ML17208B024: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
Line 1: Line 1:
{{Adams
#REDIRECT [[CNS-17-039, Relief Request 17-CN-002, Request for Alternative Method to Establish New Inservice Testing Values for a Chemical and Volume Control Pump]]
| number = ML17208B024
| issue date = 07/27/2017
| title = Relief Request 17-CN-002, Request for Alternative Method to Establish New Inservice Testing Values for a Chemical and Volume Control Pump
| author name = Simril T
| author affiliation = Duke Energy Carolinas, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000413
| license number =
| contact person =
| case reference number = CNS-17-039
| document type = Letter
| page count = 12
| project =
| stage = Request
}}
 
=Text=
{{#Wiki_filter:Tom Simril Vice President Catawba Nuclear Station Duke Energy CN01VP  I 4800 Concord Road York, SC  29745 o: 803.701.3340 f: 803.701.3221  tom.simril@duke-energy.com tom.simril@duke-energy.com CNS-17-039 10 CFR 50.55a July 27, 2017
 
U.S. Nuclear Regulatory Commission Attention:  Document Control Desk Washington, D.C. 20555-0001
 
==Subject:==
Duke Energy Carolinas, LLC (Duke Energy) Catawba Nuclear Station, Unit 1  Docket Number 50-413  Relief Request 17-CN-002 Request for Alternative Method to Establish New Inservice Testing Values for a Chemical and Volume Control Pump Pursuant to 10 CFR 50.55a(z)(2), Duke Energy hereby requests NRC approval of an alternative method to establish new reference values for testing the Train A Unit 1, Chemical and Volume Control Pump (1NVPUACC) during the 4th 10-year inservice testing interval for Catawba Nuclear Station (CNS), Unit 1. The relief is requesting an exception from the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code requirement to perform a Group A pump test at the conditions of an existing set of reference values and the results analyzed prior to establishing new reference values for pump 1NVPUACC.
Due to the degraded condition of a manual isolation valve, re-creating the conditions of an existing set of reference values and the results analyzed to test pump 1NVPUACC, would put the plant at risk of an unnecessary plant shutdown. The valve lineup needed to accomplish an online repair of the valve is prohibited by Technical Specifications. The pump has had consistent acceptable performance with no indications of degradation. Thus compliance with the OM Code method for establishing a new set of reference values, creates a hardship without a compensating increase in quality and safety as described in the enclosed relief request.
Duke Energy requests approval of the proposed alternative by August 11, 2017, in support of the planned quarterly test  that is due August 13, 2017, and no later than August 30, 2017, to avoid missing the "Late Date" (end of grace period for the test) which is September 5, 2017.
There are no regulatory commitments contained in this letter.
If you have any questions regarding this submittal, please contact Tolani E. Owusu at (803) 701-5385. 
 
U.S. Nuclear Regulatory Commission CNS-17-039 July 27, 2017 Page 2 Sincerely, Tom Simril Vice President, Catawba Nuclear Station
 
==Enclosure:==
 
Catawba Nuclear Station, Unit 1 Relief Request 17-CN-002 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
U.S. Nuclear Regulatory Commission CNS-17-039 July 27, 2017 Page 3 XC: (with Enclosure)
C. Haney,        Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J. D. Austin, Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station M. Mahoney,        NRC Project Manager (Catawba)
U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop O-8H4A Rockville, MD 20852-2738
 
S. E. Jenkins, Manager      S.C DHEC Radioactive & Infectious Waste Management jenkinse@dhec.sc.gov
 
U.S. Nuclear Regulatory Commission CNS-17-039 July 27, 2017 Page 3 bxc: (with Enclosure)
Cecil Fletcher  Jon-Michael Vandergriff Greg Harbin Randall Herring Mandi Brigman Steve Mays Tolani Owusu PMPA NCMPA-1 NCEMC RGC File Document Control File
- CN 801.01 ELL-EC02ZF
 
Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 1 of 8 Enclosure Catawba Nuclear Station, Unit 1 Relief Request 17-CN-002  Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
 
Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 2 of 8 Relief Request 17-CN-002 Proposed Alternative  in accordance with 10 CFR 50.55a(z)(2)
: 1. ASME Code Component(s) Affected Train A Unit 1, Chemical and Volume Control Pump (1NVPUACC)  Flow Diagram:  CN-1554-01.07 IST Pump Group: A  ASME Class:  2
: 2. Component Function The accident mitigation function of the Chemical and Volume Control (NV) System is to provide high pressure injection and recirculation of borated water to the Reactor Coolant (NC) System as part of the Emergency Core Cooling System  (ECCS) during a design basis accident. During an accident, the NV System is isolated except for the Centrifugal Charging Pumps (CCPs), Safety Injection (NI) and Reactor Coolant (NC) Pump seal injection flow paths. 
 
There are three primary functions of the ECCS. These functions are: (1) removing stored and fission product decay heat, (2) controlling reactivity, and (3) precluding reactor vessel boron precipitation. The CCPs have the highest discharge pressure and are first to be effective in delivering flow to the NC System. Injection of borated water from the Refueling Water Storage Tank (FWST) continues and increases with any decrease in NC System pressure. As NC pressure is reduced, sequentially lower pressure portions of the ECCS become effective in delivering flow. The Design Basis Events which impose requirements on the design of the ECCS System are: Main Feedwater Line Rupture, Main Steamline Break (including spurious steam line valve openings), Steam Generator Tube Rupture, Loss of Coolant Accident - Small Break (including spurious pressurizer valve openings), Loss of Coolant Accident - Large Break, and Control Rod Drive Mechanism Rupture. The CCPs will start automatically on a Safety Injection or on a Loss of Offsite Power (LOOP) in order to accomplish their safety related function as part of the ECCS. The CCPs are sequenced onto the diesel generator on Load Group 2.
: 3. Applicable Code Edition and Addenda ASME Operations and Maintenance Code (OM Code) 2004 through 2006 Addenda.
: 4. Applicable Code Requirement:
Establishment of Additional Set of Reference Values ASME OMb-2006, Section ISTB-3320; specifically, the requirement in paragraph one states: "If it is necessary or desirable, for some reason other than stated in ISTB-3310, to establish an additional set of reference values, a Group A or comprehensive test shall be run at the conditions of an existing set of reference values and the results analyzed." 
 
Note:  This relief is being requested to allow Catawba Nuclear Station Unit 1, to take a one-time exception from the requirements of performing the Group A test at an existing set of reference values for 1NVPUACC. 
 
Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 3 of 8
: 5. Reason for Request
 
During the Catawba Unit 1 Refueling Outage 23, (C1R23), the 1NVPUACC mini-flow isolation valve 1NV269 was identified to have a packing leak. Per that work order, the packing was adjusted (tightened) to eliminate leakage. Per the current test method and reference values, this manual isolation valve is closed during the quarterly 1NVPUACC inservice test to isolate the mini-flow line.
Following the Unit 1 outage, when attempting to perform the first quarterly test on 6/23/17, the mini-flow valve was extremely difficult to operate, indicating binding in the packing or stem. The packing gland was re-adjusted (loosened) on 7/10/17, but  the valve continues to be extremely difficult to operate and it is likely the valve problem makes it susceptible to getting stuck in the closed position. The inability to reopen the valve upon completion of future pump tests  would result in entering a Technical Specification (TS) Limiting Condition for Operation (LCO) that requires shutdown within 72 hours. An online repair of the mini-flow isolation valve would require isolation that makes both CCP trains inoperable which is not allowed by TS in Modes 1-4. Therefore, this repair must be performed during an outage. 
: 6. Proposed Alternative and Basis for Use:
Alternative Test Method Catawba Nuclear Station proposes to establish an additional set of reference values for Group A testing of pump 1NVPUACC based on data analysis that shows acceptable performance, using prior performance data, in lieu of the OM Code requirement of performing a test at existing reference values prior to establishing additional reference values.
 
Per ASME OMb-2006, ISTB-3320, states that prior to establishing additional reference values, the licensee is to perform a Group A test at the existing set of reference values and analyze the results. Establishing the conditions for performing a test at the existing reference values will likely lead to a plant shutdown. Therefore, Catawba has performed an analysis of previous test data at the existing reference values in order to verify acceptable pump performance. This analysis provides the basis for establishing an additional set of reference values  per ISTB-3300 (a) for a test configuration that does not challenge a plant shutdown.
Per ISTB-3300 (a), new reference values will be established by placing the pump in the appropriate alignment to establish the new desired test flow. Pump data will then be gathered at this new flow condition to establish the new reference values. The lower acceptable limit for differential pressure will be set at  2381 psid, as determined by the expected pump flow rate (approximately 150 GPM), which is more limiting than that of table ISTB-5121-1. Once the new reference conditions are established the first in-service test will be run at these reference conditions to verify the new reference conditions, per ISTB-3320. 
 
Basis for Use The relief is being requested for Catawba Nuclear Station Unit 1, to be excluded from the requirements of re-performing the Group A test at an existing set of reference values for pump 1NVPUACC. The equipment issue explained in Section 5, prevents the quarterly Group A test from being performed at the existing reference values, because there is a high potential that the mini-flow isolation valve would not be capable of reopening upon completion of the quarterly test. This would  lead to a TS  required shutdown of the Unit.
The current 1NVPUACC was installed in November 2000 and no work significantly affecting the hydraulic characteristics of the pump (dP/flow) has been performed. Trending data since 2006 (earliest electronic data available) indicates that pump 1NVPUACC performance has been steady and well within the limits of the IST requirements. See subsequent trending discussion below for Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 4 of 8 details. Additionally, as further evidence of no pump degradation since the last performance of the quarterly IST, Operations Rounds data was reviewed dating back to September 2016, and found that when pump flow was similar (dependent on Rx coolant system letdown), that the differential pressure was also very similar. 
 
Pump Performance Data:
Quarterly Differential Pressure:. The lower limit for quarterly IST differential pressure (at the current 90 gpm flowrate) is 2442 psid, as established by the system flow requirements. The upper limit, established by table ISTB-5121-1, is 2863 psid. The data since 2006 indicates that pump differential pressure has been relatively steady, remaining  between 2488 - 2620 psid, with an average of 2550 psid.
 
Comprehensive Differential Pressure (performed during refueling outages):  The acceptable range for the comprehensive pump test, established via table ISTB-5121-1, is 1250 - 1384 psid at flows between 461 gpm and 470 gpm (see Green Box below). The data since 2008, when the current reference point was established indicates pump differential pressure remained within the acceptance range and within 15 to 20 psid of the pump curve.
 
24002450250025502600265027002750280028502900 85 86 87 88 89 90 91 92 93 94Differential pressure (psid) Flow (gpm) Quarterly IST Data - NV Pump 1A Test Flow (gpm)Measured Diff. Pressure (psid)Lower Acceptance Limit (psid)Upper Acceptance Limit (psid)Baseline Diff. Pressure (psid)
Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 5 of 8 Similarly, pump vibration readings over time indicate acceptable pump performance with no degrading trends. The relative reference values for vibration (only) were rebaselined once in 2012 due to a rebuild of the pump's mechanical seals and the gear drive .that improved vibration characteristics. Pump hydraulic performance was unaffected by the work so hydraulic reference values remained unchanged.
12401280 132013601400460461462463464465466467468469470471 DIfferential Pressure (PSID) Flow (gpm) Comprehensive Test Results 1NVPUACC 2000 Head CurveACCEPTABLE BANDComp Test-12/1/09Comp Test-5/23/11Comp Test-12/12/12Comp Test-6/2/14Comp Test-12/7/15Comp Test-5/12/17Comp baseline-5/28/08 Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 6 of 8 Current Vibration Alert Range Criteria (rebaselined in 2012) Measurement Location Outboard BRG Vertical (V1) Outboard BRG Horiz. (H1)
Outboard BRG Axial (A1)
Inboard BRG Vertical (V2) Inboard BRG Horiz. (H2)
Quarterly Limit (inches/sec) 0.19625 0.26600 0.05825 0.15900 0.13550 
 
Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 7 of 8 Current Vibration Alert Range Criteria (rebaselined in 2012) Measurement Location Outboard BRG Vertical (V1) Outboard BRG Horiz. (H1)
Outboard BRG Axial (A1)
Inboard BRG Vertical (V2) Inboard BRG Horiz. (H2)
Comprehensive Limit (inches/sec) 0.19600 0.24800 0.04475 0.09350 0.09375 
: 7. Duration of Proposed Alternative The Proposed Alternative will be utilized to establish an additional set of reference values for the 1NVPUACC. This condition is only intended to permit Catawba Unit 1 a one-time exception from the requirements to re-perform the quarterly Group A test prior to establishing new reference values. The licensee will return to the ASME OM Code specified test method for establishing additional sets of reference values following repair of the mini-flow isolation valve. Currently the next Catawba Unit 1 refueling outage C1R24, is scheduled to begin on November 17, 2018. For any subsequent performance test, any deviation in pump performance that exceeds the newly established acceptable limits will be evaluated in accordance with ISTB-6200, Corrective Action.
: 8. 
 
==References:==
 
a)  Duke Energy, Catawba Nuclear Station ASME OM Code lnservice Testing Program,    Revision  31, dated May 30, 2017 b)  NEI white paper "Standard Format for Requests from Commercial Reactor Licensees Pursuant to Enclosure CNS Unit 1 Relief Request 17-CN-002 July 27, 2017
 
Page 8 of 8 10CFR 50.55a" Revision 1 dated June 7, 2004. c)  Duke Energy Fleet Procedure AD-EG-ALL-1720 "lnservice Testing (IST) Program Implementation"  Revision 02 dated February 2, 2017. d)  Technical Specification Surveillance Requirement 3.5.2
 
e)  NCR-02124639/WR-20069989: 1NV-269, ACTIVE BORIC ACID LEAKAGE
 
f)  NCR-02132906/WR-20076737: 1NV VA 0269: EXTREMELY DIFFICULT TO OPERATE
 
g)  CNTC-1554-NV.P003-01: Catawba Nuclear Station, Unit 1 NV System Test Acceptance Criteria-  NV (Centrifugal Charging) Pumps
 
h)  CNS-1554.NV-00-0001: Design Basis Specification for the Chemical and Volume Control System  (NV)}}

Latest revision as of 08:16, 6 April 2019

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML17208B024&oldid=1406757"