ML19347C970
ML19347C970 | |
Person / Time | |
---|---|
Site: | Summer ![]() |
Issue date: | 12/13/2019 |
From: | James Baptist NRC/RGN-II/DRS/EB1 |
To: | Stoddard D South Carolina Electric & Gas Co |
References | |
IR 2019010 | |
Download: ML19347C970 (13) | |
See also: IR 05000395/2019010
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
December 13, 2019
Mr. Daniel G. Stoddard
Senior Vice President and
Chief Nuclear Officer
South Carolina Electric & Gas Company (SCE&G)
Innsbrook Technical Center
5000 Dominion Blvd., Floor: IN-2SW
Glen Allen, VA 29060
SUBJECT: VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 - TRIENNIAL INSPECTION
OF EVALUATION OF CHANGES, TESTS AND EXPERIMENTS BASELINE
INSPECTION REPORT 05000395/2019010
Dear Mr. Stoddard:
On November 7, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at Virgil C. Summer Nuclear Station, Unit 1. On December 12, 2019, the NRC
inspectors discussed the results of this inspection with Mr. George Lippard and other members
of your staff. The results of this inspection are documented in the enclosed report.
One Severity Level IV violation without an associated finding is documented in this report. We
are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the
No NRC-identified or self-revealing findings were identified during this inspection.
If you contest the violation or the significance or severity of the violation documented in this
inspection report, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional
Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at
Virgil C. Summer Nuclear Station, Unit 1.
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
D. Stoddard 2
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. 05000395
License No. NPF-12
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
ML19347C970 SUNSI REVIEW COMPLETE FORM 665 ATTACHED
OFFICE RII:DRS/OB1 RII:DRS/EB1 RII:DRS/EB1 RII:DRS/EB1 RII:DRS/EB1
NAME M. Donithan G. Ottenberg M. Schwieg T. Su J. Baptist
DATE 12/ 13 /2019 12/ 13 /2019 12/ 13 /2019 12/ 13 /2019 12/ 13 /2019
E-MAIL COPY? YES NO YES NO YES NO YES NO
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number: 05000395
License Number: NPF-12
Report Number: 05000395/2019010
Enterprise Identifier: I-2019-010-0032
Licensee: South Carolina Electric & Gas Company (SCE&G)
Facility: Virgil C. Summer Nuclear Station, Unit 1
Location: Jenkinsville, SC
Inspection Dates: November 04, 2019 to November 07, 2019
Inspectors: M. Donithan, Operations Engineer
G. Ottenberg, Senior Reactor Inspector
M. Schwieg, Reactor Inspector
T. Su, Reactor Inspector
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 triennial inspection of evaluation of changes, tests and
experiments baseline inspection at Virgil C. Summer Nuclear Station, Unit 1, 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
Failure to Obtain License Amendment Regarding Containment Isolation Valve Technical
Specification Guidance
Cornerstone Significance Cross-Cutting Report
Aspect Section
Not Applicable NCV 05000395/2019010-01 Not Applicable 71111.17T
Open
The inspectors identified a Severity Level IV Non-cited Violation (NCV) of Section 59 of Part
50 of Title 10 of the Code of Federal Regulations (10 CFR 50.59) for the licensee's failure
to obtain Commission approval prior to making a procedure change that involved a change to
Technical Specification (TS) 3.6.4, Containment Isolation Valves.
Additional Tracking Items
None.
2
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.
REACTOR SAFETY
71111.17T - Evaluations of Changes, Tests, and Experiments
Sample Selection (IP Section 02.01) (28 Samples)
The inspectors reviewed the following evaluations, screenings, and/or applicability
determinations for 10 CFR 50.59 from November 4 through November 7, 2019.
(1) 50.59 Screening for AOP-304.4, Rev. 4, "ARG-4, Loss of All ESF AC Power While on
RHR"
(2) 50.59 Evaluation for ECR-50846K, Weld Repair Contingency for RV Head Inspection
(3) 50.59 Evaluation for ECR-50585U, A-Chiller Replacement
(4) 50.59 Evaluation for ECR-50884E, OPC Strategy for the GDC-17 Offsite Power
Sources
(5) 50.59 Screening for ECR-50890, External Flooding Protection
(6) 50.59 Screening for ECR-50906, KWH Meter Replacement
(7) 50.59 Screening for ECR-72470, Replacement of D.G. OBrien Hermetic Connector
with EGS Quick Disconnect
(8) 50.59 Screening for ECR-50753M, OCA Perimeter Upgrades - Radar Systems
(9) 50.59 Evaluation for ECR-50695E, EFW Flow Margin Improvement
(10) 50.59 Evaluation for Bases Revision Notice BRN-17-002 TS Bases 3/4 7.1.2 Change
of EFW Combined Flow of 380 Gallons
(11) 50.59 Evaluation for ECR-71781, RMA2 Guidance
(12) 50.59 Screening for ECR-50912, Modification to Platform SWP-1
(13) 50.59 Screening for ECR-50900, New Enclosure on Control Building Roof
(14) 50.59 Screening for EOP-1.0, Rev. 30, E-0, "Reactor Trip or Safety Injection"
(15) 50.59 Screening for EOP-1.1, Rev. 22, ES-0.1, "Reactor Trip Response"
(16) 50.59 Screening for ECR 50909A, RCP Oil Lift Pump Enclosure
(17) 50.59 Screening for ECR 70436, Control Room Differential Pressure Switches
(18) 50.59 Applicability Determination for ARP-045-XPN-2050, "Industrial Chiller Heat
Tracing Control Panel"
(19) 50.59 Applicability Determination for VCS-SAP-0705, "Site Switchyard Coordinator
Responsibilities"
(20) 50.59 Screening for AOP-112.1, Rev. 7, "ARG-2, Shutdown LOCA"
(21) 50.59 Evaluation ECR-50897 Replace Reactor Service Structure - Integrated Head
Assembly, 2017-001
3
(22) 50.59 Evaluation for GTP-302 Rev. 7A Change and CR 18-01032 and CR 19-02562
(23) 50.59 Screening BRN 17-001, The Basis 3/4.7.6 Is Being Revised to Clarify Which
Revisions or RG 1.196 and CEI 99-03 Is Applicable
(24) 50.59 Screening for ECR 70814, Revise Set Point for ITB00451 and ITB00452
(25) 50.59 Screening for ECR 50874V, Major Revision fpr RB Cooling Permanent Power
(26) 50.59 Screening for ECR 50905, Boron Concentration System Monitoring Sys
Elimination
(27) 50.59 Screening for ECR 72012, RWST Empty Alarm Set-point Change
(28) 50.59 Applicability Determination for ARP-016-XCP-6210-LCB1, Annunciator
Response Procedure
INSPECTION RESULTS
Failure to Obtain License Amendment Regarding Containment Isolation Valve Technical
Specification Guidance
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Severity Level IV Not 71111.17T
Applicable NCV 05000395/2019010-01 Applicable
Open
The inspectors identified a Severity Level IV Non-cited Violation (NCV) of Section 59 of Part
50 of Title 10 of the Code of Federal Regulations (10 CFR 50.59) for the licensee's failure
to obtain Commission approval prior to making a procedure change that involved a change to
Technical Specification (TS) 3.6.4, Containment Isolation Valves.
Description: In 1992, V.C. Summer requested a TS amendment from the NRC to remove
Table 3.6-1 from TS in accordance with the guidance in NRC Generic Letter (GL) 91-08,
"Removal of Component Lists from Technical Specifications," and the request was granted by
the NRC. After the amendment was granted, the licensee relocated the table to station
procedure GTP-302, Inservice Testing of Valves, revision 6A, dated March 29, 1993, and
also included other valves in the list in accordance with the guidance in GL 91-08. Prior to
the licensee's implementation of the procedure change as a result of the amendment, Table
3.6-1 was located in TS 3/4.6.4 and it identified the containment isolation valves (CIVs) to
which the TS applied. The guidance in GL 91-08 included the following description:
- "The list of containment isolation valves in the TS may not include all valves that are
classified as containment isolation valves by the plant licensing basis. Generally, the
FSAR identifies those valves that are classified as containment isolation valves. With
this TS change, the LCO, remedial actions, and surveillance requirements will apply
for all valves that are classified as containment isolation valves by the plant
licensing basis."
Because they were not previously included in Table 3.6-1, the licensee added the steam
generator power operated relief valves (SG PORVs) to the table which was now being
maintained in GTP-302. The SG PORVs function as closed system isolation valves designed
to meet the requirements of General Design Criterion (GDC) 57 in 10 CFR 50, Appendix A.
The licensee later updated GTP-302 in revision 7A, dated June 18, 1993, to include the
following 'Note (2)' which was applicable to the SG PORVs:
- "(2) Should this valve be found to be incapable of meeting its containment isolation
requirements, i.e., inoperable per Technical Specification 3/4.6.4, action statement
4
3.6.4.c may be satisfied by verifying that the connected closed system is intact inside
the Reactor Building."
TS 3.6.4 required that "Each Containment Isolation Valve Shall be OPERABLE.*" The TS
Actions (which included action statement 3.6.4.c) stated the following:
- "With one or more of the isolation valve(s) inoperable, maintain at least one isolation
valve OPERABLE in each affected penetration that is open, and either:
- a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
- b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated
automatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed
manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN
within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."
The inspectors observed that the as-approved TS action statements in TS 3/4.6.4,
Containment Isolation Valves did not contain an explicit provision in action statement 3.6.4.c
that allowed consideration of the closed system as being equivalent to a closed manual valve
or a blind flange in order to satisfy the action statement, and that Note (2) was intended to
provide appropriate action if that type of containment isolation valve was found to be
inoperable. Although the closed system would function as a passive containment barrier
similar to a closed manual valve or blind flange, the provision for its credit as such did not
exist in the approved TS. The inspectors discovered that Note (2) was added to GTP-302 in
revision 7A, and due to a revision process error did not receive a review in accordance with
10 CFR 50.59 to determine whether or not an amendment to technical specifications was
required. The inspectors determined that the licensee should have obtained Commission
approval prior to implementing the change in revision 7A to GTP-302 that included Note (2),
because it involved a change in the TS incorporated into the license for the V.C. Summer
station.
The inspectors further observed that licensee event report (LER) 2019-002-00, "Condition
Prohibited by Technical Specification 3.6.4" was submitted to the NRC on October 9, 2019,
regarding a failed SG PORV. The inspectors also noted that, prior to concluding an LER was
required, the licensee utilized the guidance contained in GTP-302, "Inservice Testing of
Valves" when they performed a past-operability review of the valve failure under action
number 14 to condition report (CR) CR-18-01032, and initially inappropriately concluded that
the failure did not constitute a condition prohibited by Technical Specifications (TS) because
the procedure allowed action statement 3.6.4.c to be considered satisfied if the closed system
inside containment was intact. Although operation of the station while the licensee may
consider action statement 3.6.4.c to be met would represent an alignment not previously
explicitly allowable by TS, the consequence of the licensee's failure to correctly consider the
change and request an amendment prior to implementing the change was considered to be
minor because the current alternate source term assumptions for radioactive release for the
events where the closed system inside containment is compromised bound the presumed
containment isolation valve failure because immediate release of the fission products with no
mitigation is assumed.
5
Corrective Actions: Upon discovery, the licensee entered the issue into their corrective action
program and indicated that Note (2) was not to be used to satisfy the action statement in TS 3.6.4.c.
Corrective Action References: CR-19-04328
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency. The licensee failed to perform a correct 50.59 screening and
safety evaluation when considering the changes to procedure GTP-302, done in
1993. Specifically, the addition of Note (2) was not originally considered within the 50.59
process when the procedure was revised, and had it been considered, the licensee should
have concluded that a TS amendment was required because it allowed an application of a TS
action statement that was not stated in the TS. The consequence of the licensee's
implementation of the change without receiving prior Commission review and approval was
considered minor because the alternate source term assumptions for radioactive release for
the events where the closed system inside containment is compromised bound the presumed
containment isolation valve failure because immediate release of the fission products with no
mitigation is assumed.
Enforcement: The ROPs significance determination process does not specifically consider
the regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter non-compliance.
Severity: The NRC's Enforcement Policy dated May 28, 2019, section 6.1.d. stated in part,
SL IV violations involve, for example: Violations of 10 CFR 50.59 result in conditions
evaluated as having very low safety significance (i.e., green) by the SDP." Further, the
NRC's Enforcement Manual, dated October 1, 2019, Part II, section 2.1.3, "Enforcement of 10 CFR 50.59 and Related FSAR Violations," subsection D.5 stated in part, "a. Violations of 10 CFR 50.59 will be considered more than minor and categorized at Severity Level IV if: the
activity or change (1) required prior Commission review and approval, and the licensee failed
to obtain Commission approval; (2) the consequence of the activity or change evaluated by
the SDP is minor, or of very low safety significance (i.e., Green); and (3) the NRC would have
likely approved the change." The inspectors determined that the change required prior NRC
review and approval, which was not obtained prior to implementing the change; the
consequence of the change evaluated by the SDP was minor; and the NRC would have likely
approved the change, if requested.
Violation: Section 59 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.59) required, in part, that The holder of a license authorizing operation of a production or
utilization facility may (i) make changes in the facility as described in the safety analysis
report, (ii) make changes in the procedures as described in the safety analysis report, and (iii)
conduct tests or experiments not described in the safety analysis report, without prior
Commission approval, unless the proposed change, test or experiment involves a change in
the technical specifications incorporated in the license or an unreviewed safety
question." Further, 10 CFR 50.59(c) required "The holder of a license authorizing operation
of a production or utilization facility who desires (1) a change in technical specifications or (2)
to make a change in the facility or the procedures described in the safety analysis report or to
conduct tests or experiments not described in the safety analysis report, which involve an
unreviewed safety question or a change in technical specifications, shall submit an
application for amendment of his license pursuant to 50.90." Contrary to the above, in 1993,
the licensee made a change to a station procedure that involved a change in the technical
6
specifications incorporated in the license and did not submit an application for an amendment
of the license pursuant to 50.90, as required by 10 CFR 50.59(c). Specifically, the addition of
Note (2) to procedure GTP-302, Inservice Testing of Valves changed the interpretation of
the TS as written but did not change the wording of the TS to match the intended usage of
Note (2).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 12, 2019, the inspectors presented the triennial inspection of evaluation of
changes, tests and experiments baseline inspection results to Mr. George Lippard and
other members of the licensee staff.
- On November 7, 2019, the inspectors presented the On site debrief inspection results to
Mr. George Lippard and other members of the licensee staff.
7
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.17T Calculations DC03290-007 RWST PIT Flood Wall Rev. 0
DC03290-008 Evaluation of Foam Seal, Drain Pipe Plugs, and Floor Mats Rev. 0
for External Flooding Event
DC03690-005 Evaluation for SWPH Pipe Rupture/Flooding Effects Rev. 2
Corrective Action CR-18-01032 While performing STP-121.002A, B PORV failed to open Dated
Documents 3/8/2018
CR-19-02562 Reportability of B SG PORV failing to open Dated
07/17/2019
Corrective Action CR-19-03964 ARP-016-XCP-6210-LCB1 Applicability Determination Check Dated
Documents Boxes Were Checked 11/06/2019
Resulting from CR-19-03972 GTP-302 May Not Be the Right Place for Note 2 Dated
Inspection 11/07/2019
CR-19-03973 Exceptions to Reg. Guide 1.75 Not Specified Dated
11/07/2019
CR-19-03974 50.59 Screening Question 3 Did Not Cover Adverse Effect of Dated:
Direct Trip to Feederer Breakers 11/07/2019
CR-19-04328 GTP-302 Addition of Note Without Licensee Amendment Dated
Request 12/05/2019
Engineering ECR 50897 Replacement Reactor Service Structure-Integrated Head Dated
Changes Assembly 02/01/2017
ECR 50905 Boron Concentration Monitoring System Elimination Dated
06/11/2018
ECR 70814 Revise Set-Point For ITB00451 and ITB00452 Dated
08/09/2019
ECR 72012 RWST Empty Alarm Setpoint Change Dated
02/16/2017
ECR-50585 A-Chiller Replacement Rev. U
ECR-50695 EFW Flow Margin Improvement Rev. E
ECR-50753 OCA Perimeter Upgrades - Radar Systems Rev. M
ECR-50884 OPC Strategy for the GDC-17 Offsite Power Sources Rev. E
ECR-50890 External Flooding Protection Rev. 0
8
Inspection Type Designation Description or Title Revision or
Procedure Date
ECR-50900 New Enclosure on Control Building Roof Rev. 0
ECR-50906 KWH Meter Replacement Rev. 0
ECR-71781 RMA2 Guidance Rev. 0
ECR-72470 Replacement of D.G. OBrien Hermetic Connector with EGS Rev. 0
Quick Disconnect
Engineering DC0295A-006 Qualification of CB Roof Enclosure Rev. 3
Evaluations
Miscellaneous 50.59 Evaluation ECR50695E, EFW Flow Margin Improvement Rev. 0
Report Log
Number 2016-
0001
Bases Revision BRN No.17-002 Dated
Notice 05/23/2017
CGE-15-70 Transmittal of LTR-PL-15-66, V.C. Summer Emergency Dated
Feedwater (EFW) Flow Margin Increase Engineering Impact 11/03/2015
Results Summary
D2018024 Incidental Submergence of Rosemount 3152N Pressure Rev. A
Transmitter with Curtiss-Wright Generation 1 EGS QDC
EGS-TR-913601- Nuclear Environmental Qualification Report of EGS 3/4 Inch Rev. B
01 Quick Disconnect Electrical Connector
Evaluation 14-10 EFW Mod ECR50695E Reliability Dated
05/09/16
FATR-079- Factory Acceptance Test Report for NLI HVAC Chiller Rev. 1
351025399-01
GTP-302 In-service Testing of Valves Rev. 7 Chg.
A
ICES # 434234 During the SG PORV Operability Test Header Power Relief Dated
Valve 04/03/2018
Procedure GTP-302 Rev 7A Inservice Testing of Valves Dated
Development 06/18/93
Form
QR-0798582-2 Qualification Report for 4" Ball Valve with Beck Actuator Rev. 0
Model 11-288-119306-04
9
Inspection Type Designation Description or Title Revision or
Procedure Date
RN 12-030 FSAR Revision Notice ECR50695B/C/E/G Dated
11/08/2017
RN 18-008 To Eliminate Boron Concentration Monitoring (BCMS) Dated
10/17/2019
RN No.16-030 Correcting Mislabeled Fire Damper Dated
03/09/2017
SAP-0107 USA 50.59 Resource Manual Rev. 3
Technical Removal of Table 3.6-1 Per Generic Letter 91-08 (TSP dated
Specifications 920003-0) 09/23/1992
Change Request
TSP 920003 Proposed Technical Specification Change Dated
09/23/1992
WCAP-7706 An Evaluation of Solid State Logic Reactor Protection in Dated
Anticipated Transients 07/1971
Procedures ARP-016-XCP- Annunciator Response Procedure Rev. 2
6210-LCB1
ECR 50874 Major Revision for RB Cooling Upgrade Dated
01/08/2015
GTP-302 Inservice Testing of Valves Revs. 6A, 7,
7A, and 17
OAP-109.1 Guidelines for Severe Weather Rev. 5
SAP-0139 Document Review and Approval Process Rev. 38
VCS-ERP-0119 ERU Flooding Response Rev. 1
10