05000416/LER-2015-003, Regarding Technical Specification Surveillance on Primary Containment Isolation Valves

Regarding Technical Specification Surveillance on Primary Containment Isolation Valves

ML15344A266
Person / Time
Site:	Grand Gulf
Issue date:	12/10/2015
From:	Nadeau J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2015/00092 LER 15-003-00
Download: ML15344A266 (6)

LER-2015-003, Regarding Technical Specification Surveillance on Primary Containment Isolation Valves

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function
4162015003R00 - NRC Website

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~Entergy GNRO-2015/00092 December 10,2015 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Operations, Inc.

P. O. Box 756 Port Gibson, MS 39150 James Nadeau Manager, Regulatory Assurance Grand Gulf Nuclear Station Tel. (601) 437-2103

SUBJECT:

Dear Sir or Madam:

Licensee Event Report (LER) 2015-003-00 Technical Specification Surveillance on Primary Containment Isolation Valves Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29 Attached is Licensee Event Report (LER) 2015-003-00, Technical Specification Surveillance am Primary Containment Isolation Valves, which is a final report. This report is submitted in accordance with Title 10 Code of Federal Regulations 50.73(a)(2)(i)(B).

This letter contains no new commitments. Should you have any questions or require additional information, please contact James Nadeau at (601) 437-2103.

JJN/tmc Attachment: Licensee Event Report (LER) 2015-003-00 cc: (See Next Page)

GNRO-2015/00092 Page 2 of 2 cc:

u.s. Nuclear Regulatory Commission ATTN: Mr. A. Wang, NRRlDORL (w/2)

Mail Stop OWFN 8 B1 Washington, DC 20555-0001 u.S. Nuclear Regulatory Commission ATTN: Mr. Marc Dapas (w/2)

Regional Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment to GNRO-2015/00092 Licensee Event Report (LER) 2015-003-00

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 (11-2015) the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

13. PAGE Grand Gulf Nuclear Station, Unit 1 05000416 1 OF 3

4. TITLE Technical Specification Surveillance on Primary Containment Isolation Valves

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.

MONTH DAY YEAR N/A 05000 N/A FACILITY NAME DOCKET NUMBER 10 14 2015 2015-003-00 12 13 2015 N/A 05000 N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)

10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) 100 o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) cgj 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi)

~ 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in =

Initial Conditions:

At the time of the event, Grand Gulf Nuclear Station (GGNS) was operating in MODE 1 at 100 percent rated core thermal power. There were no systems, structures or components that were inoperable at the start of the event that contributed to the event.

Description of Event

Grand Gulf Nuclear Station Unit 1 Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.9 requires Primary Containment Isolation Valves (PCIVs) to be hydrostatically tested at 110% of accident pressure in accordance with the Inservice Testing Program. The Inservice Testing Program is defined in TS 5.5.6. While reviewing Local Leak Rate Testing (LLRT) Procedures for outage preparation on October 14, 2015, it was discovered the Local Leak Rate Test Low Pressure Water procedure was not revised as a result of the Extended Power Uprate (EPU). A step in the procedure states the base test pressure is 12.65 psig (110% of peak accident pressure). The value of 12.65 psig is 110% of the pre-EPU peak accident pressure of 11.5 psig. The peak containment accident pressure was revised on July 18, 2012 when the Nuclear Regulatory Commission (NRC) approved the GGNS, Extended Power Uprate, License Amendment Request. The base test pressure should have been revised to 16.28 psig, 110% of the post-EPU peak accident pressure of 14.8 psig. The current testing criterion remains at the post-EPU value of 16.28 psig.

Three valves were determined to have not been Local Leak Rate Tested at or greater than the required base test pressure of 16.28 psig. These three valves included two Refuel Water Transfer Pump Suction from Suppression Pool valves and a Residual Heat Removal (RHR) Sample Return Isolation valve. Upon discovery, the valves were declared inoperable and isolated in accordance with TS 3.6.1.3. The two Refuel Water Transfer Pump Suction valves were tested at or greater than the post-EPU base test pressure on October 15, 2015 and passed with zero leakage identified. The Residual Heat Removal Isolation valve was tested at or greater than the post-EPU base test pressure on October 17. 2015 and passed with zero leakage identified. After satisfactory testing was completed, all valves were declared operable.

The Refuel Water Transfer Pump Suction valves were previously tested on June 25, 2008 and August 14, 2008 respectively with the appropriate pre-EPU test pressure of 12.9 psig. The test passed with zero leakage identified.

The RHR Isolation valve was tested on April 15, 2008 with the appropriate pre-EPU test pressure of 13.0 psig.

The test passed with zero leakage identified. Therefore, the valves were demonstrated operable at the 2008 tests.

Next testing of the valves was in 2013. The valves were inappropriately tested at or greater than the pre-EPU base test pressure and passed with zero leakage identified.

Cause of Event

The cause was determined to be inadequate incorporation of testing criteria into Surveillance Testing Procedure 06-0P-1M61-V-0003, Local Leak Rate Test Low Pressure Water. An Engineering Change was completed which evaluated the impacts of EPU on Containment Leakage Rate Testing. However, it did not identify Local Leak Rate Test Low Pressure Water as a procedure requiring change. This change was missed due to failure of personnel to follow the Engineering Change Process per station procedures. Necessary reviews were not completed and the required procedure change was not identified.

Extent of Condition:

Ten valves were identified that should have been hydrostatically tested at or above 16.28 psig per post-EPU conditions on a schedule specified by the Inservice Testing Program. Three of those valves were not tested at or above 16.28 psig during their most recent regularly scheduled surveillance test. These three

NRC FORM (11-2015) 366A U.S. NUCLEAR REGULATO~YCOMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

1. FACILITY NAME Grand Gulf Nuclear Station, Unit 1

2. DOCKET 05000416 YEAR I

3. LER NUMBER SEQUENTIAL NUMBER 2015-003-00 I

REV.

NO.

Valves are being reported under this Licensee Event Report (LER). The remaining seven valves were tested at or above 16.28 psig during their most recent regularly scheduled surveillance test. These seven valves passed the surveillance testing with satisfactory results. Therefore, there are no adverse conditions with the hydrostatic testing of the remaining seven valves.

Twenty-two valves are High/Low Interface Valves and are required to be hydrostatically tested at or above 50 psig.

The most recent water LLRT was reviewed for each valve and the test was conducted at or above 50 psig as required. Therefore, there are no adverse conditions with the hydrostatic testing of these High/Low Interface valves.

Grand Gulf Nuclear Station documented all pneumatically tested valves have been tested at or above the post-EPU accident pressure with the exception of four valves. These four valves do not require pneumatic LLRT at the post-EPU accident pressure until their next regularly scheduled test. Two valves are scheduled for testing prior to July 2016 and two valves are scheduled for testing prior to March 2017. The testing schedule for valves was addressed in the GGNS EPU License Amendment Request and acceptance of the testing schedule as documented in the EPU NRC Safety Evaluation Report. Therefore, there are no additional vulnerabilities as shown by this extent of condition evaluation summary.

Corrective Actions

Required Actions of TS 3.6.1.3 were entered and immediate actions were taken to declare the two Refuel Water Transfer Pump Suction from Suppression Pool valves and the RHR Sample Return Isolation valve inoperable.

The valves were re-tested at or greater than 16.28 psig, 110% of the post-EPU accident pressure, on October 15, 2015 and October 17, 2015 and passed with zero leakage identified. The LLRT procedure which displayed the incorrect testing pressure will be revised to reflect the appropriate post-EPU base testing pressure.

Evaluation of the human performance aspect of this event is ongoing. Correction of the causal factors of the human performance errors will eliminate reoccurrence of this type of event. A supplement to this report will be submitted in the event the evaluation leads to significant changes in the cause or corrective actions associated with this LER.

Safety Significance

There were no actual nuclear safety or industrial safety consequences related to this event.

There were no actual or potential radiological consequences as a result of this event. The October 15, 2015 and October 17,2015 post-EPU pressure surveillance testing passed with zero leakage identified for all three valves.

Passing the surveillance test at the appropriate post-EPU testing pressure demonstrated the PCIVs continuous operability. Due to the testing proving continuous safety function was maintained, this event will not be counted toward the GGNS Performance Indicator for Loss of Safety Function.

Basis of Reportability:

This LER is being submitted pursuant to Title 10 Code of Federal Regulation (10 CFR) 50.73(a)(2)(i)(B) for any operation or condition which was prohibited by plant Technical Specifications and 10 CFR 50.73 (a)(2)(v)(C) for any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) control the release of radiological material. Telephonic notification was made to the U.S. Nuclear Regulatory Commission (NRC) Emergency Notification System (ENS) on October 14,2015, pursuant to 10 CFR 50.72(b)(3)(v)(C).

~

~Entergy GNRO-2015/00092 December 10,2015 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Operations, Inc.

P. O. Box 756 Port Gibson, MS 39150 James Nadeau Manager, Regulatory Assurance Grand Gulf Nuclear Station Tel. (601) 437-2103

SUBJECT:

Dear Sir or Madam:

Licensee Event Report (LER) 2015-003-00 Technical Specification Surveillance on Primary Containment Isolation Valves Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29 Attached is Licensee Event Report (LER) 2015-003-00, Technical Specification Surveillance am Primary Containment Isolation Valves, which is a final report. This report is submitted in accordance with Title 10 Code of Federal Regulations 50.73(a)(2)(i)(B).

This letter contains no new commitments. Should you have any questions or require additional information, please contact James Nadeau at (601) 437-2103.

JJN/tmc Attachment: Licensee Event Report (LER) 2015-003-00 cc: (See Next Page)

GNRO-2015/00092 Page 2 of 2 cc:

u.s. Nuclear Regulatory Commission ATTN: Mr. A. Wang, NRRlDORL (w/2)

Mail Stop OWFN 8 B1 Washington, DC 20555-0001 u.S. Nuclear Regulatory Commission ATTN: Mr. Marc Dapas (w/2)

Regional Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment to GNRO-2015/00092 Licensee Event Report (LER) 2015-003-00

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 (11-2015) the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

13. PAGE Grand Gulf Nuclear Station, Unit 1 05000416 1 OF 3

4. TITLE Technical Specification Surveillance on Primary Containment Isolation Valves

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.

MONTH DAY YEAR N/A 05000 N/A FACILITY NAME DOCKET NUMBER 10 14 2015 2015-003-00 12 13 2015 N/A 05000 N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)

10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) 100 o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) cgj 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi)

~ 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in Initial Conditions:

At the time of the event, Grand Gulf Nuclear Station (GGNS) was operating in MODE 1 at 100 percent rated core thermal power. There were no systems, structures or components that were inoperable at the start of the event that contributed to the event.

Description of Event

Grand Gulf Nuclear Station Unit 1 Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.9 requires Primary Containment Isolation Valves (PCIVs) to be hydrostatically tested at 110% of accident pressure in accordance with the Inservice Testing Program. The Inservice Testing Program is defined in TS 5.5.6. While reviewing Local Leak Rate Testing (LLRT) Procedures for outage preparation on October 14, 2015, it was discovered the Local Leak Rate Test Low Pressure Water procedure was not revised as a result of the Extended Power Uprate (EPU). A step in the procedure states the base test pressure is 12.65 psig (110% of peak accident pressure). The value of 12.65 psig is 110% of the pre-EPU peak accident pressure of 11.5 psig. The peak containment accident pressure was revised on July 18, 2012 when the Nuclear Regulatory Commission (NRC) approved the GGNS, Extended Power Uprate, License Amendment Request. The base test pressure should have been revised to 16.28 psig, 110% of the post-EPU peak accident pressure of 14.8 psig. The current testing criterion remains at the post-EPU value of 16.28 psig.

Three valves were determined to have not been Local Leak Rate Tested at or greater than the required base test pressure of 16.28 psig. These three valves included two Refuel Water Transfer Pump Suction from Suppression Pool valves and a Residual Heat Removal (RHR) Sample Return Isolation valve. Upon discovery, the valves were declared inoperable and isolated in accordance with TS 3.6.1.3. The two Refuel Water Transfer Pump Suction valves were tested at or greater than the post-EPU base test pressure on October 15, 2015 and passed with zero leakage identified. The Residual Heat Removal Isolation valve was tested at or greater than the post-EPU base test pressure on October 17. 2015 and passed with zero leakage identified. After satisfactory testing was completed, all valves were declared operable.

The Refuel Water Transfer Pump Suction valves were previously tested on June 25, 2008 and August 14, 2008 respectively with the appropriate pre-EPU test pressure of 12.9 psig. The test passed with zero leakage identified.

The RHR Isolation valve was tested on April 15, 2008 with the appropriate pre-EPU test pressure of 13.0 psig.

The test passed with zero leakage identified. Therefore, the valves were demonstrated operable at the 2008 tests.

Next testing of the valves was in 2013. The valves were inappropriately tested at or greater than the pre-EPU base test pressure and passed with zero leakage identified.

Cause of Event

The cause was determined to be inadequate incorporation of testing criteria into Surveillance Testing Procedure 06-0P-1M61-V-0003, Local Leak Rate Test Low Pressure Water. An Engineering Change was completed which evaluated the impacts of EPU on Containment Leakage Rate Testing. However, it did not identify Local Leak Rate Test Low Pressure Water as a procedure requiring change. This change was missed due to failure of personnel to follow the Engineering Change Process per station procedures. Necessary reviews were not completed and the required procedure change was not identified.

Extent of Condition:

Ten valves were identified that should have been hydrostatically tested at or above 16.28 psig per post-EPU conditions on a schedule specified by the Inservice Testing Program. Three of those valves were not tested at or above 16.28 psig during their most recent regularly scheduled surveillance test. These three

NRC FORM (11-2015) 366A U.S. NUCLEAR REGULATO~YCOMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

1. FACILITY NAME Grand Gulf Nuclear Station, Unit 1

2. DOCKET 05000416 YEAR I

3. LER NUMBER SEQUENTIAL NUMBER 2015-003-00 I

REV.

NO.

Valves are being reported under this Licensee Event Report (LER). The remaining seven valves were tested at or above 16.28 psig during their most recent regularly scheduled surveillance test. These seven valves passed the surveillance testing with satisfactory results. Therefore, there are no adverse conditions with the hydrostatic testing of the remaining seven valves.

Twenty-two valves are High/Low Interface Valves and are required to be hydrostatically tested at or above 50 psig.

The most recent water LLRT was reviewed for each valve and the test was conducted at or above 50 psig as required. Therefore, there are no adverse conditions with the hydrostatic testing of these High/Low Interface valves.

Grand Gulf Nuclear Station documented all pneumatically tested valves have been tested at or above the post-EPU accident pressure with the exception of four valves. These four valves do not require pneumatic LLRT at the post-EPU accident pressure until their next regularly scheduled test. Two valves are scheduled for testing prior to July 2016 and two valves are scheduled for testing prior to March 2017. The testing schedule for valves was addressed in the GGNS EPU License Amendment Request and acceptance of the testing schedule as documented in the EPU NRC Safety Evaluation Report. Therefore, there are no additional vulnerabilities as shown by this extent of condition evaluation summary.

Corrective Actions

Required Actions of TS 3.6.1.3 were entered and immediate actions were taken to declare the two Refuel Water Transfer Pump Suction from Suppression Pool valves and the RHR Sample Return Isolation valve inoperable.

The valves were re-tested at or greater than 16.28 psig, 110% of the post-EPU accident pressure, on October 15, 2015 and October 17, 2015 and passed with zero leakage identified. The LLRT procedure which displayed the incorrect testing pressure will be revised to reflect the appropriate post-EPU base testing pressure.

Evaluation of the human performance aspect of this event is ongoing. Correction of the causal factors of the human performance errors will eliminate reoccurrence of this type of event. A supplement to this report will be submitted in the event the evaluation leads to significant changes in the cause or corrective actions associated with this LER.

Safety Significance

There were no actual nuclear safety or industrial safety consequences related to this event.

There were no actual or potential radiological consequences as a result of this event. The October 15, 2015 and October 17,2015 post-EPU pressure surveillance testing passed with zero leakage identified for all three valves.

Passing the surveillance test at the appropriate post-EPU testing pressure demonstrated the PCIVs continuous operability. Due to the testing proving continuous safety function was maintained, this event will not be counted toward the GGNS Performance Indicator for Loss of Safety Function.

Basis of Reportability:

This LER is being submitted pursuant to Title 10 Code of Federal Regulation (10 CFR) 50.73(a)(2)(i)(B) for any operation or condition which was prohibited by plant Technical Specifications and 10 CFR 50.73 (a)(2)(v)(C) for any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) control the release of radiological material. Telephonic notification was made to the U.S. Nuclear Regulatory Commission (NRC) Emergency Notification System (ENS) on October 14,2015, pursuant to 10 CFR 50.72(b)(3)(v)(C).