NL-15-021, Proposed License Amendment Regarding a Change to Technical Specification 3.1.4 Reactivity Control Systems Indian Point Unit Number 2
| ML15044A471 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/12/2015 |
| From: | Coyle L Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-15-021 | |
| Download: ML15044A471 (16) | |
Text
%-N-EnteW,0 Enterav Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 Lawrence Coyle Site Vice President February 12, 2015 NL-15-021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738
SUBJECT:
Proposed License Amendment Regarding a Change to Technical Specification 3.1.4 "Reactivity Control Systems" Indian Point Unit Number 2 Docket No. 50-247 License No. DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc, (Entergy) hereby requests a License Amendment to Operating License DPR-26, Docket No. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2). This change is requested to revise the acceptance criteria for the Surveillance Requirement (SR) 3.1.4.2 for the Control Rod G-3. During the last two performances of this Surveillance on September 18, 2014 and December 11, 2014, Control Rod G-3 misalignment occurred with Shutdown Bank B group movement as displayed by Individual Rod Position Indication (IRPS) and Plant Instrument Computer System (PICS). The proposed change is to defer subsequent testing of the Control Rod G-3 until repaired during the next refuel outage (March 2016) or forced outage long enough to repair the Control Rod.
Entergy has evaluated the proposed change in accordance with 10 CFR 50.91(a)(5) using the criteria of 10 CFR 50.92 (c) and has determined that this proposed change involves no significant hazards considerations. Attachment 1 includes this evaluation and describes the basis for this conclusion. The proposed Technical Specification is provided in Attachments 2. A copy of this application and the associated attachments are being submitted to the designated New York State official in accordance with 10 CFR 50.91.
Entergy requests approval of the proposed amendment by April 3, 2015. If the NRC finds that time is insufficient for the full public comment period, Entergy requests that this request be considered exigent based on the considerations discussed in Attachment 1.
ý00
NL-15-021 Docket No. 50-247 Page 2 of 2 There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February
_J z, 2015.
Sincerely, Attachments:
- 1. Analysis of Proposed Technical Specification Change Regarding Change to Surveillance Requirement 3.1.4.2
- 2.
Markup of Technical Specification Change to Surveillance Requirement 3.1.4.2 cc:
Mr. Douglas Picket, Senior Project Manager, NRC NRR DORL Mr. Daniel H. Dorman, Regional Administrator, NRC Region 1 NRC Resident Inspector's Office Mr. John B. Rhodes, President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service
ATTACHMENT 1 TO NL-15-021 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING CHANGE TO THE SURVEILLANCE REQUIREMENT 3.1.4.2 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247
NL-15-021 Docket No. 50-247 Page 1 of 11
1.0 DESCRIPTION
This letter requests an amendment to Operating License DPR-26, Docket No. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2). The proposed change is to revise the acceptance criteria for the Control Rod G-3 in the Surveillance Requirement (SR) 3.1.4.2 in Technical Specification 3.1.4 "Reactivity Control Systems" to allow deferral of the test. In accordance with the TS frequency, SR 3.1.4.2 is scheduled to be performed on March 15, 2015 plus a 25%
extension to April 7, 2015 to allow for review of the request for amendment. Considering the refuel outage is due to commence March 2016, the proposed LAR would potentially eliminate four surveillances of Control Rod G-3. Reference 1 is an example of where the NRC staff approved an amendment to defer SR 3.1.4.2 to prevent further degradation of a CRDM seal.
If NRC finds that insufficient time exists to provide a full public comment period prior to April 7, 2015, Entergy requests that this proposed change be processed as an exigent change per 10 CFR 50.91(a)(6), As demonstrated below, there is no significant hazard consideration. The change is needed because of concerns that the performance of the test would cause the control rod to drop into the core and that it would not be recoverable. The plant would shutdown to fix the control rod since continued operation would be possible but not practical.
2.0 PROPOSED CHANGE
S Revise the SR 3.1.4.2 acceptance criteria for a movement of each control rod to defer the movement of Control Rod G-3. The TS proposed change is:
From "Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core -- 10 steps in one direction."
To "Verify rod freedom of movement (trippability) by moving each rod* not fully inserted in the core ->
10 steps in one direction."
"*Control Rod G-3 need not be moved until repaired in the next forced outage of sufficient duration or the next refuel outage in 2016."
3.0 BACKGROUND
The Reactor Protection System acts to shut the reactor down by means of various reactor trips which are designed to occur when a measured plant variable exceeds predetermined limits.
The protection system consists of all instrumentation which monitors the process variables and Initiates a trip if the process variables approach safety limits. The trips function to provide rapid reduction of reactivity by the insertion of full-length Rod Cluster Control Assemblies (RCCA) assemblies under free fall into the reactor core. The full-length RCCA must be energized to remain withdrawn from the core. Automatic reactor trip occurs upon the loss of power to the full-length control rods. All power to the full-length control rod mechanisms are interlocked by duplicate series
NL-15-021 Docket No. 50-247 Page 2 of 11 connected circuit breakers. The trip breakers are opened by the under voltage coils on both breakers. The under voltage coils, which are normally energized, become de-energized by any one of the trip signals.
Overall reactivity control is achieved by the combination of chemical shim and 53 control rod clusters of which 29 are in 4 control banks and 24 are in 4 shutdown banks. Long-term regulation of core reactivity is accomplished by adjusting the concentration of boric acid in the reactor coolant. Short-term reactivity control for power changes or reactor trip is accomplished by movement of control rod clusters. The primary function of the Reactor Control System is to provide automatic control of the rod clusters in control banks during power operation of the reactor.
The four control banks are the only rods that can be manipulated under automatic control. The banks are divided into groups to obtain smaller incremental reactivity changes. All RCC assemblies in a group are electrically paralleled to step simultaneously. Position indication for each RCC assembly type is the same.
The control bank rods are divided into four banks comprising 8, 4, 8 and 9 RCC assemblies, respectively, to follow load changes over the full range of power operation. Each control rod bank is driven by a sequencing, variable speed rod drive control unit. The assemblies in each control bank are divided into two groups. The variable speed sequential rod control affords the ability to insert a small amount of reactivity at low speed to accomplish fine control of reactor coolant average temperature about a small temperature dead band.
The shutdown rods together with the control rods are capable of shutting the reactor down.
They are used in conjunction with the adjustment of chemical shim to provide shutdown margin of at least one percent following reactor trip with the most reactive control rod in the fully withdrawn position for all normal operating conditions. The shutdown banks are manually controlled during normal operation and are moved at a constant speed with staggered stepping of the groups within the banks. Any reactor trip signal causes the control and shutdown to drop by gravity into the core.
During surveillance testing to SR 3.1.4.2, the control rod drive (CRD) system moves the RCCAs based on commands from the Operator. Each RCCA is connected to a drive shaft with many grooves along their length for the working components of the control rod drive mechanism (CRDM) to engage with and physically carry their weight. The CRDM keeps the various control rods suspended within the core and will step them in or out of the core based on the commands given to the CRD system. A diagram of the CRDM is shown below. Each drive shaft has three coils associated with it: a stationary gripper coil, a moveable gripper coil, and a lift coil. These coils generate the magnetic flux to operate their associated components. Sending current to the stationary gripper coil (also referred to as the stationary coil) moves the stationary grippers to engage the drive shaft to support the weight of the rod while the rod is not in motion. Sending current to the moveable gripper coil (also referred to as the moveable coil) controls the moveable gripper to engage the drive shaft to support the weight of the rod while the rod is in motion.
Sending current to the lift coil will move the entire moveable gripper assembly up by one groove (or step) and actually facilitates the motion of the drive shaft. Removing power from all coils of all rods will drop all control and shutdown banks and provides the means to trip the reactor. The three magnetic coils surround the mechanism pressure vessel, and create a magnetic flux which passes through the vessel wall and operates the working components.
N L-15-021 Docket No. 50-247 Page 3 of 11 Grifte Coil Flux Lift Pole k~ovaleRFux
~Ring LthReturn SpringfoL Gripper Coil Armature Movable--,,._
Mvablee Latch Latch Link lvlovabl e Gripper F1lue Ring oAature (Shoyn Closed)
star lo nary Gripper Pole The S e nc Stati onary Gripper Coir d Armature Return Spring Latch Return Flux Ring SpBae ng f.o r Load Tranfer)Stationary Gripper Armature Sclationsy Etationary Latch Latch Link Drive Rod GuideTube The Surveillance Requirement 3.1.4.2 is intended to demonstrate that the control rods are operable to perfomn their design function of tripping and dropping by gravity into the core. The Bases for SR 3.1.4.2 notes that verifying each control rod is operable would require that each rod be tripped, however, during power operation tripping each control rod would result in radial or axial power tilts, or oscillations. "Exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps in one direction will not cause radial or axial power tilts, or oscillations, to occur. This SR requires that control rods be inserted or withdrawn by at least 10 steps which is sufficient to ensure that rod movement can be confirmed by individual rod position indicators. The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable but remains trippable, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.
NL-15-021 Docket No. 50-247 Page 4 of 11
4.0 TECHNICAL ANALYSIS
The Full Length Rod Control System is designed to position the Rod Cluster Control Assemblies (RCCA's) upon commands from the Operator or the Automatic Rod Control System, thus controlling reactor temperature and power distribution within the core. The RCCA's are divided into two sub groups, Control Bank and Shutdown Bank Control Rods. The primary function of Shutdown Bank RCCA's is to release the control rods which fall by gravity to the bottom of the core in response to manual or automatic reactor trip signals.
There have been instances of misalignment of Control Rod G-3, in Shutdown Bank B, during the last two Control Rod surveillance testing required by SR 3.1.4.2. On September 18, 2014, Control Rod G-3 inserted to 195 steps with Shutdown Bank B bank demand at 218 steps. On December 11, 2014, Control Rod G-3 inserted to 208 steps with Shutdown Bank B Bank demand at 221 steps.
An assessment of the September 2014 concluded that:
The Rod Control Cabinet control circuitry does not have the issue since the initial investigation of the first event found that the was no Urgent or Non Urgent alarms when the Control Rod G-3 inserted in and the fuses which were replaced were not blown and later tested as operable.
" The current rod position is being maintained as demonstrated when the Control Rod G-3 is inserted and withdrawn during the tests in September and December 2014 where the stationary coil is energized / de-energized. This demonstrated that the stationary coil is functional and therefore the rod is trippable. This is true of all previous tests since the last refuel outage.
The remaining possible causes for the rod mis-alignment were mechanical binding of the moveable or stationary grippers, degraded CRDM coilstack, or CRUD. The remaining banks that had not been tested were then tested on September 25, 2014.
Current trace data for the stationary, moveable, and lift coils were gathered for Control Rod G-3 and Rod C9 (a rod in the same group that had functioned normally) for comparison. During the continued testing (movement to 213 steps and back to 223) the current profiles are virtually identical.
Based on this and the fact that multiple rods were not affected, it was concluded that the cause was not electrical failure of the control system or mechanical binding of the CRDM.
There is a history of such intermittent rod misalignment issues in the industry that have been attributed to CRUD. Magnetic CRUD is known to deposit on CRDMs due to thermal siphoning flow. The presence of the magnetic fields required to operate the CRD mechanisms attracts ferritic crud. It is also possible that there is an intermittent mechanical issue, apart from CRUD, specific to Rod G3 that prevents either the stationary or moveable grippers from properly engaging the grooves of the Rod G3 drive shaft during movement. This type of issue would affect the motion of the moveable, lift, or stationary armatures in subsequent rod movements, and thus be seen in the current traces. Since no difference could be seen in the Rod G3 current traces, this failure mode was eliminated.
NL-15-021 Docket No. 50-247 Page 5 of 11 Additional corrective actions have been or are being taken to help alleviate this issue. The remaining corrective actions (including possible replacement of components such as drive shaft or CRDM) will require an outage to complete.
This proposed TS change is based on the concept discussed in the TS Bases "The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable the control rod(s) is considered to be OPERABLE." The following figure shows the location of Control Rod G-3 in the core:
NL-15-021 Docket No. 50-247 Page 6 of 11 CONTROL ROD LOCATIONS 15 14 13 12 R
p-N M-1l 10 9
8 7
6 5
4 3
2 FA L
K-J-
H -0'-
F E*
07S A9S SA01C[iA n
180" D
C B
A A
[qAI
4 -4 4-
.4-4.-. =- 4 -
J -
w 270' BAN K
SYMBO LD D
V 0
0 NUMBER OF ROO CLUSTERS 8
4 4
8 4
B C
0 RE.0 (BdOsJ9O)
NL-15-021 Docket No. 50-247 Page 7 of 11 For the control rods to be operable they must be able to perform their safety function, which means they must be trippable. For the control rod to be trippable, the control rods must insert on a reactor protection system (RPS) signal or on manual operator trip and the following must occur:
- 1.
The Reactor Trip Breakers open de-energize the Control Rod Drive electromagnetic gripper
- coils,
- 2.
The gripper must release,
- 3.
The mechanical components supporting the control rod must move to allow the control rod to fall into the core under the influence of gravity.
Since dropping a control rod while at power is undesirable (dropping a control rod in Modes 1 and 2 would result in radial or axial power tilts or oscillations), actions 1, 2 and 3 are tested during shutdown conditions via the control rod drop timing test. Control Rod G-3 inserted fully into the core in 1.77 to 1.83 seconds during the last three control rod drop tests (SR 3.1.4.3), well within the acceptance criteria of < 2.4 seconds. A review of rod G-3 past performance since 2010 shows no past history of binding. Additionally, there were no issues with control rod latching during refuel activities since 2010.
The rod exercising test is not intended to, and does not, test actions 1 or 2. When control rods are exercised, they are driven in 10 steps and then returned to their normal full-out position. This action assures that the drive mechanism can move the rod(s) without exceeding the rod group alignment limit. The stationary coils shown to be working based on the energization / de-energization during the last tests and its ability to hold the Control Rod G-3 in a stationary position (SR 3.1.4.1 provides a history showing whether the control rod is beginning to deviate from it's full out position). A search found no instances that Control Rod G-3 exercising / testing (rod drop testing or 10 step surveillance movements) has detected any of the occurrences in which mechanical binding of mechanical components has prevented or excessively slowed full control rod insertion.
The TS change is requested due to the likelihood of possible failures during the next surveillance test. If the TS change is approved, a temporary procedure change would include steps to open the lift-coil-disconnect switch for Control Rod G-3, during Shutdown Bank B rod motion to ensure the stationary gripper coil for Control Rod G-3 remains energized and rod alignment limits are maintained. The opening of the lift-coil-disconnect switch for Control Rod G-3 does not adversely affect the control rod design function because it does not impact shutting the reactor down by fully inserting the Control Rod G-3 into the bottom of the core because it does not affect the stationary gripper coil and grippers ability to de-energize upon opening of Reactor Trip Breakers. As discussed earlier, sending current to the stationary gripper coil moves the stationary grippers to engage the drive shaft to support the weight of the rod while the rod is not in motion and sending current to the moveable gripper coil moves the moveable gripper to engage the drive shaft to support the weight of the rod while the rod is in motion. Current to the lift coil will move the entire moveable gripper assembly up or down by one step. Removing power to the lift coil assures there will be no rod movement due to the lift coil. The stationary and movable coils will operate repetitively each step demanded during control bank motion. Connecting monitoring instrumentation on these coils will provide further assurance that the control rod remains trippable since it is loss of power to all three coils that initiates the trip and it will verify de-energizing /
energizing the stationary and movable coils and grippers. The instrumentation will also allow evaluation if a rod drop does occur since this is physically possible although not expected.
Therefore opening the lift connect disconnect does not adversely affect a method of evaluation that shows compliance with the control rod design function.
NL-15-021 Docket No. 50-247 Page 8 of 11 Access to the control rod drive mechanisms for repair of the control rod requires a plant shutdown, a cooldown to cold shutdown, and partial draining of the RCS. Eliminating this surveillance for CR G-3 should eliminate the need for an outage prior to the planned refueling outage if the control rod drops and cannot be recovered. By eliminating an additional maintenance outage the following plant challenges will not be necessary:
A significant power transient A plant thermal cycle Operation of plant safety systems (Auxiliary feedwater and shutdown cooling)
Exposure to plant conditions where the RCS pumps and steam generators are not available for decay heat removal Radiation exposure to plant personnel Generation of radioactive waste due to boration, dilution, and maintenance activities Lifting of heavy loads (reactor vessel head and missile blocks)
The Control Rod G-3 is considered operable so there is minimal additional risk to delaying the performance of the test for Control Rod G-3. In the event that Control Rod G-3 were unable to fully insert following the need for a reactor trip, the impact on ATWS risk is considered negligible. The failure criterion in NUREG/CR-5500 (Volume 2) for the RPS is conservatively assumed as 10 or more control rods failing to insert. Figure E-4 of NUREG/CR-5500 indicates that reducing the required number of failures from 10 to 9 to account for failure of Control Rod G-3 results in less than a 1E-6 increase in the overall rod failure probability. Therefore, given the aforementioned plant challenges associated with an additional plant shutdown to repair the Control Rod G-3 and the negligible risk increase associated with the potential failure of a single control rod to insert, the risk associated with a shutdown to repair rod G-3 is considered to be greater than the risk of continued full power operation for an equivalent period of time.
Based on Technical Specification 3.1.1, the shutdown margin SHALL be greater than or equal to 1.3% Ak/k. A preliminary shutdown margin calculation was performed (used Westinghouse software BEACON) based on the current core operating conditions, G-3 not inserting and the next highest reactive rod not inserting as well. The results have shown that there is greater than twice the required shutdown margin (e.g. greater than 2.6% Ak/k) with the previously stated conditions. An analysis is being performed to verify that the required shutdown margin is maintained through the end of the current operating cycle with G-3 and the most reactive rod stuck out of the core. The conclusions of this analysis will be provided as a supplement to this submittal.
5.0 REGULATORY ANALYSIS
5.1 No Significant Hazards Consideration Entergy has evaluated the proposed Technical Specification change using the criteria of 10CFR50.92 and found that no significant hazards consideration exist for the following reasons:
- 1)
Does the proposed License amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
NL-15-021 Docket No. 50-247 Page 9 of 11 The proposed change revises the requirement to perform SR 3.1.4.2 testing on Control Rod G-3 until the next refuel outage or forced outage of sufficient duration. Performing a technical specification surveillance test is not an accident initiator and does not increase the probability of an accident occurring. Since the control rod remains operable, the proposed change does not affect or create any accident initiators or precursors. The proposed revision to the test frequency is based on the ability of the control rod to continue to be able to perform its design function. The safety analyses assume control rod full insertion be de-energizing the CRDM coils and not the ability to move a full length control rod by its drive mechanism. The last rod drop test verified this ability so there is no increase in the consequences of an accident. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2)
Does the proposed License amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the requirement to perform SR 3.1.4.2 testing on Control Rod G-3 by changing the frequency of the test. The proposed change does not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed change in test frequency does not result in a change to the way that the equipment or facility is operated so that no new accident initiators are created. Therefore the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3)
Does the proposed License amendment involve a significant reduction in a margin of safety?
Response: No.
No. The conduct of performance tests on safety-related plant equipment is a means of assuring that the equipment is capable of performing its intended safety function and therefore maintaining the margin of safety established in the safety analysis for the facility. The proposed change revises the requirement to perform SR 3.1.4.2 testing on Control Rod G-3 by changing the frequency of the test. The proposed change is based the fact that there have been no problems with past tests of the Control Rod G-3 indicating the there are no problems with binding that could prevent the rod from inserting and a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance on rod position that would indicate any changes in position. There are no indications that the trip function would not work assuring the reduction in margin of safety is not significant.
Based on the above, Entergy concludes that the proposed amendment to the Indian Point 2 Technical Specifications presents no 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 5.2 Applicable Regulatory Requirements / Criteria IP2 was designed to the proposed Atomic Industrial Forum versions of the criteria issued for comment by the AEC on July 11, 1967. The plant design was later evaluated against the 10 CFR 50, Appendix A, General Design Criteria (GDC) in response to the February 11, 1980 confirmatory
NL-15-021 Docket No. 50-247 Page 10 of 11 order. The GDC that would be applicable to the Control Rod Drive System are 4, 23, 25, 26, 27, and 29.
GDC 4 requires structures, systems, and components important to safety to be designed to accommodate the effects of and to be compatible with the environmental conditions during normal plant operation as well as during postulated accidents. The proposed amendment has no effect on the design with respect to this criterion.
GDC 23 relates to the protection system failure modes such that the system shall fail into a safe state when conditions such as disconnection of system, loss of energy, or postulated adverse environment are experienced. This TS change relates to the testing of the Control Rod G-3 by movement of the rod which provides additional assurance that the rod will trip when required. The deferral of testing until the next outage will not affect the ability of the control rod to perform its function since it has performed in prior tests and there is no reason to believe the rod is stuck.
GDC 25, as it relates to the fuel design such that the specified limits are not exceeded for any single malfunction of the reactivity control system. Compliance with this GDC is maintained since Control Rod G-3 is operable.
GDC 26 requires two independent reactivity control systems of different design principles to be capable of reliably controlling reactivity changes under conditions of normal operation, including anticipated operational occurrences to assure acceptable fuel design limits are not exceeded. Compliance is maintained by continued availability of the Rod Control System and the Charging System.
GDC 27 requires the reactivity control system design to have a combined capability, in conjunction with poison addition by the emergency core cooling system to reliably control reactivity changes to assure that under postulated accident conditions the capability to cool the core is maintained. Because the TS change takes no system out of service, the capability of the IP2 systems to meet this criterion remain unchanged.
GDC 29 requires protecting against anticipated operational occurrences such that the design of the protection and reactor control systems should assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences. The TS change request involves no change to design and therefore there is no change to the extent of compliance with this GDC in the current design.
5.3 Environmental Considerations The proposed changes to the IP2 Technical Specifications do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent 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 for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
NL-15-021 Docket No. 50-247 Page 11 of 11
6.0 REFERENCES
- 1. NRC letter to Entergy Palisades "Issuance of Amendment Re: Control Rod Drive Exercise Surveillance (TAC No. ME3638), dated June 2, 2010
ATTACHMENT 2 TO NL-15-021 MARKUP OF TECHNICAL SPECIFICATION FOR CHANGE TO SURVEILLANCE REQUIREMENT 3.1.4.2 Changes indicated by lineout for deletion and Bold/Italics for additions Unit 2 Affected Pages:
TS 3.1.4 - 3 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247
Rod Group Alignment Limits 3.1.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. More than one rod not D.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within alignment limit, limits specified in the COLR.
OR D.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> required SDM to within limit.
AND D.2 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1
- NOTE -
Not required to be met for individual control rods until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after completion of control rod movement.
Verify individual rod positions within alignment limit.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.1.4.2 Verify rod freedom of movement (trippability) by 92 days moving each rod* not fully inserted in the core
_> 10 steps in one direction.
SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to criticality withdrawn position, is < 2.4 seconds from the gripper after each removal release to dashpot entry, with:
of the reactor head
- a.
Tavg >- 500OF and
- b.
All reactor coolant pumps operating.
- Control Rod G-3 need not be moved until repaired in the next forced outage of sufficient duration or the next refuel outage in 2016.
INDIAN POINT 2 3.1.4-3 Amendment No. 2-88