ML041250176
| ML041250176 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/29/2004 |
| From: | Bezilla M FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 10 CFR 50.90, 3022, LAR 03-0023 | |
| Download: ML041250176 (19) | |
Text
N FENOC
-_ _ 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Mark B. Bezilla 419-321-7676 Vice President- Nuclear Docket Number 50-346 10 CFR 50.90 License Number NPF-3 Serial Number 3022 April 29, 2004 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001
Subject:
Davis-Besse Nuclear Power Station License Amendment Application to Relocate Technical Specification Surveillance Requirement (SR) 4.4.10.l.b to the Technical Requirements Manual (TRM)
[License Amendment Request (LAR) 03-0023]
Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, the following amendment is requested for the Davis-Besse Nuclear Power Station, Unit I (DBNPS). The proposed amendment would revise Technical Specification (TS) 3/4.4.10, "Reactor Coolant System - Structural Integrity, ASME Code Class 1,2, and 3 Components," to relocate Surveillance Requirement (SR) 4.4.10.1 .b to the Technical Requirements Manual (TRM). The DBNPS TRM is a licensee-controlled document that is incorporated by reference into the DBNPS Updated Safety Analysis Report. Changes to the DBNPS TRM are performed in accordance with the regulatory requirements of 10 CFR 50.59.
TS Limiting Condition for Operation (LCO) 3.4.10.1 states that the structural integrity of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, 2, and 3 components are to be maintained in accordance with SR 4.4.10.1. SR 4.4.10.1.b requires that the reactor vessel internals vent valves be tested and inspected at least once per twenty-four months.
Relocation of SR 4.4.10.1.b from the TS to the TRM is acceptable because there is no indication provided to the operator of the positions of these valves during plant operation, and no testing of these valves that can be performed on line. Removal of the reactor pressure vessel head is necessary to perform this SR. Thus, SR 4.4.10.1 .b does not reflect requirements of immediate importance to the operator, and it is not necessary to retain SR 4.4.10.1 .b in the TS to protect the A,DA
N Docket Number 50-346 License Number NPF-3 Serial Number 3022 Page 2 public health and safety. Any changes to SRs in the TRM are made under the regulatory requirements of 10 CFR 50.59, which ensures changes are properly evaluated.
Of note are the improved Technical Specifications, NUREG-1430, StandardTechnical Specifications- Babcock and MYilcox Plants, Revision 2, which do not contain a surveillance requirement for the reactor vessel internals vent valves. These valves are common to the Babcock and Wilcox-type plant design. Accordingly, based on their absence from NUREG-1430, the existing requirement can be relocated to the TRM. to this letter contains the technical analysis for the proposed changes and the proposed no significant hazards consideration determination.
These valves are currently required to be tested by September 22, 2005. Based on historical performance data and plant chemistry conditions, the FirstEnergy Nuclear Operating Company believes it can perform a 10 CFR 50.59 evaluation, following relocation of this SR to the USAR TRM, to extend the surveillance interval by approximately four months to the end of the present operating cycle. To provide for that evaluation and outage planning, the DBNPS requests that this license amendment be approved by November 30, 2004. Once approved, the amendment will be implemented within 120 days.
The proposed changes have been reviewed by the DBNPS on-site review committee and off-site review board.
Should you have any questions or require additional information, please contact Mr. Gregory A. Dunn, Manager - Regulatory Affairs, at (419) 321-8450.
The statements contained in this submittal, including its associated enclosures and attachments are true and correct to the best of my knowledge and belief. I am authorized by the FirstEnergy Nuclear Operating Company to make this submittal. I declare under penalty of perjury that the foregoing is true and correct.
Executed on: ______.______ix__
By: : _ _ _ _
Nark B. Bezilla, Vice Preident-Nuclear
-- \11' Docket Number 50-346 License Number NPF-3 Serial Number 3022 Page 3 MSH Enclosures cc: Regional Administrator, NRC Region III J. B. Hopkins, NRC/NRR Senior Project Manager D. J. Shipley, Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
C. S. Thomas, NRC Region III, DB-1 Senior Resident Inspector Utility Radiological Safety Board
Docket Number 50-346 License Number NPF-3 Serial Number 3022 Enclosure I DAVIS-BESSE NUCLEAR POWER STATION EVALUATION FOR LICENSE AMENDMENT REQUEST NUMBER 03-0023 (14 pages follow)
\"I LAR 03-0023 Page 1 DAVIS-BESSE NUCLEAR POWER STATION EVALUATION FOR LICENSE AMENDMENT REQUEST NUMBER 03-0023
Subject:
License Amendment Application to Revise Technical Specification (TS) 3/4.4.10.1, Reactor Coolant System - Structural Integrity, ASME Code Class 1, 2, and 3 Components to Relocate TS Surveillance Requirement (SR) 4.4.10.1.b to the Technical Requirements Manual (TRM)
1.0 DESCRIPTION
2.0 PROPOSED CHANGE
3.0 BACKGROUND
4.0 TECHNICAL ANALYSIS
5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria
6.0 ENVIRONMENTAL CONSIDERATION
7.0 REFERENCES
8.0 ATTACHMENTS
LAR 03-0023 Page 2
1.0 DESCRIPTION
This is a request to amend the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS)
Facility Operating License Number NPF-3 Appendix A Technical Specifications (TS).
The proposed change revises TS 3/4.4.10.1, "Reactor Coolant System - Structural Integrity, ASME Code Class 1, 2, and 3 Components," to relocate Surveillance Requirement (SR) 4.4.10.1.b to the Technical Requirements Manual (TRM). The DBNPS TRM is a licensee-controlled document that is incorporated by reference in Section 1.5.5 of the DBNPS Updated Safety Analysis Report (USAR). Changes to the TRM are performed in accordance with the regulatory requirements of 10 CFR 50.59.
TS Limiting Condition for Operation (LCO) 3.4.10.1 states that the structural integrity of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, 2, and 3 components are to be maintained in accordance with SR 4.4.10.1. SR 4.4.10.1.b requires that the reactor vessel internals vent valves (RVVVs) be tested and inspected at least once per twenty-four months (up to 30 months when invoking TS 4.0.2). The RVVVs are 14 inch swing check valves mounted vertically between the inlet and outlet sides of the core support shield.
Their safety function is to relieve pressure that could be generated by voiding in the core following a loss of coolant accident.
Relocating SR 4.4.10.1.b from the TS to the TRM is acceptable because there is no RVVV position indication provided to the operator during plant operation, and because the RVVVs can not be tested on-line. Removal of the reactor vessel head is necessary to perform this SR. Thus, this SR does not reflect requirements of immediate importance to the operator, and it is not necessary to retain this SR in the TS to protect the public health and safety. Any changes to SRs in the TRM are made under the regulatory requirements of 10 CFR 50.59, which ensures changes are properly evaluated.
Of note are the improved Technical Specifications, NUREG-1430, Standard Technical Specifications - Babcock and Milcox Plants, Revision 2, which do not contain a surveillance requirement for the RVVVs. The RVVVs are common to the Babcock and Wilcox-type plant design. Accordingly, based on their absence from NUREG-1430, the existing requirement can be relocated to the TRM.
The RVVVs are currently required to be tested by September 22, 2005. Based on historical performance data and plant chemistry conditions, the FirstEnergy Nuclear Operating Company believes it can perform a 10 CFR 50.59 evaluation, following relocation of this SR to the TRM, to extend the surveillance interval by approximately 4 months, to the end of the present operating cycle. This would be approximately 34 months between inspections. Of note is the NRC's previous approval of a one-time interval extension to approximately 42 months under DBNPS License Amendment No. 95, dated August 20, 1986.
LAR 03-0023 Page 3 There are no associated changes to the TS Bases being proposed under this license amendment request. The affected TS Bases page is included in Attachment 3 for information, and can be revised under the DBNPS Technical Bases Control Program of TS 6.17.
2.0 PROPOSED CHANGE
The proposed change affects SR 4.4.10.1.b, and is shown on the TS markup in Attachment 1.
SR 4.4.10.1.b concerns surveillance requirements for the RVVVs and currently states:
4.4.10.1 In addition to the requirements of Specification 4.0.5:
- b. Each internals vent valve shall be demonstrated OPERABLE at least once per 24 months during shutdown by:
- 1. Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation,
- 2. Verifying the valve is not stuck in an open position, and
- 3. Verifying through manual actuation that the valve is fully open when a force of < 400 lbs. is applied vertically upward.
This requirement will be relocated from the TS to the USAR TRM with changes only made to the format. An action consistent with Limiting Condition for Operation 3.4.10.1 Action a will also be established in the TRM for these valves. There are no associated changes to the TS Bases.
3.0 BACKGROUND
The RVVVs are described in USAR Sections 4.2.2, "Reactor Vessel Internals," and 6.3.3.1.2.1, "Boron Precipitation Control". There are four fourteen-inch diameter reactor vessel internals vent valves located on the core support shield.
The proposed change revises TS 3/4.4.10.1, "Reactor Coolant System - Structural Integrity, ASME Code Class 1, 2, and 3 Components," to Relocate Surveillance Requirement (SR) 4.4.10.1.b to the Technical Requirements Manual (TRM). The proposed relocation would not by itself alter the requirements, except for format. Any changes made to the requirements after relocation to the TRM would be performed in accordance with the regulatory requirements of 10 CFR 50.59 and any other applicable regulations. The proposed relocation of the present
LAR 03-0023 Page 4 requirements of SR 4.4.10.1 .b is consistent with its absence from NUREG-1430, Standard Technical SpeciJications- Babcock and Wilcox Plants, Revision 2.
The DBNPS SR 4.4.10.1.b requires the RVVVs be demonstrated operable at least once per 24 months, with a provision under TS 4.0.2 for extension of this period by twenty-five percent (for a total of 30 months). The reactor head must be removed to complete the SR. Since the DBNPS has undergone a prolonged outage since the last inspection and exercise of the RVVVs (March 2003), the time period will expire before DBNPS will complete another 24 month operating cycle. Accordingly, following the relocation of SR 4.4.10.1.b to the TRM, the DBNPS will perform an evaluation under the requirements of 10 CFR 50.59 to determine the acceptability of extending the surveillance interval to the next refueling outage. This is currently estimated to be an extension of approximately four months, resulting in approximately 34 months between performances of the SR.
4.0 TECHNICAL ANALYSIS
The RVVVs are large swing check valves mounted vertically between the inlet and outlet sides of the core support shield. The core support shield directs cold leg (inlet) flow downward into the annular space just inside the vessel and contains core outlet flow in the central portion, directing it upward to the hot leg nozzles. The vent valve assemblies are installed so they can swing outward into the cold leg water space should pressure on the outlet side of the core exceed inlet pressure. During normal operation and most plant transient conditions, these valves are held closed by both gravity and the higher pressure on the core inlet side.
The safety function of the RVVVs is to relieve the pressure that could be generated by voiding in the core following a Loss of Coolant Accident. Steam formed in the core that would otherwise block the entry of Emergency Core Cooling System (ECCS) injection water is relieved through the valves, allowing the core to remain covered. In the unlikely event of a rupture of the reactor inlet piping, the vent valves promote expulsion of steam produced in the core directly to the break, thus enhancing the effectiveness of ECCS injection. The valves serve a safety-related function, and are subject to an inspection and manual actuation to fulfill the following purposes:
I. Ensure valve operability.
- 2. Ensure that the valves are not stuck during normal operation.
- 3. Demonstrate that the valves are fully open at the forces equivalent to the differential pressure assumed in the safety analysis.
The inspection program for these valves also ensures that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. The RVVVs were successfully inspected and exercised in March 2003.
The proposed change would relocate the inspection requirements from TS to the TRM, which is incorporated by reference in the DBNPS USAR. Since this relocation is not changing the requirements, there is no adverse impact on nuclear safety. Any future changes to the relocated requirements would require application of the regulatory requirements of 10 CFR 50.59.
LAR 03-0023 Page 5 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration The proposed change would relocate the requirements of Technical Specification Surveillance Requirement (SR) 4.4.10.1.b. Presently SR 4.4.10.1.b requires inspection and manual actuation of the reactor vessel internals vent valves at least once per 24 months during shutdown, to provide a basis for operability during operation. The present requirements of SR 4.4.10.1 .b will be relocated to the DBNPS-controlled Technical Requirements Manual (TRM) upon implementation of the approved amendment. The TRM is incorporated by reference into the DBNPS Updated Safety Analysis Report. Changes to the TRM are performed in accordance with the regulatory requirements of 10 CFR 50.59.
An evaluation has been performed to determine whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed surveillance requirement relocation from the Technical Specifications to the USAR TRM does not alter the design, operation, or testing of any structure, system, or component. No previously analyzed accident scenario is changed. Initiating conditions and assumptions remain as previously analyzed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed surveillance requirement relocation from the Technical Specifications to the USAR TRM does not alter the design, operation, or testing of any structure, system, or component. The proposed change does not introduce any new or different accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
LAR 03-0023 Page 6
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed surveillance requirements relocation from the Technical Specifications to the USAR TRM does not affect the capabilities of the Reactor Vessel Internals Vent Valves. Therefore, the proposed change will not affect a margin of safety.
Based on the above, it is concluded that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, there is a finding of "no significant hazards consideration."
- 5.2 Applicable Regulatorv Requirements/Criteria 10 CFR 50.36 contains requirements for content of operating license Technical Specifications. Specifically, 10 CFR 50.36(c)(3) states surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that operation will be within safety limits, and that the limiting conditions for operation will be met.
The requirements of SR 4.4.10.1.b for inspection and testing the RVVVs are not necessary for assuring the structural integrity of ASME Code Class 1, 2, and 3 components required by Limiting Condition for Operation (LCO) 3.4.10.1.
Relocation of this SR to the TRM is acceptable because there is no indication provided to the operator of the position of these valves during plant operation, and no testing of these valves that can be performed on-line. Removal of the reactor vessel head is necessary to perform this SR. Thus, this SR does not reflect requirements of immediate importance to the operator, and it is not necessary to retain this SR in the TS to protect the public health and safety.
Concurrent with the implementation of this requested license amendment, this SR will be relocated from the TS to the TRM with changes only made to the format.
An action consistent with LCO 3.4.10.1 Action a will also be established in the TRM for these valves. Any future change to the SR in the TRM will be made under the regulatory requirements of 10 CFR 50.59, which ensures changes are properly evaluated. Accordingly, since the criteria of 10 CFR 50.36(c)(3) is not met by this SR, it can be relocated to the TRM.
The NRC utilized 10 CFR 50.36 in the development of NUREG-1430, Standard Technical Specifications - Babock and Wilcox Plants,Revision 2. Relocating the requirements of SR 4.4.10.1.b is consistent with its absence from NUREG-1430, Revision 2, and therefore does not affect the conformance of the DBNPS Operating License TS with the requirements of 10 CFR 50.36.
LAR 03-0023 Page 7 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 NRC's regulations, and (3) the issuance of the amendment wvill not be inimical to the common defense and security or to the health and safety of the public.
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement. However, the proposed amendment 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 be prepared in connection with the proposed amendment.
7.0 REFERENCES
- 2. DBNPS Updated Safety Analysis Report through Revision 23.
- 3. DBNPS Technical Requirements Manual through Revision 19.
- 4. 10 CFR 50.36, Technical Specifications
- 5. 10 CFR 50.59, Changes, Tests and Experiments
- 6. NUREG-1430, Revision 2, Standard Technical Specifications - Babcock and Wilcox Plants 8.0 ATTACHMENTS I. Proposed Mark-Up of Technical Specification Pages
- 2. Proposed Retyped Technical Specification Pages
- 3. Technical Specification Bases Pages
LAR 03-0023 PROPOSED MARK-UP OF TECHNICAL SPECIFICATION PAGES (2 pages follow)
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1. 2 and 3 COMPONENTS LIMTRICQNITION FORQPERATlQN 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.
APPLICABILITY: All MODES.
ACTION:
- a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50TF above the minimum temperature required by NDT considerations.
- b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200'F.
- c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the component(s) to within its limit or isolate the affected component(s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5-,
ae-4nservice inspection of each reactor coolant pump flywheel shall be performed at least once every 10 years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a surface examination of exposed surfaces of the disassembled flywheel. The recommendations delineated in Regulatory Guide 1.14, Revision 1, August 1975, Positions 3, 4 and 5 of Section C.4.b shall apply.
DAVIS-BESSE, UNIT 1 3/4 4-30 Amendment No. 232,
REACTOR COOLANT SYSTEM SURYEILLANCE REQUIREMENTS (Continued)
- b. Each intemals v ent valve shall be d(menstrated OnPERALE at least once ner 24 mar during shutdoe~ by:
- 1. Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation,
- 2. Verifying the valve is not stuck in an open pesition, and
- 3. Verifying through manual actuation that the valve is fully open when a force of -
400 lbs. is applied veicrially upward.
DAVIS-BESSE, UNIT I 3/4 4-3 1 Amendment No. 23, 95, 165
LAR 03-0023 PROPOSED RETYPED TECHNICAL SPECIFICATION PAGE (1 page follows, and Page 3/4 4-31 is deleted)
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.
APPLICABILITY: All MODES.
ACTION:
- a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations.
- b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200TF.
- c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the component(s) to within its limit or isolate the affected component(s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5, inservice inspection of each reactor coolant pump flywheel shall be performed at least once every 10 years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a surface examination of exposed surfaces of the disassembled flywheel. The recommendations delineated in Regulatory Guide 1.14, Revision 1, August 1975, Positions 3, 4 and 5 of Section C.4.b shall apply.
DAVIS-BESSE, UNIT I 3/4 4-30 Amendment No. 232,
LAR 03-0023 TECHNICAL SPECIFICATION BASES PAGES (1 page follows)
Note: The Basespage isprovidedforinformation only.
I REACTOR COOLANT SYSTEM INFORMATION ONLY BASES 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASHE Code Class 1, 2 and 3 components, except steam generator tubes, ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrate that the valves are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.
3.4.4.11 HIGH POINT VENTS Deleted DAVIS-BESSE, UNIT I B 3/4 4-13 Amendment No.--W, 201
- 0 i Docket Number 50-346 License Number NPF-3 Serial Number 3022 Enclosure 2 COMMITMENT LIST THE FOLLOWING LIST IDENTIFIES THOSE ACTIONS COMMITTED TO BY THE DAVIS-BESSE NUCLEAR POWER STATION (DBNPS) IN THIS DOCUMENT. ANY OTHER ACTIONS DISCUSSED IN THE SUBMITTAL REPRESENT INTENDED OR PLANNED ACTIONS BY THE DBNPS. THEY ARE DESCRIBED ONLY FOR INFORMATION AND ARE NOT REGULATORY COMMITMENTS. PLEASE NOTIFY THE MANAGER - REGULATORY AFFAIRS (419-321-8450) AT THE DBNPS OF ANY QUESTIONS REGARDING THIS DOCUMENT OR ANY ASSOCIATED REGULATORY COMMITMENTS.
COMMITMENTS DUE DATE Relocate Reactor Vessel Internals Vent Valves Upon implementation of the Surveillance Requirement 4.4.10.1.b to the Technical proposed license amendment.
Requirements Manual and establish an Action consistent with TS LCO 3.4.10.1 Action a.