Section 5.0 Administrative Controls

Question 1

Who has responsibility for overall plant operation? What happens during their absence?

Answer 1

Section 5.1.1 - Plant superintendent shall be responsible for overall plant operation and shall delegate in writing the succession for this responsibility during his absence.

Question 2

Who approves tests, experiments or modifications to systems that affect nuclear safety?

Answer 2

Section 5.1.1 - Plant superintendent or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

Question 3

Who has control room command.

Answer 3

Section 5.1.2 - Shift Supervisor shall be responsible for the control room command function. During any absence of the SS from the control room while the plant is in MODE 1-4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the SS from the control room while the plant is is Mode 5 or 6 and individual with an active SRO license or Reactor Operator license shall be designated to assume the control room comman function.

Question 4

What Entergy procedures provide direction on Section 5.1 Responsibility areas?

Answer 4

• EN-OP-115, Attachment 9.6
• FPIP-1 Fire Protection Plan, Organization and Responsiblities
• Site Emergency Plan (SEP)

Question 5

What grants orgnizational freedom to operations training, radiation safety and quality assurance staff?

Answer 5

Section 5.2.1 [d] - The individuals who train the operating staff and those who carry out radiation safety and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

Question 6

What provides for the delination of lines of authority, responsibility, etc….

Answer 6

Section 5.2.1 - Onsite and Offsite Organizations

Onsite and offsite organization shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the Palisades plant.

Question 7

How many non-licensed operators are required?

Answer 7

Section 5.2.2[a] - A non-licensed operator shall be assigned when fuel is in the reactor and an additional non-lincesed operator shall be assigned when the reactor is operating in Modes 1-4.

Question 8

What governs minimum staffing?

Answer 8

Section 5.2.2[a] - Shift crew componsition may be less than the minimum requirement of 10CFR50.54(m)(2)(i), and 5.2.2a. and 5.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to resore the shift crew to within the requirements.

Question 9

What managers are required to hold an SRO license?

Answer 9

Section 5.2.2[f] - The operations manager or an assistant operations manager shall hold an SRO license. The individual holding the SRO license shall be responsible for directing the activities of the licensed operators.

Question 10

When is a shift engineer to be assigned?

Answer 10

Section 5.2.2[g] - When in Modes 1-4, an individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reacor engineering, and plant anaylysis with regard to the safe operations of the plant. This individual shall meet the qualifications specificed by ANSI/ANS 3.1 as endorsed by RG 1.8, Rev 3, 2000.

Question 11

What requires procedures?

Answer 11

Section 5.4.1 - Written procedures hall be established, implemented, and maintained covering the activities reference below:

1. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
2. Emergency operating oprocedures
3. Specification 5.5 Programs

Question 12

What requires and defines the purpose of the ODCM?

Answer 12

The Offsite Dose Calculation Manual (ODCM) is required by Section 5.5.1 and is

a. Contain the methodology and parameters used in calculation of offsite doeses resulting from radioactive gaseous and liquid effluents, in the calculation of faseous and liquid effluent monitoring alarma nd trip setpoints and the conduct of the radiolgical environment monitoring program; and
b. Contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be indluced in Radiological Environmental Operating Report and Radioactive Effluent Release Report required by Specification 5.6.2 and 5.6.3

Question 13

What is the limit on leakage of PCS from recirculation heat removal systems’ components?

Answer 13

Section 5.5.2 - Primary Coolant Sources Outside Containment

[e] - The maximum allowable leakage from the recirculation heat removal systems’ components (which include valve stems, flanges and pump seals) shall not exceed 0.2 gallon per minute under the normal hydrostatic head from the SIRW tank.

Question 14

What is the limit for noble gas dose rates at or beyond the site boundary?

Answer 14

Section 5.5.4 - Radioactive Effluent Controls Program

A program shall be provided conforming with 10CFR 50.36a for the control of radioactive effluents and for maintaining the doeses to members of the public from radioactive effluents as low as resonably achievable. It is contained within the ODCM.

[e] For noble gases: a dose rate <= 500 mrem/yr to the whole body and a dose rate ,= 3000 mrem/yr to skin, and

For iodine-131, idoine-133, tritium, and all radionuclides in praticulate form with half-lives greater than 8 days: a diose rate of ,=1500 mrem/yr to any organ.

Question 15

What happens if 8 or more of the dome tendons are correctively retensioned?

Answer 15

Section 5.5.5 - Containment Strucutral Integrity Surveillance Program

If, as a result of a tendon inspection, corrective retensioning of five percent (8) or more of the total number of dome tendons is necessary to restore their liftoff force to witin the limits, ad dome delamination inspection shall be performed within 90 days following such corrective retensioning.

Question 16

How often are the flywheels on the primary coolant pumps inspected?

Answer 16

Section 5.5.6 - Primary Coolant Pump Flywheel Surveillance Program

Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each 10 years.

Question 17

What is a ASME Code Class 1 component?

Answer 17

1. Reactor fuel.
2. Primary circuit components whose rupture would result in a leakage of such magnitude that it could not be compensated for by the make-up water systems of the nuclear power plant. In conformity with this principle, the following primary circuit components remain outside Safety Class 1:

• small-diameter pipes (inner diameter not more than 20 mm)
• components connected to the reactor coolant system through a passive flow-limiting device and which, if ruptured, do not cause a leak larger than that caused by the rupture of a 20 mm pipe, as well as
• components which, in the event of their failure, can be isolated from the reactor coolant system by two successive, automatically closing valves whose closing time is short enough to allow for normal reactor shutdown and cooldown.

Question 18

What is an ASME Class 2 component?

Answer 18

1. Primary circuit components not assigned to Safety Class 1.
2. Systems and components required for a reactor trip.
3. Emergency core cooling systems intended for loss-of-coolant accidents.
4. The boron supply system required to shut down the reactor or to maintain it in a sub-critical condition during a postulated accident.
5. A decay heat removal system for circulating the water of the reactor coolant system.
6. At a PWR plant, the part of the make-up water system which is bounded by make-up water pumps and the primary circuit.
7. The following parts of the steam and feed water systems
8. At a PWR plant, the part inside the reactor containment that is bounded by the outermost isolation valves
9. At a PWR plant, the part of the emergency feed water system of the steam generators that is bounded by the emergency feed water pumps and steam generators, and
10. At a BWR plant, those parts of the steam system outside the reactor containment that are bounded by the isolation valves and the subsequent shut-off valves.
11. The reactor containment and related systems required to ensure containment integrity in a postulated accident. Such systems may be for example:
12. The containment spray system
13. Other systems intended for the reduction of pressure and temperature within the containment
14. Systems to prevent the formation of an explosive mixture of gases
15. Personnel and material locks, penetrations and other equivalent structures, and
16. Isolation valves of the reactor containment other than those included in the primary circuit, and parts of the piping penetrating the containment that are bounded by the valves.
17. Supporting structures of the primary circuit
18. Structures, such as emergency restraints and missile barriers, which protect components in Safety Class 1.
19. Internals of the reactor pressure vessel that support the reactor core and are important for its coolability.
20. Storage racks for fresh and spent fuel.
21. A protective instrumentation and automation system for starting a reactor trip, reactor emergency cooling, isolation of reactor containment or other safety function necessary in a postulated accident.
22. Electrical components and distribution systems necessary for the accomplishment of safety functions of systems in Safety Class 1 and 2.
23. Electrical power supply equipment ensuring electricity supply to Safety Class 2 components upon loss of both offsite power and power supplied by the main generators.

Question 19

What is an ASME 3 component?

Answer 19

1. The boron supply system bounded by the borated water storage tank in so far as the system or parts thereof are not classified to a higher safety class.
2. At a PWR plant, those parts of the reactor volume control system that are not assigned to a higher safety class.
3. At a PWR plant, those parts of the emergency feed water system that are not assigned to Safety Class
4. Systems needed for the cooling and pressure relief of the primary circuit, if they are not classified to a higher safety class.
5. Cooling systems, including their cooling water channels and tunnels, essential for the removal of reactor decay heat, decay heat from spent fuel stored outside the reactor, heat generated by Safety Class 2 components, heat generated by the above-mentioned systems themselves into the ultimate heat sink, and which do not belong to a higher safety class.
6. Parts of the sealing water, pressurised air, lubricating, fuel, etc systems necessary for the start-up or operation of systems in Safety Classes 2 and 3.
7. Systems for treating liquids or gases containing radioactive substances the failure of which could, compared to normal conditions, result in a significant dose increase to a plant employee or a member of the public. Examples of such systems are: Reactor cooling water cleanup system, sampling systems of the primary circuit treatment and storage systems for liquid wastes, and radioactive gas treatment systems.
8. Ventilation systems that reduce the radiation exposure of employees or the releases of radioactive materials into the environment. Below are examples of the functions of these systems: maintaining of pressure differences in the reactor building and filtering of its exhaust air (including the containment with its surrounding spaces), ventilation of those rooms in the auxiliary building where radioactive contamination could occur, ventilation of the spent fuel storage, ventilation of quarters containing radioactive waste, ventilation of laboratories where considerable amounts of radioactive materials are handled, and securing of working conditions in the control room and other rooms requiring continued stay during accidents, in case the air on-site contains radioactive or other hazardous materials.
9. Air cooling and heating systems in rooms containing components classified to Safety Classes 1, 2 and 3; the systems are needed to maintain the temperature required for ensuring reliable functioning of the equipment, taking into account extreme outdoor air temperatures and the waste heat released in these rooms.
10. Those reactor pressure vessel internals not assigned to Safety Class 2.
11. Nuclear fuel handling and inspection systems whose malfunction could endanger fuel integrity.
12. The following hoisting and transfer equipment: those parts of the control rod drives that are not assigned to Safety Class 1 or 2, the reactor building main crane, equipment needed for the lifting and transfer of nuclear fuel.
13. Storages of spent fuel and liquid wastes, including pools and tanks.
14. Buildings and structures designed to protect or support equipment in Safety Classes 2 or 3 and the failure of which could endanger the integrity of the equipment, protect workers to assure their ability to maintain functions important to safety in accident conditions.
15. Concrete structures inside the reactor containment other than those assigned to Safety Class 2.
16. Instrumentation and automation systems and components required for the following functions and not classified to a higher safety class: reactor power limitation systems, control of reactor main parameters (power, pressure, coolant volume), monitoring and control of safety functions during accidents, monitoring and control of reactor power peaking, monitoring and control of safe plant shutdown from the main and standby control rooms, monitoring of reactor criticality during fuel loading, monitoring of primary circuit leaks, monitoring of hydrogen and oxygen concentrations inside the containment, monitoring of primary circuit water chemistry, on-site radiation monitoring during accidents, monitoring of radioactive releases, monitoring for radiation in rooms.
17. Electrical components and electric power distribution systems required to accomplish the safety functions of Safety Class 3 systems.
18. Systems designed to ensure the integrity of the reactor containment or to limit releases especially in a severe accident. Examples thereof are: systems limiting the containment pressure, systems intended for the control and filtering of releases out of the containment, air circulating and filtering systems that clean the containment air space, systems that prevent the formation of an explosive gas mixture, systems intended for the monitoring of the condition of the reactor and the containment, and systems and components required for cooling a molten core and for ensuring the integrity of containment penetrations and other openings.

Question 20

What defines the frequency for inservice testing? What is quarterly?

Answer 20

Section 5.5.7 Inservice Testing Program

This program provides controls for inservice testing of ASME Code Class 1 ,2 and 3 components.

Quarterly <= 31 days.

Question 21

At what through wall thickness are steam generator tubes plugged?

Answer 21

Section 5.5.8 Steam Generator Program

[c] Tubes found by inservice inspection to contain flaws with a dpeth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

Question 22

What defines Steam Generator accident induced leakage performance criterion?

Answer 22

Section 5.5.8 Steam Generator Program

b[2] - The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for individual SG. Leakage is not to exceed 0.3 gpm.

Question 23

What is the purpose of the Secondary Water Chemistry Program?

Answer 23

Section 5.5.9 Secondary Water Chemistry Program

A program shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include…..

Question 24

What defines the ventilation system filter testing for charcoal and HEPA filters?

Answer 24

Section 5.510 Ventilation Feilter Testing Program

• V-8A/B
• V-26A/B
• VF-66

Question 25

What conditions are tested as part of the Fuel Oil Testing Program?

Answer 25

Section 5.5.11 Fuel Oil Testing Program

1. API gravity or an absolute specific gravity
2. Kinematic viscosity, and
3. Water and sediment content

Covers acceptability of new fuel oil prior to addition ot the Fuel Oil Storage Tank, and acceptability of fuel oil stored in the Fuel Oil Storage Tank.

Question 26

What governs changes to the Technical Specifications Bases documents?

Answer 26

Section 5.5.12 - Technical Specifications (TS) Bases Control Program

Question 27

What is LCO 3.0.6?

Answer 27

When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with
this supported system are not required to be entered.

Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system.

In this event, an evaluation shall be performed. If a loss of safety function is determined to exist, the appropriate Conditions and Required Actions
of the LCO in which the loss of safety function exists are required to be entered.

When a support system’s Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Question 28

What is the Safety Function Determination Program?

Answer 28

Section 5.5.13 - SFDP

This program ensures loss of safety function is detected and approprate actions taken.

It contains the following:

• Provisions for cross train checks to ensure a loss of the capability ot perform the safety function assumed in the accident analysis does not go undetected;
• Provisions for ensuring the plant is maintained in a safe conditon if a loss of function condition exits;
• Provision to ensure that an inoperable supported system’s Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
• Other appropriate limitation and remedial or compensatory actions.

Question 29

When does a loss of safety function exist?

Answer 29

Section 5.5.13 - SFDP

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analyais cannot be performed.

A loss of safety function MAY exist when a support system is inoperable, and:

1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
2. A required system redundant to system(s0 in turn supported by the inoperable supported system is also inoperable; or
3. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

Question 30

What is containment design pressure and what governs testing of its leakage?

Answer 30

Section 5.5.14 - Containment Leak Rate Testing Program

Design Pressure is 55 psig

Provides direction on when and how testing is completed.

Question 31

What directs the current formula, sampling, analyses, tests, and determinations to ensure solid packaged radioactive wastes comply with 10 CFR20, COCFR 71, Federal and State Regulation, and other requriements governing the disposal of radioactive waste?

Answer 31

Section 5.5.15 - Process Control Program

Not effective until after approval by plant superintendent…

Question 32

Under what condtions is the control room maintained habitable?

Answer 32

Section 5.5.16 - Control Room Envelope Habitability Program

Shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation Filtration, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe conditon folloiwng a radiological event, hazardous chemical releaswe, or a smoke challenge.

Question 33

When is the Radiologic Environmental Operating Report due and what does it cover?

Answer 33

Section 5.6.2 - Radiological Environmental Operatirng Report

Due before May 15 of each year for the prior year.

Includes summaries, interpretations, and analysis of trends of the results of the radiological environmental montiroing program for the reporting period.

Question 34

When is the Radioactive Effluent Release Report due and what does it include?

Answer 34

Section 5.6.3 - Radioactive Effluent Release Report

Due May 1 of each year for the prior year.

Summary of the quantitites of radioactive liquid and gaseous effluents and solid waste released from the plant.

Question 35

What governs and what is provided in the Core Operating Limits Report?

Answer 35

Section 5.6.5 COLR

Established prior to each reload cyccle, or prior to any remaining portion of a reload cycle and documents:

• 3.1.1 Shutdown Margin
• 3.1.6 Regulating Rod Group Position Limits
• 3.2.1 Linear Heat Rate Limits
• 3.2.2 Radial Peaking Factor Limits
• 3.2.4 ASI Limits
• 3.4.1 DNB Limits

Question 36

When TS 3.3.7 requires a report submitted for “Post Accident Monitoring Instrumentation,” inoperable, when is it due and what does it cover?

Answer 36

Section 5.6.6 Post Accident Monitoring Report

Shall be submitted within the following 14 days.

The report shall outline the preplannned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status.

Question 37

When is the Containment Structural Integrity Surveillance Report due?

Answer 37

Section 5.6.7 Containment Structural Integrity Surveillance Report

Shall be submitted to the NRC covering Prestressing, Anchroage, and Dome Delamination tests within 90 days after completion of the tests.

Question 38

When is the steam generator tube inspection report due?

Answer 38

Section 5.6.8 Steam Generator Tube Inspection Report

Shall be submitted within 180 days after the initial entry into MODE 4 follwing completion of an inspection performed in accordance with the Specification 5..58 Steam Generator Program.

Question 39

What controls access to a high radiation area at less than 1 R/hr?

Answer 39

Section 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 Rem/hour at 30 Centimeters from the Radiation Source or from any Surface Pentrated by the Radiation.

Question 40

What controls access to high radiation areas at greater than 1 R per hour?

Answer 40

Section 5.7.2 High Radiation Areas with Dose Rates Exceeding 1.0 Rem/hour at 30 Centimeters from the Radiation Source or from any Surface Pentrated by the Radiation.