IR 05000261/2020011
| ML20356A076 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 12/21/2020 |
| From: | James Baptist NRC/RGN-III/DRS/EB1 |
| To: | Kapopoulos E Duke Energy Progress |
| References | |
| IR 2020011 | |
| Download: ML20356A076 (22) | |
Text
December 21, 2020
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000261/2020011
Dear Mr. Kapopoulos,
Jr.:
On December 15, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at H. B. Robinson Steam Electric Plant and discussed the results of this inspection with Ms. Nicole Flippin and other members of your staff. The results of this inspection are documented in the enclosed report.
No findings or violations of more than minor significance were identified during this inspection.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 05000261 License No. DPR-23
Enclosure:
As stated
Inspection Report
Docket Number: 05000261 License Number: DPR-23 Report Number: 05000261/2020011 Enterprise Identifier: I-2020-011-0019 Licensee: Duke Energy Progress, LLC Facility: H. B. Robinson Steam Electric Plant Location: Hartsville, SC Inspection Dates: October 05, 2020 to December 15, 2020 Inspectors: N. Hansing, Mechanical Engineer G. Ottenberg, Senior Reactor Inspector M. Schwieg, Reactor Inspector S. Sandal, Senior Reactor Analyst Approved By: James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (programs) inspection at H. B.
Robinson Steam Electric Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
No findings or violations of more than minor significance were identified.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), inspectors were directed to begin telework. In addition, regional baseline inspections were evaluated to determine if all or portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on site. The inspections documented below met the objectives and requirements for completion of the IP.
REACTOR SAFETY
71111.21N.02 - Design-Basis Capability of Power-Operated Valves Under 10 CFR 50.55a Requirements POV Review (IP Section 03)
The inspectors:
a. Evaluated whether the sampled POVs are being tested and maintained in accordance with NRC regulations along with the licensees commitments and/or licensing bases.
b. Evaluated whether the sampled POVs are capable of performing their design-basis functions.
c. Evaluated whether testing of the sampled POVs is adequate to demonstrate the capability of the POVs to perform their safety functions under design-basis conditions.
d. Evaluated maintenance activities including a walkdown of the sampled POVs (if accessible).
- (1) AFW-V2-14B, Steam Driven Auxiliary Feedwater Pump Feedwater Discharge to Steam Generator B
- (2) MS-V1-8C, Steam Generator C Steam Supply to Steam Driven Auxiliary Feedwater Pump
- (3) FCV-1424, Motor Driven Auxiliary Feedwater Pump A Flow Control Valve
- (4) CC-749A, Component Cooling from Residual Heat Removal (RHR) Heat Exchanger A
- (5) RHR-744B, RHR Return to Cold Legs
- (6) CC-739, Excess Letdown Heat Exchanger Outlet
- (7) SI-860B, Containment Sump Recirc Suction
- (9) FCV-478, Feedwater Regulating Valve "A"
INSPECTION RESULTS
Very Low Safety Significance Issue Resolution Process: Potential Passive 71111.21 Single Failure Design Control Issue N.02 This issue is a current licensing basis question and inspection effort is being discontinued in accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process. No further evaluation is required.
Description:
The H. B. Robinson Unit 2 emergency core cooling system (ECCS) low head safety injection (LHSI) design is such that there are two independent containment sump suction lines provided from the containment sump strainer to the suction of the residual heat removal (RHR) Pumps. The RHR system is a dual-purpose system and can be operated for the LHSI function following an accident, and the RHR pumps would take suction from the containment sump during what is known as the recirculation phase of the accident. Each train of RHR/LHSI is located in a separate RHR pit. The pits are separated by a wall which, if overtopped due to a flood in one RHR pit, could result in the flooding of the adjacent RHR pit as well. Each RHR pit contains flood level instrumentation which is used to alert operators of the need to isolate a single fluid system passive failure in the room, which according to section 6.3.2.5.1 of the current H. B. Robinson Unit 2 Updated Final Safety Analysis Report (UFSAR), is to be assumed to occur during the ECCS recirculation phase without loss of the protective function. The term "passive failure" was defined in UFSAR reference 6.3.2-2, which is mentioned in UFSAR section 6.3.2.5.1. Reference 6.3.2-2 was GID/R87038/0013, Single Failure, which defines passive failure as:
"The structural failure of a static fluid system component which prevents that component from performing its design function. Specifically, a passive failure is defined as a break in a fluid pressure boundary resulting in abnormal leakage not exceeding 50 gpm. Such leak rates are consistent with limited cracks in pipes, sprung flanges, valve packing leaks or pump seal failures and are credited historically for being the basis of the 50 gpm commitment. This definition applies only to the ECCS portion of the SIS and also does not apply to radiological dose calculations."
Per licensee procedure, if a single fluid system passive failure occurs in the RHR pit during the recirculation phase of the accident, actions would be taken to close the innermost containment sump isolation valve, SI-860A or SI-860B, depending on which train experienced the passive failure in an attempt to stop the leakage into the RHR pit and to prevent the other train of LHSI from being affected.
Mod 792, completed in 1984, added a hole to the upstream discs (the disc facing the containment sump) of the SI-860A and B valves to relieve the pressure locking effect in order to ensure the valves will successfully open upon demand. However, installation of the 3/16 hole in the upstream disc potentially violated Robinsons single passive failure design basis. Both the current version of the UFSAR, section 6.3, and the version in place at the time Mod 792 was implemented stated that:
Two independent and redundant recirculation lines are provided. Each line has two motor-operated valves. Both valves are located adjacent to the containment penetration in the RHR pit such that the line outside the containment can be isolated in the event of a passive failure.
If either valve, SI-860A or SI-860B, which are located in the RHR pit experiences an assumed passive failure, such as a packing leak following ECCS suction swap to the containment sump, closing the valve would no longer isolate the line outside containment and leakage from the containment sump would continue into the RHR pit. This is because the 3/16 hole would now result in unisolable drainage of the containment sump into the RHR pit via the hole and the passive failure location. The licensee estimated this could result in leakage rates of approximately 2.4 to 3.9 gpm, if a large passive failure of the type described in GID/R87038/0013, Single Failure, was required to be applied. This amount of leakage would require beyond design basis operator intervention to ensure the other train of ECCS remains operable due to the potential for sump vortexing due to loss of water level in the containment sump, loss of available RHR pump net positive suction head due to loss of water level in the containment sump, and/or flooding of the RHR pit. Additionally, if the leakage could not be isolated, it would exceed the amount of radioactivity released assumed in control room operator and public dose assessments. It is assumed that if the passive failure occurs anywhere downstream of the innermost containment sump isolation valve, closure of the valve would be successful at isolating the leakage due to the passive failure such that this concern is limited to the assumption of a passive failure at the innermost containment sump isolation valve itself.
Without further substantial research, the inspectors are unable to determine if a single fluid system passive failure, such as packing failure in excess of the flow expected through a 3/16" orifice, applied at either the SI-860A or B valve after transition to ECCS recirculation phase following an accident was within the current licensing basis of the facility. This is due mainly to uncertainty regarding the intent of a September 17, 1969, Atomic Energy Commission (AEC) request for information (RFI) [ADAMS ML20058C568] regarding the passive failure design of these valves. Testing and inspection requirements were subsequently incorporated into the original technical specifications (TS) section 4.4.5, Post Accident Recirculation Heat Removal System [ADAMS ML20058C527], evidently as a result of the applicant's RFI response. Additional confusion was encountered regarding the licensees definition of the passive failure in the current UFSAR section 6.3.2.5.1, and the following statement in that section of the UFSAR: This definition applies only to the ECCS portion of the SIS and also does not apply to radiological dose calculations. The inspector noted that the leakage limit of 2 gallons per hour in section 4.4.5 of the original TS corresponded to the assumptions used in the station's radiological dose calculations. An understanding of the required application of the single passive failure criterion specific to these valves is needed to determine if the hole drilled into the discs of the SI-860A/B valves in 1984 was a violation of the design control requirements in Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50), Appendix B.
The inspectors were also concerned that the licensee did not consider the existence of the holes in the valves discs, introduced by Mod 792, when they responded to a March 17, 1989, 10 CFR 50.54(f) demand for information [ADAMS ML14188A655] regarding the H. B.
Robinson Unit 2 facility compliance with the single worst failure requirement of 10 CFR 50.46 and Appendix K on May 7, 1991 [ADAMS ML14174B064]. In their evaluation of the failure modes and effects of a passive failure of the valves resulting in 50 gallons per minute of external leakage, documented in 716M-P-100, RNP ECCS Single Failure Analysis, Revision 12, the licensee incorrectly concluded that the leakage during ECCS recirculation phase would be isolated per procedure and would not affect the opposite ECCS train. Consequently, the licensee did not describe this type of passive failure mode of the SI-860A and B valves as either "not applicable", similar to their description of other single failures the licensee deemed beyond their licensing basis, or as a more damaging single failure than those previously described in the UFSAR. Without an accurate understanding of the design basis requirement, the inspector cannot ascertain if the information provided to the NRC and its materiality to the NRC's decision-making represented a violation of 10 CFR 50.9 requirements.
The design and licensing basis information the inspectors reviewed is detailed in the Licensing Basis section, below.
Following the inspectors identification of the concern, the licensee entered the issue into their corrective action program as nuclear condition report NCR 2354646, and took action to provide means to electrically backseat the valves in the event of a packing failure, considered to be the most likely of the potential passive failures at the valves. The means to electrically backseat, as opposed to local manual backseating is necessary due to the high temperature and dose rates expected from the leaking fluid post-accident. Backseating of a valve that has experienced packing failure is expected by the licensee to limit the leakage from the ECCS fluid system into the RHR pit to acceptable values. These actions taken by the licensee would result in a reduction in the risk of the issue, even below that evaluated in the Significance section below. The licensee also detailed their understanding of the design and licensing basis of the passive failure requirements for the SI-860A and B valves in an evaluation, EC EVAL 418572, (ECCS Passive Failure) CLB & Design Basis Review Evaluation, Revision 1, and provided it to the inspector for their review. The licensee's evaluation concluded that the design change implemented in Mod 792 was implemented as intended by the designer from original construction (with hole drilled in the disc facing the containment sump side), and that it was believed that the periodic leak testing and visual inspection that had been incorporated into the original TS and maintained through the subsequent years would be used to address potential failures that result in leakage.
Licensing Basis: The NRC staff reviewed many current and historical regulatory requirements and regulatory correspondence related to the single failure criterion. Several other design documents provided by the licensee that were not included on the docket were also reviewed. The main documents reviewed are detailed below. As some of the documents were listed as "non-public" in the NRC's ADAMS system, not all relevant documents are excerpted below. The NRC staff, comprised of both regional inspectors and headquarters technical staff, concluded following a reasonable review of the requirements and documentation that the issue could be closed without immediate enforcement action and treated under the very low safety significance issue resolution process in light of the evaluated risk, the licensee's immediate actions to further reduce the risk, and the lack of clarity regarding the required passive failure assumptions to be applied to the containment sump isolation valves.
Current requirements and design and licensing basis references
- 10 CFR 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors: 50.46(b)(5) "Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core."
- 10 CFR 50, Appendix K, Appendix K to Part 50ECCS Evaluation Models: D. Post-Blowdown Phenomena; Heat Removal by the ECCS 1. Single Failure Criterion. An analysis of possible failure modes of ECCS equipment and of their effects on ECCS performance must be made. In carrying out the accident evaluation the combination of ECCS subsystems assumed to be operative shall be those available after the most damaging single failure of ECCS equipment has taken place.
- 10 CFR 50, Appendix B, Criterion III, Design Control: Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.
- UFSAR Section 6.3 (current version and version in place at the time of Mod 792 development): Two independent and redundant recirculation lines are provided. Each line has two motor-operated valves. Both valves are located adjacent to the containment penetration in the RHR pit such that the line outside the containment can be isolated in the event of a passive failure.
- UFSAR Section 6.3.2.5.1 Single Failure
Analysis:
During the ECCS recirculation phase, the failure definition is expanded to consider either an active failure or a passive fluid system failure without the loss of the protective function Reference 6.3.2-2 provides a detailed description of the H. B. Robinson Unit 2 single failure criteria. [Ref 6.3.2-2 is GID/R87038/0013, Single Failure]
- [not submitted to the NRC but referenced in the UFSAR] GID/R87038/0013, Single Failure, Rev. 0, Section 1.3.8 Passive Failure: The structural failure of a static fluid system component which prevents that component from performing its design function. Specifically, a passive failure is defined as a break in a fluid pressure boundary resulting in abnormal leakage not exceeding 50 gpm. Such leak rates are consistent with limited cracks in pipes, sprung flanges, valve packing leaks or pump seal failures and are credited historically for being the basis of the 50 gpm commitment. This definition applies only to the ECCS portion of the SIS and also does not apply to radiological dose calculations.
- [not submitted to the NRC- provided by the licensee] GID/R87038/SD02, SYS DBD-Safety Injection System Rev. 23, Section 4.3.2 ECCS Layout: Each LHSI pump shall be protected against all credible leaks originating within its compartment or in adjacent compartments where a flow path to its compartment exists. Credible leak sources; i.e.,
passive failures, consist of a malfunctioning RHR pump seal, flange gasket, or a valve with blown packing, etc. and "A sump pump can be located in each compartment or a sump and sump pump can be located external to the LHSI pump compartments to receive leakage flows through a system of floor drains and drain pipes. Should the sump pump fail, each compartment should be water tight to the extent which will allow for the accumulation of leakage for a period of 30 minutes without overflow into the other LHSI pump compartment. Thirty minutes is the time assumed that the operator would take to detect and isolate the leak."
Historical requirements and design and licensing basis references
- Pre-operation Final Safety Analysis Report dated May 8, 1970 (non-public): ADAMS ML20058C389, ML20058C429, ML20058C447
- Pre-operation Technical Specifications dated May 8, 1970 (non-public): ADAMS ML20058C527
- Atomic Energy Commission Request for Information dated September 17, 1969, and Applicant Response to the Request (non-public): ADAMS ML20058C568
- Safety Evaluation Report Regarding Review of the Exxon Nuclear Company Pressurized Water Reactor Generic ECCS Codes and the H. B. Robinson Reactor ECCS Evaluation Model for Conformance to All Requirements of Appendix K to 10 CFR 50 by the Office of Nuclear Reactor Regulation (non-public): ADAMS ML14175A904
- NLS-91-120, Emergency Core Cooling System (ECCS) Failure Modes and Effects Analysis (FMEA) Summary Information, dated May 7, 1991: ADAMS ML14174B064; "The FMEA has been completed; no single failure was identified which was more damaging than those previously described in the Updated Safety Analysis Report."
and "The failure mode and effects analysis (FMEA) worksheets identified 118 potential single point system vulnerabilities that required further evaluation to determine if the postulated component failures could impact ECCS readiness during pre-accident standby operations or could cause the ECCS not to achieve the required post-accident minimum performance requirements during the short-term ECCS injection phase or during the long-term ECCS recirculation phase. Table 2.1 presents a listing of the 118 potential areas of system vulnerability and the final disposition for each potential area of vulnerability. None of the 118 areas of ECCS single point vulnerability were classified as applicable ECCS single point failures for reasons noted in the table." and "ECCS leakage in the RHR pump pit will ultimately flood the pit, disable both RHR pumps and all ECCS operation during the recirculation phase. A modification installed during Refueling Outage 13 (1990-1991) provides for remotely detecting and isolating RHR pump pit ECCS and support system leaks with equipment qualified to operate in the post-accident environment. After the leak has been isolated, one RHR pump will remain operational."
Significance: A detailed risk evaluation was performed by a regional senior reactor analyst using SAPHIRE Version 8.2.2 and NRC SPAR model Version 8.57. The analysis assumed failure of the high-pressure and low-pressure recirculation functions in the event of a passive failure of one of the RHR containment sump suction valves, 860A or 860B. The dominant sequences involved a medium break loss of coolant accident accompanied by the inability to achieve secondary side cooldown with a subsequent failure of the high-pressure recirculation function due to passive failure of the containment sump suction valves. The analysis determined that if a performance deficiency was assumed to have existed, it would have resulted in an increase in core damage frequency of <1E-6/year, representing very low safety significance (Green).
Technical Assistance Request: A technical assistance request (TAR) was not initiated.
Corrective Action Reference: NCR
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 15, 2020, the inspectors presented the design basis assurance inspection (programs) inspection results to Ms. Nicole Flippin and other members of the licensee staff.
- On October 23, 2020, the inspectors presented the onsite debrief inspection results to Ernest J. Kapopoulos Jr. and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21N.02 Calculations 1619-0002-CALC- Evaluation of Robinson Valves AFW-V2-14A, B, C Rev. 0
001 (Velan 4 Gate Valves) Using EPRI PPM v3.3 and JOG
716M-P-100 RNP ECCS SINGLE FAILURE ANALYSIS Rev. 12
95134-C-12 SEISMIC WEAK LINK ASSESSMENT MOV(s): CC- Rev. 1
749A,B
A10243-C-001 THRUST CAPACITY OF REPLACEMENT YOKE FOR Rev. 0
MOVs CC-749A/B
CSD-EG-DEP- ELECTRICAL CALCULATION OF MOTOR OUTPUT Rev. 0
0101 TORQUE FOR AC AND DC MOTOR OPERATED
VALVES (MOVs)
RNP-C/EQ-1329 Weak Link Analysis MOV(s) MS-V1-8A, B, C (Steam Rev. 2
Admission ISOL to SDAFW Turbine)
RNP-C/EQ-1346 WEAK LINK ANALYSIS MOV(S) RHR-744A RHR-744B Rev. 4
RNP-C/EQ-1406 Weak Link Analysis MOV(s) AFW-V2-14A, B, C (SDAFW Rev. 0
Pump Discharge)
RNP-E-8.042 AC MOV Protection Evaluation Based on Computer Rev. 8
Program Motorguard
RNP-F/NFSA-0267 RADIOLOGICAL CONSEQUENCE ANALYSIS OF THE Rev. 4
AST LOSS OF COOLANT ACCIDENT
RNP-M/MECH- Air Operated Valve Required Thrust and Actuator Rev. 0
288 Capabilities for FCV-478, FCV-488, and FCV-498
RNP-M/MECH- SET-UP CALCULATION FOR MOV CC-749A Rev. 13
1409
RNP-M/MECH- SET UP CALCULATION FOR MOV RHR-744B Rev. 11
1438
RNP-M/MECH- Evaluation of Static and Dynamic Test Data for RHR- Rev. 3
1472 744B
RNP-M/MECH- Set Up Calculation for AFW-V2-14B Rev. 10
21
RNP-M/MECH- Set Up Calculation For MOV MS-V1-8C Rev. 12
Inspection Type Designation Description or Title Revision or
Procedure Date
23
RNP-M/MECH- DETERMINATION OF MOV VALVE FACTORS Rev. 5
1789
RNP-M/MECH- DETERMINATION OF MOV STEM FACTORS Rev. 1
1790
RNP-M/MECH- SYSTEM MISSION TIMES Rev. 2
1892
RNP/M-MECH- Containment Analysis Inputs Rev. 16
1651
Corrective Action AR 250367, AR
Documents 267179, AR
058275, AR
00505330-21, NCR
353470, NCR
565200
Corrective Action NCR 02351725 2020 NRC DBAI Cancelled EC should be removed from
Documents EQDP 2200
Resulting from NCR 02351964 2020 RNP NRC POV DBAI EQ EDB Update
Inspection NCR 02352010 2020 RNP NRC POV DBAI RHR-744B Pre-OP valve
leakage testing
NCR 02352029 2020 RNP NRC POV DBAI CC-749A/B JOG modules
NCR 02352251 2020 RNP NRC POV DBAI Plant Licensing Documents
NCR 02352471 2020 RNP NRC POV DBAI Exceeded MOV TEST FREQ
+ GRACE
NCR 02353305 2020 RNP NRC POV DBAI Revise MOV Stem Factor
Calculation
NCR 02353397 2020 RNP NRC POV DBAI Dimension inspection not
performed
NCR 02353421 2020 RNP NRC POV DBAI Revise V1-8C Set-Up Calc
Test Frequency
NCR 02353837 2020 RNP NRC POV DBAI Demonstrate CC-739 is
Adequately Sized
NCR 02353845 2020 RNP NRC POV DBAI MS-V1-8C has rust
NCR 02354122 2020 RNP NRC POV DBAI MOV Set Up Calculations
Inspection Type Designation Description or Title Revision or
Procedure Date
Need Revision
NCR 02354143 2020 RNP NRC POV DBAI: RNP-E-8.042 Time
correction
NCR 02354646 DBAI SI-860A/B, CV SUMP RECIRC SUCTION
PASSIVE FAILURES
NCR 02361021 2020 RNP NRC POV DBAI Commitment Change not
Recognized
PRR 02352945 TMM-009 Rev. for ASME OM Code Reference to
Section ISTC-3310
Drawings 5379-01484 RESIDUAL HEAT REMOVAL SYSTEM FLOW Rev. 49
DIAGRAM
5379-01882 SERIES 300 14"-S70WDD WELD ENDS O.S. AND Y. Rev. 5
DOUBLE DISC GATE VALVE WITH SMB-0
LIMITORQUE VALVE CONTROL
5379-04375 Model D100-100 Operator 3" 150lb USA Std. Valve Rev. 3
Assembly
B-190628 Steam Driven FWP Back-Up Controls Rev. 18
SH00627
B-190628 Steam Driven FWP Valves V1-8A, V1-8B and V1-8C Rev. 11
SH00631E Computer Input
B-190628 V1-8C, SDAFW Pump Steam Isolation Rev. 14
SH00633A
B-190628 V2-14B, SDAFW Pump Discharge to S/G B Rev. 16
SH00648
B-190628 Control Wiring Diagram Rev. 5
SH00657
CWD B-190628 CC-749A, RHR Heat Exchanger A Cooling Water Outlet Rev. 11
CWD B-190628, RHR 744B, RHR Loop to RCS Cold Leg Rev. 19
SH 221
FDW-2 FDW-2 Field Inspection Isometric dated
2/27/1984
G-190188 Reactor Building Sections Rev. 16
HBR2-08436 4- 900# - V510 Globe Valve Schedule 80 Butt Weld Rev. 2
Ends With M1 Manual HWhl Act
Inspection Type Designation Description or Title Revision or
Procedure Date
HBR2-08709 Limitorque Wiring Diagram Rev. 1
HBR2-08717 2900 Socket Ends Carbon Steel Double Disc Gate Rev. 5
Valve with SMB-000-2 Limitorque Actuator
HBR2-10982 16" FIG 47 1/2 x C W.E.O.S Gate Valve with SMB-1 Rev. 3
Limit Unit
HBR2-12712 Forged Bolted Bonnet Primary Nuclear Motor Operated Rev. 0
SH00001 Gate Valve
HBR2-14189 4 900# Gate Valve With Limitorque Operator Rev. 0
Sheets 1-3
Engineering EC 279457 C,EE,Q2, Various, Grease, MOV Long Life, dated
Changes Forsythe/Cromptom,JT/JB MOVLL Grade 1 08/26/2015
EC 299154 MS-V1-8A/B/C Actuator Gear Ratio Change Rev. 0
EC 400486 Hot Short Failure Protection for MOVs AFW-V2-14A/B/C Rev. 5
M-1017 RHR modification Rev. 1
M-792 Modification of valves SI-860 A+B; SI-862A; SI-863A+B; Rev. 0
SI-865C; SI-891C+D; and RHR-759A+B
Mod M-1017 Plant Modification Field Rev.
No. 1
Engineering 730.1.140 ITT General Controls Post Seismic Functional Test dated
Evaluations Engineering Report 05/17/1983
EC 418513 EVALUATION OF ACCUMULATED LEAKAGE IN THE Rev. 0
RHR PUMP PIT RESULTING FROM PACKING
FAILURE IN SI-860B DURING LOCA LONG TERM
RECIRCULATION
EC 47656 ESR 97-00328 Rev. 006 Determination of MOV Rate-of- Rev. 6
Loading Factors
EC EVAL 418572 (ECCS PASSIVE FAILURE) CLB & DESIGN BASIS Rev. 0
REVIEW EVALUATION
MOV Post-Test MOV Post-Test Data Review Worksheet for Work Order dated
Data Review 20035512-06, Valve AFW-V2-14B 10/16/2018
Worksheet for
MOV Post-Test MOV Post-Test Data Review Worksheet for Work Order dated
Inspection Type Designation Description or Title Revision or
Procedure Date
Data Review 2232689-11, Valve MS-V1-8C 06/16/2015
Worksheet for
RNP-TR-003 MPR 1862, Part 3, Evaluation of Stem Thrust Rev. 0
Requirements for RHR-744A & RHR-7448 Using the
Miscellaneous Design Criteria for Internal Overpressurization Protection dated
for Double Disc Gate Valves 03/30/1970
SAFETY EVALUATION BY THE DIVISION OF dated
REACTOR LICENSING U.S. ATOMIC ENERGY 05/18/1970
COMMISSION IN THE MATTER OF
CAROLINA POWER AND LIGHT COMPANY H. B.
ROBINSON UNIT NO. 2 DOCKET NO. 50-261
ORDER FOR MODIFICATION OF LICENSE dated
2/27/1974
TECHNICAL SPECIFICATIONS FOLLOW-UP REVIEW dated
SECTION 4.4.3 Post Accident Recirculation Heat 11/21/1994
Removal System
H.B. Robinson TS section 4.4.3 Amendment 163
27-703-32 Manual Steam Valves, Gate- Instruction Manual Rev. 3
28-034-89 Nuclear Hydromotors NH91 thru NH98 Rev. 7
28-562-71 SMB Valve Controls Rev. 43
28-812-46 Velan Valve Instruction Manual for Manual Valves Rev. 12
Installation, Operation, and Maintenance
ADAMS SAFETY EVALUATION REPORT REGARDING dated
ML14175A904 REVIEW OF THE EXXON NUCLEAR COMPANY 09/11/1975
PRESSURIZED WATER REACTOR GENERIC ECCS
CODES AND THE H. B. ROBINSON REACTOR ECCS
EVALUATION MODEL FOR CONFORMANCE TO ALL
REQUIREMENTS OF APPENDIX K TO 10 CFR 50 BY
THE OFFICE OF NUCLEAR REACTOR REGULATION
ADAMS SUBJECT: ORDER IMPOSING A CIVIL MONETARY dated
ML14188A654 PENALTY 03/17/1989
Inspection Type Designation Description or Title Revision or
Procedure Date
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
ML14191A483 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 08/29/1990
REQUEST FOR LICENSE AMENDMENT -CONTROL
ROOM HABITABILITY
ADAMS XN-NF-83-72 REVISION 2 SUPPLEMENT 1 H.B. dated Jul.
ML14192A470 ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS 1984
REPORT, REVISION 2 DISPOSITION OF CHAPTER 15
EVENTS
ADAMS Chapter 5 UFSAR dated 1970
ADAMS Chapter 6 FSAR dated 1970
ADAMS Technical Specifications dated
ML20058C527 05/08/1970
ADAMS ADDITIONAL INFORMATION REQUIRED H. B. dated
ML20058C551 ROBINSON UNIT NO. 2 DOCKET NO. 50-261 11/05/1969
ADAMS CAROLINA POWER & LIGHT COMPANY REQUEST dated
ML20058C568 FOR INFORMATION DOCKET NO. 50 261 09/17/1969
.
ADAMS XN-NF-84-72 H. B. ROBINSON UNIT 2 LARGE BREAK dated Jul.
ML20093J554 LOCA-ECCS ANALYSIS WITH INCREASED 1984
ENTHALPY RISE FACTOR
CC-749A, MOV POST-TEST DATA REVIEW WORKSHEET dated
20015786-06, 3-7- 5/31/2017
CC-749A, WO Static Post Test Evaluation of GL 89-10 Program Rising dated
1011026-01, 6-1- Stem MOVs 7/4/2010
Contingency WO PROVIDE TEMPORARY POWER TO RHR-SMPPMP-A
20430760 01 (CONTINGENCY)
Contingency PROVIDE TEMPORARY POWER TO RHR-SMPPMP- B
WO20430761 01 (CONTINGENCY)
CPL-HBR-SPEC-5 CP&L Specification for Carbon Steel Motor Operable Rev. 1
Valves for Non-Radioactive Service
Inspection Type Designation Description or Title Revision or
Procedure Date
DBD-R87038- Design Basis Document Main Steam System Rev. 8
SD25
DBD-R87038- Design Basis Document Auxiliary Feedwater System Rev. 11
SD32
DBD/R87038/SD02 SYS DBD - SAFETY INJECTION SYSTEM Rev. 23
DBD/R87038/SD03 SYS DBD - RESIDUAL HEAT REMOVAL SYSTEM Rev. 11
DBD/R87038/SD13 COMPONENT COOLING WATER SYSTEM Rev. 13
DR 80.2 GENERIC QUALIFICATION OF SERIES EA180 LIMIT Rev. 1
SWITCHES WITH AN ELECTRICAL RECEPTACLE
AND CONNECTOR/CABLE ASSEMBLY
EQDP-2200 NAMCO EA 180 SWITCHES, EC 210 AND 290 Rev. 13
CONNECTOR ASSEMBLIES
ESR 00-00181 FCV Hydramotor supports dated
05/01/2001
ESR 00-00205 FCV-1424-AO dated
06/27/2001
GID/90-181/00/RCI GI DBD - REACTOR CONTAINMENT ISOLATION Rev. 15
GID/R87038/0013 GENERIC ISSUES DOCUMENT SINGLE FAILURE Rev. 0
HBR-02562 MS-V1-8C Purchase Order dated
10/27/1987
HBR-02766 PO for FCV-1424 dated
2/11/1982
L2-M-011 Specification for V1-8A, B, C, D, E, & F Carbon Steel Rev. 0
Gate Valves For Nonradioactive Service in the Auxiliary
Feedwater System
Letter File: NG-
3514 (R) Serial: DOCKET NO. 50-261 CONFORMANCE WITH 10/02/1974
NG-74-1122 10CFR50.46
LICENSEE EVENT POTENTIAL FOR LOSS OF RESIDUAL HEAT dated
REPORT (LER) REMOVAL CAPABILITY DUE TO PUMP FLOODING 05/04/1989
89-008-00
MOV Stem INEEL Presentation dated
Lubricant Aging 01/15/2003
Research
Inspection Type Designation Description or Title Revision or
Procedure Date
NGG-PMB-HYD-01 NGG Equipment Reliability Template, Hydramotor Rev. 0
Actuators
NLS-87-157 RESPONSE TO REQUEST FOR ADDITIONAL dated
INFORMATION REGARDING 10CFR50, APPENDIX J 9/14/1987
TEST PROGRAM EXEMPTION
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 5O-261/LICENSE NO. DPR-23 REPLY 07/15/1988
TO NOTICE OF VIOLATION AND ANSWER TO
NOTICE OF VIOLATION ENFORCEMENT ACTION 88-
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50-261/LICENSE NO. DPR-23 REPLY 04/14/1989
TO ORDER IMPOSING A CIVIL PENALTY
ENFORCEMENT ACTION 88-88 lOCFR50 APPENDIX
K VIOLATION
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50-261/LICENSE NO. DPR-23 04/19/1989
RESPONSE TO APPENDIX K SINGLE FAILURE
lOCFR50.54(f)
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50.261/LICENSE NO. DPR-23 05/19/1989
FOLLOW-UP RESPONSE TO APPENDIX K SINGLE
FAILURE lOCFR50.54(f) LETTER
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50-261/LICENSE NO. DPR-23 SINGLE 06/14/1989
ELECTRICAL FAILURES
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50-261/LICENSE NO. DPR-23 05/21/1990
CONTROL ROOM HABITABILITY TMI ITEM II.D.3.4
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 5O-261/LICENSE NO. DPR-23 05/07/1991
EMERGENCY CORE COOLING SYSTEM (ECCS)
FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
SUMMARY INFORMATION
Inspection Type Designation Description or Title Revision or
Procedure Date
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
DOCKET NO. 50-261/LICENSE NO. DPR-23 05/07/1991
EMERGENCY CORE COOLING SYSTEM (ECCS)
FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
SUMMARY INFORMATION
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. dated
- REQUEST FOR ADDITIONAL INFORMATION 01/14/1988
SAFETY INJECTION PUMP B AUTO TRANSFER
SCHEME
NRC-88-337 SUBJECT: NOTICE OF VIOLATION AND PROPOSE dated
IMPOSITION OF CIVIL PENALTY (NRC INSPECTION 06/15/1988
REPORT NOS. 50-261/88-03 AND 50-261/88-04)
NRC-89-190 SUBJECT: ORDER IMPOSING A CIVIL MONETARY dated
PENALTY 03/17/1989
NRC-91-041 SUBJECT: SINGLE FAILURE CRITERION, dated
ELECTRICAL SYSTEMS - H. B. ROBINSON STEAM 01/31/1991
ELECTRIC PLANT, UNIT No. 2 (TAC No. 72969)
NUREG-0138 STAFF DISCUSSION OF FIFTEEN TECHNICAL dated Nov.
ISSUES LISTED IN ATTACHMENT TO NOVEMBER 3, 1976
1976 MEMORANDUM FROM DIRECTOR, NRR TO
NRR STAFF
NUREG-0153 STAFF DISCUSSION OF TWELVE ADDITIONAL dated Dec.
TECHNICAL ISSUES RAISED BY RESPONSES TO 1976
NOVEMBER 3, 1976 MEMORANDUM FROM
NUREG/CP-0152, EPRI MOV Stem Lubricant Test Program, Frictional
Volume 5 Performance of Exxon Nebula and MOV Long Life in a
Stem Lubrication Application, John Hosler
NUS-8829-P-100 Design Input Requirements for Auxiliary Feedwater Rev. 4
System Control Modification
NUS-8829-P-200 Specification for Flow Control Valve and Electro- Rev. 1
Hydraulic Actuators
PLP-100 Technical Requirements Manual section 3.2.3 Revs. 0 and
Inspection Type Designation Description or Title Revision or
Procedure Date
PO 03053015 Rev. 2
PRR 02354707 APP-001 R069 Emergent Initiate Work Orders dated
10/24/2020
PRR02354708 EPP-24 R010 (Emergent, NCR 2354646) dated
10/24/2020
Rate of Loading 2003 MUG Presentation dated 2003
Flowserve Test
Program
RHR-744B, MOV POST-TEST DATA REVIEW WORKSHEET dated
212497-06, 10- 10/24/2018
2-18
SD-025 Main Steam System Rev. 9
TMM-032 MOTOR OPERATED VALVE PROGRAM Rev. 38
TMM-035 POST-TEST EVALUATION OF MOV PERFORMANCE Rev. 24
TMM-127 Post Test Evaluation of AOV Performance, CC-739 Rev. 7,
completed
9/30/2018
Procedures 8S19-P-101 Station Blackout Coping Analysis Report Rev. 11
AD-EG-ALL-1110 Design Review Requirements Rev. 9
AD-OP-ALL-0110 GENERAL EQUIPMENT OPERATING STANDARDS Rev. 3
AD-OP-RNP-0205 OPERATOR TIME CRITICAL ACTION PROGRAM Rev. 1
AD-PI-ALL-0100 Corrective Action Program Rev. 24
AOP-014 COMPONENT COOLING WATER SYSTEM Rev. 41
MALFUNCTION
AOP-020 LOSS OF RESIDUAL HEAT REMOVAL (SHUTDOWN Rev. 50
COOLING)
APP-001 MISCELLANEOUS NSSS Rev. 68
CSD-EG-DEP- MOTOR OPERATED VALVE STRUCTURAL Rev. 0
0359 EVALUATION
EGR-NGGC-0029 MIDAS-PGN MOV DESIGN CALCULATION Rev. 2
IMPLEMENTATION PROCEDURE
EGR-NGGC-0030 PGNTEST MOV TEST SETUP, EVALUATION AND Rev. 2
TRENDING
EGR-NGGC-0203 MOTOR OPERATED VALVE PERFORMANCE Rev. 17
Inspection Type Designation Description or Title Revision or
Procedure Date
PREDICTION, ACTUATOR SETTINGS, AND
DIAGNOSTIC TEST DATA RECONCILIATION
EOP-ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Rev. 2
EOP-ES-1.3 TRANSFER TO COLD LEG RECIRCULATION Rev. 1
EPP-24 ISOLATION OF LEAKAGE IN THE RHR PUMP PIT Rev. 9
EPP-24-BD EPP-24 BASIS DOCUMENT Rev. 9
OP-402 Auxiliary Feedwater System Rev. 105
OST-201-1 MDAFW System Component Test - Train A Rev. 38,
completed
08/05/2020
OST-206 Comprehensive Flow Test for the Steam Driven Auxiliary Rev. 65,
Feedwater Pump completed
09/08/2020
OST-207 Comprehensive Flow Test for the Motor Driven Auxiliary Rev. 60.
Feedwater Pumps completed
11/18/2018
OST-252-1 RHR SYSTEM VALVE TEST - TRAIN A Rev. 27,
completed
9/14/2020
OST-252-2 RHR SYSTEM VALVE TEST - TRAIN B Rev. 29,
completed
7/27/2020
OST-254 RESIDUAL HEAT REMOVAL SYSTEM AND RHR Rev. 42
LOOP SAMPLING SYSTEM LEAK TEST (REFUELING)
OST-258-1 RHR VALVE POSITION INDICATOR VERIFICATION - Rev. 11,
TRAIN A completed
2/12/2018
OST-258-2 RIIR VALVE POSITION INDICATOR VERIFICATION - Rev. 10,
TRAIN B completed
10/3/2018
OST-701-3 CC-739 INSERVICE VALVE TEST Rev. 9,
completed
7/3/2020
OST-707-3 INSERVICE VALVE POSITION INDICATOR Rev. 6,
Inspection Type Designation Description or Title Revision or
Procedure Date
VERIFICATION FOR CC-739 AND CC-832 completed
1/6/2020
OST-945 Auxiliary Feedwater Switch/Valve Position Verification Rev. 5
SP-1610 Flush of AFW-165, Motor Driven Aux Feedwater Pump Rev. 1
Discharge Flex Conn Iso
SP-837 Steam Driven AFW Pump Low Steam Pressure dated
Operation 11/13/1988
Work Orders WO 20035512-27,
1; WO 20035512-
01, M; WO 13386072-01, M;
M; WO 20035512-
EL;
EL; WO 20056710-
01, K; WO 212783-05, WO 20015786-06, WO 212497-06;
88ADBI1; WO 13326176-01; WO 20191560-01; WO 20186959
20