05000293/LER-2016-003, Regarding Spent Fuel Storage Design Feature Exceeded

Regarding Spent Fuel Storage Design Feature Exceeded

ML16202A052
Person / Time
Site:	Pilgrim
Issue date:	07/11/2016
From:	Perkins E Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.16.037 LER 16-003-00
Download: ML16202A052 (8)

LER-2016-003, Regarding Spent Fuel Storage Design Feature Exceeded

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)
2932016003R00 - NRC Website

text

~Entergx July 11, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth. MA 02360

SUBJECT:

Licensee Event Report 2016-003-00, Spent Fuel Storage Design Feature Exceeded Pilgrim Nuclear Power Station Docket No. 50-293 Renewed License No. DPR-35 LETTER NUMBER:

2.16.037

Dear Sir or Madam:

The attached Licensee Event Report 2016-003-00, Spent Fuel Storage Design Feature Exceeded, is submitted in accordance with 1 O Code of Federal Regulations 50. 73.

If you have any questions or require additional information, contact me at (508) 830-8323.

There are no regulatory commitments contained in this letter.

Sincerely,

~-?.~~

Everett P. Perkirrs,J~

Manager, Regulatory Assurance EPP/jl Attachment: Licensee Event Report 2016-003-00, Spent Fuel Storage Design Feature Exceeded (5 pages)

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station cc:

Mr. Daniel H. Dorman Regional Administrator, Region I U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 Ms. Booma Venkataraman, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8C2A Washington, DC 20555 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter No. 2.16.037 Page 2 of 2

Attachment Letter Number 2.16.037 Licensee Event Report 2016-003-00 Spent Fuel Storage Design Feature Exceeded (5 Pages)

NRG FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)

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Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

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Reported lessons learned are incorporated into the licensing process and fed back to industry.

~

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Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by "44 LICENSEE EVENT REPORT (LER) internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Pilgrim Nuclear Power Station 05000293 1OF5

4. TITLE Spent Fuel Storage Design Feature Exceeded

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED I

SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 05 12 2016 2016 -

003 -

00 07 11 2016 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that aooly)

D 20.2201(b)

D 20.2203(a)(3)(i)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

N D 20.2201(d)

D 20.2203(a)(3)(ii)

IR] 50. 73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(1)

D 20.2203(a)(4)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

D 20.2203(a)(2)(i)

D 50.36(c)(1 )(i)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

10. POWER LEVEL D 20.2203(a)(2)(ii)

D 50.36(c)(1)(ii)(A)

D 50.73(a)(2)(v)(A)

D 13.11(a)(4)

D 20.2203(a)(2)(iii)

D 5o.36(c)(2)

D 50.73(a)(2)(v)(B)

D 13.11(a)(5)

D 20.2203(a)(2)(iv)

D 5o.46(a)(3)(ii)

D 50.73(a)(2)(v)(C)

D 13.77(a)(1) 100 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50. 73(a)(2)(v)(D)

D 13.77(a)(2)(i)

D 20.2203(a)(2)(vi)

IR] 50. 73(a)(2)(i)(B)

D 50. 73(a)(2)(vii)

D 13.11(a)(2)(ii)

D 50.73(a)(2)(i)(C)

D OTHER Specify in Abstract below or in Engineering, Fuels Engineering and the BADGER test vendor.

Fuels Engineering reviewed the preliminary 2016 BADGER test results and determined that additional testing was not required at that time since 21 other panels with a similar dose and history were tested during this most recent test with a maximum cumulative gap preliminarily reported at 6. 7 inches (final value of 6.0 inches).

These 21 similar panels included all six panels in cells that share the RR35 South panel, as well as several other panels within the same rack as RR35 South. Since the 2016 BADGER measurements found only one panel out of the 22 similar panels to have a large cumulative gap beyond the analysis basis, the cumulative gap measured in panel RR35 South was considered to be an isolated occurrence.. There was no fuel in RR34, RR35 and other surrounding cells as fuel in this area was removed for the BADGER tests. Therefore, based on panel RR35 South test results, the six rack locations surrounding panel RR35 South were declared inoperable and blocked with blade guides to preclude inadvertent loading of fuel adjacent to the defective panel.

In order to evaluate the preliminary 2016 BADGER test results, a conservative Boraflex gap model was developed based on BADGER and blackness testing measurements performed at Pilgrim Nuclear Power Station (PNPS) and blackness testing measurements at Grand Gulf Nuclear Station (GGNS) over the last 28 years. Both GGNS and PNPS use fuel storage racks designed and manufactured by the same manufacturer and installed circa 1986. The model includes site specific factors that affect Boraflex degradation, such as gamma dose absorbed by each panel and the fuel storage history of each storage cell.

The other key factor that impacts Boraflex degradation is the rate which SFP water circulates through the rack volume that encapsulates the Boraflex panel. This parameter is a function of the design and manufacturing process so it is common to both sites' racks. Application of this model determined the extent of condition to be that approximately 589 panels are susceptible to gap losses of 10 inches (maximum gap loss considered in the current analysis of record) or larger. Based on these results, all Boraflex racks and any adjacent non-Boraflex cells that credit the neutron absorber in the Boraflex racks were declared inoperable.

CAUSE OF THE EVENT

The cause of the noncompliance with TS 4.3.1.1.b is Boraflex gaps greater than assumed in the criticality

safety analysis

Boraflex has been observed to be subject to in-service degradation from the combined effects of gamma radiation from spent fuel and long term exposure to the aqueous pool environment. Boraflex polymer shrinks due to irradiation beyond certain dose and forms gaps, which may have caused flow induced washout of the Boraflex material. The synergistic effects of gamma radiation and water cause the chemical composition of the polymer matrix of Boraflex to change from polydimethyl siloxane to amorphous silica, which is somewhat soluble in water. As the transformed matrix dissolves, it releases Boron and crystalline silica filler materials to the interior of the fuel racks and eventually to the SFP, resulting in gap growth. A contributing cause of the event is that water infiltration into cell RR35 South could be higher due to variation in fabrication quality. This cause would contribute to this condition by enabling boron washout.

The apparent cause of the unanalyzed condition is reliance on the method of BADGER Testing I RACKLIFE alone to determine the condition of Boraflex panels in the SFP.

CORRECTIVE ACTIONS

Corrective actions include performing a criticality evaluation to determine acceptable fuel storage configurations in Boraflex racks and achieving a SFP arrangement, in accordance with the criticality evaluation, to meet and maintain SFP Kett less than or equal to 0.95 as required by TS 4.3.1.1.b. Results of the evaluation will be provided in a supplement to this report.

The BADGER/RACKLIFE method of prediction of cumulative gap was augmented by a gap model based on PNPS and GGNS gap measurements to determine that approximately 589 locations were susceptible to gaps that could exceed the current analysis assumption. Further criticality analysis and corrective actions will utilize this gap model to determine the cumulative gap size to utilize within analysis models of the Boraflex racks. A more refined number of locations will also be calculated and provided in the supplement to this report.

SAFETY CONSEQUENCES

There was no actual safety consequence related to the event because analyses have demonstrated the current fuel storage configuration is and will be maintained substantially subcritical. This analysis utilized assumptions concerning Boraflex gap size and location based upon the results of the most recent BADGER test to conservatively model all Boraflex rack cells, including those determined to be susceptible to gaps that could exceed the current analysis assumption. In addition, SFP rack cells adjacent to panel RR35 South have been declared inoperable, blade guides have been installed into cell RR35 *and adjacent locations, and fuel movement in Boraflex racks has been stopped until completion of additional criticality analysis that supports movement of fuel within the Boraflex racks.

REPORT ABILITY A condition prohibited by TS occurred when the cumulative gap loss in SFP panel RR35 South invalidated the loss limit in the SFP Nuclear Criticality Boron Loss Study and invalidated the conclusion that the Boron loss projections would assure the SFP Kett less than 0.95 until July 28, 2016. This was a noncompliance with TS 4.3.1.1.b and is reportable in accordance with 10 Code of Federal Regulations (CFR) 50. 73(a)(2)(i)(B).

Condition Report CR-PNP-2016-02205 addresses the noncompliance. Based on initial extent of condition review, this event was reported under Event Notification 51924 on May 12, 2016, as an unanalyzed condition in accordance with 10 CFR 50. 72(b )(3)(ii)(B). Based on the unknown condition of other panels that are susceptible to gap losses, this condition is reportable as an unanalyzed condition in accordance with 10 CFR 50. 73(a)(2)(ii)(B).

PREVIOUS EVENTS None ENERGY INDUSTRY IDENTIFICATION SYSTEM (EllS) CODES YEAR 2016

3. LER NUMBER SEQUENTIAL NUMBER 003 The EllS codes for Components and Systems referenced in this report are as follows:

COMPONENTS: None SYSTEMS: None REFERENCES: CR-PNP-2016-02205 CODES: None CODES: None REV NO.

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