RNP-RA/14-0046, Report of Changes Pursuant to 10 CFR 50.59
| ML14133A007 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/29/2014 |
| From: | Peavyhouse S Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/14-0046 | |
| Download: ML14133A007 (6) | |
Text
Sharon W. Peavyhouse H. B. Robinson Steam Electric Plant Unit 2 DUKE Dir - Nuc Org Effectiveness ENERGYS Duke Energy Progress ENERGY.3581 West Entrance Road PROGRESS Hartsville. SC 29550 0: 843 857 1584 F: 843 857 1319 Sharon.Peavyhouse*,duke-energj-c 10 CFR 50.59(d)(2)
Serial: RNP-RA/14-0046 APR 2 9 2014 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Ladies and Gentlemen:
Duke Energy Progress, Inc., submits the attached report in accordance with 10 CFR 50.59(d)(2), "Changes, Tests, and Experiments," for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The report provides a description of changes that were implemented pursuant to 10 CFR 50.59 between. April 1, 2012,and April 1, 2014. A summary of the evaluation for each,item is also included in the attached report.
This letter contains no new Regulatory Commitments and no revision to existing Regulatory Commitments.
If you have any questions concerning this matter, please contact Richard Hightower, Manager -
Nuclear Regulatory Affairs at (843) 857-1329.
Sincerely, Sharon W. Peavyhouse Director - Nuc Org Effectiveness SWP/jmw Attachment cc:
Mr. K. M. Ellis, NRC Senior Resident Inspector
- Mr.,S. P. Lingam, NRC Project Manager, NRR
- Mr. V. M. McCree; NRC Region 1I Administrator u ii.
U. S. Nuclear Regulatory Commission Attachment to Serial: RNP-RA/1 4-0042 Page 1 of 5 H. B. Robinson Steam Electric Plant, Unit No. 2 Summary of Changes, Tests, and Experiments Evaluations performed for changes made in accordance with 10 VFR 50.59 during the time period from April 1, 2012 to April 1, 2014:
Evaluation No. 449684: (EC 79037) RNP Fire Safe Shutdown Analysis (SSA)
EC 79037 addresses an analysis of post-fire safe shutdown, and the analysis results require changes to engineering documents and procedures that define fire safe shutdown components, circuits and methodology. The analysis is a revalidation of the current safe shutdown capability and implements a more conservative criteria for determining credible fire induced circuit failure and component failure.
EC 79037 is strictly an analysis and no physical changes are made to any SCC's. Thus, the impact of a licensing basis fire event to safe shutdown circuits and components would be no different as a result of this EC. The implementation of more conservatism by the analysis of this EC changes the post-fire safe shutdown strategy for the various fire areas. However, the new manual actions have been assessed to be feasible. Thus, the overall result of EC 79037 is even greater assurance of achievement of post-fire safe shutdown in the event of a design basis fire in any one fire area of the plant.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 552116: (EC 83803) RCP Shutdown Seal The proposed activity replaces the Reactor Coolant Pump (RCP) Number 1 Seal Insert with a modified insert design. The modified insert will be installed on each Reactor Coolant Pump (RCP) as part of the controlled leakage seal assembly that restricts leakage along the pump shaft. The modified insert is called the Westinghouse SHIELD Shutdown Seal (SDS).
The SDS is designed to function only when exposed to an elevated fluid temperature downstream of the RCP number 1 seal, such as would occur as a result of the coincident loss of all thermal barrier heat exchanger cooling and number 1 seal injection cooling. SDS activation occurs over the temperature range of 250°F to 300 0F. In its installed and non-activated state, the SDS resides completely out of the normal seal injection and shaft seal leakage flow paths.
When activated as designed, the SDS limits RCP shaft leakoff to 1 gallon per minute (gpm) per pump.
The RCP shutdown seal does not affect nor is it associated with the RCP oil collection system, and therefore, does not affect RCP oil collection system exception. Installation of and normal functioning of the SDS does not interfere with the ability of the RCP to supply adequate core
U. S. Nuclear Regulatory Commission Attachment to Serial: RNP-RA/1 4-0046 Page 2 of 5 cooling flow during normal operation or RCP coastdown conditions, nor does it interfere with natural circulation cooling of the RCS.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 554558: (EC 86806) Permanent Cavity Seal Plate EC 86806 provides for the installation of a Permanent Cavity Seal Plate (PCSP) in the annular area between the Reactor Vessel (RV) seal ledge and the reactor cavity floor. The PCSP replaces the removable reactor cavity seal system currently used at RNP that must be installed and removed in the reactor cavity for each refueling cavity flooding and draining evolution.
The PCSP is a passive permanently installed device that will provide a leak tight seal of the annulus between the RV and reactor cavity to allow the cavity to be flooded for fuel transfer operations. The installation of the PCSP will reduce the outage time and personnel exposure involved with installing and removing the existing removable cavity seal system. The PCSP consists of two sub-assemblies; a support structure and a flexible sealing membrane. The support structure provides the support of the flexible sealing membrane and is designed to carry the weight of the canal water during refueling, loads generated by maintenance personnel and equipment, and loads generated during postulated seismic and accident conditions. The flexible membrane provides the water tight seal and allows for RV movement due to thermal expansion, pressurization, and seismic.
The PCSP is designed and examined in compliance with the general design guidelines and structural acceptance criteria of ASME Code Section III, 2007 Edition, including Addenda through 2008. Structural analyses performed to support the proposed activity demonstrate the PCSP support structure (including the access cover assemblies and port screws) and flexible membrane, respectively, comply with the structural requirements of the ASME B&PV Code Section III, Subsection ND, Table ND-3321-1. However, a fuel assembly dropped on the 1" unsupported horizontal length of the PCSP membrane may result in a corner weld failing and opening a crack approximately 0.25" x 0.38" in size. The resulting leak rate through the postulated crack in the PCSP membrane is 12.35 gal/min.
The probability that a dropped fuel assembly will fall on the 1 inch unsupported horizontal length of the PCSP flexible membrane is very low. However, since it is possible a FHA will result in leakage of the PCSP, manual action will be needed to provide make-up water to the reactor cavity using an existing plant operating procedure to maintain a minimum water level of 23 feet assumed in the FSAR Chapter 15 FHA analysis. Per FSAR Table 9.3.4-2 the capacity of each charging pump is 77 gpm. Therefore, it is shown the minimum make-up rate to the Refueling Cavity using one (1) charging pump exceeds the postulated leak rate due to a dropped fuel assembly and a water level of 23 feet will be maintained as credited in the FHA analysis and the regulatory acceptance criteria for offsite and HBR 2 Control Room TEDE doses will continue to be met. FSAR Section 15.7.4 will be updated to indicate a dropped fuel assembly may result in a leak in the PCSP that requires make-up water be provided to the Reactor Cavity through a minimum of one (1) charging pump to ensure a water level of 23 feet above the RV flange is maintained.
U. S. Nuclear Regulatory Commission Attachment to Serial: RNP-RA/1 4-0046 Page 3 of 5 NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 576518: Removal of HVS-1 Manual Operator Action Operations procedures (EOP E 0, EPP-7 Attachment 2, EPP Supplement L, AOP-005, AOP-01 3, and OP-906) currently direct that HVS-1 (Auxiliary Building Supply Fan) is secured within one hour of receipt of a Safety Injection signal, R-1 alarm, or following the occurrence of a fuel handling accident (FHA). This action was originally put into place to minimize in-leakage into the Control Room in the event that non-safety related exhaust fan HVE-7 (Auxiliary Building Exhaust Fan) fails with safety related supply fan HVS-1 remaining in service. This fan configuration would pressurize several areas in the Auxiliary Building served by HVS-1 and HVE-7 that are adjacent to the Control Room, potentially more than the Control Room when in emergency pressurization mode.
Robinson dose analyses assume elevated air in-leakage in the first hour of an accident (170 cfm) and reduced air in-leakage (100 cfm) for the remainder of the accident. Note that the Loss-of-Coolant Accident as outlined in RNP-M/MECH-1 740 is the most limiting accident for operator dose. The reduction in assumed air in-leakage following the first hour of the accident is a direct result of the manual operator action currently in place to secure HVS-1 within one hour of a Safety Injection signal, R-1 alarm, or following the occurrence of a FHA.
The proposed activity will remove the requirement to secure HVS-1 within one hour, and result in HVS-1 remaining in service for the duration of the accident. The changes to Control Room air in-leakage based on the removal of the manual operator action are within the bounds of RNP-M/MECH-1740. As shown via in-leakage testing in 2003 and 2012, Control Room air in-leakage rates with HVS-1 in service and HVE-7 out-of-service have historically been found to be less than 100 cfm. With RNP-M/MECH-1 740 assuming 170 cfm with HVS-1 in service and HVE-7 out-of-service, and 100 cfm with HVS-1 and HVE-7 out-of-service, the proposed procedure changes will not result in an increase in operator dose that deviates from the current analysis contained in RNP-M/MECH-1740.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 581311: (EC 76171) Steam Flow/Feedwater Flow Mismatch The proposed Engineering Change will delete Function 14, SG Water Level - Low, Coincident with Steam Flow/Feedwater Flow Mismatch, from Technical Specifications (TS) Table 3.3.1-1, Reactor Protection System Instrumentation. Robinson Nuclear Plant has installed median signal selector (MSS) modules during the most recent refueling outage. The installation of MSS modules enables the feedwater control system design to meet the requirements of IEEE-279 "IEEE Standard Criteria for Protection Systems for Nuclear Power Generating Stations" related to the potential for adverse control and protection system interactions and eliminates the need for the Steam Generator (SG) Water Level - Low Coincident with Steam Flow/Feedwater Flow Mismatch Reactor Protection System reactor trip function to meet IEEE-279 criteria.
U. S. Nuclear Regulatory Commission Attachment to Serial: RNP-RA/14-0046 Page 4 of 5 The change will be accomplished by removing the bistables, comparators, indications, and wiring within the Hagan racks which support this trip in the RPS. Annunciators, computer points, status lights, and Hagen rack wiring interfaces to RPS will need to be reconfigured as well as internal RPS rack relays used for this trip function and its periodic testing will be modified/deleted.
EC 75690 does not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses since the SG Water Level - Low, Coincident with Steam Flow/Feedwater Flow Mismatch trip was not credited within the UFSAR Chapter 15 accident analysis. However, the crediting of the median signal select modules installed by EC 76171 is considered to be departure from the method of evaluation described in the FSAR used in establishing the design bases function and need for the SG Water Level -
Low, Coincident with Steam Flow/Feedwater Flow Mismatch trip.
Therefore, the change to the Technical Specifications and UFSAR, as described within EC 75690 DOES require a License Amendment and NRC approval PRIOR to implementation. This change was issued via Amendment 234 to Renewed Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant, Unit No. 2.
Evaluation No. 612464: Non-bounding Analysis of Record for RNP Cycle 29 Reload Design Elements of the Analyses of Record (AOR) were not bounding for the RNP Cycle 29 reload design. That is, some of the Cycle 29 results had less margin to the limit than the Cycle 28 results. Since this results in less margin to fuel failure criteria, the proposed change is considered to adversely affect reload design function.
The previous AOR and the Cycle 29 values for those events (more limiting) were compared and revealed an increase in margin for departure from nucleate boiling ratio and linear heat generation rate/fuel centerline temperature to the respective limits. However, the analysis results of the Cycle 29 core design and supporting safety analyses demonstrated that the requirements and acceptance criteria defined in the UFSAR are satisfied for Cycle 29 operation.
Therefore, the Cycle 29 reload design, with regard to the safety analysis, will continue to meet the plant licensing basis.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 613758: RADTRAD Computer Code Error Correction RADTRAD is the computer code used for the analysis of record (AOR) for the LOCA dose; therefore, this activity is a revision to an FSAR methodology that is used in the safety analyses.
An error was discovered in the version of RADTRAD that was used in the AOR described in UFSAR 15.6.5. The correction to this error resulted in analyzed dose rates being higher (conservative) to those previously calculated. The correction in the RADTRAD code was to a design element. Per NEI 96-07 Revision 1, "In general, licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are
U. S. Nuclear Regulatory Commission Attachment to Serial: RNP-RA/1 4-0046 Page 5 of 5 essentially the same as, or more conservative than, previous results." Additionally, the results of the analysis remain within the limits of 10 CFR 50.67(b)(2) (i.e., less than 25 Rem TEDE for all cases). There were no physical changes made to any SSC. As previously stated, the results were conservative for all cases.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 622426: UFSAR Limiting Containment Analysis Replacement The proposed activity is replacing the limiting containment analyses in UFSAR Section 6.2.1,"
Containment Functional Design." Replacing the containment analyses will include the evaluation of dose consequences and instrument uncertainties affected by the containment response.
The Mass and Energy release continues to be provided by Westinghouse. The change in input data is justified by calculating acceptable results for containment temperature and pressure.
The inadvertent containment spray event continues to have acceptable results. The switch to a plant specific basis for decay heat is justified as a refinement that retains compliance with the requirements of ANS Standard 5.1. For the containment temperature and pressure response to LOCA and Main Steamline Break (MSLB), the Westinghouse methodology is being replaced with a plant specific application of the NRC-approved GOTHIC methodology. On this basis, no License Amendment is required.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).
Evaluation No. 676989: Zinc Addition Into RCS EC 69883 will implement a zinc injection program at RNP. This will include the purchase and installation of a Radiological Solutions, Inc. zinc injection skid and associated piping and electrical conduit/wiring for tie-in to the existing Chemical and Volume Control System (CVCS).
The zinc injection skid will be located in the Boric Acid Batching Tank Room on the 246' elevation in the Reactor Auxiliary Building. The tie-in to the CVCS piping is located upstream of the Volume Control Tank (VCT) in the Reactor Coolant Filter Room on the 246'8" elevation in the Reactor Auxiliary Building.
NRC review and approval of these changes prior to implementation was not required based on the conclusion that the proposed changes did not trigger any of the eight criteria listed in 10 CFR 50.59(c)(2).