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Category:Report
MONTHYEARML24103A2482024-04-12012 April 2024 Emergency Core Cooling System Performance Evaluation Models, 10 CFR 50.46(a)(3)(ii) Annual Report for 2023 ML24019A2012024-01-19019 January 2024 Fourth 10-Year Interval, Second Period Owners Activity Report Number 1R24 ML23325A1612023-10-26026 October 2023 Relief Request 70 - Proposed Alternatives in Accordance with 10 CFR 50.55a(z)(1) for Pressurizer Lower Shell Temperature Nozzle ML22308A1662022-11-0404 November 2022 23rd Refueling Outage Steam Generator Tube Inspection Report ML22091A3072022-04-0101 April 2022 Independent Spent Fuel Storage Installation (Isfsi), Evacuation Time Estimate Study ML21077A2562021-03-18018 March 2021 El Paso Electric Co., Application for Approval of Indirect Transfer of Control of Licenses Pursuant to 10 C.F.R. 50.80 and 72.50 ML21005A2712020-12-29029 December 2020 102-08208 PVNGS Communication Required by Confirmatory Order EA-20-054 ML19136A4092019-05-16016 May 2019 Emergency Core Cooling System Performance Evaluation Models, 10 CFR 50.46(a)(3)(ii) Annual Report for Calendar Year 2018 ML19046A3512019-02-15015 February 2019 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U2R21 ML18187A4172018-07-0606 July 2018 License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel ML17181A5152017-06-29029 June 2017 Flooding Focused Evaluation Summary ML17248A5262017-05-30030 May 2017 3INT-ISI-3, Revision 4, Third Inspection Interval Inservice Inspection Program Summary Manual, Unit 3, Enclosure 3 to 102-07551-MDD/MSC ML17006A2172017-01-0505 January 2017 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U3R19 ML16306A4442016-11-14014 November 2016 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation ML16321A0032016-10-31031 October 2016 WCAP-18030-NP, Revision 1, Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3. ML16221A6042016-09-13013 September 2016 Staff Assessment of Information Provided Under Title 10 of Code of Federal Regulations Part 50, Section 50.54(F), Seismic Hazard Reevaluations for Recommendation 2.1 of Near-Term Task Force Review of Insights from . ML16120A3892016-04-28028 April 2016 1-SR-2016-001-00, Fuel Building Ventilation System High Range Radioactive Gaseous Effluent Monitor Nonfunctional ML15336A0842015-11-30030 November 2015 NET-300047-07 Revision 1, Material Qualification Report of Maxus for Spent Fuel Storage. ML15188A0492015-06-30030 June 2015 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U3R18 ML15111A4302015-04-17017 April 2015 Attachments 2 & 3: Flaw Fracture Mechanics, Corrosion & Loose Parts Evaluations for One Cycle Relief (DAR-MRCDA-15-6-NP) and APS Response to NRC Request for Additional Information Dated April 14, 2015 ML15076A0732015-03-10010 March 2015 Seismic Hazard and Screening Report ML15012A4462015-01-0606 January 2015 Third 10-Year Interval, Second Period: Owner'S Activity Report Number U1R18 ML15027A1222014-12-19019 December 2014 Technical Specification (TS) Bases Revision 61 ML14184B3822014-06-27027 June 2014 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U2R18 ML14189A5332014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5 RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 52014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 ML14189A5312014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5 ML14189A5372014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5 ML14189A5002014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 5 ML14189A4982014-06-19019 June 2014 Stars Response to Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program - Vendor Information Request ML14189A5432014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 52014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 52014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 52014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 52014-06-19019 June 2014 Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5 ML14126A6362014-04-30030 April 2014 APS Supplement to the Seismic Walkdown Report Requested by the NRC Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1, 2.3, and 9.3 of the Near-Term Task Force. ML14087A1882014-04-11011 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to Fukushima Dai-Ichi Nuclear Power Plant Accident ML14034A1472014-01-24024 January 2014 Third 10-Year Interval, Second Period: Owner'S Activity Report Number U3R17 ML13329A0382013-08-31031 August 2013 WCAP-17787-NP, Rev 0, Palo Verde Nuclear Generating Station STAR Program Implementation Report. ML13252A1112013-08-31031 August 2013 WCAP-17680-NP, Supplement 1, Rev. 0, Near-Term Task Force Recommendation 2.3 Seismic Walkdown Submittal Report for Palo Verde Nuclear Generating Station Unit 2 - Supplemental Information. CY-12-009, Closure Options for Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance2013-05-16016 May 2013 Closure Options for Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance ML13120A5032013-04-10010 April 2013 Technical Requirements Manual, Revision 57 ML1307003422013-02-28028 February 2013 APS Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) ML13022A4072013-01-14014 January 2013 Third 10-Year Interval, Second Period: Owner'S Activity Report for Unit 2 Refueling Outage 17 (U2R17) ML12355A7482012-12-31031 December 2012 Evacuation Time Estimate Study; Cover Through Section 4 ML12355A7512012-12-14014 December 2012 Evacuation Time Estimate Study; Section 5 Through 12 ML12355A7522012-12-14014 December 2012 Evacuation Time Estimate Study; Appendix a Through H ML12278A1002012-09-0707 September 2012 Pressurized Water Reactor (PWR) Internals Aging Management Program Plan ML12229A1192012-08-0303 August 2012 Attachment 4, PVNGS Engineering Study 13-ES-A037, Revision 0, Fault Tree Analysis and Reliability Evaluation for Low Pressure Safety Injection (LPSI) Pump Trip at the Recirculation Actuation Signal (RAS) ML12229A1202012-08-0303 August 2012 Attachment 5, PVNGS Engineering Study 13-NS-C089, Revision 0, PRA Evaluation of LPSI Pump Failing to Trip on RAS 2024-04-12
[Table view] Category:Administrative
MONTHYEARML1025301352010-09-0101 September 2010 Comments Received from Mr. Bob Leyse to ACRS for Consideration at Palo Verde License Renewal Subcommittee Meeting 2010-09-01
[Table view] |
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ACRS Subcommittee on Plant License Renewal September 8, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. (Palo Verde)
Facts for the Subcommittee and for the record:
- 1. On May 23, 2007 Bonaca wrote Kline,
SUBJECT:
PROPOSED TECHNICAL BASIS FOR THE REVISION TO 10 CFR 50.46 LOCA EMBRITTLEMENT CRITERIA FOR FUEL CLADDING MATERIALS, ML071490090. Bonaca wrote, The requirements of 10 CFR 50.46 (a) and (b) limit the amount of embrittlement that may occur as result of a design basis accident. They specify limits for the peak clad temperature, the global oxidation of cladding, and the local oxidation of cladding. There are several deficiencies with the current regulations. The correlation specified for the rates of steam reaction with the cladding is viewed by the technical community as an anachronism. Now, Appendix K to Part 50--ECCS Evaluation Models, Item 5, specifies that the rate of energy release from the metal/water reaction shall be calculated using the Baker-Just equation and § 50.46 Acceptance Criteria, item (b)(1) specifies the peak clad temperature, 2200 degrees.
- 2. The NRC staff fiercely defends Baker-Just in its Technical Safety Analysis, ML041210109, April 29, 2004, The Baker-Just correlation using the current range of parameter inputs is conservative and adequate to assess Appendix K ECCS performance. Virtually every data set published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F.
- 3. The nuclear power industry fiercely defends Baker- Just in its Industry Comments, ML101040678, April 12, 2010, The Baker-Just correlation, using the current range of parameter inputs, has been shown to be conservative and adequate to assess Appendix K ECCS performance. Data published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F
- 4. Contrary to the exceptionally firm consistency between the NEI and NRC appraisals of Baker-Just, the pertinent data sets published since the Baker-Just correlation was developed have clearly demonstrated the non-conservatism of the Baker-Just correlation above 1800°F. The NRC has not recognized that investigations that involve heating of single specimens of zirconium alloys in steam do not yield applicable data for the temperature or range of temperatures at which thermal runaway is initiated in LWRs.
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- 5. NRC has apparently never studied Baker-Just (ML050550198) and until April 2010 it did not even have copies of the key references. Figure 16 is copied from page 37 of the Baker-Just report ML050550198.
Only the Lemmon data includes the pertinent temperature region. The Lemmon report, ML100570218, was not acquired by NRC until April, 2010.
Thus, NRC never studied Baker-Just. Figure C-1 is from Lemmon page C-4; the adjacent figure is excerpted from the flow sheet, Figure C-3 on page C-5.
Lemmon induction heated a zircaloy-2 cylinder, 2 long by 0.5 dia. in steam.
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- 6. It is absurd to license the emergency cooling of tons of zirconium alloy having thousands of square feet of interfacial surface area based on the limited investigations that yielded the Baker-Just equation. Despite this, Appendix K to Part 50--ECCS Evaluation Models, Item 5, specifies that the rate of energy release from the metal/water reaction shall be calculated using the Baker-Just equation and § 50.46 Acceptance Criteria, item (b)(1) specifies the 2200 degrees.
- 7. Data from multi-rod (assembly) severe fuel damage experiments (e.g., the LOFT LP-FP-2 experiment) show the Baker-Just equation is non-conservative for calculating the temperature at which thermal runaway would occur in the event of a LOCA.
- 8. Investigations by P. Hofmann et al. at Forschungszentrum Karlsruhe reveal that the Baker-Just equation is non-conservative for calculating the temperature at which thermal runaway will occur in a LOCA. Their report is, Physico-Chemical Behavior of Zircaloy Fuel Rod Cladding Tubes During LWR Severe Accident Reflood, Part I: Experimental results of single rod quench experiments, FZKA 5846, http://bibliothek.fzk.de/zb/berichte/FZKA5846.pdf On page 5 of 177: A series of separate-effects tests is being carried out on Zircaloy PWR fuel rod cladding to study the enhanced oxidation which can occur on quenching. In these tests, performed in the QUENCH rig, single tube specimens are heated by induction to a high temperature and then quenched by water or rapidly cooled down by steam injection.
On gage 12 of 177: No significant temperature excursion during quenching occurred such as had been observed for example in the quenched (flooded) CORA-bundle tests This absence of any temperature escalation is believed to be due to the high radiative heat losses in the QUENCH rig.
And in, CORA Experiments on the Materials Behavior of LWR Fuel Rod Bundles at High Temperatures, NUREG/CP-0119, Vol. 2, Proceedings of the Nineteenth Water Reactor Safety Information Meeting. ML042230460, P. Hofmann et al.
On page 98 of 493: The critical temperature above which uncontrolled temperature escalation takes place due to the exothermic zirconium/steam reaction crucially depends on the heat loss from the bundle; i.e., on bundle insulation. With the good bundle insulation in the CORA test facility, temperature escalation starts between 1100 and 1200°C (2012 to 2192°F), giving rise to a maximum heating rate of 15 K/sec.
- 9. It is amazing that the ACRS has never reviewed Baker-Just in the course of producing its recommendations regarding the initial licensing, the extended licensing and the licensing of power level increases of numerous American light water reactors.
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Palo Verde, Units 1, 2 and 3 - Updated Final Safety Analysis Report, Revision ML072250202 2007 14. 30 PVNGS UPDATED FSAR EMERGENCY CORE COOLING SYSTEM June 2007 6.3-76 Revision 14 6.3.3 PERFORMANCE EVALUATION 6.3.3.1 Introduction and Summary 10 CFR 50.46 provides acceptance criteria for Emergency Core Cooling Systems (ECCS) for light-water nuclear power reactors
[Reference 1]. The ECCS performance analyses described in this section demonstrate that the PVNGS ECCS design satisfies these criteria.
The PVNGS ECCS performance analyses encompass a wide range of Reactor Coolant System (RCS) break locations and sizes, including both large and small break Loss-of-Coolant Accident (LOCAs). The limiting break, which results in the closest approach to 10 CFR 50.46 acceptance criterion for peak clad temperature, is a 0.6 DEG/PD (Double-Ended Guillotine in the Reactor Coolant Pump Discharge leg) as noted in UFSAR Section 6.3.3.2. The limiting break, which results in the closest approach to 10 CFR 50.46 acceptance criterion maximum clad oxidation (or local clad oxidation), is a 0.8 DEG/PD as noted in UFSAR Section 6.3.3.2.
For these limiting breaks, the PVNGS ECCS design meets the acceptance criteria of 10 CFR 50.46 as follows:
Criterion 1: Peak Cladding Temperature. ". . .The calculated maximum fuel element cladding temperature shall not exceed 2200°F. . . ."
For the limiting break, the PVNGS ECCS performance analysis yielded a peak cladding temperature of 2110°F.
PVNGS UPDATED FSAR EMERGENCY CORE COOLING SYSTEM June 2007 6.3-130 Revision 14 6.
3.6 REFERENCES
- 1. Code of Federal Regulations, Title 10, Part 50, Section 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
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