ML16355A097
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Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Changes, Tests, and Experiments As required by 10 CFR 50.59(d)(2), the following report contains brief descriptions of changes made to the facility and/or associated documentation, and summaries of the associated 50.59 evaluations.
SS-SE-2014-0112-00: Upgrades to Unit 2 Manipulator Crane (2-QM-90)
Activity
Description:
This activity replaces the 2-QM-90 1-1/2 ton fuel manipulator crane. The scope of the Unit 2 manipulator crane modifications in EC-0000052314 consists of a total electrical/control system upgrade and changes to the top of the main hoist structure to accommodate new hoist pulleys and replacement of servo motors, load cells, pneumatics, brakes, gears, hand wheels, cable tensioners, and bumpers. Also, a new air conditioner unit and an extension of the operator platform will be added, along with hand rail modifications required to accommodate the new control cabinets. The current analog control system that is used for indication, interlock and limit switch protection of the crane and fuel assemblies in the gripper will be replaced with a digital control system.
This activity is a cross-unit implementation in Unit 2 of a previously approved change in Unit 1.
Summary of the Evaluation:
This activity is bounded by 10 CFR 50.59 Evaluation SS-SE-2014-0051-01, previously submitted to the NRC. For convenience, the Summary of the Evaluation of SS-SE-2014-0051-01 is replicated below:
The likelihood of a Fuel Handling Accident (FHA) in Containment continued to be minimal based on existing and modified crane safety features, administrative controls (operating procedures),
facility design characteristics, dependability of the new software and the enhanced control system, and the limited risk of credible failure effects. In addition to using NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, for the conduct of the 50.59 Evaluation, the dependability of the software and the associated failure modes and effects were evaluated using criteria in NEI 01-01, "Guideline on Licensing Digital Upgrades EPRI TR-102348 Rev. 1."
Manipulator crane operating procedures were enhanced and the required training was given to personnel.
These actions ensured that crane operators closely monitor crane motion during automatic and semiautomatic control modes and are ready to use the simple, one step "Emergency Stop" button to halt all crane motion in the event of any unusual indication or alarm on the control console or unexpected crane motion.
Accident consequences from a postulated Fuel Handling Accident (FHA) in Containment were found to be bounded by the previous Analysis of Record.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
SS-SE-2015-0147-00: Revise U2 Ice Basket Weight Acceptance Criteria for Cycle U2C22 Activity
Description:
The changes and issues covered by EC-0000054270 Rev. 0 are:
1.
Revising U2 Technical Specification Bases Surveillance Requirement SR 3.6.11.3 to allow use of a revised alternate Active Ice Mass Management (AIMM) approach to provide reasonable assurance that the minimum ice basket mass (600 lbm. net) will continue to be met per SR 3.6.11.3 during the planned period of operation even though the 1132 lbs. mean value for each ice basket will not be maintained in Cycle 22 due to the inability to empty and refill specific baskets during U2C22 refueling outage.
2.
Modifying Calculation ENPM-12-ICE-001-N Rev. 5, "Ice Condenser Design Basis Surveillance Requirements," to reflect ice maintenance practice changes in U2C22.
3.
Revising plant Procedure 12-EHP-4030-010-262, "Ice Condenser Surveillance and Operability Evaluation," to reflect ice maintenance practices changes for U2C22.
This activity is a cross-unit implementation in Unit 2 of a previously approved change in Unit 1.
Summary of the Evaluation:
This activity is bounded by 10 CFR 50.59 Evaluation SS-SE-2014-0454-00, previously submitted to the NRC. For convenience, the Summary of the Evaluation of SS-SE-2014-0454-00 is replicated below:
The ice condenser is a passive component serving an accident mitigation function for LOCAs and steam line breaks. It is not an accident initiator and no credible failure of the ice condenser or an individual ice basket would result in an accident.
The assumed heat removal capability of the ice baskets in the limiting LOCA Containment Integrity analysis of record conservatively assumes equal ice basket loading and no ice condenser bypass since these assumptions minimize complete ice melt-out time and maximize calculated peak containment pressure. Further, review of the distribution of ice basket weights forecast by ICEMAN at the end of Cycle 26 concludes that the dispersed lower weight baskets and the proximity of light baskets to heavy baskets do not represent significant localized degradation of the ice bed. A sensitivity evaluation that bounded the projected ice basket weights at the end of Cycle 26 confirmed that total ice mass melt-out would be slightly delayed and peak containment pressure would be slightly reduced. This supported the expectation that the current LOCA Containment Integrity analysis of record remains bounding after implementation of EC-0000053931.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
SS-SE-2015-0307-00: Revise U1 Ice Basket Weight Acceptance Criteria for U1C27 Activity
Description:
The changes and issues covered by EC-0000054414 Rev. 0 are:
1.
Revising U1 TS Bases SR 3.6.11.3 to allow use of a revised alternate AIMM approach to provide reasonable assurance that the minimum ice basket mass (600 lbm. net) required by SR 3.6.11.3 will be met during the planned period of operation even though the 1132 lbs mean value for each ice basket will not be maintained in Cycle 27. With the exception of the cycle numbers and updated EC reference, this is the same TS Bases change made during the U1C26 outage as part of EC-0000053931.
2.
Modifying Calculation ENPM-12-ICE-001-N Rev. 6, "Ice Condenser Design Basis Surveillance Requirements," to reflect changes in ice maintenance practices and administrative limits for U1C27. These changes are consistent with the changes made to this calculation for U1C26.
3.
Revising plant Procedure 12-EHP-4030-010-262, "Ice Condenser Surveillance and Operability Evaluation," to reflect changes in ice maintenance practices and administrative limits for U1C27.
These changes are consistent with the changes made to this procedure for U1C26.
4.
Preparation of new calculation MD-01-ICE-003-N, "Ice Condenser Ice Mass Evaluation for Unit 1 Cycle 27," which was developed to assess the predicted ice bed condition at the end of U1C27 operation. This calculation is incorporated by reference into EC-0000054414 and serves to provide technical justification for the proposed change. Note that the corresponding evaluation providing technical support for EC-0000053931 for U1C26 operation was included directly in the EMOD package.
This activity re-implements an activity previously implemented on Unit 1.
Summary of the Evaluation:
This activity is bounded by 10 CFR 50.59 Evaluation SS-SE-2014-0454-00, previously submitted to the NRC. For convenience, the Summary of the Evaluation of SS-SE-2014-0454-00 is replicated below:
The ice condenser is a passive component serving an accident mitigation function for LOCAs and steam line breaks. It is not an accident initiator and no credible failure of the ice condenser or an individual ice basket would result in an accident.
The assumed heat removal capability of the ice baskets in the limiting LOCA Containment Integrity analysis of record conservatively assumes equal ice basket loading and no ice condenser bypass since these assumptions minimize complete ice melt-out time and maximize calculated peak containment pressure. Further, review of the distribution of ice basket weights forecast by ICEMAN at the end of Cycle 26 concludes that the dispersed lower weight baskets and the proximity of light baskets to heavy baskets do not represent significant localized degradation of the ice bed. A sensitivity evaluation that bounded the projected ice basket weights at the end of Cycle 26 confirmed that total ice mass melt-out would be slightly delayed and peak containment pressure would be slightly reduced. This supported the expectation that the current LOCA Containment Integrity analysis of record remains bounding after implementation of EC-0000053931.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
SS-SE-2015-0322-00: Implementation of Unit 1 LOCA-Containment Integrity Analysis Using WCOBRA\\TRAC Mass and Energy Releases (WCAP-17721-P-A)
Activity
Description:
An Engineering Change was used to replace the Current Licensing Basis methodology for the Loss of Coolant Accident Mass and Energy (LOCA M&E) Releases from WCAP-10325-P-A Rev. 1 to WCAP-17721-P-A Rev. 0 for Donald C. Cook Unit 1. The new M&E Releases were then used as input to a new Unit 1-specific LOCA Containment Response analysis to demonstrate continued compliance with Containment Integrity requirements. [Previous Containment Integrity analysis bounded both Units 1 and 2.] In addition to the LOCA M&E Release methodology change, some revised input parameters were used in the new Unit 1-specific analysis. The Engineering Change did not involve any physical changes to the plant, changes to design limits or other system performance parameters, or changes to equipment operation or maintenance.
Summary of the Evaluation:
The Evaluation assessed the impacts of the revised input parameters on existing containment integrity requirements and the acceptability of replacing the LOCA M&E Release methodology. Input parameter changes were accepted based on the new analysis results that demonstrated continued compliance with Containment Integrity requirements. Use of WCAP-17721-P-A Rev. 0 was accepted for Unit 1 LOCA M&E Release methodology based on findings that 1) the LOCA M&E Release methodology is described in the UFSAR and is used in Unit 1s design bases or safety analyses, 2) the NRC has approved use of WCAP-17721-P-A for large break LOCA M&E Release analysis for ice condenser plants, 3) the new analysis adequately documents that the Limitations and the Condition identified in the NRC Safety Evaluation of WCAP-17721-P-A are met, 4) the Westinghouse personnel performing the new analysis and their supporting organization are qualified to implement the WCAP-17721-P-A methodology, and 5) the WCAP-17721-P methodology is being incorporated into the Unit 1 safety analyses en toto.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
SS-SE-2015-0479-01: Temporary Modification in Support of CCW in an event of LOCA coincident with Design Basis Earthquake Activity
Description:
The Component Cooling Water system must be able to withstand a single passive failure. It has been posited that the worst case for this scenario is a passive failure in the CCW system consisting of a valve packing leak of 11.3 GPM (worst case starting value). This failure will be detected when the CCW surge tank low-level alarm annunciates in the control room or level alarms from the sumps to which this water will drain annunciate. Per calculation MD-12-CCW-005-N (Valve and pump seal leakage from the miscellaneous train of the component cooling water system) the operators have 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 56 minutes before the surge tank would be drained to the point that the CCW pump would not retain the required net positive suction head to perform its function.
Further, the scenario is exacerbated by the fact that the CCW valves that would be used for isolating the leak are not leak tested by the IST Program (they are Category B), and therefore cannot be fully credited for isolating the leak due to seat leak-by.
Finally, the postulated leak could potentially be due to through wall pipe leak in the non-safety, non-code class CCW miscellaneous header, which is a common header to both trains on a unit.
The result of the above described scenario is that a loss of inventory of CCW will continue to occur even after the crosstie valves are closed. It has been determined that a temporary compensatory measure to provide an alternate means of making up CCW inventory will be implemented. There is no credit for crosstie to the other unit as the same scenario could also have taken out CCW make up ability on it.
In the event that the leakage continues to be observed following isolation of the leak (i.e., through the crosstie valves) the compensatory measure is proposed to be a Proceduralized Temporary Modification.
It will ensure that the system continues to function under the postulated conditions (including for the 30 day mission time of the CCW system following a design basis accident).
This Procedurally Controlled Temporary Modification will install mechanical jumper(s) to connect the Essential Service Water (ESW) system to the Component Cooling Water (CCW) system. The jumper will be a flexible hose connected between each set of valves. These jumper(s) will provide a source of makeup water to the CCW system from the ESW system, intended to be used only when there is no make-up water available from the Demineralized Water system.
Summary of the Evaluation:
As a compensatory measure to address the situation where there is a loss of inventory of CCW that will continue to occur even after the crosstie valves are closed, there are no new failure modes introduced as the only failure mode associated with the compensatory measure would be minor system leakage and CCW system leakage is already addressed by plant procedures. Further, it is intended as a mitigating action and will not be installed until it is confirmed that an ongoing depletion of CCW inventory is being experienced. Therefore it cannot be an initiator of any new malfunctions or events, and prior NRC approval is not required.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
SS-SE-2016-0148-01: NESW to upper containment ventilation unit 1-HV-CUV-3 Activity
Description:
EC-54911:
The scope of this modification includes removal of the bellow welded to both the Non-Essential Service Water (NESW) supply line to 1-HV-CUV-3, and penetration 1-CPN-85 on the Auxiliary Building side of the penetration. Additionally, the protective cover associated with Containment Penetration 1-CPN-85 is being removed.
1-TM-16-19-R0:
This Temporary Modification removes two sections of piping in the NESW supply line to 1-HV-CUV-3 (Containment Upper Compartment Quadrant #3 Ventilation Unit), and alters piping configuration at the associated piping penetration 1-CPN-85, (Non-Essential Service Water Serving Vent HV-CUV-3 and RCP
- 3 Motor Cooler Containment Penetration).
The piping sections to be removed are on either side of penetration 1-CPN-85. Removal of this piping will require that a new Containment Integrity boundary be created on the NESW line on both the Containment and Auxiliary sides of the penetration in order to meet the double isolation requirement for process piping.
As a result of the removal of the piping and the change to the penetration, NESW flow to upper Containment ventilation unit 1-HV-CUV-3 will be taken out of service. The Containment Ventilation Unit Containment Isolation Valves 1-WCR-928, 1-WCR-929, 1-WCR-930 and 1-WCR-931 will be closed to isolate the NESW supply to and return from 1-HV-CUV-3.
Note that the scope of this Temporary Modification covers only one piping line; there are four total NESW pipes that pass through penetration 1-CPN-85. The other three are not in scope, and are not to be modified.
Summary of the Evaluation:
The Sites 50.59 Review Team identified that the 50.59 Screen performed, SS-SE-2016-0148-00, did not address the potential adverse impact arising from a reduction of redundancy/diversity. The UFSAR describes the upper containment ventilation units as consisting of four units, any three of which are sufficient to provide adequate cooling, with one in standby. Although this redundancy is described in the UFSAR, the reduction should have been considered adverse and had a 50.59 Evaluation performed.
This omission has been entered into the sites Corrective Action Program under AR 2016-7569.
Preparation of a 10 CFR 50.59 Evaluation is being tracked under this AR.