Vogtle 3 and 4: Indications Of Massive Nuclear Plant Design Screw-ups

The pressure operated relief valves are(PORV)very important safety components. The bad design made the noise problem. Now this system is in a massive redesign even when this plant never ran. Is this just the beginning.  

Draft Request for Additional Information on Vogtle LAR 18-021 Main Steam PORV Noise Mitigation October 10, 2018


RAI 1 - relates to the structural stress analysis of the PORV                           ASME Boiler & Pressure Vessel Code (BPV Code), Section III, incorporated by reference in 10 CFR 50.55a, requires that piping analysis consider combinations of various loadings, including deadweight, pressure, seismic, thermal expansion and transient loads.  Southern Nuclear Operating Company (SNC) submitted Vogtle Electric Generating Plant (VEGP) Units 3 and 4 License Amendment Request (LAR) 18-021 proposing changes to the main steam (MS) branch line containing the power-operated relief valve (PORV).  These changes include relocating the PORV branch line from the MS line, increasing the size of the PORV branch line, reducing the PORV branch line length, and changing the PORV block valve size and type.  LAR 18-021 does not provide information related to the effects of the proposed changes on the structural integrity of the applicable systems, structures and components (SSCs).  To demonstrate that the structural integrity of the applicable SSCs will be maintained within acceptable design-basis limits as a result of the proposed changes in Vogtle LAR 18-021, the NRC staff requests that SNC provide the following information: 

Question 1:

(a) The MS PORV branch line had been decoupled from the MS line for the pipe stress analysis. The branch line is proposed to be changed from 6-inches to 12-inches nominal pipe size (NPS).  Please provide the minimum wall thickness or schedule of the planned 12 NPS branch line.  If the PORV branch line remains decoupled in the revised calculations, demonstrate that the proposed 12-inch branch line remains justified to be decoupled from the MS line using the current design-basis decoupling criteria. 

(b)    Describe the effects of the proposed changes in LAR 18-021 on the structural integrity of the applicable SSCs.  For the MS line and the PORV branch line, provide a brief summary of the maximum pipe stresses compared to the design-basis ASME allowable values.  In addition, provide a brief summary of the results of the evaluation of the MS containment penetration, applicable existing supports, and any additional supports to be installed.  

(c) Provide a summary of the maximum pipe break stresses compared to the design-basis break exclusion criteria as described in the FSAR.  SNC should demonstrate that the proposed changes do not result in any new postulated break locations and, therefore, there is no impact to the conclusions of the Pipe Rupture Hazard Analysis.

SNC may make applicable proprietary information to address these concerns available in its electronic reading room for NRC staff audit review.  If SNC intends to direct the NRC to the electronic reading room, please provide the exact references to documents so that they may be mentioned in an audit plan.

RAI 2 - relates to vibration and valve performance for the PORV modification
Question 1:  

General Design Criterion (GDC) 1, “Quality standards and records,” in Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50 states, in part,  that structures, systems, and components (SSCs) important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.  GDC 4, “Environmental and dynamic effects design bases,” in 10 CFR Part 50, Appendix A, states that SSCs important to safety shall be appropriately protected against dynamic effects.  Vogtle License Amendment Request (LAR) 18-021, “Power Operated Relief Valve (PORV) Noise Mitigation,”  in Section 2 of Enclosure 1 on page 5 states that the MCR noise levels will be verified as part of the AP1000 human factors engineering verification and validation program.  The NRC staff requests that SNC describe its acoustic resonance analysis (such as acoustic resonance evaluation and Strouhal analysis) to demonstrate that the planned modification for the PORV block valve and branch line will not result in vibration levels that exceed the allowable limits.  For example, SNC should determine the potential sources of acoustic resonance and the applicable plant conditions.  SNC should address the planned PORV branch line connection design and radii that could cause potential adverse flow effects.  SNC should consider any plant operating experience with the proposed PORV branch line modification.  

SNC may make applicable proprietary information to address these concerns available in its electronic reading room for NRC staff audit review.  If SNC intends to direct the NRC to the electronic reading room, please provide the exact references to documents so that they may be mentioned in an audit plan.

Question 2:  

GDC 1 in 10 CFR Part 50, Appendix A, states that SSCs important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.  GDC 1 states, in part, that appropriate records of the design, fabrication, erection, and testing of SSCs important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.  Vogtle LAR 18-021 states in Section 2 in Enclosure 1 on page 4 that the new globe valves to be used in the PORV block valves will be qualified for the same environmental and pressure/temperature conditions as the current PORV block valves.  The Vogtle Final Safety Analysis Report (FSAR) with reference to the plant-specific AP1000 Design Control Document, Tier 2, Section 3.9.3.2.2, “Valve Operability,” specifies that active valve assemblies will be qualified in accordance with ASME Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” which the NRC endorsed in RG 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants.”  The NRC staff requests that SNC describe its plans to satisfy the provisions in ASME QME-1-2007 for the dynamic, environmental, and functional qualification of the new PORV block valves.  

SNC may make applicable proprietary information to address these concerns available in its electronic reading room for NRC staff audit review.  If SNC intends to direct the NRC to the electronic reading room, please provide the exact references to documents so that they may be mentioned in an audit plan.

Question 3:  

The NRC staff requests that SNC clarify the statements in Vogtle LAR 18-021 in Section 2 of Enclosure 1 on page 11 that there will be “no change to the valve motor operator” and “no change to the valve stroke time” in light of the potential differences in stroke length and operating requirements between the original 6-inch gate valve and the new 12-inch globe valve.