ML13085A081
| ML13085A081 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/21/2013 |
| From: | Grecheck E Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13085A081 (10) | |
Text
Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion VW'-b Address: www.dom.com March 21, 2013 U.S. Nuclear Regulatory Commission Serial No.13-031 Attention: Document Control Desk NSSL/WEB 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 FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7 - CONTROL ROD DRIVE MECHANISMS Pursuant to the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests an amendment, in the form of changes to the Technical Specifications (TS) to Facility Operating License Number DPR-65 for Millstone Power Station Unit 2 (MPS2). The proposed change would revise the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal. provides a description and assessment of the proposed change. Attachment 2 provides the marked-up TS page that reflects the proposed change to TS 3.1.3.7.
The proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92. The proposed change has been reviewed and approved by the Facility Safety Review Committee.
In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.
DNC requests approval of the proposed change by March 21, 2014 to support the next scheduled refueling outage for MPS2, currently scheduled for spring, 2014.
Should you have any questions in regard to this submittal, please contact Wanda Craft at (804) 273-4687.
Sincerely, VICKI L.HULL Eugene S. Grecheck NotaryPulIc Commoneull of riorginis Ili Vice President - Nuclear Engineering and Development 14052 C My Commission Expir May 31. 2014
. .V W.. .V -VW V -
COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President - Nuclear Engineering and Development of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute belief. and file the foregoing document in behalf S of that Company, i- and that the statements in the document are true to the best of his knowledge and Acknowledged before me thic_1.ý day of AfiýA.,.
2013.
My Commission Expires: All 341.10*7j A, Notary Public k l ii I
Serial No.13-031 Docket No. 50-336 Page 2 of 2 Commitments made in this letter: None Attachments:
- 1. Evaluation of Change to Technical Specification 3.1.3.7, Control Rod Drive Mechanisms
- 2. Marked-up Technical Specifications Pages cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 J. S. Kim Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2A 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.13-031 Docket No. 50-336 ATTACHMENT 1 EVALUATION OF CHANGE TO TECHNICAL SPECIFICATION 3.1.3.7.
CONTROL ROD DRIVE MECHANISMS DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2
Serial No.13-031 Docket No. 50-336 Attachment 1, Page 1 of 5 EVALUATION OF CHANGE TO TECHNICAL SPECIFICATION 3.1.3.7.
CONTROL ROD DRIVE MECHANISMS
1.0 INTRODUCTION
Pursuant to the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests an amendment, in the form of changes to the Technical Specifications (TS) to Facility Operating License Number DPR-65 for Millstone Power Station Unit 2 (MPS2).
The proposed change would revise the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal.
2.0 PROPOSED CHANGE
The proposed change is shown below (Note: Deleted text is struck-through and added text is italicized and bold):
- The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500 OF, the pressurizer pressure is greater than 2000 psia and the hio pow,- trip is OPER;,, E= requirements of Limiting Condition for Operation for Specification 3.3.1.1, Reactor Protective Instrumentation, are met.
A mark-up of the proposed TS change is provided in Attachment 2.
3.0 BACKGROUND
On April 21, 1987, TS 3.1.3.7, Control Rod Drive Mechanisms, was added to the MPS2 TSs under License Amendment No. 116. The purpose of the amendment was to bring the TSs into conformance with the safety analysis by requiring the CRDMs to be de-energized whenever the RCS boron concentration was less than the refueling concentration specified in TS 3.9.1, Boron Concentrations, for Modes 3, 4, 5 and 6. As part of this change, a footnote was included in the TS to specify that the CRDMs could be energized in Mode 3 as long as the following conditions were met: four reactor coolant pumps operating, RCS temperature >500 0 F, pressurizer pressure > 2000 psia, and high power trip is operable.
This TS ensured that whenever the plant could experience an uncontrolled rod withdrawal from subcriticality (i.e., there was power to the CRDMs) that plant conditions were consistent with the assumptions made in the safety analysis. The change also included an addition to TS Table 3.3-1, "Reactor Protective Instrumentation," requiring the High Power Level Trip to be operable in Mode 3 unless all the CRDMs were de-energized or the RCS boron concentration was greater than or equal to the refueling concentration of TS 3.9.1.
Serial No.13-031 Docket No. 50-336 Attachment 1, Page 2 of 5 On September 25, 2003, the NRC approved License Amendment No. 282 for MPS2. This amendment resulted in several changes to TSs including changes to TS 3.3.1.1, Reactor Protective Instrumentation. As part of this change, TS Table 3.3-1, Reactor Protective Instrumentation, was revised to include new functional units: RPS Logic Matrices, RPS Logic Matrix Relays, and Reactor Trip Breakers. The change required these functional units to be operable in Modes 1, 2, and
- with the asterisk denoted as follows: "With the protective system trip breakers in the closed position and the CEA drive system [i.e.,
CRDMs] capable of CEA withdrawal."
The footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, provides a link between TS 3.1.3.7 and TS Table 3.3-1, Reactor Protective Instrumentation, to ensure the RPS operability requirements are met when the CRDMs are energized and capable of CEA withdrawal. At the time of License Amendment 282, the need to revise the footnote in TS 3.1.3.7 to include the additional RPS operability requirements in TS Table 3.3-1 (i.e., RPS Logic Matrices, RPS Logic Matrix Relays, and Reactor Trip Breakers), was not recognized.
The footnote changes proposed herein will provide consistency with the requirements of TS Table 3.3-1 when CRDMs are energized and capable of CEA withdrawal.
4.0 TECHNICAL ANALYSIS
The CRDM requirement of TS 3.1.3.7, Control Rod Drive Mechanisms, is provided to assure that the consequences of an uncontrolled CEA withdrawal from a subcritical transient will stay within acceptable levels. This TS assures that reactor coolant system (RCS) conditions exist which are consistent with the plant safety analysis prior to energizing the CRDMs. The accident is precluded when CRDMs are de-energized and cannot withdraw a CEA. The CRDMs may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate shutdown margin is maintained even if the CEAs are fully withdrawn from the core.
TS 3.1.3.7, Control Rod Drive Mechanisms, requires that CRDMs are de-energized in Modes 3*, 4, 5 and 6, whenever the RCS boron concentration is less than the refueling concentration specified in TS 3.9.1, Boron Concentrations. As described in the footnote (denoted by an asterisk) in TS 3.1.3.7, CRDMs may be energized in Mode 3 as long as the following conditions exist: 1) four reactor coolant pumps operating, 2) RCS temperature
>500 0 F, 3) pressurizer pressure > 2000 psia, and 4) high power trip is operable.
TS Table 3.3-1, Reactor Protective Instrumentation, identifies, as a minimum, the reactor protective instrumentation channels and bypasses that must be operable under specific modes and conditions. For the condition where the protective system trip breakers are in the closed position and the CEA drive system is capable of CEA withdrawal (i.e., CRDMs energized), TS Table 3.3-1 requires the high power trip (Functional Unit 2), RPS logic matrices (Functional Unit 13), RPS logic matrix relays (Functional Unit 14) and reactor trip breakers (Functional Unit 15) to be operable.
Serial No.13-031 Docket No. 50-336 Attachment 1, Page 3 of 5 Currently, the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, only requires the High Power Level Trip to be operable when CRDMs are energized and capable of CEA withdrawal. It does not include the additional operability requirements (i.e., RPS logic matrices, RPS logic matrix relays and reactor trip breakers) that were later added to TS Table 3.3-1, Reactor Protective Instrumentation, under License Amendment 282. To correct this inconsistency between TS 3.1.3.7 and TS Table 3.3-1, the footnote in TS 3.1.3.7 will be revised to delete the reference to: high power trip is OPERABLE and replace it with the requirements of Limiting Condition for Operation for Specification 3.3.1.1, Reactor Protective Instrumentation, are met. This change will ensure the operability requirements are met when the CRDMs are energized in Mode 3.
5.0 REGULATORY ANALYSIS
5.1 No Significant Hazards Consideration The NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. DNC has evaluated whether or not a significant hazards consideration (SHC) is involved with the proposed change. A discussion of these standards as they relate to this change request is provided below.
Criterion 1 Will operation of the facility in accordance with the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to revise the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when CRDMs are energized and capable of withdrawal. The proposed change to the footnote in TS 3.1.3.7 does not modify the physical design or operation of the plant and does not increase the probability or consequences of an accident previously evaluated.
The proposed change has no impact on the operation of the CRDMs. In addition, the design basis accident remains unchanged for the postulated events described in the MPS2 Final Safety Analysis Report (FSAR). Since the initial conditions and assumptions included in the safety analyses are unchanged, the consequences of the postulated events remain unchanged. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Serial No.13-031 Docket No. 50-336 Attachment 1, Page 4 of 5 Criterion 2 Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the physical configuration of the plant (no new or different type of equipment will be installed) or introduce any operating configurations not previously evaluated. The proposed change does not alter the way any system, structure, or component (SSC) functions and does not alter the manner in which the plant is operated. The proposed change does not introduce any new failure modes and no new accident precursors are generated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
Criterion 3 Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?
Response: No.
The proposed change to the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, does not involve a change in the operational limits or physical design of the plant. The proposed change does not alter the function or operation of plant equipment or affect the response of that equipment if it is called upon to operate. The proposed change does not decrease the scope of equipment currently required to operate or subject to surveillance testing, nor does the proposed change affect any instrument setpoints or equipment safety functions.
The ability of operable SSCs to perform their designated safety function is unaffected by this proposed change. The proposed change does not reduce the margin of safety since it does not affect the assumptions in any accident analysis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Conclusion Based on the above, DNC concludes that the proposed change does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).
5.2 Applicable Regulatory Requirements/Criteria In Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.
Serial No.13-031 Docket No. 50-336 Attachment 1, Page 5 of 5 The proposed change does not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed change does not involve a change to any safety limits, limiting safety system settings, limiting control settings, limiting conditions for operation, surveillance requirements, design features, or administrative controls required by 10 CFR 50.36.
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.0 ENVIRONMENTAL CONSIDERATION
DNC has evaluated the proposed amendment and has determined that it does 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 to be prepared in connection with the proposed amendment.
Serial No.13-031 Docket No. 50-336 Attachment 2 Marked-up Technical Specifications Page DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2
Marceh 16, 2006 REACTIVITY CONTROL SYSTEMS I *CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized.
APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3.9.1.
ACTION:
With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or immediately open the reactor trip circuit breakers.
SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The control rod drive mechanisms may be energized for'ODE 3 as long as 4 reactor coolant pumps are OPERATINQ the reactor coolant system t lperature is greater than 5000 F, the pressurizer pressure is greater than 2000 psia and the high pwcr trip is OPE?6B,1ALE. I MILLSTONE - UNIT 2 3/4 1-31 Amendment No. +4-6,2-9+