Revision as of 14:58, 6 April 2019

Text

2016301 Final SRO Written Exam - Delay Release 2 Years
	ML17017A385
Person / Time
Site:	Brunswick
Issue date:	01/27/2017
From:	; NRC/RGN-II/DRS/OLB
To:	;
References
	Download: ML17017A385 (450)
	v • d • e

SRO Written Exam Reference Index

1. 0EOP-01-NL , EOP/SAMG Numerical Limits and Values , Attachment 3, Containment Parameters, Secondary Containment Area Temperature Limits, Table 3

-B 2. 0EOP-01-SBO-01, Attachment 4, Drywell Temperature Calculation Using RSDP Recorder Inputs

3. 0EOP-01-UG, User's Guide, Attachment 7, Heat Capacity Temperature Limit 4. 0EOP-01-UG, User's Guide, Attachments 19 (RPV Saturation Limit), 22 Shutdown Range Level Instrument (N 027A, B) Caution), and 31 (RPV Level Caution, pages 1 & 2) 5. 0EOP-01-UG, User's Guide, Attachment 26, Unit 2 RPV Level at LL4

6. 0OI-01-07, Notifications, Attachment 1, Reportability Evaluation Checklist

7. NUREG 1022, Event Report Guidelines, Table 1, Reportable Events

8. 1 OP-27, Attachment 2, Estimated Capability Curves

9. ODCM 7.3.10, Gaseous Radwaste Treatment System

10. Tech Spec 3.2.1

, Average Planar Linear Heat Generation Rate

11. Tech Spec 3.2.2, Minimum Critical Power Ratio

12. Tech Spec 3.2.3, Linear Heat Generation Rate

13. Tech Spec 3.3.1.1, Reactor Protection System (RPS) Instrumentation

14. Tech Spec 3.3.5.1, Emergency Core Cooling System (ECCS)

Instrumentation

15. Tech Spec 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

16. Tech Spec 3.8.1, AC Sources

- Operating 17. 1-FP-05887, Auto Depressurization System Elementary Diagram Unit

18. 1 0PEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions

0EOP-01-NL Rev. 27 Page 158 of 258 ATTACHMENT 3 Page 73 of 87 Containment Parameters Secondary Containment Area Temperature Limits Table 3-B PLANT AREA PLANT LOCATION DESCRIPTION MAX NORM OPERATING VALUE (°F) MAX SAFE OPERATING VALUE (°F) AUTO GROUP ISOLATION N CORE SPRAY N CORE SPRAY ROOM 120 175 N/A S CORE SPRAY S CORE SPRAY ROOM 120 175 N/A RWCU PMP ROOM A PMP ROOM B HX ROOM 140 225 3 N RHR N RHR EQUIP ROOM 175 295 N/A S RHR S RHR EQUIP ROOM RCIC EQUIP ROOM 175 165 295 295 N/A 5 HPCI HPCI EQUIP ROOM 165 165 4 STEAM TUNNEL RCIC STM TUNNEL HPCI STM TUNNEL 190 190 295 295 5 4 20 FT 2O FT NORTH 20 FT SOUTH 140 140 200 200 N/A N/A 50 FT 50 FT NW 50 FT SE 140 140 200 200 N/A N/A REACTOR BLDG MULTIPLE AREAS ANNUN.

A-02 5-7 ALARM SETPOINT N/A 3, 4, AND/OR 5 REACTOR BLDG MSIV PIT ANNUN.

A-06 6-7 ALARM SETPOINT N/A 1 PLANT MONITORING (PAS) 0EOP-01-SBO-01 Rev. 0 Page 16 of 18 ATTACHMENT 4 Page 1 of 1 Drywell Temperature Calculation Using RSDP Recorder Inputs Values obtained from Recorder CAC

-TR-778 Above 70' Elevation PT 1 x 0.141 =

F Between 28

' and 45' Elevation PT 3 x 0.404 =

F Between 10

' and 23' Elevation PT 4 x 0.455 =

F Average Drywell Temperature F (Sum of 3 Regional Weighted Areas)

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 74 of 156 ATTACHMENT 7 Page 1 of 1 << Heat Capacity Temperature Limit

>> Torus water temperature is determined by:

CAC-TR-4426-1A, Point Wtr Avg OR CAC-TR-4426-2A, Point Wtr Avg OR Computer point G050 OR Computer point G051 OR CAC-TY-4426-1 OR CAC-TY-4426-2 Select graph line immediately below torus water level as the limit.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 87 of 156 ATTACHMENT 19 Page 1 of 1 << RPV Saturation Limit

>>

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 90 of 156 ATTACHMENT 22 Page 1 of 1 << Shutdown Range Level Instrument (N027A, B) Caution

>>

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 99 of 156 ATTACHMENT 31 Page 1 of 4 << RPV Level Caution

>> Caution 1 A RPV level instrument may be used to determine RPV level only when the conditions for use specified below are satisfied for that instrument.

NOTE Reference leg area drywell temperature is determined using Attachment 18, Level Instrument Reference Leg Area Drywell Temperature Calculations, ERFIS or Instructional Aid based on Attachment

18. ...............................

If the temperature near any instrument run is in the UNSAFE region of the Attachment 19, RPV Saturation Limit, the instrument may be unreliable due to boiling in the run.

...........................................................................................................

Immediate reference leg boiling is not expected to occur for short duration excursions into the unsafe region due to heating of the drywell.

The thermal time constant associated with the mass of metal and water in the reference leg will prohibit immediate boiling of the reference leg.

Reference leg boiling is an obvious phenomenon.

Large scale oscillations of all water level instruments associated with the reference leg that is boiling will occur.

This occurrence will be obvious and readily observable by the operator.

Additionally, if the operator is not certain whether boiling has occurred, he can refer to plant history as provided on water level recorders or ERFIS

. Reference leg boiling is indicated by level oscillations without corresponding pressure oscillations.

................................

. Instrument Conditions for Use Narrow Range Level Instruments C32-LI-R606A, B, C (N004A, B, C)

C32-LPR-R608 (N004A, B)

Indicating Range 150

-210 Inches Cold Reference Leg Unit 1 Only

The indicated level is in the SAFE region of Attachment

20. Unit 2 Only

The indicated level is in the SAFE region of Attachment

21. Shutdown Range Level Instruments B21-LI-R605A, B (N027A, B)

Indicating Range 150

-550 Inches Cold Reference Leg The indicated level is in the SAFE region of Attachment

22.

To determine RPV level at the Main Steam Line Flood Level (MSL), see Attachment

30. Attachment 30 has two curves:

The upper curve is for reference leg area drywell temperature equal to or greater than 200 F. The lower curve is for reference leg area drywell temperature less than 200 F.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 100 of 156 ATTACHMENT 31 Page 2 of 4 << RPV Level Caution >>

Caution 1 (Continued)

Instrument Conditions for Use Wide Range Level Instruments B21-LI-R604A, B (N026A, B)

C32-PR-R609 (N026B)

Indicating Range 0

-210 Inches Cold Reference Leg Temperature on the Reactor Building 50' below 140F (B21-XY-5948A A2-4, B21-XY-5948B A2-4, ERFIS Computer Point B21TA102, OR B21TA103)

AND IF the reference leg area drywell temperature is in the UNSAFE region of Attachment 19, RPV Saturation Limit , THEN the indicated level is greater than 20 inches OR IF the reference leg area drywell temperature is in the SAFE region of Attachment 19, RPV Saturation Limit , THEN the indicated level is greater than 10 inches.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 94 of 156 ATTACHMENT 26 Page 1 of 1 << Unit 2 RPV Level at LL 4 (Minimum Steam Cooling RPV Level)

>> When RPV pressure is less than 60 psig, use indicated level. LL

-4 is -27.5 inches.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 26 of 46 ATTACHMENT 1 Page 1 of 8 << Reportability Evaluation Checklist

>> NOTE If the answer to the following question is YES, then Accelerated Verbal Notification to the NRC is required within 15 minutes. Reference 0AOP

-40.0, Security Events, for notification content.

15 MINUTE REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION NA Has a Hostile Action occurred?

[NRC Bulletin 2005

-02] NOTE NUREG-1022, Rev. 3 is a reference to provide additional guidance on reportability.

If the answer to any of the following questions is YES, the event is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If all answers to the following questions are NO, the event is not reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 1 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 1.1 Is the event a deviation from technical specifications as per 10 CFR 50.54(X)?

[10 CFR 50.72(b)(1)]

1.2 1.2.1 Does the event involve by

-product, source or special nuclear material possessed by the licensee that might have or threatens to cause:

Any individual's exposure to reach or exceed 25 Rems total effective dose equivalent (TEDE); 75 Rems eye dose equivalent; or 250 Rads shallow

-dose equivalent to the skin or extremities? [10 CFR 20.2202(a)(1)]

1.2.2 The release of radioactive material inside or outside of a restricted area, such that, had an individual been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the individual could have received an intake 5 times the occupational annual limit on intake? [10 CFR 20.2202(a)(2)]

1.3 ISFSI

- Does the event involve accidental criticality or loss of any special nuclear material?

[10 CFR 72.74(a)]

1.4 Does the event involve the discovery of a cyber attack that adversely impacted safety related or important

-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that compromised support systems and equipment resulting in adverse impacts to safety, security, or emergency preparedness functions within the scope of 10 CFR 73.54?

(Note 1) [10 CFR 73.77(a)(1)]

Notes: 1. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 27 of 46 ATTACHMENT 1 Page 2 of 8 << Reportability Evaluation Checklist >>

NOTE If the answer to any of the following questions is YES, the event is reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If all answers to the following questions are NO, the event is not reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 2.1 Is plant shutdown required by technical specifications being initiated?

(Note 1) [10 CFR 50.72(b)(2)(i)]

2.2 Has the event resulted in or should have resulted in an Emergency Core Cooling System (ECCS) discharge into the Reactor Coolant System as a result of a valid signal, except when the actuation resulted from and was part of a pre

-planned sequence during testing or reactor operation? [10 CFR 50.72(b)(2)(iv)(A)] 2.3 Did the event or condition result in actuation of the reactor protection system (RPS) when the reactor was critical, except when the actuation resulted from and was part of a pre-planned sequence during testing or reactor operation?[10 CFR 50.72(b)(2)(iv)(B)]

2.4 Is the event a situation, as related to the health and safety of the public or on

-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made?

(Note 2) (Note 3)

[10 CFR 50.72(b)(2)(xi)]

[10 CFR 72.75(b)(2)]

Notes: 1. Includes any Safety Limit violation (Tech Spec 2.2)

2. Such an event may include an on

-site fatality or an inadvertent release of radioactively contaminated materials.

3. The North Carolina Wildlife Commission's Sea Turtle Coordinator (NCSTC) is notified of each sea turtle recovery. A report per 10 CFR 50.72(b)(2)(xi) is required 1) when a dead turtle is recovered OR 2) when, after consultation with t he NCSTC, it is determined that an injured turtle requires rehabilitation versus release. The NRC notification is required no later than 4 hrs after consultation with the NCSTC when either of these conditions is met.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 28 of 46 ATTACHMENT 1 Page 3 of 8 << Reportability Evaluation Checklist >>

4 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 2.5 Has any licensed material been lost, stolen, or missing in an aggregate quantity equal to or greater than 1,000 times the quantity specified in 10 CFR 20 Appendix C under such circumstances that it appears that an exposure could result to persons in unrestricted areas? (Note 1) [10 CFR 20.2201(a)(i)]

2.6 ISFSI

- Departure from License Condition.

Has an action been taken in an emergency that departs from a condition or a technical specification contained in a license or certificate of compliance issued under 10 CFR 72 when the action was immediately needed to protect the public health and safety and no action consistent with license conditions or technical specifications that could provide adequate or equivalent protection was immediately apparent as per 72.32(d)?

[10 CFR 72.75(b)(1)]

2.7 Does the event involve discovery of a cyber attack that could have caused an adverse impact to safety related or important

-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that could have compromised support systems and equipment, which if compromised could have adversely impacted safety, security, or emergency preparedness functions within the scope of 10 CFR -73.54? (Note 2) [10 CFR 73.77(a)(2)(i)]

2.8 Does the event involve discovery of a suspected or actual cyber attack initiated by personnel with physical or electronic access to digital computer and communication systems and networks within the scope of 10 CFR 73.54?

(Note 2) [10 CFR 73.77(a)(2)(ii)]

2.9 Does the event involve notification of a local, State, or other Federal agency (e.g., law enforcement, FBI, etc.) of an event related to implementation of the cyber security program for digital computer and communication systems and networks within the scope of 10 CFR 73.54 that does not otherwise require a notification under paragraph (a) of this section?

(Note 2) [10 CFR 73.77(a)(2)(iii)]

Notes: 1. Further information is located in AD

-SY-ALL-0150, Reporting Safeguards, Security, and Fitness for Duty Events.

2. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 29 of 46 ATTACHMENT 1 Page 4 of 8 << Reportability Evaluation Checklist >>

NOTE If the answer to any of the following questions is YES, the event is reportable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If all the answers to the following questions are NO, the event is not reportable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.1 Has the event or condition resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degrad ed? [10 CFR 50.72(b)(3)(ii)(A)]

3.2 Has the event or condition resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety?

[10 CFR 50.72(b)(3)(ii)(B)]

3.3 3.3.1 Did the event or condition result in valid actuation of any of the systems listed below except when the actuation resulted from and is part of a pre

-planned sequence during testing or reactor operation?

(Note 1) [10 CFR 50.72(b)(3)(iv)(A)]

These systems are:

Reactor protection system (RPS) including: reactor scram and reactor trip.

[10 CFR 50.72(b)(3)(iv)(B)(1)] Notes: 1. Automatic OR Manual initia tion of the system listed is reportable. NUREG

-1022, Section 3.2.6 discussion, should be referenced for additional information.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 30 of 46 ATTACHMENT 1 Page 5 of 8 << Reportability Evaluation Checklist >>

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION

3.3.2 General

containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

Main Steam Isolation.

Main Steam Line Drain Isolation.

HPCI Steam Line Isolation.

RCIC Steam Line Isolation

. RWCU Suction Isolation.

Primary Containment Isolation.

Secondary Containment Isolation.

SGTS Actuation.

Combustible Gas Control (CAD).

[10 CFR 50.72(b)(3)(iv)(B)(2)]

3.3.3 Emergency

core cooling systems (ECCS), including:

Core Spray (CS)

High Pressure Coolant Injection (HPCI)

Low Pressure Coolant Injection (LPCI) function of the Residual Heat Removal (RHR)

Automatic Depressurization (ADS) System

[10 CFR 50.72(b)(3)(iv)(B)(4)]

3.3.4 Reactor

Core Isolation Cooling (RCIC)

[10 CFR 50.72(b)(3)(iv)(B)(5)]

3.3.5 Containment

heat removal and depressurization systems including containment spray and fan cooler systems.

RHR Suppression Pool Cooling.

Drywell Spray System Actuation.

[10 CFR 50.72(b)(3)(iv)(B)(7)]

3.3.6 Emergency

Diesel Generators (DGs)

[10 CFR 50.72(b)(3)(iv)(B)(8)]

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 31 of 46 ATTACHMENT 1 Page 6 of 8 << Reportability Evaluation Checklist >>

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.4 3.4.1 Could the event or condition at the time of discovery have prevented the fulfillment of the safety function of structures or systems that are needed to:

[10 CFR 50.72(b)(3)(v)]

These criteria cover an event or condition in which scoped in SSCs could have failed to perform their intended function because of one or more personnel errors, including procedure violations; equipment failures; inadequate maintenance; or design, analysis, fabrication, equipment qualification, construction, or procedural deficiencies and no redundant equipment in the same system was OPERABLE. However, individual component failures need not be reported if redundant equipment in the same system was OPERABLE and available to perform the required safety function. (Note 1) [10 CFR 50.72(b)(3)(vi)]

Shut down the reactor and maintain it in a safe shutdown condition?

(Note 1) [10 CFR 50.72(b)(3)(v)(A)]

3.4.2 Remove

residual heat?

[10 CFR 50.72(b)(3)(v)(B)]

3.4.3 Control

the release of radioactive material?

[10 CFR 50.72(b)(3)(v)(C)]

3.4.4 Mitigate

the consequences of an accident?

(Note 2) [10 CFR 50.72(b)(3)(v)(D)]

3.5 Does the event require the transport of a radioactively contaminated person to an off-site medical facility for treatment?

[10 CFR 50.72(b)(3)(xii)]

[10 CFR 72.75(c)(3)]

Notes: 1. No Event Notification

[i.e., per 10 CFR 50.72(b)(3)(v)

] is required for conditions which could have prevented fulfillment of the saf ety function that are discovered when the affected system is INOPERABLE or when the affected system is INOPERABLE but considered available. If the condition is discovered when the system is OPERABLE, an EN will be made per 10 CFR 50.72(b)(3)(v).

2. RCIC INOPERABILITY is not reportable as a single train system per 10 CFR 50.72(b)(3)(v)(d). TS Basis 3.5.3 states that the RCIC System is not an ESF system and no credit is taken in the safety analysis for RCIC System operation. As such, consistent with Example 2 on NUREG 1022, Revision 3, RCIC Failure is not reportable under 10 CFR 50.72(b)(3)(v)(d).

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 32 of 46 ATTACHMENT 1 Page 7 of 8 << Reportability Evaluation Checklist >>

NOTE Additional reportability guidance concerning loss of emergency preparedness capabilities is contained in Section 3.2.13 of NUREG

-1022, Rev. 3.

Consultation with an Emergency Preparedness representative is advised when assessing the significance of the loss of capability.

0PLP-37, Equipment Important to Emergency Preparedness and ERO Response, is a reference for assistance in determining equipment important to Emergency Preparedness and whether planned or unplanned OPERABILITY of the equipment may be reportable.

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.6 Has the event resulted in a major loss of emergency assessment capability, off

-site response capability, or communications capability (i.e., significant portion of the Main Control Room indication, emergency notification system, or off

-site notification system)? [10 CFR 50.72(b)(3)(xiii)]

Major loss of emergency or off

-site notification system is considered to be/but not limited to:

a. Loss of: 1) DEMNET; OR NRC Emergency Notification System (ENS);

AND 2) Commercial telephone network.

b. INOPERABILITY for greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of: 1) Seven or more off

-site sirens; OR 2) All off-site sirens in one county.

3.7 ISFSI

-Important to Safety Defect Has a defect been discovered in any Independent Spent Fuel Storage structure, system, or component that is important to safety?

[10 CFR 72.75(c)(1)]

3.8 ISFSI

- Reduction in Effectiveness Has a condition been discovered which results in a significant reduction in the effectiveness of any Independent Spent Fuel Storage cask confinement system during use?

[10 CFR 72.75(c)(2)]

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 33 of 46 ATTACHMENT 1 Page 8 of 8 << Reportability Evaluation Checklist >>

NOTE Additional cyber security event reportability guidance is contained in NEI 15-09, Revision 0. Consultation with Nuclear Information Technology and Licensing is advised when assessing the significance of cyber security events.

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.9 Does the event involve receipt or collection of information regarding observed behavior, activities, or statements that may indicate intelligence gathering or pre-operational planning related to a cyber attack against digital computer and communication systems and networks within the scope of 10 CFR 73.54?

(Note 1) [10 CFR 73.77(a)(3)]

NOTE If the answer to any of the following questions is YES, the event is reportable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

24 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 4.1 4.1.1 Does the incident involve the loss of control of licensed material possessed by BNP which might have caused or threatens to cause:

Any individual's exposure in a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to exceed: 5 Rems total effective dose equivalent (TEDE); or 15 Rems eye dose equivalent; or 50 Rems shallow-dose equivalent to the skin or extremities?

[10 CFR 20.2202(b)(1)]

4.1.2 The release of radioactive material inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the individual could have received an intake in excess of one occupational annual limit on intake?

[10 CFR 20.2202(b)(2)]

4.2 ISFSI

- Equipment Important to Safety Disabled or Failed to Function Does the event involve equipment important to safety which is disabled or fails to function as designed when:

The equipment is required by certificate of compliance to be available and

OPERABLE to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident; and, No redundant equipment was available and OPERABLE to perform the required safety function. [10 CFR 72.75(d)(1)]

Notes: 1. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

Event Report Guidelines 10 CFR 50.72 and 50.73 Final Report

Manu script Completed: January 2013 Date Published: January 2013

Prepared by

Aron Lewin

Office of Nuclear Reactor Regulation NUREG-1022, Rev. 3

vii EXECUTIVE

SUMMARY

Two of the many elements contributing to the safety of nuclear power are emergency response and the feedback from operating experience into plant operations. These are achieved partly by the licensee event reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors

," and 10 CFR 50.73, "Licensee Event Report System

." In 10 CFR 50.72, the U.S.

Nuclear Regulatory Commission (NRC) provides for immediate notification requirements through the emergency notification system

, and in 10 CFR 50.73 provides for 60

-day written licensee event reports. The NRC staff uses the information reported under 10 CFR 50.72 and 10 CFR 50.73 in responding to emergencies, monitoring ongoing events, confirming licensing bases, studying potentially generic safety problems, assessing trends and patterns of operational experience, monitoring performance, identifying precursors of more significant events, and providing operational experience to the industry.

NUREG-1022 contains guidelines that the staff of the NRC considers acceptable for use in meeting the requirements of 10 CFR 50.72 and 10 CFR 50.73. Several identified reporting issues could not be quickly resolved given certain ambiguities in the guidance in Revision 2 of NUREG-1022. In developing Revision 3 to NUREG

-1022, the NRC held numerous public and internal meetings to solicit stakeholder input and feedback.

In resolving the ambiguities, the NRC consider ed the provisions of the rule itself, the associated statements of consideration, and other available guidance in that hierarchal order

. Revision 3 to NUREG

-1022 revises the event reporting guidelines to provide clearer guidance.

3 Table 1 Reportable Events Declaration of an Emergency Class (See Section 3.1.1 of this report)

§ 50.72(a)(1)(i) "The declaration of any of the Emergency Classes specified in the licensee

's approved Emergency Plan." Plant Shutdown Required by Technical Specifications (See Section 3.2.1 of this report)

§ 50.72(b)(2)(i)

"The initiation of any nuclear plant shutdown required by the plant

's Technical Specifications."

§ 50.73(a)(2)(i)(A)

"The completion of any nuclear plant shutdown required by the plant

's Technical Specifications."

Operation or Condition Prohibited by Technical Specifications (See Section 3.2.2 of this report)

§ 50.73(a)(2)(i)(B)

"Any operation or condition which was prohibited by the plant

's Technical Specifications except when:

(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event." Deviation from Technical Specifications Authorized under § 50.54(x) (See Section 3.2.3 of this report)

§ 50.72(b)(1)

"... any deviation from the plant

's Technical Specifications authorized pursuant to

§ 50.54(x) of this part."

§ 50.73(a)(2)(i)(C)

"Any deviation from the plant

's Technical Specifications authorized pursuant to

§ 50.54(x) of this part."

Degraded or Unanalyzed Condition (See Section 3.2.4 of this report)

§ 50.72(b)(3)(ii)

"Any event or condition that 50.73(a)(2)(ii)

"Any event or condition that

4 Table 1 Reportable Events (continued) results in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety."

resulted in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety."

External Threat or Hampering (See Section 3.2.5 of this report)

§ 50.73(a)(2)(iii)

"Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

" System Actuation (See Section 3.2.6 of this report)

§ 50.72(b)(2)(iv)(A)

"Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre

-planned sequence during testing or reactor operation."

§ 50.72(b)(2)(iv)(B)

"Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation."

§ 50.72(b)(3)(iv)(A)

"Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre

-planned sequence during testing or reactor operation."

§ 50.73(a)(2)(iv)(A)

"Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragr aph (a)(2)(iv)(B) of this section, except when:

(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or

5 Table 1 Reportable Events (continued)

§ 50.72(b)(3)(iv)(B)

"The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are:

(1) Reactor protection system (RPS) including: reactor scram and reactor trip.

5 (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including:

high-head, intermediate

-head, and low

-head injection systems and the low pressure injection function of residual (decay) heat removal systems. (4) ECCS for boiling water reactors (BWRs) including:

high-pressure and low

-pressure core spray systems; high

-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system. (7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

__________

5 Actuation of the RPS when the reactor is critical is reportable under § 50.72(b)(2)(iv)(B).

(2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed."

§ 50.73(a)(2)(iv)(B)

"The systems to which the requirements of paragraph (a)(2)(iv)(A) of this section apply are:

(1) Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including:

high-head, intermediate

-head, and low

-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs) including:

high-pressure and low

-pressure core spray systems; high

-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system. (7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

6 Table 1 Reportable Events (continued)

(8) Emergency ac electrical power systems, including:

emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

" (8) Emergency ac electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs. (9) Emergency service water systems that do not normally run and that serve as ultimate heat sinks." Event or Condition that Could Have Prevented Fulfillment of a Safety Function (See Section 3.2.7 of this report)

§ 50.72(b)(3)(v)

"Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident."

§ 50.72(b)(3)(vi)

"Events covered in paragraph (b)(3)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies.

However, individual component failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function."

§ 50.73(a)(2)(v)

"Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident."

§ 50.73(a)(2)(vi)

"Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies.

However, individual component failures need not be reported pursuant to paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function."

Common Cause Inoperability of Independent Trains or Channels (See Section 3.2.8 of this report)

§ 50.73(a)(2)(vii)

"Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition;

7 Table 1 Reportable Events (continued)

(B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident."

Radioactive Release (See Section 3.2.9 of this report)

§ 50.73(a)(2)(viii)(A)

"Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in append ix B to part 20, table 2, column 1." § 50.73(a)(2)(viii)(B)

"Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases." Internal Threat or Hampering (See Section 3.2.10 of this report)

§ 50.73(a)(2)(x)

"Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases."

Transport of a Contaminated Person Offsite (See Section 3.2.11 of this report)

§ 50.72(b)(3)(xii)

"Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment."

News Release or Notification of Other Government Agency (See Section 3.2.12 of this report)

§ 50.72(b)(2)(xi)

"Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for

8 Table 1 Reportable Events (continued) which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials."

Loss of Emergency Preparedness Capabilities (See Section 3.2.13 of this report)

§ 50.72(b)(3)(xiii)

"Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system)." Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems (See Section 3.2.14 of this report)

§ 50.73(a)(2)(ix)(A)

"Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident."

§ 50.73(a)(2)(ix)(B)

"Events covered in paragraph (ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy.

However, licensees are not required to report an event pursuant to paragraph (ix)(A) of this section if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation."

GENERATOR AND EXCITER SYSTEM OPERATING PROCEDURE 1OP-27 Rev. 63 Page 60 of 70 ATTACHMENT 2 Page 1 of 1 << Estimated Capability Curves

>>

GASEOUS RADWASTE TREATMENT SYSTEM 7.3.10 Brunswick Units 1 and 2 7.3.10-1 Rev. 25 7.3.10 GASEOUS RADWASTE TREATMENT SYSTEM ODCMS 7.3.10 The GASEOUS RADWASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY:

Whenever the Main Condenser Air Ejector (evacuation) System is in operation.

COMPENSATORY MEASURES CONDITION REQUIRED COMPENSATORY MEASURE COMPLETION TIME A. GASEOUS RADWASTE TREATMENT SYSTEM not in operation.

A.1 Place GASEOUS RADWASTE TREATMENT SYSTEM in operation.

7 days B. NOTE Required Compensatory Measure B.1 shall be completed if this Condition is entered. Required Compensatory measure and associated Completion Time not met.

B.1 Submit a Special Report to the NRC that identifies the required inoperable equipment and the reasons for th e inoperability, corrective actions taken to restore the required inoperable equipment to OPERABLE status, and a summary description of the corrective actions taken to prevent recurrence.

30 days TEST REQUIREMENTS TEST FREQUENCY TR 7.3.10.1 Verify GASEOUS RADWASTE TREATMENT SYSTEM in operation by checking the readings of the relevant instruments.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months APLHGR 3.2.1 Brunswick Unit 2 3.2-1 Amendment No. 247 3.2 POWER DISTRIBUTION LIMITS

3.2.1 AVERAGE

PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR. APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIMEA. Any APLHGR not within limits. A.1 Restore APLHGR(s) to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion Time

not met. B.1 Reduce THERMAL POWER to < 23% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits specified in the COLR.

Once within

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter MCPR 3.2.2 Brunswick Unit 2 3.2-2 Amendment No. 247

3.2 POWER

DISTRIBUTION LIMITS

3.2.2 MINIMUM

CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR. APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIMEA. Any MCPR not within limits. A.1 Restore MCPR(s) to within limits. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion

Time not met. B.1 Reduce THERMAL POWER to < 23% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the limits specified in the COLR.

Once within

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

LHGR 3.2.3 Brunswick Unit 2 3.2-4 Amendment No. 274 3.2 POWER DISTRIBUTION LIMITS

3.2.3 LINEAR

HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within limits. A.1 Restore LHGR(s) to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion

Time not met. B.1 Reduce THERMAL POWER to < 23% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits specified in the COLR.

Once within

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-1 Amendment No. 243 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS -----------------------------------------------------------NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIMEA. One or more required channels inoperable. A.1 Place channel in trip.

OR A.2 ---------------NOTE-------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Place associated trip system in trip.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 12 hours (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-2 Amendment No. 247 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE----------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

One or more Functions with one or more required

channels inoperable in both trip systems. B.1 Place channel in one trip system in trip.

OR B.2 Place one trip system in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months C. One or more Functions with RPS trip capability not maintained. C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. Required Action and associated Completion Time

of Condition A, B, or C not

met. D.1 Enter the Condition referenced in

Table 3.3.1.1-1 for the

channel. Immediately E. As required by Required Action D.1 and referenced in

Table 3.3.1.1-1. E.1 Reduce THERMAL POWER to < 26% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-3 Amendment No. 243 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMEF. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. F.1 Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months G. As required by Required Action D.1 and referenced in

Table 3.3.1.1-1. G.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months H. As required by Required Action D.1 and referenced in

Table 3.3.1.1-1. H.1 Initiate action to fully insert all insertable control rods in

core cells containing one or more fuel assemblies.

Immediately I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. I.1 Initiate alternate method to detect and suppress

thermal hydraulic instability

oscillations.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months J. Required Action and associated Completion Time

of Condition I not met. J.1 Reduce THERMAL POWER to < 20% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-4 Amendment No. 247 SURVEILLANCE REQUIREMENTS ----------------------------------------------------------NOTES---------------------------------------------------------- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function. 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.)

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2 Perform CHANNEL CHECK.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.1.3 --------------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

THERMAL POWER 23% RTP.

Adjust the average power range monitor (APRM) channels to conform to the calculated power while

operating at 23% RTP.

7 days SR 3.3.1.1.4 --------------------------------NOTE--------------------------------

Not required to be performed when entering MODE 2

from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

7 days (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-5 Amendment No. 282 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram contactor.

7 days SR 3.3.1.1.6 Verify the source range monitor (SRM) and intermediate range monitor (IRM) channels overlap.

Prior to withdrawing SRMs from the fully

inserted position SR 3.3.1.1.7 --------------------------------NOTE--------------------------------

Only required to be met during entry into MODE 2 from

MODE 1.

Verify the IRM and APRM channels overlap.

7 days SR 3.3.1.1.8 Calibrate the local power range monitors.

2000 effective full power hours SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.1.1.10 Calibrate the trip units.

92 days (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-6 Amendment No. 243 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 --------------------------------NOTES------------------------------ 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2. 2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow

transmitters.

Perform CHANNEL FUNCTIONAL TEST.

184 days SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.

24 months SR 3.3.1.1.13 --------------------------------NOTES------------------------------ 1. Neutron detectors are excluded. 2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2. 3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION.

24 months SR 3.3.1.1.14 (Not used.)

SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-7 Amendment No. 247 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL

POWER is 26% RTP.

24 months SR 3.3.1.1.17 --------------------------------NOTES---------------------------- 1. Neutron detectors are excluded. 2. For Functions 3 and 4, the sensor response time may be assumed to be the design sensor response time. 3. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED

TEST BASIS Frequency. 4. For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and Oscillation Power Range Monitor (OPRM) outputs shall alternate.

Verify the RPS RESPONSE TIME is within limits.

24 months on a

STAGGERED TEST

BASIS SR 3.3.1.1.18 Adjust the flow control trip reference card to conform to reactor flow.

Once within 7 days

after reaching

equilibrium

conditions following

refueling outage (continued)

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-8 Amendment No. 243 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated Thermal Power is 25%and recirculation drive flow is 60%. 24 months

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-9 Amendment No. 247 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 1. Intermediate Range Monitors

a. Neutron Flux-High

2

3

G

SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.15

5 (a) 3 H SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.13 SR 3.3.1.1.15

b. Inop 2

5 (a) 3

3 G

H SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.15

SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.15

2. Average Power Range Monitors

a. Neutron Flux-High (Setdown)

2 3(c)

G

SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13

b. Simulated Thermal Power-High 1 3(c) F SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 (b)

(continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) [0.55 (W - W) + 62.6% RTP] when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." The value of W is defined in plant procedures. (c) Each APRM channel provides inputs to both trip systems.

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-10 Amendment No. 243 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2. Average Power Range Monitors (continued)

c. Neutron Flux-High

1

3(c)

F

SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13

d. Inop 1,2 3(c) G SR 3.3.1.1.5 SR 3.3.1.1.11

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17

f. OPRM Upscale 20% RTP 3(c) I SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19

3. Reactor Vessel Steam Dome Pressure-High 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

4. Reactor Vessel Water Level-Low Level 1 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

5. Main Steam Isolation Valve-Closure 1 8 F SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

6. Drywell Pressure-High 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15

(continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM period based detection algorithm (PBDA) setpoint limits.

RPS Instrumentation 3.3.1.1 Brunswick Unit 2 3.3-11 Amendment No. 247 Table 3.3.1.1-1 (page 3 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 7. Scram Discharge Volume Water Level-High 1,2 2 G SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15

5 (a) 2 H SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15

8. Turbine Stop Valve-Closure 26% RTP 4 E SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17

9. Turbine Control Valve Fast Closure, Control Oil Pressure-Low 26% RTP 2 E SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17

10. Reactor Mode Switch- Shutdown Position 1,2 1 G SR 3.3.1.1.12 SR 3.3.1.1.15

5 (a) 1 H SR 3.3.1.1.12 SR 3.3.1.1.15

11. Manual Scram 1,2 1 G SR 3.3.1.1.9 SR 3.3.1.1.15

5 (a) 1 H SR 3.3.1.1.9 SR 3.3.1.1.15

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-35 Amendment No. 233

3.3 INSTRUMENTATION

3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.5.1-1.

ACTIONS -----------------------------------------------------------NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIMEA. One or more channels inoperable. A.1 Enter the Condition referenced in

Table 3.3.5.1-1 for the

channel. Immediately B. As required by Required Action A.1 and referenced in

Table 3.3.5.1-1.

B.1 -------------NOTES------------- 1. Only applicable in MODES 1, 2, and 3. 2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.

Declare supported feature(s) inoperable when

its redundant feature ECCS

initiation capability is

inoperable.

AND

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

initiation capability

for feature(s) in both

divisions

(continued)

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-36 Amendment No. 233

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 --------------NOTE--------------

Only applicable for

Functions 3.a and 3.b.

Declare High Pressure Coolant Injection (HPCI)

System inoperable.

AND B.3 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

HPCI initiation

capability

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. As required by Required Action A.1 and referenced in

Table 3.3.5.1-1.

C.1 ------------NOTES-------------- 1. Only applicable in MODES 1, 2, and 3. 2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.

Declare supported feature(s) inoperable when

its redundant feature ECCS

initiation capability is

inoperable.

AND C.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

initiation capability

for feature(s) in both

divisions

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued)

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-37 Amendment No. 233

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMED. As required by Required Action A.1 and referenced in

Table 3.3.5.1-1 D.1 --------------NOTE--------------

Only applicable if HPCI

pump suction is not aligned

to the suppression pool.

Declare HPCI System inoperable.

AND D.2.1 Place channel in trip.

OR D.2.2 Align the HPCI pump suction to the suppression

pool.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

HPCI initiation

capability

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued)

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-38 Amendment No. 233

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMEE. As required by Required Action A.1 and referenced in

Table 3.3.5.1-1. E.1 Declare Automatic Depressurization System (ADS) valves inoperable.

AND E.2 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

ADS initiation

capability in both trip systems

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from

discovery of

inoperable channel

concurrent with

HPCI or reactor core

isolation cooling (RCIC) inoperable

AND 8 days (continued)

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-39 Amendment No. 233

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMEF. As required by Required Action A.1 and referenced in

Table 3.3.5.1-1. F.1 Declare ADS valves inoperable.

AND F.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of

ADS initiation

capability in both trip systems

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from

discovery of

inoperable channel

concurrent with

HPCI or RCIC

inoperable

AND 8 days G. Required Action and associated Completion Time

of Condition B, C, D, E, or F

not met. G.1 Declare associated supported feature(s)

inoperable.

Immediately ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-40 Amendment No. 233

SURVEILLANCE REQUIREMENTS ----------------------------------------------------------NOTES---------------------------------------------------------- 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as

follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Function 3.c; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other

than 3.c provided the associated Function or the redundant Function maintains ECCS

initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.5.1.3 Calibrate the trip unit.

92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION.

24 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.5.1.6 Perform CHANNEL FUNCTIONAL TEST.

24 months ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-41 Amendment No. 233 Table 3.3.5.1-1 (page 1 of 4) Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

a. Reactor Vessel Water Level-Low Level 3

1,2,3, 4 (a), 5 (a)

4

B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

13 inches b. Drywell Pressure-High 1,2,3 4 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

1.8 psig c. Reactor Steam Dome Pressure-Low 1,2,3 4 C SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

402 psig and 425 psig 4 (a), 5 (a) 4 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

402 psig and 425 psig d. Core Spray Pump Start-Time Delay Relay 1,2,3, 4 (a), 5 (a) 2 1 per pump C SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

14 seconds and 16 seconds 2. Low Pressure Coolant Injection (LPCI) System a. Reactor Vessel Water Level-Low Level 3

1,2,3, 4 (a), 5 (a)

4

B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

13 inches

b. Drywell Pressure-High 1,2,3 4 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 1.8 psig

(continued) (a) When associated subsystem(s) are required to be OPERABLE.

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-42 Amendment No. 233 Table 3.3.5.1-1 (page 2 of 4) Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

2. LPCI System (continued)

c. Reactor Steam Dome Pressure-Low

1,2,3

4

C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

402 psig and 425 psig 4 (a), 5 (a) 4 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

402 psig and 425 psig d. Reactor Steam Dome Pressure-Low (Recirculation Pump Discharge Valve Permissive) 1 (b),2 (b), 3 (b) 4 C SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

302 psig e. Reactor Vessel Shroud Level 1,2,3 2 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

-50 inches f. RHR Pump Start-Time Delay Relay 1,2,3, 4 (a), 5 (a) 4 1 per pump C SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

9 seconds and 11 seconds 3. High Pressure Coolant Injection (HPCI) System a. Reactor Vessel Water Level-Low Level 2

1, 2(c), 3(c)

4

B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

101 inches b. Drywell Pressure-High 1, 2(c),3(c) 4 B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 1.8 psig

(continued) (a) When associated subsystem(s) are required to be OPERABLE.

(b) With associated recirculation pump discharge valve or recirculation pump discharge bypass valve open.

(c) With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-43 Amendment No. 233 Table 3.3.5.1-1 (page 3 of 4) Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

3. HPCI System (continued)

c. Reactor Vessel Water Level-High

1, 2(c), 3(c)

2

C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

207 inches d. Condensate Storage Tank Level-Low 1, 2(c), 3(c) 2 D SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

23 feet 4 inches

e. Suppression Chamber Water Level-High 1, 2(c), 3(c) 2 D SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

-2 feet 4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water Level-Low Level 3

1, 2(c), 3(c)

2

E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

13 inches b. ADS Timer 1, 2(c), 3(c) 1 F SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

108 seconds c. Reactor Vessel Water Level-Low Level 1 1, 2(c), 3(c) 1 E SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

153 inches d. Core Spray Pump Discharge Pressure-High 1, 2(c), 3(c) 2 F SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

102 psig and 130 psig e. RHR (LPCI Mode) Pump Discharge Pressure-High 1, 2(c), 3(c) 4 2 per pump F SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 102 psig and 130 psig (continued)

(c) With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Brunswick Unit 2 3.3-44 Amendment No. 233 Table 3.3.5.1-1 (page 4 of 4) Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5. ADS Trip System B

a. Reactor Vessel Water Level-Low Level 3

1, 2(c), 3(c)

2

E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

13 inches b. ADS Timer 1, 2(c), 3(c) 1 F SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

108 seconds c. Reactor Vessel Water Level-Low Level 1 1, 2(c), 3(c) 1 E SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

153 inches d. Core Spray Pump Discharge Pressure-High 1, 2(c), 3(c) 2 F SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

102 psig and 130 psig e. RHR (LPCI Mode) Pump Discharge Pressure-High 1, 2(c), 3(c) 4 2 per pump F SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

102 psig and 130 psig (c) With reactor steam dome pressure > 150 psig.

PCIVs 3.6.1.3 Brunswick Unit 2 3.6-7 Amendment No. 233

3.6 CONTAINMENT

SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." ACTIONS ----------------------------------------------------------NOTES---------------------------------------------------------- 1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Acti ons for systems made inoperable by PCIVs. 4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance

criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. --------------NOTE-------------

Only applicable to

penetration flow paths with

two PCIVs.

One or more penetration flow paths with one PCIV

inoperable except for MSIV

leakage not within limit. A.1 Isolate the affected penetration flow path by

use of at least one closed

and de-activated automatic

valve, closed manual valve, blind flange, or check valve

with flow through the valve

secured. AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

(continued)

PCIVs 3.6.1.3 Brunswick Unit 2 3.6-8 Amendment No. 233

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------------NOTE--------------

Isolation devices in high

radiation areas may be

verified by use of

administrative means.

Verify the affected penetration flow path is

isolated.

Once per 31 days

for isolation devices

outside primary

containment

AND Prior to entering

MODE 2 or 3 from

MODE 4, if primary

containment was

de-inerted while in

MODE 4, if not

performed within the

previous 92 days, for

isolation devices

inside primary

containment (continued)

PCIVs 3.6.1.3 Brunswick Unit 2 3.6-9 Amendment No. 233

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTE--------------

Only applicable to

penetration flow paths with

two PCIVs.

One or more penetration flow paths with two PCIVs

inoperable except for MSIV

leakage not within limit. B.1 Isolate the affected penetration flow path by

use of at least one closed

and de-activated automatic

valve, closed manual valve, or blind flange.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C. --------------NOTE--------------

Only applicable to

penetration flow paths with

only one PCIV.

One or more penetration flow paths with one PCIV

inoperable. C.1 Isolate the affected penetration flow path by

use of at least one closed

and de-activated automatic

valve, closed manual valve, or blind flange.

AND C.2 --------------NOTE--------------

Isolation devices in high

radiation areas may be

verified by use of

administrative means.

Verify the affected penetration flow path is

isolated.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months except for excess flow check

valves (EFCVs)

AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for EFCVs

Once per 31 days (continued)

PCIVs 3.6.1.3 Brunswick Unit 2 3.6-10 Amendment No. 233

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMED. One or more penetration flow paths with one or more

MSIVs not within MSIV

leakage rate limits. D.1 Restore leakage rate to within limit.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months E. Required Action and associated Completion Time

of Condition A, B, C, or D

not met in MODE 1, 2, or 3. E.1 Be in MODE 3.

AND E.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Required Action and associated Completion Time

of Condition A, B, C, or D

not met for PCIV(s) required

to be OPERABLE during

MODE 4 or 5. F.1 Initiate action to suspend operations with a potential

for draining the reactor vessel (OPDRVs).

OR F.2 Initiate action to restore valve(s) to OPERABLE

status. Immediately

Immediately

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-2 Amendment No. 205 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIMEB. --------------NOTES--------------1. Only applicable when Unit 2 is in MODE 4 or 5. 2. Condition B shall not be entered in conjunction with Condition A.

Two Unit 2 offsite circuits inoperable due to one Unit 2

balance of plant circuit path

to the downstream 4.16 kV

emergency bus inoperable

for planned maintenance.

AND DG associated with the affected downstream

4.16 kV emergency bus

inoperable for planned

maintenance. B.1 Declare required feature(s) with no power available

inoperable when the

redundant required

feature(s) are inoperable.

AND B.2 Perform SR 3.8.1.1 for OPERABLE offsite circuit(s).

AND B.3 Restore both Unit 2 offsite circuits and DG to OPERABLE status.

Immediately from

discovery of

Condition B

concurrent with

inoperability of

redundant required

feature(s)

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter

7 days AND 10 days from discovery of failure

to meet LCO 3.8.1.a

or b (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-3 Amendment No. 264 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIMEC. One offsite circuit inoperable for reasons other than Condition A or B. C.1 Perform SR 3.8.1.1 for OPERABLE offsite

circuit(s).

AND C.2 Declare required feature(s) with no offsite power available inoperable when

the redundant required

feature(s) are inoperable.

AND C.3 Restore offsite circuit to OPERABLE status.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from

discovery of no

offsite power to one

4.16 kV emergency

bus concurrent with

inoperability of

redundant required

feature(s)

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 17 days from discovery of failure

to meet LCO 3.8.1.a

or b (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-4 Amendment No. 264 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIMED. One DG inoperable for reasons other than Condition B. D.1 Perform SR 3.8.1.1 for OPERABLE offsite

circuit(s).

AND D.2 Evaluate availability of supplemental diesel generator (SUPP-DG)

AND D.3 Declare required feature (s), supported by the inoperable DG, inoperable

when the redundant

required feature (s) are

inoperable.

AND D.4.1 Determine OPERABLE DG(s) are not inoperable

due to common cause

failure. OR D.4.2 Perform SR 3.8.1.2 for OPERABLE DG(s).

AND 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of

Condition D

concurrent with

inoperability of

redundant required

feature (s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-5 Amendment No. 264 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIMED. (continued) D.5 Restore DG to OPERABLE status.

7 days from discovery of

unavailability of

SUPP-DG AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of

Condition D entry 6 days concurrent with unavailability of

SUPP-DG AND 14 days AND 17 days from discovery of failure

to meet LCO 3.8.1.a

or b E. Two or more offsite circuits inoperable for reasons other

than Condition B. E.1 Declare required feature(s) inoperable when the

redundant required

feature(s) are inoperable.

AND E.2 Restore all but one offsite circuit to OPERABLE status. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from

discovery of

Condition E

concurrent with

inoperability of

redundant required

feature(s)

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-6 Amendment No. 264 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIMEF. One offsite circuit inoperable for reasons other than Condition B.

AND One DG inoperable for reasons other than

Condition B.

NOTE-------------------

Enter applicable Conditions and

Required Actions of LCO 3.8.7, "Distribution Systems-Operating,"

when Condition F is entered with no

AC power source to any 4.16 kV

emergency bus.

F.1 Restore offsite circuit to OPERABLE status.

OR F.2 Restore DG to OPERABLE status.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 12 hours G. Two or more DGs inoperable. G.1 Restore all but one DG to OPERABLE status.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months H. Required Action and associated Completion Time

of Condition A, B, C, D, E, F

or G not met. H.1 Be in MODE 3.

AND H.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months I. One or more offsite circuits and two or more DGs inoperable.

OR Two or more offsite circuits and one DG inoperable for

reasons other than

Condition B. I.1 Enter LCO 3.0.3. Immediately

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-7 Amendment No. 264 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.

7 days SR 3.8.1.2 -------------------------------NOTES------------------------------- 1. All DG starts may be preceded by an engine prelube period. 2. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR. When modified start procedures are

not used, the time, voltage, and frequency

tolerances of SR 3.8.1.7 must be met. 3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG starts from standby conditions and achieves steady state voltage 3750 V and 4300 V and frequency 58.8 Hz and 61.2 Hz.

31 days (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-8 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3 -------------------------------NOTES------------------------------- 1. DG loadings may include gradual loading. 2. Momentary transients outside the load range do not invalidate this test. 3. This Surveillance shall be conducted on only one DG at a time. 4. This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7. 5. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG is synchronized and loaded and operates for 60 minutes at a load 2800 kW and 3500 kW.

31 days SR 3.8.1.4 Verify each engine mounted tank contains 150 gal of fuel oil.

31 days SR 3.8.1.5 Check for and remove accumulated water from each engine mounted tank.

31 days SR 3.8.1.6 Verify the fuel oil transfer system operates to transfer fuel oil from the day fuel oil storage tank to the engine mounted tank.

31 days (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-9 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 -------------------------------NOTES------------------------------- 1. All DG starts may be preceded by an engine prelube period. 2. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG starts from standby condition and achieves, in 10 seconds, voltage 3750 V and frequency 58.8 Hz, and after steady state conditions are reached, maintains voltage 3750 V and 4300 V and frequency 58.8 Hz and 61.2 Hz.

184 days (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-10 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.8 -------------------------------NOTES------------------------------- 1. SR 3.8.1.8.a shall not be performed in MODE 1 or 2 for the Unit 1 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR. 2. SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit. 3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify:

a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsite

circuit.

24 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-11 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9 -------------------------------NOTES------------------------------- 1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR. 2. If performed with the DG synchronized with offsite power, it shall be performed at a power factor 0.9. 3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG rejects a load greater than or equal to its associated core spray pump without tripping.

24 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-12 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.10 -------------------------------NOTE---------------------------------

A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG's automatic trips are bypassed on an actual or simulated ECCS initiation signal except: a. Engine overspeed;

b. Generator differential overcurrent;

c. Low lube oil pressure;

d. Reverse power;

e. Loss of field; and

f. Phase overcurrent (voltage restrained).

24 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-13 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.11 -------------------------------NOTES------------------------------- 1. Momentary transients outside the load and power factor ranges do not invalidate this test. 2. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG operating at a power factor 0.9 operates for 60 minutes loaded to 3500 kW and 3850 kW.

24 months SR 3.8.1.12 -------------------------------NOTE---------------------------------

A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify an actual or simulated ECCS initiation signal is capable of overriding the test mode feature to return

each DG to ready-to-load operation.

24 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-14 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 -------------------------------NOTE---------------------------------

This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with

DG 1 and DG 2. However, credit may be taken for

unplanned events that satisfy this SR.

Verify interval between each sequenced load block is within +/- 10% of design interval for each load sequence

relay.

24 months (continued)

AC Sources-Operating

3.8.1 Brunswick

Unit 1 3.8-15 Amendment No. 205 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.14 -------------------------------NOTES------------------------------- 1. All DG starts may be preceded by an engine prelube period. 2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.

Verify, on actual or simulated loss of offsite power signal in conjunction with an actual or simulated ECCS

initiation signal: a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and: 1. energizes permanently connected loads in 10.5 seconds, 2. energizes auto-connected emergency loads through load sequence relays, 3. maintains steady state voltage 3750 V and 4300 V, 4. maintains steady state frequency 58.8 Hz and 61.2 Hz, and 5. supplies permanently connected and auto-connected emergency loads for 5 minutes.

24 months

@@ Line 2: / Line 2: @@
 | number = ML17017A385
 | issue date = 01/27/2017
-| title = Brunswick 2016301 Final SRO Written Exam - Delay Release 2 Years
+| title = 2016301 Final SRO Written Exam - Delay Release 2 Years
 | author name =
 | author affiliation = NRC/RGN-II/DRS/OLB

ML17017A385: Difference between revisions

Revision as of 14:58, 6 April 2019

Contents

Text

1.3 ISFSI

2.6 ISFSI

3.3.2 General

3.3.3 Emergency

3.3.4 Reactor

3.3.5 Containment

3.3.6 Emergency

3.4.2 Remove

3.4.3 Control

3.4.4 Mitigate

3.7 ISFSI

3.8 ISFSI

4.2 ISFSI

SUMMARY

3.2.1 AVERAGE

3.2 POWER

3.2.2 MINIMUM

3.2.3 LINEAR

3.3 INSTRUMENTATION

3.6 CONTAINMENT

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

Navigation menu

ML17017A385: Difference between revisions

Revision as of 14:58, 6 April 2019

Text

1.3 ISFSI

2.6 ISFSI

3.3.2 General

3.3.3 Emergency

3.3.4 Reactor

3.3.5 Containment

3.3.6 Emergency

3.4.2 Remove

3.4.3 Control

3.4.4 Mitigate

3.7 ISFSI

3.8 ISFSI

4.2 ISFSI

SUMMARY

3.2.1 AVERAGE

3.2 POWER

3.2.2 MINIMUM

3.2.3 LINEAR

3.3 INSTRUMENTATION

3.6 CONTAINMENT

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

3.8.1 Brunswick

Navigation menu

Search