View source for PNP 2018-014, Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel

{{Adams
| number = ML18086A097
| issue date = 03/27/2018
| title = Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel
| author name = Hardy J
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000255
| license number = DPR-020
| contact person = 
| case reference number = PNP 2018-014
| document type = Letter type:PNP, Response to Request for Additional Information (RAI)
| page count = 9
| project = EPID:L-2017-LLR-0142
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:*
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
  --- Entergy.
27780 Blue Star Memorial Highway Covert, MI 49043-9530 Tel 269 764-2000 Jeffery A. Hardy Regulatory Assurance Manager PNP 2018-014 March 27, 2018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

==SUBJECT:==
Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel (EPID L-2017-LLR-0142)
Palisades Nuclear Plant Docket 50-255 Renewed Facility Operating License No. DPR-20

==References:==
: 1. Entergy Nuclear Operations, Inc. letter to NRC, PNP 2017-068, Proposed Alternative - Relief Request Number RR 5-6, Alternative to the Reexamination Frequency for a Relevant Condition - Foreign Material in Reactor Vessel, dated December 1,2017 (ADAMS Package Accession Number ML17335A013)
: 2. NRC e-mail to Entergy Nuclear Operations, Inc., Palisades Nuclear Plant-Request for additional information regarding proposed alternative for relevant condition (EPID L-2017-LLR-0142), dated February 28,2018 (ADAMS Accession Number ML18059A820)

==Dear Sir or Madam:==

Entergy Nuclear Operations, Inc. (ENO) submitted Reference 1 to the Nuclear Regulatory Commission (NRC) requesting authorization for the Palisades Nuclear Plant of proposed alternative, relief request number RR 5-6, Alternative to the Reexamination Frequency for Relevant Condition - Foreign Material in Reactor Vessel. ENO received an electronic request for additional information (RAI) from the NRC in Reference 2.
Attached is the ENO response to the RAI.
This letter contains no new or revised commitments.
Sincerely, 9A.\~f)
JAH/jpm

PNP 2018-014 Page 2 of 2 :     Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel cc:    Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel A request for additional information regarding Palisades Nuclear Plant alternative to the reexamination frequency for a relevant condition, foreign material lodged in the reactor pressure vessel, from the U.S. Nuclear Regulatory Commission, was received by electronic mail on February 28, 2018. The RAI stated:
By letter dated December 1, 2017 (Agencywide Documents Access and Management System Accession No. ML17335A013), Entergy Nuclear Operations, Inc. (the licensee),
submitted Request No. RR 5-6 for Palisades Nuclear Plant (PNP) to the U.S. Nuclear Regulatory Commission (NRC) for review and approval, pursuant to the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(z)(2). The licensee's application (also referred to as RR 5-6) requested that the NRC authorize its proposed alternative to the successive inspection requirement of Paragraph IWB-2420(b) of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, "Rules for Inservice Inspection [lSI] of Nuclear Power Plant Components," (also referred to as the Code), for a "relevant condition"- a piece of primary coolant pump impeller that is lodged in the interior of the reactor pressure vessel (RPV). The proposed alternative is applicable for the remainder of the fifth 10-year lSI interval at PNP, which commenced on December 13, 2015 and ends on December 12, 2025. In accordance with 10 CR 50.55a(z)(2), the licensee submitted its proposed alternative based on its determination that compliance with the specified Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has determined that additional information is required in order to complete its review of this proposed alternative. The staff's request for additional information (RAI) is provided below.
Regulatory and Technical Basis for RAI-1 and RAI-2 The licensee's analytical basis for its proposed alternative relies on the results of its 2014 operability evaluation for meeting the analytical evaluation requirement of Code Paragraph IWB-3142.4. For acceptance of conditions by analytical evaluation, IWB-3142.4 also requires that reexaminations of such conditions be performed during successive inspection periods in accordance with IWB-2420 to determine whether any changes to the conditions have occurred that would require further corrective action. The staff must review certain information from the 2014 analytical evaluation in order to determine whether this condition will remain acceptable for continued service for the duration of this proposed alternative (through December 2025).
NRC Request (RAI-1)
: 1. Please provide the following information (summary description) for demonstrating that the lodged impeller piece will not affect the functionality of the RPV or the flow skirt during normal plant operations through December 2025:
: a. Impeller piece dimensions; Page 1 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel ENO Response (RAI-1a)
: a. The impeller piece is of asymmetrical shape with approximate dimensions of 13 inches on the longest side and 6 inches on the widest side. The thickness varies from approximately 1/4 of an inch at the longest side to as much as 1 inch on the side opposite the longest side.
NRC Request (RAI-1)
: 1. Please provide the following information (summary description) for demonstrating that the lodged impeller piece will not affect the functionality of the RPV or the flow skirt during normal plant operations through December 2025:
: b. Description of the structural evaluation for determining that the impact of the impeller piece wedged between the RPV and the flow skirt would not exceed structural integrity criteria for the RPV wall or the flow skirt support welds; ENO Response (RAI-1 b)
: b. The 2014 analytical evaluation considered the impact of the wedged impeller piece relative to the structural integrity of the RPV wall and flow skirt support welds, and concluded that there are no impacts on the current structural analyses. This determination was made by calculating the flow induced forces acting on the impeller piece and demonstrating that the impeller piece would not impose a significant load on the RPV wall or flow skirt support welds. Although not specifically mentioned in the 2014 analytical evaluation, these insignificant flow induced forces are also not large enough to further wedge the piece into the gap, between the RPV wall and flow skirt, such that no additional loading on the RPV wall or flow skirt support welds is expected to occur. Additionally, displacement induced loading on the RPV wall and flow skirt support welds due to thermal and pressure effects were dispositioned as inSignificant, citing that the high stiffness of the RPV wall effectively maintained the designed gap between the RPV wall and flow skirt. This ensures that the design gap would be relatively unaffected by thermal or pressure effects, and the impeller piece would not impose significant loads on the RPV wall or flow skirt support welds. Therefore, based on the 2014 analytical evaluation, the stuck impeller piece has an insignificant impact on the RPV wall and flow skirt support weld stresses, and thus is not expected to affect their functionality during normal plant operations through December 2025.
NRC Request (RAI-1)
: 1. Please provide the following information (summary description) for demonstrating that the lodged impeller piece will not affect the functionality of the RPV or the flow skirt during normal plant operations through December 2025:
Page 2 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel
: c. Considering the material types identified in the UFSAR for the RPV cladding (308/309 stainless steel), flow skirt (lnconel), and impeller piece (ASTM A 351, Grade CF8 or Grade CF3), address the potential for corrosion at the interfaces of the lodged piece with the RPV and flow skirt and the effects of corrosion on the structural integrity of the RPV and flow skirt.
ENO Response (RAI-1c)
: c. A corrosion analysis concluded that general corrosion was controlled by passivation of the surfaces, and galvanic corrosion was limited by the reducing environment of the reactor coolant system and the low galvanic potential of the metallic pairs such that the RPV cladding, lodged impeller piece, and flow skirt will not be significantly impacted by corrosion during normal plant operation.
During refueling outage periods, which represents a small portion of the total service time <<10%) of the RPV, the lodged impeller piece (cast austenitic stainless steel ASTM A 351, Grade CF8 or Grade CF3), the flow skirt (Alloy 600, Inconel),
and the RPV cladding (stainless steel weld filler, Type 308 or 309 stainless steel) will be exposed to warm << 212&deg;F) air saturated boric acid solution. Research, which has been previously conducted on galvanic corrosion and crevice corrosion for coupled RPV carbon steel base metal and Type 304 stainless steel in warm, oxygenated boric acid environments, has shown that galvanic and crevice effects are negligible between low alloy (RPV carbon steel base metal) and stainless steels. This research conservatively bounds the impeller piece to RPV cladding and the impeller piece to flow skirt interfaces because it assumes an interface between stainless steel and low alloy steel (RPV carbon steel base metal) when in fact the subject interfaces are between RPV stainless steel cladding and the cast stainless steel lodged impeller piece and the flow skirt Alloy 600 stainless steel and the cast stainless steel lodged impeller.
To further support the negligible effects of corrosion at the subject interfaces, measurements of electrochemical potential (ECP) against a standard hydrogen electrode potential in air saturated solutions of boric acid at 203&deg;F are available for Alloy 600, Type 308 stainless steel (similar to ASTM A351 cast stainless steel impeller piece), Type 304 stainless steel, and ASTM A533 Grade 3 low alloy RPV carbon steel base metal. These measurements give quantitative data on the relative ECPs of the materials of interest in warm, oxidizing conditions. This data shows that the ECP difference between low alloy steel and austenitic; stainless steels bound any differences in ECP between the Alloy 600 (flow skirt) and austenitic stainless steels (impeller piece) interface and the ECP differences between the austenitic stainless steel Type 308 (RPV cladding) and austenitic stainless steel Grade CF3 (impeller piece) interface.
From the above discussions it can be reasonably concluded that the galvanic and crevice corrosion effects between stainless steel (the impeller piece) and Alloy 600 (the flow skirt) and stainless steel (the impeller piece) and stainless steel (the RPV Page 3 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel cladding or more conservatively the RPV carbon steel base metal) in an air saturated boric acid solution would also be negligible. Therefore, since the corrosion effects are negligible, they are also inconsequential to the structural integrity of the RPV and flow skirt at the lodged impeller interfaces, and hence are not expected to affect the functionality of the RPV or the flow skirt during normal plant operations through December 2025.
NRC Request (RAI-2)
: 2. Please provide the following information (summary description) for demonstrating that the lodged impeller piece will not generate loose parts that would adversely affect reactor safety during normal plant operations through December 2025:
: a. Description of the fracture analysis for determining, based on assumed initiating crack sizes in the piece, that the crack growth rate would reduce and essentially stop once the crack depth approached 75 percent of the thickness of the piece; ENO Response (RAI-2a)
: a. A fatigue crack growth rate analysis of the impeller piece using the flaw growth methods of Appendix C (C-3000) of Section XI of the ASME Code was performed.
The analysis modeled the fragment as a rectangular section measuring 14.4 inches long and having three assumed thicknesses of 0.25, 0.5, and 1 inch.
Several thicknesses were evaluated in order to bound the potential effects of a crack in a fragment with varying thicknesses. In order to determine the forces and stresses, the crack model was treated as a simply-supported beam with a displacement applied in the center, in line with the assumed crack. A conservatively assumed displacement was used to represent a load cycle on the fragment. The chosen displacement value was determined to be conservative based on minimal design clearance changes between the RPV shell and flow skirt due to thermal and pressure expansion (Le., load cycle). The forces and stresses were calculated using the applied displacement and classic beam equations.
For the limiting case of an assumed 1 inch thickness, the crack growth rate is very low (less than 5x10-4 inches per loading cycle), and essentially arrests at a depth of approximately 0.75 inches (75% through wall). This supports the conclusion that it is unlikely that, if a crack were to exist in the lodged impeller piece, it would cause the piece to break into smaller pieces. Therefore the impeller piece is not expected to generate loose parts that would adversely affect reactor safety during normal plant operations through 2025.
NRC Request (RAI-2)
: 2. Please provide the following information (summary description) for demonstrating that the lodged impeller piece will not generate loose parts that would adversely affect reactor safety during normal plant operations through December 2025:
Page 4 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel
: b. If the piece could fragment into smaller pieces, please address the impacts of the smaller fragments on the fuel, control rod functionality, and RPV integrity.
ENO Response (RAI-2b)
: b. The impeller piece is not expected to fragment into smaller pieces. The analysis of the forces acting on the piece shows that there is insufficient force to cause the piece to fragment. The piece will remain stuck between the flow skirt and the reactor vessel wall.
In the unlikely event that the lodged impeller piece breaks into smaller pieces, flow forces will, first, direct the pieces through holes in the flow skirt or through the gap between the flow skirt and the reactor vessel. If the pieces are thin enough to pass through the gap or small enough to pass through the flow skirt, the remaining flow forces are small and, based on PNP operating experience, would result in the pieces settling in the lower reactor vessel head under the core support structure.
For this case, these smaller pieces impact on the integrity of the reactor pressure vessel is bounded by the 2014 analytical evaluation discussed in ENO's response to RAI 1 above.         .
If by chance a fragment is caught in the main reactor flow field directed towards the core support plate, the larger fragments will be filtered by the holes in the bottom plate of the core support structure and are likely to be lodged in or pinned against this structure. Smaller fragments that can pass through the flow holes in the bottom support plate are typically trapped in the lower end of the fuel assembly. They are trapped at the lower end of the fuel assembly because the fuel assembly bottom plates have a fuel guard debris filter for added protection against fuel failures from loose parts. Additionally, potential flow area blockage from trapped impeller fragments in the lower internals will have an insignificant effect on core performance since the flow will be redistributed downstream of the blockage and in the lower span of the fuel assemblies.
A fragmented piece is unlikely to affect control rod functionality as the piece would have to travel first through a hole in the core support bottom plate. It must then be lifted to the upper core support plate and pass through a hole in this core support plate. It must then turn horizontally underneath the fuel bundle and turn again vertically to get into the gap between fuel bundles. It is highly unlikely that fragments from the impeller piece would be of a size and shape that would permit them to navigate this path and then turn up and pass through the gaps between fuel bundles. Additionally, PNP's control rod drive mechanisms (CRDMs) are located at the top of the reactor vessel and, since the smaller impeller pieces are highly likely to be filtered by the bottom core support structure and fuel assembly bottom plates they are not expected to reach the CRDMs. Since the CRDMs are located at the top of the reactor vessel, and it is highly unlikely that fragments from Page 5 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel the impeller piece will lodge in the gap between fuel bundles, control rod functionality will not be compromised.
Therefore, based on the above discussion, even if the impeller could fracture into smaller fragments it is expected that there would be negligible impacts on the fuel, control rod functionality, and RPV integrity.
Regulatory and Technical Basis for RAI-3 The application appears to rely on other inspections during normal outages (without removal of the core support barrel) that "provide an opportunity to identify a change in the wedged impeller piece's condition" in the unlikely event the condition of the wedged impeller piece would change. As examples, the licensee cited foreign material inspections of the top of the core, inspections of select fuel bundles inside the core, and inspections of select discharged fuel assemblies during each refueling outage. The staff noted that these other inspections are not specified as part of the proposed alternative under Section 5 of the application.
NRC Request (RAI-3)
Please provide more detail regarding the specific visual examinations of the RPV interior that are implemented during normal outage activities (without removal of the core barrel),
and describe how they could identify whether there is a change in the condition of the lodged impeller piece. If these examinations may provide indications of changing conditions in the wedged impeller piece, please include these other examinations as part of your proposed alternative, or justify why additional examinations do not need to be included as part of the proposed alternative request.
ENO Response (RAI-3)
END did not intend for the routine foreign material exclusion inspections that are conducted during refueling outages to be included in the proposed alternative since they do not directly inspect the stuck impeller piece. The inspections were included in the relief request because they are required by site procedures (Le., RFL-V-4, Foreign Object Search and Retrieval (FOSAR) Prior to Fuel Movement), they provide an opportunity to identify foreign material, and they would evaluate its impact on the fuel. Further, foreign material discovered during these inspections is entered into END's corrective action process (EN-L1-102, Corrective Action Program) and evaluated as to the potential source which may conclude that the foreign material is attributed to a change in the stuck impeller piece condition.
No additional inspections are needed as part of the proposed alternative because, based on the END's RAt responses to RAI 2a and RAI 2b above, it is highly unlikely that the impeller piece will become dislodged or fracture into a small pieces. Further, if this were to occur, it is unlikely that the fragmented pieces would get caught in a significant enough flow stream that would result in them traversing from the bottom of the flow skirt, up through the core support structure and then through the fuel lower tie plate debris filters. Finally, to pass through the debris filter, the fragments would have to be of such a small size (less than 0.1 inches) that Page 6 of 7

PNP 2018-014 ATTACHMENT 1 Response to Request for Additional Information -
Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel their impact on the fuel cladding and control rod function would be nonconsequential. For fragments to bypass the fuel bundle debris filters, they would have to travel a circuitous path from under the fuel bundle then up through the 0.105 or 0.365 inch gap between adjacent fuel bundles. It is highly unlikely that fragments from the impeller piece would be of a size and shape that would permit them to navigate this path and then turn up and pass through the gaps between fuel bundles. This provides high confidence that fuel cladding and control rod functionally will not be compromised by the presence of the stuck impeller piece.
Therefore no additional examinations are needed or are included as part of the proposed alternative described in relief request number RR 5-6 (Reference 1).
References
: 1. Entergy Nuclear Operations, Inc. letter to NRC, PNP 2017-068, Proposed Alternative-Relief Request Number RR 5-6, Alternative to the Reexamination Frequency for a Relevant Condition - Foreign Material in Reactor Vessel, dated December 1, 2017 (ADAMS Package Accession Number ML17335A013)
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