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:
Report Number:
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
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 Single Failure Design Control Issue 71111.21 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
Procedure
Type
Designation
Description or Title
Revision or
Date
71111.21N.02 Calculations
1619-0002-CALC-
001
Evaluation of Robinson Valves AFW-V2-14A, B, C
(Velan 4 Gate Valves) Using EPRI PPM v3.3 and JOG
Rev. 0
RNP ECCS SINGLE FAILURE ANALYSIS
Rev. 12
95134-C-12
SEISMIC WEAK LINK ASSESSMENT MOV(s): CC-
749A,B
Rev. 1
A10243-C-001
THRUST CAPACITY OF REPLACEMENT YOKE FOR
MOVs CC-749A/B
Rev. 0
CSD-EG-DEP-
0101
ELECTRICAL CALCULATION OF MOTOR OUTPUT
TORQUE FOR AC AND DC MOTOR OPERATED
VALVES (MOVs)
Rev. 0
RNP-C/EQ-1329
Weak Link Analysis MOV(s) MS-V1-8A, B, C (Steam
Admission ISOL to SDAFW Turbine)
Rev. 2
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
Pump Discharge)
Rev. 0
RNP-E-8.042
AC MOV Protection Evaluation Based on Computer
Program Motorguard
Rev. 8
RNP-F/NFSA-0267
RADIOLOGICAL CONSEQUENCE ANALYSIS OF THE
AST LOSS OF COOLANT ACCIDENT
Rev. 4
RNP-M/MECH-
288
Air Operated Valve Required Thrust and Actuator
Capabilities for FCV-478, FCV-488, and FCV-498
Rev. 0
RNP-M/MECH-
1409
SET-UP CALCULATION FOR MOV CC-749A
Rev. 13
RNP-M/MECH-
1438
SET UP CALCULATION FOR MOV RHR-744B
Rev. 11
RNP-M/MECH-
1472
Evaluation of Static and Dynamic Test Data for RHR-
744B
Rev. 3
RNP-M/MECH-
21
Set Up Calculation for AFW-V2-14B
Rev. 10
RNP-M/MECH-
Set Up Calculation For MOV MS-V1-8C
Rev. 12
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
23
RNP-M/MECH-
1789
DETERMINATION OF MOV VALVE FACTORS
Rev. 5
RNP-M/MECH-
1790
DETERMINATION OF MOV STEM FACTORS
Rev. 1
RNP-M/MECH-
1892
SYSTEM MISSION TIMES
Rev. 2
RNP/M-MECH-
1651
Containment Analysis Inputs
Rev. 16
Corrective Action
Documents
267179, AR
058275, AR
00505330-21, NCR
353470, NCR
565200
Corrective Action
Documents
Resulting from
Inspection
20 NRC DBAI Cancelled EC should be removed from
EQDP 2200
20 RNP NRC POV DBAI EQ EDB Update
20 RNP NRC POV DBAI RHR-744B Pre-OP valve
leakage testing
20 RNP NRC POV DBAI CC-749A/B JOG modules
20 RNP NRC POV DBAI Plant Licensing Documents
20 RNP NRC POV DBAI Exceeded MOV TEST FREQ
+ GRACE
20 RNP NRC POV DBAI Revise MOV Stem Factor
Calculation
20 RNP NRC POV DBAI Dimension inspection not
performed
20 RNP NRC POV DBAI Revise V1-8C Set-Up Calc
Test Frequency
20 RNP NRC POV DBAI Demonstrate CC-739 is
Adequately Sized
20 RNP NRC POV DBAI MS-V1-8C has rust
20 RNP NRC POV DBAI MOV Set Up Calculations
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Need Revision
20 RNP NRC POV DBAI: RNP-E-8.042 Time
correction
DBAI SI-860A/B, CV SUMP RECIRC SUCTION
PASSIVE FAILURES
20 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
DIAGRAM
Rev. 49
5379-01882
SERIES 300 14"-S70WDD WELD ENDS O.S. AND Y.
DOUBLE DISC GATE VALVE WITH SMB-0
LIMITORQUE VALVE CONTROL
Rev. 5
5379-04375
Model D100-100 Operator 3" 150lb USA Std. Valve
Assembly
Rev. 3
B-190628
SH00627
Steam Driven FWP Back-Up Controls
Rev. 18
B-190628
SH00631E
Steam Driven FWP Valves V1-8A, V1-8B and V1-8C
Computer Input
Rev. 11
B-190628
SH00633A
V1-8C, SDAFW Pump Steam Isolation
Rev. 14
B-190628
SH00648
V2-14B, SDAFW Pump Discharge to S/G B
Rev. 16
B-190628
SH00657
Control Wiring Diagram
Rev. 5
CWD B-190628
CC-749A, RHR Heat Exchanger A Cooling Water Outlet
Rev. 11
CWD B-190628,
SH 221
RHR 744B, RHR Loop to RCS Cold Leg
Rev. 19
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
Ends With M1 Manual HWhl Act
Rev. 2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
HBR2-08709
Limitorque Wiring Diagram
Rev. 1
HBR2-08717
2900 Socket Ends Carbon Steel Double Disc Gate
Valve with SMB-000-2 Limitorque Actuator
Rev. 5
HBR2-10982
16" FIG 47 1/2 x C W.E.O.S Gate Valve with SMB-1
Limit Unit
Rev. 3
HBR2-12712
SH00001
Forged Bolted Bonnet Primary Nuclear Motor Operated
Gate Valve
Rev. 0
HBR2-14189
Sheets 1-3
900# Gate Valve With Limitorque Operator
Rev. 0
Engineering
Changes
C,EE,Q2, Various, Grease, MOV Long Life,
Forsythe/Cromptom,JT/JB MOVLL Grade 1
dated
08/26/2015
MS-V1-8A/B/C Actuator Gear Ratio Change
Rev. 0
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;
SI-865C; SI-891C+D; and RHR-759A+B
Rev. 0
Mod M-1017
Plant Modification
Field Rev.
No. 1
Engineering
Evaluations
730.1.140
ITT General Controls Post Seismic Functional Test
Engineering Report
dated
05/17/1983
EVALUATION OF ACCUMULATED LEAKAGE IN THE
RHR PUMP PIT RESULTING FROM PACKING
FAILURE IN SI-860B DURING LOCA LONG TERM
RECIRCULATION
Rev. 0
ESR 97-00328 Rev. 006 Determination of MOV Rate-of-
Loading Factors
Rev. 6
EC EVAL 418572
(ECCS PASSIVE FAILURE) CLB & DESIGN BASIS
REVIEW EVALUATION
Rev. 0
MOV Post-Test
Data Review
Worksheet for
MOV Post-Test Data Review Worksheet for Work Order 20035512-06, Valve AFW-V2-14B
dated
10/16/2018
MOV Post-Test
MOV Post-Test Data Review Worksheet for Work Order
dated
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Data Review
Worksheet for
232689-11, Valve MS-V1-8C
06/16/2015
RNP-TR-003
MPR 1862, Part 3, Evaluation of Stem Thrust
Requirements for RHR-744A & RHR-7448 Using the
Rev. 0
Miscellaneous
Design Criteria for Internal Overpressurization Protection
for Double Disc Gate Valves
dated
03/30/1970
SAFETY EVALUATION BY THE DIVISION OF
REACTOR LICENSING U.S. ATOMIC ENERGY
COMMISSION IN THE MATTER OF
CAROLINA POWER AND LIGHT COMPANY H. B.
ROBINSON UNIT NO. 2 DOCKET NO. 50-261
dated
05/18/1970
ORDER FOR MODIFICATION OF LICENSE
dated
2/27/1974
TECHNICAL SPECIFICATIONS FOLLOW-UP REVIEW
SECTION 4.4.3 Post Accident Recirculation Heat
Removal System
dated
11/21/1994
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
Installation, Operation, and Maintenance
Rev. 12
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
dated
09/11/1975
SUBJECT: ORDER IMPOSING A CIVIL MONETARY
PENALTY
dated
03/17/1989
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23
REQUEST FOR LICENSE AMENDMENT -CONTROL
ROOM HABITABILITY
dated
08/29/1990
XN-NF-83-72 REVISION 2 SUPPLEMENT 1 H.B.
ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS
REPORT, REVISION 2 DISPOSITION OF CHAPTER 15
EVENTS
dated Jul.
1984
Chapter 5 UFSAR
dated 1970
Chapter 6 FSAR
dated 1970
Technical Specifications
dated
05/08/1970
ADDITIONAL INFORMATION REQUIRED H. B.
ROBINSON UNIT NO. 2 DOCKET NO. 50-261
dated
11/05/1969
CAROLINA POWER & LIGHT COMPANY REQUEST
FOR INFORMATION DOCKET NO. 50 261
.
dated
09/17/1969
XN-NF-84-72 H. B. ROBINSON UNIT 2 LARGE BREAK
LOCA-ECCS ANALYSIS WITH INCREASED
ENTHALPY RISE FACTOR
dated Jul.
1984
CC-749A,
20015786-06, 3-7-
MOV POST-TEST DATA REVIEW WORKSHEET
dated
5/31/2017
CC-749A, WO 1011026-01, 6-1-
Static Post Test Evaluation of GL 89-10 Program Rising
Stem MOVs
dated
7/4/2010
Contingency WO 20430760 01
PROVIDE TEMPORARY POWER TO RHR-SMPPMP-A
(CONTINGENCY)
Contingency
WO20430761 01
PROVIDE TEMPORARY POWER TO RHR-SMPPMP-B
(CONTINGENCY)
CPL-HBR-SPEC-5
CP&L Specification for Carbon Steel Motor Operable
Valves for Non-Radioactive Service
Rev. 1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
DBD-R87038-
SD25
Design Basis Document Main Steam System
Rev. 8
DBD-R87038-
SD32
Design Basis Document Auxiliary Feedwater System
Rev. 11
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
SWITCHES WITH AN ELECTRICAL RECEPTACLE
AND CONNECTOR/CABLE ASSEMBLY
Rev. 1
EQDP-2200
NAMCO EA 180 SWITCHES, EC 210 AND 290
CONNECTOR ASSEMBLIES
Rev. 13
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
Gate Valves For Nonradioactive Service in the Auxiliary
Feedwater System
Rev. 0
Letter File: NG-
3514 (R) Serial:
NG-74-1122
DOCKET NO. 50-261 CONFORMANCE WITH
dated
10/02/1974
LICENSEE EVENT
REPORT (LER)
89-008-00
POTENTIAL FOR LOSS OF RESIDUAL HEAT
REMOVAL CAPABILITY DUE TO PUMP FLOODING
dated
05/04/1989
MOV Stem
Lubricant Aging
Research
INEEL Presentation
dated
01/15/2003
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
NGG-PMB-HYD-01 NGG Equipment Reliability Template, Hydramotor
Actuators
Rev. 0
RESPONSE TO REQUEST FOR ADDITIONAL
INFORMATION REGARDING 10CFR50, APPENDIX J
TEST PROGRAM EXEMPTION
dated
9/14/1987
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 5O-261/LICENSE NO. DPR-23 REPLY
TO NOTICE OF VIOLATION AND ANSWER TO
NOTICE OF VIOLATION ENFORCEMENT ACTION 88-
dated
07/15/1988
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23 REPLY
TO ORDER IMPOSING A CIVIL PENALTY
ENFORCEMENT ACTION 88-88 lOCFR50 APPENDIX
K VIOLATION
dated
04/14/1989
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23
RESPONSE TO APPENDIX K SINGLE FAILURE
lOCFR50.54(f)
dated
04/19/1989
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50.261/LICENSE NO. DPR-23
FOLLOW-UP RESPONSE TO APPENDIX K SINGLE
FAILURE lOCFR50.54(f) LETTER
dated
05/19/1989
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23 SINGLE
ELECTRICAL FAILURES
dated
06/14/1989
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23
CONTROL ROOM HABITABILITY TMI ITEM II.D.3.4
dated
05/21/1990
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 5O-261/LICENSE NO. DPR-23
EMERGENCY CORE COOLING SYSTEM (ECCS)
FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
SUMMARY INFORMATION
dated
05/07/1991
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
DOCKET NO. 50-261/LICENSE NO. DPR-23
EMERGENCY CORE COOLING SYSTEM (ECCS)
FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
SUMMARY INFORMATION
dated
05/07/1991
- H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.
- REQUEST FOR ADDITIONAL INFORMATION
SAFETY INJECTION PUMP B AUTO TRANSFER
SCHEME
dated
01/14/1988
SUBJECT: NOTICE OF VIOLATION AND PROPOSE
IMPOSITION OF CIVIL PENALTY (NRC INSPECTION
REPORT NOS. 50-261/88-03 AND 50-261/88-04)
dated
06/15/1988
SUBJECT: ORDER IMPOSING A CIVIL MONETARY
PENALTY
dated
03/17/1989
SUBJECT: SINGLE FAILURE CRITERION,
ELECTRICAL SYSTEMS - H. B. ROBINSON STEAM
ELECTRIC PLANT, UNIT No. 2 (TAC No. 72969)
dated
01/31/1991
STAFF DISCUSSION OF FIFTEEN TECHNICAL
ISSUES LISTED IN ATTACHMENT TO NOVEMBER 3,
1976 MEMORANDUM FROM DIRECTOR, NRR TO
NRR STAFF
dated Nov.
1976
STAFF DISCUSSION OF TWELVE ADDITIONAL
TECHNICAL ISSUES RAISED BY RESPONSES TO
NOVEMBER 3, 1976 MEMORANDUM FROM
dated Dec.
1976
Volume 5
EPRI MOV Stem Lubricant Test Program, Frictional
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
System Control Modification
Rev. 4
NUS-8829-P-200
Specification for Flow Control Valve and Electro-
Hydraulic Actuators
Rev. 1
PLP-100
Technical Requirements Manual section 3.2.3
Revs. 0 and
Inspection
Procedure
Type
Designation
Description or Title
Revision or
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
Flowserve Test
Program
2003 MUG Presentation
dated 2003
RHR-744B,
212497-06, 10-
2-18
MOV POST-TEST DATA REVIEW WORKSHEET
dated
10/24/2018
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
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
COMPONENT COOLING WATER SYSTEM
MALFUNCTION
Rev. 41
LOSS OF RESIDUAL HEAT REMOVAL (SHUTDOWN
COOLING)
Rev. 50
APP-001
MISCELLANEOUS NSSS
Rev. 68
CSD-EG-DEP-
0359
MOTOR OPERATED VALVE STRUCTURAL
EVALUATION
Rev. 0
EGR-NGGC-0029
MIDAS-PGN MOV DESIGN CALCULATION
IMPLEMENTATION PROCEDURE
Rev. 2
EGR-NGGC-0030
PGNTEST MOV TEST SETUP, EVALUATION AND
TRENDING
Rev. 2
EGR-NGGC-0203
MOTOR OPERATED VALVE PERFORMANCE
Rev. 17
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
PREDICTION, ACTUATOR SETTINGS, AND
DIAGNOSTIC TEST DATA RECONCILIATION
LOSS OF EMERGENCY COOLANT RECIRCULATION
Rev. 2
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
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
Feedwater Pump
Rev. 65,
completed
09/08/2020
OST-207
Comprehensive Flow Test for the Motor Driven Auxiliary
Feedwater Pumps
Rev. 60.
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
LOOP SAMPLING SYSTEM LEAK TEST (REFUELING)
Rev. 42
OST-258-1
RHR VALVE POSITION INDICATOR VERIFICATION -
TRAIN A
Rev. 11,
completed
2/12/2018
OST-258-2
RIIR VALVE POSITION INDICATOR VERIFICATION -
TRAIN B
Rev. 10,
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
Procedure
Type
Designation
Description or Title
Revision or
Date
VERIFICATION FOR CC-739 AND CC-832
completed
1/6/2020
OST-945
Auxiliary Feedwater Switch/Valve Position Verification
Rev. 5
Flush of AFW-165, Motor Driven Aux Feedwater Pump
Discharge Flex Conn Iso
Rev. 1
Steam Driven AFW Pump Low Steam Pressure
Operation
dated
11/13/1988
Work Orders
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