View source for RNP-RA/17-0068, Response to NRC Request for Additional Information Related to License Amendment Request Regarding Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles

{{Adams
| number = ML17272A015
| issue date = 09/28/2017
| title = Response to NRC Request for Additional Information Related to License Amendment Request Regarding Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles
| author name = Kapopoulos E
| author affiliation = Duke Energy Progress, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000261
| license number = DPR-023
| contact person = 
| case reference number = RNP-RA/17-0068
| document type = Letter type:RNP, Response to Request for Additional Information (RAI)
| page count = 28
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:Ernest J. Kapopoulos, Jr.
H. B. Robinson Steam Electric Plant Unit 2 Site Vice President Duke Energy 3581 West Entrance Road Hartsville, SC 29550 0 . 843 8571704 F: 843 8571319 Ernie.Kapopoulos@duke-energy.com 10 CFR 50.90 Serial: RNP-RA/17-0068 SEP 2 8 2017 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES Ladies and Gentlemen:
By letter dated August 31, 2017, the NRC requested that Duke Energy Progress, LLC respond to a request for additional information (RAI) regarding Technical Specification Surveillance Requirement Frequencies to support 24-Month Fuel Cycles at H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2.
The Duke Energy Progress, LLC response to this request (General RAl-1, SRXB RAl-1 and SBPB RAl-1) is provided in Enclosures 1, 2 and 3 to this letter.
There are no regulatory commitments made in this submittal. If you have any questions regarding this submittal, please contact Mr. Tony Pile, Manager - Nuclear Regulatory Affairs, at (843) 857-1409.
I declare under penalty of perjury that the foregoing is true and correct.
Executed On:      'Z-fS' 5~ f''f~PJ'itl 2o 17 Sincerely, Ernest J. Kapopoulos, Jr.
Site Vice President

United States Nuclear Regulatory Commission Serial: RNP-RA/17-0068 Page 2 of 2 EJK/am : RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (GENERAL RAl-1) : RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (SRXB RAl-1) : RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (SBPB RAl-1) cc: NRC Regional Administrator, NRC, Region II D. Galvin, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP Unit No. 2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)
A Gantt, Chief, Bureau of Radiological Health (SC)
A Wilson, Attorney General (SC)

United States Nuclear Regulatory Commission  to Serial: RNP-RA/17-0068 19 Pages (including cover page)
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (GENERAL RAl-1)

General   BAI -  1 In Enclosure 1 of Generic Letter (GL) 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML031140501), it is stated, in part:
Licensees should evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small. Licensees should confirm that historical plant maintenance and surveillance data support this conclusion.
By letter dated April 3, 2017, as supplemented by letters dates April 3, 2017 and May 2, 2017, (ADAMS Accession Nos. ML17093A787, ML17093A796, and ML17122A223) Duke Energy Progress, LLC (the licensee) submitted a license amendment request (LAR) for H. B. Robinson Steam Electric Plant Unit No. 2 (Robinson). Attachment 8, "Non-Calibration Surveillance Failure Analysis" to the LAR lists the same three unique failure histories, as listed below, for a significant number of Surveillance Requirements (SRs):
: a. On 5/3/2007, CB-1 on Battery Charger A-1 failed to trip when power was removed from the charger.
: b. On 10/29/2008, input breaker CB1 on Battery Charger B-1 failed to trip on loss of alternating current voltage.
: c. On 6/20/2015, when placing the PZR HTR BACK-UP GROUP "A" switch to ON, the RTGB ON indication did not illuminate as expected.
The LAR states that the test procedure and preventive maintenance task implementing these SRs are very large and test a wide range of equipment. It is not clear how these failures are representative enough to demonstrate the lack of impact of the change in SR frequency for the wide range of and diversity of the SRs.
(1.) Clarify the shared characteristics of these SRs that allow them to be treated as a group.
(2.) Confirm that the same failure histories applies to all the SRs.

(3.) Alternatively, provide the appropriate failure histories for the SRs.
The SRs with the same unique failure histories are below.
TS 3.3.1 Reactor Protection System (RPS) Instrumentation SR 3.3.1.14 Perform TADOT. Note: Verification of setpoint is not required .
Table 3.3.1-1, Function 16: Safety Injection (SI) Input from Engineered Safety Feature Actuation System (ESFAS)
TS 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation SR 3.3.2.3 Perform MASTER RELAY TEST.
SR 3.3.2.5 Perform SLAVE RELAY TEST.
Table 3.3.2-1, Function 1.b: Safety Injection -Automatic Actuation Logic and Actuation Relays Table 3.3.2-1, Function 3.a.2: Containment Isolation - Phase A Isolation -Automatic Actuation Logic and Actuation Relays Table 3.3.2-1, Function 5.a: Feedwater Isolation -Automatic Actuation Logic and Actuation Relays SR 3.3.2.6 Perform TADOT. Note: Verification of setpoint is not required.
Table 3.3.2-1, Function 1.a: Safety Injection - Manual Initiation Table 3.3.2-1, Function 3.a.1: Containment Isolation - Phase A Isolation - Manual Initiation TS 3.3.3 Post Accident Monitoring (PAM) Instrumentation SR 3.3.3.3 Perform TADOT. Note: Verification of setpoint is not required.
Table 3.3.3-1, Function 9: Containment Isolation Valve Position TS 3.3.6 Containment Ventilation Isolation Instrumentation SR 3.3.6.3 Perform MASTER RELAY TEST.
SR 3.3.6.5 Perform SLAVE RELAY TEST.
Table 3.3.6-1, Function 2: Automatic Actuation Logic and Actuation Relays SR 3.3.6.6 Perform TADOT. Note: Verification of setpoint is not required .
Table 3.3.6-1, Function 1: Manual Initiation TS 3.3.8 Auxiliary Feedwater (AFW) System Instrumentation SR 3.3.8.3 Perform TADOT.
Table 3.3.8-1 , Function 3: Automatic Actuation Logic and Actuation Relays

TS 3.4.9 Pressurizer SR 3.4.9.2 Verify capacity of required pressurizer heaters   is~  125KW SR 3.4.9.3 Verify required pressurizer heaters are capable of being powered from an emergency supply TS 3.5.2 Emergency Core Cooling Systems (ECCS) - Operating SR 3.5.2.4 Verify each ECCS automatic valve in the flow path that is locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal SR 3.5.2.5 Verify each ECCS pump starts automatically on an actual or simulated actuation signal TS 3.6.3 Containment Isolation Valves SR 3.6.3.5 Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal TS 3.6.6 Containment Spray and Cooling Systems SR 3.6.6. 7 Verify each containment cooling train starts automatically on an actual or simulated actuation signal TS 3.6.8 Isolation Valve Seal Water (IVSW) System SR 3.6.8.4 Verify each automatic valve in the IVSW System actuates to the correct position on an actual or simulated actuation signal TS 3.7.4 Auxiliary Feedwater (AFW) System SR 3.7.4.3 Verify each AFW isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. Note: Not applicable in MODE 4 when steam generator is being used for heat removal.
SR 3.7.4.4 Verify each AFW pump starts automatically on an actual or simulated actuation signal - Note 1. Not required to be performed for the steam driven AFW pump until 24 hours after ~ 100 psig in the steam generator. 2.
Not applicable in MODE 4 when steam generator is being used for heat removal.
TS 3.7.7 Service Water System (SWS)
SR 3.7.7.3 Verify each SWS pump and SWS booster pump starts automatically on an actual or simulated actuation signal.

TS 3.7.9 Control Room Emergency Filtration System (CREFS)
SR 3.7.9.3Verify each CREFS train actuates on an actual or simulated actuation signal.
TS 3.8.1 AC Sources - Operating SR 3.8.1 .8 Verify each DG rejects a load greater than or equal to its associated single largest post-accident load and does not trip on overspeed. Notes: 1. This Surveillance shall not be performed in MODE 1 or 2. 2. If performed with the DG synchronized with offsite power, it shall be performed at a power factors 0.9.
SR 3.8.1 .10 Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and: a. Ins 10 seconds after auto-start achieves voltage~ 467 V, and after steady state conditions are reached, maintains voltage~ 467 V ands 493 V; b. Ins 10 seconds after auto-start achieves frequency~ 58.8 Hz, and after steady state conditions are reached, maintains frequency ~ 58.8 Hz and s 61.2 Hz; c. Operates for~ 5 minutes; d. Permanently connected loads remain energized from the offsite power system; and e. Emergency loads are energized through the automatic load sequencer from the offsite power system. Notes: 1. All DG starts may be preceded by prelube period. 2. This Surveillance shall not be performed in MODE 1 or 2. 3. During periods when a diesel generator is being operated for testing purposes, its protective trips need not be bypassed after the diesel generator has properly assumed the load on its bus.
SR 3.8.1.9 Verify on an actual or simulated loss of offsite power signal: a.
Deenergization of emergency buses; b. Load shedding from emergency buses; c. DG auto-starts from standby condition and: 1. energizes permanently connected loads in s 10 seconds, 2. energizes auto-connected shutdown loads through automatic load sequencer, 3. maintains steady state voltage ~ 467 V and s 493 V, 4. maintains steady state frequency ~ 58.8 Hz ands 61.2 Hz, and 5. supplies permanently connected and auto-connected shutdown loads for~ 5 minutes - Notes: 1. All DG starts may be preceded by an engine prelube period. 2. This Surveillance shall not be performed in MODE 1, 2, 3, or4.
SR 3.8.1.15 Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ESF actuation signal: a. De-energization of emergency buses; b. Load shedding from emergency buses; and c. DG auto-starts from standby condition and: 1. energizes permanently connected loads in s 10 seconds, 2. energizes auto-connected emergency loads through load sequencer, 3. achieves steady state voltage ~ 467 V and s 493 V, 4.
achieves steady state frequency~ 58.8 Hz ands 61.2 Hz, and 5. supplies permanently connected and auto connected emergency loads for ~ 5 minutes
               - Notes: 1. All DG starts may be preceded by an engine prelube period. 2.
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. 3. During periods when a diesel generator is being operated for testing purposes, its

protective trips need not be bypassed after the diesel generator has properly assumed the load on its bus.
TS 3.9.3 Containment Penetrations SR 3.9.3.2 Verify each required containment ventilation valve actuates to the isolation position on an actual or simulated actuation signal.
General RAl-1 Response:
For the review of the acceptability of Surveillance failures, a study was performed to ensure that there is no evidence of repetitive surveillance test failures which would potentially result in a decrease in the system availability from the change to a 24-month surveillance interval.
To demonstrate this, a comprehensive review of the surveillance test history for 18-month Technical Specification required surveillances was performed. The study included a minimum of five (5) surveillance performances, providing a minimum of seven (7) years of data which has been proven acceptable by the USNRC in previous license submittals for interval extension [Hatch Nuclear Plant (ML012700177), DC Cook Nuclear Plant (ML041200298), and Oconee Nuclear Station (ML101330499)]. In most cases, seven (7) surveillance test performances were used for conservatism (minimum 10 years). During the study period, all 18-month surveillance interval test failures were identified and dispositioned. To simplify the evaluations necessary for this study, surveillance test failures were classified into categories which have been generically justified, with the exception of Category D, Unique Failures. The classification technique is shown in the graphic shown as 'Appendix A - Categorization Flow Chart'. All the failures were evaluated per an Engineering Change. The failures in Categories A, B, and C can be demonstrated to not change the conclusion that the impact on system availability, if any, will be small as a result of the change to a 24-month surveillance test interval. The reasons for this conclusion are listed in the flowchart. Each Category D failure was evaluated individually and collectively to ensure that repetitive failures of similar components tested by different surveillance tests would be identified. These two evaluations (failure categorization and review of unique-Category D failures) did not identify any repetitive failures which could potentially invalidate the conclusion that the impact on system availability and safety would be small from the change to a 24-month surveillance test interval.
Engineering Evaluation EC 407942 evaluated all failures. Attached is an excerpt (subset) from Appendix C of Engineering Evaluation - EC 407942. This excerpt includes the portion that evaluated all failures found for OST-163 [Safety Injection Test and Emergency Diesel Generator Auto Start On Loss OF Power and Safety Injection (Refueling)] which is categorized and dispositioned for the 24 month surveillance interval (Partial Appendix C). This section was chosen as an example as it contained the three (3) unique failures listed in the RAI and is one of the largest 18-month surveillance tests at H. B. Robinson Nuclear Plant (RNP). This type of analysis and documentation was utilized for all surveillances potentially affected by extension of the surveillance interval to 24 months.
If a surveillance test was failed, the failure was conservatively applied to the failure history of all SRs which the test credited, even when only one section of the test was failed. Most failures were classified into categories that can be generically justified and shown not to change the conclusion that the impact on system availability, if any, would be small. The exception was "unique" failures. To ensure that repetitive failures of similar components tested by different surveillance tests would be identified, a separate evaluation was performed of all "unique" failures. This evaluation concluded the impact, if any, on system availability would be small.
Since it was determined that there were no time based nor repetitive failures that would have a significant effect on system availability, it was not necessary to further analyze any smaller subset of the surveillance test.

In conclusion, a comprehensive review of surveillance test history was performed and all failures were systematically evaluated. This review did not identify any repetitive failures which could potentially invalidate the conclusion that the impact on system availability and safety would be small from the change to a 24-month surveillance test interval.

SUBSET OF APPENDIX C of EC 407942 TECHNICAL SPECIFICATION NON-CALIBRATION PMID-RQ FAILURE HISTORYEVALUATION

APPENDIX A - CATEGORIZATION FLOW CHART Define the Surveillance Test Failure Category A Since the failure does not impact any NO       safety function, the failure does not invalidate the conclusion that the increased surveillance interval will have a small, if any, impact on system availability.
Category B Since the failure would be detected by a YES       more frequent activity, there is no potential impact from an increase to the surveillance interval since the failure will not go undetected for a longer period of time.
CategoryC Since the failure was caused by a specific activity, which if performed requires the YES   performance of the corresponding test, the failure does not invalidate the conclusion that the impact, if any, from the increased test interval on system availability is small.
NO                                      Category D If the failure mechanism can be proven to not be related to a time-based NO   degradation and unique, then the failure does not invalidate the conclusion that the increased operating cycle will have a small, if any, impact on system availability. This conclusion should be supported by an additional review of the Category E                                                surveillance test history.
Perform additional evaluation to determine if the extended surveillance interval can be justified. If not, evaluate the possibility to perform the surveillance at power.

0001373801            OST-163 (RM) SI TEST & EOG AUTO START ON LOSS OF POWER &                                    SR 3.3.1.14(16)
SI&EDG                                                                                      SR 3.3.2.3(1.b)
SR 3.3.2.3(3.a.2)
Reviewed By BLC              # Performances: 7          # Failures: 15 SR 3.3.2.3(5.a)
SR 3.3.2.5(1.b)
SR 3.3.2.5(3.a.2)
SR 3.3.2.5(5.a)
SR 3.3.2.6(1.a)
SR 3.3.2.6(3.a.1)
SR 3.3.3.3{9)
SR 3.3.6.3(2)
SR 3.3.6.5(2)
SR 3.3.6.6(1)
SR 3.3.8.3(3)
SR3.4.9.2 SR3.4.9.3 SR3.5.2.4 SR3.5.2.5 SR3.6.3.5 SR3.6.6.7 SR3.6.8.4 SR3.7.4.3 SR3.7.4.4 SR3.7.7.3 SR3.7.9.3 SR 3.8.1.10 SR3.8.1.15 SR3.8.1.8 SR3.8.1.9 SR3.9.3.2 24 Month Justification:
Two failures are identified as event driven failures which are not indicative of a repetitive time based failure mechanism. Three failures are identified as unique failures which are not indicative of repetitive time based failure mechanisms. The other failures do not impact any safety function; therefore, the surveillance test history supports the proposed 24 month surveillance interval.
Failure Review:
Perf. Date            Fail Cat. Description of Failure                                             Justification of Failure 10/20/2005                A       Fuse was blown on the DS UPS when starting up the UPS            The identified failure(s) would not have prevented the IAW OP-602. Additional completion comments state:                performance of the required safety function of the

             "Fuse was blown in the DS UPS when starting up. Status       equipment. Therefore, this failure will have no impact on an light for CC-7 l 6B tested SAT." WO 1077253 l is             extension to a 24 month surveillance interval.
identified as written to address issue with a description of "PS A has failed on startup following OST-163." WO 1077253 l instructions state: "When charger was started up per OP-602, output voltage could not be adjusted greater than 123 VDC with an output current of 12 A.
Over the course of an hour, voltage was within the proper band. These same indications were present upon last startup when one phase of fuses had blown. Suspect same problem." WO 10772531 replaced blown KAA200 fuses in power supply "A" and returned power supply to service. The failure occurred during the restoration of the power supply following the completion of the testing of the required safety function.
10/20/2005 A Procedure completion comments states: "Flag will not         The identified failure(s) would not have prevented the reset due to failed relay." WR 10213292 written to           performance of the required safety function of the equipment.
document finding and states: "Following OST-163, APP         Therefore, this failure will have no impact on an extension to 036-H8 alarm is locked in with DS Bus parameters in          a 24 month surveillance interval.
Normal Range. Refer to WO #554666 from RF0-022.
Suspect same UV relay problem." WO 00772530 is referenced. WO 00772530 Task 01 completion comments state: "After comparing HBR2-7707, PIC-843 and I.L.41-201 for CV-2 relays, the determination was made that no problem exists with the relay 27-1-DSB. The relay has a built-in ICS seal-in contact which will not allow an automatic reset of the alarm (APP-036-H8). HBR2-7707 implies that the seal-in is lifted, but this requires modifying the relay by lifting a lead internally. Per Tim Halker, the alarm circuitry will remain as is and the alarm must be reset manually by momentarily recycling knife switch #IO (the red handled one) open and then closed.
This will de-energize the JCS coil (if normal DS Bus Voltage has been restored) causing the alarm to clear.
The ICS flag can then be reset." The failure occurred during the restoration of the relay following the completion of the testing of the required safety function.

10/20/2005 c Chart recorder for EDG B failed during loss of power to       This is an event driven failure in that the test equipment EI. The chart recorders are utilized as Measurement and       (chart recorder) being utilized to document the test results Test Equipment to record Emergency Diesel Generator           failed during the procedure performance. As a result, the parameters during a loss of power test. As a result of the    procedure acceptance criteria was not able to confirmed as failure of the chart recorder, the test results for Emergency acceptable. Therefore, this failure will have no impact on an Diesel Generator B were not annotated for review and          extension to a 24 month surveillance interval.
analysis to acceptance criteria.
5/3/2007   A The pink/blue ECCS status light for RHR-758 on Safety         The identified failure(s) would not have prevented the Injection Status Panel did not illuminate Pink as             performance of the required safety function of the equipment.
expected. RHR-758 was in the closed position per test         Therefore, this failure will have no impact on an extension to conditions but there was a problem with its position          a 24 month surveillance interval.
indication as documented in WR 10287007. Work Order 0 I 057 417 was generated for the repair of the issue. WO 01057417 Task 01 instructions state: "During Section 8.2 ofOST-163, the Pink/Blue ECCS status light for RHR-758 did not illuminate. Valve was in the closed position during the test." WO completion comments state: "HCV-758 was demanded shut (0.0%). FCV-605 was controlling flow at approximately 3200 gpm. The system pressure with 2 RHR pumps sometimes running was forcing (or could be forcing) HCV-758 to stay open slightly off its closed limit switch. Therefore, its indication would remain blue since the valve was slightly open due to system pressure. OST was run on the simulator and the program has the valve go shut during OST-163. It appears that the program doesn't allow for system pressure being high enough to force the valve open. The RO on the RTGB at the time agrees that the problem may be due to system pressure during the test with flow set at 3200 gpm and two RHR pumps sometimes running. Adjusted limit switches for proper indication. Completed SAT." The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.

6/24/2010 A Valve FCV- I 932A did not indicate full closed during        The identified failure(s) would not have prevented the OST- I 63. Per WR I 0438956, PCV- I 7 I 6, Instrument Air    performance of the required safety function of the equipment.
Isolation to CV, did not show closed. Indication on          Therefore, this failure will have no impact on an extension to ERFIS shows transition. Valve was verified to be closed      a 24 month surveillance interval.
locally. WR 10438956 written to address issue. WO I I 777330 states: "During performance ofOST-163, FCV-1932A did not show closed. Indication on ERFIS shows transition. Valve was verified to be closed locally.
Suspect limit switch issue." WO I 1777330 Task 01 completion comments state: Adjusted limit switch SAT."
The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.
6/24/2010 A Indicating light on Safety Injection Panel for SI-863B is    The identified failure(s) would not have prevented the not indicating Pink as expected per procedure. WR            performance of the required safety function of the equipment.
I 0438958 written to address issue. WO states: "During       Therefore, this failure will have no impact on an extension to performance of OST- I 63, it was noted that SI-863B, RHR     a 24 month surveillance interval.
Loop Recirc, pink/blue status lights did not show closed.
Attempts to change bulb did not return indication. RGTB indication does show closed." WO I I 77734 I referenced on WR. WO I I 77734 I Task 0 I completion comments state: "Verified fuses are good; suspect contacts dirty at limit torque valve. Inspected and cleaned limit switch contact for status light. Prior to cleaning the contact, terminal points 22 to 23 on Aux panel "JD" measured open. After cleaning the contact, terminal points measured closed (approx. zero ohms)." The failure was an "indication only" issue as a result of dirty contacts on the limit switch; the actual valve position was as expected.

5/3/2007   D CB-1 on Battery Charger A- I failed to trip when power       The identified failure is unique and does not occur on a was removed from charger. WO 1056863 written to              repetitive basis and is not associated with a time-based failure address issue. WO states that Relay K-6 is suspected to be   mechanism. Therefore, this failure will have no impact on an the problem. Fuses F-14 and F-15 were checked and            extension to a 24 month surveillance interval.
found to be satisfactory. Replaced the K-6 relay. Post       It should also be noted that Battery Charger A-1 was replaced in 2012 Maintenance Test (PMT) of the battery charger breaker        per EC 280186 (Master 277389) and work order 1747165 (CAS #
was perfonned on the same WO.                                11747165). This replacement was to replace aging battery chargers, implement the capability to restart all four safety-related batteries when energizing after a loss of power (Appendix R related), and to add load sharing capability to pennit parallel operation.
10/29/2008 A Indicating lights on CONTAINMENT ISOLATION                   The identified failure(s) would not have prevented the PHASE A panel for PRT TO ANALYZE VA-516 and VA-              perfonnance of the required safety function of the equipment.
553 SHUT (RC-553) did not display Pink as expected.          Therefore, this failure will have no impact on an extension to Completion comments state "RC-553 appeared to stroke         a 24 month surveillance interval.
closed but was just off the closed limit switch locally.
ERFIS indicated "transition" which confinns the valve stroked closed from the open position when required.
Appears to be a limit switch issue." WR I 0356264 states:
             "when signaled closed during OST-163, valve travelled closed but did not close completely or give closed indication." WO 11439851 written and states: "Adjusted limits for RC-553. Perfonned two complete strokes of valve and verified proper indications. Visually verified valve movement during the process. Operations notified." The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.

10/29/2008 A Indicating lights on SAFETY INJECTION status panel          The identified failure(s) would not have prevented the for SI-869 did not display Pink as expected. Completion     performance of the required safety function of the equipment.
comments state "SI-869 pink status was not illuminated. Therefore, this failure will have no impact on an extension to Valve was closed by indication at control switch.           a 24 month surveillance interval.
Changing the bulb did not correct the problem." WR 10356457 states: "SI-869 is closed and the SI status light on the RTGB does not indicate. Valve was closed by indication at the control switch." WO 11440701 written and states: "Pink and Blue status light is not lit. This ticket is for the Pink and Blue status light on the RTGB.
SI-869 is closed and the safety injection light on the RTGB does not light." WO Completion comments state:
             "Found fuse holder that connects to TB-1-3 loose.
Removed both circuit #2 fuses, one at a time, cleaned, checked continuity - good. Squeezed both fuse connection points to make a tight connection on the fuse. Reinstalled fuses . Checked indication in Control Room - SAT."
The failure was an "indication only" issue as a result of the loose fuse holder for the indicating lights; the actual valve position was as expected.
10/29/2008 D Input Breaker CB l on Battery Charger B-1 failed to trip    The identified failure is unique and does not occur on a on loss of AC Voltage. WR 10356292 initiated to             repetitive basis and is not associated with a time-based failure document finding. WO 11440102 completion comments           mechanism. Therefore, this failure will have no impact on an state: "Obtained new K-6 relay from stock and bench         extension to a 24 month surveillance interval. It should also tested SAT. Replaced relay in charger and functionally      be noted that Battery Charger B-1 was replaced in 2012 per tested SAT."                                                EC 280187 (Master EC 277389) and work order 1747166 (CAS # 11747166), This replacement was to replace aging battery chargers, implement the capability to restart all four safety-related batteries when energizing after a loss of power
{Appendix R related), and to add load sharing capability to permit parallel operation.

612412010 A PCV-1716 had dual indication when tested. WR               The identified failure(s) would not have prevented the 10435692 written to document finding. WO 11764784          performance of the required safety function of the equipment.
written for corrective maintenance. WO title: "Closed      Therefore, this failure will have no impact on an extension to limit switch seems to be out." WO instructions state:      a 24 month surveillance interval.
            "Fully closed ERFIS indicates transition and the status light on RTGB never indicated closed." Limit switch was adjusted SAT per the WO completion comments. WO Task 03 written to perform a PMT of the closed limit switch. PMT completed satisfactory. The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.
6/24/2010 A RC-553 had dual indication when tested. WR 10439106        The identified failure(s) would not have prevented the written to document finding. WO 11777320 written for       performance of the required safety function of the equipment.
corrective maintenance. WO title: "RC-553 shows dual       Therefore, this failure will have no impact on an extension to indication." WO instructions state: "RC-553 was closed     a 24 month surveillance interval.
during Phase A testing during the conduct ofOST-163.
Valve now shows dual indication." Limit switch was adjusted SAT per the WO completion comments. WO Task 02 written to perform a PMT ofRC-553 limit switch. PMT completed satisfactory. The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.

10/28/2013 A Not all indicating lights on CONTAINMENT                     The identified failure(s) would not have prevented the ISOLATION PHASE A panel displayed Pink as                    performance of the required safety function of the equipment.
expected. Comments state "Valve verified to move in          Therefore, this failure will have no impact on an extension to closed direction and locally indicate shut. Limit switch     a 24 month surveillance interval.
actuated approximately 15 minutes later." Comments on Attachment I0.7 state "PS-956C, PRZ LIQUID SPACE SAMPLE, indicated mid-position, valve was verified to be locally closed, did not reposition on reset signal." WR 11602322 and WO 13309283 written to document finding. WR 11602322 states: "PS-956C did not indicate closed on ERFIS during OST-163. Local indication at the panel in the Aux Bldg near the PASS panel indicated that the valve was in the intermediate position. This may indicate a problem with the limit switch for the valve."
WO 13309283 completion comments state: "Found PS-956C not indicating closed on ERFIS and switch was out of adjustment. Adjusted closed switch to closed position only. Left switch operating SAT. Closed position showing closed; open position showing open. Performed FME- closeout SAT." The failure was an "indication only" issue as a result of misalignment of the limit switch; the actual valve position was as expected.

6/20/2015 c Indicating lights on CONTAINMENT ISOLATION              This is an event driven failure in that the replacement of relay PHASE A panel for FP-249 and FP-258 did not display     CA23X during R0-29 (in 2014) and the subsequent incorrect Pink as expected. WR 11676193 (FP-249) and WR           wiring of the relay during replacement contributed directly to 11676192 (FP-258) written to address findings . WO      the As Found condition. Therefore, this failure will have no 13535461 initiated to perform troubleshooting. WO       impact on an extension to a 24 month surveillance interval.
completion comments state: "Performed troubleshooting per WO instructions. Found relay CA23X wired incorrectly. NCR 755233. Wires that were supposed to be landed on terminals 23 and 24 were landed on terminals 7 and 8. Task 02 of this WO was initiated for needed repairs. "Task 02 rewired relay CA23X per WO instructions IA W plant drawings per PM-049. Wires were found landed on terms 7 and 8 and were moved (re-landed) to terminals 23 and 24 per plant drawings.
CWDs 749, 751, and 5379-03235. Completed SAT."
Action Request 00755033 states that this relay was replaced during R0-29 on WO 2101372-0 I.

6/20/2015 D When placing the PZR HTR BACK-UP GROUP "A"                The identified failure is unique and does not occur on a switch to ON, the RTGB ON indication did not illuminate   repetitive basis and is not associated with a time-based failure as expected. Notes in procedure state: "52/2C did not     mechanism. Therefore, this failure will have no impact on an close from RTGB. WR 11676262 submitted.                   extension to a 24 month surveillance interval.
NCR755 I I 2. 52/2C closed from local control station in Rod Control Room." WO 13535588 written to address corrective action. WO completion comments state:
            "Removed control power fuses for 52/2C. Lifted cable C2 I 3 I G at Aux Panel "AF". Measured continuity between B to W, R to G, and 0 to U. All readings were OPEN.
Relanded cable C213 I G and reinstalled fuses. Racked 52/2C to test, cycled breaker SAT. Removed fuses, verified circuit at remote switch station from RTGB to breaker - all SAT. Checked breaker alarm switches - all SAT. Opened back-up "A" heater breakers, racked 52/2C in and had Ops cycle breaker three times - no problems.
Found wiring discrepancies in local HTR control box; indicating lights are energized in Remote and drawing B-190628 sht. 131 shows they should be OFF." The Correct Only Evaluation (COE) of NCR 00755112 states that although the breaker did not close from the RTGB, it would close when operated from the local position. The COE goes on to state that no issues were found during troubleshooting and the breaker was cycled several times from the RTGB with no issues. NCR 00755505 documents a similar condition in 2013 (RF0-28) on this same breaker. NCR 00755505 concluded that although there was a repeat failure of the breaker in both RF0-28 and RF0-29, they were not identical failures.

United States Nuclear Regulatory Commission  to Serial: RNP-RA/17-0068 3 Pages (including cover page)
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (SRXB RAl-1)

SRXB RAl-1 In Enclosure 1 of GL 91-04 it is stated, in part:
Licensees should evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small. Licensees should confirm that historical plant maintenance and surveillance data support this conclusion.
The staff noted that no failure histories were provided in Attachment 6 of the LAR for the following SR:
SR 3.4.1.4 Verify by precision heat balance that RCS total flow rate is;;:: 97.3 x 106 lbm/hr  of the LAR states in part for SR 3.4.1.4, 'There is no evaluation required for extension of this SR." However, no justification for this statement was provided. The NRC staff requests the licensee to provide the failure histories of the above mentioned SR in accordance with the GL 91-04, or to provide justification as to why it is not required.
SRXB BAI - 1 Response:
Correction for RAI: The information quoted in the RAI is from Attachment 6 instead of .
The following information is provided as a clarification to that already provided in Attachment 6:
Equipment used to perform this SR can be grouped into two categories: equipment calibrated online and equipment calibrated during refueling outages.
The following equipment used to perform this SR are calibrated online: Computer points .
RCF0400A, RCF0401A, RCF0402A, RCF0420A, RCF0421A, RCF0422A, RCF0440A, RCF0441A, RCF0442A, RCP0481A, FWT0418A, FWT0438A, FWT0458A, EMT0013, sensors TE-3004, TE-3005, and TE-3006, indicators Fl-414, Fl-415, Fl-416, Fl-424, Fl-425, Fl-426, Fl-434, Fl-435, Fl-436, FIT-1328A, FIT-13288, and FIT-1328C, and rack components FM-414, FM-415, FM-416, FM-424, FM-425, FM-426, FM-434, FM-435, FM-436, and PM-456.
The following equipment used to perform this SR are calibrated during refueling outages:
sensors TE-41181, TE-41102, TE-41103, TE-412C, TE-42101, TE-42102, TE-42103, TE-422C, TE-43101, TE-43102, TE-43283, TE-4320, Fl-476A, Fl-486A, Fl-496A, FT-414, FT-415, FT-416, FT-424, FT-425, FT-426, FT-434, FT-435, FT-436, and PT-456.
Online calibrations will remain at the same frequency as currently performed. Therefore, the interval between those calibrations and performance of SR 3.4.1.4 will not be affected. As an example, blowdown indicating transmitters FIT-1328C, FIT-13288 and FIT-1328A are calibrated online, on an 18-month frequency. Currently the largest duration between those calibrations and performance of SR 3.4.1.4 is 18-months nominal. Assuming SR 3.4.1.4 is revised to a 24-month surveillance frequency, FIT-1328C, FIT-13288 and FIT-1328A will remain on an 18-

month nominal, online calibration frequency; therefore, the duration between equipment calibrations and performance of SR 3.4.1.4 will never exceed 18-months nominal. Since there is no change in the interval between calibrations and performance of this surveillance requirement, for on line calibrated components, review of the historical surveillances is not required.
Instrumentation calibrated during refueling outages will remain on a refueling outage calibration schedule, which will be extended from 18-months to 24-months nominal. This will not result in an increase in duration of the interval between calibration of the equipment and performance of SR 3.4.1.4 since the SR is only performed when the plant is increasing reactor thermal power, following a refueling outage. The interval between calibration of outage related equipment and performance of SR 3.4.1.4 will not increase. As an example, pressurizer pressure transmitter PT-456 is currently calibrated at some point during each refueling outage per SR 3.3.1.10.
When the plant is increasing in reactor thermal power following each refueling outage, SR 3.4.1.4 is performed; therefore, the interval between equipment calibration and performance of the SR is no greater than the duration of the refueling outage. Assuming SR 3.4.1.4 is revised to a 24-month surveillance frequency, PT-456 will still be calibrated during each refueling outage per SR 3.3.1 .10 and SR 3.4.1.4 will still be performed when the plant is increasing reactor thermal power following each refueling outage; therefore, the duration between equipment calibration and performance of the SR will be no greater than the duration of the refueling outage (typically one or two months). Since there is no change in the interval between calibrations and performance of this surveillance requirement for outage related calibrated components, review of the historical surveillances is not required.

United States Nuclear Regulatory Commission  to Serial: RNP-RA/17-0068 4 Pages (including cover page)
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT FREQUENCIES TO SUPPORT 24-MONTH FUEL CYCLES (SBPB RAl-1)

SBPB RAl-1 (a) The LAR is proposing to change the test frequency in TS 5.5.17 "Control Room Envelope Habitability Program," item d from 18 to 24 months. However, the LAR did not propose a corresponding change in TS 5.5.17, item c.
The LAR is proposing to revise the frequency in TS 5.5.17, Item d, from 18 months to 24 months for taking measurements (e.g. Control Room envelope (CRE) pressure and Control Room emergency filtration system (CREFS) flow rate) on one of two trains of the CREFS on a staggered test basis. The two trains would then be tested on a 48 month or 4 year frequency TS 5.5.17, Item c references Regulatory Guide (RG) 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, dated May 2003 (ADAMS Accession No. ML031490664). TS 5.5.17 Item c (ii) requires assessing CRE habitability at the frequencies specified in Sections C.1 and C.2 of RG 1.197. RG 1.197, Section C.1, states that all CR Es should be tested on a performance-based periodic frequency consistent with Figure 1. RG 1.197, Figure 1, has an assessment frequency of 3 years. With the proposed change to TS 5.5.17, Item d, the two trains would then be tested on a 48 month or 4 year frequency instead of the 3 year frequency in RG 1.197.
The NRC staff requests the licensee to address this conflict in TS 5.5.17. The licensee should identify deviations taken from RG 1.197, if any, as part of its response.
SBPB RAl-1(a) Response:
RNP Technical Specifications 5.5.17 Item c (ii) requires assessing CRE Habitability at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197. Regulatory Guide 1.197, section C.1 states that all CRE's should be tested on a performance based periodic frequency consistent with Figure 1. Regulatory Guide 1.197 Figure 1 has an assessment frequency of 3 years following an integrated inleakage test. Figure 1 also shows that an integrated inleakage test is to be scheduled 3 years following an assessment. Following the Regulatory Guide 1.197 Figure 1 flow chart, the minimum possible years between program assessments is 6 years (with exception to the initial assessment performed 3 years after the initial baseline inleakage test) .
EST-023 (RNP Control Room Emergency Ventilation System Filter and Performance Testing) is currently performed every 18 months and satisfies RNP TS 5.5.17 item d for testing adjacent area pressures to the Control Room. Note, EST-023 is NOT the integrated inleakage test required by RG 1.197. EST-023 is a test required by RNP TS 5.5.11 as a part of the Ventilation Filter Testing Program (VTFP). During the test, pressures of external areas adjacent to the Control Room envelope are measured while the system is in emergency pressurization mode after verification of proper air flow, and compared to Control Room envelope pressures. These adjacent areas to the Control Room should be at a lower pressure to ensure air is not leaking into the Control Room envelope. As stated in RNP TS 5.5.17 item d, the emergency pressurization trains are staggered. 'A' Train Control Room Emergency Filtration System is tested in EVEN number outages, and 'B' Train Control Room Emergency Filtration System is tested in ODD number outages. See section 7.4 of EST-023 for further details.
By extending the frequency of testing adjacent area pressures with each Control Room emergency pressurization train from 3 years (again, due to alternating frequency) to 4 years,

data would still be collected (for both trains) in that 6 year time frame that the program assessment is required to evaluate per RNP TS 5.5.17 item d. That data will be compared to, and trended against, historical adjacent area pressures (recorded by historical EST-023 performances) to determine if any Control Room Envelope deficiencies are apparent.
Therefore, no deviations to Regulatory Guide 1.197 specified frequencies outlined in sections C.1, C.2, or Figure 1 will be taken.

SBPB RAl-1 (b) While TS 5.5.17 references RG 1.197 in TS 5.5.17, the licensee has not included RG 1.197 in Section 1.8, "Conformance to NRC Regulatory Guides," of the Robinson Updated Final Safety Analysis Report (UFSAR). Also, based on the response to part (a) of this question, the licensee may identify deviations from RG 1.197. The staff notes that this is contrary to the approach the licensee took regarding RG 1.52 "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety- Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants" Revision 2, dated March 1978 (ADAMS Accession No. ML003740139). Regarding RG 1.52, for TS 5.5.11, "Ventilation Filter Testing Program," the LAR proposes to change in filter testing frequency from 18 months to 24 months as an exception to RG 1.52. The LAR describes the exception in the proposed change to TS 5.5.11. In addition, in LAR Attachment 5, "Summary of License Commitments," the licensee made a commitment that UFSAR Section 1.8 will be modified to include the exception to RG 1.52 conformance.
Please clarify if conformance to RG 1.197, including any deviations, as applicable, will be included in Section 1.8 of the UFSAR?
SBPB RAl-1(b) Response:
RNP currently conforms to Regulatory Guide 1.197 as documented in RNP Technical Specifications 5.5.17. No deviations will be taken from the current commitments. It is noted that Reg Guide 1.197 is not currently documented in the RNP UFSAR.}}