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{{Adams
#REDIRECT [[RA-16-043, LER 16-002-00 for Oyster Creek Regarding Control Rod Drive Cooling Water System Isolation Scram Time Testing Not Performed]]
| number = ML16139A033
| issue date = 05/12/2016
| title = LER 16-002-00 for Oyster Creek Regarding Control Rod Drive Cooling Water System Isolation Scram Time Testing Not Performed
| author name = Gillin M
| author affiliation = Exelon Generation Co, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000219
| license number = DPR-016
| contact person =
| case reference number = RA-16-043
| document report number = LER 16-002-00
| document type = Letter, Licensee Event Report (LER)
| page count = 6
}}
 
=Text=
{{#Wiki_filter:t. / Exelon Generation RA-16-043 May12,2016 U.S. Nuclear Regulatory Commission One White Flint North Attn: Document Control Desk or 0-881 11555 Rockville Pike Rockville, MD 20852 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219 10 CFR 50.73
 
==Subject:==
 
Licensee Event Report (LER) 2016-002-00, "Control Rod Drive Cooling Water System Isolation Scram Time Testing Was Not Performed" Enclosed is LER 2016-002-00 reporting the operation or condition that was prohibited by the plant's Technical Specifications associated with the failure to perform Scram Time Testing following the isolation of cooling water flow to three Control Rod Drive Mechanisms (CROM). A supplement to this report will be issued following the determination of the apparent and contributing causes of the event and all associated corrective actions.
This event did not affect the health and safety of the public or plant personnel.
This event did not result in a safety system functional failure.
There are no regulatory commitments made in this LER submittal.
Should you have any questions concerning this report, please contact Mike McKenna, Regulatory Assurance
: Manager, at (609) 971-4389.
Respectfully, Michael Gillin Plant Manager Oyster Creek Nuclear Generating Station w/Enclosure cc: Administrator, NRC Region I NRC Senior Resident Inspector
-Oyster Creek Nuclear Generating Station NRC Project Manager -Oyster Creek Nuclear Generating Station NRCFORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 (11-2015)
Estimated burden per response to comply with this mandatory collection request 80 hours . .,+ \. Reported lessons learned are incorporated into the licensing process and fed back to industry.
" ; Send comments regarding burden estimate to the FOIA, Privacy and Information Collections
\ LICENSEE EVENT REPORT {LER) Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by .. ...... internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory
: Affairs, NEOB-10202, (3150-0104),
Office of Management and Budget, Washington, digits/characters for each block) DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Oyster Creek 05000219 1 OF 5 4. TITLE Control Rod Drive Cooling Water System Isolation Scram Time Testing Was Not Performed
: 5. EVENT DATE 6. LEA 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 FACILITY NAME DOCKET NUMBER 03 16 2016 2016 -002 -00 05 12 2016 N/A 05000 9. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS*OF 10 CFR §: {Check all that apply) 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(dl D 20.2203(a)(3)(iil D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)<1>
D 20.2203(a)(4)
D so.13(a)(2)(m)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(il 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)(iil D 50.36(c)(1)(ii)(A)
D 50.73(a)(2)(v)(A)
D 13.11(al<4l D 20.2203(a)(2)(iiil D so.3a(c)(2)
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D 73.11(a)<s>
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D so.4a(a)(3)(n)
D 50.73(a)(2)(v)(C)
D 13.11(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)
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D OTHER Specify in* Abstract below or in NRC Form 366A "'. ;,. "' '"' ;.*' .*' 12. LICENSEE CONTACT FOR THIS LEA LICENSEE CONTACT I TELEPHONE NUMBER (Include Area Code)
* Michael McKenna, Regulatory Assurance Manager (609) 971-4389
: 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANU-REPORTABLE CAUSE SYSTEM COMPONENT MANU-REPORTABLE
'FACTURER TOEPIX FACTURER TO EPIX E AA SEAL G080 N N/A N/A N/A N/A N/A 14. SUPPLEMENTAL REPORT EXPECTED 15.EXPECTED MONTH DAY YEAR [8J YES (If yes, complete
: 15. EXPECTED SUBMISSION DATE) 0No SUBMISSION DATE 06 20 2016 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) On 03/16/2016, it was identified that isolating or reducing cooling water to the Hydraulic Control Units (HCUs) for three control rods should have been considered a modification since it had the potential to impact the scram times of the control rods. Even though scram time penalties were applied for the three control rods where the cooling water flow was either isolated or reduced, failing to identify the isolation of cooling water to the control rods as a modification as described by the Technical Specifications resulted in the Plant not taking the action to scram time test the affected rods. By not completing scram time testing for the control rods, whose cooling water was isolated or reduced, the station was in violation of the requirements of Technical Specifications Section 3.2, since the issue was not identified previously and the affected control rods were not declared inoperable and isolated.
This event resulted in an Operation or Condition that was Prohibited by the Plant's Technical Specifications (TS) and is therefore being reported under 1 OCFR50.73(a)(2)(i)(B).
NRC FORM 366 (11-2015)
Page 2 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LER) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and led back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Colleetions Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnlocollects.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 NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LEA NUMBER VEAR Oyster Creek 05000-219 SEQUENTIAL NUMBER REV NO. 2016 -002 -00 NARRATIVE Description of Event On 03/16/2016, the NRC identified that scram time testing had not been pertormed following the isolation of cooling water flow to control rods 18-47 and 42-27, and the reduction in cooling water flow to control rod 30-03 as a result of a failed isolation valve to fully open. With the exception of control rod 30-03, the valves were isolated as a compensatory action to mitigate leakage from the Control Rod Drive Mechanism (CROM) seals that were being quantified as unidentified leakage within the drywell.
The isolated, or reduced, cooling water results in a hot CROM condition that is expected to impact scram times as documented in Generic Electric (GE) Service Information Letter (SIL) 173, Supplement 1, Revision 1, "Control Rod Drive High Operating Temperature."
Technical Specification (TS) Section 4.2.C.2.
states: "For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:
a_ 1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90% insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power. b. Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure."
The isolation of cooling water flow to a control rod, while it does not affect the operability or functionality, does reduce the flow to the CROM and can impact the scram time. Accordingly, scram time testing should have been pertormed per the TS due to a system modification that could impact the scram time. Since the testing was not performed, TS. 4.0.1 was also applicable as a surveillance requirement that was not met. In accordance with this TS section, if the surveillance requirements are not satisfied, this would require entry into the appropriate Limiting Condition for Operations (LCO) as described under TS Section 3.2.8.4, which would have required the control rods be declared inoperable, fully inserted, and isolated.
Additionally, TS Sections 3.2.A.2 and 3.2.A.3 are also applicable and would require a determination that adequate shutdown margin would be maintained within six hours of declaring the rods inoperable.
Since this was not accomplished, this resulted in an Operation or Condition that was prohibited by the Plant's TS. Equipment Description The control rod and drive mechanism provides control of reactor power, including the ability to provide a sufficiently rapid insertion of control rods (scram) so that no fuel damage results from any abnormal operating transient and limits fuel damage under accident conditions.
The 137 control rods for the Oyster Creek reactor are located uniformly throughout the core. The control rods are operated by CRDMs. The Hydraulic Control Units (HCUs) for the control rods supply and control the pressure and flow NRC FORM 366A (11-2015)
Page 3 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LEA) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate
.to the FOIA, Privacy and Information Collections Branch (f-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by 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 NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LER NUMBER YEAR Oyster Creek 05000-219 2016 -NARRATIVE SEQUENTIAL NUMBER 002 REV NO. -00 requirements to the Control Rod Drives (CRDs). The HCUs provide hydraulic power to be able to position control rods in the reactor core. HCU scram accumulators are designed with a limited nitrogen pressure and volume, which are sufficient to initiate control rod scram motion. The CRD System supplies water to the CRD HCUs for manual control rod movement, scram, control rod mechanism
: cooling, and to the head spray cooling system. The system provides the operators the ability to control core reactivity through control rod movement, both manually and by scram. On a control rod insertion, drive water flows up this riser to the under-piston area of the CROM at a pressure high enough to drive the CRDM against reactor pressure (typically 1260 psig at full reactor power). On a rod withdrawal, exhaust water at reactor pressure flows down this riser from the piston area of the CROM. During periods of no rod motion, a small amount of cooling water continuously flows up this riser at just over reactor pressure.
On a reactor scram, the scram inlet valve opens a flow path from the accumulator to the under-piston area of the CRDM via this header. A cooling water flow of 0.3 gpm to the CRD provides protection for the graphitar seals and elastomer rings. The CRD will perform its design function without cooling water supplied, as described in the Updated Final Safety Analysis Report (UFSAR) Section 3.9.4.2.4.
Analysis of Event The isolation of or reduced cooling water flow to a CRD can result in a high temperature control rod. As the corrective action for this condition, scram time penalties were applied to control rods 18-47, 42-27 and 30-03. By not identifying that reduced or isolated CRD cooling water flow is a modification, scram time testing as not performed as required by TS 4.2. This resulted in a violation of Tech spec sections 4.01, as a surveillance requirement not met, which would have required compliance with TS section 3.2. Failing to take the action required under tech spec section 3.2, resulted in the violation being reported by this LER. At the time of discovery, there were three control rods that had scram time penalties applied:
18-47, 42-27, and 30-03. The scram time testing of these control rods had not been completed as required since the isolation of cooling water was not deemed to be a "modification" to the system since it was considered bounded under the UFSAR description as not impacting the operability of the control rod. While this is true, isolating the cooling water to the control rod still has an impact on the scram time of the rod as described in the GE SIL. On 03/18/2016, control rods 18-47, 30-03 and 42-27 were scram time tested to evaluate their performance with elevated control rod drive temperature.
GE SIL 173, Supplement 1, Revision 1, scram time penalties had previously been applied to the individual scram times for each of the three high temperature control rods. When the control rods were scram timed on 03/18/2016, both 30-03 and 42-27 had scram times that were faster than the previous times with the high temperature penalties applied.
The scram time for control rod 18-47 was significantly longer than the previous time with penalties applied.
The degraded scram time for control rod 18-47 was not related to the high temperature NRG FORM 366A (11-2015)
Page 4 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018
[ LICENSEE EVENT REPORT (LEA) \ ** ; ..... .t CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
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 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 currenijy valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LER NUMBER YEAR Oyster Creek 05000-219 2016 -NARRATIVE SEQUENTIAL NUMBER 002 REV NO. 00 condition;
: however, it did indicate an abnormal condition with the rod which was subsequently declared inoperable, fully inserted, and isolated.
Assessment of Safety Consequences The isolation of the cooling water to a CRD does not directly affect the ability of a control rod to perform " its design function;
: however, isolating or reducing cooling water will increase the control rod temperature and has the potential to impact the scram time of a control rod.
* UFSAR Section 4.6 states that rapid shutdown of the reactor is accomplished through actuation of the Reactor Protection System (or via manual scram) which opens the scram valves and permits water under pressure to be applied to the drive mechanism.
The action exerts a pressure on the CRD piston mechanisms, and causes all rods to be fully inserted into the reactor core. Any control rod which is fully withdrawn will be fully inserted in approximately five seconds.
A control rod shall also be considered operable if the Control Rod is valved in service, can be moved with normal CRD pressure, and its accumulator is valved in service, with a minimum nitrogen tank pressure of 940 psig, fulfilling TS Section 3.2.B.4 requirements.
The control rod moves at a normal speed with normal CRD system parameters.
Control Rod 18-47 scram time testing was re-performed on 03/18/2016 and the times were slightly slower than the mean rod insertion times stated in the UFSAR, which is considered a degraded condition.
The individual control rod scram time is an attribute of this component that is not controlled by TS. Control Rod 18-47 maintains its functionality since it was capable of performing its specified function of scramming and notching, as set forth in the Current Licensing Basis (CLB). The degrading scram time, although not directly resulting in an inoperable control rod per the TS, did indicate a degraded condition, as understood by the operators.
As a result, the control rod was proactively removed from service and isolated.
A review of the TS Section 3.2 Bases was performed to ensure that the issue is not indicative of a common mode failure.
In addition to this particular control rod being the only one that was exhibiting the behavior described above, collet housing/collet finger type failures were also researched.
These types of failures would be demonstrated in either the control rod not inserting when.the control rod is scrammed, or in the control rod not latching after movement.
As control rod 18-47 was able to be scrammed, and able to being fully inserted to its 00 position, and stay at the 00 position, this was not indicative of failure of the collet housing.
TS Section 3.2.B.3 contains the requirements for the core average scram time and scram times of the fastest 3-out-of-4 control rods within a 2x2 array. Individual control rod scram times themselves do not have specifications.
The 2x2 array and core average scram time requirements were both met with control rod 18-47 in the degraded condition.
NRC FORM 366A (11-2015)
Page 5 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LER) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request:
80 hours. Reported lessons learned are incorporated into the licensing process and led back to industry.
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 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 valid OMB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LEA NUMBER YEAR Oyster Creek 05000-219 SEQUENTIAL NUMBER REV NO. 2016 -002 -00 NARRATIVE Based on review of the recent data, previous performance data, equipment design, and a fleet technical call with Subject Matter Experts (SMEs) and GE representatives, reasonable assurance of future scram time continuing to support steady values could not be technically justified based on the observed condition.
As result, control rod 18-47 was fully inserted and isolated.
Cause of Event The apparent and contributing causes of the event and the associated extent of condition and extent of cause reviews are still being determined at this time and will be included in the supplement to this LER. The following immediate actions were taken:
* On 03/18/2016 control rod 18-47 was inserted to its 00 position, valved-out of service and isolated.
Corrective Actions
* Executed repairs on control rods 18-47 and 42-27 and restored cooling water flow.
* Corrective actions will be determined by the Apparent Cause Evaluation and included in the supplement to this LER. Previous Occurrences There have been no similar, previous events resulting from the isolation of cooling water to a CRD or failing to perform scram time testing at Oyster Creek. Component Data Component IEEE 805 System ID IEEE 803A Function Control Rod Drive System AA SEAL NRG FORM 366A{11-2015) 
: t. / Exelon Generation RA-16-043 May12,2016 U.S. Nuclear Regulatory Commission One White Flint North Attn: Document Control Desk or 0-881 11555 Rockville Pike Rockville, MD 20852 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219 10 CFR 50.73
 
==Subject:==
 
Licensee Event Report (LER) 2016-002-00, "Control Rod Drive Cooling Water System Isolation Scram Time Testing Was Not Performed" Enclosed is LER 2016-002-00 reporting the operation or condition that was prohibited by the plant's Technical Specifications associated with the failure to perform Scram Time Testing following the isolation of cooling water flow to three Control Rod Drive Mechanisms (CROM). A supplement to this report will be issued following the determination of the apparent and contributing causes of the event and all associated corrective actions.
This event did not affect the health and safety of the public or plant personnel.
This event did not result in a safety system functional failure.
There are no regulatory commitments made in this LER submittal.
Should you have any questions concerning this report, please contact Mike McKenna, Regulatory Assurance
: Manager, at (609) 971-4389.
Respectfully, Michael Gillin Plant Manager Oyster Creek Nuclear Generating Station w/Enclosure cc: Administrator, NRC Region I NRC Senior Resident Inspector
-Oyster Creek Nuclear Generating Station NRC Project Manager -Oyster Creek Nuclear Generating Station NRCFORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 (11-2015)
Estimated burden per response to comply with this mandatory collection request 80 hours . .,+ \. Reported lessons learned are incorporated into the licensing process and fed back to industry.
" ; Send comments regarding burden estimate to the FOIA, Privacy and Information Collections
\ LICENSEE EVENT REPORT {LER) Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by .. ...... internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory
: Affairs, NEOB-10202, (3150-0104),
Office of Management and Budget, Washington, digits/characters for each block) DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Oyster Creek 05000219 1 OF 5 4. TITLE Control Rod Drive Cooling Water System Isolation Scram Time Testing Was Not Performed  
: 5. EVENT DATE 6. LEA 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 FACILITY NAME DOCKET NUMBER 03 16 2016 2016 -002 -00 05 12 2016 N/A 05000 9. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS*OF 10 CFR &sect;: {Check all that apply) 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(dl D 20.2203(a)(3)(iil D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)<1>
D 20.2203(a)(4)
D so.13(a)(2)(m)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(il 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)(iil D 50.36(c)(1)(ii)(A)
D 50.73(a)(2)(v)(A)
D 13.11(al<4l D 20.2203(a)(2)(iiil D so.3a(c)(2)
D 50. 73(a)(2)(v)(B)
D 73.11(a)<s>
D 20.2203(a)(2)(iv)
D so.4a(a)(3)(n)
D 50.73(a)(2)(v)(C)
D 13.11(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)
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D 50.73(a)(2)(vii)
D 73.77(a)(2)<n>
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**. ,:'*:, i D 50.73(a)(2)(i)(C)
D OTHER Specify in* Abstract below or in NRC Form 366A "'. ;,. "' '"' ;.*' .*' 12. LICENSEE CONTACT FOR THIS LEA LICENSEE CONTACT I TELEPHONE NUMBER (Include Area Code)
* Michael McKenna, Regulatory Assurance Manager (609) 971-4389
: 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANU-REPORTABLE CAUSE SYSTEM COMPONENT MANU-REPORTABLE
'FACTURER TOEPIX FACTURER TO EPIX E AA SEAL G080 N N/A N/A N/A N/A N/A 14. SUPPLEMENTAL REPORT EXPECTED 15.EXPECTED MONTH DAY YEAR [8J YES (If yes, complete
: 15. EXPECTED SUBMISSION DATE) 0No SUBMISSION DATE 06 20 2016 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) On 03/16/2016, it was identified that isolating or reducing cooling water to the Hydraulic Control Units (HCUs) for three control rods should have been considered a modification since it had the potential to impact the scram times of the control rods. Even though scram time penalties were applied for the three control rods where the cooling water flow was either isolated or reduced, failing to identify the isolation of cooling water to the control rods as a modification as described by the Technical Specifications resulted in the Plant not taking the action to scram time test the affected rods. By not completing scram time testing for the control rods, whose cooling water was isolated or reduced, the station was in violation of the requirements of Technical Specifications Section 3.2, since the issue was not identified previously and the affected control rods were not declared inoperable and isolated.
This event resulted in an Operation or Condition that was Prohibited by the Plant's Technical Specifications (TS) and is therefore being reported under 1 OCFR50.73(a)(2)(i)(B).
NRC FORM 366 (11-2015)
Page 2 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LER) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and led back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Colleetions Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnlocollects.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 NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LEA NUMBER VEAR Oyster Creek 05000-219 SEQUENTIAL NUMBER REV NO. 2016 -002 -00 NARRATIVE Description of Event On 03/16/2016, the NRC identified that scram time testing had not been pertormed following the isolation of cooling water flow to control rods 18-47 and 42-27, and the reduction in cooling water flow to control rod 30-03 as a result of a failed isolation valve to fully open. With the exception of control rod 30-03, the valves were isolated as a compensatory action to mitigate leakage from the Control Rod Drive Mechanism (CROM) seals that were being quantified as unidentified leakage within the drywell.
The isolated, or reduced, cooling water results in a hot CROM condition that is expected to impact scram times as documented in Generic Electric (GE) Service Information Letter (SIL) 173, Supplement 1, Revision 1, "Control Rod Drive High Operating Temperature."
Technical Specification (TS) Section 4.2.C.2.
states: "For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:
a_ 1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90% insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power. b. Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure."
The isolation of cooling water flow to a control rod, while it does not affect the operability or functionality, does reduce the flow to the CROM and can impact the scram time. Accordingly, scram time testing should have been pertormed per the TS due to a system modification that could impact the scram time. Since the testing was not performed, TS. 4.0.1 was also applicable as a surveillance requirement that was not met. In accordance with this TS section, if the surveillance requirements are not satisfied, this would require entry into the appropriate Limiting Condition for Operations (LCO) as described under TS Section 3.2.8.4, which would have required the control rods be declared inoperable, fully inserted, and isolated.
Additionally, TS Sections 3.2.A.2 and 3.2.A.3 are also applicable and would require a determination that adequate shutdown margin would be maintained within six hours of declaring the rods inoperable.
Since this was not accomplished, this resulted in an Operation or Condition that was prohibited by the Plant's TS. Equipment Description The control rod and drive mechanism provides control of reactor power, including the ability to provide a sufficiently rapid insertion of control rods (scram) so that no fuel damage results from any abnormal operating transient and limits fuel damage under accident conditions.
The 137 control rods for the Oyster Creek reactor are located uniformly throughout the core. The control rods are operated by CRDMs. The Hydraulic Control Units (HCUs) for the control rods supply and control the pressure and flow NRC FORM 366A (11-2015)
Page 3 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LEA) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate
.to the FOIA, Privacy and Information Collections Branch (f-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by 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 NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LER NUMBER YEAR Oyster Creek 05000-219 2016 -NARRATIVE SEQUENTIAL NUMBER 002 REV NO. -00 requirements to the Control Rod Drives (CRDs). The HCUs provide hydraulic power to be able to position control rods in the reactor core. HCU scram accumulators are designed with a limited nitrogen pressure and volume, which are sufficient to initiate control rod scram motion. The CRD System supplies water to the CRD HCUs for manual control rod movement, scram, control rod mechanism
: cooling, and to the head spray cooling system. The system provides the operators the ability to control core reactivity through control rod movement, both manually and by scram. On a control rod insertion, drive water flows up this riser to the under-piston area of the CROM at a pressure high enough to drive the CRDM against reactor pressure (typically 1260 psig at full reactor power). On a rod withdrawal, exhaust water at reactor pressure flows down this riser from the piston area of the CROM. During periods of no rod motion, a small amount of cooling water continuously flows up this riser at just over reactor pressure.
On a reactor scram, the scram inlet valve opens a flow path from the accumulator to the under-piston area of the CRDM via this header. A cooling water flow of 0.3 gpm to the CRD provides protection for the graphitar seals and elastomer rings. The CRD will perform its design function without cooling water supplied, as described in the Updated Final Safety Analysis Report (UFSAR) Section 3.9.4.2.4.
Analysis of Event The isolation of or reduced cooling water flow to a CRD can result in a high temperature control rod. As the corrective action for this condition, scram time penalties were applied to control rods 18-47, 42-27 and 30-03. By not identifying that reduced or isolated CRD cooling water flow is a modification, scram time testing as not performed as required by TS 4.2. This resulted in a violation of Tech spec sections 4.01, as a surveillance requirement not met, which would have required compliance with TS section 3.2. Failing to take the action required under tech spec section 3.2, resulted in the violation being reported by this LER. At the time of discovery, there were three control rods that had scram time penalties applied:
18-47, 42-27, and 30-03. The scram time testing of these control rods had not been completed as required since the isolation of cooling water was not deemed to be a "modification" to the system since it was considered bounded under the UFSAR description as not impacting the operability of the control rod. While this is true, isolating the cooling water to the control rod still has an impact on the scram time of the rod as described in the GE SIL. On 03/18/2016, control rods 18-47, 30-03 and 42-27 were scram time tested to evaluate their performance with elevated control rod drive temperature.
GE SIL 173, Supplement 1, Revision 1, scram time penalties had previously been applied to the individual scram times for each of the three high temperature control rods. When the control rods were scram timed on 03/18/2016, both 30-03 and 42-27 had scram times that were faster than the previous times with the high temperature penalties applied.
The scram time for control rod 18-47 was significantly longer than the previous time with penalties applied.
The degraded scram time for control rod 18-47 was not related to the high temperature NRG FORM 366A (11-2015)
Page 4 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018
[ LICENSEE EVENT REPORT (LEA) \ ** ; ..... .t CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
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 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 currenijy valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LER NUMBER YEAR Oyster Creek 05000-219 2016 -NARRATIVE SEQUENTIAL NUMBER 002 REV NO. 00 condition;
: however, it did indicate an abnormal condition with the rod which was subsequently declared inoperable, fully inserted, and isolated.
Assessment of Safety Consequences The isolation of the cooling water to a CRD does not directly affect the ability of a control rod to perform " its design function;
: however, isolating or reducing cooling water will increase the control rod temperature and has the potential to impact the scram time of a control rod.
* UFSAR Section 4.6 states that rapid shutdown of the reactor is accomplished through actuation of the Reactor Protection System (or via manual scram) which opens the scram valves and permits water under pressure to be applied to the drive mechanism.
The action exerts a pressure on the CRD piston mechanisms, and causes all rods to be fully inserted into the reactor core. Any control rod which is fully withdrawn will be fully inserted in approximately five seconds.
A control rod shall also be considered operable if the Control Rod is valved in service, can be moved with normal CRD pressure, and its accumulator is valved in service, with a minimum nitrogen tank pressure of 940 psig, fulfilling TS Section 3.2.B.4 requirements.
The control rod moves at a normal speed with normal CRD system parameters.
Control Rod 18-47 scram time testing was re-performed on 03/18/2016 and the times were slightly slower than the mean rod insertion times stated in the UFSAR, which is considered a degraded condition.
The individual control rod scram time is an attribute of this component that is not controlled by TS. Control Rod 18-47 maintains its functionality since it was capable of performing its specified function of scramming and notching, as set forth in the Current Licensing Basis (CLB). The degrading scram time, although not directly resulting in an inoperable control rod per the TS, did indicate a degraded condition, as understood by the operators.
As a result, the control rod was proactively removed from service and isolated.
A review of the TS Section 3.2 Bases was performed to ensure that the issue is not indicative of a common mode failure.
In addition to this particular control rod being the only one that was exhibiting the behavior described above, collet housing/collet finger type failures were also researched.
These types of failures would be demonstrated in either the control rod not inserting when.the control rod is scrammed, or in the control rod not latching after movement.
As control rod 18-47 was able to be scrammed, and able to being fully inserted to its 00 position, and stay at the 00 position, this was not indicative of failure of the collet housing.
TS Section 3.2.B.3 contains the requirements for the core average scram time and scram times of the fastest 3-out-of-4 control rods within a 2x2 array. Individual control rod scram times themselves do not have specifications.
The 2x2 array and core average scram time requirements were both met with control rod 18-47 in the degraded condition.
NRC FORM 366A (11-2015)
Page 5 of 5 NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2018 LICENSEE EVENT REPORT (LER) CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request:
80 hours. Reported lessons learned are incorporated into the licensing process and led back to industry.
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 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 valid OMB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET NUMBER 3. LEA NUMBER YEAR Oyster Creek 05000-219 SEQUENTIAL NUMBER REV NO. 2016 -002 -00 NARRATIVE Based on review of the recent data, previous performance data, equipment design, and a fleet technical call with Subject Matter Experts (SMEs) and GE representatives, reasonable assurance of future scram time continuing to support steady values could not be technically justified based on the observed condition.
As result, control rod 18-47 was fully inserted and isolated.
Cause of Event The apparent and contributing causes of the event and the associated extent of condition and extent of cause reviews are still being determined at this time and will be included in the supplement to this LER. The following immediate actions were taken:
* On 03/18/2016 control rod 18-47 was inserted to its 00 position, valved-out of service and isolated.
Corrective Actions
* Executed repairs on control rods 18-47 and 42-27 and restored cooling water flow.
* Corrective actions will be determined by the Apparent Cause Evaluation and included in the supplement to this LER. Previous Occurrences There have been no similar, previous events resulting from the isolation of cooling water to a CRD or failing to perform scram time testing at Oyster Creek. Component Data Component IEEE 805 System ID IEEE 803A Function Control Rod Drive System AA SEAL NRG FORM 366A{11-2015)}}

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