L-PI-09-025, Response to Request for Additional Information Regarding License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 0.422-Inch 14x14 Vantage+ Fuel
| ML090510691 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/20/2009 |
| From: | Wadley M Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-09-025, TAC MD9142, TAC MD9143 | |
| Download: ML090510691 (21) | |
Text
@ Xcel Energya L-PI-09-025 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Response To Request For Additional lnformation Regarding License Amendment Request For Technical Specifications Changes To Allow Use Of Westinqhouse 0.422-Inch 14x14 Vantage+ Fuel (TAC Nos. MD9142 and MD9143)
References:
- 1) Letter from M. Wadley (Nuclear Management Company) to Document Control Desk (NRC), L-PI-08-047, License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 0.422-inch OD 14x14 VANTAGE+ Fuel, dated June 26,2008 (ADAMS Accession No. ML081820137)
- 2) Letter from T. Wengert (NRC) to M. Wadley (Northern States Power - Minnesota), Prairie Island Nuclear Generating Plant, Units 1 and 2 Request For Additional lnformation Related to License Amendment Request For Technical Specifications Changes to Allow Use of Westinghouse 0.422-Inch OD 14x14 Vantage+ Fuel (TAC Nos. MD9142 and MD9143), dated December 15,2008 (ADAMS Accession No. ML083300210)
By letter dated June 26,2008 (Reference I), Nuclear Management Company, LLC, (now Northern States Power, a Minnesota corporation (NSPM))
requested approval of amendments to the Operating Licenses and associated Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, as well as certain supporting analyses, in support of the transition from Westinghouse 0.400-inch outer diameter (OD) VANTAGE+
(hereinafter referred to as 400V+) fuel to 0.422-inch OD VANTAGE+ (hereafter referred to as 422V+) fuel.
On December 15, 2008, the NRC staff notified NSPM (Reference 2) that additional information was necessary for the staff to complete its review. The enclosed response addresses Question 3. NSPM responded earlier to Questions 1 and 2 of the request for additional information in a letter dated January 30, 2009 (ADAMS Accession No. ML090300684).
171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
Document Control Desk Page 2 The supplemental information provided in this letter does not impact the conclusions of the Determination of No Significant Hazards Consideration and Environmental Assessment presented in the June 26,2008 submittal as supplemented by letters dated August 4,2008, August 26,2008, November 14,2008, January 12,2009, and January 30,2009.
In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of this License Amendment Request supplement by transmitting a copy of this letter and enclosure to the designated State Official.
Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on:
FFR 2 0 2009 Michael D. Wadley u
Site Vice-President Prairie Island Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota
ENCLOSURE Response to Request for Additional Information License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 422V+ Fuel Page 1 of 19 By letter dated June 26, 2008 (ADAMS Accession No. ML081820137), Nuclear Management Company, LLC, (now Northern States Power, a Minnesota corporation (NSPM)) requested approval of amendments to the Operating Licenses and associated Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, as well as certain supporting analyses, in support of the transition from Westinghouse 0.400-inch outer diameter (OD) VANTAGE+ (hereinafter referred to as 400V+) fuel to 0.422-inch OD VANTAGE+ (hereafter referred to as 422V+) fuel. On December 15, 2008 (ADAMS Accession No. ML083300210), the NRC staff notified NSPM that additional information was necessary for the staff to complete its review. The NRC request for additional information (RAI) is repeated below with the NSPM response following:
- 3.
The NRC staff reviewed documents during an October 1-2, 2008, audit at the Westinghouse Energy Center in support of the NRC staff's review of the requested fuel upgrade. During its audit, the NRC staff identified potentially non-conservative assumptions made regarding the [Prairie Island Nuclear Generating Plant (PINGP)] capability for post-[Loss of Coolant Accident (LOCA)], long-term core cooling. Please re-evaluate the post-LOCA, long-term core cooling at PINGP, and demonstrate acceptable safety injection capability. The re-evaluation should at least consider the following:
- a.
The acceptance criteria set forth in 10 CFR 50.46(b).
- b.
The 10 CFR Part 50, Appendix K, decay heat requirements.
- c.
Differences in lattice pitch over generically studied plants referenced in the analysis (Westinghouse Proprietary Calculation CN-LIS-07-126, "Prairie Island Units 1 and 2 (NSP/NRP) Post-LOCA Long-Term Cooling Analysis in Support of the 422V+ Fuel Transition Program").
- d.
Differences in post-LOCA decay power shape compared to the power shape evaluated in CN-LlS-07-126 (Westinghouse Proprietary) and the effect that this difference could have on core boiloff.
NSPM Response:
Acceptance Criterion The relevant acceptance criterion for demonstration of acceptable safety injection capability for post-LOCA, long-term cooling is 10 CFR 50.46(b)(5) which states:
Long-term cooling After any calculated successful initial operation of the
[emergency core cooling system (ECCS)], the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
ENCLOSURE Response to Request for Additional Information License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 422V+ Fuel Page 2 of 19 The acceptance criterion lacks specificity in defining long-term cooling and what an acceptably low core temperature value is. Some insight to the Staffs position regarding this criterion can be gained from [3-1]. The response to Question 1 of [3-1] states that long-term cooling refers to the sump recirculation mode of ECCS operation and the second paragraph further states that criterion (b)(5) is considered to be satisfied when the core is quenched, the switch from injection to recirculation phases is complete, and recirculation flow rate is large enough to match the boiloff rate. Therefore, criterion (b)(5) is applicable to the scenario being evaluated. With regard to fuel temperature, the same reference defines an acceptable heat-up to be on the order of 212 - 400 °F but allows for heat-ups to be higher if justified. Per [3-2], a fuel cladding temperature of 800 °F for a duration of up to 30 days has been justified (after the initial heat-up and quench). Again, higher fuel cladding temperatures and time at temperature may be acceptable if justified. To summarize:
The core is quenched.
The switch to recirculation is complete.
Recirculation flow is enough to match boiloff.
Fuel cladding temperature of 800 °F can be justified for 30 days (after the initial heat-up and quench).
Evaluation Due to the evaluation method (described below) chosen to demonstrate acceptable safety injection capability, items 3.c and 3.d of the question are no longer considered applicable. The evaluation method neither relies upon the generic studies nor the power shape assumed therein. The WCOBRA/TRAC thermal-hydraulic computer code was used as the evaluation tool. The PINGP-specific base model used a full core model of Westinghouse 0.422-inch outside diameter Vantage+ 14x14 fuel, high peaking factors (FQ = 2.327, FH = 1.798), and a peaked to the top (Axial Offset = +13.9%) axial power distribution shown in Figure 3-1 to minimize voiding in the lower portion of the core and thereby minimize the two-phase mixture level swell. Modifications to the base model include the use of a more conservative decay heat standard (as requested in Item 3.b of the question) defined in 10 CFR 50, Appendix K shown in Figure 3-2, and reduced interfacial drag in the axial direction within the core region to better predict the void fraction and two-phase mixture level swell for low pressure boiloff [3-3].
The low head safety injection (LHSI) flow to the reactor vessel upper plenum was stopped for transfer to recirculation at the earliest anticipated time (1200 seconds) based upon draindown of the refueling water storage tank (RWST) to the low level setpoint. The LHSI was interrupted for the maximum permissible time of 864 seconds, after which it was modeled in recirculation mode at a flow rate of 440 gpm and a temperature of 212 °F starting at 2,064 seconds. During the period of LHSI interruption, the intact loop cold leg high head safety injection (HHSI) was modeled at a flow rate of 290 gpm consistent with the requirement specified in Updated Safety Analysis Report Table 14.10-1 at the maximum (RWST) temperature of 120 °F. At the end of the LHSI interruption, the HHSI pump could continue taking suction from the RWST for approximately 1320 additional seconds based upon draindown of the RWST to the low low setpoint; however, only 300 seconds of continued HHSI injection was credited in the
ENCLOSURE Response to Request for Additional Information License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 422V+ Fuel Page 3 of 19 evaluation. HHSI was terminated at 2,364 seconds and the transient was terminated at 2,964 seconds. Table 3-1 is a table of the sequence of ECCS alignments and flow rates for this transient. Figures 3-3 and 3-4 show the LHSI flow and HHSI flow versus time, respectively.
Figure 3-5 is a plot of the hot rod peak cladding temperature (PCT) versus time. It can be seen that there is a fuel cladding heatup during the transfer to sump recirculation during the period of only HHSI that is quickly terminated once LHSI is restarted in recirculation mode. The fuel cladding temperature peaks at <750 °F and the total duration of the reheat is approximately 720 seconds and core quench is demonstrated by observing that, upon the restart of LHSI, the hot rod peak cladding temperature has been reduced to approximately the saturation temperature of the liquid in the core region. The void fraction at the top of the hot assembly that contains the hot rod is shown in Figure 3-6, and the collapsed liquid levels for the WCOBRA/TRAC core channels are shown in Figures 3-7 through 3-10. Before the LHSI interruption, the void fraction has dropped to approximately 0.90 and the core collapsed liquid levels are either stable or increasing in all the channels. During the LHSI interruption, the top of the hot assembly void fraction begins to increase and eventually reaches 1.0 at which point the hot rod fuel cladding temperature begins to increase. Also, during the LHSI interruption, the collapsed level in all channels drops significantly, approaching 3 ft in the hot assembly. When the LHSI is restarted in recirculation mode, and the HHSI continues to draw from the RWST, the top of the hot assembly void fraction quickly reduces to approximately 0.96, the hot rod re-quenches, and all of the liquid levels slowly begin to increase; they stabilize once cold leg HHSI is terminated several minutes after LHSI is restarted. The maximum calculated hot rod fuel cladding temperature during the LHSI interruption is < 750 °F occurring within a reheat period of approximately 720 seconds. At the end of the transient, the core levels are all stable.
The collapsed liquid level in the counter current flow limitation (CCFL) region (the region between the top of active fuel and the upper core plate) follows the same trend as the core collapsed liquid levels as shown in Figure 3-11. The pool established in the upper plenum completely drains during the period of upper plenum LHSI interruption (upper plenum collapsed liquid levels plotted in Figures 3-12 and 3-13), but recovers after the LHSI is restarted in recirculation mode. The vessel fluid inventory plotted in Figure 3-14 is also stable at the end of the transient.
Conclusion The re-evaluation of post-LOCA long term cooling at PINGP demonstrates acceptable safety injection capability with respect to the acceptance criteria set forth in 10 CFR 50.46(b) with consideration of additional information provided in [3-1] and [3-2].
During the transfer to recirculation while LHSI is interrupted, a fuel cladding reheat to
< 750 °F for a duration of approximately 720 seconds occurs. Once LHSI is restarted in recirculation mode, the core re-quenches. The cold leg HHSI is terminated several minutes following the restart of LHSI thereby completing the transfer to recirculation.
After the transfer is complete, the core and upper plenum region collapsed liquid levels and reactor vessel liquid mass inventory remain stable indicating that boiloff is met and core quench will be sustained.
ENCLOSURE Response to Request for Additional Information License Amendment Request for Technical Specifications Changes to Allow Use of Westinghouse 422V+ Fuel Page 4 of 19 References 3-1 Martin, T. O. (USNRC) to Gresham, J. A. (Westinghouse), Nuclear Regulatory Commission Response to Westinghouse Letter LTR-NRC-06-46 Dated July 24, 2006, Regarding Pressurized Water Reactor (PWR) Containment Sump Downstream Effects, ADAMS Accession No. ML062070451, August 2006.
3-2 Schiffley, F. P. (Pressurized Water Reactor Owners Group) to Document Control Desk (USNRC), Pressurized Water Reactor Owners Group Responses to the NRC Request for Additional Information (RAI) on WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (PA-SEE-0312), ADAMS Accession No. ML073050471, October 2007.
3-3 AP1000 Code Applicability Report, WCAP-15644-P Rev. 2 (Proprietary) and WCAP-15644-NP Rev. 2 (Non-Proprietary), March 2004.
3-4 Prairie Island Updated Safety Analysis Report, Revision 29, May 2007.
Page 5 of 19 Table 3 Sequence of Events Event Time (s)
Mode Start of Transient 0.0 Low Head Safety Injection Begins 32.2 Injection High Head Safety Injection Begins 32.2 Injection Low Head Safety Injection Stops 1200.0 Low Head Safety Injection Restarted 2064.0 Recirculation High Head Safety Injection Stops 2364.0 End of Transient 2964.0
Page 6 of 19 Figure 3-1. Axial Power Distributions of Fuel Rods Modeled in WCOBRA/TRAC
Page 7 of 19 Figure 3-2. Normalized Decay Heat Power Fraction
Page 8 of 19 Figure 3-3. Upper Plenum Low Head Safety Injection Mass Flow Rate
Page 9 of 19 Figure 3-4. Cold Leg High Head Safety Injection Mass Flow Rate
Page 10 of 19 Figure 3-5. Hot Rod Peak Cladding Temperature
Page 11 of 19 Figure 3-6. Top of Hot Assembly Void Fraction
Page 12 of 19 Figure 3-7. Collapsed Liquid Level in the Hot Assembly Channel
Page 13 of 19 Figure 3-8. Collapsed Liquid Level in the Low Power Channel
Page 14 of 19 Figure 3-9. Collapsed Liquid Level in the SC/OH/SP Average Rod Channel
Page 15 of 19 Figure 3-10. Collapsed Liquid Level in the Guide Tube Average Rod Channel
Page 16 of 19 Figure 3-11. Collapsed Liquid Level in the Counter Current Flow Limitation Region
Page 17 of 19 Figure 3-12. Collapsed Liquid Level in the Upper Plenum Inner Global Region
Page 18 of 19 Figure 3-13. Average Collapsed Liquid Level in the Upper Plenum Outer Global Regions
Page 19 of 19 Figure 3-14. Reactor Vessel Fluid Inventory