ML20206N533
| ML20206N533 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/12/1999 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | Sellman M WISCONSIN ELECTRIC POWER CO. |
| References | |
| TAC-MA1082, TAC-MA1083, NUDOCS 9905170258 | |
| Download: ML20206N533 (9) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.c. 30se6 cerH I$,
May 12, 1999 Mr. Michael B. Sellman Senior Vice President and Chief Nuclear Officer Wisconsin Electric Power Company 231 West Michigan Street Milwaukee,WI 53201 s
SUBJECT:
POINT BEACH NUCLEAR POWER PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RE: TECHNICAL SPECIFICATION CHANGE -
REQUEST 204 (TAC NOS. MA1082 AND MA1083)
Dear Mr. Sellman:
l By letter dated February 26,1998, and supplemented by oral presentation during a meeting held on June 4,1998, at Nuclear Regulatory Commission's headquarter offices, the Wisconsin Electric Power Company submitted a license amendment request for the Point Beach Plant Units 1 and 2 to revise Technical Specifications 15.3.12, " Control Room Emergency Filtration and Primary Auxiliary Building Exhaust Filtration," and 15.4.11, " Control Room Emergency Filtration and Primary Auxiliary Building Exhaust Filtration." The amendment was submitted to i
address issues with control room habitability and was required by a license condition issued with amendment no.174 on July 9,1997. Based upon our review of your submittal, the staff has developed the enclosed request for additional information (RAI).
The enclosed request was discussed with Mr. Jack Gadzala and other members of your staff during several conference calls, the most recent of which occurred on May 5,1999. A mutually agreeable target date of 60 days of the date of this letter, for your response was established, if circumstances result in the need to revise the target date, please contact me at (301) 415-1355 at the earliest opportunity.
9905170258 990512 PDR ADOCK 05000266 P
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Mr. Michael B. Sellman May 12, 1999 The NRC staff also discussed performance issues associated with this amendment during the May 5,1999, conference call. The technical content of your submittal was not complets and contained errors which necessitated numerous RAI questions and phone calls between the NRC staff and your technical staff to clarify the application.
Sincerely, Original signed by:
Beth A. Wetzel, Senior Pro,iect Manager, Section 1 Project Directorate Ill Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Request for Additional Information cc w/ enc!: See next page DISTRIBUTION Docket File '
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Mr. Michael B. Sellman May 12, 1999 The NRC staff also discussed performance issues associated with this amendment during the May 5,1999, conference call. The technical content of your submittal was not complete and j
contained errors which necessitated numerous RAI questions and phone calls between the NRC staff and your technical staff to clarify the application.
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Sincerely, Original signed by:
Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate ill Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Request for AdditionalInformation cc w/ encl: See next page QlSTFIBUTION:
Docket File PUBLIC PD3-1 Reading C. Thomas /G. Dick j
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ACRS G. Grant, Rill J. Zwolinski/S. Black j
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Mr. Michael B. Sellman
- i The NRC staff also discussed performance issues associated with this amendment during the May 5,1999, conference call. The technical content of your submittal was not complete and contained errors which necessitated numerous RAI questions and phone calls between the NRC staff and your technical staff to clarify the application.
l Sincerely,
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Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate lli
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Division of Licensing Project Management j
Office of Nuclear Reactor Regulation j
l Docket Nos. 50-266 and 50-301 l
Enclosure:
Request for AdditionalInformation cc w/ encl: See next page i
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Mr. Michael B. Sellman Point Beach Nuclear Plant Wisconsin Electric Power Company Units 1 and 2 cc:
Mr. John H. O'Neill, Jr.
Ms. Sarah Jenkins Shaw, Pittman, Potts & Trowbridge Electric Division 2300 N Street, NW Public Service Commission of Wisconsin Washington, DC 20037-1128 P.O. Box 7854 Madison, Wisconsin 53707-7854 Mr. Richard R. Grigg President and Chief Operating Officer Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201 Mr. Mark E. Reddemann Site Vice President Point Beach Nuclear Plant Wisansin Electric Power Company 6610 Nuclear Road Two Rivers, Wisconsin 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, Wisconsin 54228 Chairman Publ c Service Commission of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 Regional Administrator, Region lil U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Resident inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, Wisconsin 54241 o m ises I
REQUEST FOR ADDITIONAL INFORMATION Questions from Plant Systems Branch:
1.
As discussed during our conference call on March 4,1999, please provide design flow rates of fans W-14A and W-14B, and revise proposed TS 15.3.12.1.c to include these flow rates.
2.
With respect to the control room emergency filtration system, how do you maintain the ratio of makeup to recirculation flow to maintain 500 cfm i 10% makeup flow and 4450 cfm i10% recirculation flow?
3.
The staff is concemed that the unfiltered air provided to the mechanical equipment room will enter the control room envelope as unfiltered inleakage. Therefore, we are requesting that you identify the Wisconsin code goveming the replacement of the equipment room chillers discussed with the Nuclear Regulatory Commission's (NRC's) staff in the meeting on June 4,1998, in addition, provide the name and telephone number of a Wisconsin official familiar with the identified code.
4.
The staff expects that the primary auxiliary building filtration system will maintain the primary auxiliary building at a negative pressure with respect to adjacent areas (typically negative 1/8-inch water gauge (WG)) to ensure that all effluents leaving the primary auxiliary building are filtered. If the primary auxiliary building will be maintained at a negative pressure other than 1/8-inch WG during accident conditions, identify the negative pressure at which the primary auxiliary building will be maintained and discuss the acceptability of that pressure.
5.
The proposed change to delete the requirement contained in TS 15.4.11.1 to verify the pressure drop across the combined HEPA and charcoal filters for the control room filtration system is unacceptable. The purpose of verifying the pressure drop across the filters is to demonstrate that the filters are sufficiently clean to permit the requisite air flow to the control room and ensure that the filtration system remains capable of functioning as required. However, Wisconsin Electric (WE) may propose an attemative that corresponds to the maximum pressure drop allowed by the system while maintaining the required TS flow rate through the filters. In addition, a similar surveillance requirement should be included in TS 15.4.11.4 for the primary auxiliary building filtration system.
6.
In your proposed TS amendment you have committed to ASTM D3803-1989 for laboratory testing of a representative charcoal sample. However, the statements that
" velocities within 20% of design" in TS 15.4.11.3.d and 15.4.11.4.c are not consistent with the tolerances permitted in Table 1 of ASTM D3803-1989. Therefore, these statements should be removed. In addition, as discussed in our conference call on March 4,1999, these TSs should be revised to state that the test shall be performed in accordance with ASTM D3803-1989.
ENCLOSURE
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7.
The proposed change for the control room in TS 15.4.11.4.e should be revised to include a statement that a flow test without verification of pressurization shall be performed following maintenance, in accordance with Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Post-accident Engineered-Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."
8.
As discussed during our conference call on March 4,1999, if the primary auxiliary building has HEPA filters, revise TS 15.4.11.4 to include bypass test requirements for the HEPA filters.
9.
In accordance with Regulatory Guide 1.52, the staff acceptance criteria for in-place charcoal filter bypass testing is 0.05% or less bypass leakage through the activated carbon adsorber section. Proposed TS 15.3.12.C.1 Indicates 1% bypass leakage through the charcoal adsorber banks for the primary auxiliary building. The NRC staff finds this unacceptabie.
Questions from Probabilistic Safety Assessment Branch:
General Questions:
1.
Provide a copy of the specifics of the methodology, assumptions and inputs used to calculate the control room X/Q values for the postulated release point / intake pairs.
2.
Briefly describe how the source terms for the core inventory were determined. Please list bum-up, fuel enrichment, and other assumptions utilized to determine the core inventory. Also please state any computer codes utilized.
3.
The locked rotor, main steam line break, and steam generator tube rupture accidents utilize a pre-accident iodine spike of 48 micro Cl/g of DEI 131. The Technical Specifications should reflect this change. Markups of the Technical Specifications would help this review.
4.
The secondary volume utilized in the dose calculations is 7.18E+7 grams. The va!ue utilized in the WE Technical Specifications is 6.225E+7 grams. Please explain the reason for the use of the number 718E+7 grams and why it is different from the number glwn on page 15.3.4-3 of the WE Technical specifications (Unit 1, Amendment 173 and Unit 2 Amendment 177).
5.
How is the Technical Specification primary to secondary leakage limit determined? Is the measured value determined at room temperature or at normal operating pressure and temperature? The maximum allowable volumetric leak rate in your dose calculations is calculated at operating temperature. Typically, leakage measurements are made at room temperature. Please verify that the methodology utilized in your operational surveillances is consistent with the methods utilized in your rod ejection, locked rotor, steam generator tube rupture and main steam line break dose calculations.
For additional information concoming an instance when this was not considered, see
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Information Notice 97-79. If these surveillances are performed at room temperature, the j
calculations should reflect this or please provide adequate justification for your modeling.
6.
For several of the accidents the switchover time (the time at which the control room is assumed to actuate from normal to emergency mode) is delayed. Please specify the j
sources of the signals or operator actions utilized to perform this initiation of the emergency mode makeup. It would be beneficial to put these values into the Safety Analysis Report tables which describe these analyses.
7.
Provide the bases for stopping the coincident lodine spike release at a time which is less than the time at which the steaming release is completed. For example, in the Steam Generator Tube Rupture accident, the release is cut off at 1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but the accident release (steaming) does not end until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. For all accidents which include the coincident iodine spike, the releases should be for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (the duration of the accident) unless adequate justification is provided.
Loss-of-Coolant Accident Questions:
1.
Please explain why the control room whole body dose does not seem to include the dose due to an extemal cloud of radiation. The control room whole body dose on page 44 of of the submittal states the whole body dose as 0.7 rem. Page 41 of the same attachment states that the dose due to an extemal cloud of radiation is 1.31 rem.
2.
On page 41 of Attachment 2 the calculation for the control room dose due to extemal cloud needs to be explained in more detail. Please provide Wisconsin Electric Calculation 97-0115, " Point Beach Nuclear Plant Control Room Intemal Dose Rates Due to Extemal Cloud."
3.
The removal coefficient for particulate plate out is given in Table 14, page 43. The value given is 5.57/hr. The WE calculation utilizes 6.08/hr. Please clarify which value is to be used.
4.
In Table 14 the lodine chemical species for methyl and particulate iodine are not consistent with Standard Review Plan, Section 15.6.5. Please make them consistent or provide adequate justification for this deviation.
5.
The WE offsite doses due to Emergency Core Cooling System (ECCS) equipment leakage do not seem to be consistent with the assumption that no primary auxiliary building exhaust filtration is assumed for the offsite dose analyses. Provide adequate justification for this deviation or adjust the model utilized in the calculation to remove credit for this filtration.
6.
The WE control room doses for ECCS equipment leakage do not appear to adequately model the 1% bypass leakage of proposed TS 15.3.12.C.1. Please justify the doses predicted by the model or adjust this calculation accordingly to be consistent with the proposed TS. It would be benefic!al to include this adjustment in the description of filter efficiencies in the Safety Analysis Report.
4 Steam Generator Tube Rupture Questions:
7.
The value given in Table 8 of the submittal for steam release from the steam generators to the environment appears to be incorrect. The post-trip intact steam generator mass j
release is given as 2.326E+6 lbm (trip-2 hrs.). Should this number be 2.326E+5 lbms?
8.
The post-trip intact steam generator mass release is given as 593,000 lbm (2-8 hrs.).
' The value utilized in Westinghouse Calculation CN-CRA-98-010, Revision 0, page 17 is 5.29E+5 lbm for 2-8 hours. Please confirm which numbers were utilized in the j
calculation.
j 9.
The break flow rate to the ruptured steam generator utilized in Calculation CN-CRA-98-010, Revision 0, page 17 is 60.5 lbm/sec (trip-30 min). The value reported in Table 8 is 55 lbm/sec. Please confirm the correct value.
10.
Table 9, page 29 gives the control room and LPZ dose for the SGTR at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. WE needs to insure that the value given is for 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> or 30 days.
Main Steam Line Break Question:
j 1.
The WE doses due to the release of radioactivity from the ruptured loop do not seem to adequately model the release of all radioactivity in the steam generator of the ruptured.
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loop. It appears that these doses correlate to approximately 63% of the initial activity in this steam generator. The steam generator is assumed to boil dry in 15 minutes. Please adequately justify the reasons for this model and assumption or adjust the current model and assumption. Please note that typically this release is modeled as an instantaneous release of all the steam generator activity, and is typically treated as a puff or instantaneous release due to uncertainties in the timing of the accident scenarios.
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