from LRA Section 2.1.2.1.2 and it will provide a technical justification as to why the safety-related SSCs are capable of withstanding the effects of spray and leakage. The applicant also stated that it will include a technical justification in the LRA annual update under the section Components Qualified/Designed for Environment.
During a meeting on February 15, 2005, the staff indicated and the applicant agreed, that this response required further clarification. In its response, a clarification letter dated March 15, 2005, the applicant committed to provide details of the 10 CFR 54.4(a)(2) scoping methodology changes, including specific exceptions, and how these will impact the LRA. The staff agreed with the applicants proposed methodology changes. However, the applicant committed to provide detailed information with regard to these changes by the end of April 2005.
In its response, by letter dated April 29, 2005, the applicant provided additional information on scoping and screening of non-safety-related (NSR) components. The portions of the response discussed during the telephone conference are as follows:
The original scoping process at PBNP for 10 CFR 54.4(a)(2) included evaluating potential effect on vulnerable safety-related (SR) components, assuming a finite spray/leakage duration from a failed non-safety-related (NSR) system or component (SC). The spray/leakage duration assumed that the spray/leak would be identified by normal work activities within the plant (walkdowns, system parameter changes, sump alarms and trends, etc.). Based on feedback from the staff, NMC is changing the scoping methodology to remove the assumption of leakage duration and to eliminate the limitation of vulnerable targets to only electrical components.
The revised methodology invokes a plant ?spaces approach that assumes a spatial interaction can potentially occur if SR and NSR SCs are located within the same space. For the purposes of this process, a space is defined by the room in which the SR and NSR components are located. Physical barriers (e.g., walls, ceilings, and floors) enclose the space. This revised methodology evaluates the effect of sprays and leaks on mechanical as well as electrical SR SCs, with no limitation on the duration of the sprays/leaks. NMC considers all liquid or steam bearing NSR SCs to be in the scope of the criterion 10 CFR 54.4(a)(2), provided the NSR SCs are located in the same space as a SR SC, and the NSR SCs are in proximity where spray or leakage from the NSR SCs could contact a SR SC.
NMC has re-evaluated the SC at PBNP using the revised 10 CFR 54.4(a)(2) methodology. A number of configurations have been identified where the failure of NSR SCs would not result in the loss of intended function of the SR SCs in the ?space. The exceptions to the revised methodology are as follows:
Exceptions 1.
NSR SCs in containment were not re-evaluated. SR SCs within containment are already evaluated for post-accident environments including spray and/or steam.
As such, the existing current licensing basis (CLB) has addressed the bounding environmental conditions for SR SCs within containment.
2.
NSR components in rooms or cubicles where there are no SR components do not need to be in-scope. These rooms or cubicles have also been evaluated to ensure that SR piping does not run through them.
The following cubicles or rooms have no SR equipment in them, and therefore NSR components in these cubicles or rooms are not in-scope:
Demineralizer Cubicles and Demineralizer Valve Gallery Sump Tank Cubicle Gas Stripper Building Blowdown Evaporator Building Drumming Area PAB Truck Bay
?B and ?C Hold Up Tank (HUT) Cubicles (See note on ?A HUT in Exception 5)
Laundry Tank Room 3.
Only NSR SCs containing liquid or steam are considered to pose any potential for spatial interaction. NSR SCs containing gases (e.g., plant air systems, ventilation systems) pose no potential for aging effects on SR SCs due to leakage of air or gas.
NSR portions of plant air systems (instrument air, service air) are not in-scope per this exception.
NSR portions of ventilation systems are not in-scope per this exception.
NSR portions of gas systems (nitrogen, hydrogen) are not in-scope per this exception.
NSR portions of systems that are normally vented, or connected to the vent header are not in-scope per this exception.
4.
Abandoned or manually isolated and drained NSR SCs are not considered to pose any potential for aging effects on SR SCs, and therefore do not need to be in-scope.
NSR glycol drain tank in G-03/G-04 Emergency Diesel Generator Rooms is manually isolated, vented, and drained when the engine is operational. Therefore, is not in-scope per this exception.
NSR portions of the Heating Steam system that are isolated and control tagged as abandoned are not in-scope per this exception.
- 5.
Spray is not postulated from unpressurized systems, however leakage still is a potential. Leakage can only affect SR SCs that are physically below the unpressurized NSR components. If SR components are above or beside the unpressurized NSR components, the NSR components would not need to be in-scope.
NSR chemical addition pots in the auxiliary feedwater pump rooms are normally vented and isolated and are mounted near the floor where they cannot leak on any SR equipment. Therefore, these chemical addition pots are not in-scope.
?A HUT Cubicle has some SR piping that passes through this cubicle and exits the ceiling to supply SR components in a cubicle above. NSR piping that is above or adjacent to this SR piping has been included in-scope, but the HUT tank itself is at very low pressure (normally 2.5 psig), and as such, it could not spray on the SR piping. Therefore, the ?A HUT is not in-scope.
6.
NSR SCs in large open areas (e.g., turbine building, facade) are eliminated from scope if it can be shown that there is no possible effect on the SR SCs.
The NSR Reactor Makeup Water (RMW) Tank is on the 6.5 elevation of the facade. This tank is not pressurized. This tank is approximately 1/3 the size of the Refueling Water Storage Tank (RWST), and therefore is bounded by the flooding analysis that assumed the RWST would fail. There are also two short runs of pressurized pipe (-5 long) where the pipe exits the adjacent RMW pump room and crosses into the Primary Auxiliary Building (PAB).
Intervening structures exist between these pipes and any SR equipment (nearest would be the containment penetrations on 26 elevation). Failure of any of these RMW components on the 6.5 elevation could not affect any SR equipment, and therefore, these NSR components are not in-scope.
SR Crossover Steam Dump components are located within the 66 fan room. SR equipment in the 66 fan room includes SG pressure transmitters and the main steam lines themselves (in the overhead). The transmitters are environmental qualification (EQ) qualified for harsh environment based on main steam line break potential, and this bounds the energy level of the crossover steam dump system by a significant margin. The crossover steam dump components are a minimum of 50 away from SR equipment and failure of these NSR components will not affect the function of these SR components. Therefore, the NSR crossover steam dump components would not be in-scope.
The SR Main Feedwater Regulating Valves (MFRVs), bypass valves, and associated solenoid operated valves (SOVs), are located on the 26 elevation of the Unit 1/Unit 2 Turbine Building.
The safety function is for the MFRVs and bypass valves to close. The SOVs are EQ qualified for harsh environments. The piping on either end of these valves is NSR. All other equipment on this elevation of the turbine building is NSR. The only potential failure that could cause a failure of the safety function of these components, would be a flow accelerated corrosion (FAC) failure where a pipe-whip impact could bend the actuator stem and prevent the valve from closing. Simple leakage or spray would not affect the safety function of these valves, as external aging effects would only create fail-safe failures of the SR valves (through-wall leakage would divert flow from SGs which is the fail-safe direction). Therefore, NSR high energy piping sections that pose the potential for pipe-whip on these valves and/or SOVs are included in-scope. Portions of high energy piping that cannot physically reach, or are shielded from, the SR components by structures or other larger piping, are not included in-scope. Major components such as feedwater heaters and the condenser are anchored and do not have the potential for pipe-whip, and therefore are also not included in-scope.
Discussion: The applicant stated that all NSR SCs containing liquid or steam were in scope
... provided the NSR SCs are located in the same space as a SR SC and the NSR SCs are in proximity where spray or leakage from the NSR SCs could contact a SR SC. The staff requested that the applicant clarify the use of the word ?proximity.
During the conversation, the applicant stated that proximity referred to the NSR SCs located in large open areas and included the exceptions listed in Section 6 of the response. No additional information is required at this time.
Discussion: For Exception 3, the applicant stated that... NSR portions of gas systems (nitrogen, hydrogen) are not in-scope per this exception. The staff requested that the applicant explain why leakage from external corrosion was not considered as a failure mechanism that could cause a possible hydrogen fire/explosion.
During the conversation, the applicant stated that industry experience did not identify external corrosion as an aging effect for dry gas systems. The staff requested that the applicant provide further clarifications regarding operating experience.
Discussion: For Exception 4, the applicant stated that abandoned or manually isolated and drained NSR SCs are not considered to pose any potential for aging effects on SR SCs, and therefore do not need to be in-scope. The staff requested that the applicant clarify whether the manually isolating component (e.g., valve) was in-scope, and identify the safety-related boundary for the manual isolation.
During the conversation, the applicant stated that the isolating component (valve, flange, etc.)
was in the scope of license renewal and represented the system boundary. The applicant agreed to revise the response and submit it formally.
Discussion: For Exception 6, the applicant stated that... The crossover steam dump components are a minimum of 50 away from SR components. Therefore, the NSR crossover steam dump components would not be in-scope. The staff requested that the applicant provide a technical justification for the 50 exclusion value and consider the effects of impingement.
During the conversation, the applicant stated that the steam dump components are low pressure, and non-safety-related components are within the jet impingement cone. The applicant agreed to revise the response and submit it formally.
Discussion: For Exception 6, the applicant also stated that... Portions of high-energy piping that can not physically reach, or are shielded from, the SR components by structures or other larger piping, are not included in-scope. The staff requested that the applicant to clarify physically reach, and provide qualification for the barriers.
During the conversation, the applicant stated that the phrase can not physically reach referred to pipe whip. The applicant agreed to provide additional information on jet impingement.