ML052280059

Tech Spec Pages for Amendment 226 Administrative Changes

ML052280059
Person / Time
Site:	Vermont Yankee
Issue date:	08/15/2005
From:	NRC/NRR/DLPM
To:
References
TAC MC5243
Download: ML052280059 (12)

Text

E. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level Entergy Nuclear Operations, Inc. is authorized to operate the facility at reactor core power levels not to exceed 1593 megawatts thermal in accordance with the Technical Specifications (Appendix A) appended hereto.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment 2 2 vare hereby incorporated in the license. Entergy Nuclear Operations, Inc. shall operate the facility in accordance with the Technical Specifications.

C. Reports Entergy Nuclear Operations, Inc. shall make reports in accordance with the requirements of the Technical Specifications.

D. This paragraph deleted by Amendment No. 226-E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:

Amendment No. 26, 208, 226

VYNPS BASES: 3.4 & 4.4 (Cont'd)

B. Operation With Inoperable Components Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. Assurance that the system will perform its intended function is obtained from the results of the pump and valve testing performed in accordance with the Requirements of Specification 4.6.E.

C. Standby Liquid Control System Tank - Borated Solution The solution saturation temperature varies with the concentration of sodium pentaborate. The solution shall be kept at least 10OF above the saturation temperature to guard against boron precipitation. The 101F margin is included in Figure 3.4.2. Temperature and liquid level alarms for the system are annunciated in the Control Room.

Once the solution has been made up, boron concentration will not vary unless more boron or water is added. Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change. Considering these factors, the test interval has been established.

Sodium pentaborate concentration is determined within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> following the addition of water or boron, or if the solution temperature drops below specified limits. The 24-hour limit allows for 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> of mixing, subsequent testing, and notification of shift personnel.

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Isotopic tests of the sodium pentaborate are performed periodically to ensure that the proper boron-10 atom percentage is being used.

10CFR50.62(c) (4)requires a Standby Liquid Control System with a minimum flow capacity and boron content equivalent to 86 gpm of 13 weight percent natural sodium pentaborate solution in the 251-inch reactor pressure vessel reference plant. Natural sodium pentaborate solution is 19.8 atom percent boron-10. The relationship expressed in Specification 3.4.C.3 also contains the ratio M251/M to account for the difference in water volume between the reference plant and Vermont Yankee. (This ratio of masses is 628,300 lbs./401,247 lbs.)

To comply with the ATWS rule, the combination of three Standby Liquid Control System parameters must be considered: boron concentration, Standby Liquid Control System pump flow rate, and boron-10 enrichment.

Fixing the pump flow rate in Specification 3.4.C.3 at the minimum flow rate of 35 gpm conservatively establishes a system parameter that can be used in satisfying the ATWS requirement, as well as the original system design basis. If the product of the expression in Specification 3.4.C.3 is equal to or greater than unity, the Standby Liquid Control System satisfies the requirements of 10CFR50.62(c)(4).

Amendment No. 41G, -8w, -75, G-, 226 4-19, 98

VYNPS BASES:

3.5 CORE AND CONTAINMENT COOLANT SYSTEMS A. Core Spray Cooling System and Low Pressure Coolant Injection System This Specification assures that adequate standby cooling capability is available whenever irradiated fuel is in the Reactor Vessel.

Based on the loss-of-coolant analyses, the Core Spray and LPCI Systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit the accident-caused core conditions as specified in 10CFR50, Appendix K. The analyses consider appropriate combinations of the two Core Spray Subsystems and the two LPCI Subsystems associated with various break locations and equipment availability in accordance with required single failure assumptions.

(Each LPCI Subsystem consists of the LPCI pumps, the recirculation pump discharge valve, and the LPCI injection valve which combine to inject torus water into a recirculation loop.)

The LPCI System is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system is completely independent of the Core Spray System; however, it does function in combination with the Core Spray System to prevent excessive fuel clad temperature. The LPCI and the Core Spray Systems provide adequate cooling for break areas up to and including the double-ended recirculation line break without assistance from the high pressure emergency Core Cooling Subsystems.

Specification 3.5.A.1 is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor pressure less than the RHR shutdown cooling permissive pressure, if capable of being manually realigned (remote) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during Hot Shutdown, if necessary.

The intent of these specifications is to prevent startup from the cold condition without all associated equipment being operable. However, during operation, certain components may be out of service for the specified allowable repair times. Assurance that the systems will perform their intended function is obtained from the results of the pump and valve testing performed in accordance with the requirements of Specification 4.6.E.

B. and C. Containment Spray Cooling Capability and RHR Service Water System The containment heat removal portion of the RHR System is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specified, the containment long-term pressure is limited to less than 5 psig and, therefore, the flow is more than ample to provide the required heat removal capability.

Reference:

Section 14.6.3.3.2 FSAR.

Each Containment Cooling Subsystem consists of two RHR service water pumps, 1 heat exchanger, and 2 RHR (LPCI) pumps. Either set of equipment is capable of performing the containment cooling function. In fact, an analysis in Section 14.6 of the FSAR shows that one subsystem consisting of 1 RHR service water pump, 1 heat exchanger, and 1 RHR pump has sufficient capacity to perform the cooling function. Assurance that the systems will perform their intended function is obtained from the results of the pump and valve testing performed in accordance with the requirements of Specification 4.6.E.

Amendment No. 2-, b4i, 4-2-,4-9-, 24 226 110

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION D. Safety and Relief Valves D. Safety and Relief Valves

1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification 4.6.E.

greater than 150 psig and The lift point of the temperature greater than safety and relief valves 3500 F, both safety valves shall be set as specified and at least three of the in Specification 2.2.B.

four relief valves shall be operable.

2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 150 psig and 350'F within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing The structural integrity and 1. Inservice inspection of the operability of the safety-related components safety-related systems and shall be performed in components shall be accordance with maintained at the level Section XI of the ASME required by the original Boiler and Pressure acceptance standards Vessel Code and throughout the life of the applicable Addenda as plant. required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.

Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternate measures approved by NRC Staff.

Amendment No. -3, 44, i'G,9 42, 4-G,44G, 4-2, 4W7, 19G, -19, 226 120

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

2. Operability testing of safety-related pumps and valves shall be performed in accordance with the Code of Record as required by 10 CFR 50.55a, except where specific written relief has been granted by the NRC.

F. Jet Pumps F. Jet Pumps

1. Whenever the reactor is 1. Whenever there is in the startup/hot recirculation flow with standby or run modes, all the reactor in the jet pumps shall be intact startup/hot standby or and all operating jet run modes, jet pump pumps shall be operable. integrity and operability If it is determined that shall be checked daily by a jet pump is inoperable, verifying that the an orderly shutdown shall following two conditions be initiated and the do not occur reactor shall be in a simultaneously:

cold shutdown condition within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. a. The recirculation pump flow differs by more than 10% from the established speed-flow characteristics.

b. The indicated total core flow is more than 10% greater than the core flow value derived from established power-core flow relationships.

2. Flow indication from each 2. In the event that the jet of the twenty jet pumps pump(s) fail the tests in shall be verified prior Specifications 4.6.F.l.a to initiation of reactor and 4.6.F.l.b, determine startup from a cold their operability by shutdown condition. verifying that each individual jet pump AP%

deviation from average loop AP does not vary from its normal established deviation by more than 10%.

Amendment No. 4a, 9, 4-44, 4-9., 226 121

VYNPS BASES: 3.6 and 4.6 (Cont'd) throughout plant life. The inservice inspection and testing programs are performed in accordance with 10CFR50, Section 50.55a except where specific relief has been granted by the NRC. These inspection and testing programs provide further assurance that gross defects are not occurring and ensure that safety-related components remain operable.

The type of inspection planned for each component depends on location, accessibility, and type of expected defect. Direct visual examination is proposed wherever possible since it is sensitive, fast, and reliable. Magnetic particle and liquid penetrant inspections are planned where practical, and where added sensitivity is required.

Ultrasonic testing and radiography shall be used where defects can occur on concealed surfaces.

Generic Letter 88-01 established the NRC position for in-service inspection of BWR austenitic stainless steel piping susceptible to Intergranular Stress Corrosion Cracking (IGSCC).

By letter dated November 9, 1998 (NVY 98-155), NRC approved use of ASME Code Case N-560 in association with inservice inspection of Class 1, Category B-J, piping welds under ASME Section XI. VY's ASME Category B-J piping welds are also Category A piping welds as defined in GL 88-

01. The Code Case reduces the inspection sample, while stipulating selection of that sample in accordance with a risk-informed analytical methodology.

The in-service inspection and testing programs are based on a thorough evaluation of present technology and state-of-the-art inspection and testing techniques.

F. Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.

Therefore, if a failure occurred, repairs must be made.

The following factors form the basis for the surveillance requirements:

• A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.

• The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.

• The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Specifications 4.6.F.l.a and b.

Amendment No 4-, 4-2&, 439, 4-44-, 1-96, 226 14 3

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

i. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and ii. Suspend core alterations; and iii. Initiate action to suspend operations with the potential for draining the reactor vessel.

C. Secondary Containment System C. Secondary Containment System

1. Secondary Containment 1. Secondary containment Integrity shall be capability to maintain a maintained during the 0.15 inch of water vacuum following modes or under calm wind conditions: (2<0<5 mph) conditions with a filter train flow

a. Whenever the reactor rate of not more than is in the Run Mode, 1,500 cfm, shall be Startup Mode, or Hot demonstrated at least Shutdown condition*; quarterly.

or

• NOTE: The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:

1. Reactor coolant temperature is < 2120 F;

2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and

3. No core alterations are in progress.

Amendment No. 444, 4-4-7-,a-9%, 42, 226 155a

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

b. During movement of irradiated fuel assemblies or the fuel cask in secondary containment; or

c. During alteration of the Reactor Core; or

d. During operations with the potential for draining the reactor vessel.

Amendment No. 4-4, 1-97, G-2, 226 156

rYNPS 3.7 LIMITING CONDITIONS FOR l 4.7 SURVEILLANCE REQUIREMENTS OPERATION

2. With Secondary 2. Intentionally blank.

Containment Integrity not maintained with the reactor in the Run Mode, Startup Mode, or Hot Shutdown condition, restore Secondary Containment Integrity within four (4) hours.

3. If Specification 3.7.C.2 3. Intentionally blank.

cannot be met, place the reactor in the Hot Shutdown condition within 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> and in the Cold Shutdown condition within the following 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

4. With Secondary 4. Intentionally blank.

Containment Integrity not maintained during movement of irradiated fuel assemblies or the fuel cask in secondary containment, during alteration of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and

b. Suspend alteration of the Reactor Core; and

c. Initiate action to suspend operations with the potential for draining the reactor vessel.

Amendment No. 44, -97, 226 157

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS OPERATION

4. 480 V Uninterruptible Power Systems From and after the date that one Uninterruptible Power System or its associated Motor Control Center are made or found to be inoperable for any reason, the requirements of Specification 3.5.A.4 shall be satisfied.

5. RPS Power Protection From and after the date that one of the two redundant RPS power protection panels on an in-service RPS MG set or alternate power supply is made or found to be inoperable, the associated RPS MG set or alternate supply will be taken out of service until the panel is restored to operable status.

C. Diesel Fuel C. Diesel Fuel There shall be a minimum of 1. The quantity of diesel 36,000 usable gallons of generator fuel shall be diesel fuel in the diesel logged weekly and after fuel oil storage tank. each operation of the unit.

2. Once a month a sample of diesel fuel shall be taken and checked for quality. The quality shall be within the applicable limits specified on Table I of ASTM D975-02 and logged.

Amendment No. 2-6, 44-2, 4@G, -44,226 218

VYNPS BASES: 4.10 (Cont'd) for the associated batteries. The results of these tests will be logged and compared with the manufacturer's recommendations of acceptability.

The Service Discharge Test (4.10.A.2.c) is a test of the batteries ability to satisfy the design requirements of the associated dc system. This test will be performed using simulated or actual loads at the rates and for the durations specified in the design load profile (battery duty cycle).

Assurance that the diesels will meet their intended function is obtained by periodic surveillance testing and the results obtained from the pump and valve testing performed in accordance with the requirements of Specification 4.6.E. Specification 4.10.B.l.a provides an allowance to avoid unnecessary testing of the operable emergency diesel generator (EDG). If it can be determined that the cause of the inoperable EDG (e.g., removal from service to perform routine maintenance or testing) does not exist on the operable EDG, demonstration of operability of the remaining EDG does not have to be performed. If the cause of inoperability exists on the remaining EDG, it is declared inoperable upon discovery, and Limiting Condition for Operation 3.10.B.1 requires reactor shutdown within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

Once the failure is repaired, and the common cause failure no longer exists, Specification 4.10.B.l.a is satisfied. If the cause of the initial inoperable EDG cannot be confirmed not to exist on the remaining EDG, performance of Surveillance Requirement (SR) 4.10.B.l.b suffices to provide assurance of continued operability of that EDG.

In the event the inoperable EDG is restored to operable status prior to completing either SR 4.10.B.l.a or SR 4.10.B.l.b, the plant corrective action program will continue to evaluate the common cause possibility. This continued evaluation, however, is no longer under the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> constraint imposed while in the condition of SR 4.10.B.1 or SR 4.10.B.3.b.2.

According to NRC Generic Letter 84-15, 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> is a reasonable time to confirm that the operable EDG is not affected by the same problem as the inoperable EDG.

Verification of operability of an off-site power source within one hour and once per eight hours thereafter as required by 4.10.B.3.b.1 may be performed as an administrative check by examining logs and other information to determine that required equipment is available and not out of service for maintenance or other reasons. It does not require performing the surveillance needed to demonstrate the operability of the equipment.

C. Logging the diesel fuel supply weekly and after each operation assures that the minimum fuel supply requirements will be maintained. During the monthly test for quality of the diesel fuel oil, a viscosity test and water and sediment test will be performed as described in ASTM D975-02. The quality of the diesel fuel oil will be acceptable if the results of the tests are within the limiting requirements for diesel fuel oils shown on Table 1 of ASTM D975-02.

Amendment No. 1-25, A4 , BVY 01 40, 2G-, 214, -44, 226 223

VYNPS 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY A. The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during absences.

B. The plant manager or designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect nuclear safety.

C. The shift supervisor shall be responsible for the control room command function. During any absence of the shift supervisor from the control room while the unit is in plant startup or normal operation, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command control room while the unit is in cold shutdown or refueling with fuel in the reactor, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

6.2 ORGANIZATION A. Onsite and Offsite Organizations Organizations shall be established for unit operation and corporate management. These organizations shall include the positions for activities affecting safety of the nuclear power plant.

1. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organizational positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Program Manual. The plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the Technical Requirements Manual.

2. The plant manager shall be responsible for overall unit safe operation and shall have control over-those on-site activities necessary for safe operation and maintenance of the plant.

3. The site vice president shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

Amendment No. A6-, s, 4 4-, 71, 214, 226 255