NRC 2002-0080, Response to Request for Additional Information License Amendment Request 223 Technical Specification LCO 3.6.4, Containment Pressure
| ML022680142 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/12/2002 |
| From: | Cayia A Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2002-0080 | |
| Download: ML022680142 (6) | |
Text
Nuclear Management Company, LLC MCPoint Beach Nuclear Plant Committed to Nuclear Ellence 6610 Nuclear Road Two Rivers, WI 54241 NRC 2002-0080 10 CFR 50.90 September 12, 2002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Dockets 50-266 and 50-301 Point Beach Nuclear Plant, Units 1 and 2 Response to Request for Additional Information License Amendment Request 223 Technical Specification LCO 3.6.4, Containment Pressure By submittal dated January 11, 2002, Nuclear Management Company, LLC (NMC) requested an amendment to the Technical Specifications (TS) for Point Beach Nuclear Plant, Units 1 and 2. The proposed amendment would revise TS 3.6.4, Containment Pressure, to reduce the maximum allowable pressure from three psig to two psig. In a letter dated August 19, 2002, NMC provided a response to a request for additional information from NRC staff regarding the original submittal.
During a conference call between NMC representatives and NRC staff on September 4, 2002, NRC staff requested additional information regarding certain aspects of the submittal.
Attachment I of this letter provides the NMC response to the staff's questions. The information provided does not require any changes to the proposed Technical Specifications nor their Bases.
We have determined that this additional information for the proposed amendments does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure. Therefore, we conclude that the proposed amendments meet the categorical exclusion requirements of 10 CFR 51.22(c)(9) and that an environmental impact appraisal need not be prepared.
NMC requests approval of the proposed License Amendment by October 2002, with the amendment being implemented within 45 days. This proposed amendment supports implementation of proposed License Amendment 226, Calorimetric Power Uprate.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Wisconsin Official.
AOD(
NRC 2002-0080 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on September 12, 2002.
J.C Site e Pres' nt JG/kmd
Attachment:
1 - Response to Request for Additional Information cc: NRC Regional Administrator NRC Project Manager NRC Resident Inspector PSCW
NRC 2002-0080 Page 1 of 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 223 TECHNICAL SPECIFICATION LCO 3.6.4, CONTAINMENT PRESSURE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2
NRC 2002-0080 Page 2 of 4 The following information is provided in response to the Nuclear Regulatory Commission staff's request for additional information (RAI) on NMC's January 11, 2002 License Amendment Request (LAR), as discussed during a telephone conference between NRC and NMC staff on September 4, 2002. The LAR proposes to modify Technical Specification (TS) 3.6.4, Containment Pressure, to reduce the maximum allowable pressure from three psig to two psig.
The NRC staff's questions are restated below, with the NMC response following.
NRC Question 1:
Please explain why the current main steam line break core analysis remains valid if the single failure of a feedwater regulation valve was not considered.
NMC Response:
The effect of a feedwater regulating valve failure on the main steamline break core response was described in PBNP Licensee Event Report LER 88-008-00, "Steam Line Break with Continued Feedwater Addition", dated August 12, 1988. The single failure of a feedwater regulating valve was considered and the current main steam line break core analysis remains valid.
NRC Question 2:
The containment volume is given as lx106 ft3 . Is there any conservatism in this value?
NMC Response:
Yes. The containment free volume was recently recalculated in support of the MAAP 5 containment response analysis effort. An in-house calculation note was prepared to determine the free volume in containment. The resulting value is slightly larger than lx106 cubic feet.
Several methods were used to include some conservatism in the calculation. The approach used to estimate the free volume modeled many of the structures and components as solid displacement volumes, even though they may be hollow or otherwise available for volumetric expansion of steam. Component insulation on components was assumed to be part of the displacing volume, even though some types of insulation are essentially empty volume available for steam expansion (e.g., reflective foil). This methodology is believed to provide an accurate, yet reasonably conservative, estimation of the containment net free volume.
NRC 2002-0080 Page 3 of 4 NRC Question 3:
How significant is the conservative assumption that only one train of ECCS is included in the calculation?
NMC Response:
There are two aspects to the assumptions used for the ECCS injection; 1) one train of ECCS is assumed, which reduces the rate of borated water delivered to the core; and, 2) an assumed boron concentration of 0 ppm in the SI piping increases the time delay before boron reaches the core.
It is estimated that if both trains of SI were credited, the effect would be an additional 100 to 200 pcm (0.1% Ak/k to 0.2% Ak/k). The additional negative reactivity would delay any return to criticality, but is judged to have a small reduction on the core heat flux, which would slightly lower the break flowrate and peak containment pressure.
In addition, a sensitivity was performed that demonstrates that if the SI piping were assumed to contain 2000 ppm of borated water, the core would remain approximately 500 pcm (0.5% Ak/k) more subcritical. Although a case can be made that the SI piping probably does contain a high concentration of boric acid (based on periodic testing of the SI pumps from the RWST, a minimum 2700 ppm source), it would likely require a surveillance to verify that concentration.
Therefore, it was conservatively assumed that the boric acid concentration was 0 ppm.
NRC Question 4" The results of TRANFLO calculations of steam generator blowdown, including entrainment were used in the calculations discussed in the January 11, 2002 letter. The statement is made (page 6 of 11) that the results of large double-ended breaks were found to be largely insensitive to the steam generator design. Please provide justification for this statement. Provide an example.
Appendix A to WCAP 8822, "Mass and Energy Release Following a Steam Line Rupture", page 5, lists steam generator design in a list of plant variables which affect the steam line break analysis. Page 12 states that the variations in steam generator design between the preheat and non-preheat steam generators were considered in the calculations discussed in Section 4.0 of WCAP 8822.
NRC 2002-0080 Page 4 of 4 NMC Response:
The justification for modeling of liquid entrainment in the break flow for Point Beach is based on the results of WCAP-8821, "TRANFLO Steam Generator Code Description" (Reference 7 of our January 11th submittal). It is noted on page 5-41 of WCAP-8821, that there is a lack of sensitivity of blowdown entrainment to the separator equipment model. This is attributed primarily to flow reversal through the separator drains during the blowdown for a full double ended rupture. It is this work, and the additional studies found in WCAP-8822, that form the basis for the conclusion that entrainment is largely insensitive to steam generator design.
As stated in your question, Appendix A to WCAP-8822 (page 5) does list steam generator design as a plant variable which influences the significant parameters affecting steam line break mass and energy releases. Furthermore, Appendix A to WCAP-8822 stated (page 12) that the effects of variations in steam generator design on mass and energy releases have been considered in the analysis, and are conservative for both of the steam generator models analyzed (Models D and 51). These statements indicate that steam generator design was considered in the mass and energy release analysis. More specifically, the design features related to steam separation equipment and feedwater inlet locations were the main features of interest.
The effect of the steam generator design on the liquid entrainment aspects of the WCAP-8822 analysis is discussed in Section 2.2.1. It can be seen in Figures 2.2.1-1 and 2.2.1-2 that the effluent quality is very similar for the two steam generator designs with the exception of the peak that occurs briefly at about 15 seconds (for a duration of 5 seconds) for the full power case. The peak is attributed to the difference in location of feedwater addition in the two types of steam generators. The conclusion is that in the steam generator with a preheater (i.e., Model D), low quality fluid reaches the steam nozzles earlier than in non-preheater steam generators (i.e.,
Model 51), but is of a lesser quantity. It is concluded that the blowdown from a preheat type steam generator is of higher quality (i.e., dryer), and therefore more conservative for containment calculations. The Point Beach steam generators are the non-preheater type.
For the Point Beach analysis, a table of break effluent quality versus time was generated that conservatively bounds the data from WCAP-8822. As stated in the January 11th submittal, the LOFTRAN model predicts less than 60% of the integrated liquid flowrate than that predicted by the TRANFLO code. Based on the above discussion, it is judged that for the Point Beach steam generators (Models 44F and A47), this effluent quality versus time is a reasonable and conservative estimation.