ML17090A110

Millstone Power Station, Unit 2 - ASME Section Xi Relief Request RR-04-25

ML17090A110
Person / Time
Site:	Millstone
Issue date:	03/29/2017
From:	Sartain M D Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
17-127, RR-04-25
Download: ML17090A110 (10)

Text

Dominion Nuclear Connecticut, Inc. 5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 March 29, 2017 DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 ASME SECTION XI RELIEF REQUEST RR-04-25 Serial No. NLOSfTFO Docket No. License No.17-127 R1 50-336 DPR-65 Pursuant to 10 CFR 50.55a(z)(2), Dominion Nuclear Connecticut, Inc. (DNC) requests relief from IWB-3142 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Millstone Power Station Unit 2 (MPS2). The relief request is needed to allow MPS2 to shutdown and conduct refueling activities with the "B" boric acid pump available which has a through-wall leak at the stuffing box cover. DNC has assessed the condition of the stuffing box cover as structurally acceptable for continued service. However, ASME Section XI requires the defect be corrected prior to returning the affected component to service. In a letter dated December 14, 2016 (ML 16354A424), DNC informed the Nuclear Regulatory Commission (NRC) that control element assembly (CEA)-39 associated with MPS2, has a potentially degraded upper gripper coil, and requested approval to revise the Technical Specifications to exempt CEA-39 from the last remaining quarterly freedom of movement test in operating Cycle 24. In a letter dated February 7; 2017 (ML 17018AOOO), the NRC approved the request to revise the Technical Specifications. Due to the condition of CEA-39, plant shutdown to start the next refueling outage will be accomplished by injection of boron. The "B" boric acid pump is one of two redundant pumps provided to support boron injection to the reactor coolant system via the suction of the charging pump. The continued availability of the "B" boric acid pump is needed to maximize the boric acid system reliability and defense in depth during the remainder of the operating cycle and for execution of the plant shutdown and core offload. Due to a long lead time for a replacement part, a code repair of the "B" boric acid pump stuffing box cover is not possible prior to the scheduled unit shutdown. Consequently, compliance with the requirements of 10 CFR 50.55a would result in a hardship without a compensating increase in the level of quality or safety. Attachment 1 to this letter describes the engineering evaluation of the defect, actions that will be implemented by DNC, and the basis for the proposed relief request. A permanent code repair will be completed when the replacement part is received. DNC requests approval of this relief request by March 30, 2017 to support planned unit shutdown activities. This relief request has been approved by the Millstone Facility Safety Review Committee.

Serial No.17-127 Docket No. 50-336 Page 2 of 2 If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687. Sincerely, Mark D. Sartain Vice President-Nuclear Engineering and Fleet Support

Attachment:

1. ASME Section XI Relief Request RR-04-25, Boric Acid Pump P-19B Stuffing Box Cover Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C 2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station ATTACHMENT 1 ASME SECTION XI RELIEF REQUEST RR-04-25 BORIC ACID PUMP P-19B STUFFING BOX COVER MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC. Serial No.17-127 Docket No. 50-336

---Proposed Relief Request RR-04-25 In Accordance with 10 CFR 50.55a(z)(2) --Hardship or Unusual Difficulty Serial No.17-127 Docket No. 50-336 Attachment 1, Page 1 of 7 Without Compensating Increase in Level of Quality or Safety--1. ASME Code Components Affected ASME Code Class: Code Class 2

Reference:

ASME Section XI, IWC-2500 Examination Category: Table IWC-2500-1, Category C-H Item Number: C7 .1 O

Description:

Components: Millstone Power Station Unit 2 (MPS2) "B" boric acid pump. The "B" boric acid pump stuffing box cover is a casting of type 316 austenitic stainless steel per American Society for Testing and Materials (ASTM) A351 CF-8M. 2. Applicable Code Edition and Addenda MPS2 is currently in the fourth 10-year inservice inspection (ISi) interval, which started on April 1, 2010, and ends on March 31, 2020. The ASME Section XI, 2004 Edition (No Addenda) applies to the ISi program. 3. Applicable Code Requirement The ASME Code requirements for which this relief request is being submitted are those associated with Section XI, 2004 Edition, No Addenda, (Reference 8.1) and contained in Article IWB-3142, Acceptance. 4. Reason for Request On February 28, 2017, during a visual (VT-2) examination of the "B" boric acid system, dry boric acid residue was identified on the pump casing. The source of the boric acid could not be identified. This condition was entered into the corrective action program. During the following shift, the boric acid residue was cleaned and the pump was run for two hours in an attempt to determine the leak location and leak rate. No evidence of leakage was observed by the plant equipment operator who was stationed at the pump during the two-hour run. On March 1, 2017, a follow up ISi walkdown was performed and dry boric acid residue was found at the same location, but in a smaller quantity. The pump was declared non-functional and corrective action was initiated to identify the source of the leak.

Serial No.17-127 Docket No. 50-336 Attachment 1, Page 2 of 7 On March 2, 2017, the "B" boric acid pump was tagged out, but not drained, to support an informational liquid penetrant (LP) examination. The examination found no relevant or recordable indications. Following the LP examination, the "B" boric acid pump was cleaned and then run for an additional time. After 10 minutes, a small accumulation of boric acid residue appeared as a wet, translucent spot. After 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of run time, the wet spot had increased in size to about 1/8 inch in diameter. Although a measurable leak rate could not be determined, growth of the wet spot confirmed the presence of a through-wall leak. The pump was subsequently isolated on March 9, 2017. DNC has researched the original purchase order to identify initial testing associated with the boric acid pump. The pump was constructed in accordance with the draft ASME Code for Pumps and Valves for Nuclear Power, dated November 1968, as Class 2, and Combustion Engineering specification 18767-PE-404, Revision 1, dated February 9, 1972. The Combustion Engineering specification refers to the ASME code for testing requirements. Section B of the draft ASME code contains the requirements for Class 2 pumps and valves. Article 23, sub-article 2314 specifies special requirements related to Class 2 components. Normally, this sub-article would require that radiography be performed for a cast product. However, article 2314 identifies that the special requirements do not apply to materials for pumps and valves with inlet piping connections of four inches and less in nominal pipe size unless otherwise noted in the design specification. The inlet piping to the boric acid pump is less than four inches and the design specification did not invoke a radiography requirement for this pump. Initial testing did include an LP examination and a hydrostatic test and were found satisfactory The 2004 ASME Section XI code was reviewed for acceptance standards associated with the stuffing box cover leak. Table IWC-3410-1, Acceptance Standards, indicates for pressure retaining components (examination category C-H), the acceptance standard is IWC-3516. IWC-3516 refers to IWB-3522. IWB-3522 refers to IWB-3142 for corrective actions. Article IWB-3142, Acceptance, specifies that a component with a relevant condition is unacceptable for continued service unless the requirements of IWB-3142.2, 3 or 4 have been satisfied. All three options have been considered. IWB-3142.2 is related to acceptance by performing supplemental examinations. This section is about sizing of defects for comparison to ASME criteria to allow leaving a flaw in service. This option is not applicable to a through-wall condition. IWB-3142.4 is related to determining acceptance by analytical evaluation. This option is not possible because there is not a code-specified methodology for analyzing the condition and . flaw characterization is limited. Of the three options available for acceptance of a relevant condition, a repair/replacement per IWB-3142.3 is the only viable option for addressing the current observed condition. However, a replacement part is not readily available. A purchase order has been submitted to the pump vendor for fabrication of a

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Serial No.17-127 Docket No. 50-336 Attachment 1, Page 3 of7 replacement part with expedited delivery. Replacement of the degraded part can be accomplished with MPS2 in service. Based on engineering assessment of the tound condition, there is high confidence that structural integrity of the stuffing box cover will be maintained for the duration of the required period of continued operation. In a letter dated December 14, 2016 (Reference 8.2), DNC informed the Nuclear Regulatory Commission (NRG) that control element assembly (CEA)-39 has a potentially degraded upper gripper coil, and requested approval to revise the Technical Specifications to exempt CEA-39 from the last quarterly freedom of movement test in operating Cycle 24. In a letter dated February 7, 2017 (Reference 8.3), the NRG approved the request to revise the Technical Specifications. Due to the condition of CEA-39, plant shutdown to start the next refueling outage will be accomplished by injection of boron into the reactor coolant system. The "B" boric acid pump is one of two redundant pumps provided to support boron injection to the reactor coolant system. The continued availability of the "B" boric acid pump is needed to maximize the boric acid system reliability and defense in depth during the remainder of the operating cycle and for execution of the plant shutdown and core offload. Due to a long lead time for a replacement part, a code repair of the "B" boric acid pump stuffing box cover is not possible prior to the scheduled shutdown date. Maintaining the "B" boric acid pump available provides additional assurance that the boric acid injection function will remain available while bringing the unit offline and while maintaining shutdown margin requirements during the early stages of the refueling outage. Compliance with the requirements of 10 CFR 50.55a would result in a hardship without a compensating increase in the level of quality or safety. The function of the "B" boric acid pump is for reactivity control, but the pump is not credited in the Final Safety Analysis Report, Chapter 14 accident analyses. 5. Proposed Alternative and Basis for Use 5.1 Proposed Alternative The pump will be removed from service following completion of core offload. DNC will inspect the affected pump stuffing box cover each shift for leakage until the pump is removed from service. The inspection results will be documented. Should leakage *increase, an engineering evaluation will be performed to reassess structural integrity. If confidence in structural integrity cannot be maintained, the pump will be isolated. 5.2 Basis for Use Specific Considerations Flaw Characterization and Results -Because of limited access and complex surface geometry, direct volumetric examination of the leakage location from Serial No.17-127 Docket No. 50-336 Attachment 1, Page 4 of 7 the outside surface of the assembled pump is not possible. However, to characterize the leak path and condition of the material surrounding the defect and area in which the wetness was observed, the following inspections were performed: o An LP examination was performed on the affected area of the stuffing box cover, no recordable indications were identified. o Leakage was confirmed visually following completion of the LP examination and extended pump run. This leak location is tracked under the boric acid control program, which manages the impact of the leakage on any potential targets in its vicinity. Based on the results of the above inspections, the stuffing box cover pressure boundary leakage is likely due to small casting void defects or porosity that enable a through-wall pathway for leakage. Alternative explanations have been considered, but determined to be unlikely, as discussed below. For conservative characterization of the leakage pathway and its effects on structural integrity and functionality, it is ,assumed to be bounded as a 1/8 inch diameter straight through-wall defect. Structural Evaluation Leak Location and Operating Conditions -The location of the leakage is on the portion of the stuffing box cover that surrounds the pump shaft. At this location, the stuffing box seal cover pressure cannot be measured but a typical rule for estimating stuffing box pressures is: P = (P outlet -Pinlet) I 4 + Pinlet P =stuffing box pressure, psig p outlet = pump design discharge pressure, psig = 150 psig Pinlet = pump suction pressure, psig = 4.5 psig The pump design discharge pressure is 150 psig and the suction pressure is 4.5 psig. These values give an estimate for the stuffing box pressure of 41 psig using the above rule. For conservatism, 50 psig is assumed.

Material Performance and Degradation Potential -The figure at right shows the configuration of the boric acid pump. The boric acid pump has been in service since initial plant operation. The cross-hatched area on the left is the stuffing box cover casting, and the impeller is the cross-hatched component on the right. The stuffing box cover is made from cast type 316 austenitic stainless steel per specification ASTM A351 CF-8M, as shown on the vendor drawing. Type 316 stainless steel has excellent general corrosion resistance to the boric acid solution that is the process fluid for this pump. Based on testing done for Electric Power Research Institute (Reference 8.4), the general corrosion rate in the process fluid is too small to measure. The other internal parts of the pump are made from cast or wrought type 316 stainless steel. Type 316 stainless steel, either cast or wrought, is not susceptible to pitting or stress corrosion cracking in this environment. The cast version of type 316 stainless steel is not susceptible to thermal aging at the low temperature at which this pump Serial No.17-127 Docket No. 50-336 Attachment 1, Page 5 of 7 operates. Additionally, mechanical loads due to normal operation, including pump vibration, are low and would not be anticipated to result in service induced flaw growth. With these considerations, it is concluded that there are no active aging degradation mechanisms for this component that would cause initiation and through-wall growth of a planar flaw. The remaining reasonable explanation for the leakage is that there is a leakage pathway via small voids, or porosity, originally present in the cast material. Other instances of this phenomenon in the industry are referenced in the section 7, Precedents. As discussed below, such small imperfections do not significantly affect the structural integrity of the component. The absence of an active degradation mechanism also supports the conclusion that the currently observed leak rate will remain nearly constant for the duration of this requested relief. Structural Integrity -There is no code-specified methodology for evaluating the structural integrity of this type of component when through-wall leakage is detected. However, the stuffing box cover in the area of concern can be conservatively bounded as a cylindrical section, similar to a cylindrical vessel with localized leakage. The Code of Construction for the pump is the draft ASME Code for Pumps and Valves for Nuclear Plants, 1968. The stuffing box cover material is ASTM A351 CF-8M which has an allowable design stress of 17.5 ksi. The inside diameter of this section is approximately 2.5 inches.

Serial No.17-127 Docket No. 50-336 Attachment 1, Page 6 of 7 Considered as equivalent to a cylindrical vessel, the minimum wall thickness would be calculated per ASME Ill NC-3324 as: tmin = P RI (S -0.6 P) The values for the parameters in this formula are: P = stuffing box pressure, psig = 50 psig R = radius of the stuffing box= 2.5/2 = 1.25 inches S =stuffing box allowable stress= 0.8 x 17.5 = 14 ksi (including a casting quality factor of 0.8) The result is tmin = 0.0045 inch. The actual wall thickness of this section in the area of the observed leakage is not specified on drawings but is estimated to be approximately 0.250 inch based on the minimum wall thickness required for valves by the construction code. The evaluation result of 0.0045 inch required wall thickness, in comparison to the approximately 0.250 inch estimated actual wall thickness, results in a factor of about 56. This demonstrates that this portion of the pump stuffing box cover is not challenged by the pump stuffing box pressure. Additionally, this portion of the stuffing box cover is not required to withstand any significant mechanical loading. This portion of the casting supports only the mechanical shaft seal. The pump shaft is supported independently of the stuffing box cover and seismic loading from the shaft, impeller, and piping nozzles are transferred through the pump frame adaptor and pump casing to anchorage feet attached to the pump casing. The loads do not pass through this portion of the stuffing box cover. Thus the minimum wall thickness calculation reasonably demonstrates the structural integrity of the pump stuffing box cover in the area of the observed leakage. Functionality -The above structural integrity evaluation supports the conclusion that the pump is capable of performing its intended function of transferring concentrated boric acid from the boric acid storage tank to the suction of the charging pumps and will retain this capability for the duration of the requested relief. The overall mechanical integrity of the pump will be maintained such that the pump will be able to generate sufficient head for the required flow, and since the pressure boundary is maintained there would be no significant diversion of boric acid intended for injection into the charging pump suction. Conclusion The structural integrity of the stuffing box cover cannot be demonstrated in accordance with a code-specified methodology. However, by comparison with typical ASME code design rules, it is concluded that there is reasonable Serial No.17-127 Docket No. 50-336 Attachment 1, Page 7 of 7 assurance that the structural integrity and functional requirements of the pump will be maintained during the requested period of relief. 6. Duration of Proposed Alternative This relief is requested to be effective upon approval and until completion of core offload during the spring 2017 MPS2 refueling outage. 7. Precedents A search of the industry OE was performed for similar instances of casting voids associated with relief requests. The search did not find any that directly applied. However, the following items are related to cast voids and are listed for reference. 7.1 NRC letter from J. Quichocho (USNRC) to W. Gideon (CP&L), dated December 12, 2012*, "H. B. Robinson Steam Electric Plant, Unit No. 2 -Relief Request-07 from Immediate ASME Code Repair of Refueling Water Storage Tank Drain Valve (Safety lnjection-837) for Fifth10-year lnservice Inspection Program Plan (TAC NO. ME9747)." [ADAMS Accession Number ML 12325A612] 7.2 NRC Letter from J. B. Martin (USNRC) to C. A. Schrock (WPSC) dated November 5, 1993, related to Notice of Enforcement Discretion 'associated with Residual Heat Removal Pump Casting Void. (As identified in WPSC letter from C. A. Shrock to USNRC, dated December 6, 1993, "Reportable Occurrence 93-019-00"). 8. References 8.1 ASME Code Section XI, Division 1, 2004 Edition with No Addenda. 8.2 DNC Letter to NRC [ADAMS Accession Number ML 16354A424] dated December 14, 2016 -Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 2, License Amendment Request to Revise Technical Specification Surveillance Requirement 4.1.3.1.2 for Control Element Assembly 39 for the Remainder of Cycle 24. 8.3 NRC Letter to DNC [ADAMS Accession Number ML 17018AOOO] dated February 7, 2017 -Millstone Power Station, Unit No. 2 -Issuance of Amendment, Re: Technical Specification Surveillance Requirement 4.1.3.1.2 for Control Element Assembly 39 (CAC No. MF8935). 8.4 2012 EPRI Technical Report: Materials Reliability Program: Boric Acid Corrosion Guidebook, Revision 2: Managing Boric Acid Corrosion Issues at PWR Power Stations (MRP-058, Rev 2).