ML16354A424

License Amendment Request to Revise Technical Specification Surveillance Requirement 4.1.3.1.2 for Control Element Assembly 39 for the Remainder of Cycle 24

ML16354A424
Person / Time
Site:	Millstone
Issue date:	12/14/2016
From:	Mark D. Sartain Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Guzman R
References
16-454
Download: ML16354A424 (18)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com

~~

Dominion December 14, 2016 U.S. Nuclear Regulatory Commission Serial No 16-454 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 4.1 .3.1.2 FOR CONTROL ELEMENT ASSEMBLY 39 FOR THE REMAINDER OF CYCLE 24 In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) is submitting a license amendment request to amend Operating License No. DPR-65 for Millstone Power Station Unit 2 (MPS2). Specifically, DNC proposes to revise technical specification (TS) surveillance requirement (SR) 4.1.3.1 .2, Control Element Assembly (CEA) freedom of movement surveillance, such that CEA 39 may be excluded from the last remaining quarterly performance of the SR in MPS2 Cycle 24.

This amendment is necessary due to a potentially degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) that normally holds CEA 39 in place.

The UGC is also used during the performance of testing to satisfy SR 4.1.3.1.2, which verifies CEA freedom of movement. On November 1, 2016, during a review of data from the most recent CEA freedom of movement surveillance performed on October 27, 2016, the UGC low current ripple for CEDM 39 was found to have increased . Based on industry operating experience and recommendations from Westinghouse, CEA 39 was moved from the UGC to the lower gri"pper coil. Upon transfer to the lower gripper coil , the UGC was de-energized and remains de-energized, unless needed for urgent plant response.

CEA 39 remains capable of meeting the requirements of TS 3.1 .3.1 because it remains trippable and within 10 steps of other CEAs in its group. If DNC is required to perform the SR 4.1.3.1.2 freedom of movement surveillance on CEA 39, the UGC would need to be re-energized to move the CEA. If the UGC failed, the CEA could drop into the core, resulting in a reactivity transient and subsequent power reduction , and a plant shutdown if the CEA is unrecoverable. Therefore, DNC requests a one-time allowance to not perform SR 4.1.3.1.2 for CEA 39.

Specifically, the one-time allowance would allow CEA 39 to not be exercised during the last remaining quarterly performance of SR 4.1.3.1.2 in Cycle 24. Repairs to the CEDM for CEA 39 will be completed during the next MPS2 refueling outage.

Serial No: 16-454 Docket No. 50-336 Page 2 of 3 Attachment 1 to this letter describes the proposed changes and provides justification for the changes. Attachment 2 provides the marked-up TS page.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.

The next freedom of movement test for CEA 39 is scheduled for January 19, 2017.

In accordance with TS 4.0.2, this test can be extended to February 11, 2017.

Therefore, DNC requests approval of this license amendment request by February 1,2017.

In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of Connecticut.

Should you have any questions in regard to this submittal, please contact Wanda Craft at (804) 273-4687.

Sincerely, N<i',l~~*htl~~IC CommorivJealth of Virginia Mark D. Sartain . . Re~: # 140~42 .* .  :

Vice President - Nuclear Engineering 1:1v!.c;ommiss1.o.n: Exp[res May 31,..2Q18 COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this I.Bay of btCtm btv , 2016.

My Commission Expires: 5 -3I- /~

Notary Public Attachments:

1. Evaluation of Proposed License Amendment

2. Marked-Up Technical Specification Page Commitments made in this letter: None

Serial No: 16-454 Docket No. 50-336 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.16-454 Docket No. 50-336 ATTACHMENT 1 EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 1 of 12 EVALUATION OF PROPOSED LICENSE AMENDMENT 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a license amendment request to amend Operating License No.

DPR-65 for Millstone Power Station Unit 2 (MPS2). Specifically, DNC proposes to revise technical specification (TS) surveillance requirement (SR) 4.1.3.1.2, Control Element Assembly (CEA) freedom of movement surveillance, such that CEA 39 may be excluded from the last remaining quarterly performance of the SR in MPS2 Cycle 24.

This amendment is necessary due to a potentially degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) that normally holds CEA 39 in place. The UGC is also used during the performance of testing to satisfy SR 4.1.3.1.2, which verifies CEA freedom of movement. On November 1, 2016, during a review of data from the most recent CEA freedom of movement surveillance performed on October 27, 2016, the UGC low current ripple for CEDM 39 was found to have increased. Based on industry operating experience and recommendations from Westinghouse, CEA 39 was moved from the UGC to the lower gripper coil. Upon transfer to the lower gripper coil, the UGC was de-energized and remains de-energized, unless needed for urgent plant response.

CEA 39 remains capable of meeting the requirements of TS 3.1.3.1 because it remains trippable and within 10 steps of other CEAs in its group. If DNC is required to perform the SR 4.1.3.1.2 freedom of movement surveillance on CEA 39, the UGC would need to be re-energized to move the CEA. If the UGC failed, then the CEA could drop into the core, resulting in a reactivity transient and subsequent power reduction, and a plant shutdown if the CEA is unrecoverable. Therefore, DNC requests a one-time allowance to not perform SR 4.1.3.1.2 for CEA 39. Specifically, the one-time allowance would allow CEA 39 to not be exercised during the last remaining quarterly performance of SR 4.1.3.1.2 in Cycle 24. Repairs to the CEDM for CEA 39 will be completed during the next MPS2 refueling outage.

2.0 DESCRIPTION

OF THE PROPOSED CHANGE The proposed amendment would add the following note to SR 4.1.3.1.2 such that CEA 39 may be excluded from the last remaining quarterly performance of the SR in MPS2 Cycle 24. The note would read as follows:

Not required to be performed for CEA 39 for the remainder of Cycle 24.

A markup of the proposed TS change is provided in Attachment 2.

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 2 of 12 3.0 DETAILED DESCRIPTION 3.1 Description of Control Element Assembly Groups MPS2 has 61 CEDMs which hold a total of 73 CEAs (the shutdown (dual) CEAs are comprised of 24 CEAs connected to 12 CEDMs). The CEAs are divided into nine control groups: two Shutdown Groups A and B, and seven Regulating Groups 1 through 7. CEA 39 is in Regulating Group 7.

Shutdown Groups A and B ensure that sufficient negative reactivity is available to support a reactor trip or normal shutdown. The shutdown CEAs must be within their insertion limits (i.e., fully withdrawn) any time the reactor is critical or approaching criticality. Insertion limits on Regulating Groups 1-7 are also established and the CEA positions are monitored and controlled during initial criticality and power operation to ensure that the power distribution and reactivity limits are preserved. During reactor startup, regulating CEA groups are withdrawn and operated in a predetermined sequence with a predetermined amount of position overlap.

During power operations, the CEAs are normally fully withdrawn, except to complete SR 4.1.3.1.2 to demonstrate CEA freedom of movement, or as required for plant maneuvers or to respond to certain abnormal plant conditions.

3.2 Description/Operation of Control Element Drive Mechanism Control System 3.2.1 General System Description The Control Element Drive System (CEDS) positions the CEAs in the core to control reactivity during reactor startups and shutdowns, to make quick reactivity adjustments in controlling plant transients, and to control power distribution within the core during normal power operations. The use of many CEAs keeps the individual CEA reactivity worth low enough to prevent prompt criticality in the event of a continuous CEA withdrawal or a CEA Ejection accident. The safety function of the system is to fully insert all of the CEAs when the Reactor Protection System (RPS) detects conditions that could possibly lead to core damage.

The CEDS manually controls the direction and motion of the CEAs. The CEAs can be moved individually or as part of pre-assigned control groups. There are three different modes of control for CEAs:

• Manual Individual

• Manual Group

• Manual Sequential The active interface between the RPS and the CEDS is at the trip circuit breakers located at the Reactor Trip Switchgear. A reactor trip initiated by the RPS causes the

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 3 of 12 power to be removed from the CEDS by opening the reactor trip circuit breakers (RTCBs), which removes coil voltage and releases the spring-loaded grippers, allowing the CEAs to insert into the core by gravity.

3.2.2 Control Element Drive Mechanisms The CEDM is an electromechanical device that converts electrical energy into mechanical motion. The CEDM coils provide the magnetic flux that operates the mechanical parts of the drive within the pressure housing. Motion of these parts engage, lift, and release the latching devices, which translate the motion of the gripper assembly to the CEDM drive shaft.

Each CEDM has a lift coil, upper gripper coil, pull down coil, load transfer coil, and lower gripper coil, which comprise the coil stack assembly. These coil stack assemblies produce magnetic fields which control the magnetic jack assemblies causing the jacks to engage, hold, move, or release the CEAs. Each CEDM is capable of withdrawing, inserting, holding or tripping (releasing) its CEA from any point within its 137-inch stroke.

Under normal operating conditions, a CEA is not in motion and the CEA is held in place by the upper gripper coil which is continuously energized.

The CEDM Cooling system provides forced air cooling to the CEDMs to maintain these within specified operating temperatures.

4.0 TECHNICAL EVALUATION

In 2015, MPS2 experienced a failure of CEDM 40 due to UGC degradation. Following that failure, DNC initiated a monitoring and trending program to assess the ripple or currenUcurrent noise for the CEAs during quarterly rod motion or freedom of movement testing. The monitoring activity trends the UGC current ripple, which is the range of oscillation above and below the set UGC current value. The trending data is used to identify coil degradation to reduce the potential for a CEA drop event.

The purpose of SR 4.1.3.1.2 is to verify that the CEAs are free to move (i.e., trippable).

This is accomplished by moving each CEA in the Manual Individual mode (i.e., only one CEA is moved at a time by the control room operator). Successful movement of the CEA confirms no mechanical binding exists.

In addition, the design of the CEAs provides for freedom of movement. The CEDS is designed to ensure that electrical problems will not prevent insertion of a CEA into the core when the RTCBs are opened. Results from the last quarterly performance of SR 4.1.3.1.2 on October 27, 2016 showed freedom of movement.

Limiting Condition for Operation (LCO) 3.1.3.1 states in part that: A// CEAs shall be OPERABLE with each CEA of a given group positioned within 10 steps (indicated

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 4of12 position) of all other CEAs in its group ... . With the UGC de-energized and the lower gripper coil energized, CEA 39 will remain aligned and withdrawn with its group as required by LCO 3.1.3.1.

On November 1, 2016, during review of the MPS2 quarterly rod motion data obtained on October 27, 2016, the UGC low current ripple for CEA 39 was found to have increased.

Specifically, the data obtained on October 27, 2016 showed the low current ripple for CEA 39 to be 2 to 3 amps, which was a step change from the CEA 39 low current ripple values observed between 1st quarter and 2nct quarter 2016. The observed UGC low current ripple for other CEAs was 1 amp or Jess. Figure 1 below shows the step change in UGC low current ripple:

Figure 1 Change in UGC Low Current Ripple for CEA 39 6 .. **-*-*-*-*-*-****-**-******-* ...... **-*-****-**---*-**-*-*-*-*-*-*-*-*- *-*--*-... - ... ----*-*--.. *****-*-*-*-** *****-*-******-****--*-*-*-*---*---*----- --*-.. *-----.. -*-*-*-----.. *****-****-***** -*-*-*-*-****-****-*--*-**--*---*-** .. ***-- ---*-----*--*--..*--

4 . -** *--**- -** - -*--*-*-*-*-*-** **-* *-*----- ---- ------*--*-- ---- **--*--*-****-*- -** **-*-*---** ... **-----*--- - ----*---*- *-* **-*--*-*-- . .. . . .. **-*-*-*- **- --**-..**-*--- --

3 .5 - *- -*--*-*-*--*--***-*--*-*-*-*-**-*--*-*-*-*-*--*-*-*--*"*"*--*---**-*-*-**---*--*-*-*-****-*-***-*****-*--*-*-*-*--****-*-*-*---*--*-*"*"-""""*----*--- -*-*-**--****-*-*-***********-****-*-*-*-*****-****-*-*-********--**-****-*--

-CEOM: 39 obtained on 10*-.27-16 (After Chanie) 3 .. -*-*-*--*-**-****-*-*-*-*-*-*- *-*-**-*-*-*---**-*-*-****--**-*-**-*-*-*****-*-*-*-*-*****-*-*- -*-*-*-*--*-*-----*-*-"'"

-CEOM: 39 obtained on 02*18*16 (Sefore Chanae) 2.5 .. --*-*-*--***********--**--**-*-*--**-----*----********-****-********-*****-*****-*-*--*-* .. *-*-*-*--*-*--**"-*--**----**--------*----*---**-****-**---**-*--*-*-***-**-*.. *-* .. *- *-*-*--*.. *-*-*-*-*-*---**-*-*---**--*-*-*-**-*--*-*-**-*--*--*--*--*---

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 5 of 12 Troubleshooting the cause of the step change in CEDM 39 UGC low current ripple has not determined a definitive cause. There are three potential causes that have been assessed, including power switch components, a capacitor, and the UGC, only one of which could be repaired with the unit online. Further troubleshooting would require use of the hold bus to hold the CEA in place on the UGC. The risk of a dropped and unrecoverable CEA is considered high and not in the interest of plant safety due to potential reactivity transient and/or plant shutdown. The most likely cause, based on industry experience and consultation with Westinghouse, is degradation of the UGC.

The primary failure modes of a CEDM coil are:

• One or more winding shorts

• An open coil Overheating is the primary driver of both of these failure modes. Heating is primarily internal and largely a result of the coil being energized.

The majority of industry failures are related to coil insulation breakdown. A turn-to-turn short in the coil winding is the most common cause of coil failure. It occurs when insulation resistance degrades within a winding, allowing a secondary, or parasitic, current path. Although a single shorted turn in a winding may not have an immediate effect on a coil's performance, the point of insulation degradation becomes a source of additional heat. This localized heat buildup causes further insulation breakdown.

Furthermore, the shorted turns reduce the overall circuit resistance resulting in additional current draw and heat generation, and reduced coil magnetic holding power.

The increase in UGC ripple current for CEDM 39 most likely indicates a turn-to-turn short in the coil winding due to degraded insulation and consequential localized heating. The coil will further degrade with continued energization. The degradation is typically not linear or predictable. This localized heating can result in further turn-to-turn shorts which will increase current draw until either the Automatic CEA Timer Module (ACTM) auto transfers to the lower gripper or the associated breaker opens (releasing the CEA to drop into the core). An open coil condition has not been observed for CEDM 39.

The postulated failure described above is associated with the coil and associated control circuitry. The RTCBs are upstream of the control and power switch assemblies. The RTCBs open upon an automatic or manual reactor trip signal, removing all power from both control and holding circuitry. All coils on each CEDM subsequently de-energize, resulting in all CEAs inserting into the core. This design is fail-safe in that a loss of power, regardless of whether a reactor trip signal has been generated, will result in the CEAs inserting into the core.

Since each CEA is magnetically coupled to its associated gripper coil; there is no physical (mechanical) coupling between the CEDM circuit/coil and the associated gripper. Heating is the primary driver for coil degradation. This heating is associated with the internal heating of the coil windings. Any postulated failure mechanism that could prevent rod insertion (such as mechanical binding of the CEA itself) is not

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 6 of 12 influenced or impacted by coil failure or, control or holding circuitry failures. Therefore, there are no postulated failure mechanisms where the coil or associated circuitry could physically prevent rod insertion once the RTCBs have opened.

In addition, the postulated failure mode would not affect the RTCBs. Although not directly related to the identified condition of CEA 39, the system is designed in accordance with single failure criteria such that all circuitry will be de-energized even if one RTCB fails to open upon a reactor trip.

The control circuitry does not impact the RTCBs, nor is there a failure mode associated with the control circuitry which would prevent a loss of holding power when the RTCBs open. The control circuitry and/or power source interacts with the lifting device magnetically and is not electrically connected to the mechanical grippers which perform the actual movement of the CEA. The CEAs are designed to fail safe on loss of power.

Therefore, once power is lost (normally by opening of the RTCBs), no failure mode exists within the CEDM control or power circuitry that could interfere with releasing a CEA to drop into the core.

Further exercising of CEA 39 increases the potential for an inadvertent CEA drop, which would result in a reactivity transient and subsequent power reduction, and would lead to a reactor shutdown if the CEA is unrecoverable. Therefore, DNC requests NRC approval of a one-time allowance to not perform SR 4.1.3.1.2 for CEA 39. This will allow MPS2 to avoid the potential for an inadvertent drop, should the UGC fail during the next scheduled surveillance. Repair of CEDM 39 will occur during the next MPS2 refueling outage.

4.1 Reactivity Impact CEA 39 is trippable and is expected to remain so for the remainder of Cycle 24 operation. Because CEA 39 is trippable, the current FSAR Chapter 14 Safety Analysis is unaffected. The current FSAR Chapter 14 Safety Analysis assumes the most reactive CEA fails to insert on a reactor trip signal.

As discussed above, CEA 39 remains trippable, will insert following receipt of a reactor trip signal, and the existing FSAR Chapter 14 Safety Analysis remains bounding.

Consistent with the industry precedents identified in Section 5.2 of this LAR, the potential impact on Shutdown Margin (SOM) of the hypothetical failure of the highest reactivity combination of CEA 39 and a second CEA failing to insert on reactor trip is being provided for additional information. The TS definition of SOM is:

Shutdown Margin shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 7 of 12 TS 3.1.1.1 specifies that the SOM shall be greater than or equal to that specified in the Core Operating Limits Report (COLR). The COLR SOM operating limit for MPS2 Cycle 24 is 3.6% LiK/K in Modes 3 through 5; this value being the maximum SOM requirement over the temperature range. A parametric study was conducted from a Cycle 24 core exposure of 11,500 MWD/MTU to the end of Cycle 24 to determine the minimum SOM that would exist following a reactor trip assuming the highest reactivity worth combination of CEA 39 and a second CEA fails to insert. The calculated minimum SOM for this scenario is 3.7% LiK/K, which is above the 3.6% LiK/K SOM requirement in the COLR.

The calculated SOM value bounds operation for the remainder of MPS2 Cycle 24 operation.

Based on the above results, it can be shown analytically that SOM in excess of the COLR limit of 3.6% LiK/K exists for the remainder of MPS2 Cycle 24 operation, even if CEA 39 fails to insert into the core during a reactor trip. The calculations were performed using NRG-approved methodologies (Reference 7.6) used to generate the COLR and to perform the TS surveillances.

4.2 Nuclear Safety Risk Insights SR 4.1.3.1.2 verifies that the CEAs are not mechanically bound. The Probabilistic Risk Assessment model presumes insertion of one-half or more of the control rods is needed to achieve hot, zero power (

Reference:

NUREG/CR-5500, Vol. 10). Therefore, a common cause failure of roughly 35 CEAs is necessary to fail the reactivity control function.

Since only one of the 73 CEAs will not be exercised during the last remaining quarterly surveillance prior to the next MPS2 refueling outage, the impact on the reactivity control function and thus, Core Damage Frequency and Large Early Release Frequency, is negligible. The remaining CEAs will be tested which provides confidence that a common cause condition does not exist.

4.3 Administrative Controls The following administrative controls and compensatory actions have been established to minimize the frequency of energizing the CEA 39 UGC and potentially causing further degradation during the remainder of Cycle 24 operation. The administrative controls ensure operator movement of CEA 39 is not performed without knowledge of the current condition of the CEA 39 UGC degradation.

• MPS2 Operations has issued a standing order to limit but not prohibit the use of CEAs. Guidance has been provided for changing power levels while limiting Regulating Group 7 motion for the remainder of Cycle 24 operation. Specifically, the use of CEDM 39, and therefore Regulating Group 7, will be limited to a plant

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 8 of 12 response as directed by abnormal operating procedures or planned power reductions in excess of 15%.

• MPS2 Operations has placed a tag on the CEA motion control switch identifying that CEA 39 is on the lower gripper coil and the potential for actuating the ACTM Trouble alarm upon motion of Regulating Group 7.

• Reactivity plans have been developed by Reactor Engineering for downpowers to 95, 90, and 85% power without the use of CEAs for Axial Shape Index (ASI) control. The reactivity plans utilize a combination of RCS boration/dilution and ramp rate control in order to minimize axial xenon perturbations and maintain ASI within its COLR limits.

5.0 REGULATORY EVALUATION

5.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, "Technical specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2)

LCOs, (3) SRs, (4) design features, and (5) administrative controls.

SRs in 10 CFR 50.36 are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that LCOs will be met.

On February 20, 1971, the Atomic Energy Commission published in the Federal Register the General Design Criteria [GDC] for Nuclear Power Plants. The GDC, which are contained in Appendix A of 10 CFR 50, establish minimum requirements for the principal design criteria for water-cooled nuclear power plants. Although MPS2 was designed and licensed to the GDC, as issued on July 11, 1967, DNC has attempted to comply with the intent of the newer GDC to the extent possible, recognizing previous design commitments.

The GDC requirements applicable to the proposed LAR are as follows:

GDC-26, "Reactivity control system redundancy and capability." Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 9 of 12 acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions; GDC-27, "Combined reactivity control systems capability." The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained; GDC-28, "Reactivity limits." The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition; and GDC-29, "Protection against anticipated operational occurrences." The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

5.2 Precedent The proposed change is similar to the changes previously approved by the Nuclear Regulatory Commission (NRC) for Palo Verde Unit 2 (Reference 7.2), Arkansas Nuclear One, Unit 2 (References 7.3 and 7.5), and Palisades (References 7.1 and 7.4).

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 1Oof 12 5.3 No Significant Hazards Consideration Determination In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a license amendment request to amend Operating License No.

DPR-65 for Millstone Power Station Unit 2 (MPS2). Specifically, DNC proposes to revise technical specification (TS) surveillance requirement (SR) 4.1.3.1.2, Control Element Assembly (CEA) freedom of movement surveillance, such that CEA 39 may be excluded from the last remaining quarterly performance of the SR in MPS2 Cycle 24.

DNC has determined that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c). This determination is based on an evaluation with respect to the specific criteria of 10 CFR 50.92(c) as follows:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. The function of CEA 39 is to provide negative reactivity addition into the core upon receipt of a signal from the Reactor Protection System (RPS). CEA 39 was demonstrated to be moveable and trippable during the last performance of SR 4.1.3.1.2. Since the functionality of CEA 39 has not been affected, the assumptions and conclusions of the Final Safety Analysis Report (FSAR) Chapter 14, Safety Analysis, are not affected by this license amendment request.

The misoperation of a CEA, which includes a CEA drop event, has been evaluated in the MPS2 FSAR and found acceptable. The proposed change would minimize the potential for inadvertent insertion of CEA 39 into the core by eliminating the requirement to place the CEA on the UGC to perform freedom of movement testing.

The proposed change does not significantly increase the probability of a failure of a CEA to insert into the core on a reactor trip or the probability of an inadvertent CEA drop into the core at power.

No modifications are proposed to the RPS or associated Control Element Drive Mechanism (CEDM) system logic.

Based on the above, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 11 of 12 Response: No.

The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. CEA 39 was demonstrated to be moveable and trippable during the last performance of SR 4.1.3.1.2; therefore, the functionality of CEA 39 has not been affected. The proposed change will not introduce any new design changes or systems that can prevent the CEA from performing its specified safety function to insert on a reactor trip. The current MPS2 FSAR safety analysis considers the drop of a CEA into the core as an initiating event. This change does not alter assumptions made in the FSAR Chapter 14 safety analysis.

Based on the above, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in the margin of safety?

Response: No.

The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. SR 4.1.3.1.2 is intended to verify freedom of movement of CEAs (i.e., trippable). CEA 39 was demonstrated to be moveable and trippable during the last performance of SR 4.1.3.1.2. The physical and electrical design of the CEAs, and past operating experience, provides high confidence that CEAs remain trippable whether or not exercised during each SR interval. Eliminating further exercise of CEA 39 for the remainder of MPS2 Cycle 24 operation does not directly relate to the potential for CEA binding to occur. The current MPS2 FSAR safety analysis is unaffected by this license amendment request and there is no reduction in the margin of safety.

There is no known failure mechanism (e.g., crud deposition) that would preclude the CEA from inserting into core on a valid trip signal or loss of power.

Based on the above, the proposed amendment does not involve a significant reduction in the margin of safety.

Based on the above, DNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

The proposed amendment does not represent a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, nor does it introduce a significant increase in individual or cumulative occupational radiation exposure.

Serial No.16-454 Docket No. 50-336 Attachment 1, Page 12of12

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change an inspection or surveillance requirement. However, as established above, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion of categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statements or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 NRC letter to Entergy [ADAMS Accession Number ML16281A498] dated October 28, 2016 - Palisades Nuclear Plant - Issuance of Amendment Re: Control Rod Drive Exercise Surveillance (GAG No. MF8297).

7.2 NRC letter to Arizona Public Service Company [ADAMS Accession Number ML15266A005] dated September 25, 2015 - Palo Verde Nuclear Generating Station - Issuance of Amendment to Amend Technical Specification Surveillance Requirement to Eliminate Movement of Control Element Assembly 88 for the Remainder of Unit 2, Operating Cycle 19.

7.3 NRC letter to Entergy [ADAMS Accession Number ML15096A381] dated April 29, 2015 - Arkansas Nuclear One, Unit 2 - Issuance of Amendment Re: Revise Technical Specifications Surveillance Requirement to Eliminate Movement of Control Element Assembly 18 for the Remainder of Operating Cycle 24 (TAC No.

MF5698).

7.4 NRC letter to Entergy [ADAMS Accession Number ML101380534] dated June 2, 2010 - Palisades Nuclear Plant- Issuance of Amendment Re: Control Rod Drive Exercise Surveillance (TAC No. ME3638).

7.5 NRC letter to Entergy [ADAMS Accession Number ML012960550] dated October 22, 2001 - Arkansas Nuclear One, Unit 2 - Issuance of Amendment Re:

Allowance to Eliminate Movement of Control element Assembly 43 for the Remainder of Cycle 15 (TAC No. MB2779).

7.6 EMF-96-029(P)(A) Volumes 1 and 2, Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 - Benchmarking Results, Siemens Power Corporation, January 1997.

Serial No.16-454 Docket No. 50-336 ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATION PAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.16-454 Docket No. 50-336 Attachment 2, Page 1 of 1 Gs!seer 29, i!Qll REACTIVITY CONTROL SYSTEMS AC'ITON: (Continued):

C. CEA Deviation Circuit C, 1 Verify the indicated position of each CEA to be within inoperable. 10 steps of all -other CEAs in its group within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter or otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. One or more CEAs untrippable. D.l Be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

OR Two or more CEAs misaligned by

~ 20 steps.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 Verify the indicated position of each CEA to be within 10 steps of all other CEAs in its group at the frequency specified in the Surveillance Frequency Control Program AND \Vithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following any CEA n1ovement larger than 10 steps.

4.1.3.1.2 erify CEA freedotn ofn1ove1nent (trippability) by moving each individual CEA that is not fully inserted into the reactor core 10 steps in either direction at the t

frequency specified in the Surveillance Frequency Control Program.

4.1.3.1.3 Verify the CEA Deviation Circuit is OPERABLE at the frequency specified in the Surveillance Frequency Control Program by a functional test of the CEA group Deviation Circuit \Vhich verifies that the circuit prevents any CEA from being misaligned from all other CEAs in its group by more than 10 steps (indicated position).

4.1.3.14 Verify the CEA Motion Inhibit is OPERABLE by a functional test \Vhich verifies that the circuit tnaintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit prevents regulating CEAs fro111 being inserted beyond the Transient Insertion Li1nits specified in the CORE OPERATING LIMITS REPORT:

a. Prior to each entry into MODE 2 from MODE 3, except that such verification need not be perfonned more often than once per 31 days, and

b. At the frequency specified in the Surveillance Frequency Control Program. -+-

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Not required to be performed for CEA 39 for the remainder of Cycle 24 MILLSTONE - UNIT 2 3/4 1-21 Amend1nent No.~ UQ., ~