ML19347C970

Triennial Inspection of Evaluation of Changes Tests and Experiments Baseline Inspection Report 05000395/2019010

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

Enforcement Policy.

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

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