ML18135A263
| ML18135A263 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/21/2018 |
| From: | Wrona D J Plant Licensing Branch III |
| To: | Entergy Nuclear Operations |
| Green K J NRR/DORL/LPL3 415-1506 | |
| References | |
| EPID L-2017-LLR-0142 | |
| Download: ML18135A263 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Vice President, Operations Entergy Nuclear Operations, Inc. Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043-9530 May 21, 2018
SUBJECT:
PALISADES NUCLEAR PLANT -RELIEF REQUEST NO. RR 5-6 ALTERNATIVE TO THE REEXAMINATION FREQUENCY FOR A RELEVANT CONDITION (EPID L-2017-LLR-0142)
Dear Sir or Madam:
By letter dated December 1, 2017, as supplemented by letters dated March 27 and April 30, 2018, Entergy Nuclear Operations, Inc. (ENO, the licensee), submitted Relief Request No. RR 5-6 for the 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).
ENO 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 lnservice Inspection
[ISi] of Nuclear Power Plant Components," for a "relevant condition" -a piece of primary coolant pump (PCP) impeller that is lodged in the interior of the reactor pressure vessel. In accordance with 10 CFR 50.55a(z)(2), ENO submitted its proposed alternative based on its determination that compliance with the specified ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that ENO has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC authorizes the alternative to the ASME Code examination requirements for the PCP impeller piece that is lodged in the interior of the reactor pressure vessel for the remainder of the fifth 10-year ISi interval at PNP, which commenced on December 13, 2015, and ends on December 12, 2025. The NRC staff notes that all other ASME Code requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Energy Nuclear Operations, Inc. If you have any questions, please contact the Senior Project Manager, Kimberly Green at (301) 415-1627.
Docket No. 50-255
Enclosure:
Safety Evaluation cc: Listserv Sincerely, David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1.0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. RR 5-6 REGARDING REEXAMINATION FREQUENCY FOR FOREIGN MATERIAL LODGED IN THE REACTOR PRESSURE VESSEL PALISADES NUCLEAR PLANT ENTERGY NUCLEAR OPERATIONS, INC DOCKET NO. 50-255 INTRODUCTION By letter dated December 1, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 17335A013), as supplemented by letters dated March 27 and April 30, 2018 (ADAMS Accession Nos. ML 18086A097 and ML 18120A218, respectively), Entergy Nuclear Operations, Inc. (ENO, the licensee), submitted Relief Request No. RR 5-6 (referred to as RR 5-6) for the Palisades Nuclear Plant (PNP) to the U.S. Nuclear Regulatory Commission (NRC or Commission) 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 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 lnservice Inspection
[ISi] of Nuclear Power Plant Components," for a "relevant condition" -a piece of primary coolant pump (PCP) impeller that is lodged in the interior of the reactor pressure vessel (RPV). The proposed alternative is applicable for the remainder of fifth 10-year ISi interval at PNP, which commenced on December 13, 2015, and ends on December 12, 2025. In accordance with 1 O CFR 50.55a(z)(2), the licensee submitted its proposed alternative based on its determination that compliance with the specified ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), ISi of ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year ISi interval and subsequent intervals comply Enclosure with the latest edition and addenda of Section XI of the ASME Code that was incorporated by reference in 10 CFR 50.55a(a)(1)(ii), 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. Regulation 10 CFR 50.55a(z) states that alternatives to the requirements of 1 O CFR 50.55a(g) may be used when authorized by the NRC. A licensee's proposed alternative must be submitted to and authorized by the NRC prior to implementation.
The licensee must demonstrate that: (1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality and safety. Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Based on the above regulatory requirements, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to propose and the Commission to authorize this alternative to the requirements of the ASME Code,Section XI. 3.0 3.1 TECHNICAL EVALUATION Licensee's Proposed Alternative (RR 5-6) Code of Record for PNP The Code of Record for the fifth 10-year ISi interval at PNP is the 2007 Edition through 2008 Addenda of the ASME Code,Section XI, as conditioned by 10 CFR 50.55a. The fifth 10-Year ISi interval began on December 13, 2015, and ends on December 12, 2025. Component Covered by the Proposed Alternative This proposed alternative is applicable to the RPV interior at PNP. The RPV interior is an ASME Code Class 1 component.
The ISi criteria for the RPV interior are provided in the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-1, Item No. 813.10. Applicable Code Requirements Table IWB-2500-1, Examination Category B-N-1, Item No. 813.10, requires a VT-3 visual examination of accessible areas of the RPV interior, using the VT-3 visual examination acceptance standard of ASME Code,Section XI, Paragraph IWB-3520.2, to determine the presence of relevant conditions.
ASME Code,Section XI, Paragraph IWB-2420(b), Successive Inspections, states that if a component with relevant conditions is accepted for continued service by analytical evaluation in accordance with ASME Code,Section XI, Paragraph IWB-3142.4, the areas of the component containing relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the Inspection Program of ASME Code,Section XI, Paragraph IWB-2400. Licensee's Statement of Condition In accordance with the ASME Code,Section XI, Paragraph IWB-3520.2, acceptance standard for VT-3 visual examination, the licensee identified a PCP impeller piece stuck inside the RPV during the 2014 refueling outage. The licensee stated that this condition was found during the third period of the fourth 10-year ISi interval, and the condition was accepted for continued service by analytical evaluation in accordance with ASME Code,Section XI, Paragraph IWB-3142.4.
The licensee identified that ASME Code,Section XI, Paragraph IWB-2420(b), requires that it perform successive reexaminations of this condition in the first period of the fifth ISi interval ( ending December 12, 2018), the second period of the fifth ISi interval ( ending December 12, 2022), and the third period of the fifth ISi interval (ending December 12, 2025). Licensee's Proposed Alternative Pursuant to 10 CFR 50.55a(z)(2), the licensee submitted a proposed alternative to the successive reexamination requirement of Paragraph IWB-2420(b) for this relevant condition on the basis that performance of reexaminations during the three successive inspection periods presents an undue hardship without a compensating increase in the level of quality and safety. As an alternative to the IWB-2420(b) requirement for three successive reexaminations, the licensee proposed to perform one reexamination of the wedged impeller piece during either the second period or the third period of the fifth 10-year ISi interval.
Licensee's Basis for the Proposed Alternative The licensee identified that the impeller piece was from PCP P-50C. The piece was found wedged between the inside wall of the RPV and the bottom of the flow skirt in the annulus section of the RPV. The licensee stated that attempts were made to remove the piece without success. The licensee reported that the location of the wedged piece precludes examination using available techniques without fully off-loading the core of fuel and removing the core support barrel. The licensee stated that attempting to perform the examination with the core support barrel in place, using a "yet to be developed alternate inspection technique," would increase the risk of introducing additional foreign material into the RPV. The licensee also stated that performing a full core off-load and removing the core support barrel for the sole purpose of conducting the IWB-2420(b) successive reexaminations would result in increased plant risk and a significant amount of additional radiation does to plant personnel, estimated to be approximately 12.7 rem (roentgen equivalent man). The licensee stated that this condition was accepted by analysis in accordance with ASME Code requirements, as supported by an operability evaluation described in an NRC integrated inspection report (IR) for PNP, dated May 7, 2014 (ADAMS Accession No. ML 14127A543).
The licensee stated its evaluation considered the wedged impeller piece location, the likelihood that it could become dislodged, the likelihood that it could fracture into smaller pieces, and its impacts on the adjacent RPV and flow skirt. The licensee determined that the impeller piece will not dislodge and is not expected to fragment over the remainder of the plant life due to low hydraulic loads acting on the piece. The licensee determined that leaving the impeller piece in place will have no adverse effect on the RPV or any structure, system, or component associated with the primary coolant system. Based on the location of the impeller piece in the bottom of the RPV below the core support cylinder, and not blocking any of the flow holes in the flow skirt, the licensee identified that reactor coolant flow and control rod motion are not affected. The licensee's evaluation determined that the impeller piece will not dislodge because it is tapered in thickness from three-sixteenth-inch to roughly one-inch thick, and the maximum gap size between the RPV wall and the flow skirt where the piece is wedged is one-half inch. Based on a fluid dynamics analysis, the licensee reported that the maximum hydraulic lift force acting on the piece is 350 pounds, which is significantly less than the 3000 pounds of force that it applied in 2014 when attempting to dislodge and remove the piece using hydraulic tools. The licensee also reported that during heatup and cooldown, expansion and contraction of the RPV and flow skirt would be sufficiently similar to ensure that the gap size between the RPV and flow skirt would remain constant.
The licensee stated that part of its evaluation included a fracture analysis of the impeller piece, which determined that the piece is unlikely to break up into smaller pieces during plant operation.
The licensee reported that its fracture analysis determined that for all assumed initiating crack sizes in the piece, the crack growth rate would reduce and essentially stop once the crack depth approached 75 percent of the thickness of the piece. The licensee's evaluation also determined that the RPV interior cladding under the wedged impeller piece is not removed, and no damage to RPV cladding has occurred or is expected to occur during plant operations.
The licensee identified that the wedged impeller piece does not impact the pressure-retaining function of the RPV. The licensee also briefly discussed other inspections that occur each refueling outage that "provide an opportunity to identify a change in the wedged impeller piece's condition";
however, these other inspections are not specifically identified as part of the proposed alternative 1* As examples, the licensee stated that it conducts a visual inspection of the top of the reactor core, select fuel bundles inside the core, and select discharged fuel assemblies for foreign material.
3.2 NRC Staff Evaluation of Proposed Alternative ASME Code Requirements -VT-3 Visual Examination of Pressurized-Water Reactor (PWR) Vessel Interiors and Evaluation of Relevant Conditions The ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-1, Item No. 813.10, requires a VT-3 visual examination of accessible areas of the RPV interior every 3 or 4 year inspection period, as defined in Table IWB-2411-1.
Note (1) of Item No. 813.10 specifies that the areas to be examined shall include regions above and below the core that are made accessible for examination by removal of components.
Examination Category B-N-3, Item No. 813.70, of the ASME Code requires the removal of PWR core support structure (CSS) components for visual examination once every 10-year ISi interval.
In PWRs, RPV interior regions below the core (the location of the lodged impeller piece at PNP) may be considered inaccessible for VT-3 visual examination during normal refueling outages with the CSS in place; however, removal of CSS components for the Item No. 813.70 examinations provides access for VT-3 visual examination of these RPV interior regions once every 10-year ISi interval.
Therefore, for these plants, VT-3 visual examination of RPV interior regions below the CSS is performed once every 10-year ISi interval when CSS components have been removed from the RPV in order to meet the requirements of ASME Code, Item Nos. B 13.10 and B 13. 70. 1 This issue was addressed through the staff's evaluation of supplemental information provided in the licensee's March 27, 2018 letter and is documented in Section 3.2 of this safety evaluation. If the VT-3 visual examination required by Item No. 613.10 detects any of the relevant conditions described in the acceptance standard of ASME Code,Section XI, Paragraph IWB-3520.2, the condition shall require corrective action through the performance of supplemental examinations, repair/replacement activity, or analytical evaluation in order to meet ASME Code requirements for continued service. The relevant condition that is applicable to the lodged PCP impeller piece at PNP is specified in Subparagraph IWB-3520.2(c)-
specifically, the discovery of foreign material that could interfere with control rod motion or could result in blockage of coolant flow through fuel. In ASME Code,Section XI, Paragraph IWB-3142.4, Acceptance by Analytical Evaluation, specifies that a component containing relevant conditions is acceptable for continued service if an analytical evaluation 2 demonstrates the component's acceptability.
Paragraph IWB-3142.4 also specifies that a component accepted for continued service based on analytical evaluation shall be subsequently examined in accordance with the successive inspection requirements of Paragraph IWB-2420.
Paragraph IWB-2420(b) specifies that the areas containing relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the inspection program of Paragraph IWB-2400.
Therefore, reexamination of relevant conditions accepted by analytical evaluation shall be performed every 3 or 4 year inspection period for the next three periods following that when the relevant condition was found and analyzed.
Pursuant to 10 CFR 50.55a(z), alternatives to this successive reexamination requirement shall be formally submitted to and authorized by the NRC prior to the date of the first reexamination required to satisfy Paragraph IWB-2420(b).
Staff Evaluation of the Licensee's Hardship Argument In 2014, during the fourth 10-year interval ISi of the lower RPV interior with the core support barrel removed, the licensee discovered the lodged impeller piece by VT-3 visual examination per ASME Code, Item No. 613.10, and accepted this condition for continued service per IWB-3142.4 analytical evaluation.
Since the 2014 VT-3 visual examination and analytical evaluation occurred during the third period of the fourth 10-year ISi interval, the licensee determined that compliance with ASME Code,Section XI, Paragraph IWB-2420(b), would require three successive reexaminations of this condition, each to occur no later than the end of the first, second, and third periods of the fifth 10-year ISi interval, corresponding to December 12, 2018, December 12, 2022, and December 12, 2025, respectively.
The NRC staff reviewed this information and verified that this sequence of reexaminations would be necessary in order to comply with Paragraph IWB-2420(b
). In RR 5-6, the licensee provided photos and figures of the lodged piece illustrating its location within the RPV and in relation to the internal and CSS components.
Based on its review of this information, the NRC staff verified that the piece is lodged between the lower corner of the flow skirt and the RPV interior and is located in the lower RPV annulus region below the core support assembly.
Based on its review of operating experience (OpE) pertaining to performance of qualified in-vessel visual examinations for PWR RPVs, the staff determined that the lodged piece is located in a region that is generally considered inaccessible for VT-3 visual examinations during normal refueling outage activities that do not involve the removal of the core support barrel. Based on the lack of qualified techniques for performing VT-3 visual 2 IWB-3144(b) specifies that IWB-3142.4 analyses of examination results shall be submitted to the regulatory authority having jurisdiction at the plant site. IWB-3142.4 analyses are therefore provided to NRC inspectors for independent review to verify component acceptability for continued service as part of NRC reactor oversight activities.
However, formal NRC approval of IWB-3142.4 analyses for determining operability is not applicable. examinations of this region with the core support barrel in place, the NRC staff confirmed that any attempt to perform these exams with the core support barrel in place would present an unusual difficulty for the licensee, and it would increase the risk of introducing additional foreign materials into the RPV. Furthermore, the staff confirmed the licensee's statement that performing a full core off-load and removing the core support barrel during successive inspection periods for the sole purpose of conducting the three reexaminations per Paragraph IWB-2420(b) would result in significant additional radiation dose to plant personnel 3* Finally, the staff noted that the increase in hardship associated with examination activities that require removal of CSS components is implicit in the ASME Code Item No. B13.10 requirement for VT-3 visual examinations of accessible areas of RPV interiors every 3 or 4 period because Note (1) of this ASME Code item specifically requires examination of these regions only when they are made accessible by removal of CSS components for the ASME Code Item No. B13.70 inspection once every 10-year interval.
Therefore, the NRC staff determines that compliance with the successive reexamination requirement of Paragraph IWB-2420(b) for the sole purpose of examining the condition of the lodged impeller piece represents an unusual difficulty and would result in undue hardship for the licensee.
Staff Evaluation of Safety Basis for Proposed Alternative The proposed alternative RR 5-6 identifies that the lodged impeller piece was accepted for continued service based on a 2014 analytical evaluation (also referred to as an operability evaluation) in accordance with ASME Code requirements.
NRC inspectors reviewed these analyses to verify the acceptability of the condition for continued service, but the inspection did not itself form a basis for any alternative to ASME Code requirements under 10 CFR 50.55a(z).
The inspectors' review of the analytical evaluation is documented in the May 7, 2014, PNP IR 05000255/2014002 (ADAMS Accession No. ML 14127A543), which states that NRC inspectors independently reviewed the following information from the licensee's evaluation for the lodged piece: 1. The licensee's determination that the impact of the impeller piece wedged between the interior wall of the RPV and the flow skirt did not exceed the structural integrity of the RPV wall or the flow skirt support welds; 2. The analytical basis for determining that the lodged impeller piece is expected to remain in place; 3. The licensee's analysis for determining that the lodged impeller piece is not expected to break up into smaller pieces, and in the unlikely scenario that it did, the impact of the smaller fragments of fuel cooling, fuel cladding, and RPV structure;
- 4. The licensee's assessment of the potential for corrosion at the interface of the lodged impeller piece, RPV, and flow skirt; and, 5. The licensee's assessment of a worst-case scenario accident that could result in the impeller piece impacting the reactor vessel or affecting fuel integrity.
3 The NRC staff must consider whether the performance of the three successive reexaminations is warranted based, in part, on the competing requirement of 10 CFR Part 50, Section 20.1101 (b ), which specifies that licensees shall use, to the extent practical, procedures and engineering controls to achieve occupational doses and doses to members of the public that are "as low as is reasonably achievable (ALARA)." The inspection report IR 05000255/2014002 documents that the licensee's evaluation was sufficient for demonstrating the acceptability of the lodged impeller piece for continued service; thus, the licensee was in compliance with the analytical evaluation clause of ASME Code,Section XI, Paragraph IWB-3142.4.
However, because Paragraph IWB-3142.4 also requires successive reexaminations of the lodged piece per Paragraph IWB-2420(b) to look for changes requiring corrective action, the NRC staff determined that the licensee's application of these analyses for its proposed alternative must demonstrate that performance of the three reexaminations would not result in a compensating increase in the level of plant safety'\ considering the hardship associated with meeting this ASME Code requirement.
Therefore, the NRC staff independently reviewed the licensee's description of these analyses to determine whether they provide reasonable assurance that the lodged impeller piece would remain acceptable for continued service with only one reexamination during either the second or third period of the fifth 10-year ISi interval -or equivalently, no later than the third period of the fifth ISi interval, ending on December 12, 2025. Analyses for Acceptance of the Condition In its March 27 and April 30, 2018, letters, the licensee provided supplemental information requested by the NRC staff to address the analytical basis for acceptance of this condition with one reexamination no later than December 12, 2025. The NRC staff's review of this information is addressed below. Impact of the Lodged Impeller Piece on the RPV and Flow Skirt The licensee's March 27, 2018, letter provided a discussion of its evaluation of the impact of the lodged impeller piece on the structural integrity of the RPV and flow skirt, taking into consideration the potential loading effects of the lodged piece and the potential for corrosion at the interfaces of the lodged piece with the RPV cladding and flow skirt. With respect to the loading effects of the lodged impeller piece, the licensee stated that its 2014 analytical evaluation determined that the wedged impeller piece has no impact on the structural analyses for the RPV wall or the flow skirt support welds. The licensee stated that its evaluation included analysis of flow-induced forces acting on the impeller piece and displacement-induced loading on the RPV wall and flow skirt support welds due to thermal and pressure effects. The licensee explained that the flow-induced forces acting on the piece are 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. The licensee also explained that its analysis of displacement-induced loading showed that the high stiffness of the RPV wall effectively maintains the gap between the RPV wall and flow skirt, such that the lodged impeller piece would not impose significant loads on the RPV wall or flow skirt support welds due to pressure and temperature changes. The licensee stated that its evaluation of these forces and loads showed that the impeller piece has an insignificant impact on the stresses in the RPV wall and the flow skirt support welds, thus, demonstrating that it is not expected to affect their functionality during normal plant operations through December 2025. The NRC staff determined that the licensee's description of its 2014 structural evaluation for the lodged impeller piece provides reasonable assurance that the loading imposed by the piece would have no significant impact on the integrity of the RPV wall and flow skirt support welds. The staff determined that the licensee's description of the flow-induced forces acting on the 4 10 CFR 50.55a(z)(2), Hardship without a compensating increase in quality and safety. piece and displacement-induced loading on the RPV wall and flow skirt show that the licensee adequately considered the loading effects of the piece. The staff noted that the insignificant structural impact of the lodged impeller piece is further supported by the licensee's statement in its submittal that no removal or damage to the RPV interior stainless steel cladding has occurred, based on the 2014 evaluation.
This also supports the licensee's determination that operating stresses beneath the RPV cladding, within the low allow steel pressure boundary, are not significantly affected by the wedged impeller piece. Therefore, the NRC staff determines that the licensee's description of its structural evaluation of the impact of the lodged impeller piece on the RPV wall and flow skirt support welds is acceptable.
While the loading imposed by the lodged impeller piece would not significantly impact the structural integrity of the RPV wall and flow skirt support welds, the NRC staff noted that the effects of localized corrosion at the interfaces with the RPV cladding and the flow skirt could potentially result in time-dependent degradation at the points of contact. Therefore, considering the material types identified in the updated final safety analysis report for the RPV cladding (Type 308 or 309 stainless steel), flow skirt (Alloy 600, lnconel), and impeller piece (ASTM A351, Grade CF8 or Grade CF3), the staff also requested that the licensee address the potential for corrosion at the interfaces of the lodged piece with the RPV cladding and flow skirt and potential effects of corrosion on the structural integrity of the RPV and flow skirt. With respect to the potential effects of corrosion at the interfaces, the licensee's identified that general corrosion and galvanic corrosion are controlled by surface passivity and low galvanic potential of the metallic pairs. The licensee stated that during refueling outage periods, which is less than 10 percent of service time, the subject materials are exposed to warm air-saturated boric acid solution.
The licensee noted that previous research has shown that the effects of galvanic and crevice corrosion are negligible for metallic pairs of RPV low allow steel base metal and austenitic stainless steel in warm, oxygenated, boric acid environments.
The licensee also described an assessment of the differences in electrochemical potential (ECP) for the metallic pairs to demonstrate that this research conservatively bounds the interfaces of the impeller piece with the RPV cladding and flow skirt. The licensee identified that the ECP difference between RPV low alloy steel and austenitic stainless steel bounds any difference in ECP at the subject interfaces.
Based on this evaluation, the licensee concluded that the effects of galvanic and crevice corrosion at the interfaces of the lodged piece with the RPV cladding and flow skirt would be negligible and inconsequential to the structural integrity of the RPV and flow skirt at the interfaces; thus, corrosion would not affect RPV and flow skirt functionality through 2025. The NRC staff determined that the licensee's evaluation of corrosion at the interfaces of the lodged piece with the RPV cladding and flow skirt provides reasonable assurance that the effects of corrosion at these interfaces would be insignificant.
The staff confirmed that all free surfaces in the vicinity of the interfaces would remain protected from general corrosion to the extent that surface passivity is maintained.
The staff verified that the specific metallic pairs of RPV cladding (Type 308/309 stainless steel) and flow skirt (Alloy 600) in contact with impeller material (Grade CF8 or Grade CF3 cast austenitic stainless steel) do not present a galvanic corrosion concern. Considering the licensee's statement regarding previous research on galvanic and crevice corrosion of coupled RPV low alloy steel and stainless steel in warm, oxygenated, boric acid environments, the NRC staff cannot confirm on a generic basis that crevice corrosion would always be negligible in aerated conditions.
However, a review of PWR OpE with RPV cladding indications shows that crevice corrosion has not been an issue for small regions of exposed low alloy steel due to the nearly oxygen deficient PWR primary coolant environment during at-power conditions and the absence of aggressive anion species. Furthermore, extensive study of PWR internals aging mechanisms involving passive austenitic material pairs (stainless steels and nickel-chromium alloys in tight contact, similar to the subject interfaces) has shown that crevice corrosion is not an active degradation concern in PWR primary coolant environments.
These findings provide reasonable assurance that localized corrosion at the interfaces would be insignificant and, thus, inconsequential to the structural integrity of RPV and flow skirt. Therefore, the NRC staff determines that the licensee's description of its corrosion evaluation for the interfaces of the lodged piece with the RPV cladding and flow skirt is acceptable.
Based on its review of the supplemental information provided in the licensee's March 27, 2018, letter, the NRC staff finds that there is reasonable assurance that that the lodged impeller piece will have no significant impact on the structural integrity of the RPV or flow skirt during normal plant operations through December 2025. Loose Parts Evaluation The NRC staff determined that the location of the lodged impeller piece between the lower corner of the flow skirt and the RPV interior ensures that it would not interfere with fuel cooling or the safe function of other structures and components within the RPV, assuming that it does not relocate and does not break up into smaller pieces. The RR 5-6 identifies that the licensee applied 3000 pounds of force to the wedged impeller piece in attempts to dislodge and remove it, and the piece could not be dislodged.
Therefore, the licensee performed a fluid dynamics evaluation as part of its 2014 analytical evaluation and determined that the maximum net fluid force acting on the piece is 350 pounds. In addition, RR 5-6 specifies that the maximum thickness of the piece is one inch, whereas, the maximum gap size between the RPV and flow skirt where the piece is wedged is only one-half inch. Analysis of heatup and cooldown effects showed that this gap size would remain constant.
The NRC staff reviewed this information and determined that there is reasonable assurance that the whole piece is unlikely to dislodge since the net fluid force acting on the piece is less than 12 percent of the force that was applied to the piece in the attempt to dislodge it. Moreover, even if the whole piece could dislodge, it would not fit through the one-half inch gap given its maximum thickness of one inch. Therefore, there is reasonable assurance that the whole piece will not dislodge and relocate to another region inside the vessel. The RR 5-6 documents that the 2014 analyses included a fracture evaluation of the lodged impeller piece, which determined that the piece is unlikely to break up into smaller pieces during plant operation.
RR 5-6 also states that in the unlikely event that the piece were to break up, the smaller pieces would settle to the bottom of the vessel and not affect the safe operation of the plant. With respect to these statements, the licensee provided supplemental information in its March 27 and April 30, 2018, letters that was requested by the NRC staff to demonstrate that the lodged impeller piece would not be expected to generate loose parts that would adversely affect reactor safety during normal plant operations through December 2025. In RR 5-6, the licensee identified that its fracture evaluation for the lodged piece concluded that for all assumed initiating crack sizes in the piece, the crack growth rate would reduce and essentially stop once the crack depth approached 75 percent of the thickness of the piece. Therefore, the NRC staff requested that the licensee provide a description of the fracture analysis for determining that the crack growth rate would reduce and essentially stop once the crack depth approached 75 percent. In its March 27, 2018, letter, the licensee stated that a fatigue crack growth rate analysis of the impeller piece was performed using the methods provided in Article C-3000 of the ASME Code,Section XI, Appendix C. The licensee indicated that the crack model was treated as a simply-supported beam with a displacement applied in the center, in line with the assumed crack. The licensee stated the stresses in the piece for a load cycle were calculated using a conservatively assumed displacement and classic beam equations.
The licensee reported that the crack growth rate is less than 5 X 10-4 inch per load cycle, and the crack essentially arrests at a depth of approximately 75 percent through the thickness of the piece. Based on this analysis, the licensee determined that if a crack were to exist in the lodged impeller piece, the piece would be unlikely to break into smaller pieces and is, therefore, not expected to generate loose parts that would adversely affect reactor safety during normal plant operations through 2025. In its April 30, 2018, letter, the licensee provided further explanation regarding its assumption that the fracture analysis of the lodged piece could be modeled as a beam loaded by a fixed displacement.
The licensee stated that the loading used in its analysis conservatively assumed the maximum force needed to bend the piece to the maximum displaced shape. The licensee explained that this loading scenario is most appropriately considered as a force loading on a three-point bending specimen with a displacement limit due to the finite size of the gap between the RPV and flow skirt. The licensee also stated that the gap size limits the amount that the piece can bend and provides an upper bound on the applied force that causes the piece to bend. The licensee noted that an applied load higher than the upper bound bending force would be balanced by reaction forces where the piece contacts the adjacent components.
The licensee calculated the force needed to bend the piece to the maximum amount and determined it is significantly greater than the steady state fluid force of 350 pounds for crack depths up to 75 percent of the piece thickness.
The licensee confirmed that this bending force would reduce as crack depth increases through the thickness of the piece. The licensee clarified that its crack growth analysis does not show the postulated flaw actually arresting under cyclic loading because the stress intensity factor range (~K,) for the load cycle would still be above the flaw growth threshold for stainless steel. However, the licensee confirmed that the flaw growth rate would start to diminish significantly for postulated cracks greater than 75 percent through the thickness of the piece. The licensee reported that the number of load cycles (specifically, fluctuation cycles between no-flow and flow conditions due to PCP actuation) needed to grow the postulated crack past 75 percent of thickness is on the order of thousands of cycles, which far exceeds the estimated 100 PCP actuation cycles from when this piece was discovered through 2025. The NRC staff reviewed this information and determined that the licensee provided a reasonable explanation of its basis for assuming that the fracture of the piece could be modeled as a beam loaded by a fixed displacement.
Considering the piece as a three-point bend specimen with maximum bending displacement constrained by the gap size between the RPV and flow skirt, the staff confirmed that the bending stress in the piece would decrease as the depth of the postulated crack increases through the thickness of the piece. While the crack may not actually arrest, the staff confirmed that the crack growth rate would reduce as a function of crack depth through the thickness of the piece due to decreasing bending stress, and the number of load cycles to continue growing the crack would be expected to increase.
Considering the number of PCP actuation cycles required to grow the crack beyond 75 percent of the thickness, the staff determined that the licensee adequately demonstrated that postulated cracks would not be expected to grow though the thickness of the piece and result in the piece breaking up. Therefore, the NRC staff determines that the licensee's fracture evaluation of lodged impeller piece is acceptable.
In the unlikely event that the piece were to fragment into smaller pieces, the licensee provided a description of the potential impact of smaller fragments of fuel and control rod functionality.
The licensee indicated that, considering the low fluid forces and prior plant OpE with foreign material, any fragments small enough to pass through the flow skirt would be expected to settle in the lower RPV and not affect other structures and components.
The licensee stated that if a fragment is caught in the main reactor flow field directed towards the core support plate, larger fragments would be filtered by flow holes in the bottom plate of the CSS. The licensee indicated that smaller fragments that could pass through the bottom support plate flow holes would be trapped at the lower end of the fuel assembly by the fuel guard debris filter; this would have an insignificant impact on core cooling since the flow is redistributed downstream of the filter in the lower span of the fuel assemblies.
The licensee also addressed how a fragmented piece would be unlikely to affect control rod functionality because the fragment, even if it could pass through the core support plate, would be highly unlikely to navigate the multi-directional flow path in order to bypass the debris filter and pass into the small gaps between fuel bundles. The licensee concluded that its evaluation demonstrates that, even if the impeller piece could fracture into smaller pieces, there is high confidence that the fragments would not be expected to have an impact on fuel cooling, fuel cladding, or control rod functionality.
The NRC staff determined that the licensee's description of the fluid forces, flow paths and filtering effects in the core region provide reasonable assurance that, if the piece could fragment, it is unlikely that the smaller pieces would be transported into core locations where they could interfere with fuel or the function of control rods. Therefore, any assumed smaller pieces would not be expected to have an adverse impact on fuel and control rod functionality.
Accordingly, the staff determines that the licensee's evaluation of the potential for loose fragments inside the RPV (in the unlikely event of the piece breaking up) is acceptable.
Based on its review of the supplemental information provided in the licensee's March 27 and April 30, 2018, letters, the NRC staff finds that there is reasonable assurance that the lodged impeller piece is unlikely to break up into smaller pieces. However, if it were to break up, the smaller pieces would not be expected to have an adverse impact on fuel or control rod functionality.
Therefore, the lodged impeller piece is not expected to generate loose parts that would adversely affect reactor safety during normal plant operations through December 2025. Foreign Material Inspections The RR 5-6 addresses visual 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." 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 NRC staff noted that these other inspections are not specified as part of the proposed alternative under Section 5 of the request. Therefore, the staff requested that the licensee address how visual inspections of the RPV interior that are implemented during normal outage activities (without removal of the core barrel) could identify whether there is a change in the condition of the impeller piece. If these inspections could identify changing conditions for the piece, the staff requested that the licensee include them as part of its proposed alternative, or justify why they do not need to be included. In its March 27, 2018, letter, the licensee clarified that it did not intend for the routine foreign material exclusion inspections to be included in the proposed alternative since it does not directly inspect the stuck impeller piece. The licensee explained that these inspections were cited in the request because they are required by site procedures, and they provide an opportunity to identify foreign material and evaluate its impact on the fuel. The licensee noted that foreign material discovered during these inspections is entered into ENO's corrective action program and evaluated as to the potential source. The licensee emphasized that no additional inspections need to be included as part of its proposed alternative because it is highly unlikely that the lodged impeller piece would become dislodged or break up into smaller pieces that could interfere with fuel or control rod functionality.
Based on its review of this information, the NRC staff determined that the foreign material inspections mentioned in RR 5-6 are not needed as part of the proposed alternative because they would not actually be capable of observing the condition (or change in the condition) of the lodged impeller piece. If foreign material is observed during these inspections, it would be appropriately addressed by the PNP Corrective Action Program to determine the source. Given its evaluation of the loose parts issues discussed above, the NRC staff determined that the piece is not expected to be a source of foreign material that would interfere with safe plant operation.
Therefore, the NRC staff considers that the routine foreign material inspections conducted without removal of the core support barrel, as cited in RR 5-6, are actually not relevant for determining the condition of the stuck piece, and they are not relied on for the staff's review of this alternative to the ASME Code. Technical Conclusion Based on the foregoing evaluation of RR 5-6, as supplemented by letters dated March 27 and April 30, 2018, the NRC staff has determined that there is reasonable assurance of the following: ( 1) The lodged impeller piece will not affect the structural integrity of the RPV or the flow skirt during normal plant operations through December 2025. (2) The lodged impeller piece is not expected to generate loose parts that would adversely affect reactor safety during normal plant operations through December 2025. Therefore, the NRC staff finds that the lodged impeller piece is acceptable for continued service with one reexamination of its condition to occur no later than the third period of the fifth 10-year ISi interval, ending on December 12, 2025.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the licensee's proposed alternative provides reasonable assurance of structural integrity and adequate protection from loose parts. The NRC staff finds that compliance with the ASME Code,Section XI, Paragraph IWB-2420(b), requirement to reexamine the lodged impeller piece during the next three inspection periods would result in hardship and unusual difficulty for the licensee without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC staff authorizes the alternative described in relief request No. RR 5-6 for PNP until the end of the fifth 10-year ISi interval, which concludes on December 12, 2025. All other ASME Code,Section XI, requirements for which this alternative was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor:
C. Sydnor, NRR Date of issuance:
May 21, 2018
SUBJECT:
PALISADES NUCLEAR PLANT -RELIEF REQUEST NO. RR 5-6 ALTERNATIVE TO THE REEXAMINATION FREQUENCY FOR A RELEVANT CONDITION (EPID L-2017-LLR-0142)
DATED MAY 21, 2018 DISTRIBUTION:
PUBLIC RidsAcrs_MailCTR Resource RidsNrrDorl Resource RidsNrrDorlLpl3 Resource RidsNrrPMPalisades Resource CSydnor, NRR PM Reading File RidsNrrDmlr Resource RidsNrrLASRohrer Resource RidsRgn3MailCenter Resource ADAMS A ccess1on N ML 18135A263 o.: OFFICE DORULPL3/PM DORULPL3/LA DMLR/MVIB/BC*
NAME KGreen SRohrer DAIiey DATE 05/17/18 05/16/18 05/10/18 OFFICIAL RECORD COPY *b *1 1yema1 DORULPL3/BC DWrona 05/21/18