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: x. LESSON BODY X. INSTRUCTOR NOTES letdown is greater than charging, the PZR pressure will drop, the vessel void will grow and the PZR level will rise. In the same way when charging is greater than letdown, the PZR pressure will rise, the vessel void will shrink, and the level will drop. If inventory shrink is used to reduce PZR level, careful attention should be paid to the Technical Specification pressure-temperature limits. | : x. LESSON BODY X. INSTRUCTOR NOTES letdown is greater than charging, the PZR pressure will drop, the vessel void will grow and the PZR level will rise. In the same way when charging is greater than letdown, the PZR pressure will rise, the vessel void will shrink, and the level will drop. If inventory shrink is used to reduce PZR level, careful attention should be paid to the Technical Specification pressure-temperature limits. | ||
: 7) MONITOR RVLIS greater than 69%. Objective 18 I | : 7) MONITOR RVLIS greater than 69%. Objective 18 I | ||
Purpose: To ensure the void does not enter Operator Fundamentals: | |||
==Purpose:== | |||
To ensure the void does not enter Operator Fundamentals: | |||
the hot legs and disrupt natural circulation. Understanding of plant design and system operation. | the hot legs and disrupt natural circulation. Understanding of plant design and system operation. | ||
Basis: If steam enters the hot legs, it would NOTE: RVLIS at 69% | Basis: If steam enters the hot legs, it would NOTE: RVLIS at 69% | ||
| Line 1,803: | Line 1,805: | ||
An uncertainty is applied to the reading to ensure that the void will actually be above the hot legs, even when assuming worst case channel inaccuracies. | An uncertainty is applied to the reading to ensure that the void will actually be above the hot legs, even when assuming worst case channel inaccuracies. | ||
: 8) DETERMINE if cold leg accumulators should be isolated. | : 8) DETERMINE if cold leg accumulators should be isolated. | ||
Purpose: To determine if appropriate plant conditions exist for locking out SI. | |||
==Purpose:== | |||
To determine if appropriate plant conditions exist for locking out SI. | |||
S1. | S1. | ||
Basis: The cold leg accumulator isolation valves should be closed and their power supplies locked out to prevent the dumping of borated water into the Res when Res pressure drops below accumulator pressure. | Basis: The cold leg accumulator isolation valves should be closed and their power supplies locked out to prevent the dumping of borated water into the Res when Res pressure drops below accumulator pressure. | ||
| Line 5,531: | Line 5,535: | ||
Page 76 | Page 76 | ||
TVA Title: | TVA | ||
==Title:== | |||
LIQUID RADWASTE PROCESSING SYSTEM N3-77C-4001 3.3.5 Passive Sump Instrumentation The passive sump located in the auxiliary building is designed to receive shield building annulus floor drains and in the event of a pipe break, receive water from the containment spray pump room and/or pipe chase on elevation 676.0. The passive sump instrumentation is designed to detect eleva'tion two types of leaks. to detect a leakage rate First, the instrumentation is to. | |||
of 50 gal/min into the passive sump from areas throughout the auxiliary building. Secondly, the level indicators are part of the postaccident monitoring system which provide sump level indication to the operator in the control room. Water in the sump is drained to the Auxiliary Building floor and equipment drain sump (ABF&EDS) via 6" drain valve DRV-77-916 bottom of the sump at elevation 665'-9" with which is located at the 'bottom mechanical linkage provided to a hand wheel on the floor above on elevation 676'-2". | of 50 gal/min into the passive sump from areas throughout the auxiliary building. Secondly, the level indicators are part of the postaccident monitoring system which provide sump level indication to the operator in the control room. Water in the sump is drained to the Auxiliary Building floor and equipment drain sump (ABF&EDS) via 6" drain valve DRV-77-916 bottom of the sump at elevation 665'-9" with which is located at the 'bottom mechanical linkage provided to a hand wheel on the floor above on elevation 676'-2". | ||
The sump level indication is provided by a qualified differential pressure transmitter with capillary tubes mounted to the sump wall. The upper capillary tap is mounted 13 inches below the top of the passive sump at elevation 672'-6.5", and the lower tap is mounted near the floor of the sump at elevation 667' -6'" | The sump level indication is provided by a qualified differential pressure transmitter with capillary tubes mounted to the sump wall. The upper capillary tap is mounted 13 inches below the top of the passive sump at elevation 672'-6.5", and the lower tap is mounted near the floor of the sump at elevation 667' -6'" | ||
Revision as of 21:25, 6 December 2019
| ML100830075 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 02/11/2010 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-390/09-302 | |
| Download: ML100830075 (341) | |
Text
{{#Wiki_filter:11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) 76. 022 AA2.03 076 . Given the following
- Unit at 100% power, steady state conditions.
- MCR Controller for 1-HIC-62-89A, CHRG-HDR-RCP SEALS FLOW MeR CONTROL, fails to "0" output.
- Operator respond by taking manual control of charging flow and report annunciators LIT directing the use of AOI-20, "Malfunctions of the Pressurizer Level Controller System" and AOI-24, "Rep "RCP Malfunctions During Pump Operation."
Which ONE of the following identifies how the controller failure will affect the the RCP seal injection flow and which AOI should receive priority for implementation Rep because it provides the directions to restore the RCP seal injection flow to normal? RCP Seal Injection Flow Procedure implementation A. Increase "RCP Malfunctions During Pump AOI-24, "Rep Operation" B. Increase AOI-20, "Malfunctions of the Pressurizer Level Controller System" C. Decrease C. "RCP Malfunctions During Pump AOI-24, "RCP Operation" D!' D'!'" Decrease AOI-20, "Malfunctions of the Pressurizer Level Controller System" Page 1
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR ANAL YSIS: A. Incorrect, If 1-HIC-62-89A failed to '0' output, valve 1-FCV-62-89 will be driven full open which would not result in an increase in RCP seal injection flow (the flow would decrease) and while AOI-24 is the procedure for RCP seal issues, it is not the procedure to address low seal injection flow. Plausible because* because the change in seal injection flow would be correct if the applicant concluded '0" output on the controller caused the fail to go closed and AOI-24 does address RCP seal issues. B. Incorrect, If 1-HIC-62-89A failed to '0' output, valve 1-FCV-62-89 will be driven full open which would not result in an increase in RCP seal injection flow (the flow would decrease) and AOI-20 is the procedure that contains the actions to restore flow Plausible because the change in seal injection flow would be the seal injection flow correct if the applicant concluded '0" output on the controller caused the fail to go closed and A AOI-20 01-20 is correct. C. Incorrect, If 1-HIC-62-89A failed to '0' output, valve 1-FCV-62-89 will be driven full open resulting lower injection flow to the RCP seals but while AOI-24 is the procedure for RCP seal issues, it is not the procedure to address low seal prooedure seal injection flow. Plausible because the change in seal injection flow is correct and A AOI-24 01-24 does address RCP seal issues. D. Correct, If 1-HIC-62-89A failed to '0' output, valve 1-FCV-62-89 will be driven full open resulting in less backpressure being applied to the charging line causing RCP deal flow to decrease but more water to be directed through the charging supply line and increased overall flow to the RCS. Pressurizer level would increase causing 1-FCV-62-93 to throttle more closed to maintain pressure level which would drop RCP seal injection flow more. AOI-20 has the steps to restore the seal flow by taking manual control of valve 1-FCV-62-89 or bypassing the valve. ) Page 2
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) Question Number: 76 Tier: _1_ Group* 1 KIA: 022 AA2.03 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Failures of flow control valve or controller. Importance Rating: 3.1 / 3.6 10 CFR Part 55: 43.5 / 45.13 10CFR55.43.b: .5 KIA Match: Applicant must determine how a failure of a valve controller will affect operation of the system and then select the appropriate procedure to be used to correct the condition. SRO because following a valve controller failure in the CVCS multiple alarms are lit and the applicant is required to determine which procedure has priority for stabilizing the plant. Technical
Reference:
AOI-20, Malfunction of the Pressurizer Level Control ) System, Rev 31 ARI-95-101, ARI-95-1 01, Reactor Coolant Pumps, Rev 31 Window 101-E ARI-102-108, HVAC & CVCS, Rev 25 Window 108-A Proposed references None to be provided: Learning Objective: 3-0T-AOI2000
- 06. Demonstrate ability to determine causes for pressurizer level Malfunctions.
Question Source: New X Modified Bank Bank Question History: New question Comments ) Page 3
101-E SOURCE SET POINT SETPOINT RCP1: RCP 1: 1-FS-62-1 6.5 6.S gpm RCP 2: 1-FS-62-14 Rep SEAL \ RCP 3: 1-FS-62-27 SUPPLY RCP 4: 1-FS-62-40 FLOW La Probable Cause: 1. Failure of FCV-62-89, Seal Water Supply Back Pressure Control valve (valve fails open)
- 2. Seal water injection filter dirty
- 3. Valve misalignment to individual RCP
- 4. Charging pumps shut down Corrective Action: [1] CHECK seal water flow to each pump [1-M-5].
[1-M-S].
- RCP 1: FI-62-1A
- RCP 2: FI-62-14A
)
- RCP 3: FI-62-27A
- RCP 4: FI-62-40A
[2] IF all seal water flows are low, THEN [a] CHECK FCV-62-89 for proper operation. [b] CHECK Window 101-0 101-D in alarm for indication of dirty filter. [3] IF only one pump's seal water flow is low, THEN [a] MONITOR pump leakoffs. [b] MONITOR pump vibrations. [4] REFER TO AOI-24, RCP MALFUNCTIONS DURING PUMP OPERATION.
References:
1-47W61 0-62-1 AOI-24 ) WBN I Page 46 of 50 I ARI-95-101 _._ _ _ _----1. Rev 29
108-A SOURCE SETPOINT 1-FS-62-93-AlB 150 gpm ) PLANT COMPUTER 47 gpm with BOTH 75 gpm orifice valves CLOSED CHARGING OR FLOW 55 gpm with EITHER 75 gpm HilLa orifice valve OPEN Probable Cause: 1. System pipe break
- 2. Charging pump tripped
- 3. Malfunction of PZR level control system.
Corrective Action: [1] IF ALL the following conditions exist:
- Any RCP Thermal Barrier Out-Of-Service,
- In-Service Charging pump trips,
- RCP seal injection flow required, THEN IMMEDIATELY START available charging pump to restore seal flow.
) [2] CHECK 1-FI-62-93A [1-M-5] to determine if flow is high or low. [3] CHECK PZR level indication on 1-M-4. [4] IF PZR level control system malfunction, THEN GO TO AOI-20, MALFUNCTION OF PRESSURIZER LEVEL CONTROL CHANNEL. [5] IF charging flow is low, THEN CHECK letdown temperature and CONSIDER increasing charging flow, or ISOLATE letdown. [6] IF charging is lost, THEN . IMMEDIATELY ISOLATE letdown. [7] DETERMINE cause of problem and INITIATE corrective action. [8] REFER TO SOI-62.01, CVCS - CHARGING AND LETDOWN.
References:
1-47W61 0-62-2 1-47W809-1 AOI-20 SOI-62.01 ) WBN Page 42 of 50 08 I ARI-1 02-1 08 1
. Rev 25 .
AOI-20 MALFUNCTION OF PRESSURIZER Revision 31 WBN LEVEL CONTROL SYSTEM PaQe 5 of 17 Page 3.0 OPERATOR ACTIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION Charging and letdown must be in service together. If letdown isolates or charging is lost, the other must be isolated.
- 1. CHECK pzr level program PERFORM the following:
signal NORMAL:
- a. PLACE charging valve controller
- 1-LR-68-339 1-HIC-62-93A in MAN, and (green pen). RESTORE level to normal.
- b. IF letdown in service, THEN
** GO TO Step 10.
- c. IF letdown NOT in service,
)) THEN
** GO TO Step 6.
NOTE 1-XS-68-339E selects one channel to control level to program and one backup channel for control interlocks.
- 2. CHECK if 1-XS-68-339E is IF pzr level is low OR selected to FAILED channel dropping, (control or backup): THEN
** GO TO Step 12.
** L1-68-339, LI-68-339, IF pzr level is high OR OR rising, THEN
** LI-68-320, ** GO TO Step 14.
OR IF non-selected channel failure,
** LI-68-335. THEN
**
** GO TO Step 8.
)
AOI-20 WBN MALFUNCTION OF PRESSURIZER Revision 31 LEVEL CONTROL SYSTEM Page PaQe 10 of 17 ) 3.0 OPERATOR ACTIONS (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED RESPONSE.
- 13. PLACE 1-HIC-62-93A, CHARGING Locally CONTROL 1-FCV-62-93 FLOW PZR LEVEL CONTROL, OR 1-FCV-62-89:
in MANUAL, and RESTORE pzr level to program
- REFER TO SOI-62.01, USING 1-FCV-62-93 CVCS - Charging and and/or 1-FCV-62-89. Letdown.
- 14. CHECK letdown IN SERVICE: ESTABLISH letdown:
- 1-FCV-62-69 OPEN.
- REFER TO Attachment 1.
- 1-FCV-62-70 OPEN. IF letdown can NOT be established,
- 1-FCV-62-77 OPEN. THEN PERFORM the following:
)
- Letdown orifice OPEN.
- a. CLOSE charging valves 1-FCV-62-90 and 1- FCV-62-91.
- b. MAINTAIN RCP seal flow between 8 and 13 gpm with charging valve controller 1-HIC-62-93A.
- c. PLACE excess letdown in service:
* . REFER TO SOI-62.01 SOI-62.01,,
CVCS - Charging and Letdown.
- d. ENSURE pzr control heater bank D0 red light LIT.
)
AOI-20 WBN MALFUNCTION OF PRESSURIZER Revision 31 LEVEL CONTROL SYSTEM -) _L-pag~_J1 Page 11 of 17
/
3.0 OPERATOR ACTIONS (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 15. CHECK pzr level RETURNING IF pzr level RISING to program. THEN
** GO TO STEP 17.
IF pzr level DROPPING THEN PERFORM the following:
- a. ISOLATE letdown.
- b. CLOSE charging valves 1-FCV-62-90 and 1- FCV-62-91.
- c. MAINTAIN RCP seal flow
) between 8 and 13 gpm with charging valve controller 1-HIC-62-93A.
- d. PLACE excess letdown in service:
- REFER TO SOI-62.01, CVCS - Charging 'and and Letdown.
- e. ENSURE PZR control heater bank D red light LIT.
- 16. ** GO TO Step 18.
)
AOI-20 MALFUNCTION OF PRESSURIZER - Revision 31 WBN LEVEL CONTROL SYSTEM
- Pa~2 Page 12 of 17 ) ~- -----
3.0 OPERATOR ACTIONS (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 17. PLACE 1-HIC-62-93A, CHARGING Locally CONTROL 1-FCV-62-93 FLOW PZR LEVEL CONTROL, OR 1-FCV-62-89:
in MANUAL, and RESTORE pzr level to program
- REFER TO SOI-62.01, USING 1-FCV-62-93 cves - Charging and and/or 1-FCV-62-89.
1-FCV-62":89. Letdown.
- 18. NOTIFY Work Control to initiate corrective action, if necessary.
- 19. EVALUATE system alignment/status:
- REVIEW actions performed in
)) this Instruction.
- REFER TO SOI-62.01 ,
CVCS - Charging and Letdown.
- 20. RETURN TO instruction in effect.
- END- )
11/2009 Watts Bar SRO NRC Exam- Exam - As submitted 10/2/2009
- 77. 029 G 2.2.22 77.029 2.2.22077 077 Given the following:
Time 0500 - Unit 1 was operating at 100% power when a turbine trip occurred. The reactor did not automatically automatically trip and the crew was unable to manually trip the reactor from the Main Control Room Room Trip SWjtches. Switches. 0501 - FR-S.1, "Nuclear Power Generation 1 ATWS," was entered to mitigate the event. 0503 - RCS pressure increase to a maximum of 2715 psig during the event. 0505 - AUO opened both Reactor Trip Breakers locally and all rods fully insert. 0506 - RCS pressure is restored to 2235 psig and stabilized. 0509 - Shift Manager determines the EAL had been met for a Site Area Emergency due to the reactor failing to trip. Which ONE of the following identifies ... (1) the action the .Shift Shift Manager Manager will take in accordance with the Radiological Emergency Plan, . and (2) if the RCS Pressure Safety Limit was exceeded during the event? (1 ) (2) Condition Safety Limit ) A. Declare the Site Area Exceeded Emergency B~ Declare the Site Area B:I NOT exceeded Emergency
- c. Report conditions had existed C. Exceeded for but NOT declare the Site Area Emergency.
D. Report conditions had existed NOT exceeded for but NOT declare the Site Area Emergency. Page 4
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR ANAL YSIS: A. Incorrect, the Site Area Emergency must be declared. An A TWS condition is an exception to the provision of reporting that conditions for an event classification had existed but are no longer are present before classifying an event but with the RCS pressure at a maximum of 2715 psig the limit was not exceeded. Plausible because reporting the event but not making a declaration would be correct when the conditions are resolved prior to making the declaration (with A TWS being an exception) and the pressure did peak at greater than the RCS system designed pressure. B. Correct, the Site Area Emergency must be declared. An A TWS condition is an "exception to the provision of reporting that conditions for an event classification had exception existed but are no longer are present before classifying an event. Tech Spec Section 2.0, Safety Limits, identifies the RCS pressure Safety Limit to be 2735 psig and with the RCS pressure at a maximum of 2715 psig the limit was not exceeded. C. Incorrect, the Site Area Emergency would be declared even thoughthoughthethe reactor has been tripped and the condition no longer exist but with the RCS pressure at a maximum of 2715 psig the limit was not exceeded. Plausible because normally if the condition causing the classification is resolved prior to classifying the event, the event is reported but not declared ( An A TWS TWS condition is an exception.) and the pressure did peak at greater than the RCS system designed pressure.
)
D. Incorrect, the Site Area Emergency would be declared even though the reactor has been tripped and the condition no longer exist. Tech Spec Section 2.0, Safety Limits, identifies the RCS pressure safety Limit to be 2735 psig and with the RCS pressure at a maximum of 2715 psig the limit was not exceeded. Plausible because normally if the condition causing the classification is resolved prior to classifying the event, the event is reported but not declared. An A TWS condition is an exception. Also because both the RCS not exceeding the Safety Limit is correct. Question Number: 77 Tier: 1 _1_ Group _1_ 1 KIA: 029 G 2.2.22 Anticipated Transient Without Scram (ATWS) Knowledge of limiting conditions for operations and safety limits. Importance Rating: 4.0/4.7 4.0 / 4.7 10 CFR Part 55: 41.5 / 43.2 / 45.2 )) 10CFR55.43.b: 2 Page 5
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 KIA Match: Applicant must demonstrate knowledge of the criteria in the ) Emergency Plan Implementing Procedures including conditions where events would be reported and not declared compared to conditions where the event must be declared. Applicant must also demonstrate knowledge of Technical Specification Safety Limits. SRO because the applicant must have knowledge of the REP criteria for classification of an event when the condition for classifying did exist but the condition is resolved before the declaration could be made. Technical
Reference:
FR-S.1, Nuclear Power Generation I A ATWS, TWS, Rev 19 Technical Specifications, 2.0, Safety Limits EPIP-1, Emergency Plan Classification Flowchart, Rev 31 Proposed references None to be provided: Learning Objective: 3-0T-PCD-048C
- 1. Classify emergency events.
- 5. Use the WBN Emergency Plan Implementing Procedures (EPIPs).
3-0T-TS-0200
- 2. Identify the Reactor Safety Limits.
Question Source: New X X Modified Bank Bank Question History: New question Comments: Page 6
WBN EMERGENCY PLAN EPIP-l EPIP-1
) CLASSIFICAnON Revision 31 FLOWCHART Page 3 of47 4
1.0 PURPOSE4 This Procedure provides guidance in determining the classification and declaration of an emergency based on plant conditions. 2.0 RESPONSIBILlTY 22;4,4 The responsibility of declaring an Emergency based on the guidance within this procedure belongs to the Shift Manager/Site Emergency Director (SM/SED) or designated Unit Supervisor (US) when acting as the SM or the TSC Site Emergency Director (SED). These duties CAN NOT be delegated. 4 3.0 INSTRUCTIONS 4 3.1 The criteria in WBN EPIP-1 are given for GUIDANCE ONLY: knowledge of actual plant conditions or the extent of the emergency may require that additional steps be taken. In all cases, this logic procedure should be combined )) with the sound judgment of the SM/SED and/or the TSC SED to arrive at a classification for a particular set of circumstances. 3.2 The Nuclear Power (NP) Radiological Emergency Plan (REP) will be activated when anyone of the conditions listed in this logic is detected. 3.3 Classification Determination 3.3.1 To determine the classification of the emergency, review the Initiating Conditions of the Events described in this procedure with the known or suspected conditions and CARRY OUT the notifications and actions referenced. 3.3.2 If a Critical Safety Function (CSF) is listed as an Initiating Condition: the respective status tree criteria will be monitored and used to determine the Event classification for the modes listed on the classification flowchart. 3.3.3 The highest classification for which an Emergency Action level (EAL) currently exists shall be declared.
WBN EMERGENCY PLAN EPIP-l
) CLASSIFICAnON CLASSIFICATION Revision 3311 I FLOWCHART Page 1_QfJ]_
. _page 4 of 47 3.0 INSTRUCTIONS (continued) 3.3.4 After an Event classification, if the following investigation shows that Initiating Conditions were met that dictate a higher Event classification, the new event classification shall be declared at the clock time of the determination.
3.3.5 IF an EAL for a higher classification ~ was exceeded but the present situation indicates a lower classification, the fact that the higher classification occurred SHALL be reported to the NRC and Central Emergency Control Center (CECC), but should not be declared. (Refer to SPP-3.5, Regulatory Reporting Requirements) 3.3.6 IF the Parameter is indeterminate due to instrument malfunction and the existence of the condition CAN NOT be reasonably discounted (i.e., spurious or false alarm that can be substantiated within 15 minutes) the condition is considered MET and the SM/SED SHALL follow the indications provided until such time as the alarm is verified to be false. Note IF an EAL was exceeded, but the emergency has been totally resolved ) prior to declaration, then EPIP 2, 3,4 and 5 are not applicable. 3.3.7 IF an EAL was exceeded, but the emergency has been totally resolved (prior to declaration), the emergency condition that was appropriate shall not be declared but reported to the NRC and Operations Duty Specialist (ODS) within one hour using SPP-3.5, Regulatory Reporting Requirements. 3.3.8 The ACCEPTABLE time frame for notification to the Operation Duty Specialist (ODS) is considered to be five (5) minutes. This is the time period between declaration of the emergency and notifying the ODS. 4.0 RECORDS 4.1 Non-QA Records None 4.2 QA Records None )
EPIP-1 EPIP-l Revision 31
) Page 130f 47 130f47 2.4 Fuel Clad Degradation Mode Initiating/Condition 1,2
.Loss Loss of Core cooling capability and VALID Trip Signals did not result in a reduction of Rx power to <5% and s
decreasing (1 and 2) Refer to "Fission Product Barrier Matrix" y
- 1. (a or b)
S
- a. CSF status tree indicates Core Cooling Red T
- b. CSF status tree indicates Heat Sink Red E
M
- 2. FR-S.1 entered and subsequent actions Did Not result in a Rx Power of <5% and decreasing Rx power Not <5% and decreasing after VALID Auto Refer to "Fission Product Barrier Matrix" D
1,2 and Manual trip signals (1 and 2 and 3) E
- 1. VALID Rx Auto T rip signal received or required Trip G
R
- 2. Manual Rx Trip from the MCR was Not successful. A
- 3. FR-S.1 has been entered. D A
T I o N
)) Automatic Rx trip did not occur after VALID Trip signal and manual trip from MCR was successful Refer to "Fission Product Barrier Matrix" U
1,2 (1 and 2) 1
- 1. VALID.Rx VALID Rx Auto Trip signal received or required
- 2. Manual Rx Trip from the MCR ~ was successful and power is <5% and decreasing.
- - - -_ _ _
- _ _ _ _ _ _ _ _ _ _ _ _ _ - I _ _ _ __
-*--**1--------*-*---- .... - - - - - - - - - - - --.-------
Not Applicable Reactor Coolant System specific activity exceeds LCO (Refer to WBN Tech. Spec. 3.4.16) 1,2, 1. Radiochemistry analysis indicates (a or b) 3,4, a. Dose equivalent Iodine (1-131) >O.26511Cilgm
>0.265 pCilgm 5 for >48 Hours QI
>21 ~lCi/gm.
flCi/gm.
>100/[ ~lCi/gm
- b. Specific activity >100/E pCilgm
))
TENNESSEE VALLEY AUTHORITY NP-REP APPENDlXC APPENDIX C WBN NUCLEAR POWER Page C-54 RADIOLOGICAL EMERGENCY PLAN Revision 89 Section 2.0 SYSTEM DEGRADATION Event 2.3 FAILURE OF RX PROTECTION Classification SITE AREA EMERGENCY Mode 1,2 1.2 Description Reactor power Not <5% and decreasing after VALID Auto and Manual Trip signals (1 and 2 and 3)
- 1. RX Auto Trip signal received or required.
VALID RXAuto
- 2. Manual RX Trip from the MCR was Not successful.
- 3. FR-S.1 has been entered.
Basis This IC indicates a failure of the automatic and main control room manual signals to scram the reactor. reactor. I Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as
)) anticipatory to the Fission Product Barrier Degradation IC, its inclusion is necessary to better beUer assure timely recognition and emergency response.
FR-S.1 lists actions intended to shutdown the reactor. This includes actions in the main control room and in other areas of the plant. FR-S.1 is utilized within the EAL to discriminate between those situations in which immediate manual reactor trip was not possible from the control room. The Unit 1 control room has two trip control locations on the main control board. Both are "Within vvithin immediate access for the reactor operator. If both fail to result in a reactor trip EOP E-O di rects the operator to FR-S.1. directs FR-S.1 . An indication or report or condition is considered to be VALID when it is conclusively verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel.. Implicit in this definition is the need for timely assessment, assessment. I.e., i.e., within 15 minutes. Escalation Escalation of this event would be based on the inability to trip the Rx and indications of Heat Sink Red or Core Cooling Red. References NUMARC/NESP-007, SS2, Rev. 2,1/92 NUMARC/NESP-007. T.S.3.3.1 Reactor Trip System (RTS) Instrumentation FR-S.1 Nuclear Power GenerationJATWS GenerationlATWS FR-H.1. Loss of Secondary Heat Sink FR-H.1, )
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1. 2.1. 2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained S~ 2735 psig. 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: ) 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.2.3 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.4 Within 24 hours, notify the Plant Manager and Site Vice President. 2.2.5 Within 30 days a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the NSRB, the Plant Manager, and Site Vice President. 2.2.6 Operation of the unit shall not be resumed until authorized by the NRC. Watts Bar-Unit 1 2.0-1
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) 78. 2.2.44 078 038 G 2.2.44078 . Given the following plant conditions:
- A tube rupture occurred on Unit 1 in Steam Generator #2.
- The operating crew is performing E-3, "Steam Generator Tube Rupture."
- After terminating safety injection, the operating crew observes the following conditions.
- RCS pressure is dropping
- Pressurizer level drops to 12% and is trending down with charging flow at maximum.
- Containment pressure, temperature, and radiation rising.
The correct operating crew action would be to _ _ _ _ _ ___ '
"Re-diagnosis."
A. Manually initiate Safety Injection and go to ES-O.O, Re-diagnosis. B. Manually initiate Safety Injection and return to E-O, "Reactor Trip or Safety Injection." C. Manually start ECCS pumps as necessary and continue in E-3, "Steam ) Generator Tube Rupture," until a transition is directed. D~ Manually start ECCS pumps as necessary and transition to ECA-3.1 D!' ECA-3.1,, Recovery ."
"SGTR and LOCA - Subcooled Recovery."
) Page 7
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DIS TRA CTOR ANAL YSIS: DISTRACTOR A. Incorrect, the conditions to not require the initiation of a safety injection nor do they need to implement ES-O. O. However, this distractor is plausible considering that the crew is responding to one casualty when another casualty occurs and ES-O. 0 can be used because a safety injection has occurred and a transition has been made from E-O.. B. Incorrect, Incorrect, the conditions to not require the initiation of a safety injection nor a return to E-O. However, this distractor is plausible considering that the crew is responding to one casualty when another casualty occurs and the manually initiation of safety injection would establish EGGS flow but it would realigned other plant component unnecessarily and delay correct mitigation actions .. G. Incorrect, Although manually starting the EGGS pumps as necessary is correct, the foldout page directs an immediate transition to EGA-3. 1 not to continue in E-3. D. Incorrect, Given the stated conditions, specifically, PZR level unable to be maintained> 15%, the foldout page of E-3 directs operators to manually start EGGS pumps as necessary and since Step 33 has been completed, a transition is to be made to EGA-1.1. Page 8
11/2009 Watts Bar SRO NRC NRC Exam - As submitted 10/2/2009 ) Question Number: 78 Tier: _1_ _1_ Group 1 KIA: 038 G 2.2.44 Steam Generator Tube Rupture Equipment Control Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. Importance Rating: 4.2 / 4.4 10 CFR Part 55: 41.5/43.5/45.12 10CFR55.43.b: 5 KIA Match: Applicant is required to assess the control room indications and information to determine the proper action and direction to be provided to the control room crew. . SRO because the applicant must assess plant conditions and prescribe an action and procedure to mitigate and recover. Technical
Reference:
E-3, Steam Generator Tube Rupture, Rev 22 Proposed references None to be provided: Learning Objective: 3-0T -EOP0300
- 5. Given a set of plant conditions, use E-3, ES-3.1,
. ES-3.2, and ES-3.3 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.
Question Source: New Modified Bank Bank x Question History: SQN bank question 038 G2.1.32 009, previously on North Anna Exam 2006 Comments: Question reworded to identify plant conditions instead of stating that a LOCA had occurred and to match WBN names and setpoints. Page 9
WBN STEAM GENERATOR TUBE RUPTURE E-3 Rev 22 FOLDOUT for E-3 SI REINITIATION CRITERIA Manually START ECCS pumps as necessary
- Pzr level cannot be maintained greater than 15% [33% ADVj OR
- RC S subcoollng less than 65°F [85°F ADVl RCS IF Step 33 has been completed AND SI Reinitiation occurs.
occurs THEN
..... GO TO ECA-31, ** ECA-3 1, SGTR and LOCA - Subcooled Recovery RCP TRIP CRITERIA
- Phase B Isolation, OR
- One charging pump OR one SI Sl pump Injecting AND RCS press dropped UNCONTROLLED to less than 1500 pSlg EVENT DIAGNOSTIC TRANSITIONS
- IF any S/G press low or dropping uncontrolled AND has NOT been Isolated.
THEN
** GO TO E-2, Faulted Steam Generator Isolation. unless that S/G IS needed for cool down cooldown
- IF intact S/G radiation abnormal or S/G level rising uncontrolled.
THEN
** GO TO E-3, E-3, Steam Generator Tube Rupture, Step 1 SUMP RECIRC SWITCHOVER CRITERIA
- IF RWST level less than 34%,
THEN
** GO TO ES-1 3, Transfer to RHR Containment Sump AFW OPERATION
- IF CST volume less than 5000 gal THEN MONITOR AFW pumps to ensure suction transfer
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 79. 058 AA2.02 079 Given the following plant conditions:
- Unit 1 is at 100% power
- The following alarm is received in the control room:
- Window 18-A, 125 DC VITAL CHGRIBATT II ABNORMAL.
- Battery Board II Voltage indicates 130v DC and lowering.
- Battery Charger II DC Output Breaker is tripped open.
Which ONE of the following identifies (1) why the 125V DC Vital Battery Channel II is inoperable, and (2) how long the battery is designed to be able to maintain greater than the minimum terminal voltage for the current plant conditions? A. (1) Declare Battery Channel II INOPERABLE because of the voltage is less than required by Technical Specifications; (2) 30 minutes B. (1) Declare Battery Channel II INOPERABLE because of the voltage is less than required by Technical Specifications; (2) 4 hours C. (1) Declare Battery Channel II INOPERABLE because the battery board is ) not connected to a charger; (2) 30 minutes D~ (1) Declare Battery Channel II INOPERABLE because the battery board is D!' not connected to a charger; (2) 4 hours Page 10
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR ANAL YSIS: A. Incorrect, the voltage is above the minimum of 128v, the battery channel is inoperable because a battery charger is not connected to the battery board and the time the battery can supply the design bases loads without an accident is 4 hours, not 30 minutes. Plausible because the 125v DG battery must be 132v to be operable and 30 minutes is the time the battery can supply the design bases loads with an accident present present: B. Incorrect, the voltage is above the minimum of 128v, the battery channel is inoperable because a battery charger is not connected to the battery board but the battery will supply the design bases loads during normal operation (without an accident) for 4 hours. Plausible because the 125v DG battery must be 132v to be operable and 4 hours being the time the battery can supply the design bases loads with an accident present is correct. C. Incorrect, the battery channel is inoperable because a battery charger is not connected to the battery board but the time the battery can supply the design bases loads without an accident is 4 hours, not 30 minutes. Plausible because being inoperable because a battery charger is not connected is correct and 30 minutes is the time the battery can supply the design bases loads with an accident present. ) D. Correct, the battery channel is inoperable because a battery charger is not connected to the battery board. Tech Spec Bases identifies that an operable DC sub system will have the required battery and respective charger operating and connected to the DC buses. The Bases and the annunciator response also indicate the battery will supply the design bases loads during normal operation (without an accident) for 4 hours. Question Number: 79 Tier: _1 __ Group _1_ 1 KIA: 058 AA2.02 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power: 125V de bus voltage, low/eritieallow, alarm.
-
Importance Rating: 3.3* / 3.6 ) 10 CFR Part 55: 43.5/45.13 43.5 / 45.13 10CFR55.43.b: 2 Page 11
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ')
, KIA Match:
r Applicant is required to assess conditions on a station battery with dropping voltage and determine why the battery channel is inoperable and recall the required coping time that the battery is designed to maintain greater than the minimum voltage while supplying design bases loads. SRO because the question requires knowledge of the Tech Spec bases as to what constitutes operability and the coping time as included in the bases for the LCO. Technical
Reference:
ARI-15-21, CNTL PWR & FIRE PROT, Rev 22 Technical Specifications 3.8.4, DC Sources - Operating Technical Specification Bases, 3.8.4, DC Sources - Operating Proposed references None to be provided: Learning Objective: 3-0T-SYS057P "
- 10. Given the condition/status of the 125V DC Vital system/component and the appropriate sections of Tech Specs, determine if operability requirements are met and what actions, if any, are required.
- 11. State the 125V DC Vital system parameters
) governed by TS. Question Source: New Modified Bank Xx Bank Question History: SON question used on 2007 exam modified Comments: ) Page 12
) 18-A SOURCE Amp Relay for Batt. UV Relay on Batt. feed to Batt. Bd. II 125 DC VITAL UV Relay on AC supply to Charger CHGRIBA CHGRIBATT TT II Contacts on Charger DC output breaker ABNORMAL OV Relay on Charger DC output Probable Cause: 1. Charger failure or charger out of service
- 2. Battery overcharging
- 3. Charger DC supply breaker open or tripped .
- 4. Bkr 36 on 120V ac Vital Instrument Power Bd 1-11 open Note 125V DC Vital Battery II isable is able to supply design basis loads for 4 hours while maintaining a minimum terminal voltage of 1 05V DC.
105V The minimum terminal voltage during the first minute is 113V DC. Corrective Action: [1] IF Vital Battery Bd II is lost, THEN GO TO AOI-21.02, LOSS OF OF 125V DC BATTERY BD II. [2] CHECK Battery II voltage and amps [1-M-1]. [3] DISPATCH Operator to determine cause of alarm. [4] IF needed, THEN 801-236.02, 125V DC PLACE spare charger in service per SOI-236.02, BA TTERY BOARD II. BATTERY [5] REFER TO Tech Specs.8pecs. [st [6] NOTIFY Work Control to initiate corrective action, if necessary.
References:
1-45W600-55-3 1-45W703-2 AOI-21.02 801-236.02 SOI-236.02 WB-DC-30-27 ARI-15-21 ARI-1S-21 ) WBN Page 22 of 48 Rev 22
) 18-B 18-8 SOURCE SETPOINT Batt Bd II Ground 125 DC VITAL Any Batt Bd II fuse Blown BATT BD II Any Batt Bd II breaker Tripped ABNORMAL 120 VAC Instrument Pwr Tripped Bd 1-11 Bkr Probable Cause: 1. Feeder overload
- 2. Ground on bus or feeder
- 3. Blown fuse
- 4. Any Batt Bd II breaker tripped
- 5. Bkr 36 on 120 VAC Vital Instrument Power Bd 1-11 open CAUTION Power loss to some feeders could cause a unit trip.
) Corrective Action: [1] IF Vital Battery Bd II is lost, THEN* THEN GO TO AOI-21.02, LOSS OF 125V DC BATTERY BD II. [2] DISPATCH Operator to determine cause of alarm. [3] 801-236.02, 125V DC BA TTERY BOARD II, for REFER TO SOI-236.02, system operating instructions. [4] NOTIFY Work Control to initiate corrective action, if necessary. [5] IF High or Low Voltage condition exist, THEN REFER TO 0-SI-236-22 0-81-236-22 for performance.
References:
1-45W600-55-3 1-45W703-2 AOI-21.02 801-236.02 SOI-236.02 ) WBN I Page 23 of 48 ARI-15-21 Rev 22
DC Sources - Operating 3.8.4 ) 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4 Four channels of vital DC and four Diesel Generator (DG) DC electrical power power subsystems shall be OPERABLE.
-----------------------------------------------N 0 IEES----------------------------------------------
------------------------------------------------NOT S------------------------------------------------
- 1. Vital Battery V may be substituted for any of the required vital batteries.
- 2. The C-S DG and its its associated DC electrical power subsystem may be substituted for any of the required DGs and their associated DC electrical power subsystem.
APPLICABILITY: MODES 1,2, 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One vital DC electrical A.1 Restore vital DC electrical 2 hours power subsystem power subsystem to inoperable. OPERABLE status. B. Required Action and B.1 Be in MODE 3. 6 hours )) Associated Completion Time of Condition A not AND met. B.2 Be in MODE 5. 36 hours C. One DG DC electrical C.1 Restore DG DC electrical power 2 hours power subsystem subsystem to OPERABLE inoperable. status.
---- -------
(continued) Watts Bar-Unit 1 3.8-24 )
DC Sources - Operating 3.8.4 ) ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and 0.1 D.1 Declare associated DG Immediately associated Completion inoperable. Time of Condition C not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 ~ 128 V Verify vital battery terminal voltage is :2 (132 V V(132 7 days for vital battery V) on float charge charge..
. SR 3.8.4.2 Verify DG battery terminal voltage is :2 ~ 124 V on float 7 days charge.
SR 3.8.4.3 Verify for the vital batteries that the alternate feeder 7 days breakers to each required battery charger are open. ) SR 3.8.4.4 Verify correct breaker alignment and indicated power 7 days availability for each DG 125 V DC distribution panel and associated battery charger. (continued) Watts Bar-Unit 1 3.8-25
DC Sources-Operating B 3.8.4 ') BASES BACKGROUND Subsystem (continued) 125 V Vital DC Electrical Power Subsystem Additionally, battery boards I, II, III, and IV have manual access to the fifth vital battery system. The fifth 125V DC Vital Battery System is intended to serve as a replacement for anyone of the four 125V DC vital batteries during their testing, maintenance, and outages with no loss of system reliability under any mode of operation. Each of the vital DC electrical power subsystems provide the control power for its associated Class 1 E AC power load group, 6.9 kV switchgear, and 480 V load centers. The vital DC electrical power subsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses. Additionally, they* they power the emergency DC lighting system. . The vital DC power distribution system is described in more detail in Bases for LCO 3.8.9, "Distribution System - Operating," and LCO 3.8.10, "Distribution Systems - Shutdown." Each vital battery has adequate storage capacity to carry the required load continuously for at least 4 hours in the event of a loss of all AC power (station blackout) without an accident or for 30 minutes with an accident considering a single failure. Load shedding of nonrequired loads will be performed to achieve the required coping duration for station blackout conditions. Each 125 VDC vital battery is separately housed in a ventilated room apart from
) its charger and distribution centers, except for Vital Battery V. Each subsystem is located in an area separated physically and electrically from the other subsystem to ensure that a single failure in one subsystem does not cause a failure in a redundant subsystem. There is no sharing between redundant Class 1 E subsystems, such as batteries, battery chargers, or distribution panels.
The batteries for the vital DC electrical power subsystems are sized to produce required capacity at 80% of nameplate rating, corresponding to warranted capacity at end of life cycles, derated for minimum ambient temperature and the
,.
(continued)
) Watts Bar-Unit 1 B 3.8-55
DC Sources-Operating B 3.8.4 BASES APPLICABLE The OPERABILITY of the DC sources is consistent with the initial assumptions SAFETY ANALYSES of the accident analyses and is based upon meeting the design basis of the (continued) (continued) plant. This includes maintaining the DC sources OPERABLE during accident conditions in the event of:
- a. An assumed loss of all offsite AC power or all onsite AC power; and
- b. A worst case single failure.
The DC sources satisfy Criterion 3 of the NRC Policy Statement. LCO Four 125V vital DC electrical power subsystems, each vital subsystem channel consisting of a battery bank, associated battery charger and the corresponding control equipment and interconnecting cabling supplying power to the associated DC bus within the channel; and four DG DC electrical power subsystems each consisting of a battery, a battery charger, and the corresponding control equipment and interconnecting cabling are required to be OPERABLE to ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (AOO) (ADO) or a postulated DBA. Loss of any DC DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4). An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses. The LCO is modified by two Notes. Note 1 indicates that Vital Battery V may be substituted for any of the required vital batteries. However, the fifth battery cannot be declared OPERABLE until it is connected electrically in place of another battery and it has satisfied applicable Surveillance Requirements. Note 2 has been added to indicate that the C-S DG and its associated DC subsystem may be substituted for any of the required DGs. However, the C-S DG and its associated DC subsystem cannot be declared OPERABLE until it is connected electrically in place of another DG, and it has satisfied applicable Surveillance Requirements. (continued) Watts Bar-Unit 1 B 3.8-57
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 80. 077 AA2.07 080 80.077 Given the following plant conditions:
Unit 1 is at 100% power. Diesel Generator 1B-B is operating and paralleled to the grid for a surveillance.
- A switching error at the Hydro Unit causes a Loss of Offsite Power to 'D'
'0' CSST.
- The resulting grid disturbance causes the 6.9kV Shutdown Board 1B-B Emergency Feeder Breaker 1914 to be tripped by the overcurrent (50) relay on the breaker.
Which ONE of the following describes ... (1) the expected status of CCP 1B-B and RHR pump 1B-B 30 seconds after the event, and (2) the required Tech Spec completion time for restoring the offsite power supply? (1 ) (2) Pump Status Required Action Completion Time A. Only the CCP pump running 12 hours B. Both CCP and RHR pump running 12 hours C~ Only the CCP pump running 72 hours D. Both CCP and RHR pump running 72 hours Page 13 Page
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ') DISTRACTOR ANAL YSIS: A. Incorrect. Only the CCP would be running but the required time to restore the off-site power supply is 72 hours not 12 hours. Plausible, since the only the CCP running is correct and 12 hours for the Required Action Completion Time given in numerous places throughout the same Tech. Spec. 8; B. Incorrect, both the CCP and the RHR pump being started is not correct. Only the CCP will be running after the shutdown board is restored by the diesel generator. There would have to be a safety injection signal present for the RHR pump to start. The required time to restore the off-site power supply is 72 hours not 12 hours. Plausible because the RHR would be running if a safety injection signal was present and 12 hours for the Required Action Completion Time is also plausible since this time is given in numerous places throughout the same Tech. Spec. C. Correct, after the shutdown board was energized by the diesel generator, the CCP would automatically be sequenced on. The RHR pump would not start because there is no safety injection signal present. With a loss of one of the offsite power supplies, Technical Specifications require the power supply to be restored within 72 hours.
-D. ~,
Incorrect, both the CCP and the RHR pump being started is not correct. Only the CCP will be running after the shutdown board is restored by the diesel generator. There would have to be a safety injection signal present for the RHR pump to start. The required time to restore the off-site power supply is 72 hours.. Plausible because the RHR would be running if a safety injection signal was present and 72 hours for the Required Action Completion Time is correct. Question Number: 80 Tier: _1_ _1 _ Group 1 KIA: 077 AA2.07 Generator Voltage and Electric Grid Disturbances Determine and interpret operational status of engineered safety features. Importance Rating: 3.6 1/ 4.0 10 CFR Part 55: 41.5 arid and 43.5/45.5, 43.5/45.5,45.7, 45.7, and 45.8 10CFR55.43.b: 2 KIA Match: Applicant must apply knowledge of relay operations in the context of
) a grid disturbance and determine the effect that has on the . operational status of ECCS pumps being supplied from the Shutdown Board. Applicant also is required to assess provided Page 14
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 KIA Match: Applicant must apply knowledge of relay operations inthe in the context of that has on the a grid disturbance and determine the effect thathas operational status of ECCS pumps being supplied from the Shutdown Board. Applicant also is required to assess provided information to make a decision on completion time allowed to restore the operational status of the 1E electrical distribution system and allow exiting the LCO. Question is SRO because it requires detailed knowledge of the application of Tech Specs, including required completion time for conditions requiring LCO entry (knowledge below the double line). Technical
Reference:
Technical Specification LCO 3.8.1 3-0T-SYS201 3-0T-SYS201B, B, Blackout and Load Shed Logic, Rev 5 Proposed references None to be provided: Learning Objective: 3-0T-STG-201A, "6.9kV System,"
- 21. Identify consequences of abnormal conditions in or improper work on electrical distribution systems, and include what those conditions are. SOER 90-001, Rec. 4.
3-0T-SYS201 3-0T -SYS20 1B
- 16. Identify the equipment and associated (sequence
) timer) ST setpoint that will start with the following:
- a. A blackout with a return of voltage.
- b. blackout with a return of voltage together with a safety injection signal.
Question Source: New xX Modified Bank Bank Question History: New question Comments:
)
Page 15
3-0T-SYS201 B Rev 8 Page 45 of 79 pages Equipment Name Start Time Start Time ) BO only BO with SI Miscellaneous Loads 0 0 Centrifugal Charging Pump & Rm Clr 5 sec. 5 sec. Safety Injection Pump & Rm Clr NJA 10 sec. Residual Heat Removal Pump & Rm Clr NJA 15 sec. Essential Raw Cooling Water Pump 20 sec. 20 sec. Component Cooling System Pump 35 sec. 35 sec. Thermal Barrier Booster Pump 35 sec. 35 sec. Auxiliary Feedwater Pump 25 sec. 25 sec. Fire Pump 40, 40, 45, 55 sec. 40,40, 45, 55 sec. Pressurizer Heaters (Backup Group) 90 sec. 90 sec. Containment Spray Pump & Rm Clr 184 sec. 184 sec. Shutdown BO Rm Chiller Pkg 360 sec. 360 sec. Main Control Rm Chiller Pkg 360 sec. 360 sec. Electrical Bd Rm Chiller Pkg 360 sec. 360 sec. Blackout Load Sequence Objective 16 Discuss the Blackout Load sequence for equipment that times back on after the DG restores voltage to the SD Bd.
- The table on the slide depicts the loads that are sequenced back on the 480V and 6.9kV shutdown boards following a loss and restoration of power to the boards.
- Note that some loads require an additional signal, not just blackout signal, to initiate their ST timers:
- SI pumps & Rm Clr require an SI signal with the blackout
- RHR pumps & Rm Clrs require an SI signal with the blackout
- Fire pumps require a fire pump start signal from the fire protection system to start.
- Containment Spray pumps & & Rm Clrs require a Hi Hi Cntmt pressure signal to start.
- Should a safety injection signal occur after a blackout with voltage return on the 6.9kV SD Bds, all Start Time (ST) timers will reset or start timing - in the case of those loads that start only on an SI signal. Should any of the equipment that sequences back onto the shutdown boards have their timers already timed out and their breakers already closed following the blackout sequence, then they will not trip due to receiving a safety injection signal.
Equipment Name Start Time Start Time BO only BO with SI Miscellaneous Loads 0 0 Centrifugal Charging Pump & Rm Clr 5 sec. 5 sec. Safety Injection Pump & Rm Clr N/A 10 sec. Residual Heat Removal Pump & Rm Clr N/A 15 sec. Essential Raw Cooling Water Pump 20 sec. 20 sec. Component Cooling System Pump 35 sec. 35 sec. Thermal Barrier Booster Pump 35 sec ... 35 sec. Auxiliary Feedwater Pump 25 sec. 25 sec. Fire Pump 40, 40,45, 55 sec. 40, 40, 45, 55 sec. Pressurizer Heaters (Backup Group). Group) 90 sec. 90 sec. Containment Spray Pump & Rm Clr 184 sec. 184 sec. Shutdown B'DBO Rm Chiller Pkg 360 sec. 360 sec. Main Control Rm Chiller Pkg 360 sec. 360 sec. Electrical BdRmChilier Bd Rm Chiller Pkg 360 sec. 360 sec.
AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:
- a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and
- b. Four diesel generators (DGs) capable of supplying the onsite Class 1 E AC Electrical Power Distribution System.
-----
----------
--- --- --------
----- --- --- -----
--- -- --- --- ---- ----------
-- ------------ -NOT E-- --
---NOT --- ---
----- ------
------ -- -------
-------- ----- ------- --- ---
---------- ----------
----- --- -------
--
The C-SCoS DG may be substituted for any of the required DGs. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS
-------------------
-------- ----------- ------------- ---------
---- ------ ------ ----------
----- ------- ------ ----
----------- --NOT
----- NOT E------- --- --
------ -- ----
---- --- ----------------
--- --------- --------------- ----- -----------
-- -----------------------------
LCO 3.0.4.b 3.0A.b is not applicable to DGs. CONDITION REQUIRED ACTION COMPLETION TIME A. One offsite circuit A.1 Perform SR 3.8.1.1 for 1 hour inoperable. OPERABLE offsite circuit. AND Once per 8 hours thereafter AND A.2 Declare required feature(s) with 24 hours from no offsite power available discovery of no inoperable when its redundant offsite power to one required feature(s) is train concurrent with inoperable. inoperabilityof redundant required feature(s) AND ((continued) continued) ) Watts Bar-Unit 1 3.8-1 Amendment 55
AC Sources -Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ((co continued) ntinued) A.3 Restore offsite circuit to 72 hours OPERABLE status. AND 6 days from discovery of failure to meet LCO B. One required DG B.1 Perform SR 3.8.1.1 for the 1 hour inoperable. offsite circuits. AND
,.
Once per 8 hours thereafter AND B.2 Declare required feature(s) 4 hours from supported by the inoperable discovery of DG inoperable when its Condition B required redundant feature(s) concurrent with is inoperable. inoperabilityof inoperability of redundant required feature(s) AND B.3.1 Determine OPERABLE DGs 12 hours are not inoperable due to common cause failure. OR B.3.2 Perform SR 3.8.1.2 for 12 hours OPERABLE DGs. OGs. AND ((continued) continued) Watts Bar-Unit 1 3.8-2 Amendment 39
AC Sources -Operating 3.8.1
)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.4 Restore required DG to 14 days OPERABLE status. AND
\\
17 days from discovery of failure to meet LCO C. Two required DGs in Train C.1 Perform SR 3.8.1.1 for the 1 hour A inoperable. offsite circuits. AND OR '"" Once per 8 hours Two required DGs in Train thereafter B inoperable. AND C.2 Declare required feature(s) 4 hours from supported by the inoperable discovery of DGs inoperable when its Condition C ) required redundant feature(s) is concurrent with inoperable. inoperability of inoperabilityof redundant required feature(s) AND C.3.1 Determine OPERABLE DGs are not inoperable due to 12 hours common cause failure. OR C.3.2 Perform SR 3.8.1.2 for OPERABLE DGs. 12 hours AND (continued)
- ----
Watts Bar-Unit 1 3.8-2a Amendment 39
( >iNVl8 >1NVl8 l.:l3l 1.:l3l AllVNOI1N31NI 38Vd SIHl
AC Sources -Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.4 Restore at least one required 72 hours DG to OPERABLE status. AND 6 days from discovery of failure to meet LCO D. Two offsite circuits D.1 Declare required feature(s) 12 hours from inoperable. inoperable when its redundant discovery of required feature(s) is Condition D inoperable. concurrent with
.. inoperability of redundant required features AND D.2 Restore one offsite circuit to 24 hours OPERABLE status.
(continued) Watts Bar-Unit 1 3.8-3 Amendment 30, 39
AC Sources -Operating 3.8.1
)
ACTIONS
.. -_
- .. _.. - _(continued)
...... _,....--
CONDITION REOUIRED REQUIRED ACTION COMPLETION COMPLETION TIME E. One offsite circuit ------- -------- ------- -------N--- ---NOT 0 T E-----------------------
------ ----- --------- ---
inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, AND "Distribution Systems - Operating," when Condition E is entered with no AC One or more required power source to any train. DG(s) in Train A --_ -_
....... -_ -.. --
.... .... --_ .. -----
.. - ............
-- ..........
... ------ ---- ... ----_
__ .... _--- -_..........
-- _......... _ -_ ............
- ....-----_ ..... ---
inoperable. E.1 Restore offsite circuit to 12 hours OR OPERABLE status. One or more required OR DG(s) in Train B inoperable. E.2 Restore required DG(s) to 12 hours OPERABLE status. F. One or more required F.1 Restore required DGs in Train 2 hours DG(s) in Train A A to OPERABLE status. inoperable. OR AND F.2 Restore required DGs in Train One or more required B to OPERABLE status 2 hours DG(s) in Train B
..
, inoperable.
inoperable . G. Required Action and G.1 Be in MODE 3. 6 hours Associated Completion Time of Condition A, B, C, AND D, E, or F not met. G.2 Be in MODE 5. 36 hours
- --- ---- --------_ .. - ---
(continued) ) Watts Bar-Unit 1 3.8-4 Amendment 39
AC Sources -Operating 3.8.1 ACTIONS .. - .. - .. - (continued)
- .. - . . . -- --
CONDITION REQUIRED REOUIRED ACTION COMPLETION TIME H. Two offsite circuits H.1 Enter LCO 3.0.3. Immediately inoperable. AND One or more required DG(s) in Train A inoperable. OR One or more required DG(s) in Train B inoperable. I. One offsite circuit 1.1 Enter LCO 3.0.3. Immediately inoperable. AND One or more required DG(s) in Train A inoperable. AND One or more required requi.red DG(s) in Train B inoperable. Walls Watts Bar-Unit 1 3.8-5 3.5-5 Amendment 39
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7 days availability for each offsite circuit. SR 3.8.1.2 ------
---- --------
-- --- ---------
---------- --------
--- ------- NOT ES ----
--NOT ------------- ----- ----------
--------------- --- ------
- 1. Performance of SR 3.8.1.7 satisfies this SR.
- 2. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
Verify each DG starts from standby conditions and As specified in achieves steady state voltage ~ 6800 V and and::; S 7260 V, Table 3.8.1-1 and frequency ~ 58.8 Hz and s 61.2 Hz. and::; (continued)
)
) Watts Bar-Unit 1 3.8-6
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS REOUIREMENTS (continued) SURVEILLANCE FREQUENCY FREOUENCY SR 3.8.1.3 ----------
------ -----------------------
---------- --- --- ------- ---- NOT EES----------------------------
S------------------- ---------
- 1. DG loadings may include gradual loading as recommended by the manufacturer.
- 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.
Verify each DG is synchronized and loaded and As specified in operates for ~ 60 minutes at a load ~ 3960 kW and Table 3.8.1-1
- s; 4400 kW.
SR 3.8.1.4 Verify each skid mounted day tank contains 31 days ) ~ 218.5 gal of fuel oil. SR 3.8.1.5 Check for and remove accumulated water from each 31 days skid mounted day tank. SR 3.8.1.6 Verify the fuel oil transfer system operates to 31 days automatically transfer fuel oil from 7 day storage tank to the skid mounted day tank. (continued) Watts Bar-Unit 1 3.8-7
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS REOUIREMENTS(continued2 (continued) SURVEILLANCE FREQUENCY FREOUENCY SR 3.8.1.7 Verify each DG starts from standby condition and I 184 days achieves in :s; ::; 10 seconds, voltage ~ 6800 V, and frequency ~ 58.8 Hz. Verify after DG fast start from standby conditions that the DG achieves steady state voltage ~ 6800 V and and::;:s; 7260 V, and frequency ~ 58.8 Hz andand::; :s; 61.2 Hz. SR3.8.1.8 SR 3.8.1.8 --------
--- --- -- ----- ------
--- -------------- ---- -----
-------NOT NOTE E------ ----- --
----------- ------- -- ---
----- -------- --- ---
------
This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify automatic and manual transfer of each 6.9 kV 18 months shutdown board power supply from the normal offsite circuit to each alternate offsite circuit. (continued) (continued) ) Watts Bar-Unit 1 3.8-8
AC Sources-Operating 3.8.1
)
SURVEILLANCE REQUIREMENTS kontinued) (continued) SURVEILLANCE FREQUENCY SR 3.8.1.9 ----- -- --------------- ---- ------NOT
-------------------------------- N OT ES ---- ------- ----- --- --- ------
----------------------------
- 1. This Surveillance shall not be performed in MODE 1 or 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;::
factor 2: 0.8 and:; and::; 0.9. Verify each DG rejects a load greater than or equal to 18 months its associated single largest post-accident load, and:
- a. Following load rejection, the frequency is
-
- 66.75 Hz;
-
- b. Within 3 seconds following load rejection, the voltage is ;:: 2: 6555 V and:;
and::; 7260 V; and
- c. Within 4 seconds following load rejection, the frequency is;:: is 2: 58.8 Hz and:;
and::; 61.2 Hz. SR 3.8.1.10 -- ------------
--- -- -----
--- --- ------- ---------------------------------
--- -- ---- ---NOT E---------- ----- ------- --- ----- ---
This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify each DG operating at a power factor;:: factor 2: 0.8 and I 18 months
-
- 0.9 does not trip and voltage is maintained
- ; maintained::; 8880 V during and following a load rejection of ;:: 2: 3960 kW and
-
- 4400 kW and;
- : and 2: 2970 kVAR and:; and::; 3300 kVAR.
-
(continued) ) Watts Bar-Unit 1 3.8-9
AC Sources-Operating 3.8.1
)
SURVEILLANCE REQUIREMENTS continued) SURVEILLANCE FREQUENCY SR 3.8.1.11 ---- ----- --------
----------- --- --- ----- ----- -- ----NOT
---- -------- --- -NOT E--- ----
-- --- --------
---- ---------- --------
----- ------ --
-- ---
This Surveillance shall not be performed in MODE 1, 2,3, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR. Verify on an actual or simulated loss of offsite power 18 months signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses;
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected loads in ~ 10 seconds,
- 2. energizes auto-connected shutdown loads through automatic load sequencer,
)
- 3. maintains steady state voltage
~ 6800 V and ~ 7260 V,
- 4. maintains steady state frequency
~ 58.8 Hz and ~ 61.2 Hz, and
- 5. supplies permanently connected and auto-connected shutdown loads for
~ 5 minutes.
(continued) ) Watts Bar-Unit 1 3.8-10
AC Sources-Operating 3.8.1 SURVEILLANCE FREOUENCY FREQUENCY SR3.8.1.12 ---
----------- --- --- -----------
-------------- ----NOT
------------- NOT E -- -- --------- ------- ---
--------------------- -------
---------
This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify on an actual or simulated Engineered Safety 18 months Feature (ESF) actuation signal each Unit 1 DG auto-starts from standby condition and:
- a. In s~ 10 seconds after auto-start and during tests, achieves voltage 2'~: 6800 V and frequency 2'~: 58.8 Hz;
- b. After DG fast start from standby conditions the DG achieves steady state voltage 2'~: 6800 V and s~ 7260 V, and frequency 2'~: 58.8 Hz and s~ 61.2 Hz.
- c. Operates for 2'~: 5 minutes;
- d. Permanently connected loads remain energized from the offsite power system; and
- e. Emergency loads are energized from the offsite power system.
(co nti nued) (continued) Watts Bar-Unit 1 3.8-11
AC Sources-Operating 3.8.1 SURVEILLANCE REOUIREMENTS (continued) SURVEILLANCE FREOUENCY SR 3.8.1.13 ----- ----------
---- ------- ------------ ----NOT
---------------- --- -NOT E------------------------------
---- -- -- ------------------ ----
This SUNeiliance SUNeillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify each DG's automatic trips are bypassed on 18 months automatic or emergency start signal except:
- a. Engine overspeed; and
- b. Generator differential current.
SR3.8.1.14 SR 3.8.1.14 ---- -------------- ------ --- --- --NOT E S----- ES- ---- ----- --- --- --- ----------
- 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
- 2. For performance of this test in MODE 1, 2, 3 or 4, three DGs must be maintained operable and in a standby condition.
- 3. Credit may be taken for unplanned events that satisfy this SR.
Verify each DG operating at a power factor ~ 2': 0.8 and 18 months s:~ 0.9 operates for ~2': 24 hours:
- a. For ~2 2 hours loaded ~2 4620 kW and s:~ 4840 kW and ~2 3465 kVAR and s: ~ 3630 kVAR; and
- b. For the remaining hours of the test loaded
~
2': 3960 kW and ~s: 4400 kW and ~22970 2970 kVAR and ~s: 3300 kVAR. ((continued) continued) Watts Bar-Unit 1 3.8-12 Amendment 12
AC Sources-Operating 3.8.1
)
SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.15 ------
--- --------------
--------------- -------------NOT
-- --- ------------------------------
---- -- -- --NOT E---- -- ------- -----------------
This Surveillance shall be performed within 5 minutes of shutting down the DG OG after the DG OG has operated
~ 2 hours loaded 2:
2: ~ 3960 kW and ~ 4400 kW. Momentary transients outside of load range do not invalidate this test. Verify each DG OG starts and achieves, in ~ 10 seconds, I 18 months voltage 2: ~ 6800 V, and frequency ~ 2: 58.8 Hz. Verify after DG OG fast start from standby conditions that the DG OG achieves steady state voltage 2: ~ 6800 V and ~ 7260 Y, V, and frequency ~ 2: 58.8 Hz and ~ 61.2 Hz. SR 3.8.1.16 -- ---- ------------ -- ----
------------------ -------- NOT E----------
-- -NOT
----- --- --- ------ ---- --- ----
---------------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for 2,3, unplanned events that satisfy this SR. Verify each DG: OG: 18 months
- a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
- b. Transfers loads to offsite power source; and
- c. Returns to ready-to-Ioad operation.
(continued) ) Watts Bar-Unit 1 3.8-13 3.8~13
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE SU RVEI LLANCE FREQUENCY SR3.8.1.17 ------------ ------ --------- -- -- NOT E------
------------------------------- -- -- --- --- ---------- ---- --
--------------------------------
This Surveillance shall not be performed in MODE 1, 2,3, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR. Verify, with each Unit 1 DG operating in test mode and I 18 months connected to its bus, an actual or s'imulated ESF actuation signal overrides the test mode by:
- a. Returning DG to ready-to-Ioad operation; and
- b. Automatically energizing the emergency load from offsite power.
SR 3.8.1.18 ------- ------------------N 0 T E-------------------------------
-------------------------N
--------------
This Surveillance shall not be performed in MODE 1, 2, 2,3,3, or 4. However, credit may be taken for unplanned events that satisfy this SR. " Verify the time delay setting for each sequenced load I 18 months block is within limits for fo r each accident condition and non-accident condition load sequence. (continued) Watts Bar-Unit 1 3.8-14
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.19 ----------------------
----- ----- --- ---- -------- -------- N OT E--
-----NOT --------------------------------
-- -- ------ -- ---- ---- ---- -- ----
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR. Verify on an actual or simulated loss of offsite power 18 months signal in conjunction with an actual or simulated ESF actuation signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses;
- c. DGs of the same power train auto-start from standby condition and:
- 1. energizes permanently connected loads in s; s 10 seconds,
- 2. energizes auto-connected emergency loads through load sequencer,
- 3. achieves steady state voltage:
;:::~ 6800 V and S;s 7260 V,
- 4. achieves steady state frequency
;:::~ 58.8 Hz and s 61.2 Hz, and S;
- 5. supplies permanently connected and auto-connected emergency loads for
;:::~ 5 minutes.
SR 3.8.1.20 Verify during idle operation that any automatic or 18 months emergency start signal disables the idle start circuitry and commands the engine to full speed. (continued) Watts Bar-Unit 1 3.8-15
AC Sources-Operating 3.8.1 FREQUENCY SR 3.8.1.21 Verify when started simultaneously from standby 10 years condition, each DG achieves, in ~ 10 seconds, in:::: voltage ~ 6800 V and frequency ~ 58.8 Hz. Verify after DG fast start from standby conditions that the DG achieves steady state voltage ~ 6800 V and :::: ~ 7260 V, and frequency ~ 58.8 Hz and ~ 61.2 Hz. and:::: Watts Bar-Unit 1 3.8-16
AC Sources-Operating 3.8.1 ') Table 3.8.1-1 (page 1 of 1) Diesel Generator Test Schedule NUMBER OF FAILURES IN LAST 25 VALID TESTS(a) FREQUENCY
- 3
~3 31 days
~4 7 days(b)
(but no less than 24 hours) (a) Criteria for determining number of failures and valid tests shall be in accordance with Regulatory Position C.2.1 of Regulatory Guide 1.9, Revision 3, where the number of tests and failures is determined on a per DG basis. (b) This test frequency shall be maintained until seven consecutive failure free starts from standby conditions and load and run tests have been performed. If, subsequent to the 7 failure free tests, 1 or more additional failures occur, such that there are again 4 or more failures in the last 25 tests, the testing interval shall again be reduced as noted above and maintained until 7 consecutive failure free tests have been performed. Watts Bar-Unit 1 3.8-17
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) 81. W/E05 G 2.4.3081 , Given the following: Unit 1 initially at 100% power with Shutdown Board 2A-A out of service.
- A Reactor Trip/Safety Injection occurs and due to complications entry into FR-H.1, "Loss of Secondary Heat Sink" was required.
Which ONE of the following identifies ... (1) a reason steam generator Wide Range levels are used instead of Narrow Range levels to determine if feed and bleed criteria is met, and (2) a condition that, if the condition existed, would require FR-H.1 to be exited prior to determining if bleed and feed was required? A'! (1) Because the Narrow Range Level Instruments would be off-scale low. At! (2) RCS pressure lower than any Intact steam generator pressures. B. (1) Because the Narrow Range Level Instruments would be off-scale low. (2) 6.9 kV Shutdown Board 1B-B trips due to electrical fault. C. (1) Because the Narrow Range Level instruments are NOT Post Accident Monitors. (2) RCS pressure lower than any Intact steam generator pressures. ) D. (1) Because the Narrow Range Level instruments are NOT Post Accident Monitors. (2) 6.9 kV Shutdown Board 1B-B trips due to electrical fault. Page 16
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Correct, the Narrow Range Level Instruments would be off-scale low and if RGS pressure was lower than any Intact steam generator pressure a RCS transition to the procedure in effect prior entering FR-H. 1 would be required. B. Incorrect, the Narrow Range Level Instruments would be off-scale low but if the 6.9 kV Shutdown Board 1B-B tripped no transition would be required. Plausible because the Narrow Range Level Instruments being off-scale low is correct and if the 6.9 kV Shutdown Board 1A-A had been out of service initially, a transition to EGA-D. ECA-O.ODwould would be required. G. C. Incorrect, the Narrow Range Level Instruments are post accident instruments (but they would be off-scale low) but if the RCS RGS pressure was lower than any Intact steam generator pressure a transition to the procedure in effect prior entering FR-H. 1 would be required. Plausible (i. e. Steam because all steam generator instrumentation is not post accident (i.e. Flow and Feed Flow) and if RCS RGS pressure was lower than any Intact steam generator pressure a transition to the procedure in effect prior entering FR-H. 1 is correct. D. Incorrect, the Narrow Range Level Instruments are post accident instruments (but they would be off-scale low) and if the 6.9 kV Shutdown ) Board 1B-B tripped no transition would be required. Plausible because all steam generator instrumentation is not post accident (i.e. Steam Flow and Feed Flow) and if the 6.9 kV Shutdown Board 1 A-A had been out of service 1A-A initially, a transition to EGA-D. ECA-O.OD would be required. ) Page 17
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 81 Tier: _1_ 1 Group 1 KIA: WIEOS IE05 G 204.3 2.4.3 Loss of Secondary Heat Sink Ability to identify post-accident instrumentation. Importance Rating: 3.7/3.9 3.7 I 3.9 10 CFR Part 55: 41.6 14S.4 I 45.4 10CFR55.43.b: S 5 KIA Match: Applicant must identify the steam generator narrow range level instruments which are post accident instrumentation. Identification of procedure transition makes the question SRO because the applicant must assess plant conditions and prescribe a procedure to mitigate, recover, or which to proceed. Technical
Reference:
FR-H.1, Loss of Secondary Heat Sink, Rev 17 MeR photograph Proposed references None to be provided: Learning Objective: 3-0T-FRH0001
- 6. Given a set of plant conditions, use procedures FR-H.1, .2, .3, .4, & .5 to correctly identify any
& .S required procedure transition.
3-0T-T/S0303 Obj. 1 Demonstrate the ability to extract specific information from the Technical Specification, and Technical Requirements, as they pertain to Instrumentation Systems. Question Source: New x Modified Bank Bank Question History: New question Comments Page 18
WBN LOSS OF SECONDARY HEAT SINK FR-H.1 Rev 17 I1Step I1 I1Action/Expected Response II 11 Response Not Obtained CAUTION
- If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
- If an Intact S/G is available, feed flow should NOT be reestablished to any faulted S/G.
- 1. CHECK if secondary heat sink is required:
- a. RCS pressure a. RETURN TO Instruction in effect.
greater than any Intact S/G pressure.
- b. RCS temperature
- b. PLACE RHR System in service greater than 375°F while continuing in this
[360°F ADV]. instruction.
- REFER TO SOI-74.01, Residual Heat Removal System.
WHEN adequate RHR shutdown cooling established, . THEN RETURN TO Instruction in effect.
- 2. ENSURE at least one IF at least one charging pump NOT charging pump RUNNING. RUNNING, THEN STOP all RCPs AND
** GO TO Cautions prior to Step 18 to initiate RCS bleed and feed.
30f 3 of 31
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 82. 033 AG2.2.40 082 Given the following:
Unit 1 at 3% reactor power being returned to service following a refueling outage. Intermediate Range Monitor, NI-35, fails low. Which ONE of the following is correct in regard to the actions that are required to comply with Tech Spec LCO 3.3.1 "RTS Instrumentation," LCO 3.3.3 "Post Accident Monitoring (PAM) Instrumentation" and the requirement for performance of a Risk assessment for Mode 1 entry? A. Reduce power to less than P-6 within 2 hours because Mode 1 can NOT be entered without performance of a Risk Assessment for the failed instrument. B:t Raise power to greater than 10% within 2 hours. Performance of a Risk Assessment is not required prior to entering Mode 1 to be in compliance with LCO 3.3.3. C. Place NI-35 HIGH LEVEL TRIP bistable in the trip position and reduce power to less than P-6 within 2 hours because Mode 1 can NOT be entered without a completed Risk Assessment for the failed instrument. D. Place NI-35 HIGH LEVEL TRIP bistable in the trip position and raise power to greater than 10% within 2 hours. A Risk Assessment is not required to allow entry into Mode 1 to be in compliance with LCO 3.3.3. Page 19
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, Power is not required to be reduced to less than P-6 due to the requirement for a Risk Assessment. Plausible because reducing power is an option to meet LCO 3.3.1 and LCO 3.3.3 but it is not required due to the need for a Risk Assessment. B. Correct, Tech Spec LCO 3.3.1 requires that within 2 hours the power level be raised to greater than P-10 (10%) or reduced to less than P-6. Raising power to get to a condition where the LCO is not applicable requires a Mode change to Mode 1. In accordance with LCO 3.0.4 a, the Mode change can be made if the LCO will permit continued operation in the Mode or other specified condition for an unlimited time. The IRM is not required above 10% power and therefore is not applicable. The same is true for Tech Spec 3.3.3. While there is a time limit for the failed IRM because it is a Post Accident Monitor, it is also not applicable when the power is greater than 10%. The provisions of LCO requirement for a risk assessment is applicable if the provisions 3.0.4.b are being relied on and this case 3. 0.4. a is applied for the mode change. C. Incorrect, NI-35 HIGH LEVEL TRIP bistable is not required to be placed in the trip position (and doing so would cause a reactor trip at the current power level) and reduce power to less than P-6 within 2 hours is not required due to Mode 1 entry being not being allowed without a completed Risk Assessment for the failed instrument. Plausible because Power Range instrument failure do place the high flux bistable to the trip position and during response to the IRM failure the Level Trip switch will be placed to the bypass position. Also plausible because a Risk Assessment is required for mode changes where the LCO does not provide for continued operation. D. Incorrect, NI-35 HIGH LEVEL TRIP bistable is not required to be placed in the trip position (and doing so would cause a reactor trip at the current power level) and entering Mode 1 without a completed Risk Assessment for the failed instrument is allowed and in compliance with LCO 3.3.3. Plausible because Power Range instrument failures do place the high flux bistable to the trip position and during response to the IRM failure the Level Trip switch will be placed to the bypass position and also because entering Mode 1 without a Risk Assessment is correct. Page 20
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 82 Tier: 1 _1_ Group _2_ 2 KIA: 033 AG2.2.40 Loss of Intermediate Range Nuclear Instrumentation Ability to apply Technical Specifications for a system. Importance Rating: 3.4 / 4.7 10 CFR Part 55: 41.10/43.2/43.5/45.3 10CFR55.43.b: 2 KIA Match: Applicant must apply the Technical Specification requirements of a failed Intermediate Range Nuclear Instrument during a startup. SRO because of the application of required actions on an LCO. Technical
Reference:
Technical Specification LCO 3.3.1 Technical Specification LCO 3.3.1 Bases (page 3.3-43) Technical Specification LCO 3.3.3 Technical Specification LCO 3.0.4 AOI-4, Nuclear Instrument Malfunction, Rev 28 ) Proposed references None to be provided: Learning Objective: 3-0T-SYS092A
- 23. Given a failure of one or more components of the Excore Instrumentation System, identify and/or explain the applicable technical specification LCO and required actions.
Question Source: New x X Modified Bank Bank Question History: New question Comments ) Page 21
AOI-4 WBN NUCLEAR INSTRUMENTATION MALFUNCTIONS Revision 28 Page 12 of 27 3.3 Intermediate Range Monitor (IRM) Failure ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. IF greater than P-6 and less then P-10 with BOTH IRM channels failed, THEN
. STOP positive reactivity changes.
NOTE Placing the affected channel in bypass will cause either window 64B or 65B to alarm.
- 2. PLACE failed channel LEVEL TRIP switch to BYPASS [1-M-13].
- 3. ENSURE 1-NR-92-145 recording an operable IRM.
- 4. REFER TO Tech Spec 3.3.1, Rx Trip System Instrumentation and 3.3.3, PAM Instruments.
- 5. NOTIFY Operations Duty Manager and Rx Engineering of any failed channel.
- 6. INITIATE repair of IRM.
AOI-4 WBN NUCLEAR INSTRUMENTATION MALFUNCTIONS Revision 28 Page 13 of 27 3.3 Intermediate Range Monitor (IRM) Failure (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. DO NOT CONTINUE UNTIL repairs are complete.
- 8. CHECK intermediate range monitor operable:
- CHECK INSTRUMENT POWER ON and CONTROL POWER ON lights LIT at NIS racks [1-M-13].
** CHECK NON-OPERATE light DARK [1-M-13].
** COMPARE channel output.
- 9. PLACE failed channel LEVEL TRIP switch in NORMAL
[1-M-13].
- 10. RETURN TO Instruction in effect.
- END OF SUBSECTION -
LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified specified conditions in the Applicability, except as provided in LCO 3.0.2. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LC03.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:
- a. MODE 3 within 7 hours;
- b. MODE 4 within 13 hours; and
- c. MODE 5 within 37 hours.
Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:
- a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;
((continued) continued) Watts Bar-Unit 1 3.0-1 Amendment 55
LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 b. After performance of a risk assessment addressing inoperable (continued) systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this in the individual Specifications, or Specification are stated in
- c. When an allowance is stated in the individual value, parameter, or other Specification.
This Specification shall not prevent changes in MODES or other specified that are required to comply with ACTIONS or that conditions in the Applicability that are part of a shutdown of the unit. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. ) LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions 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, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. (continued) (continued) ) Watts Bar-Unit 1 3.0-2 Amendment 55
LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.6 LC03.0.6 When a support system's Required Action directs a supported system to be (continued) (continued) 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. LCO 3.0.7 LC03.0.7 Test Exception LCOs 3.1.9 and 3.1.10 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. ) Watts Bar-Unit 1 3.0-3
RTS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.3.1 The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1-1. ACTIONS
NOTE---------------------------------------------------------------
NOT E---------------------------------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME A. A One or more Functions with A.1 A1 Enter the Condition referenced Immediately one or more required channels in Table 3.3.1-1 for the inoperable. channel(s). B. One Manual Reactor Trip B.1 Restore channel to 48 hours channel inoperable. OPERABLE status. OR B.2.1 Be in MODE 3. 54 hours AND B.2.20pen reactor trip breakers 55 hours (RTBs). ((continued) continued) Watts Bar-Unit 1 3.3-1
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One channel or train C.1 Restore channel or train to 48 hours inoperable. OPERABLE status. OR C.2 Open RTBs. 49 hours D. One Power Range Neutron ------------------------ NOT E------------------- Flux - High channel The inoperable channel may be inoperable. bypassed for up to 12 hours for surveillance testing and setpoint adjustment of other channels.
----------------------------------------------------
0.1.1 Place channel in trip. 72 hours AND 0.1.2 Reduce THERMAL POWER 78 hours to::; 75% RTP. to:o;; OR 0.2.1 Place channel in 72 hours trip. AND
----------------------NOTE---------------------
Only required to be performed when the Power Range Neutron Flux input to QPTR is inoperable.
----------------------------------------------------
0.2.2 Perform SR 3.2.4.2. Once per 12 hours OR 0.3 Be in MODE 3. 78 hours (continued) Watts Bar-Unit 1 3.3-2 Amendment 68
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One channel inoperable. ------------------------NOTE------------------- The inoperable channel may be bypassed for up to 12 hours for surveillance testing of other channels.
----------------------------------------------------
E.1 Place channel in trip. 72 hours OR E.2 Be in MODE 3. 78 hours F. THERMAL POWER> P-6 F.1 Reduce THERMAL POWER 2 hours and < P-10, one Intermediate to < P-6. Range Neutron Flux channel inoperable. OR F.2 Increase THERMAL 2 hours POWER to> to > P-10. P-1 O. G. THERMAL POWER> P-6 and G.1 Suspend operations Immediately
< P-10, two Intermediate involving positive reactivity Range Neutron Flux channels additions.
inoperable. AND G.2 Reduce THERMAL POWER 2 hours to < P-6. H. THERMAL POWER < P-6, H.1 Restore channel(s) to Prior to increasing one or two Intermediate OPERABLE status. THERMAL POWER Range Neutron Flux channels to> P-6 inoperable. (continued) Watts Bar-Unit 1 3.3-3 Amendment 68
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME I. One Source Range Neutron 1.1 Suspend operations involving Immediately Flux channel inoperable. positive reactivity additions. J. Two Source Range Neutron J.1 Open RTBs. Immediately Flux channels inoperable. K. One Source Range Neutron K.1 . Restore channel to 48 hours Flux channel inoperable. OPERABLE status. OR K.2 Open RTBs. 49 hours L. Required Source Range L.1 Suspend operations involving Immediately Neutron Flux channel positive reactivity additions. inoperable. AND 1 L.2 un borated water source Close unborated isolation valves. 1 hour AND L.3 Perform SR 3.1.1.1. 1 hour AND Once per 12 hours thereafter ((continued) continued) Watts Bar-Unit 1 3.3-4 )
RTS Instrumentation 3.3.1 ) ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME M. One channel ----------------------NOT
----------------------N OTE E---------------------
inoperable. The inoperable channel may be bypassed for up to 12 hours for surveillance testing of other channels.
----------------------------------------------------
------ ...--------------------------------------------
M.1 Place channel in trip. 72 hours OR M.2 Reduce THERMAL POWER to 78 hours
< P-7.
N. One Reactor Coolant Flow-- ---------------------NOT
---------------------NO TEE----------------------
Low channel inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
----------------------------------------------------
N.1 Place channel in trip. 72 hours ) OR N.2 Reduce THERMAL POWER to 78 hours
. < P-7.
- _..
(continued) (continued) Watts Bar-Unit 1 3.3-5 Amendment 68
\
I I
RTS Instrumentation 3.3.1 ) ACTIONS (continued) CONDITION . REQUIRED ACTION COMPLETION TIME O. One Low Fluid Oil Pressure -----------------------NOT
---------------------N OT E--------------------
Turbine Trip channel The inoperable channel may be inoperable. bypassed for up to 12 hours for surveillance testing of other channels.
----------------------------------------------------
---------------------------------------------------
0.1 Place channel in trip. 72 hours OR 0.2 Reduce THERMAL POWER to 76 hours
< P-9.
P. One train inoperable. ----------------------NOT
---------------------N OTE E---------------------
One train may be bypassed for up to 4 hours for surveillance testing provided the other train is OPERABLE.
----------------------------------------------------
P.1 Restore train to OPERABLE 24 hours status. OR P.2 Be in MODE 3. 30 hours (continued) Watts Bar-Unit 1 3.3-6 Amendment 68
\1 /
RTS Instrumentation 3.3.1 ) ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME o. Q. One RTB train inoperable. -~--------------------N 0 T E------------------
-----------------------N -------------------
One train may be bypassed for up to 4 hours for surveillance testing, provided the other train is OPERABLE.
----------------------------------------------------
----------------------------------------------------
Q.1 0.1 Restore train to OPERABLE 24 hours status. OR 0.2 Be in MODE 3. Q.2 30 hours R R. One channel R1 R.1 Verify interlock is 1 hour inoperable. in required state for existing unit conditions. OR R.2 Be in MODE 3. R2 7 hours S. One channel S.1 Verify interlock is 1 hour inoperable. in required state for existing unit conditions. OR S.2 Be in MODE 2. 7 hours (continued) (continued) Watts Bar-Unit 1 3.3-7 Amendment 68
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME T. One trip mechanism T.1 Restore inoperable trip 48 hours inoperable for one mechanism to OPERABLE RTB. status. OR c T.2.1 Be in MODE 3. 54 hours AND T.2.2 Open RTB. 55 hours U. One Steam Generator Water -----------------------N 0 T E-------------------- Level Low-Low channel One channel may be bypassed inoperable. for up to 12 hours for surveillance testing.
---------------------------------------------------
U.1.1 Place channel in trip. 72 hours AND U.1.2 For the affected protection set, 72 hours set the Trip Time Delay (TS) to match the Trip Time Delay (T M)* M)' OR U.2 Be in MODE 3. 78 hours (continued) Watts Bar-Unit 1 3.3-8 Amendment 68
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME V. L'!T One Vessel b. T channel --------------------~NOTE----------------------
---------------------NOTE ----------------------
inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
----------------------------------------------------
V.1 Set the Trip Time Delay 72 hours threshold power level for (TS) and (T M) to 0% power. OR V.2 Be in MODE 3. 78 hours W. One channel --------------------N 0 T E ----------------------- inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
---------------------------------------------------
W.1 Place channel in trip. 72 hours OR W.2 Be in MODE 3. 78 hours X. One channel --------------------NOTE----------------------- inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
----------------------------------------------------
X.1 Place channel in trip. 72 hours OR X.2 Reduce THERMAL POWER to 78 hours
< P-7.
( continued) Watts Bar-Unit 1 3.3-9 Amendment 68
RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME Y. One, two or three Turbine Y.1 Place channel(s) in trip. 72 hours Stop Valve Closure channels inoperable. OR Y.2 Reduce THERMAL POWER to 76 hours
< P-9.
z. Z. Two RTS Trains inoperable Z.1 Enter LCO 3.0.3. Immediately SURVEILLANCE REQUIREMENTS
---------------------------------------------------------------------N OTE ---------------------------------------------------------- ------------------------------------------------------------------N ------------------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function. ) SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.2 ----------------------------------N 0 TE S-------------------------------
----------------------------------NOTE
- 1. Adjust NIS channel if absolute difference is > 2%.
- 2. Required to be performed within 12 hours after TH THERMAL ERMAL POWER is ~ 15% RTP.
Compare results of calorimetric heat balance calculation to Nuclear 24 hours Instrumentation System (NIS) channel output. (continued) Watts Bar-Unit 1 3.3-10 Amendment 68
)
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.3 -------------------------------- NOT ES-----------------------------------
--------------------------------NOTE
- 1. Adjust NIS channel if absolute difference is;;:: is :::: 3%.
- 2. Required to be performed within 96 hours after THERMAL POWER is;;:: 15% RTP.
is ::::15% Compare results of the incore detector measurements to 31 effective full NIS AFD. power days (EFPD) SR 3.3.1.4 -----------------------------------N 0 T E --------------------------------- This Surveillance must be performed on the reactor trip
. bypass breaker prior to placing the bypass breaker in service.
Perform T ADOT. TADOT. 62 days on a STAGGERED TEST BASIS SR 3.3.1.5 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.1.6 ---------------------------------N OTE
---------------------------------NOT E----------------------------------
Required to be performed within 6 days after THERMAL POWER is::::is ;;:: 50% RTP. Calibrate excore channels to agree 92EFPD with incore detector measurements. (continued) Watts Bar-Unit 1 3.3-11 Amendment 68
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.7 ---------------------------------N 0 TE T E------------------------------
---------------------------------
For Functions 2 and 3 (Power Range Instrumentation), this Surveillance shall include verification that P-1 0 is in the required state for existing unit interlock P-10 conditions. Perform COT. 184 days SR 3;3.1.8 3.3.1.8 --------------------------------N 0 TE S----------------------------
---------------------------------NOTES------------------------------- ---------NOT E--------
---------NOTE--------
- 1. Not required to be performed for Source Range Only required when not instrumentation prior to entering MODE 3 from MODE 2 performed until4 until 4 hours after entry into MODE 3. within previous 31 days
- 2. This Surveillance shall include verification that interlock P-6 is in the required state for existing unit conditions.
Prior to Perform COT. reactor startup ) AND Four hours after reducing power below P-10 for intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND Every 31 days thereafter (continued) (continued) Watts Bar-Unit 1 3.3-12 Amendment 68 )
RTS Instrumentation 3.3.1 ') SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.9 -------------------------------N 0 TE --------------------------------
-------------------------------NOTE Verification of setpoint is not required.
Perform TADOT. 92 days SR 3.3.1.1 0 ------------------------------NOTE--------------------------------- This Surveillance shall include verification that the time constants are adjusted to the prescribed values. Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.11 -------------------------------NOTE--------------------------------- Neutron detectors are excluded from CHANNEL CALIBRATION. CALI BRATION. Perform CHANNEL CALIBRATION. 18 months
)
SR 3.3.1.12 Perform COT. 18 months SR 3.3.1.13 -------------------------------NOTE------------------------------------- Verification of setpoint is not required. Perform TADOT. 18 months (continued) Watts Bar-Unit 1 3.3-13 )
RTS Instrumentation 3.3.1 ) SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.14 -------------------------------------N 0 TE ------------------------------
------------------------------------NOTE-----------------------------
Verification of setpoint is not required. Perform T TADOT. ADOT. Prior to exceeding the P-9 interlock whenever the unit has been in Mode 3, if not performed within the previous 31 days SR 3.3.1.15 --------------------------------NOTE--------------------------------
-----------------------------------NOTE---------------------------------
Neutron detectors are excluded from response time testing. Verify RTS RESPONSE TIME is within limits. 18 months on a
. STAGGERED TEST BASIS
) Watts Bar-Unit 1 3.3-14 Amendment 68 )
RTS Instrumentation 3.3.1 of9) Table 3.3.1-1 (page 1 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT CONDITIONS
- 1. Manual Reactor 1,2 2 8B SR 3.3.1.13 SR3.3.1.13 NA NA Trip 3(a), 4(a), 5(a) 2 C SR3.3.1.13 NA NA
- 2. Power Range Neutron Flux
- a. High 1,2 4 D SR3.3.1.1 SR 3.3.1.1 :s:
- S 111.4% 109% RTP SR 3.3.1.2 RTP SR 3.3.1.7 SR3.3.1.7 SR 3.3.1.11 SR 3.3.1.15
- b. Low 1(b), 2 4 E SR3.3.1.1 SR 3.3.1.1 :S 27.4% RTP
- s:27.4% 25% RTP SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.15
- 3. Power Range Neutron Flux Rate
- a. High Positive 1,2 4 E SR 3.3.1.7 :s:
- S 6.3% RTP 5% RTP with Rate SR 3.3.1.11 with time time constant constant
;::: 2 sec
?: ;::: 2 sec
?:
- b. High Negative Rate - Deleted
- 4. Intermediate Range 1(b),2(C) 2 F, G F,G SR 3.3.1.1 :S 40% RTP
- S:40% 25% RTP Neutron Flux SR 3.3.1.8 SR 3.3.1.11 2(d) 2 H SR 3.3.1.1 :S:40%
SR3.3.1.1 :S 40% RTP 25% RTP SR 3.3.1.8 SR 3.3.1.11 SR3.3.1.11 (continued) (a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal. (b) Below the P-10 (Power Range Neutron Flux) interlocks. (c) Above the P-6 (Intermediate Range Neutron Flux) interlocks. (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks. Watts Bar-Unit 1 3.3-15 Amendment 18
RTS Instrumentation 3.3.1 'J Table 3.3.1-1 (page 2 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS .VALUE SETPOINT CONDITIONS
- 5. Source Range 2(d) 2 I,.J I, J SR 3.3.1.1 ::;; 1.5 E5
- S: 1.0 E5 cps Neutron Flux SR 3.3.1.8 cps SR 3.3.1.11 3(a), 4(a), 5(a) 2 J, K SR 3.3.1.1 SR3.3.1.1 ::;;
- S: 1.5 E5 1.0 E5 cps SR 3.3.1.8 cps SR 3.3.1.11 SR 3.3.1.15 SR3.3.1.15 3(e), 4(e), 5(e) 1 L SR 3.3.1.1 N/A N/A SR 3.3.1.11
- 6. Overtemperature il
!1T 1,2 4 W SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 1 Note 1 SR 3.3.1.6 (Page 3.3- (Page 3.3-
- 21) 21)
SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR3.3.1.15
- 7. Overpower il!1T 1,2 4 W SR 3.3.1.1 Refer to Refer to SR 3.3.1.7 Note 2 Note 2 SR 3.3.1.10 (Page 3.3- (Page 3.3-
) SR 3.3.1.15
- 22) 22)
- 8. Pressurizer Pressure
- a. Low 1(f) 1(f) 4 X X SR 3.3.1.1 ;::
;::: 1964.8 1970 psig SR 3.3.1.7 SR3.3.1.7 psig SR 3.3.1.10 SR3.3.1.10 SR 3.3.1.15 SR3.3.1.15
- b. High 1,2 4 W SR 3.3.1.1 ::;; 2390.2
- S: 2385 psig SR 3.3.1.7 SR3.3.1.7 psig SR 3.3.1.10 SR 3.3.1.15 SR3.3.1.15 (continued)
(a) With RTBs closed and Rod Control System capable of rod withdrawal. (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks. (e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide indication. (f) Above the P-7 (Low Power Reactor Trips Block) interlock. Watts Bar-Unit 1 3.3-16 Amendment No. 52
)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT CONDITIONS
- 9. Pressurizer Water 1(I) 3 X SR 3.3.1.1 ::;92.7%
~ 92.7% 92% span Level-High SR 3.3.1.7 span SR 3.3.1.10 10 Reactor Coolant 1(I)
(f) 3 per N SR 3.3.1.1 <:2': 89.7% 90% flow Flow-Low loop SR3.3.1.7 SR 3.3.1.7 Flow SR 3.3.1.10 SR 3.3.1.15 11 Undervoltage 1(1) 1(f) 1 per bus M SR 3.3.1.9 <:2': 4734 V 4830 V RCPs SR 3.3.1.10 SR 3.3.1.15 12 Underfrequency 1(I) (f) 1 per bus M SR 3.3.1.9 <:2': 56.9 Hz 57.5 Hz RCPs SR 3.3.1.10 SR 3.3.1.15 (continued) (f) Above the P-7 (Low Power Reactor Trips Block) interlock. Watts Bar-Unit 1 3.3-17 Amendment 47, 68
RTS Instrumentation 3.3.1 ) (page4of9) Table 3.3.1-1 (page 4 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT CONDITIONS
- 13. SG Water Level -- 1,2 3/SG U SR 3.3.1.1 ~
2: 16.4% of 17% of Low-low SR3.3.1.7 SR 3.3.1.7 narrow narrow SR 3.3.1.10 range span range span Coincident with: SR 3.3.1.15 SR3.3.1.15 a) Vessel LlT t..T 1,2 3 V SR 3.3.1.7 Vessel LlT Vess.el t..T Vessel LlT t..T Equivalent SR 3.3.1.10 variable input variable input to power :';52,6% s: 52.6% RTP 50% RTP
- ,;50%
s: 50% RTP With a time :';1.01Ts s:1.01Ts Ts (Refer delay (Ts) (Refer to to Note 3, if one steam ifone Note 3, Page 3.3-23) generator is Page 3.3-23) affected or A time delay :,; s: 1.01 Tm Tm (Refer (Tm) if two (Refer to to Note 3, or more steam Note 3, Page 3.3-23) generators are Page 3.3-23) affected
) OR
/1 b) t..T Vessel LlT 1,2 3 V SR 3.3.1.7 Vessel LlTt..T Vessel LlT t..T Equivalent to SR 3.3.1.10 variable input variable input power> 50% s: 52.6% RTP
- ';52.6% 50% RTP RTP with no time delay (Ts and Tm = =
0)
- 14. Turbine Trip
- a. Low Fluid 1(i) 3 0 SR 3.3.1.10 ~
2: 43 psig 45 psig Oil pressure SR 3.3.1.14
- b. Turbine Stop 1(i)
(il 4 Y SR 3.3.1.10 ~ 2: 1% open 1% open Valve Closure SR 3.3.1.14 (continued) (i) Above the P-9 (Power Range Neutron Flux) interlock. Watts Bar-Unit 1 3.3-18 Amendment 7 Amendment?
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT CONDITIONS
- 15. Safety Injection (SI) 1,2 1.2 2 trains P SR 3.3.1.13 NA NA Input from Engineered Safety Feature Actuation System (ESFAS)
- 16. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, Flux. P-6 (1) Enable 2(d) 2 R SR 3.3.1.11 SR3.3.1.11 NA 1.66E-04%
Manual Block SR 3.3.1.12 RTP of SR Trip (2) Auto Reset 2(d) 2 R SR3.3.1.11 2: 7.65E-5%
?: 0.47E-4%
(Unblock SR 3.3.1.12 RTP RTP below Manual Block setpoint of SR Trip)
- b. Low Power 1 per S SR3.3.1.11 SR 3.3.1.11 NA NA Reactor Trips train SR 3.3.1.12 Block, Block. P-7
- c. Power Range 4 S SR 3.3.1.11 ~
- 50.4% 48% RTP Flux. P-8 Neutron Flux, SR3.3.1.12 SR 3.3.1.12 RTP
- d. Power Range 4 S SR 3.3.1.11 ~
- 52.4% 50% RTP Neutron Flux, P-9 SR 3.3.1.12 RTP
- e. Power Range 1,2 1.2 4 R SR3.3.1.11 SR 3.3.1.11 2: 7.6% RTP
?: 10% RTP Flux. P-10 Neutron Flux, SR 3.3.1.12 and
~
- 12.4% RTP
- f. Turbine Impulse 2 S SR3.3.1.10 ~
- 12.4%
12.4% 10% Pressure, P-13 SR 3.3.1.12 full- power full-power pressure pressure (continued) (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks. Watts Bar-Unit 1 3.3-19
RTS Instrumentation 3.3.1 ) Table 3.3.1-1 (page 60f9) 6 of 9) Reactor Trip System Instrumentation APPLICABLE MODES NOMINAL OR OTHER REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT CONDITIONS
- 17. Reactor Trip 1,2 2 trains Q SR 3.3.1.4 NA NA Breakers (j)
U) 3(a), 4(a) , Sea) 2 trains C SR 3.3.1.4 NA NA
- 18. Reactor Trip 1,2 1 each T SR 3.3.1.4 SR3.3.1.4 NA NA Breaker per RTB Undervoltage and Shunt Trip 3(a), 4(a) , Sea) 1 each C SR 3.3.1.4 NA NA Mechanisms per RTB
- 19. Automatic Trip 1,2 2 trains P SR 3.3.1.5 NA NA Logic 3(a), 4(a) , Sea) 2 trains C SR 3.3.1.5 NA NA (a) With RTBs closed and Rod Control System capable of rod withdrawal.
U)
- 0) Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.
) Watts Bar-Unit 1 3.3-20
RTS Instrumentation B 3.3.1 Bases ACTIONS F.1 and F.2 (continued) (continued) Condition F applies to the Intermediate Range Neutron Flux trip when THERMAL POWER is above the P-6 setpoint and below the P-10 setpoint and one channel is inoperable. Above the P-6 setpoint and below the P-10 setpoint, the NIS intermediate range detector performs the monitoring Functions. If THERMAL POWER is greater than the P-6 setpoint but less than the P-10 setpoint, 2 hours is allowed to reduce THERMAL POWER below the P-6 setpoint or increase THERMAL POWER above the P-10 setpoint. The NIS Intermediate Range Neutron Flux channels must be OPERABLE when the power level is above the capability of the source range, P-6, and below the capability capability of the power range, P-10. If THERMAL POWER is greater than the P-10 setpoint, the NIS power range detectors perform the monitoring and protection functions and the intermediate range is not required. The Completion Times allow for a slow and controlled power adjustment above P-10 or below P-6 and take into account the redundant capability afforded by the redundant OPERABLE channel, and the low probability of its failure during this period. This action does not require the inoperable channel to be tripped because the Function uses one-out-of-two logic. Tripping one channel would trip the reactor. Thus, the Required Actions specified in this Condition are only applicable when channel failure cfoes does not result in reactor trip. G.1 and G.2 Condition G applies to two inoperable Intermediate Range Neutron Flux trip channels in MODE 2 when THERMAL POWER is above the P-6 setpoint and below the P-10 setpoint. Required Required Actions specified in this Condition are only applicable when channel failures do not result in reactor trip. Above the P-6 setpoint and below the P-10 setpoint, the NIS intermediate range detector performs the monitoring Functions. With no intermediate range channels OPERABLE, the Required Actions are to suspend operations involving positive reactivity additions immediately. This will preclude any power level increase since there are no OPERABLE Intermediate Range Neutron Flux channels. The operator must also reduce THERMAL POWER below the P-6 setpoint within two hours. Below P-6, the Source Range Neutron Flux channels will be able to monitor the core power level. The Completion Time of 2 hours will allow a slow and controlled power reduction to less than the P-6 setpoint and (continued) Watts Bar-Unit 1 B 3.3-43
PAM Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.3-1. ACTIONS
NOTE---------------------------------------------------------
NOTE ---------------------------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME A. ----------N 0 T E-------------- A.1 Restore required channel to 30 days Not applicable to Functions OPERABLE status. 3,4, 14, and 16.
--------------------------------
One or more Functions with one required channel inoperable. B. Required Action and B.1 Initiate action in accordance Immediately associated Completion with Specification 5.9.8. Time of Condition A not met. (continued) Watts Bar-Unit 1 3.3-41 Amendment 72
PAM Instrumentation 3.3.3 ) _ _._- -_ ACTIONS
.. .. lcontinued)
.. _... _._-
CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions with C.1 Restore one channel to 7 days
-two required channels OPERABLE status.
inoperable. OR Functions 3, 4,14,4, 14, and 16 with one required channel inoperable. D. Required Action and D.1 Enter the Condition referenced Immediately associated Completion Table 3.3.3-1 for the in Table Time of Condition C not channel. met. E. As required by Required E.1 Be in MODE 3. 6 hours Action D.1 and referenced in Table 3.3.3-1. AND ) E.2 Be in MODE 4. 12 hours F. As required by Required F.1 Initiate action in accordance Immediately Action D.1 and referenced with Specification 5.9.8. in Table 3.3.3-1. Watts Bar-Unit 1 3.3-42 Amendment 72 )
PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS
NOTE-------------------------------------------------------------
N 0 T E -------------------------------------------------------------
SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1. SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized. SR 3.3.3.2 ---------------------------N 0 T ES-------------------------------
- 1. Neutron detectors are excluded from CHANNEL CALIBRATION.
- 2. Not applicable to Functions 11 and 16.
Perform CHANNEL CALIBRATION. 18 months SR 3.3.3.3 --------------------------N 0 T ES-------------------------------
--------------------------NOTES-------------------------------
- 1. Verification of relay setpoints not required.
- 2. Only applicable to Functions 11 and 16.
Perform TADOT. 18 months Watts Bar-Unit 1 3.3-43 Amendment 72
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2) Post Accident Monitoring Instrumentation APPLICABLE MODES CONDITION OR OTHER REFERENCED FROM SPECIFIED REQUIRED REQUIRED FUNCTION CONDITIONS CHANNELSrTRAINS CHANNELSITRAINS ACTION 0.1
- 1. Intermediate Range Neutron 1(a), 2(b), 3 2 E Flux(g)
- 2. Source Range Neutron Flux 2(C),3 2 E
- 3. Reactor Coolant System (RCS) Hot 1,2,3 1 per loop E Leg Temperature (T-Hot)
- 4. RCS Cold Leg Temperature (T- 1,2,3 1 per loop E Cold)
- 5. RCS Pressure (Wide Range) 1,2,3 3 E
- 6. Reactor Vessel Water Level (I) (9) 1,2,3 2 F
- 7. Containment Sump Water Level 1,2,3 2 E (Wide Range)
- 8. Containment Lower Compo 1,2,3 2 E Atm. Temperature
- 9. Containment Pressure (Wide 1,2,3 2 E Range) (g)
- 10. Containment Pressure (Narrow 1,2,3 4 E Range)
- 11. Containment Isolation Valve 1,2,3 2 per penetration flow E Position (g) path (d)(i)
- 12. Containment Radiation (High 1,2,3 2 upper containment F Range) 2 lower containment
- 13. RCS Pressurizer Level 1,2,3 3 E
- 14. Steam Generator (SG) Water Level 1,2,3 1/SG E Range)(g)
(Wide Range )(9)
- 15. Steam Generator Water Level 1,2,3 3/SG E (Narrow Range)
- 16. AFW Valve Status m Q) 1,2,3 1 per valve E
- 17. Core Exit Temperature- 1,2,3 2 (e) E Quadrant 1(f)(I)
(continued) Watts Bar-Unit 1 3.3-44 Amendment 72
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2) Post Accident Monitoring Instrumentation APPLICABLE MODES CONDITION OR OTHER REFERENCED FROM SPECIFIED REQUIRED REQUIRED FUNCTION CONDITIONS CHANNELSITRAINS ACTION D.1
- 18. Core Exit Temperature- 1,2,3 2 (e) E Quadrant 2(1)
- 19. Core Exit Temperature- 1,2,3 2 (e) E Quadrant 3(1)
- 20. Core Exit Temperature- 1,2,3 2 (e) E Quadrant 4(1)
- 21. Auxiliary Feedwater Flow 1,2,3 2/SG
*2/SG E
- 22. Reactor Coolant System 1,2,3 2 E Subcooling Margin Monitor (h)
- 23. Refueling Water Storage Tank 1,2,3 2 E Water Level
- 24. Steam Generator Pressure 1,2,3 2/SG E
- 25. Auxiliary Building Passive Sump 1,2,3 2 E Level 0)
(a) Below the P-10 (Power Range Neutron Flux) interlocks. (b) Above the P-6 (Intermediate Range Neutron Flux) interlocks. (c) Below the P-6 (Intermediate Range Neutron Flux) interlocks. (d) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, pressure relief valve, or check valve with flow through the valve secured. (e) A channel consists of two core exit thermocouples (CETs). (f) The ICCM provides these functions on a plasma display. (g) Variables. Regulatory Guide 1.97, non-Type A, Category 1 Variables. (h) This function is displayed on the ICCM plasma display and digital panel meters. (i) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel. 0) U) Watts Bar specific (not required by Regulatory Guide 1.97) non-Type A Category 1 variable. Watts Bar-Unit 1 3.3-45 Amendment 72 )
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 83. 067 AA2.12 083 83.067 Given the following:
- Unit 1 was operating at 100% power.
EI 713 pipe chase near the BIT. A fire developed in the EI713
- The electrical cables for BIT outlet valves 1-FCV 63-25 and 1-FCV-63-26 have been severely damaged.
- ShiftManager The Shift Manager declared an Appendix R fire and the crew entered AOI-30.2, "Fire Safe Shutdown."
- The operating crew determined the fire area to be A28 and are now performing AOI-30.2 C.59, "Fire Safe Shutdown Room 676-A 1, 676-A 16, 692-A8,713-A28."
692-A8, 713-A28." Which ONE of the following identifies .. (1) the action required for the unit, and (2) the impact of the BIT outlet valves being damaged due to the fire? A. (1) Unit must be tripped in accordance with AOI-30.2 C.59. (2) Require reopening the charging flow isolation valves to establish high pressure injection if a safety injection was actuated. B~ (1) Unit must be tripped in accordance with AOI-30.2 C.59. 1 J (2) No impact because the BIT outlet valves are required to remain closed during an Appendix R fire. C. (1) Unit must be shutdown in accordance with AOI-39, "Rapid Load Reduction" due to entry into Technical Specification LCO 3.0.3. (2) Require reopening the charging flow isolation valves to establish high pressure injection if a safety injection was actuated. D. (1) Unit must be shutdown in accordance with AOI-39, "Rapid Load Reduction" due to entry into Technical Specification LCO 3.0.3. (2) No impact because the BIT outlet valves are required to remain closed during an Appendix R fire. J
. Page 22
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ') DISTRACTOR ANAL YSIS: A. Incorrect, the reactor would be tripped but while BIT flow could not be established, the realignment of charging to establish high pressure injection would not be required. Plausible because tripping the unit is correct and the BIT outlet valves do normally open on a Safety Injection and the CCPs could inject through the charging line if the isolation valves were reopened. B. Correct, the AOI requires the reactor to unit tripped and there is no impact because the BIT outlet valves are required to remain closed during an Appendix R fire. C. Incorrect, the unit would not be shutdown in accordance with A AOI-39 01-39 (reactor trip required) and while BIT flow could not be established, AOI-30.2 requires the BIT outlet valves to remain closed to better ensure protection for the RCP seals, thus realignment of charging to establish high pressure injection would not be required. Plausible because with both BIT outlet valves inoperable, TIS 3.0.3 would require a shutdown, the BIT outlet valves do normally open on a Safety Injection and the CCPs could inject through the charging line if the isolation valves were reopened. D. Incorrect, the unit would not be shutdown in accordance with A 01-39 (reactor trip required) and there is no impact because the BIT outlet valves are required to remain closed during an Appendix A fire. Plausible because with both BIT outlet valves inoperable, TIS 3.0.3 would require a shutdown and the BIT outlet valves ) being required to remain closed during an Appendix A fire is correct. Page 23
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 83 Tier: _1_ Group 2 KIA: 067 AA2.12 Plant Fire On Site Ability to determine and interpret the following as they apply to the Plant Fire on Site: Location of vital equipment within fire zone. Importance Rating: 2.9/3.9 10 CFR Part 55: 43.5 / 45.13 43.5/45.13 10CFR55.43.b: 5 KIA Match: Applicant is required to interpret the impact of the damage to the BIT outlet valves as their loss applies to the implementation of the Fire Safe Shutdown procedure. SRO because of the selection of appropriate procedure to remove the unit from service. Technical
Reference:
AOI-30.2, Fire Safe Shutdown. Rev 0000 AOI-30.2 C.59, "Fire Safe Shutdown Room 676-A 1, 676-A16, 692-A8, 713-A28, Rev 0000 Proposed references None to be provided: Learning Objective: 3-0T -AO 13000 3-0T-AOI3000
- 9. List the assumptions (3) made for analysis as described in AOI-30.2 with respect to an Appendix R fire.
Question Source: New X Modified Bank Bank Question History: New question Comments Page 24
WBN Fire Safe Shutdown AOI-30.2 C.S9 Unit 0 Room 676-A1, 676-A16, 692-A8, 713- Rev. 0000 A28 Page 4 of21 of 21 1.0 MAIN CONTROL ROOM OPERATOR ACTIONS CAUTION Handswitches are positioned to to ensure equipment isin is in its required safe shutdown configuration. This ensures the proper signal is maintained on the correct device. The equipment may appear to be in the condition called for in the procedure step, but the indicated condition may be showing a false indication due to faulted control circuits or instrumentation. NOTE Based on the location and severity of the fire, Operations personnel must determine which Train of equipment is reliable for plant cooldown. Some action steps (such as VCT isolation) require this information be passed to AUOs outside the MCR for use in performing assigned manual actions.
- Alternate means (flow, pressure, amps, etc.) may be needed for checking component or system status.
- Table 1 is a listing of instrumentation which should remain available during a fire in this location.
- Table 2 is a summary of AUO assignments.
- This procedure is applicable for a fire in any of the following rooms:
o 676-A1 o 676-A16 o 692-A8 o 713-A28 [1] ENSURE Reactor trip:
- Reactor trip and bypass breakers OPEN.
- RPls at bottom of scale.
- Neutron flux dropping.
[2] ENSURE turbine trip and steam valves CLOSED: A. Turbine stop valves CLOSED. B. MSIVs CLOSED. C. MSIV bypasses CLOSED.
WBN Fire Safe Shutdown AOI-30.2 Unit 0 Rev. 0000 Page 9 of19 of 19 ) 5.1 Background and Assumptions (continued) G. The Safe Shutdown Logic Diagram also shows the paths available to provide the safety functions for the safe shutdown conditions described in 5.1 E and 5.1F. Paragraphs 5.1E 5.1 F. For each safety function, the equipment required to accomplish the safety function has been divided into "Keys" which represent the safety groups of functionally-related equipment necessary to accomplish the-safety function. These are also represented on the Safe Shutdown Logic Diagram. H. At least one path of equipment or components needed to achieve safe shutdown is required to remain operable or capable of being operated for 72 hours following a postulated fire (to establish long-term heat removal via RHR). I. No other accident is assumed to occur concurrently with a fire therefore, a valid SI signal is assumed not to be present at the time of an Appendix R fire. However, spurious SI signal actuation could occur as a result of the effects of However,' the fire. Since many of the actions in the Safe Shutdown analysis require components to be in positions opposite that required by SI, a spurious SI would require these components to be repositioned. For example, the BIT outlet valves are required to be closed for an Appendix R fire. The purpose of this is to: .
- 1. Guarantee flow to the RCP seal line for boron injection
- 2. Prevent pressurizer overfill (no RCS break is assumed and normal charging/letdown may not be available due to fire or loss of air).
charginglletdown
- 3. Prevent damage to the charging pump due to fast drawdown of the VCT (automatic circuit for the swap over to RWST on low VCT level is not (automatiC guaranteed).
J. In general it is assumed that shutdown of the plant will be performed from the Main Control Room for a postulated fire elsewhere in the plant. For shutdown from outside the Control Building, it is essential that, functionally, the same equipment and instrumentation be available from the Aux Control Room or remote stations or otherwise be justified. Loss of offsite power is assumed concurrently with MCR fires. K. The possibility of shutdown and cooldown of the plant from the auxiliary control room was considered in the manual actions of the approved Site Engineering calculation. L. Where the spurious operation of a component could defeat the required system safety function, manual actions are taken to address the effects of spurious component operation. Components identified as those which could prevent a safe shutdown should they spurious operate are those that:
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 \ 84. WE06 EG2.1.23 084
) Given the following conditions:
- A LOCA is in progress with all RCPs stopped.
- The crew is performing E-1, "Loss of Reactor or Secondary Coolant."
- The STA reports the following:
- Core exit TCs: 734 of 734°F
- RCS Pressure 1385 psig
- RVLlS: 50%
Which ONE of the following describes the procedure required to be entered and the restrictions on starting the RCPs during the performance of the procedure? Procedure RCP start restrictions A'I FR-C.2, "Response to A"! Even with support equipment available Degraded Core Cooling"; an RCP would NOT be started. B. FR-C.2, "Response to RCP would be started without all Degraded Core Cooling"; normal support conditions available. C. FR-C.1 "Response to RCP would NOT be started without Inadequate Core Cooling"; support equipment available. ) D... FR-C.1 "Response to D RCP would be started without all support Inadequate Core Cooling"; conditions available. Page 25
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR ANAL YSIS: A. A. Correct, with incore temperature less than 1200 0 12000FF but greater than 72rF, 72JDF, subcooling less than 65°F, no RCP running, and RVL/SRVLlS greater than 33% there is an Orange path to FR-C.2, Degraded Core Cooling requiring a transition. While an RCP would be directed to left in service even if normal support equipment was not established or maintained, there is no direction in FR-C.2 to restart an RCP with or without support equipment available. B. Incorrect, The condition requires a transition to FR-C.2 due to an Orange path but an RCP running without all support equipment available would be left in service however if the RCPs are off, there is no action to start an RCP in FR-C.2. Plausible because the transition to FR-C.2 is correct and while a running RCP would remain in service without all normal support conditions available, an idle RCP would not be restarted while while performing FR-C.2 even if the normal support equipment was available. C. Incorrect, The condition does not require a transition to FR-C. FR-C.11 (it is requires a transition to FR-C.2 due to an Orange path) and the RCP support equipment is required if an RCP is to be restarted. Plausible because if the RVLlS RVL/S and been lower, a transition to FR-C.1 would be correct and the RCP support equipment would be still be required prior to restating an RCP for the conditions. ) D. Incorrect, The condition does not require a transition to FR-C.1 FR-C. 1 (it is requires a transition to FR-C.2 due to an Orange path) and the RCP cannot be restarted in unless all support equipment equipment exist. Plausible because if the RVLlS RVL/S and been lower, a transition to FR-C. 1 would be correct and the RCPs can be started in FR-C.FR-C.11 without support equipment under different conditions. ) Page 26
11/2009 Watts Bar SRO NRC NRC Exam - As submitted 10/2/2009 Question Number: 84 Tier: _1_ __ Group _1 2 KIA: WE06 EG2.1.23 Degraded Core Cooling Ability to perform specific system and integrated plant procedures during all modes of plant operation. Importance Rating: 4.3 4.3// 4.4 10 CFR Part 55: 41.10/43.5/45.2/45.6 10CFR55.43.b: 5 KIA Match: Applicant must determine the procedure to be implemented from conditions provided and then determine the restrictions placed on any restart of the RCPs during the performance of the procedure. SRO because the question requires assessing plant conditions and then prescribing a procedure to mitigate, recover, or with which to proceed. Then it requires knowledge of specific actions within the procedure. ) Technical
Reference:
FR-O, Status Trees, Rev. 13
. FR-C.2, Degraded Core Cooling, Rev. 11 Proposed references None to be provided:
Learning Objective: 3-0T -FRCOOO 1 3-0T-FRC0001
- 1. Given a set of plant conditions use FR-O Core Cooling Status Tree to identify and implement the correct procedure.
- 3. Given a set of plant conditions use FR-C.1, FR-C.2, and FR-C.3 to identify and implement: Action Steps, RNOs, Foldout Pages, Notes, and Cautions.
Question Source: New Modified Bank Xx Bank Question History: Watts Bar bank question WE06EA2.1 001 modified Comments ).
)
Page 27
WBN RESPONSE TO DEGRADED CORE COOLING FR-C.2 ') Rev 11
*1., Step I" IAction/Expected Response II" Response*
Response Not Obtained NOTE RCPs should remain in service even if normal operating conditions can NOT be established or maintained. 11.. 11 EVALUATE RCP status:
- a. CHECK at least one RCP a. IF all four RCPs stopped, RUNNING. ~.
I~O ? THEN
** GO TO Step 14.
ESTABLISH and maintain RCP support conditions:
- REFER TO Table 1, RCP Emergency Restart Criteria.
') 12. CHECK if RCS inventory being restored:
- a. CHECK RVLlS greater than 44%. a. IF RVLlS rising, THEN
** GO TO Step 1.
IF RVLlS stable or dropping, THEN
** GO TO Step 13.
- b. RETURN TO Instruction in effect.
)
11 of 22
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ') 85. WE09 EA2.1 085
. Given the following plant conditions:
- A reactor trip occurred on Unit 1.
- The crew was performing .ES-0.1, ES-0.1, "Reactor Trip Response," when a loss of off-site power occurred.
- A transition is made to ES-0.2, "Natural Circulation Cooldown," to cool down the plant.
Which ONE of the following identifies ... (1) a reason the crew would transition to ES-0.3, " Natural Circulation Cooldown with Steam Void In Vessel (with RVLlS)." and (2) why RVLlS is maintained greater than 69% during performance of ES-0.3? A. (1) Detected presence of a void in the reactor head (2) To prevent loss of natural circulation due to steam voiding in the SG tubes. B. (1) Detected presence of voiding in the reactor vessel head (2) To ensure pressurizer level is adequate to accommodate void collapse. C~ (1) Plant conditions require a cooldown rate faster than allowed by ES-0.2. C!' (2) To prevent loss of natural circulation due to steam voiding in the SG tubes.- tubes: D. (1) Plant conditions require a cooldown rate faster than allowed by ES-0.2. (2) To ensure pressurizer level is adequate to accommodate void collapse. ) Page 28
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR DISTRA CTOR ANAL YSIS: A. Incorrect, Detecting a void while performing ES-O.2 would not be a reason to transition to ES-O.3 but RVLlS is maintained greater than 69% to ensure the steam void is maintained above the top of the hot legs to prevent steam more getting into the SG tubes and impeding natural circulation. Plausible because head voiding is not desired in ES-O.2 and if a void is detected actions are taken to collapse the void. Also because maintaining RVLlS above 69% to ensure the steam void is maintained above the top of the hot legs to prevent steam more getting into the SG tubes and impeding natural circulation is correct B. Incorrect, Detecting a void while performing ES-O.2 would not be a reason to transition to ES-O.3 but RVLlS is maintained gre-ater greater than 69% to ensure the steam void is maintained above the top of the hot legs to prevent steam more getting into the SG tubes and impeding natural circulation. Plausible because head voiding is not desired in ES-O.2 and if a void is detected actions are taken to collapse the void. Also*because Also-because pressurizer level is maintained above normal to provide inventory to accommodate the collapse of a head void. C. Correct, the transition to ES-O.3 would be made if plant conditions required a faster cooldown rate than allowed by ES-O.2 and RVLlS is maintained greater than 69% to ensure the steam void void is maintained above the top of the hot legs to prevent steam more getting into the SG tubes and impeding natural circulation. \) D. Incorrect, the transition to ES-O.3 would be made if plant conditions required a faster cooldown rate than allowed by ES-O.2 but maintaining RVLlS above 69% is not to ensure the pressurizer has adequate inventory to accommodate the collapse of the void in the head. Plausible because transitioning to ES-O.3 if plant conditions required a faster cooldown rate than allowed by ES-O.2 is correct and because the pressurizer level is maintained above normal to provide inventory to accommodate the collapse of a head void. Question Number: 85 Tier: _1_ _1__ Group 2 KIA: WE09 EA2.1 Ability to determine and interpret the following as they apply to the (NaturarCirculation (Natural Circulation Operations) - Facility conditions and selection of appropriate procedures during abnormal and emergency operations. Importance Rating: 3.1 13.8
/3.8 10 CFR Part 55: 41.7/45.5/45.6
) 10CFR55.43.b: 5 Page 29
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 KIA Match: Applicant must determine conditions that would require the selection of and transition to a different procedure while a natural circulation cooldown is in progress. SRO because the applicant is required to understand conditions that would cause another procedure to be selected and then demonstrate knowledge of the bases of an action in the procedure selected. Technical
Reference:
ES-0.2, Natural Circulation Cooldown. Rev 20 ES-0.3, Natural Circulation Cooldown with Steam Void In Vessel (with RVLlS), Rev 10 3-0T-EOPOOOO, E-O, Reactor Trip or Safety Injection, R15 Proposed references None to be provided: Learning Objective: 3-0T-EOPOOOO
- 13. Given plant conditions occurring as a result of depressurization during a natural circulation cooldown, determine whether or not RCS voiding is taking place.
- 15. Explain the purpose for and basis of each step in E-O, ES-O.O, ES-0.1, ES-0.2, ES-0.3, and ES-OA.
- 16. Explain the purpose of procedures ES-0.3 and ES-OA including when their use might be required.
Question Source: New x Modified Bank Bank Question History: New question Comments: Page 30
WBN NATURAL CIRCULATION COOLDOWN ES-O.2 Rev 20 IStep III IAction/Expected Response II Response Not Obtained NOTE If at any time it is determined that a natural circulation cooldown and depressurization must be performed at a rate that may form a steam void in the reactor vessel, instruction ES-0.3, Natural Circulation Cool down With Cooldown Steam Void In Vessel (With RVLlS), or ES-OA, Natural Circulation Cooldown With Steam Void In Vessel (Without RVLlS), should be used.
- 15. IF cooldown and depressurization must be performed at a rate which may form steam voids in the reactor vessel, THEN
** GO TO appropriate cool cooldown down instruction:
- ES-0.3, Natural Circulation Cool down With Steam Void in Cooldown Vessel (With RVLlS),
OR
- ES-OA, Natural Circulation Cool down With Steam Void in Vessel (Without RVLlS).
11 of 19
WBN NATURAL CIRCULATION COOLDOWN ES-O.3 WITH STEAM VOID IN VESSEL (WITH RVLlS) Rev 10 IStep III[Action/Expecteg Action/Expected Response II Response Not Obtained
- 6. CONTROL pzr level during cooldown and depressurization:
- a. CHECK pzr level greater a. CONTROL charging and letdown than 29%. to establish pzr level greater than 29%.
- b. CHECK pzr level less than 90%. b. PERFORM the following:
- 1) STOP ReS depressurization.
- 2) MAINTAIN pzr press stable with pzr heaters.
- 3) REDUCE pzr level to less than 90%
90% by one of the following:
- CONTROL charging and letdown as necessary.
OR
- CONTINUE ReS cooldown to shrink ReS inventory.
- 4) WHEN pzr level less than 90%,
THEN CONTINUE depressurization.
- 7. MONITOR RVLlS greater REPRESSURIZE ReS to maintain than 69%. RVLlS greater than 69%.
** GO TO Step 5.
7 of 12
3-0T-EOPOOOO Revision 15 Page 66 of 103
- x. LESSON BODY X. INSTRUCTOR NOTES letdown is greater than charging, the PZR pressure will drop, the vessel void will grow and the PZR level will rise. In the same way when charging is greater than letdown, the PZR pressure will rise, the vessel void will shrink, and the level will drop. If inventory shrink is used to reduce PZR level, careful attention should be paid to the Technical Specification pressure-temperature limits.
- 7) MONITOR RVLIS greater than 69%. Objective 18 I
Purpose:
To ensure the void does not enter Operator Fundamentals: the hot legs and disrupt natural circulation. Understanding of plant design and system operation. Basis: If steam enters the hot legs, it would NOTE: RVLIS at 69% most likely be condensed by by the subcooled corresponds to top of hot legs hot leg fluid. However, there may be the plus allowances for normal potential for some steam to enter the hot channel accuracy. legs and reach the top of the SG U-tubes, ofthe thereby disrupting the natural circulation flow circuit. By monitoring RVLIS and limiting the void growth to the top of the hot legs (repressurizing the Res if necessary), the potential for introducing voids into the SG U-tubes is minimized. An uncertainty is applied to the reading to ensure that the void will actually be above the hot legs, even when assuming worst case channel inaccuracies.
- 8) DETERMINE if cold leg accumulators should be isolated.
Purpose:
To determine if appropriate plant conditions exist for locking out SI. S1. Basis: The cold leg accumulator isolation valves should be closed and their power supplies locked out to prevent the dumping of borated water into the Res when Res pressure drops below accumulator pressure. The pressure criteria from the appropriate T.S. are used in this step.
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 86. 008 G2.2.25 086 Given the following plant conditions:
- Unit 1 is being cooled down for a refueling outage.
- All RCPs were required to be removed from service.
- All 4 SG levels are at 38% NR.
- The RCS is 220°F with RHR Train A in service.
- Component Cooling Water (CCS) pump 1A-A trips due to motor failure.
- The crew is performing actions of AOI-15, "Loss of Component Cooling Water."
Which ONE of the following identifies how the trip of the CCS pump 1A-A will affect Technical Specification ... (1) LCD LCO 3.7.7 Component Cooling System and (2) LCD LCO 3.4.6 RCS Loops - Mode 4? A. (1) LCDLCO 3.7.7 is required to be entered until Mode 5 is reached. LCO 3.4.6 entry NOT required because of (2) LCD of the loss of a support system as provided for by LCD LCO 3.0.6. B~ (1) LCD B:' LCO 3.7.7 is required to be entered until Mode 5 is reached. (2) LCD LCO 3.4.6 is required to be entered because of the loss of a support LCO 3.0.6 can NOT be used. system and the provisions of LCD C. (1) LCDLCO 3.7.7 entry NOT required because the plant is in Mode 4. LCO 3.4.6 entry NOT required because of the loss of a support system (2) LCD as provided for by LCD LCO 3.0.6. D. (1) LCDLCO 3.7.7 entry NOT required because the plant is in Mode 4. (2) LCD LCO 3.4.6 is required to be entered because of the loss of a support LCO 3.0.6 can NOT be used. system and the provisions of LCD Page 31
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, LCO 3.7.7 is required to be entered for the inoperability of one train of Component Cooling Water until the plant is cooled down Mode 5 conditions where the CCS TIS is no longer applicable but not entering LCO 3.4.6 by applying the provision of 3. o. O. 6 is not correct. There is a note in the action statement for 3. 3.7.7
- 7. 7 identifying the required entry into LCO 3.4.6 due to the loss CCS. Plausible because entering LCO 3.7.7 until the plant reaches Mode 5 is correct and because CCS is a support system for RHR. The provisions of LCO 3.0.6 normally allowed system to not be declared inoperable due to a support system inoperability.
B. Correct, LCO 3.7.7 is required to be entered for the inoperability of one train of Component Cooling Water until the plant is cooled down Mode 5 conditions where the CCS TIS is no longer applicable. CCS is a support system to the RHR system and normally to loss of a support system would not require the cascading to TIS to the supported system as allowed by LCO 3.0.6, 3. O. 6, if this case the action statement is LCO 3.7.7 contains a note that LCO 3.4.6 must also be entered and this is an exception to the allowance of LCO 3. o. O. 6 C. Incorrect, LCO 3.7.7 is applicable in Mode 4 making not entering the LCO wrong and not entering LCO 3.4.6 by applying the provision of 3.0.6 is not correct. There is a note in the action statement for 3.7.7 3.7. 7 identifying the required entry into LCO 3.4.6 due to the loss CCS. Plausible because some Tech Specs are not applicable in Mode 4 and because CCS is a support system for RHR. The provisions of LCO
- 3. O. 6 normally allowed system to not be declared inoperable due to a support 3.0.6 system inoperability.
D. Incorrect, LCO 3.7.7 is applicable in Mode 4 making not entering the LCO wrong but entering LCO 3.4.6 as identified in the note in the action statement for 3. 7. 7 is required. Plausible because some Tech Specs are not applicable in Mode 4 and entering LCO 3.4.6 is correct due to the note that prevents using the provisions of LCO 3.0.6. Question Number: 86 Tier: _2_ Group 1 KIA: 008 G2.2.25 Loss of Component Cooling Water (CCW) Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. Importance Rating: 3.2 / 4.2 10 CFR Part 55: 41.5/41.7/43.2 Page 32
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
) 10CFR55.43.b: 2 KIA Match: Applicant is required to recall the Tech Spec bases for the Component Cooling Water LCO and identify from a set of plant conditions the Tech Spec LCO entries that are applicable actions as explained in the TS Bases.
SRO because it requires the knowledge of Tech Spec Bases. Technical
Reference:
Tech Spec 3.7.7 Component Cooling Water (CCS) Bases. Tech Spec 3.4.4 RCS loops - Mode 4, Amendment 59 Proposed references None to be provided: Learning Objective: 3-0T-SYS070A 3-0T -SYS070A
- 16. Regarding Technical Specifications and Technical Requirements for this system:
- a. Identify the conditions and required actions with completion time of one hour or less.
- b. Explain the Limiting Conditions for Operation, Applicability, and Bases.
- c. Given a status/set of plant conditions, apply the
) appropriate Technical Specifications and Technical Requirements. Question Source: New X X Modified Bank Bank Question History: Comments
)
Page 33
RCS Loops-MQDE 4 3.4.6 ) 3.4 REACTOR COOLANT SYSTEM .((RCS) RCS) 3.4.6 RCS Loops-Loops - MODE 4 LCO 3.4.6 Two loops shall be OPERABLE, and consist of either:
- a. Any combination of RCS loops and residual heat removal (RHR) loops, and one loop shall be in operation, when
-the the rod control system is not capable of rod withdrawal; or
- b. Two RCS loops, and both loops shall be in operation, when the rod control system is capable of rod withdrawal.
----------------------------NOTE---------------------------
No RCP shall be started with any RCS cold leg temperature
- s;
~ 350°F unless the secondary side water temperature of each steam generator (SG) is:s; is ~ 50°F above each of the RCS cold leg temperatures.
APPLICABILITY: MODE 4. ) ACTIONS -------- CONDITION REQUIRED ACTION COMPLETION TIME A. Only one RCS loop A.l Initiate action to Immediately OPERABLE. restore a second loop to OPERABLE status. AND Two RHR loops inoperable.
.,
. . "
'.
B. One required RHR lqop lQop B.l B.1 Be in MODE 5. 24 hours inoperable. AND No RCS loops OPERABLE. (continued)
)
Watts Bar-Unit 1 3.4-11* 3.4-11
RCS Loops-MODE 4 3.4.6 " ........ _.. - ACTIONS (continued)
\ _" .. v * " ___
CONDITION REQUIRED ACTION COMPLETION TIME C. One required RCS loop C.1 Restore required RCS 1 hour not in operation, and loop to operation~ operation. reactor trip breakers closed and Rod Control OR System capable of rod withdrawal. C.2 Deenergize Oeenergize all 1 hour control rod drive mechanisms (CRDMs). (CROMs). D. O. Required RCS or RHR D.1 0.1 Deenergize Oeenergize all CRDMs. CROMs. Immediately loops inoperable. AND OR D.2 0.2 Suspend all Immediately No required RCS or RHR operations involving loop in operation. a reduction of RCS boron concentration. AND D.3* 0.3' Initiate action to Immediately restore one loop to OPERABLE status and operation. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3'.4.6.1 3.4.6.1 Verify two RCS loop are in operation when 12 hours the rod control system is capable of rod withdrawal. (continued) Watts Bar-Unit 1 3.4-12
RCS Loops Mode 4 3.4.6 SURVEILLANCE REQUIREMENTS (continued) REQUIREMENTSfcontinued) SURVEILLANCE FREQUENCY SR 3.4.6.2 Verify one RHR or RCS loop is in operation when the rod I 12 hours control system is not capable of rod withdrawal. SR 3.4.6.3 Verify SG secondary side water levels are greater than or I 12 hours equal to 32% narrow range for required RCS loops. SR 3.4.6.4 Verify correct breaker alignment and indicated power are I 7 days available to the required pump that is not in operation. Watts Bar-Unit 1 3.4-13 Amendment 61
- CCS 3.7.7
) 3.7 'PPLANT LANT SYSTEMS 3.7.7 Component Cooling System (CCS) LCO 3.7.7 Two CCS trains shall be OPERABLE. NV1 (aE;G\.l NIJ1 {l.fG\.{ ( a£0 () APPLICABILITY: MODES 1, 2, 3, and 4. (N Moo Eof. t N MoO .r r ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CCS train A.l --------NOTE--------- inoperable. Enter applicable Conditions Condit ions and Required Actions of LCO 3.4.6, "RCS lCO Loops - MODE 4," for residual res i dua 1 heat removal loops made inoperable by CCS.
---------------------
Restore CCS train to 72 hours OPERABLE status. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 5. 36 hours ) Watts Bar-Uni Bar-Unitt .1 3.7-17
. CCS 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify that the alternate feeder breaker to I 7 days the C-S pump is open.
SR 3.7.7.2 -------------------NOTE-------------------- Isolation of CCS fl~w flow to individual components does not render the CCS inoperable. Verify each CCS manual, power operated, and 13131 days automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position. SR 3.7.7.3 Verify each CCS automatic valve in the flow I 18 months ) path that is not. locked, sealed, or otherwise secured in position, actuates to the correct .position on an actual or simulated actuation signal. SR 3.7.7.4 Verify each CCS pump starts automatically I 18 months on an actual or simulated actuation signal. Watts Bar-Unit 1 3.7-18
CCS B 3.7.7
) BASES LCO , A CCS train is considered OPERABLE when:
(continued)
- a. The pump and associated surge tank are OPERABLE; and
- b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE.
isolation of CCS from other components or systems not The isolation.of required for safety may render those components or systems inoperable but does not affect the OPERABILITY of the CCS. APPL I CAB I LITY APPLICABILITY In MODES 1, 2, 3, and 4, the CCS is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger. In MODE 5 or 6, the OPERABILITY requirements of the CCS are determined by the systems it supports. ) ACTIONS A.I Required Action A.l A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," be entered if an inoperable CCS train results in an inoperable RHR loop. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. If one CCS train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE CCS train is adequate to perform the heat removal function. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period. B.1 and B.2 B.l If the CCS train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be (continued) Watts Bar-Unit .1 B 3.7-40 B
LCO Applicability 3.0
)
3.0 LCO APPLICABILITY LCO 3.0.4 b. After performance of a risk assessment addressing inoperable (continued) systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or
- c. When an allowance is stated in the individual value, parameter, or other Specification.
This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY OPi::RABILlTY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO 3.0.6 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. this event, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. ) (continued) Watts Bar-Unit 1 3.0-2 Amendment 55
LCO Applicability
) 3.0 3.0 LCO APPLICABILITY LCO 3.0.6 When a support system's Required Action directs a supported system to be (continued) 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.
LCO 3.0.7 Test Exception LCOs 3.1.9 and 3.1.10 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. ) Watts Bar-Unit 1 3.0-3
RCS Loops - MODE 4 B 3.4.6 B B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.6 RCS Loops - MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid. The reactor coolant is circulated through four RCS loops connected in parallel to the reactor vessel, each loop containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure, level, and temperature instrumentation for control, protection, and indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and to prevent boric acid stratification. In MODE 4, with the reactor trip breakers open and the rods not capable of withdrawal, either RCPs or RHR loops can be used to provide forced circulation. The intent in this case is to provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The flow provided by one RCP loop or RHR loop is adequate for decay heat removal. The other intent is to require that two paths be available to provide redundancy for decay heat removal. In MODE 4, with the reactor trip breakers closed and the rods capable of withdrawal, two RCPs must be OPERABLE and in operation to provide forced circulation. APPLICABLE In MODE 4, with the reactor trip breakers open and the rods SAFETY ANALYSES not capable of withdrawal, Res RCS circulation is considered in determination of the time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation. (continued) Watts Bar-Unit 1 B 3.4-27
RCS Loops - MODE 4 B 3.4.6 BASES APPLICABLE Whenever the reactor trip breakers (RTBs) are in the closed SAFETY ANALYSES position and the control rod drive mechanisms (CRDMs) are (continued) energized, an an inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a malfunction of the rod control system. In addition, the possibility of a power excursion due to the ejection of an inserted control rod is possible with the breakers closed or open. Such a transient could be caused by the mechanical failure of a CRDM. Therefore, in MODE 4 with RTBs in the closed position and Rod Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be OPERABLE and in operation to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal, any combination of two RCS or RHR loops are required to be OPERABLE, but only one loop is required to be in operation to meet decay heat removal requirements. RCS Loops - MODE 4 have been identified in the NRC Policy Statement as important contributors to risk reduction. LCO The purpose of this LCO is to require that at least two loops be OPERABLE. In MODE 4 with the RTBs in the closed position and Rod Control System capable of rod withdrawal, two RCS loops must be OPERABLE and in operation. Two RCS loops are required to be in operation in MODE 4 with RTBs closed and Rod Control System capable of rod withdrawal due to the postulation of a power excursion because of an inadvertent control rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met for all of the postulated accidents. With the RTBs in the open position, or the CRDMs de-energized, the Rod Control System is not capable of rod withdrawal; therefore, only one loop in operation is necessary to ensure removal of decay heat from the core and homogenous boron concentration throughout the RCS. In this case, the LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. An additional loop is required to be OPERABLE to provide redundancy for heat removal. (continued) Watts Bar-Unit 1 B 3.4-28
RCS Loops - MODE 4 B 3.4.6 BASES LCO The Note requires that the secondary side water temperature (continued) of each SG be :0; 50°F above each of the RCS cold leg temperatures before the be:::: temperature:::: start of an RCP with any RCS cold leg temperature :0; 350°F. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started. An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG, which has the minimum water level specified in SR 3.4.6.3. Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required. APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of either RCS or RHR provides sufficient circulation for these purposes. However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure considerations. Operation in other MODES is covered by: LCO 3.4.4, "RCS Loops - MODES 1 and 2";
\I LCO 3.4.5, "RCS Loops - MODE 3";
LCO 3.4.7, "RCS Loops.;. Loops - MODE 5, Loops Filled"; LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled"; LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Water Level" (MODE 6). ) , (continued) Watts Bar-Unit 1 B 3.4-29 Revision 72, 82 Amendment 65
RCS Loops - MODE 4 B 3.4.6 BASES (continued) ACTIONS 6.J. If only one RCS loop is OPERABLE and both RHR loops are inoperable, redundancy for heat removal is lost. Action must be initiated to restore a second RCS or RHRloop RHR loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
!L1 If one required RHR loop is OPERABLE and in operation and there are no RCS loops OPERABLE, an inoperable RCS or RHR loop must be restored restored to OPERABLE status to provide a redundant means for decay heat removal.
If the parameters that are outside the limits cannot be restored, the plant must be brought to MODE 5 within 24 hours. Bringing the plant to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 (~ (:::; 200°F) rather than MODE 4 (200 to 350°F). The Completion Time of 24 hours is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems. C.1 and C.2 If one required RCS loop is not in operation, and the RTBs are closed and Rod Control System capable of rod withdrawal, the Required Action is either to restore the required RCS loop to operation or to de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets. When the RTBs are in the closed position and Rod Control System capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation, the RTBs must be opened. The Completion Times of 1 hour to restore the required RCS loop to operation or de-energize all CRDMs is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period. (continued) Watts Bar-Unit Bar-Unit 1 B 3.4-30
RCS Loops - MODE 4 B B3.46 3.4.6 BASES ACTIONS 0.1. 0.1.020.2 and 0.3 (continued) If no loop is OPERABLE or in operation, all CROMs must be decenergized de-energized by opening the RTBs or de-energizing the MG sets. All operations involving a reduction of RCS boron concentration must be suspended, and action to restore oneRCS one RCS or RHR loop to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and the margin to criticality must not be reduced in this type of operation. Opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation. SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours that two RCS loops are in operation when the rod control system is capable of rod withdrawal. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance. SR 3.4.6.2 This SR requires verification every 12 hours that one ReSRCS or RHR loop is in operation when the rod control system is not capable of rod withdrawal. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance. (continued) Watts Bar-Unit 1 B 3.4-31
RCS Loops - MODE 4 B 3.4.6 BASES
)
SURVEILLANCE SR 3.4.6.3 REQUIREMENTS (continued) SR 3.4.6.3 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is greater than or equal to 32% (value accounts for instrument error, Ref. 1). If the SG secondary side narrow range water level is less than 32%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level. SR 3.4.6.4 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available ayailable to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES 1.
- 1. Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance Tables."
) Watts Bar-Unit 1 B 3.4-32 Revision 29, 79 Amendment 61
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) 87. 010 A2.02 087 Given the following: Unit 1 in Mode 3 with the RCS at normal operating temperature and pressure with the following conditions: Loop #2 Pressurizer spray valve unable to be opened. RCP #4 out of service.
- A Safety injection occurs due to a Steam Generator #3 tube rupture.
- The operating crew is ready to initiate RCS depressurization following the Rapid cooldown in accordance with E-3, "Steam Generator Tube Rupture."
Which ONE of the following identifies ... (1) a condition that would result in the pressurizer sprays being ineffective in depressurizing the RCS, and (2) the action required to accomplish the RCS depressurization? (1 ) (2) Condition Action required to depressurize the RCS is ... A~ A'I Trip of the RCP #2 toiise to-ttse a pressurizer PORV in accordance with E-3.
'with
) B. Trip of the RCP #2 to use auxiliary,sprays auxiliaryspr:ays in accordance with ECA-3.3, "SGTR without PZR Pressure ControL" C. Loss of 120v Vital to use a pressurizer PORV in accordance Instrument Power Board 1-1 with E-3. D. Loss of 120v Vital to use auxiliary sprays in accordance with with Instrument Power Board 1-1 ECA-3.3, "SGTR without PZR Pressure ControL" ') Page 34
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) DISTRACTOR ANAL YSIS: A. Correct, the trip of the RCP #2 would leave only RCP's #1 and #3 running. For the spray to be effective in accordance with the note in E-3, if RCP #2 is not running then the other 3 RCPs are to be running and if the sprays are not available, E-3 directs to use of a pressurizer PORV. B. Incorrect, the trip of the RCP #2 would make the sprays ineffective but use of the auxiliary sprays in accordance with ECA-3.3 is not the correct mitigating action. Plausible because the trip of RCP #2 making the sprays ineffective is correct and implementing ECA-3.3 is an action directed in the RNO column for when sprays are not available but would only be performed if other actions were unsuccessful and
. using the auxiliary spray is an action directed in the ECA as well as in E-3.
C. Incorrect, the loss of the 120v AC Vital Instrument Power Board would not make the sprays ineffective (manual control of the Loop 1 spray would be available) and use of a PORV in accordance with E-3 is the action directed in E-3. Plausible because the loss of the 120v AC Vital Instrument Power Board would prevent operation of spray controls except for manual control of Loop 1 and the use of a PORV in accordance with E-3 is correct. D. Incorrect, the loss of the 120v AC Vital Instrument Power Board would not make the sprays ineffective (manual control of the Loop 1 spray would be available) and use of the auxiliary sprays in accordance with ECA-3.3 is not the correct mitigating action. Plausible because the loss of the 120v AC Vital Instrument Power Board would prevent operation of spray controls except for manual control of Loop 1 and implementing ECA-3.3 is an action directed in the RNO column for when sprays are not available but would only be performed if other actions were unsuccessful and using the auxiliary spray is an action directed in the ECA as well as in E-3. Question Number: 87 Tier: _2_ Group 1 KIA: 010 A2.02 010A2.02 Pressurizer Pressure Control System (PZR PCS) Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures. Importance Rating: 3.9/3.9 3.9/ 3.9 ) 10 CFR Part 55: 41.5/43.5/45.3/45.13 10CFR5.5.43.b: 10CFR55.43.b: 5 Page 35
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 KIA Match: Applicant must determine the malfunction that would cause the pressurizer spray valves to be ineffective at reducing RCS pressure during an attempted RCS depressurization condition and how procedures would be used as a result of the condition to mitigate and control the plant. SRO because the question requires assessing plant conditions and then prescribing a procedure to mitigate, recover, or with which to proceed. Technical
Reference:
E-3, Steam Generator Tube Rupture, Rev 22 ECA-3.3, SGTR without PZR Pressure Control, Rev 10 AOI-25.01, Loss of 120v AC Vital Instrument Power Board, Rev 27 Proposed references None to be provided: Learning Objective: 3-0T-EOP-0300
- 5. Given a set of plant conditions, use E-3, ES-3.1, ES-3.2, and ES-3.3 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.
Question Source: New X X Modified Bank Bank Question History: New question Comments: Page 36
WBN STEAM GENERA GENERATORTOR TUBE RUPTURE E-3 Rev 22 IStep I IAction/Expected Response I IResponse Not Obtained CAUTION Cycling of the pzr PORV should be minimized to improve PORV reliability_ reliability. NOTE
- If RCPs are not running, the upper head region may void during RCS depressurization. This will result in a rapidly rising pzr level.
- Either Loop 1 or 2 pzr spray valve is effective for Loop 2 RCP in service-w for Loops 1,3, &
serviceQ[ & 4 RCPs in service.
- 30. INITIATE RCS depressurization to minimize break flow, and REFILL pzrpzr to greater than 15%
[33% ADV].
- a. CHECK pzr level a. ** GO TO Caution prior less than 63% to Step 32.
[58% ADV].
- b. MAINTAIN subcooling greater than 65°F
[85°F ADV].
- c. DEPRESSURIZE RCS with c. IF normal sprays NOT available, normal sprays at maximum rate. THEN USE one pzr PORV, and MONITOR the following:
- Vessel head void formation.
- PRT rupture.
IF both normal sprays AND pzr PORVs NOT available, THEN ALIGN aux spray USING Appendix A (E-3) ALIGN AUX SPRAY. IF RCS pressure control can NOT be established, THEN
** GO TO ECA-3.3, SGTR Without Pzr Pressure Control.
19 of 47
WBN SGTR WITHOUT PZR PRESSURE CONTROL ECA-3.3 ) Rev 10 IStep II Action/Expected Response II Response Not Obtained
- 4. ESTABLISH aux spray control:
- a. ENSURE at least one SI pump a. IF NO SI pump can be started, running. THEN
** GO TO Step 5.
- b. ENSURE at least one Charging b. IF NO charging pump can be pump running, AND started, ALIGN aux spray USING THEN Appendix A (ECA-3.3), ** GO TO Step 5.
ALIGN AUX SPRAY. IF aux spray can NOT be established, THEN
** GO TO Step 5.
- c. ** GO TO E-3, Steam Generator Tube Rupture, Step 30.
- 5. MAINTAIN Intact S/G NR levels:
- a. MONITOR levels a. MAINTAIN total feed flow greater than 29% [39% AOV]. greater than 410 gpm UNTIL level is greater than 29% [39% AOV] in at least one S/G.
- b. CONTROL intact S/G levels b. IF level in any intact S/G between 29% and 50% continues to rise without feed
[39% and 50% AOV]. flow, THEN
** GO TO E-3, Steam Generator Tube Rupture.
- 6. CHECK pzr level ** GO TO Step 1.
greater than or equal to 15% [33% AOV].
)
50f29 5 of 29
AOI-25.01 WBN LOSS OF 120VAC 120V AC VITAL INSTRUMENT Revision 27 POWER BOARDS 1-1 AND 2-1 Page 17of24 17 of24 - - - - APPENDIX A Page 4 of 4 120V AC VITAL INSTRUMENT POWER BOARD 1-1 LOADS (continued) SYSTEM 67 - ERCW:
- 1-TCV-67-84,-85, -92, & & -93, Train A (A &
& C) Lower Containment and CRDM Coolers, go open (temperature controllers NOT operable).
SYSTEM 68 - RCS:
- RVLlS [1-R-184].
RVLlS[1-R-184].
- RCP UV and UF Relays on RCP 1 Sensor Panel (to SSPS).
- 1-LT-68-339, PZR Level Indication fails low.
- 1-PT-68-340, PZR Pressure Indication fails low.
- Only Loop 1 spray valve operable in Manual.
SYSTEM 70 - CCS:
- 1B-B CCS Pump starts.
SYSTEM 74 - RHR:
- 1-FCV-74-1 will require placing 1-XS-74-1 to AUX position [Rx MOV Bd 1A1-A, Compt 5B] to open valve when RCS pressure is less than 380 psig.
- 1-FCV-74-16 & -32 will require placing 1-XS-74-16 & -32 to AUX position
[1-L-11A Aux Control Room] to control RHR cooldown temperature. SYSTEM 85 - CERPI
- 1-MON-85-5000/1, CERPI Monitor 1.
SYSTEM 88 - Containment Isolation:
- Train A CRI and CVI will occur after restoration of the board.
- If the refueling logic switch, 1-HS-90-410-A [1-R-73], is in the REFUEL position, then an "A" Train ABI (partial ABI associated with high rad in refuel area) will also occur when the CVI is initiated.
SYSTEM 90 - Radiation Monitors:
- 1-RM-90-106, Lower Containment Air Monitor control power NOT operable.
- 0-RM-90-125, MCR Train A Air Intake Monitor control power NOT operable.
- 1-RM-90-130, Containment Train A Purge Exhaust Monitor control power NOT operable.
SYSTEM 92 - Excore Nuclear Instrumentation:
- All Channell NIS Control and Instrument Power.
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 88. 026 A2.03 088 Given the following:
Unit 1 at 100% power.
- Containment Pressure Transmitter "1-PDT-30-43" (Channel III) failed and is out of service with the channel bistables positioned as required in accordance with Technical Specifications.
- The Surveillance Instruction to prove the Channel Operability Test (COT) requirements for Containment Pressure Transmitter "1-PDT-30-44" "1-PDT 44" (Channel II) is required to be performed prior to the return of transmitter "1-PDT-30-43."
Which ONE of the choices below correctly completes the following statement? During performance of the Surveillance Instruction for 1-PDT-30-44, the containment HI-Hi bistables for ... A. 1-PDT-30-43 will be placed to trip to provide for automatic actuation of Phase B isolation and Containment Spray if a Hi-Hi containment pressure developed during the surveillance test. B. 1-PDT-30-43 will be placed to trip because Technical Specification while allowing for dedicated operator action for a Phase B isolation, does NOT allow for dedicated operator action for Containment Spray actuation. O! C'!' both channels will be placed to bypass and if a containment Hi-Hi pressure condition developed during the test, both the Phase B isolation and the Containment Spray actuation would occur automatically. D. both channels will be placed to bypass and while a Phase B isolation would automatically occur, a Dedicated Operator would be assigned to manually initiate a Containment Spray actuation if a containment Hi-Hi pressure condition developed during the test. Page 37
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, The Hi-Hi bistable for 1-PDT-30-43 will not be placed to the trip position (to provide for automatic actuation of Phase B isolation Hi-Hi containment pressure developed during the surveillance test) because the other 2 channel still provide for the protection of the function. 'f!'lausible fflausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position. ) B. Incorrect, The Hi-Hi bistable for 1-PDT-30-43 will not be placed to the trip position (to provide for automatic actuation of Phase B isolation Hi-Hi containment pressure developed during the surveillance test) because the other 2 channel still provide for the protection of the function and Tech Specs do not contain provisions for use of a dedicated operator for this application. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met (and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position) and because the use of a Dedicated Operator is provided for in other Containment Tech Specs. C. Correct, Both channels would have the Hi-Hi stables placed in the bypass position as identified in the Tech Spec 3.3.2 Bases. The Required Action for LCO 3.3.2 Condition E has a Note that allows a channel to be bypassed for up to 12 hours for surveillance testing. This Note is explained in the Tech Spec Bases. There are 4 channels provided for the Hi-Hi containment function to actuate and it takes 2 of the 4 to generate the signal. D. Incorrect, Both channels would have the Hi-Hi stables placed in the bypass position as identified in the Tech Spec 3. 3. 228ases Bases but and Tech Specs do not contain provisions for use of a Dedicated Operator for this application. Plausible because the positioning of the bistables is correct and use of a Dedicated Operator is provided for inin~nother another Containment Tech Spec. Page 38
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ) Question Number: 88 Tier: 2 _2_ Group 1 KIA: 026 A2.03 Containment Spray System (CSS) Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or Failur~ of ESF operations: Failure
~, """-""
---~
Importance Rating: 4.1 / 4.4 10 CFR Part 55: 41.5/43.5/45.3/45.13 10CFR55.43.b: 2,5 KIA Match: Applicant must determine how the bistables will be concurrently configured on a failed Containment Spray System actuation ESF transmitter and an ESF transmitter which is required to be tested in order to run the surveillance instruction and how the function is maintained as identified in the Technical Specification bases. SRO because the question requires knowledge of Tech Spec bases that is required to analyze Tech Spec required actions and terminology. Technical
Reference:
Tech Spec LCO 3.3.2, ESFAS Instrumentation(Amendment 68) and Bases (Revision 90) 1-47W611-88-1, R23 Proposed references None to be provided: Learning Objective: 3-0T-SYS072A
- 20. Explain Tech Spec bases for Containment Spray components and parameters governed by Tech Specs.
Question Source: New x X Modified Bank Bank Question History: New Question Comments: Page 39
ESFAS Instrumentation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.2-1. ACTIONS
------------------------------------N 0 TE ---'-----------~---------------------
-------------------------------------NOTE--------------~----------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with one A.1 Enter the Condition referenced Immediately or more required channels or in Table 3.3.2-1 for the trains inoperable. channel(s) or train(s). B. One channel or train inoperable. B.1 Restore channel or train to 48 hours OPERABLE status. OR B.2.1 Be in MODE 3. 54 hours AND B.2.2 Be in MODE 5. 84 hours (continued) Watts Bar-Unit 1 3.3-24
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One train inoperable. C.1 OTE-----------------
-------------N 0 TE ----------------
One train may be bypassed for up to 4 hours for surveillance testing provided the other train is OPERABLE.
---------------------------------------
--------------------------------------
Restore train to OPERABLE status. 24 hours OR C.2.1 Be in MODE 3. 30 hours AND C.2.2 Be in MODE 5. 60 hours ) D. One channel inoperable. 0.1 D.1 -------------NOTE---------------
--------------NOT E----------------
One channel may be bypassed for up to 12 hours for surveillance testing.
---------------------------------------
Place channel in 72 hours trip. OR 0.2.1 D.2.1 Be in MODE 3. 78 hours AND 0.2.2 D.2.2 Be in MODE 4. 84 hours
------
(continued) Watts Bar-Unit 1 3.3-25 Amendment 68
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One Containment E.1 ---------------N 0 TE ----------------
---------------NOTE Pressure channel inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
----------------------------------------
Place channel in bypass. 72 hours OR E.2.1 Be in MODE 3. 78 hours AND E.2.2 Be in MODE4. MODE 4. 84 hours F. One channel or train F.1 Restore channel or 48 hours inoperable. train to OPERABLE status. OR F.2.1 Be in MODE 3. 54 hours AND F.2.2 Bein MODE 4. 60 hours --- ._---_.- - - - - - _ ... -- (continued) Watts Bar-Unit 1 3.3-26 Amendment 68
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. One train inoperable. G.1 -----------------N OT E---------------
------------------NOT One train may be bypassed for up to 4 hours for surveillance testing provided the other train is OPERABLE.
-----------------------------------------
Restore train to OPERABLE 24 hours status. OR G.2.1 Be in MODE 3. 30 hours AND G.2.2 Be in MODE 4. 36 hours H. One train inoperable. H.1 ---------------N 0 TE
---------------NOT E-----------------
One train may be bypassed for up to 4 hours for surveillance testing provided the other train is OPERABLE.
----------------------------------------
Restore train to OPERABLE 24 hours status. OR H.2.1 Be in MODE 3. 30 hours AND H.2.2 Be in MODE 4. 36 hours (continued) (continued) Watts Bar-Unit 1 3.3-27 Amendment 68 ~
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME I. One Steam Generator Water 1.1 ---------------NOT E-----------------
---------------N 0 TE-----------------
Level--High High channel One channel may be bypassed inoperable. for up to 12 hours for surveillance testing.
----------------------------------------
Place channel in trip. 72 hours OR 1.2.1 Be in MODE 3. 78 hours OR 1.2.2 Be in MODE 4. 84 hours J. One or more Turbine Driven J.1 Restore channel to 48 hours Feedwater Pump trip channel(s) OPERABLE status. inoperable. OR J.2 Be in MODE 3. 54 hours K. One channel inoperable. K.1 ----------------NOT E----------------- One channel may be bypassed for up to 12 hours for surveillance testing.
-----------------------------------------
Place channel in bypass. 72 hours OR I~~ (continued) (continued) Watts Bar-Unit 1 3.3-28 Amendment 68, 75
ESFAS Instrumentation 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME K. (continued) (continued) K.2.1 Be in MODE 3. 78 hours AND K.2.2 Be in MODE 5. 108 hours L. One P-11 interlock channel L1 L.1 Verify interlock is in required 1 hour inoperable. state for existing unit condition. OR L.2.1 Be in MODE MODE3. 3. 7 hours AND L2.2 L.2.2 Be in MODE 4. 13 hours ((continued) continued) Watts Bar-Unit 1 3.3-29 Amendment 68
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME M. One Steam Generator Water ------------------N OTE 0 TE ---------------------- Level--Low--Low channel One channel may be bypassed inoperable. for up to 12 hours for surveillance testing.
------------------------------------------------
M.1.1 Place channel in 72 hours trip. AND M.1.2 For the affected protection set, 72 hours set the Trip Time Delay (Ts) to match the Trip Time Delay (Tm) OR M.2.1 Be in MODE 3. 78 hours AND M.2.2 MODE4. Be in MODE 4. ) 84 hours N. One Vessel L1t-.T T channel --------------------N OTE-------------------- 0 TE ------------------- inoperable. One channel may be bypassed for up to 12 hours for surveillance testing.
------------------------------------------------
-------------------------------------------------
N.1 Set the Trip Time Time. 72 hours Delay threshold power (Ts) level for (T 5) and (T m) to 0% power. OR N.2 Be in MODE 3. 78 hours
----- ---------_ .. _ - - - - - - - - - ----
(continued) Watts Bar-Unit 1 3.3-30 Amendment 68
ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
~~~
High ~~a~e~~~~:aa~;::. channel inoperable. Level I ~~-~-~~~~~~~~I~~~:n~-~~-~~~-~~------ O. One MSW Room Water Level --------------------NOTE------------------- O. The inoperable channel may be bypassed for up to 12 hours for surveillance testing of other channels. 0.1 Place channel in trip 72 hours OR 0.2 Be in MODE 3 78 hours SURVEILLANCE REQUIREMENTS NOT E------------------------------------------------------
NOTE Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.
SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.2.2 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.2.3 Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.2.4 Perform COT. 184 days ( continued) Watts Bar-Unit 1 3.3-31 Amendment 68
ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.5 ----------------------------N 0 TE TE------------------------ Slave relays tested by SR 3.3.2: 3.3.2.77 are excluded from this surveillance. Perform SLAVE RELAY TEST. 92 days OR 18 months for Westinghouse type AR relays SR 3.3.2.6 -------------------------~---
----------------------------- NOTE NOT E----------------------------
Verification of relay setpoints not required. Perform TADOT. 92 days SR 3.3.2.7 Perform SLAVE RELAY TEST on slave relays K603A, I 18 months K603B,K604A, K604B,K607A,K607B,K609A, K609B, K612A, K625A, and K625B. SR 3.3.2.8 ----------------------------- NOTE--------------------------
-----------------------------NOT Verification of setpoint not required.
Perform TADOT. 18 months SR 3.3.2.9 -----------------------------NOTE--------------------------
-----------------------------NOT E--------------------------
This Surveillance shall include verification that the time constants are adjusted to the prescribed values. Perform CHANNEL CALIBRATION. 18 months (continued) Watts Bar-Unit 1 3.3-32 Amendment 17
ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.10 ---------------------------NOTE -------------------------------- Not required to be performed for the turbine driven AFW pump until 24 hours after;:::after;:: 1092 psig in the steam generator. Verify ESFAS RESPONSE TIMES are within 18 months on a limit. STAGGERED TEST BASIS SR 3.3.2.11 ---------------------------N 0 TE --------------------------------
---------------------------NOTE ---------------------------------
Verification of setpoint not required. Perform T TADOT. ADOT. Once per reactor trip breaker cycle
)
Watts Bar-Unit 1 3.3-33 Amendment 13
')
I /
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT
- 1. Safety Injection
- a. Manual 1,2,3,4 2 B SR 3.3.2.8 NA NA Initiation h.
- b. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 33.23 3.3.2.3 and Actuation SR3.3.2.5 SR 3.3.2.5 Relays SR 3.3.2.7
- c. Containment 1,2,3 3 D SR 3.3.2.1 33.2.1 ::::;
- 0; 1.6 psig 1.5 psig Pressure- SR 3.3.2.4 High SR 33.2.9 SR3.3.2.9 SR 3.3.2.10
- d. Pressurizer 1,2,3(a) 3 D SR 3.3.2.1 ~
2: 1864.8 psig 1870 psig Pressure-Low SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
- e. Steam Line 1, 2, 3(a) 3(a) 3 per steam D SR 3.3.2.1 ~
2: 666.6(b) psig 67S(b)psig 675(b)psig Pressure-Low line SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10 \
)
J 2. Containment Spray
- a. Manual 1,2,3,4 2 per train, B SR 3.3.2.8 NA NA Initiation 2 trains h.
- b. Automatic 1, 2, 3, 4 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 SR3.3.2.3 and Actuation SR 3.3.2.5 Relays
- c. Containment 1,2,3 4 E . SR 3.3.2.1 :0; 2.9 psig
- 2.8 psig Pressure- SR33.2.4 SR3.3.2.4 High High SR 3.3.2.9 SR3.3.2.9 SR 3.3.2.10 (continued)
(a) Above the P-11 (Pressurizer Pressure) Interlock. (b) Time constants used in the lead/lag controller are t1 ;::: 2 50 seconds and t2 ::::;
- 5 seconds.
Watts Bar-Unit 1 3.3-34
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDmONS CONDITIONS REQUIREMENTS VALUE SETPOINT
- 3. Containment Isolation
- a. Phase A Isolation (1) Mannal Manual 1,2,3,4 2 B SR 3.3.2.8 NA NA Initiation (2) Antomatic Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation SR3.3.2.3 SR 3.3.2.3 Logic and SR 3.3.2.5 Actuation ActUation SR 3.3.2.7 Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
h.
- b. Phase B Isolation (1) Mannal Manual 1,2,3,4 2 per train, B SR 3.3.2.8 SR3.3.2.8 NA NA Initiation 2 trains (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.3 Logic and 3.3~2.5 SR 3.3.2.5
) Actuation SR 3.3.2.7 J Relays (3) Con- 1,2,3 4 E SR 3.3.2.1 ::;;2.9
- S 2.9 psig 2.8 psig tainment taimnent SR 3.3.2.4 Pressure-- SR 3.3.2.9 SR3.3.2.9 High High SR 3.3.2.10
- 4. Steam Line Isolation
- a. Manual 1, 2(e), 3(e) I/valve lIvalve F SR 3.3.2.8 NA NA Initiation h.
- b. Automatic i 1, 2(e), 3(e)
C), 3(c) 2 trains G SR 3.3.2.2 NA NA Actuation SR 3.3.2.3 SR3.3.2.3 Logic and SR 3.3.2.5 Actuation Relays (continued) (continued) (C) Except when all MSIVs are closed and de-activated. Watts Bar-Unit 1 3.3-35
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT
- 4. Steam Line Isolation (continued)
(continued)
- c. Containment 2(e), 3(e) 1, le), 4 E SR3.3.2.1 SR 3.3.2.1 :$; 2.9 psig
- 2.8 psig Pressure- SR 3.3.2.4 High High SR3.3.2.9 SR 3.3.2.9 SR 3.3.2.10
- d. Steam Line Pressure (1) Low 1, 2(e), 3(a)(e) 3(a) (e) 3 per steam D SR 3.3.2.1 ~
;?: 666.6(b) psig 675(b) psig line SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 3(d) (e) :$; 108.5(e) psi
- 100(e) psi.
psi (2) Negative 3 per steam D SR 3.3.2.1 Rate-High line SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
- 5. Turbine Trip and Feedwater F eedwater Isolation
\ ) a. Automatic 1,2(t),3(t) 1, 2(t), 3(1) 2 trains H SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays
- b. SGWater 1, I, 2(t), 3(t) 3 per SG SR 3.3.2.1 :0;
- $; 83.1% 82.4%
Level-High SR 3.3.2.4 High(P-14) SR 3.3.2.9 (h) SR 3.3.2.10 SR 3.3.2.10 (h)
- c. Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
(continued)
.(a) Above the P-11 (Pressurizer Pressure) interlock interlock..
(b) Time constants used in the lead/lag controller are t1 :?: ~ 50 seconds and t2 :$; :0; 5 seconds. (c) Except when all MSIVs are closed and de-activated. (d) Function automatically blocked above P-11 (Pressurizer Interlock) setpoint and is enabled below P-11 when safety injection on Steam Line Pressure Low is manually blocked. (e) Time constants utilized in the rate/lag controller are ta and 4 hand t4 :?:
;::o: 50 seconds.
(f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve. (g) MODE 2 if Turbine Driven Main Feed Pumps are operating. (h) For the time period between February 23, 2000, and prior to turbine restart (following the next time the turbine Forthe is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1::.FSV 1~FSV 027. Watts Bar-Unit 1 3.3-36 Amendment 23 /)
ESFASlnstrumentation 3.3.2 Table 3.3.2-1 (page 4 of 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALlIE VALUE SETPOINT
- d. North MSV Vault 1, 2(f),
2(t), (g) 3/vault 0 SR 3.3.2.6 ~ S; 5.31 inches 4 inches Room Water Room SR 3.3.2.9 LeveI-Level- High
- e. South MSV Vault 1, 2(t)(g) 1,2(1).(g) 3/vault 0 SR3.3.2.6 SR 3.3.2.6 ~
S; 4.56 inches 4 inches Room Water Room SR 3.3.2.9 LeveI-Level- High
- 6. Auxiliary Feedwater
- a. Automatic 1,2,3 2 trains G SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays
- b. SG Water Level- 1,2,3 3 perSG per SG M SR3.3.2.1 SR 3.3.2.1 ~
;::: 16.4% 17.0%
Low Low SR 3.3.2.4 SR 3.3.2.9 SR3.3.2.9 Coincident with: SR3.3.2.10 SR 3.3.2.10 1) I) Vessel ~T Vesselll.I 1,2 3 N SR 3.3.2.4 Vessel ~T Vesselll.I ~T Vessel LlI equivalent SR 3.3.2.9 variable variable to power input ~S; 52.6% input 50%
~50%RIP S;50% RTP RIP RTP RIP RTP With a time ~
S; l.OlTs LOlTs Is Ts (Note 1, I, delay (Is) (Ts) (Note 1, Page 3.3-40) if one S/G Page 3.3-40) is affected or A time delay ~ S; LOlTm l.OlTm 1m Tm (Note 1, I, (1m) (Tm) if two (Note 1, I, Page 3.3-40) or more S/G's Page 3.3 -40) 3.3-40) are affected. OR
- 2) Vessel ~T VesseIll.I 1,2 3 N SR 3.3.2.4 Vessel ~T Vesse1ll.I ~T Vessel LlI equivalent SR3.3.2.9 SR 3.3.2.9 variable variable to power> input ~ S; 52.6% input 50%
50% RIP RTP RIP RTP RIP RTP with no time delay (Is (Ts and Im~O Tm=O (continued) (f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve. (g) MODE 2 if Turbine Driven Main Feed Pumps are operating. Watts Bar-Unit 1 3.3-37 Amendment 7
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION COJ\l)ITIONS CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT
- 6. Auxiliary Feedwater (continued)
- c. Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
- d. Loss of Offsite 1( i) , 2( j ) 4 per bus F Refer to Function 4 of Table 3.3.5-1 for Power SRs and Allowable Values e..
- e. . Trip of all 1,2 1 per pump J SR 3.3.2.8 ::::
2: 48 psig 50 psig Turbine Driven SR 3.3.2.9 Main Feedwater SR 3.3.2.10 Pumps
- f. Auxiliary AuxilialY 1,2,3 3 F SR 3.3.2.6 2: 0.5 A) 2': A) 1.2 Feedwater SR 3.3.2.9 pSlg psig pSlg psig Pumps SR3.3.2.1O SR 3.3.2.10 Train A andB AandB 2: 1.33 B) :::: B) 2.0 2.0 Suction Transfer pSlg psig psig pSlg on Suction Pressure - Low
- 7. Automatic Switchover Switch over to Containment Sump
- a. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays (continued)
(i) Entry into Condition J J may be suspended for up to 4 hours when placing the second Turbine Driven Main Feedwater (TDMFW) Pump in service or removing one of two TDMFW pumps from service. (j) When one or more Turbine Driven Feedwater Pump(s) are supplying feedwater to steam generators. Watts Bar-Unit 1 3.3-38 Amendment 1, 75
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of of7) 7) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP TRlP FUNCTION CO]\<'DITIONS CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT
- 7. Automatic Switchover to Containment Sump (continued)
- b. Refueling 1,2,3,4 4 K SR 3.3.2.1 ::::;::: 155.6 inches 158 inches Water Storage SR 3.3.2.4 from from Tank (RWST) SR 3.3.2.9 Tank Base Tank Base Level-Low SR 3.3.2.10 Coincident Refer to Function 1 (Safety Injection) for all initiation with Safety functions and requirements.
Injection and Coincident 1,2,3.4 1,2,3,4 4 K SR 3.3.2.1 ;::: 37.2 in
- 38.2 in.
with SR 3.3.2.4 above above Containment SR 3.3.2.9 el. 702.8 ft el. 702.8 ft Sump SR 3.3.2.10 Level- High 8 ESFAS Interlocks
- a. Reactor 1,2,3 1 per train, F SR 3.3.2.11 SR3.3.2.11 NA NA Trip, P-4 2 trains
- b. Pressurizer Pressure, P-II (1)
(I) Unblock 1,2,3 3 L SR 3.3.2.1 ::;
- S: 1975.2 psig 1970 psig (Auto SR 3.3.2.4 33.2.4 Reset of SR3.3.2.9 SI Block)
(2) Enable 1,2,3 3 L SR 3.3.2.1 ::::
;::: 1956.8 psig 1962 psig Manual SR 3.3.2.4 Block of SR 3.3.2.9 SI Watts Bar-Unit 1 3.3-39
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 7 of 7) Engineered Safety Feature Actuation System Instrumentation NOTE 1: Steam Generator Water Level Low-Low Trip Time Delay: Ts == A(P fl + B(P li + C(P}C(P) + D Tm == E(P l Tn! + F(P l/ + G(P}G(P) + H Where: P = Vessel L'lT llT Equivalent to power (% RTP), P ::;
- <:; 50% RTP.
Ts = Time Delay for Steam Generator Water Level--Low-Low Reactor Trip, one Steam Generator affected. Tm = Time Delay for Steam Generator Water Level--Low-Low Reactor Trip, two or more Steam Generators affected.
=
A =-0.0085041 = E =-0.0047421
=
B = 0.9266400 = F = 0.5682600
=
C = -33.85998 G = -23.70753 D = 474.6060 H = 357.9840 Watts Bar-Unit 1 3.3-40
ESFAS Instrumentation B 3.3.2 B BASES ACTIONS 0.1. 0.2.1. and 0.2.2 (continued) 0.1.0.2.1. Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE. The Required Actions have been modified by a Note that allows placing an inoperable channel in the bypassed condition for up to 12 hours while performing routine surveillance testing of other channels. The Note also allows a channel to be placed in bypass for up to 12 hours for testing of the bypassed channel. However, only one channel may be placed in bypass at anyone time. The 12 hours allowed for testing are justified in Reference 17. E.1, E.2.1, and E.2.2 Condition E applies to:
- Containment Spray Containment Pressure-High High;
- Steam Line Isolation Containment Pressure-High High; and
- Containment Phase B Isolation Containment Pressure-High High.
None of these signals has input to a control function. Thus, two-out-of-three logic is necessary to meet acceptable protective requirements. However, a two-out-of-three design would require tripping a failed channel. This is undesirable because a single failure would then cause spurious containment spray initiation. Spurious spray actuation is undesirable because of the cleanup problems presented. Therefore, these channels are designed with ((continued) continued) Watts Bar-Unit 1 B 3.3-104 Revision 90 Amendment 68
ESFAS Instrumentation B3.3.2 ') BASES ACTIONS E.1. E.2.1. and E.2.2 (continued) two-out-of-four logic so that a failed channel may be bypassed rather than tripped. Note that one channel may be bypassed and still satisfy the single failure criterion. Furthermore, with one channel bypassed, a single* single instrumentation channel failure will not spuriously initiate containment spray. To avoid the inadvertent actuation of containment spray and Phase B containment isolation, the inoperable channel should not be placed in the tripped condition. Instead it is bypassed. Restoring the channel to OPERABLE status, or placing the inoperable channel in the bypass condition within 72 hours, is sufficient to assure that the Function remains OPERABLE and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The Completion Time is further justified based on the low probability of an event occurring during this interval. Failure to restore the inoperable channel to OPERABLE status, or place it in the bypassed condition within 72 hours, requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE. The Required Actions are modified by a Note that allows placing one channel in bypass for up to 12 hours while performing .routine surveillance testing. The ) channel to be tested can be tested in bypass with the inoperable channel also in bypass. The time limit is justified in Reference 17. F.1, F.2.1, and F.2.2 F.1. Condition F applies to:
- Manual Initiation of Steam Line Isolation;
- Loss of Offsite Power;
- Auxiliary Feedwater Pump Suction Transfer on Suction Pressure-Low; and (continued)
Watts Bar-Unit 1 B 3.3-105 B Revision 90 Amendment 68
L-Illl- L L9MLt-1 2 3 4 5 6 7 10 11 12 A A A ACTUATE TRAINA/8 SIS Q~~eAL Q~~SAL 1-47.611-63-1 1-47\1'611-63-' B B cC cC TR A S~~12 CONTAINIoIENTSPRAY CONTAINt.!ENT SPRAY OISCHARGE DISCHARGE VALVE CONTAINI.IENTSPRAY l-FCV-7Z-2 P\,It.lP1-47We11-72-'
,**H1611-72-' t.lAIN STEAt.I LINE SYSTEM SHEET
'.1 ,
'-471611-72-1
..
ISOLATION VALVES
'-4711611-1-1 1-47W611-JO 1,~
1-4711611-88 1-47W611-88 SYSTEt.I S'rSTEt.I 1-47'611-30 1-471611-30 SHEET I, 6 1,6 2 SYSTEhi SYSTEM 1-47'/11611-30 1-47111611-30 ,, J ** J NOTES, I.
- 1. FOR TRAINH!AIN B, l-RE-90-1Jl.
'-RE-90-131. l-HS-90-41S.
l-HS-90-415, AND 0-HS-90-1J6A2. 0-HS-90-136A2 ARE USE:D. USED.
'-47'611-62 1 0-HS-90-1J6Al 0-HS-90-136Al .. a. lJ6A2 136A2 ARE MULTI SELECT SW'S 5W'S USED TO TEST ONE: ONE
'-471611-63 J, 5, 6, 8 1-471611-32 CHANNEL AT A THolE.
TtI.lE. D 1-47W611-65 1-47'111611-65 I, 2, J 1,2, 1-47W611-32 1-47W611-67
'-47W611-67 D
'-471611-68
'-47W611-68 1 .3. ALL ISOLATION SIGNALS AND CIRCUITRY THAT ARE PART OF ENGINEERED
- 3. ,l,LL 1-471611-70 1-47W611-70 J SAFEGUAROS SAFEGUARDS ARE REDUNDANT.
- 4. NoTENOTE DELETED.
1-4711611_77 1-47W611_77 1, 5 1,5 1-471611-81 1
- 5. EACH TRAIN A RESET HA5 HAS AN ASSOCIATED TRAIN B RESET. SEE TVA 1-4"611-81 '-47W610-30-' FOR TRAIN B "RESET" HANOSWITCH DRAWING 1-47'610-.30-1 HANDSWITCH NU"'BERS.
NUMBERS.
'-471611-'
'-47'611-' J 6. VALVES l-FCV-4J-22 1-f'CV-43-22 AND l-FCV-4J-2J l-FCV-4J-Z3 ""'YMAY BE NORI.tALL'I' NORMALLY OPEN OPtN 'ITH WITH
'-471611-Jl
'-47W611-31 l-F5V-4.3-2J l-FSV-43-23 ENERGIZED TO ALLO\l ALLOW CONTINUOUS Res RCS FLOW DUE TO THERI.tAL THERMAL 1-47'11611-88 1-47W611-88 CYCLING CONCERNS (RET. (REF. EOC-511.34).
EDC-51134). 1-47'11'611-26 1-47'611-26
- 8. l-RE-90-1.30 l-RE-90-130 .. a. lJl.
131, WHICH HOUS<: HOUSE GAS RADIATION DETECTORS, OETECTORS, I.IONITOR MONITOR CONTAINMENT PURGE EXHAUST. THE CONTAINI.tENT E 9. DIGITAL AND ANO ANALOG LOGIClOGIC SYI.lBOLS SYMBOLS ARE USED ON LCCIC TO FUNCTIONALLY DESCRIBE THE PROCESS CONTROL. REFER lOGIC OIAGR,\/,IS OIAGRAMS REF'ER TO THE E ASSOCIATED 'IIIRINC WIRING SCHEMATIC FOR THE ELECTRICAL Cou.PONENTS C0t.4PONENTS USEO TO IMPLEI.tENT USED IMPLEMENT THE CONTROL SCHEI.tE. SCHEME.
- 11. FOR SYMBOLS SEE INSTRUI.tENTION AIDENTlFICATION INSTRUMENT ION .. IDENTIFICATION STANDARDS, LATEST ISSUE.
REFERENCEORAWING5, REFERENCE DRAWINGS: 1-47w611-0-1---------- 1-41111611-0-' ----------
.78601-S(RI(5-------
<4- SERIES -------
1-45W600-4~-2 1_ 2 --------- I.4ATlCDIAGRA]..4 1-4~W600-90-1, --------- 1- --------- DIAGRAM 1-1-47W610-30-1 1-8--------- 1-4~W600-S7-1! --------- 1 THRU THRU6-- 6 -- DIACRAMS 1-47\1610-90-1 1-4711151 1 THRUJ-- THRIJ 3 -- GRAM J, -5, 1-47'11610-4.3-.3, 1-4711161 DIAGRAM 1-4711611-1-1---------- 1-471611-1-1 ---------- MAIN 1-47\1611-6.3-1---------S 1-47W611-6J-l --------- SAFE 1_47.611_72_1 1-47W61 1-72-1 --------- C CONTA F CONTAINMENT ISOLATION SIGNAL F PHASE A 23 l-Il-08 CONTAINMENT CONTAINI.tENT VENT ISOLATION TRAIN 9 TRAINS C-l1 COORD C-ll l-HS-4~-202A dd l-HS-~.3-202"" l-HS-4J-207A
'-HS-4J-20? A l-HS-4J-208A
'-HS-43-208A PROJECT FACIl.ITY FACILITY TYPICAL POWERHOUSE G FOR "lOrE 6 UNITS 1 & 2 PUIdP TITLE RUNNING TYPIC,I,L SAMPLING TYPICAL X
S,!,I.\PLINC SYSTEI.I SYSTEM ISOLATION VALVES ELECTRICAL FOR REACTOR COOLANT. AND ACCUM COOLANT, PRESSURIZER LID
,l,CCUI.4 TANKS. (P\lR
'-'7'610*43*' ..J (PIR TRAIN AS NOTED
.. GAS, LIO ...
NOTtD ON
'-451600*43-2).
1-471610-43-1. -3" -5 .. 1-45'1600-43-2), 1'."'.90_108 LOGIC DIAGRAM j CONTAINMENT ISOLATION
* *, .
'-FeV-90-' 08 dd l-FCV-90-109 l-Fev-90-109 l_FS 1-F5 l-FCV_90_110 1 -FeV-50-11 0 S 5
'*FCV-4J-4J4
'-FCV-4J-4.34 TYPICAL I-FCV-90_111 90-107 l-FCV-90-111 WATTS BAR NUCLEAR PLANT l-FCV-43-4,'S~
l-FCV-4J-4J~ l-FCV-4J-4J6 l-FCV-43-436 TYPICAL FOR FOR l:~g~:~g:n~ l-FCV-90-11,'S l-FCV-SO-!14 l-FCV-90-11S l-FCV-SO-115 l-FCV-90-116 I-FCV-90-116 l-FCv-90-117 1-'CV-90-117 TENNESSEE VALLEY AUTHORITY aQ NOTE 6 FO. FOR DESIGN INITIAL ISSUE ENGINEERING APPROVAL H xX TYPICAL TYPICAL. CONTAINMENT BLDG X BLOC UPPER ... L'II'R UPPER" LWR RO ISSUE PER 1 C.A. ATKINS
,OR S",I.jPLlNG SYSTEM ISOLATION VALVES TYPICAL S"'IoIPLINI; REo\CTOR COOLANT, PRESSURIZER LID FOR REACTOR LiQ a.ol GAS. eOI.lPARTMENT AIR I.IONlTOR COI.I.PARTI.IENT MONITOR ISOL VLV 1-47W610-90-3 ..
(PIfR TRAIN AS NOTED ON 1-471610-90-J (PIR FSAR FIG 6.2.4-21 & EAI 3.10 .l a:. RIMS 2 O. MURR AND ACCUU. ACCUM TANKS. (PWR 1-471610-4J-1, 1-47'11'610-43-1, -J (PIR TRAIN AS NOTED ON
- a. -5
-3 .. a.
~ 1-451600-4J-2).
1-4511600-43-2). 1-451f600-50-1 ) 1-451600-90-1) FSAR FIG 7.3-3 SH 4 T28 '92 0227 860 ROBERT D. IotURR
- 33 c.e. L YKE FOR IACB C.C. LYKE IdeB OFf' OFF' ISSUED BY, 8Y, DATE
________ ~:~._ J..9~~~ __
~.~._~t!t!~~ 1-47W611-88-1 R23
~RI!'ITSREQ'OR_
~~IIITS ~EC'O R _ Sl~
SIZ r S!Z~ sz1 II 2 3 4 5 6 7 8 BRORPIIOJAaCEE:EIoIENEF£IIESETEBLSF.BSO 9RORPRQJABCEE[t.I£NEF'£IIESET£BLU.esc HE FE HE 011 HOJ AS CE E[ ME NE UORPROJABCEEEIoiE B~ HESESE Tt eF .8 TE BI. Sf WBSQSC I CAD MAINTAINED DRAWING (CONFIGURATION CONTROL DRAWING)
11/2009 Watts 8ar SRO NRC Exam - As submitted 10/2/2009
- 89. 063 G 2.4.20089 2.4.20 089 Given the following plant conditions:
- A loss of power has occurred to 125V DC Vital 8attery 8d I.
- Operators are in the process of stopping 18-8, 2A-A and 28-8 DGs and placing in standby.
- AUO's have been dispatched by the SRO to perform the applicable portion of SOI-82 for each DG.
- All 43TL switches have been placed in the "TEST" position in accordance with the SOl's.
- Power to 125V DC Vital 8attery 8d I has NOT been restored.
Which ONE of the following describes the direction the SRO will give the AUO with regard to the position of each logic panel 43TL switch? A. Place ALL 43TL switches in "NORMAL" when directed by the SOL SOL) ru /) Il___
';:' 0./.
---'
.-----' f 8~
8:' Place 43TL switches for 18-8, 2A-A and 28-8 in "NORMAL" and leave 1A-A
>-
1A-A 43TL switch in "TEST" until the board is reenergized. ) C. Leave ALL A~ 43TL switches in "TEST" until board is reenergized. (~
~L D. Leave 43TL switches for 18-8, 2A-A and 28-8 in "TEST" until the board is)is.)
reenergized and place 1 A-A 43TL switch in "NORMAL." 1A-A 7 l~l'-~ , l.r-~l* Page 40
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, All 43TI switches would not be placed in normal because it would result is a restart of the other diesel generators if the 1A-A board '43TL' where placed to start. Plausible because during a normal shutdown when no loss of power existed the procedure would have the AUO place all 43TL switches in normal. B. Correct, The 43TL switch for the 1A-A board must be left in the TEST position to prevent restarting the other DGs when their respective 43TL switch is placed in NORMAL. The other 3 DGs would have their '43TL' switch placed in normal. C. Incorrect, Normal shutdown with no loss of power would have the AUO place the 43TL switches for each of the 4 DG in normal (not in the TEST position. Plausible because TEST is a possible switch placement and the applicant may confuse the purpose of the switch positions. D. Incorrect, the 1A-A board '43TL'
'43 TL , switch for board 1A-A would not be placed to normal and the others would not be left in test. The opposite is true. To shutdown with a loss of power the board with the power loss is left in test to prevent starting any other DG whose 43TL switch was placed to normal. Plausible because TEST is a possible switch placement and the applicant may confuse the purpose of the switch positions and reverse the required positions.
Page 41
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 89 Tier: _2_ Group 1 KIA: 063 G 2.4.20 D.C. Electrical Distribution Knowledge of the operational implications of EOP warnings, cautions, and notes. ------- Importance Rating: 3.8/4.3 10 CFR Part 55: 41.10 / 43.5 / 45.13 41.10/43.5/45.13 10CFR55.43.b: 5 KIA Match: Applicant is required to understand the conflict of direction between two in progress procedures and comply with the note in the AOI that identifies the need to mark the step as not applicable when emergenGY start signals performing the SOl to prevent undesired emergenoy being sent to the diesel generators. SRO because the question involves the hierarchy or control of procedure step implementation when one procedure has a directed step or steps of another procedure not performed and marked as not ( applicable. Technical
Reference:
AOI-21.01, Loss of 125V DC Vital Battery BD I, Rev 21 SOI-82.02, Diesel Generator (DG) 1B-B, Rev 0065 Proposed references None to be provided: Learning Objective: 3-0T-AOI2100
- 5. Describe basic procedure for shutting down unneeded DGsDGs after loss of a Vital Battery Bd.
Question Source: New Modified Bank Bank x Question History: WBN bank question SYS201 B.18 002 Comments SYS201. B.18 002 question was used on the 2006 exam. SYS201 ( Page 42
AOI-21.01 WBN LOSS OF 125V DC OC VITAL BATTERY BD BO I Revision 21 (( Page 10 of25 of 25 3.0 Loss Of 125V DC Vital Battery Bdl Bd I (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 18. SUSPEND Core Alterations.
- 19. TRANSFER Shutdown Board 125V DC control power to alternate supply USING Appendix 8. B.
- 20. PERFORM Appendix E to place ventilation systems in service.
- 21. PLACE Handswitch 43TL, in 6.9KV Panel 1A-A, in the TEST position.
Logic Panel' ( NOTE Placing the 43TL handswitch for DIG 1A-A to the NORMAL position in the following SOls, should be noted as N/A when 125V DC Vital Battery Board I is deenergized, and only performed when the vital battery board' board has been reenergized.
- 22. STOP the following DIGs, and PLACE in standby:
- a. 1B-B USING SOI-82.02, Diesel Generator (DG) 1B-B.
- b. 2A-A USING SOI-82.03, Diesel Generator (DG) 2A-A.
- c. 2B-B USING SOI-82.04, Diesel Generator (DG) 2B-8.
2B-B. (
WBN W8N Diesel Generator (DG) 18-8 1B-B SOI-82.02 501-82.02 Unit 1 Rev. 0065 Page 49 of 88 Pag_e Date _ _ __ Date* Initials 8.3 Removing DG from Service after Emergency Start NOTE This section provides instructions for for removing the DG from service and returning it to Standby Alignment after starting from either a Blackout, Safety Injection, manual emergency start or inadvertent emergency start.
- Section 8.3.1 provides instructions for removing DG from service when tied on to SD SO Bd.
- Section 8.3.2 provides instructions for removing DG from service when NOT tied on to SD SO Bd.
8.3.1 Removing DG from Service-DG Tied on to SD 8d Bd [1] IF SI signal is present, THEN RESET SI Signal before removing DG from service. [2] IF amber light, 1-XI-82-49, EMERG START LOCKOUT RELAY 86 LOR MONITOR [O-M-26] is already ILLUMINATED and 43T(L) is in NORMAL, THEN GO TO Step 8.3.1 [7] NOTE All four 43T(L) switches must be in TEST simultaneously [one in each logic cabinet for SD SO Boards]. [3] PLACE 43T(L) switch in each of the four Logic Cabinets to TEST. SO LOGIC PANEL 6.9 KV SO BO UNIO POSITION PERF VERIF INITIAL INITIAL 1A-A Logic Relay Panel (1-PNL-211-A-A) Shutdown Board 1A-A 43T(L) TEST CV Shutdown Board 2A-A Logic Relay Panel (2-PNL-211-A-A) 43T(L) TEST CV c Shutdown Board 1 1B-B B-B Logic Relay Panel (1-PNL-211-B-B) 43T(L) TEST CV Shutdown Board 2B-B Logic Relay Panel (2-PNL-211-B-B) 43T(L) TEST CV
WBN Diesel Generator (DG) 1 B-B 501-82.02 Unit 1 Rev. 0065 Page 50 of 88 Date _ _ __ Initials
----
8.3.1 Removing DG from Service-DG Tied on to SD Bd (continued) NOTE The emergency start start signal is reset when at least one of the cabinets' amber lights is illuminated. [4] ENSURE at least one of the four cabinets' amber lights illuminates after the last 43T(L) switch is placed in TEST. [5] WHEN amber light has illuminated, THEN RETURN all 43T(L) switches to NORMAL. 6.9 KV 50 SO SO LOGIC PANEL UNIO POSITION PERF VERIF INITIAL INITIAL Shutdown Board 1A-A Logic Relay Panel (1-PNL-211-A-A) 43T(L) NORMAL CV Shutdown Board 2A-A Logic Relay Panel (2-PNL-211-A-A) 43T(L) NORMAL CV Shutdown Board 1B-B Logic Relay Panel (1-PNL-211-B-B) 43T(L) NORMAL CV Shutdown Board 2B-B Logic Relay Panel (2-PNL-211-B-B) 43T(L) NORMAL CV CAUTION Prematurely resetting 86 Lockout Relay (LOR) (resetting with red light illuminated) may cause the solenoid to burn out. NOTE Voltage/frequency may change when 86 LOR is reset. [6] PERFORM the following at the DG building: [6.1] CHECK 1-RLY-82-86LOR2, Y-82-86LOR2, DG 1B-B EMERGENCY START LOCKOUT (red) indicating light is NOT illuminated [1-ARB-82-B/1, Diesel Generator 1B-B Relay Board]. [6.2] RESET 1-RLY-82-86LOR2, DG 1B-B EMERGENCY START LOCKOUT, [1-ARB-82-B/1, Diesel Generator 1B-B Relay Board].
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 , \
- 90. 103 A2.03 090 Given the following conditions:
- The plant is in Mode 6 with the reactor head set.
Draining of the Refueling Transfer Canal inside containment is in progress.
- The following Sis are on the schedule to be completed during the shift.
For Phase A 1-SI-99-605-A, "Response Time Test - Containment Isolation Phase A Slave Relay K606 - Train A." For Phase B 1-SI-99-625-A, "Response Time Test - Containment Isolation Phase B Slave Relay K625 - Train A." Which one of the following describes the implications of performing these surveillance procedures with the above plant conditions? A":I A'I The Phase A surveillance procedure poses a risk of oVerflowing overflowing the RCDT, unless the Transfer Canal drain is terminated. B. The Phase A surveillance procedure may result in the requirement to reclose ice condenser doors within 4 hours, per LCD 3.6.12, Ice Condenser Doors, due to expected cycling of fans inside containment. ( C. The Phase B surveillance procedure poses a risk of overflowing the RCDT, unless the Transfer Canal drain is terminated. D. The Phase B surveillance procedure. procedure may result in the requirement to reclose ice condenser doors within 4 hours, per LCD 3.6.12, Ice Condenser Doors, due to expected cycling of fans inside containment. Page 43
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. CORRECT. Per the listed surveillance procedure, the Reactor Coolant Drain Tank pumps will trip (if they are operating) when 1-FCV-77-10, RCDT Pump Disch Flow Control Valve, closes during the performance of the surveillance. Since the drain of the Refueling Transfer Canal goes to the RCDT, this tank may overflow without the RCDT pumps f10wpath f/owpath available. a. B. Incorrect. It is plausible for the applicant to confuse component responses to a a, Phase A actuation with those of a Phase B, and believe that performance of the Phase A A surveillance will cause changes in containment ventilation f10wpaths f/owpaths and fan operation, resulting in the pressure transients discussed. However, it is the a Phase B surveillance that contains a cautionary statement that the cycling of fans (due to the expected actuations when performing the surveillance) could cause pressure transients inside containment which will wiff result in opening of the ice condenser doors. C. Incorrect. It is plausible for the applicant to confuse component responses to a a Phase B actuation with those of a Phase A A and think that performance of the Phase aB surveiffance surveillance willwiff cause isolation of the RCDT pump f1owpath. f/owpath. However, it is the Phase A surveillance which causes this. The Reactor Coolant Drain Tank pumps wiff will trip (if they are operating) when 1-FCV-77-10, RCDT Pump Disch Flow Control ( Valve, closes during the performance of the surveillance. surveiffance. Since the drain of the Refueling Transfer Canal goes to the RCDT, this tank may overflow without the RCDT pumps f10wpathf/owpath available. D. a Incorrect. Plausible, sine performance of the Phase B surveiffance surveillance does actually result in pressure transients inside containment (due to cycling of fans caused by the surveillance), surveiffance), and the potential for undesired ice condenser door operation. However, the applicant must recognize that thfj thfJ given plant mode is one in which the LCO does not apply. LCO 3.6.12 is applicable in Modes 1, 2, 3, and 4, not Mode 6. Question Number: 90 Tier: _2_ _2__ Group 1 KIA: 103 A2.03 Containment Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: ~se Phase A...and B isolation
-~-.----.-
-----~,
Page 44
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 10 CFR Part 55: 41.5 I/ 43.5 I/ 45.3 I/ 45.13 10CFR55.43.b: 5 KIA Match: The KIA specified Phase A AND B isolation. This question presents the applicant with two plant conditions involving surveillances which test Phase A and B isolation actuation. The applicant must apply system knowledge, and make a prediction on the impact of the actuations, including a procedurally directed action to address the impact. SRO only - see comments below Technical
Reference:
1-SI-99-605-A, "Response Time Test - Containment Isolation Phase A Slave Relay K606 - Train A." 1-SI-99-625-A, "Response Time Test - Containment Isolation Phase B Slave Relay K625 - Train A." LCO 3.6.12, "Ice Condenser Doors." Proposed references None to be provided: Learning Objective: 3-0T-SYS088A
- 9. Given a set of plant conditions, determine the correct
( response of the Containment Isolation System. Question Source: New x X Modified Bank Bank
--
Question History: . New question Comments: KIA Match and SRO only: The applicant must have at least a basic familiarity with several procedures, including the surveillance procedures listed in the stem, and the procedure for draining the Refueling Transfer Canal, including where Canal,including the drain water is routed to; to: The procedure usage aspect of the KIA is further addressed in the fact that the surveillance procedures contain cautions regarding the potential consequences as detailed in each of the distractors. The SRO only aspect of this question is to assess plant conditions and to make a decision on which procedure should be used; in this case, which surveillance procedure can be authorized and with minimal adverse impact on containment conditions; i.e., overflowing a tank, as compared to changing airflows causing pressure Page 45
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Comments: KIA Match and SRO only: The applicant must have at least a basic familiarity with several procedures, including the surveillance procedures listed in the stem, and the procedure for draining the Refueling Transfer Canal, including where the drain water is routed to. The procedure usage aspect of the KIA is further addressed in the fact that the surveillance procedures contain cautions regarding the potential consequences as detailed in each of the distractors. The SRO only aspect of this question,is to assess plant conditions and to make a decision on which procedure should be used; in this case, which surveillance procedure can be authorized and with minimal adverse impact on containment conditions; i.e., overflowing a tank, as compared to changing airflows causing pressure transients which may result in undesired ice condenser door operation. ( Page 46
WEN WBN RESPONSE TIME TEST - CONTAINMENT I-SI-99-605-A 1 ISOLATION PHASE A SLAVE Revision 5 RELAY K605 - TRAIN A Page 6 of 32
,
3.0 PRECAUTIONS AND LIMITATIONS A. This procedure has been classified as having CRITICAL sections or steps. (i.e. the actions/inactions of an individual if performed incorrectly, will affect reactivity, generation and/or harm plant equipment). The sections or steps will be identified prior to section or step performance with **** CRITICAL SECTION **** or
**** CRITICAL STEP ****.
B. This Instruction can only be performed in Modes 5 and 6. C. Slave Relay Test Panel (SRTP) must be installed at 1-R-52, l-R-52, Safeguards Test Cabinet, in accordance with TI-25. D. SSPS Train B must be in service with GENERAL WARNING lamp off while SSPS Train A is removed from service. E. SLAVE RELAY RESET green light, located on SRTP, must be LIT before turning slave relay selector switches SlOl, S101, S102, and S103. F. To prevent pump damage when 1-FCV-77-128, REAC BLDG SUMP DISCH FLOW CONTROL, closes, the following pumps will be required to be off during performance of this Instruction:
- 1. 1-PMP-77-125A, RB FLR/EQ DRAIN SUMP PUMP 1A.
- 2. 1-PMP-77-125B, RB FLR/EQ DRAIN SUMP PUMP lB.
( The sump level should be lowered before the pumps are secured. G. Reactor Coolant Pump seal water can overflow the RCP No. 3 seal splashguard and spill onto the Reactor Building floor if the Reactor Coolant Drain Tank (RCDT) pressure exceeds 5.0 psig. H. The Reactor Coolant Drain Tank pumps will trip (if operating) when 1-FCV-77-10, RCDT PUMP DISCH FLOW CONTROL VALVE, closes during the performance of this Instruction. I. This Instruction should not be performed if the refueling canal is being drained or the pressurizer relief tank drain and excess letdown flow are required to be imposed simultaneously on RCDT pumps. J. 1-PMP-3-118-A AUX FEEDWATER PUMP 1B-B, must be off in order to open steam generator blowdown sample valves. 1-SI-99-605-A
WBN RESPONSE TIME TEST - CONTAINMENT 1-SI-99-625-A 1 ISOLATION PHASE B SLAVE RELAY Revision 7 K625 - TRAIN A Page 70f28 3.0 PRECAUTIONS AND LIMITATIONS (Continued) F. Performance of this Instruction will:
- 1. Isolate the following coolers:
- a. Lower Containment Vent Cooler A.
- b. Lower Containment Vent Cooler B.
- c. Lower Containment Vent Cooler C.
- d. Lower Containment Vent Cooler D.
- e. CRDM Vent Cooler A.
- f. CRDM Vent Cooler B.
- g. CRDM Vent Cooler C.
- h. CRDM Vent Cooler D.
- i. Upper Containment Vent Cooler A.
j . Upper Containment Vent Cooler C.
- k. RCP Motor 1 Coolers.
- l. RCP Motor 2 Coolers.
- m. RCP Motor 3 Coolers.
- n. RCP Motor 4 Coolers.
- 2. The following fans will be secured during performance of this Instruction:
- a. l-FAN-30-74, LWR CNTMT CLR A-A 1-FAN-30-74,
- b. l-FAN-30-77, LWR CNTMT CLR C-A 1-FAN-30-77,
- c. l-FAN-30-83, CRDM CLR A-A 1-FAN-30-83,
( d. l-FAN-30-88, CRDM CLR C-A 1-FAN-30-88,
- e. l-FAN-30-95, UPR CNTMT CLR A 1-FAN-30-95,
- f. l-FAN-30-99, UPR CNTMT CLR C 1-FAN-30-99,
- g. l-FAN-30-38-A, CNTMT AIR RETURN FAN 1 A:'" A 1-FAN-30-38-A, G. Starting and stopping fans inside containment, particularly lower containment, can cause pressure transients resulting in opening of ice condenser doors. Caution should be observed when performing this Instruction.
__ 1-SI-99-625-A _ _ __ __ __ __ ____ ____ __ ___ __ ______ ____ ____ __ __ __ __ ___
WBN RESPONSE TIME TEST - CONTAINMENT 1-SI-99-625-A 1 ISOLATION PHASE B SLAVE RELAY Revision 7 K625 - TRAIN A Page 60f28 3.0 PRECAUTIONS AND LIMITATIONS A. This Instruction can only be performed in Modes 5 and 6. Technical Specification Section 3.3.6, Condition C and Section 3.9.8 contain limitations on testing SSPS '1ith with unit in Mode 6 and movement of irradiated fuel assemblies within containment in progress. B. Slave Relay Test Panel (SRTP) must be installed at l-R-52, Safeguards Test Cabinet, in accordance with TI-25. C. SSPS Train B must be in service with GENERAL GENERAL WARNING lamp off while SSPS Train A is removed from service. D. SLAVE RELAY RESET (green light), located on SRTP, must be LIT before turning slave relay selector switches SlOl, S101, Sl02, S102, and Sl03. S103. E. The following valves will receive an ESF actuation signal to close when Slave Relay K625 Train A is energized:
- 1. I-FCV-67-87, LWR CNTMT A CLRS DISCH ISOL VLV INSIDE CNTMT.
1-FCV-67-87,
- 2. 1-FCV-67-95, I-FCV-67-95, LWR CNTMT C CLRS DISCH ISOL VLV INSIDE CNTMT.
- 3. 1-FCV-67-97, I-FCV-67-97, LOWER CNTMT C COOLER SUPPLY ISLN VLV IC.
- 4. I-FCV-67-99, LWR CNTMT B COOLERS SUPPLY ISOL VLV.
1-FCV-67-99, ( 5. 1-FCV-67-107, I-FCV-67-107, LWR CNTMT D COOLERS SUPPLY ISOL VLV.
- 6. I-FCV-67-130, UPPER CNTMT VENT CLR A SUPPLY ISOL VLV.
1-FCV-67-130,
- 7. 1-FCV-67-133, I-FCV-67-133, UPPER CNTMT VENT CLR C SUPPLY ISOL VLV.
- 8. 1-FCV-67-295, I-FCV-67-295, UPPER CNTMT VENT CLR A ISOL VLV INSIDE CNTMT.
- 9. 1-FCV-67-296, I-FCV-67-296, UPPER CNTMT VENT CLR C ISOL VLV INSIDE CNTMT.
1-SI-99-625-A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
WBN RESPONSE TIME TEST - CONTAINMENT 1-SI-99-625-A 1 ISOLATION PHASE B SLAVE RELAY Revision 7 ( K625 - TRAIN A Page 70f28 3.0 PRECAUTIONS AND LIMITATIONS (Continued) F. Performance of this Instruction will: l.
- 1. Isolate the following coolers:
- a. Lower Containment Vent Cooler A.
- b. Lower Containment Vent Cooler B.
- c. Lower Containment Vent Cooler C.
- d. Lower Containment Vent Cooler D.
- e. CRDM Vent Cooler A.
- f. CRDM Vent Cooler B.
- g. CRDM Vent Cooler C.
- h. CRDM Vent Cooler D.
- i. Upper Containment Vent Cooler A.
j . Upper Containment Vent Cooler C.
- k. RCP Motor 1 Coolers.
- l. RCP Motor 2 Coolers.
- m. RCP Motor 3 Coolers.
- n. RCP Motor 4 Coolers.
- 2. The following fans will be secured during performance of this Instruction:
- a. 1-FAN-30-74, l-FAN-30-74, LWR CNTMT CLR A-A
- b. 1-FAN-30-77, l-FAN-30-77, LWR CNTMT CLR C-A
- c. 1-FAN-30-83, l-FAN-30-83, CRDM CLR A-A
( d. 1-FAN-30-88, l-FAN-30-88, CRDM CLR C-A
- e. l-FAN-30-95, UPR CNTMT CLR A 1-FAN-30-95,
- f. 1-l-FAN-30-99, FAN-30-99, UPR CNTMT CLR C
- g. 1-FAN-30-38"':A, l-FAN-30-38-A, CNTMT AIR RETURN FAN 1A-A lA-A G. Starting and stopping fans inside containment, particularly lower containment, can cause pressure transients resulting in opening of ice condenser doors. Caution should be observed when performing this Instruction.
1-SI-99-625-A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
Ice Condenser Doors 3.6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Ice Condenser Doors LCO 3.6.12 The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and closed. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS
-------------------------------------NOTE------------------------------------- -------------------------------------NOTE---------------------------~---------
Separate Condition entry is allowed for each ice condenser door. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ice A.l A.1 Restore inlet door to 1 hour condenser inlet OPERABLE status. doors inoperable due to being physically restrained from opening. B. One or more ice B.l B.1 Verify maximum ice bed Once per condenser doors temperature is S 27 o F. 4 hours inoperable for reasons other than -AND
-
Condition A or not closed. B.2 Restore ice condenser 14 days door to OPERABLE status and closed positions. (continued) Watts Bar-Unit 1 3.6-31 (
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 91. 028 A2.01 091 Given the following:
- Unit 1 at 100% power with the following equipment out of service:
- Auxiliary Building Gas Treatment System (ABGTS) Train A.
- Hydrogen Recombiner A.
- j1-FCV-68-332,
-FCV-68-332, Block valve for PORV 334, closed due to the PORV leaking through.
- A LOCA occurs and the operating crew is in the process of placing the Hydrogen Recombiner B in service in accordance with the Emergency Instructions.
- Containment pressure currently at 4.5 psig.
Which ONE of the following identifies ... (1) the Large Early Release Frequency (LERF) radiological risk level that existed before the LOCA occurred, in accordance with TI-124, "Equipment To Plant Risk Matrix," and (2) the required power level that would be established on the Hydrogen Recombiner B when it was placed in service? (, REFERENCE PROVIDED LERF risk Recombiner Power Level A'I Yellow A'! 63kW 63 kW B. Yellow 72 kW 72kW C. Orange 72 kW D. Orange 63 kW 63kW Page 47
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Correct, With the Hydrogen Recombiner and a pressurizer PORV out of service, there is a single train of 2 systems out of service which yields a yellow yel/ow risk for the Large Early Release Frequency (LERF) radiological risk level and the graph in T/-83.01 identifies when the Recombiner B is placed in service the power level TI-83.01 should be established at 63kW 63kW. B. Incorrect, With the Hydrogen Recombiner and a pressurizer PORV out of service, there is a single train of 2 systems out of service which yields a yellow yel/ow risk for the Large Early Release Frequency (LERF) radiological risk level and the graph in T/-83.01 identifies when the Recombiner B is placed in service the power level should not be established at 72kW (it should be 63kW). Plausible because the risk level is correct and the setting the power at 72kW is the power level that would be determined if the graph for the Recombiner A was referenced. C. Incorrect, With the Hydrogen Recombiner and a pressurizer PORV out of service, there is a single train of 2 systems out of service an orange risk is not present for the Large Early Release Frequency (LERF) radiological risk level (the risk is yel/ow) yellow) and the graph in T/-83.01 TI-83.01 identifies when the Recombiner B is placed in service the power level should not be established at 72kW (it should be 63kW). Plausible because the risk level being orange could be arrived at if the focus was on the ABGTS or the PORV being out of service because they both appear in multiple ( the. correct use of the table) and the setting the system listings (but that is not the power at 72kW is the power level that would be determined if the graph for the Recombiner A was referenced. D. Incorrect, With the Hydrogen Recombiner and a pressurizer PORV out of service, there is a single train of 2 systems out of service an orange risk is not present for the Large Early Release Frequency (LERF) radiological risk level (the risk is yellow) yel/ow) but the graph in T/-83.01 identifies when the Recombiner B is placed in service the power level should be established at 63kW. Plausible because the risk level being orange could be arrived at if the focus was on the ABGTS or the PORV being out of service because they both appear in multiple system listings (but thatis that is not the correct use of the table) and the setting the power at 63kW correct. Page 48
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 91 Tier: _2_ Group 2 KIA: 028 A2.01 Hydrogen Recombiner and Purge Control System (HRPS) Malfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Hydrogen recombiner power setting, determined by using plant data book. Importance Rating: 3.4* / 3.6* 10 CFR Part 55: 41.5/43.5/45.3/45.13 10CFR55.43.b: 4, 5 KIA Match: Applicant is required to determine the required power setting for a Hydrogen Recombiner using the plant reference graph (plant data book) and question raised to the SRO level by using the Risk Matrix procedure to evaluate risk when one of the Hydrogen Recombiners is out of service. ( Technical
Reference:
TI-124, Equipment Eq uipment to Plant Risk Matrix, Rev 17 TI-83.01, Hydrogen Recombiner Required Power-VS-Containment Pressure Curves, Rev 1 Proposed references TI-124, Equipment Eq uipment to Plant Risk Matrix, Rev 17 to be provided: - pages 21-23 TI-83.01, Hydrogen Recombiner Required Power-VS-Containment Pressure Curves, Rev 1
-ALL Learning Objective: 3-0T-SYS083A
- 11. Describe how to place the Hydrogen Recombiners in service. (Ref. SOI-83.01 Question Source:
New xX Modified Bank Bank Question History: New question Comments: Page 49
WBN Equipment to Plant Risk Matrix TI-124 Unit 1 Rev. 0017 Page 21 of of46 46 6.2.2 Large Early Release Frequency The following systems are important to reducing the "beyond design basis" radiological risk to public/personnel but are not significant contributors to the risk of core damage. These systems impact the radiological release profiles for WBN. For those activities modeled in the site PSA that could impact the probability of radiological releases the Large Early Release Frequency (LERF) should be considered. The following thresholds are established:
- RED Incremental Large Early Release probability (ILERP) greater than 1E-06 should not be entered voluntarily. If such conditions are entered it should be for very short periods of time and only with a clear detailed understanding of events that caused the risk level.
- ORANGE ILERP greater than SE-07 5E-07 but less than 1E-06, assess non quantifiable factors, establish risk management actions.
- YELLOW ILERP greater than 1E-07 but less than 5E-07, SE-07, assess non quantifiable factors, establish risk management actions.
- GREEN ILERP less than 1E-07, no separate risk management plans or approval are required.
Only use the category/group for which the top system is unavailable. These risk levels are based upon deterministic evaluation; unavailability is bounded by Tech Spec allowable out of service time. SSCs that are modeled in the PSA are identified by an asterisk. Otherwise, the functions are not modeled in the PSA. Obtain risk evaluations from Corporate PRA analysts per SPP-9.11-2 PRA Evaluation Request to validate the risk configuration based upon actual unavailability. A single train for anyone systems unavailable in a given category should be considered as GREEN. A single train for any two systems unavailable in a given category should be considered as YELLOW. A single train of any three or more in a category should be considered as ORANGE. Example: For Group 268; if A Train Hydrogen Analyzer (Permanent Hydrogen Mitigation System) is unavailable with B Train Hydrogen Recombiner unavailable the likelihood of increased post accident fission product productrelease release would be YELLOW or one incremental color change. If either train of Air Return Fans was also ( unavailable, the likelihood would be increased to ORANGE or 2 incremental color changes.
WBN Equipment to Plant Risk Matrix TI-124 Unit 1 Rev. 0017 of46 Page 22 of 46 6.2.2 Large Early Release Frequency (continued) 030 Auxiliary Building Gas Treatment SysSys Emergency Gas Treatment System Auxiliary Bldg Sec Cont Envelope Annulus Vacuum Control System Control Room Emer Ventilation Sys Auxiliary Control Air System* 030 Air Return Fan System* Hydrogen Recombiner Permanent Hydrogen Mitigation System (Igniters*, Analyzers) Ice Condenser System* 030 Control Room Emer Ventilation Sys Emergency Gas Treatment System Auxiliary Building Gas Treatment Sys Annulus Vacuum Control System Auxiliary Control Air System* ( 065 Annulus Vacuum Control System* Emergency Gas Treatment System Auxiliary Building Gas Treatment Sys Control Room Emer Ventilation Sys Auxiliary Control Air System* 065 Emergency Gas Treatment System Auxiliary Building Gas Treatment Sys Containment Isolation System* Annulus Vacuum Control System Auxiliary Control Air System* 074 Containment Spray System including recirc path* Lower Compartment Coolers Air Return Fan System* Containment Isolation System* RHR System (containment cooling via CS Hx)/RHR Spray)* Safety Injection (CS to empty RWST for Charging Bleed & Feed)*
WBN Equipment to Plant Risk Matrix TI-124 Unit 1 Rev. 0017 Page 23 of46 of 46 6.2.2 Large Early Release Frequency (continued) 083 Hydrogen Recombiner System Pressurizer PORVS* Reactor Head Vent System Air Return Fan System* Permanent Hydrogen Mitigation Sys (Igniters*, Analyzers) 268 Perm H2 Mitigation Sys (Igniters)* Hydrogen Recombiners Reactor Head Vent System Air Return Fan System* Hydrogen Analyzers Pressurizer PORVS* 268 Perm H2 Mitigation Sys (Analyzers) PHMS (lgniters)* ( Reactor Head Vent System Air Return Fan System* Hydrogen Recombiners Pressurizer PORVS* 061 Ice Condenser System* Containment Spray System including recirc path* Air Return Fans System* Containment Isolation* Residual Heat Removal (Spray and containment cooling)* 6.2.3 ORAM-SENTINEL WBN uses the ORAM-Sentinel computer program as one of the tools to assess the risk of core damage or radioactive releases to the environment for work on risk significant equipment. The ORAM-Sentinel model is a WBN Unit 1 model. The WBN Sentinel model originated from the WBN PRA, and the WBN Technical Specifications.
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT Technical Instruction TI-83.01 HYDROGEN RECOMBINER REQUIRED POWER-VS-CONTAINMENT PRESSURE CURVES Units 1 ( Revision 1 QUALITY RELATED PREPARED BY: Ernest 1. T. Haston SPONSORING ORGANIZATION: OEM APPROVED BY: Charles R. Allen EFFECTIVE DATE: 05/22/2006 LEVEL OF USE: INFORMATION (
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 2 of 14 REVISION lOGLOG REVISION OR AFFECTED CHANGE EFFECTIVE PAGE NUMBER DATE NUMBERS DESCRIPTION OF REVISION/CHANGE Rev 0 All 10CFRSO.S9 Screening Review Initial Issue. A 10CFRSO.S9 has been prepared for this procedure (Engineering Tracking # WBP-LMN-OS-019-0). Rev 1 OS/22/06 05/22/06 1,2,6, 7, 8, 9, Recombiner 1-RCB-83-1-A curve is being 10,11,13 10, 11, 13 revised since the power has changed from 48 kW to 44 kW which is outside the range of +/- 2 kW. Section S.1 and 6.1 was updated, Section 6.3 was revised to clarify the equations, Table updated, and Attachment 1 was revised to 6.3 Updated, change psig to psia. (
- ----~-- ----- -----~-
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 3 of 14 ( TABLE OF CONTENTS Section Title Page Revision Log 2 Table of Contents 3
1.0 INTRODUCTION
4 1.1 Purpose 4 1.2 Scope 4 1.3 Frequency and Conditions 4 1.4 Background 4
2.0 REFERENCES
4 2.1 Performance References 4 2.2 Developmental References 5 3.0 PRECAUTIONS AND LIMITATIONS 5 4.0 PREREQUISITE ACTIONS 6 5.0 ACCEPTANCE CRITERIA 6 ( 5.1 Acceptance Criteria for Hydrogen Recombiner A 6 5.2 Acceptance Criteria for Hydrogen Recombiner B 7 6.0 PERFORMANCE 7 6.1 Hydrogen Recombiner A 7 6.2 Hydrogen Recombiner B 8 6.3 Required Power-vs-Containment Pressure Setting After a LOCA 9 7.0 POST PERFORMANCE ACTIVITY 10 8.0 RECORDS 10 ATTACHMENT 1 Recombiner Calibration Factor Curve 11 ATTACHMENT 2 Recombiner Power Correction Factor Curve 12 ATTACHMENT 3 Required Power-vs-Containment Pressure Curve 13 for Hydrogen Recombiner A ATTACHMENT 4 Required Power-vs-Containment Pressure Curve 14 for Hydrogen Recombiner B
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE POWER-V8-CONTAINMENT Revision 1 CURVES Page 4 of 14
1.0 INTRODUCTION
1.1 Purpose This Technical Instruction (TI) provides guidelines for development and revision of Hydrogen Recombiner "Required Power vs. Containment Pressure Curve" following a Design Basis Event (DBE). These curves provide the operator with an initial setting for the recombiners, assist operations in assessing the heatup test results and reduce operator action during a DBE. 1.2 Scope A. Operation Test Instruction OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test." B. Operation Test Instruction OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test." C. Post Performance Activity D. SOI-83.01, "Containment Hydrogen System Operating Instruction 801-83.01, ( Recombiners." 1.3 Frequency and Conditions A. This instruction is to be used with the performance of OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" and OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test." B. This instruction may be used in any Mode
. 1.4 Background
- WBN has generated the "Required Power vs. Containment Pressure" curves for use in determining the Hydrogen Recombiner post-LOCA power setting. The curves were generated by multiplying the C p value determined from Figure 4-5 in the Vendor Technical Manual by the power setting determined from the Operation Test Instruction OTI-83.01 and OTI-83.02.
2.0 REFERENCES
2.1 Performance References A. None
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 5 of 14 2.2 Developmental References A. System Description Manual N3-83-4001, "Combustible Gas Control System" B. Tech Specs 3.6.5, Containment Air Temperature C. Tech Specs 3.6.7, Hydrogen Recombiners D. TVA Drawings 45W755-1 and -3 E. WBN-VTM-W120-2298, Vendor Technical Manual for Westinghouse Electric Hydrogen Recombiner System F. Westinghouse Energy System letter, WTD-D-9913 (RIMS # T33 950114 806), dated 01/13/95 and entitled "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1& 2 Hydrogen Recombiner Required Power Vs Containment Pressure Curve" G. Westinghouse Energy System letter, WTD-D-9946 (RIMS # T33 950217 ( 801), dated 02/17/95 and entitled "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1& 2 Hydrogen Recombiner Required Power Vs Containment Pressure Curve" H. Westinghouse Energy System letter, WTD-D-9964 (RIMS # T33 950309 812), dated 03/08/95 and entitled "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1& 2 Hydrogen Recombiner Required Power" 3.0 PRECAUTIONS AND LIMITATIONS None
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 6 of 14 4.0 PREREQUISITE ACTIONS 4.1 PRELIMINARY ACTIONS Performance of the following procedures: A. OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" B. OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" 4.2 SPECIAL TOOLS AND SUPPLIES None 4.3 FIELD PREPARATIONS None 4.4 APPROVALS AND NOTIFICATIONS ( None 5.0 ACCEPTANCE CRITERIA 5.1 ACCEPTANCE CRITERIA FOR HYDROGEN RECOMBINER A A. IF the POWER OUT kW rating recorded in OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" is 42 to 46 kW and the REFERENCE POWER recorded is 50.263 to 54.263 kW, THEN a revision to the "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner A" is not required. B. IF the POWER OUT kW rating recorded in OTI-83.01 OTI-83.01,, "Hydrogen Recombiner A 18 Month Heatup Test" is less than 42 or greater than 46 kW, THEN a revision to the "Required Power-vs-Containment Pressure PressureCurve Curve for Hydrogen Recombiner A" is required and new acceptance criteria ' established (+/- 2 kW of new POWER OUT). C. IF the REFERENCE POWER kW rating recorded in OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" is less than 50.263 or greater than 54.263 kW, THEN a revision to the "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner A" is required and a new acceptance criteria established (+/- 2 kW of new REFERENCE POWER).
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 7 of 14 ( 5.2 ACCEPTANCE CRITERIA FOR HYDROGEN RECOMBINER B A. IF the POWER OUT kW rating recorded in OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" is 46 to 50 kW and the REFERENCE POWER recorded is 43.936 to 47.936 kW, THEN a revision to the "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner B" is not required . required.
. B. IF the POWER OUT kW rating recorded in OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" is less than 46 or greater than 50 kW, THEN a revision to the "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner B" is required and new acceptance criteria established (+/- 2 kW of new POWER OUT).
C. IF the REFERENCE POWER kW rating recorded in OTI-83.02, "Hydrogen Recombiner B B 18 Month Heatup Test" is less than 43.936 or greater than 47.936 kW, THEN a revision to the "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner B" is required and new acceptance criteria established (+/- 2 kW of new REFERENCE POWER). 6.0 PERFORMANCE 6.1 HYDROGEN RECOMBINER A [1] DETERMINE Calibration Factor (Cd (C c ) from Attachment 1 using Containment Dry Bulb Temperature and Absolute Pressure (psia) recorded on OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" [2] RECORD the determined Calibration Factor (Cd (Cc) [3] RECORD the POWER OUT wattmeter kW reading recorded on OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" [4] CALCULATE REFERENCE POWER (PR), by multiplying the POWER OUT wattmeter kW reading by the Calibration Factor (Cd (Cc) [5] IF POWER OUT kW rating recorded in OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" is 42 to 46 kW
'OR OR
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 8 of 14 ( the REFERENCE POWER recorded is 50.263 to 54.263 kW, THEN a revision to Attachment 3 "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner A" is NOT required. [6] IF POWER OUT kW rating recorded in OTI-83.01, "Hydrogen Recombiner A 18 Month Heatup Test" is less than 42 kW or greater than 46 kW OR the REFE"RENCE REFERENCE POWER recorded is less than 50.263 kW or greater than 54.263 kW, THEN a revision to Attachment 3 "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner A" for Hydrogen Recombiner A is required. The new curve for Hydrogen Recombiner A is generated as shown in Section 6.3. 6.2 HYDROGEN RECOMBINER B [1] DETERMINE Calibration Factor (C c ) from Attachment 1 using Containment Dry (Cc) Bulb Temperature and Absolute Pressure (psia) recorded on OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" [2] RECORD the determined Calibration Factor (C (Cc) c) [3] RECORD the POWER OUT wattmeter kW reading recorded on OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" [4] CALCULATE REFERENCE POWER (PR), by multiplying the POWER OUT wattmeter kW reading by the Calibration Factor (Cc) (C c ) [5] IF POWER OUT kW rating recorded in OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" is 46 to 50 kW OR the REFERENCE POWER recorded is 43.936 to 47.936 kW, THEN a revision to Attachment 4 "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner B" is NOT required.
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 9 of 14 ( [6] IF POWER OUT kW rating recorded in OTI-83.02, "Hydrogen Recombiner B 18 Month Heatup Test" is less than 46 kW or greater than 50 kW OR the REFERENCE POWER recorded is less than 43.936 kW or greater than 47.936 kW, THEN a revision to Attachment 4 "Required Power-vs-Containment Pressure Curve for Hydrogen Recombiner B" is required. The new curve for Hydrogen Recombiner B is generated as shown in Section 6.3. 6.3 REQUIRED POWER-VS-CONTAINMENT PRESSURE SETTING AFTER A LOCA Required Power vs. Containment Pressure" curves for use in determining the Hydrogen Recombiner post-LOCA power setting curves were generated by multiplying the Cp value determined from Attachment 2 (Figure 4-5 in the Vendor Technical Manual) by the REFERENCE POWER (P R) determined from OTI-83.01 and OTI-83.02. Curves may need to be revised based on results of the applicable OTI performance. ( Post-accident Required Power (PAl For example, the new Post-accident Required Power (PAA ) is calculated (Ref. 2.2.E) as follows: Power Out during calibration ==48 kW Temperature in Containment during calibration = = 75° F Pressure in Containment during calibration ==14.0 psia Reference Power (P R ) ==(Calibration Factor (Cd) (Cd) * (Power Out during calibration) Cc ==0.92 (See Attachment 1) PR =0.92
- 48 =44.2 kW
= (Power Correction Factor after a DBE (C p >>))
- PR Required Power (P A ) =
=Pressure in containment after a Design Basis Event Pc =
From Plant Operating Records, assume Pre-LOCA Containment Pressure 14.7 0 psia pSia and Pre-LOCA Containment Temperature 110 0 F.
WBN HYDROGEN RECOMBINER REQUIR~D REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 10 of 14 Cp ==Pressure Factor determined using Attachment 2 (Figure 4-5 in the WBN-VTM-W120-2298 (Ref 2.2.E)) and WAT-D-9913 (Ref. 2.2.F) that stated C p = = 1.52 @ 24.7 psia. PAA = 1.52
- 44.2 =67.18 kW Post-accident Required Power (P A) for Recombiner A and B Post-accident Required Power (PAA ) setting for containment pressures 14.7 thru 24.7 psia are calculated and tabulated in Table 6.3 for Hydrogen Recombiners A
& B. This table uses the calibration data from performance of OTI-83.01
&
(Recombiners A) and OTI-83.02 (Recombiners B). Table 6.3 Hydrogen Hydrogen Pc Cp C Recombiner A Recombiner B. Psia Psig PA PA 14.7 0.0 1.2 62.715 55.123 16.9 2.2 1.31 68.464 60.176 18.7 4.0 -1.37 71.600 62.932 20.7 6.0 1.43 74.736 65.688 22.7 8.0 1.48 77.349 67.985 24.7 10.0 1.52 79.439 69.823 The "Required Power-vs-Containment Pressure" curves for Hydrogen Recombiner A and B are developed and included in Attachment 3 & & 4 respectively. 7.0 POST PERFORMANCE ACTIVITIES None 8.0 RECORDS (
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 11 of 14 1I ATTACHMENT 1 RECOMBINER CALIBRATION FACTOR CURVE Page 1 of 1 1.8-.-----..---r----.,----.-----r------, 1.8-.------r---..----..,----,...-----r------, Containment pressure during 1.6-1-----f.---+----i-----+----f.----I 1 .6-1----+---t----I----I----+-----l calibration 9.7 psia U 1.4-1-----f.---+----i-...""..-==T-----f.-...,..---I 1.4-1----+---t----I--:=:::;o-9----+-....,.----l
~
!l:
e ()
~ 1.2-+----+---+-----+----+----+--."".,.+- 12.2 psia 1.2-+----t---t---+----+----t----::::::::=ool_
z o
~
aI 1.0-+---::::1:;:::ooo"...:::::~t---+----+---t-_=:;:::::oo~ 14.7 psia 1.0"+---:::::1:_""""'--1---+----+---;-::::::::::::=0'1-
- J
~ ~-.-
___- r - 17.2 psia c==t==+---1---r---r--T 0.8-+----t---t---+=-"""'1'==---t-----i 0.8-1----+---t----I-=-""'f==---+-----l ( t==t==+~--1---r--r--T 19.2 psia 0.6-1-- --
----+ t---t +-------+I ---
-----I
--+ ------+--t---- -Il 60 70 80 90 100 110 120 CONTAINMENT TEMPERATURE (OF)
Figure 4-1 (Reference 2.2E) NOTE A straight edge and numerical interpolation should be used if the containment pressure falls between the constant pressure lines. For example: If containment pressure is 14.00 psia and containment temperature is 75° F: Cc @ 14.70 psi psia = a = .87 and =1.05 psi a = Cc @ 12.20 psia Therefore,
=
psi a = .87 + [(1.05 - .87) X (14.70 - 14.00)] = 0.92 Cc @ 14.00 psia (14.70 - 12.20)
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 ( CURVES Page 12 of 14 I ATTACHMENT 2 RECOMBINER POWER CORRECTION FACTOR CURVE Page 1 of 1 1.7 Pre-Loca Containment Temp. 11 0°0 in Lower {ompartment 110 Compartment and 75°F in Upper Compart-1.6 ment
./ ,
1.5 V
../ --
0.. Co.. w Ll
/.,/'"
V
/
IX 0= 1.4 0 r-to- V/" V
/
w u cr ct: I.J... IoU IX 0=
;;;) ;:)
1.3 V1 VI Vl VI ( w W ex: 0:: 0.. Co.. 1.2 1.1 1.0 14.7 16.7 18.7 20.7 22.7 24.7 26.7 28.7 POST-lOCA POST-LOCA CONTAINMENT PRESSURE (PSIA) Figure 4-5 Ice Condenser Containment Recombiner Power Correction Factor Versus Containment Pressure, 14.7 PSIA Initial Pressure
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 Page 13 of 14 ( CURVES I ATTACHMENT 3 REQUIRED POWER-VS-CONTAINMENT PRESSURE CURVE FOR HYDROGEN RECOMBINER A Page 1 of 1
-- ---
80 80 I ~~
----- ---
75 i; ! I ---~ :
~
---
I I I I ! 70 v---- : VC
!
65
/ I I,
V
/ !
I 60 I I
;
I. : 55 I I I I ( 50 I 45 I I I i 40 40 ... o 2 4 6 9 10 II 11 12 o x; 12 Containment Pressure (psig) Note: Curve based on a calculated Reference Power (PR) of 52.263 kW which is obtained from the eighteen month performance of OTI-83.01. OTl-83.01. This curve may need to be revised based on results of the OTI-83.01 performance.
WBN HYDROGEN RECOMBINER REQUIRED TI-83.01 1 POWER-VS-CONTAINMENT PRESSURE Revision 1 CURVES Page 14 of 14 I ATTACHMENT 4 REQUIRED POWER-VS-CONTAINMENT PRESSURE CURVE FOR HYDROGEN RECOMBINER B Page 1 of 1 80 75 70r-~----r---~--~--~--~--r---~--~~~--~~ 70 65 65~--~--~--~--~--~~~--~--~--~--~--~~ VV .......... ,----~---~
~....
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~
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/
50 45 40~--~--~--~--~--~--~--~--~--~--~--~~ o 2 3 4 5 6 7 8 9 10 11 12 Containment Pressure (psig) Required Power-vs-Containment Pressure For Hydrogen Recombiner B Note: Curve based on a calculated Reference Power (PR) of 45.936 kW which is obtained from the eighteen month performance of OTI-83.02. This curve may need to be revised based on results of the OTI-83.02 performance.
Watts Bar Nuclear Plant' Plant Unit 1 Operating Test Instruction OTI-83.01 Hydrogen Recombiner A 18 Month Heatup Test Revision 0008 Quality Related Level of Use: Continuous Use ( Effective Date: 09-04-2007 Responsible Organization: OPS, Operations Prepared By: A. K. Keefer Approved By: A. K. Keefer
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 2 of of1919 \ Revision Log Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change 6 06/22/05 1,2 Minor/Editorial* Changes. Minor/Editorial 3, 16, 17 Procedure was revised to utilize DEM'sOEM's newly created TI-83.01 for PER 6057 and EDC 51420. 5,14 This procedure's Figure A is reproduced and 16 maintained by DEM.OEM. This revision removes 15, 16 usage and reference to Figure A and points pOints to its relocation in TI-83.01. 16 Removed Figure A from TOC, Section 8.1"s 8.1 "s "Figures", Figure A. Added TI-83.01 to Performance Reference as 2.1.A. Revised Step 6.1 [26] to utilize newly created curve in TI-83.01, TI-83.01 , Attachment 1. Deleted use of previous revision "Figure A" which has been relocated. ( Revised Steps 7.1[1] 7.1 [1] and 7 7.1.2[b]
.1.2[b] to reflect relocation of curve in SOI-83.01 to TI-83.01, Attachment 3.
Step 7.1.2[b], corrected typo from "Recombiner B" 8" to "Recombiner A." and changed WR to WO 7.1 [3] in 7.1[3] 7 2/16/06 All This procedure has been converted from Word 95 to Word 2002(XP) using rev 6, by Austin Norris. 14,17 Changed verify to check or ensure PER 95269, removed MIG from sign-off for Data Sheet 1 step 6.1 [7.3] and Data Sheet 3 step 6.1[20.7]. 6.1 [20.7]. Correction of Data sheets not addressed in Rev 5. 8 09104/07 09/04/07 8, 10, 14, 8,10,14, Added MIG or OPS for sign off to data sheets. 17 (PER 126239).
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 Page 3 of of1919 Table of Contents
1.0 INTRODUCTION
.......................................................................................................... 4 1.1 Purpose ........................................... :............................................................................ 4
.........................................................................................................................
1.2 Scope ...................................................................................... :.... :................................ 4
............................................................................................................................
1.3 Frequency and Conditions ............................................... ,............................................ 4
............................................................................................
2.0 REFERENCES
............................................................................................................. 4 2.1 Performance Reference ................................................................................................ 4 2.2 Developmental References ........................................................................................... 4 3.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 5 4.0 PREREQUISITE ACTIONS .......................................................................................... 6 4.1 Preliminary Actions ....................................................................
................................................................... ;................................... 6 4.2 Special Tools and Equipment ....................................................................................... 6 4.3 Field Preparations ......................................................................................................... 6 4.4 Approvals and Notifications .......................................................................................... 6
( 5.0 ACCEPTANCE CRITERIA ........................................................................................... 7 6.0 PERFORMANCE .......................................................................................................... 7 6.1 Recombiner A Heatup Test (1-H2C-83-1) ..................................................................... 7 7.0 POST PERFORMANCE ACTIVITY ............................................................................ 12 7.1 Reference Power Comparisons .................................................................................. 12 8.0 RECORDS .................................................................................................................. 13 8.1 QA Records ................................................................................................................ 13 8.2 Non-QA Records ........................................................................................................
......................................................................................................... 13 Data Sheet 1: Initial Data, Recombiner A (1-H2C-83-1) ................................................. 14 Data Sheet 2: Heatup Data, Recombiner A (1-H2C-83-1) (1-H2C-83-1 )............................................... 15 Data Sheet 3: Final Data, Recombiner A (1-H2C-83-1) .................................................. 17 Source Notes ............................................................................................. 19
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 Page 4 of19 of 19
1.0 INTRODUCTION
1.1 Purpose To provide instructions for heatup testing the A Hydrogen Recombiner. This test meets a Westinghouse recommendation that the recombiners are heatup tested every 18 months. The results of this test are used by Nuclear Engineering to update chart curves in other instructions as necessary. 1.2 Scope This Instruction includes the following Sections: A. Recombiner A Heatup Test B. Post Performance Activity 1.3 Frequency and Conditions A. This instruction is to be performed on an 18 month basis. B. This instruction may be performed in any Mode. (
2.0 REFERENCES
2.1 Performance Reference A. TI-83.01, "Hydrogen Recombiner Required Power-Vs-Containment Pressure Curves" 2.2 Developmental References A. Systems Description Manual, N3-83-4001, "Combustible Gas Control System" B. Tech Specs 3.6.5, Containment Air Temperature C. Tech Specs 3.6.7, Hydrogen Recombiners D. TVA Drawings: 45W755-1 and-3 and -3 E. WBN-VTM-W120-2296, Vendor Technical Manual for Westinghouse Electric Hydrogen Recombiner System F. Westinghouse Energy Systems letter, WAT-D-9913, (RIMS # T33 950114 806), dated 01/13/95 and entitled "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 & 2 Hydrogen Recombiner Required Power vs Containment Pressure Curve"
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 5 of 19 2.2 Developmental References (continued) G. Westinghouse Energy Systems letter, WAT-D-9946, (RIMS # T33 950217 801), dated 02/17/95 and entitled "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 && 2 Hydrogen Recombiner Required Power vs Containment Pressure Curve" H. Westinghouse Energy Systems letter, WAT-D-9964, (RIMS # T33 950309 812), dated 03/08/95 and entitled "Tennessee
Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 && 2 Hydrogen Recombiner Required Power Curve".
3.0 PRECAUTIONS AND LIMITATIONS A. If the smallest difference in temperature (between any 2 thermocouples) after recombiner temperature has stabilized, exceeds 50°F, then the thermocouple is to be considered inaccurate and inoperable. B. If all 3 thermocouples are operable, an average of the 3 temperatures must be used. C. If only 2 of the thermocouples are operable, an average of the 2 temperatures must be used, disregarding the inoperable thermocouple. ( D. Recombiner temperature of 1400°F is not to be exceeded on any operable thermocouple. E. A maximum of 75 kW per recombiner is not to be exceeded exceeded... F. A maximum of 110°F average containment upper compartment air temperature is not to be exceeded. G. Portions of this test are conducted in or around hot and electrically energized equipment. H. During performance of the test, access is to be limited to test personnel in the general area of the recombiner by placing a barricade at the bottom of the stairwell leading up to the landing where the recombiner is located. A "Test in Progress Do not Enter" sign should be used to alert personnel. I. Work in a Radiological Control Area (RCA) requires the use of existing RWPs, and may require additional ALARA Preplans. Failure to follow posted Rad control requirements can cause unnecessary radiation exposure. Rad Con should be notified of work having the potential to change radiological conditions.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 6 of19 of 19 Date_ __ INITIALS 4.0 PREREQUISITE ACTIONS 4.1 Preliminary Actions [1] [1 ] RECORD Start Date and Time on Surveillance Task Sheet. [2] NOTIFY Electrical Maintenance (MEG) that assistance will be required to perform steps in this procedure. 4.2 Special Tools and Equipment [1 ] [1] ENSURE the following M&TE is AVAILABLE: A. Psychrometer, Range 60 - 120°F (as a minimum), Accuracy +/- 2°F (each thermometer), maximum difference in thermometers is 2°F B. Barometer, Range 22 - 31.5 in. Hg, Accuracy +/- 0.5% of full scale or better ( c. Voltmeter, Range 0 - 600 Vac, Accuracy +/- 0.5% of full scale or better MEG [2] ENSURE required M&TE is within its current calibration cycle as evidenced by an affixed calibration sticker. 4.3 Field Preparations [1 ] [1] ENSURE Precautions and Limitations, Section 3.0, REVIEWED. [2] REVIEW Plant procedures, processes, and programs in progress to ensure accurate configuration of components necessary for System operation/testing. 4.4 Approvals and Notifications [1 ] [1] OBTAIN SM/UNIT SRO approval to perform this Instruction on Surveillance Task Sheet.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 Page 7 of 19 Date._ __ Date INITIALS 5.0 ACCEPTANCE CRITERIA A. All required Hydrogen Recombiner controls and instrumentation function properly. B. The Reference Power (Pref), is less than 54 kW with the Recombiner operating at a steady-state average heater temperature of 1215 to 1235°F. 6.0 PERFORMANCE 6.1 Recombiner A Heatup Test (1-H2C-83-1) CAUTIONS
- 1) Temperature of 1400°F is not to be exceeded on any operable thermocouple.
- 2) When on DG power, DG load should remain 4400 kW or less.
( 3) If there is any indication Recombiner A is not operating properly (through recombiner instrumentation), the Recombiner should be shutdown, a WR written, and the other Recombiner tested.
- 4) Obtain MIG support as needed for recording wet bulb, dry bulb, and barometric pressures.
[1] CHECK 1-BKR-83-1, ELEC H2 RECOMBINER HTR 1A-A (1-HTR-83-1), [Reactor Vent Board 1A-A, Cmpt. 20] ON. [2] ENSURE POWER ADJUST potentiometer [1-M-10] set at 000. [3] CHECK the White POWER IN AVAILABLE light LIT. [4] ENSURE TEMPERATURE CHANNEL (thermocouple selector) selected to channel 1, 2, or 3. NOTE The green light, on the TEMPERATURE OUT (indicator dial), is lit when temperature is below set point. The red light is lit when temperature is above the set pOint. point. [5] ENSURE TEMPERATURE OUT (indicator dial) set on 1400°F.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 8 of 19 Date_ __ INITIALS 6.1 Recombiner A Heatup Test (1-H2C-83-1) (continued) [6] RECORD the DATE and TIME on Data Sheet 1. [7] RECORD the following on Data Sheet 1: [7.1 [7.1]] Containment Dry Bulb Temperature (near recombiner A) OPS/MIG [7.2] Containment Wet Bulb Temperature (near-recombiner A) OPS/MIG [7.3] Containment Barometric Pressure (in. Hg), (near-recombiner A) OPS/MIG [7.4] Recombiner Input Voltage (phase-to-phase), [1-PNL-83-L 159-A, A4V/782, Control Rod Drive Room] MEG ( [8] CONVERT barometric pressure from Step 6.1 [7.3] to Absolute Pressure (psia) on Data Sheet 1. [9] TURN POWER OUT (MS Starter) switch to ON, AND CHECK switch plate Red light LIT. [10] ADJUST POWER ADJUST potentiometer clockwise to obtain 5 kW on POWER OUT meter, AND MAINTAIN for 10 minutes. [11] ADJUST POWER ADJUST potentiometer to obtain 10 kWon POWER OUT meter, AND MAINTAIN for 10 minutes. [12] ADJUST POWER ADJUST potentiometer to obtain 15 kWon POWER OUT meter, AND MAINTAIN for 10 minutes. [13] ADJUST POWER ADJUST potentiometer to obtain 30 kWon POWER OUT meter, AND MAINTAIN for 10 minutes.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 9 of19 of 19 Date __ Date,_ __ __ INITIALS 6.1 Recombiner A Heatup Test (1-H2C-83-1) (continued) [14] ADJUST POWER ADJUST potentiometer to obtain 48 kWon POWER OUT meter. [15] RECORD the DATE and TIME on Data ,Sheet Sheet 2. [16] RECORD the following every 20 minutes on Data Sheet 2: [16.1] Ch 1, 2, and 3, TEMPERATURE OUT [16.2] Average TEMP OUT [16.3] POWER OUT (kW) [16.4] POWER ADJUST setting [17] MAINTAIN recombiner at 48 kW for 5 hours to allow heater temperatures to stabilize. NOTES
- 1) Periodic potentiometer adjustment is necessary to maintain the required temperature setting. Temperature is obtained by selecting an average temperature of all operable thermocouples. Temperature should rise to 1150-1400°F in approximately 4 hours depending on initial containment temperature.
- 2) Recombiner temperature should stabilize before making power changes due to the lag time between adjustment and actual temperature change.
- 3) 1 kW power change is approximately equal to 20°F recombiner temperature change.
- 4) Obtain MIG support as needed for recording wet bulb, dry bulb, and barometric pressures.
[18] IF the average heater temperature is NOT 1215 to 1235°F at the end of the 5 hour stabilization period, THEN READJUST the POWER ADJUST potentiometer UNTIL an average heater temperature of 1215 to 1235°F is obtained (allow at least 2 hour temperature stabilization between each adjustment). [19] CONTINUE recording information on Data Sheet 2 every 20 minutes.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 Page 10 of 19 Date_ __ INITIALS 6.1 Recombiner RecombinerA A Heatup Test (1-H2C-83-1) (continued) [20] WHEN Recombiner average heater temperature is stable at 1215 to 1235°F, THEN RECORD the following on Data Sheet 3: [20.1] DATE and TIME [20.2] POWER OUT wattmeter [20.3] POWER ADJUST setting [20.4] Ch 1,2,3 TEMPERATURE OUT [20.5] Containment Dry Bulb Temperature (near recombiner A) OPS/MIG [20.6] Containment Wet Bulb Temperature (near recombiner A) OPS/MIG ( [20.7] Containment Barometric Pressure (in. Hg), (near recombiner A) OPS/MIG [20.8] Recombiner Input Voltage (phase-to-phase), [1-PNL-83-L 159-A, A4V/782, Control Rod Drive Room] MEG [21] REDUCE the POWER ADJUST potentiometer to 000, AND CHECK POWER OUT meter reducing to O. [22] TURN POWER OUT (MS Starter) switch to OFF, AND CHECK switch plate Red light NOT lit. [23] ENSURE TEMPERATURE CHANNEL (thermocouple selector) selected to channel 1, 2, or 3, AND CHECK temperature is REDUCING. [24] CALCULATE the Average Temperature Out from all operable thermocouples on Data Sheet 3. [25] CONVERT barometric pressure from Step 6.1 [20.7] to Absolute Pressure (psia) on Data Sheet 3. ( \
,
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 11 of 19 Date_ __ INITIALS 6.1 Recombiner A Heatup Test (1-H2C-83-1) (continued) NOTE The Shift Technical Advisor may be requested to assist in determination of the Calibration Factor (C ec ). [26] DETERMINE Calibration Factor (Cec ) from TI-83.01 , HYDROGEN RECOMBINER REQUIRED POWER-VS-CONTAINMENTPRESSURE POWER-VS-CONTAINMENT PRESSURE CURVES, Attachment 1, Recombiner Calibration Factor Curve, using Containment Dry Bulb Temperature and Absolute Pressure (psia) recorded on Data Sheet 3, THEN RECORD the determined Calibration Factor (Cec ) on Data Sheet 3. [27] On Data Sheet 3, CALCULATE Reference Power (P ref), by (Pref), multiplying the POWER OUT wattmeter kW reading ( (6.1 [20.2]) by the Calibration Factor (Cec ) from Step 6.1 [26].
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 12 of 19 Date. Date_ __ INITIALS 7.0 POST PERFORMANCE ACTIVITY 7.1 Reference Power Comparisons [1] IF the POWER OUT kW rating recorded in Step 6.1[20.2] IFthe is 42 to 46 kW, AND the Reference Power recorded in Steps 6.1 [27] is 39.18 to 43.18 kW, THEN A revision to the REQUIRED POWER-vs-CNTMT PRESSURE curve in TI-83.01, Attachment 3, is NOT required and, Steps 7.1[2] 7.1 [2] and 7.1[3] 7.1 [3] may be marked N/A. NOTE Steps 7.1 [2] and 7.1 [3] are to be marked N/A if no ranges were exceeded as determined in (, Step 7.1[1]. [2] IF the POWER OUT kW ratings recorded in Step 6.1 [20.2] is less than 42 or greater than 46 kW, OR the Reference Power recorded in Steps 6.1 [27] is less than 39.18 or greater than 43.18 kW, THEN PERFORM the following: [2.1] WRITE a Problem Evaluation Report (PER) containing as a minimum the following information:
- This procedure number and the date it was performed.
- Results of this Heatup Test including which parameter and what range it failed to fall within.
RECORD PER number: [2.2] NOTIFY NE that the existing REQUIRED POWER-vs-CONTAINMENT PRESSURE curve in TI-83.01, Appendix 3, may need to be revised based on the results of testing Hydrogen Recombiner A.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 13 of 19 Date_ __ __ _
.....,- INITIALS 7.1 Reference Power Comparisons (continued)
[2.3] PROVIDE a copy of the results of this test to Nuclear Engineering (NE). [3] IF Reference Power recorded in Steps 6.1 [27] is 54 kW or greater, THEN
- NOTIFY SM/Unit SRO that the recombiner has failed this test and REFER TO Tech Spec. 3.6.7 for applicable LCO.
- Originate WO to Troubleshoot and Repair Recombiner.
[4] NOTIFY SM/Unit SRO that this Instruction is complete. [5] RECORD completion date and time on Surveillance Task Sheet. 8.0 RECORDS ( 8.1 QA Records The following documents are QA records and handled in accordance with the Document Control and Records Management (DCRM) program: Completed Instruction Sections Completed Data Sheets 8.2 Non-QA Records None
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 14 of 19 Data Sheet 1 (Page 1 of 1) Initial Data, Recombiner A (1-H2C-83-1) 6.1 [6] DATE: TIME: 6.1 [7.1] Containment Dry Bulb Temperature: OF
-------
6.1[7.2] 6.1 [7.2] Containment Wet Bulb Temperature: OF of
-------
6.1 [7.3] Containment Barometric Pressure: in. Hg
-------
Data Recorded by: OPS/MIG Date 6.1 [7.4] Recombiner Input Voltage: Phase A-B: Vac Phase A-C: Vac ( Phase B-C: Vac Data Recorded by: MEG Date 6.1 [8]
-----
in. Hg) 2.04 = = - - - - Absolute Pressure (psia) Step 6.1[7.3] 6.1 [7.3] TEST EQUIPMENT TVA 10 ID NO. CAL DUE DATE RANGE INITIALS NOTES: (
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 15 of 19 Data Sheet 2 (Page 1 of 2) Heatup Data, Recombiner A (1-H2C-83-1) 6.1[15] 6.1 [15] DATE: TIME: TEST TEMPERATURE OUT (OF) Average POWER POWER POWER Recorded By LENGTH TEMP OUT OUT ADJUST Hrs. Min. CH 1 CH2 CH 2 CH3 CH 3 (OF) * (kW) Setting INITIAL I DATE 0:00 0:20 0:40 1:00 1:20 1:40 2:00 2:20 2:40 3:00 3:20 3:40 4:00 4:20 4:40 5:00 5:20 5:40 6:00 6:20 6:40
* =
Average TEMP OUT = (CH 1 + CH 2 + CH 3) + 3 NOTE Once data collection is complete, line-out, initial, and date unused blank data spaces.
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 Page 16 of 19 ( Data Sheet 2 (Page 2 of 2) Heatup Data, Recombiner A (1-H2C-83-1) TEST TEMPERATURE OUT (OF) Average POWER POWER Recorded By LENGTH CH 1 CH2 CH 2 CH3 CH 3 TEMP OUT OUT ADJUST INITIAL I DATE Hrs. Min. (OF) * (kW) Setting 7:00 7:20 7:40 8:00 8:20 8:40 9:00 9:20 9:40 , 10:00 ( 10:20 10:40 11:00 11:20 11:40 12:00 12:20 12:40 13:00 13:20 13:40
* =
Average TEMP OUT = (CH 1 + CH 2 + CH 3) + 3 NOTE Once data collection is complete, line-out, initial, and date unused blank data spaces. (
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 17 of 19 Data Sheet 3 (Page 1 of 2) Final Data, Recombiner A (1-H2C-83-1) 6.1 [20.1] 6.1[20.1] DATE: TIME:
-------
6.1 [20.2] POWER OUT wattmeter: kW 6.1 [20.3] POWER ADJUST setting: 6.1 [20.4] TEMPERATURE OUT: Ch 1: OF
------- of Ch 2:
Ch 3: of Data Recorded by: SIGNATURE Date 6.1 [20.5] Containment Dry Bulb Temperature: OF
------
6.1 [20.6] Containment Wet Bulb Temperature: of ( ------ 6.1 [20.7] Containment Barometric Pressure: in. Hg
------
Data Recorded by: OPS/MIG Date 6.1 [20.8] Recombiner Input Voltage: Phase A-B: Vac Phase A-C: Vac Phase B-C: Vac Data Recorded by: MEG Date (
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 18 of 19 Data Sheet 3 (Page 2 of 2) Final Data, Recombiner A (1-H2C-83-1) 6.1[24] 6.1 [24] Average Temperature Out: of (Average Temperature Out = =(CH 1 + CH 2 + CH 3) +3) 6.1[25] 6.1 [25] ( in. Hg) + 2.04 = = Absolute Pressure (psia) Step 6.1 [20.7] 6.1[26] 6.1 [26] = Calibration Factor (Cec ) = 6.1 [27] 6.1[27] (kW) X X (Cec ) =
= (Pref)
Power Out (kW) X X Calibration Factor (Cec ) =
= Reference Power (P ref )
Step 6.1 [20.2] X X Step 6.1 [26] Step 6.1 [27] Data Recorded by: ( SIGNATURE Date TEST EQUIPMENT TVA ID NO. CAL DUE RANGE INITIALS DATE NOTES:
WBN Hydrogen Recombiner A OTI-83.01 Unit 1 18 Month Heatup Test Rev. 0008 ( Page 19 of 19
\\
Source Notes (Page 1 of 1) Implementing Requirements Statement Source Document Statement None (
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 92. 92.068 068 A2.04 092 A planned release of the Monitor Tank is in progress when the following valid alarm annunciates: 181-A WDS RELEASE LINE O-RM-122 LlQ RAD HI The release does NOT automatically terminate. O-RCV-77-43 0-RCV-77-43 remains open. Which one of the following identifies ... (1) how the Liquid Radwaste System will respond if there was no operator action taken, and theTVA (2) the TVA internal notification the Shift Manager is required to make, in addition to Site Operations Management, in accordance with SPP-3.5, "Regulatory Reporting Requirements?" A. (1) The entire contents of the Monitor Tank will be released. (2) Duty Plant Manager. B. (1) The entire contents of the Monitor Tank will be released. (2) Corporate Duty Officer. ( C!' (1) Release will continue until the Monitor Tank level lowers to approximately 10%. (2) Duty Plant Manager. D. (1) Release will continue until the Monitor Tank level lowers to approximately 10%. (2) Corporate Duty Officer. Page 50
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. It is plausible that an applicant could think that since a tank has been sampled, recirculated, and approved for release, that ALL of the tank will be released. Many tanks at power plants are capable of being completely drained. However, in this case, there is an interlock which trips the pumps on a low level. Further plausibility of this distractor is added by the fact that the required notification is correct. B. Incorrect. It is plausible that an applicant could think that since a tank has been sampled, recirculated, and approved for for release, that ALL of the tank will be released. Many tanks at power plants are capable of being completely drained. However, in this case, there is an interlock which trips the pumps on a low level. The Corporate Duty Officer is plausible since this individual does have to be notified, but NOT by the Shift Manager. Per the procedure for Regulatory Reporting Requirements, the Shift Manager notifies the Duty Plant Manager, who then makes further notifications (including the Corporate Duty Officer). C. Correct. The Monitor Tank pumps automatically trip when level in the Monitor Tank lowers to the low level alarm setpoint setpoint (10%), per the System Operating Procedure and the Alarm Response Instruction for Monitor Tank HilLo Hi/Lo level. There is a table notification matrix in Appendix D of SPP-3.5, SPP-3. 5, "Regulatory Reporting Requirements," Requirements, " ( which specifies that the Duty Plant Manager must be notified for "accidental, "accidenta/, unplanned, or uncontrolled off-site radioactive release. D. Incorrect. Plausible, since the amount of Monitor Tank contents released is correct. However, the applicant does not recognize conditions which require that the Shift Manager notify the Duty Plant Manager. The Corporate Duty Officer is plausible since this individual does have to be notified, but NOT by the Shift Manager. Per the procedure for Regulatory Reporting Requirements, the Shift Manager notifies the Duty Plant Manager, who then makes further notifications (including the Corporate Duty Officer). Question Number: 92 Tier: _2__ Group _2_ 2 KIA: 068 A2.04 Liquid Radwaste System (LRS) Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of automatic isolation. Importance Rating: 3.3 Page 51
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 10 CFR Part 55: 41.5/ 43.5 / 45.3 / 45.13 41.5/43.5/45.3/45.13 10CFR55.43.b: 5 KIA Match: The applicant must evaluate conditions involving a planned release from the Liquid Radwaste System, and determine that the release should have automatically terminated, but did NOT. As an SRO, the applicant must then apply knowledge of a procedure for reporting requirements and recognize who is to be notified by the SRO. Technical
Reference:
(Standard Program and Process) SPP-3.5, Regulatory Reporting Requirements, Rev 0021, Section 3.2.4 and Appendix D, page 2 of 2. ARI181-A, ARI 181-A, "WDS RELEASE LINE 0-RM-122 LlQ RAD HI," Rev. 31. ARI 0-l-2A, 0-L-2A, "Holdup Tank Panel," window 4, "Monitor HilLo level Tank Hillo Level Units 1&2," 1&2," Rev. 3. SOI-77.01, "Liquid Waste Disposal," Section 6.7.B., Rev. 62 Proposed references None to be provided: ( Learning Objective: 3-0T-SPP0305
- 8. Identify the _criteria criteria requiring notification per Appendix D and Appendix E of SPP-3.5. .
3-0T-SSYS077 A
- 19. discuss how processed water is released.
Question Source: New X Modified Bank Bank Question History: New question Comments: Page 52
181-A (( SOURCE SETPOINT 0-RM-90-122A determined by by* Chemistry WDS RELEASE LINE O-RM-122 LlQ RAD HI Probable Cause: 1. High activity in release line
- 2. Loss of power to control relay (WDRS)
- 3. Loss of power to ratemeter
- 4. Background radiation rise at monitor
-Background NOTE 0-RM-90-122 has associated ICS computer point R1022A.
R 1022A. Corrective Action: [1] ENSURE 0-RCV-77-43, WASTE CONDENSATE DISCHARGE RADIATION ISOLATION VALVE closed. [2] CHECK 0-RE-90-9 (ARM) [3] NOTIFY Chemistry to to-perform perform CM-9.09 "Effluent Radiation Monitor Monitor Alarm Response Guidelines". [4]* [4] NOTIFY RADCON to investigate alarm. (5] IF release was in progress, THEN [5] EVALUATE discharge lineup. [6] REFERTO REFER TO AOI-31, Abnormal Release Of Radioactive Material. [7] REFER TO EPIP-1.
References:
1-45W600-57-27 1-45W600-77-1 1-4 7W61 0-90-2 1-47W61 AOI-31 EPIP-1 CM-9.09 WBN Page 9 of 47 ARI-180-187 Rev 31
(( SOURCE SETPOINT 0-L1S-77 -206A/B 0-LlS-77-206A/B* HI: 145 in. H 2 0 (93%)-increasing (93%)increasing MONITOR 0-L1S-77-206B/A 0-LlS-77 -206B/A LO: 12 in. H 2 0 (10%) decreasing TANK HI-LO LEVEL UNIT511 &2 UNITS &2 4 NOTE Monitor Tank (MT) pumps auto stop at 10% decreasing. Probable Cause: 1.
- 1. MT ready for release
- 2. MT release complete
- 3. Drain valve open Corrective Action: [1] DETERMINE cause for alarm.
[2] IF MT level is high, THEN ENSURE MT influent is stopped/isolated, and and REFER TO SOI-77.01, Liquid Waste Disposal, for ( processing. [3] IF MT level is low, THEN ENSURE MT Pumps are off: [a] 0-HS-77-2904B1, MONITOR TANK PUMP A [0-L-2A] [b] 0-HS-77-2906B1, MONITOR TANK PUMP B [0-L-2A]
References:
1-47W61 0-62-5 1-45W760-62-7 1-47W830-1 SOI-77.01 I WBN Page 7 of 21 ARI-O-L-2A Rev 3
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 ( Page 15 of 146 6.6 Normal Operation of the Tritiated Drain Collector Tank and Pumps The two Tritiated Drain Orain Collector Tank (TOCT) Pumps have no auto start logic but will auto stop on low level in the TOCr. TOCT. The TOCT is normally aligned/configured for for its contents to be placed on recirculation with either pump. Tank operating setpoints (values listed are nominal values and have uncertainty related to setpoint and scaling tolerances): A. High level alarm at 90% B. Both pumps off at 10% C. Low level alarm at 5% Oue to the pump's physical elevation, approximately 30% level must be available in Due the TOCT to ensure sufficient suction pressure. 6.7 Normal Operation of the Monitor Tank and Pumps The two pumps have NO auto start logic but will auto stop on low level in Monitor Tank. The Monitor Tank collects processed water from the CNSI Oemin and is normally aligned/configured for its contents to be placed on recirculation with either pump. Tank operating setpoints are (values listed are nominal values and have uncertainty related to setpoint and scaling tolerances): A. High level alarm at 93% B. Both pumps off at 10% C. Low level alarm at 10% To release the Monitor Tank contents to the Cooling Tower Blowdown, the release must be processed in accordance with 0-001-90-1.
NPG Standard Regulatory Reporting Requirements SPP-3.S Programs and Rev. 0021 Processes Page 11 of 74 ( 3.2 Event or Condition Reporting (continued)
- 4. Appendix D, "Site Event Notification Matrix" contains the internal management notification requirements for plant events. If designated TVA Manager is unavailable for notification due to temporary assignment, (e.g., INPO, loanee) notification should be made to designee or next higher manager. The Operations Shift Manager is responsible for notifying Site Operations management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.
- 5. Appendix E, "Other Regulatory Reporting" contains the criteria for reporting of events or conditions to Federal and State regulatory agencies other than the NRC. Operations is responsible for making the reportability determinations and notifications for these non-NRC Federal and State regulatory agency reporting requirements.
NOTE Additional reporting guidance for deficiencies is contained in SPP-3.1.
- 6. Appendix G, "Determination of Reportability under §Part 21" contains the criteria for reporting defects in basic components in accordance with §Part 21. Licensing is responsible for making the final reportability determination and written report to NRC in accordance with §Part 21.
( 7. Appendix H, "Reporting of Decommissioning Funding" contains the criteria for notifying NRC when permanently shutting down the operation of a reactor, as required by §50.54(bb). Licensing is responsible for making the written notification to NRC in accordance with §50.54(bb).
- 8. Appendix I, "Communication with the NRC Following a Significant Operational Event" contains guidance on communications that needs to be established with the NRC within 24 to 36 hours following a significant operational event that could result in an incident investigation by the NRC. Site Licensing coordinates this communication with NRC.
- 9. Appendix J, "Internal Notification of Events Requiring Serious Accident Investigations" provides internal management notification requirements for serious accidents, as as prescribed in TVA-SPP-18.010, TVA-SPP-18.01 0, "Conduct Serious Accident Investigation."
- 10. Appendix K, "Registration Requirements for Spent Fuel Storage Cask Placed into Service," provides the minimum reporting necessary to register the use of a loaded spent fuel storage cask. Licensing is responsible for developing (with input from appropriate organizations) and submitting the letter to NRC in accordance with the CoCo
(
~ ,....-.,' NPG Standard Regulatory Reporting Requirements SPP-3.S Programs and Rev. 0021 Processes Page 43 of 74 P,!ge Appendix D 0 (Page 2 of 2) Site Event Notification Matrix Notification Requirements Event/Condition Duty Plant Plant Manager Ops. Duty Spec. Site VP Corporate Manager (ODS) Duty Officer* NRC 1 hour, 4 hour, or 8 hour phone cans. calls. Yes Yes Yes for reactor Yes Yes for 1 hour trips, shutdowns, and 4 hour transport of calls. contaminated or potentially contaminated victim to hospital and for loss of Prompt Notification System. Any unusual radiation exposure to personnel. Yes Yes No Yes Yes Accidental, unplanned or uncontrolled off-site radioactive release. Yes Yes No Yes Yes Any reasonable threat to generation. Yes Yes No Yes Yes
,
Outage critical path extensions exceeding 6 hours. Yes Yes No Yes Yes Any reactivity event or unplanned reactivity change. Yes Yes No Yes Yes
- -----
NOTE: (1) Consider items specified in Appendix Estep 2.2 of ofthis this procedure.
# Plant Manager should ensure the Plant Managers at the other NPG sites are notified so the Shift Managers and Work Week Managers at the unaffected sites can review their schedules for potential generation-risk activities that may need to be deferred.
If the Corporate Duty Officer cannot be contacted within 15 minutes, minutes, the Chief Nuclear Officer should be notified. Corporate Duty Officer (COO) The COO position is always available and is not part of the REP. This position pOSition is not intended to conflict or duplicate any responsibilities of the Central Emergency Control Center (CECC) Director under the REP or the Site Vice President.
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ( 93. 071 G 2.4.8093 Given the following Unit 1 at 100% power.
- A release of the Waste Gas Decay Tank 'C' in progress.
The following conditions occur: . RCS pressure and Pressurizer level dropping.
- TURB/AUXlRX BLDG FLOODED alarm comes in.
RHR 1 / CVCS HI TEMP PIPE BREAK alarm comes in.
- O-RM-90-1 01, Auxiliary Building Stack, trending up and alarms on Hi Rad.
1-RM-90-400, Shield Building Stack, trend up and alarms on Hi Rad.
- Area monitor recorder 1-RR-90-1 indicates radiation increasing in the Auxiliary Building.
0-RM-90-118, Waste Gas Rad Monitor, steady with no radiation increase. The operating crew manually trips the reactor and initiates Safety Injection. Which ONE of the following identifies the status of the release from the Waste Gas Tank? The Waste Gas Decay Tank Release would ... ( A. have been automatically been terminated, and ECA-1.2, "LOCA Outside Containment," would ensure that the release had terminated. B. have been automatically been terminated, and AOI-31, "Abnormal Release of Radioactivity," would ensure that the release had terminated. C. continue until a transition was made to ECA-1.2, "LOCA Outside Containment," which would direct the release to be terminated. D~ continue until AOI-31, "Abnormal Release of Radioactivity," was D!' implemented and the release directed to be terminated: terminated. ( Page 53
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 ( DISTRACTOR ANAL YSIS: A. Incorrect, the Waste Gas Decay Tank release would not have been automatically terminated (it will continue until it is terminated manually.) but ECA-1.2 does not contain the direction to terminate an on-going release. Plausible because automatic termination of the Waste Gas Decay Tank release would have occurred if radiation monitor O-RM-90-11B O-RM-90-118 had detected high radiation and the conditions support a transition to ECS-1.2, "LOCA Outside Containment" after the emergency procedure network is entered. B. Incorrect, the Waste Gas Decay Tank release would not have been automatically terminated (it will continue until it is terminated manually.) but AOI-31 does contain the direction to terminate an on-going release. Plausible because automatic termination of the Waste Gas Decay Tank release would have occurred if radiation O-RM-90-118 had detected high radiation and AOI-31 containing the steps monitor O-RM-90-11B to manually terminate the release is correct. C. Incorrect, the Waste Gas Decay Tank release will continue until it is terminated manually but ECA-1.2 does not contain the direction to terminate an on-going release. Plausible because the Waste Gas Decay Tank release continuing is correct and the conditions support a transition to ECS-1.2, "LOCA Outside Containment" after the emergency procedure network is entered. ( D. Correct, High Radiation on neither the Auxiliary Building stack nor the U-1 Shield Building Stack will cause the Waste Gas Decay Tank release to be automatically terminated but when AOI-31 is implemented it has a steps that directs the termination of anyon-going release. Question Number: 93 Tier: _2_. _2_ _ Group 2 KIA: 071 G 2.4.8 Waste Gas Disposal System (WGDS) Knowledge of how abnormal operating procedures are used in conjunction with EOPs. Importance Rating: 3.8 / 4.5 3.8/4.5 10 CFR Part 55: 41.10/43.5/45.13 10CFR55.43.b: 4, 5 4,5 KIA Match: The question requires knowledge knowledge of how AOI-31 and ECA-1.2 would ( be used in conjunction to mitigate radioactive releases from the site if a LOCA outside containment occurred while a Waste Gas release was in progress. Page 54
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 KIA Match: The question requires knowledge of how AOI-31 and ECA-1.2 would be used in conjunction to mitigate radioactive releases from the site if a LOCA outside containment occurred while a Waste Gas release was in progress. SRO because the question assesses how an AOI and the ECA will be used in conjunction when the step in E-O directs the use of both in the same step RNO column. The applicant must assess plant conditions and understand how the procedures would be used to mitigate and recover, using the knowledge of the actions directed both by the ECA and the AOI to limit the radioactive release when both are directed to be used in conjunction .. Technical
Reference:
ECA-1.2, LOCA Outside Containment, Rev. 4 AOI-31, Abnormal Release of Radioactive Material, Rev 22 E-O, Reactor Trip or Safety Injection, Rev 27 Drawing 1-47W611-77-4 Proposed references None to be provided: Learning Objective: 3-0T-AOI3100 3-0T-AOI31 00
- 2. Identify Automatic Actions that may occur during an
. Abnormal Release of Radioactive Material.
( 3. Explain Operator Actions on Abnormal Release of Radioactive Material. Question Source: New xX Modified Bank Bank Question History: New question Comments: Page 55
E-O ( WBN REACTOR TRIP OR SAFETY INJECTION Rev 27 IStep IIIIAction/Expected Response III IResponse Not Obtained
- 23. CHECK secondary side radiation IF secondary side radiation high, NORMAL: THEN
** GO TO E-3,
- NOTIFY RADPROT to survey sUNey main Steam Generator Tube Rupture.
-Steam steamlines and S/G blowdown.
- NOTIFY Chemistry to sample S/G for activity.
- OPEN S/G blowdown isolation valves, and CHECK 1-RM-90-120 and 1-RM-90-121
[M-12].
- 24. CHECK Aux Bldg radiation for loss of EVALUATE the following:
RCS inventory outside cntmt:
- RHR pipe break alarm,
- a. Area monitor recorders
- Pipe chase flooding alarm, 1-RR-90-1 and O-RR-90-12A
(
- ECCS pump room flooding alarm.
Aux Bldg pOints points NORMAL.
)J
- b. Vent monitor recorder ~j IF loss of RCS inventory outside O-RR-90-101 O-RR-90-1 01 cntmt is indicated, NORMAL trend prior to isolation. \ THEN:
- 1) REFER TO AOI-31, Abnormal Release of Radioactive Material.
- 2) ** GO TO ECA-1.2, LOCA Outside Containment.
\
-
11 of 28
AOI-31 WBN ABNORMAL RELEASE OF Revision 22 RADIOACTIVE MATERIAL Page 7 of 27 3.2 Abnormal Release of Radioactive Material in Auxiliary Building ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. EVACUATE affected area.
- 2. NOTIFY Radiological Protection to monitor affected area for radiological hazards.
- 3. DETERMINE pointpOint of release, and TERMINATE release.
- 4. IF planned release in progress, THEN:
( a. STOP release. b.NOTIFY
- b. NOTIFY Chemistry to resample batch.
- 5. EVALUATE plant release rate:
- Plant Computer
[EFF1 on TSC menu].
- EPIP-13, Initial Dose Assessment For Radiological Emergencies.
- 6. REFER TO EPIP-1, Emergency Plan Classification Flowchart.
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS ECA-1.2 LOCA OUTSIDE CONTAINMENT Revision 4 Unit 1 QUALITY RELATED REQUESTED BY: S. M. Baker SPONSORING ORGANIZATION: OPERATIONS APPROVED BY: A. K. Keefer _~1~1/2=2::....=./O-=--6_ EFFECTIVE DATE: ._~1~1/2=2;;...;;.IO...;;.-6_ LEVEL OF USE: CONTINUOUS
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 1.0 PURPOSE This Instruction provides actions to identify and isolate a LOCA outside containment. 2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications Abnormal Auxiliary Building radiation due to LOCA outside containment. 2.2 Transitions A. E-O, Reactor Trip or Safety Injection. B. E-1, Loss Of Reactor' Reactor Or Secondary Coolant. 3.0 OPERATOR ACTIONS ( 2of5
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IStep II Action/Expected Response II Response Not Obtained
- 1. ENSURE RHR suction from RCS CLOSED:
- 1-FCV-74-1 and 1-FCV-74-9.
AND
- and 1-FCV-74-8.
1-FCV-74-2 and1-FCV-74-8.
- 2. ENSURE SI pumps hot leg injection injection.
1-FCV-63-156 and 1-FCV-63-157 CLOSED.
- 3. ENSURE RCS letdown isolated:
(
- Letdown isolation 1-FCV-62-69 and 1-FCV-62-70 CLOSED.
- Excess letdown isolation .
1-FCV-62-54 and 1-FCV-62-55 CLOSED.
- 4. ENSURE RHR hot leg injection 1-FCV-63-172 CLOSED.
- 5. CHECK RCS press ** GO TO Step 14.
DROPPING or stable.
- 6. CLOSE RHR crosstie valve 1-FCV-74-33 or 1-FCV-74-35.
( 3 of 5 30f5
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IStep II Action/Expected Response II Response Not Obtained
- 7. CLOSE RHR Train A cold leg injection IF 1-FCV-63-93 failed OPEN, valve 1-FCV-63-93. THEN
** GO TO Step 10.
- 8. CHECK LOCA isolated: OPEN 1-FCV-63-93.
- RCS press rising. .** GO TO Step 1010..
- 9. ISOLATE RHR Train A:
- a. STOP RHR pump A-A, and PLACE in PULL TO LOCK.
- b. CLOSE RHR suction valve 1-FCV-74-3.
(
- c. ** GO TO Step 15.
- 10. CLOSE RHR Train B cold leg IF 1-FCV-63-94 failed OPEN, injection valve 1-FCV-63-94. THEN
** GO TO Step 13.
- 11. CHECK LOCA isolated:
CHECK: OPEN 1-FCV-63-94.
- RCS press rising. ** GO TO Step 13.
( 4 of 5 4of5
( WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IStep step II Action/Expected Response II Response Not Obtained
- 12. ISOLATE RHR Train B:
- a. STOP RHR pump B-B, and PLACE in PULL TO LOCK.
- b. CLOSE RHR suction valve 1-FCV-74-21.
- c. ** GO TO Step 15.
- 13. ENSURE RHR crosstie valves 1-FCV-74-33 and 1-FCV-74-35 OPEN.
- 14. IDENTIFY break location: NOTIFY TSC of failure to identify break location.
(
- RADPROT surveys.
- RHR pipe break I,ights lights [M-6].
- ECCS pump flows.
- Aux bldg flood alarms [M-15; light panel, Aux Bldg 757].
- Radiation area monitor recorders 1-RR-90-1 and O-RR-90-12A.
- 15. DETERMINE appropriate Instruction: NOTIFY TSC of failure to isolate break.
- IF LOCA outside cntmt isolated, THEN **
** GO TO ECA-1.1,
** GO TO E-1, Loss of RHR Sump Recirculation, Recirculation.
Loss of Reactor or Secondary Coolant.
-End -
-End-5 of 5 50f5
t-LL- L L9I1.Lt- L I "I 3 12 NOTES,
~ fOR Syt.lBOLS
- 1. FOR SYI,jBOLS AND GENERAl. \-471611-77-\,
GENERAL NOTES SEE 1-47W611-77-I. ONE_ WASTE GAS COMPRESSOR IIlL
- 2. ONE WILL BE RUN CONTINUOUSLY
~01B CDX-2A- r------.., WITH THE OTHER SERVING AS A BACKUP TO BE STARTED "HEN lITH HEADER PRESSURE EXCEEOS EXCEEDS 2 PSIG. THE COMPRESSORS 'WILL WHEN Ill BE Al.TERNATED UNIFOR~ WEAR.
E~~O~os.
~~~o~osw CDX-JA-GOX-JA- ALTERNATED FOR UNIfORM II : C 0 E 1 J. ONE DECAY TANK PRESSURE ISOLATION lSOLAT10N VALVE WILL NORIolALLY NORI.IALLY GO)(-4,o,_
GOX-+A_ a BE OPEN WITH ANOTHER SELECTED FOR STANDBY. ALL OTHER A A
'"
(LS-l0JOO!J) (L5-10300/J) A I AS F : ALL TANKS IILLBECLD5ED. WILL BE CLOSED.
.-----~~~ r-i'~J O-L5-77-950/E D-L5-77-9SDjE GDX-5A_~
GDX-,5J.,_~ rij" rif" cox-a'&'_ V GOX-SA_V .
._lC:::J-
- lC:::J- I AUTO "'-"l-- I H G C I L__'&P~p___ J J Fk;t~iCLO" COX-7A-CDX-7A- ___
--IFk2ji GDX-SA-CDX-8A-f2~
(O~~ -- GDX-S ... _ GDX-9/1_ CLOSE AUTO AUlD
-g> R ~ ANU"'L ANUAL
~
~
--g>
( LO LO SF LEVEL <LOLOSP OPEN ffo-HSf fa-HS SAt.4PLE SAI.4PLE (1.5-10308/G) (L5-10JOB/G)
~~K/L 77-11' 77-115
~~
O-LS-77-95G!F 0-LS-77-95G/F LEVEL> HI HI SP SF O-LS II (LS-IDJOE/K) (LS-10JOEjIO WOP rf FLO. IN UNIT 1 SHIELD! } - FLOWINUNIT2SHIELO FLOW IN UNIT 2 SHIELD HI SP "ISP 77-95H/J-------l 77-9~H7J-------l FLOfINUNIT1SHIELD!}- BUILOING 8UILDING EXHAUST VENT - - { BUILDING eUILDING EXHAUST VENT
> SP (1-FS-JD-150) > SP (2-FS-JO-165)
- { >SP(2-fS-JD-t65)
B (LS-IOJOA/F) (lS-l OJOA/F) III lAS ANALYZER lASANALYZER
>SP(I-FS-JO-150)
B O-L5 O-LS Ir SICNAL '5 SIGNAL 1 ~47W611-77-.5 1-4111611-77-5 iIII
~~J
§'
17-403,1, 77-403,1, UNIT 1 t SHIELD BUILDING UNIT 2.2 SHIELD BUILDING I EXHAUST VENT SELECTED EXHAUST VENT SELECTED
- I I
{0-FCV-77-245) (Q-FCV-77-245) (D-FCV-77-245) (O-,CV-77-24S) MOISTURE SEPARATOR B T I COAX1 GAXI _ _ _ _ _ ....::.SP...:'...:".:.:IC::.":...-_ _::.SP...:>...:".:.:"::.",-_ '-471611-77-8, COORD 0-10':>- 1-+71611-77-8, I
-- iIII COX-IA GDX-IA
- I
- I I
- TO DRAI~
DRAIN iI
- I I
I C-r--
- START!/o-H~
;/o-H~ ~ r-!PRESSURE SET POIMT iI
~ ~
U &-L-2 < SET START
-r---t.-=~~;~:~
-r---t.-="~~;~:~ S r-!PRESSURE POIMT I
_ F77s-"~, F77,_l1~,t~ps t~PS(TYP AL::-l~:7im~~~~IN'
"::-l~:7jm~~~~IN' "AIN
.l 1.
77-~IB 77-~lB 5 Ir a-psv O-PSV r\~
- I 77-115B 77-1158 I
I t-- I-"7L
--"7L
~;
I INSTRUMENT
- r--TO TO C-!i C,:,S I (lYP MALFUNCTION -~f-~-=:--r--'
-~f--~-=:---r--'
--
- ~ I
_ VENT 77-115
,h§~.
h§~. I
~04B
~046
-
-
0-P5 O-PS o-pcv O-Pe'" I O-PCV D-PCV r--.--Ih~""'~'"'-l~~::?
-
-
r:2~ r:2~ ,
~u",o;
\Cj.u...... 11_'0'~: ~: J ~;::===~==:::;_=--7i7~'
1 ________~;:====~==:::;_=--7i7~' ~'
~9 77-05 11-99 I Ir Ir
....'+t'.'-...'"
77-115' h-\f'......
.....\f' DECA Y GAS DECAY
~~
~;~~~58
~;~~~58
~VENf
~vw
~
~
TO DRAIN 7HO< O-PS AUTO u XX T,I,NK A TANKA D- -po
-PO I
~
[1- 77-BSD/E 77-8I1D/E GAS DECAY Df:CAY TANK A 77-.Y r------------{~ r---------------------{~ PRESS. ISOLATION
- <. I---
f--- 0-P5 O-PS (peV-IOJa ... ) VALVE (PCV-,OJSA)
,
~;,-~'_~"_ -;: : := +-+- ~Et-~ :;:~-:9S~l'V: :,"O_:7;GL~fT ~;~;~9
<'1-47.611-77-8, H*7W611-7?-!!, COORD C-J 77 -88EIO 77,-88EIO OPEN !/o-H5
!/o-HS CLOSE 77-95
- <l 77-9se 77-956 I
'------"
'-------' ~~~pt~~~
~~~pt~~h Y~l~~E
- 1 i I
(PCV-IOJ6B) iPCV-IO~6B)
- i I
( --- iI Ir O-FSV O-LSV ~~f~~~~R
~~;~~~~R ;:-
;-
i 1
...I9'-( r-
- G~r¥ yeT ::-- r-----.J 77-90 77-956 77-959 I
!I! ~~ ~~ :;~~~9 S
- 1-471611-62-3 I
!
J::?ih
- I O*PIC D-PIC 77'),9 77~9 5s o-'cv yl7 ~~ 77;J");5,....,.--~1>',<;.&
,c 6~
~~
~~
-
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O-LCV f
~~ -9~9 C-Ley E-
-
-
eves EVAP
~~n'~-62-!i ---
r' O-PCV O-PCV 77-89 77-89 O-Fev O-Fey 77-90
'Ii
""
CO~PRESSOR COMPRESSOR A VENT VENT 7 7 -'" X
'
DRAIN D R A I N ! I ~ ]1"'92) ~~~:~j:~:;~~:"':Rc:o':'rr:~:~~~~'~ET~l~~CV~~:~?~~~:r:IO~NT------f),~~~...~...~ E
,,~( ~( ~:
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- evcs eves EVAI'
-
I II J .4~J UNIT 22. 1111
~JJ~IH___ ~LEVEl
- 111I r-t--
)~ ~LEVEL ~;~~"79----t ~----------GAS --------~I~- ~:~~
wee Ba
- IGC 1-47'611-77-8, COORD E-3 1-47W611-77-8, r
-
-
_L------1 ___ L-+________________ . !L",EV",EL,- ,->"'SP'--{~;~!5BfA
'-+,----------------
< SPSP DECAY TANK a
'YO (9304)
'---}
I-r I-t.....- ccs CCS --- TO GAS ANALYZER rF
~
~~;~~~
D-F5V . --- ..!L..,EV",,,,--,->"'SP'----{~;~!5B/A I F- IOSSRST t....-
-
-_
WOS SRST ......
'77_ O-FSV S 77-86 1 IiI 1
~----------GAS DECAY TANK C --------~
'r-I-(9;50J)
(9J003) 1-471511-77-5, C-l1 I-
~
~ I ~ ~SP ;~~;:::----1 r-55 VCT UNIT 2 ........
S5 ....... rvI ~----------CAS
~----------GAS --------~
- LEV" >
LEVEl , DECAY TANK 0
*.JI 51355 510355 I I LD7'r --=- - - 55 VCT UNIT 1 _ _
S5 VCT UNIT I ........
~;~~2
~;~;2 IlIL..______' _________-..l -" I ~ LEVEL) 51' LEVEL> sp** TO I:
H 1 - r-I-- ADIIUN ADtdlN REVISED PER DCA 51 ~S5-0;5-0 OJ,
~55-0J-D AND DELETED SYSTEtd OLO SYSTEM BOUNOARY.
BOUNDARY.
- I ~----------GAS --------~
COORD B5 85 I1 DECAY TANK E ---------'J-- REV REV CHANGE REF PREPARER CHECKER APPROVED O... TE DATE
-- ~¥~~1 ~~62-5--
> 5P
>SP I H 1 - r-I-- SCALE: NTS EXCEPT AS NOTED
- f AUTO O-pSV O-PSV iI PROJECT FACILITy FACILITY G-G- f- -
- - -- -- -- -- -- -- -, , ~----------GAS POWERHOUSE
,..- r-77-92 r------+-----------CAS DECAr DECA'! TANK F ----------1-- UNITS 1 a:. 2
_ liDT -.- O-H~OPEN O-H~ OPEN I
~~i 5 I
- CLOSE
'r--
ONTROLS
'--'~-PCV 77-9.5A 77-9511 I TITLE
,--,~-PCV ELECTRICAL
~
- ONTROLS I--
kh !I SIMILAR TO r-.
-
-
-
~~!j:~~~-77-,....
'IDS ReOT
~~HW~11-77-""" ~
0-LCV-71-403 O-LCV-77-4DJ
~:OORO
~:OOfW T
TO O ! ~----------GAS DECAr DECAY TANK G --------~ LOGIC DIAGRAM VENT 77-92 M
,.. -
'r-- r-I-- WASTE DISPOSAL SYSTEM
-- ~05B---
~Q5e---
O-LSV eves TO cvcs < 51' J II LEVEL O-LS 0-L5 ______ 77-95A 77-95,1,
- X x ~;:~~5 a-LCV 77-405 ,
S 0-LCV-77-9SA O-LCV-77-95,1,
~~i~V~-~~~~
~~i2V~-~~~~ ---,---.---'!--1><:J--t'
---,----r--'!--1><:J--t'
~VEL 77-950/E 71-950/£ r-----+-----------CAS r - - - - - - + - - - - - - - - - - - C A 5 DECAY TANK H H--------~ WATTS BAR NUCLEAR PLANT
-
-
FULLY OPEN L ______ L
--
< SP5P O-LS 0-L5 77-95EIO 77-95E/O ,.. -
'r-- r-I-- 1 I TENNESSEE VALLEY AUTHORITY
--
TO OR' TO 0'" D""""""T D...".." ~~ MOISTURE SEPARATOR 8 ~----------CAS
~----------GAS --------~
DESIGN INITIAL ISSUE ENGINEERING ENCINEERING
'l",
VENT ~ -#- Vf:NT~~ DECAY TANK J APPROVAL H- 1.....(1-47'SII-77-8, L.....(1-47W6'1-77-8. COORD C-S G-S DRAFTER ICHECKER RO ISSUE PER I----r-1--1-- :,A.
!.A. TALLEY TAI::EY .'- ATKIN~_
_'- C.A. ATKIN.!_ EA.I-3.10 EAI-3.10" ~ RI~S RIMS 1 GARY BARNARD
'--- ---- -- ---- --- -- ---- ---- -- --- --
-- -- -- --!!..,,....! !- -- -- -- - ---- ..
X
..
,,PP R Rllh td lA AR RY Y WATER DESIGNER I!REVIEWER T28 '92 0225 859 2 ROBERT D.o. h!URR MURR J
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11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 94. G 2.1.41094 Given the following plant conditions:
Unit is in Mode 6 with Containment Equipment Hatch held closed by 4 bolts. Upper Containment Air Lock doors are both open.
- 50 fuel assemblies have been reloaded after a complete core off-load.
RHR Train A is in service maintaining RCS temperature at 103°F. RHR Train A loop boron concentration at last sample: 2950 ppm.
- Source Range N-131 indicates 10 cps and is selected for audible count rate indication.
- Source Range N-132 indicates 8 cps.
In accordance with FHI-7, "Fuel Handling and Movement," which ONE of the following conditions would require fuel movement to be stopped until 'condition is explained or resolved? A. Source Range N-131 and N-132 rise to 22 cps and 14 cps, respectively. B. Updated log reading on RHR Train A inlet temperature indicates 111°F. C~ Auxiliary Building Gas Treatment (ABGTS) Train A is declared inoperable. D. Updated RHR Train A boron concentrations on two consecutive samples indicate 2935 ppm. Page 56
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, NI131 doubled but NI-131 did not. So both did not increase by a factor of 2 which would have required fuel movement to be stopped. Plausible because there is criteria for stopping fuel movement if both SRMs increase by a factor of 2 or 1 SRM increases by a factor of 5. B. Incorrect, the RHR temperature changing from 103°F to 111°F111 of remains within the of 70-120°F for refueling operations. Plausible because there is an range of70-120°F acceptable temperature range for refueling operation that could be exceeded causing fuel movement to be required to be stopped. C. Correct, If the containment or annulus is open to the auxiliary building during movement of irradiated fuel inside containment, both trains of ABGTS must be operable. D. Incorrect, the boron change of 15 ppm would not require the fuel movement to be stopped. It is less that the required change of 20 ppm. Plausible because there is if the samples had indicated a decrease in boron concentration by 20 ppm then fuel movement would be required to be stopped. Question Number: 94 Tier: _3_ Group n/a KIA: G 2.1.41 Knowledge of the refueling process. Importance Rating: 2.8/3.7 10 CFR Part 55: 41.2/41.10/43.6/45.13 41.2/41.10143.6/45.13 10CFR55.43.b: 2,7 KIA Match: Applicant must determine a condition that would require the movement of irradiated fuel assemblies to be stopped during refueling operations. SRO only because the questions requires an assessment of plant lineup and component conditions with detailed information, not only in the MCR but locally on the refuel floor by the refueling SRO to determine the impact on fuel handling evolutions. Also requires knowledge of Tech Spec Bases information. Technical
Reference:
FHI-7, Fuel Handling and Movement, Rev 0031 GO-7, Refueling Operations, Rev 0030 Technical Specification 3.9.4 Bases Amendment 74 Page 57
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Proposed references None to be provided: Learning Objective: . 3-0T-SYS079A
- 12. Identify the Tech Specifications!
Specifications/ Tech. Requirements relative to Fuel Handling with regard to:
- a. Boron concentration h.
- b. Source Range Neutron Monitoring
- c. Decay time
- d. Water level over the core
- e. Communications
- f. Definition: Refueling Mode Question Source:
New Modified Bank X Bank Question History: WBN question FH10700.01 004 modified. Comments: Changed to stem conditions, new correct answer added . '\1 and located in different place than bank question correct answer, former correct answer made incorrect, and distractor wording and values modified. Page 58
QUESTIONS REPORT for ILT Bank Current wo pw forlL FHI0700.01 004 Given the following plant conditions:
- Unit is in Mode 6.
- 50 fuel assemblies have been reloaded after a complete core off-load.
- RHR Train A is in service maintaining RCS temperature at 90°F.
- RHR Train A loop boron concentration at last sample: 2950 ppm.
- SR N-131 indicates 10 cps and is selected for audible count rate indication.
- SR N-132 indicates 5 cps.
In accordance with FHI-7, "Fuel Handling and Movement", which ONE of the following conditions requires suspension of core alterations and an investigation by the Fuel Handling Supervisor (FHS) prior to resuming core alterations?
- a. SR N-131 rises to 40 cps.
- b. Updated log reading on RHR Train A inlet temperature indicates 98°F.
- c. There is a small amount of cavity seal leakage that does not prevent maintaining proper cavity level.
dd.¥'
..... Updated RHR Train A boron concentrations on two consecutive samples indicate 2920 ppm.
The correct answer is D Monday, July 13, 200912:43:37 2009 12:43:37 PM 1
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 Page 8 of 41 3.0 PRECAUTIONS AND LIMITATIONS A. ABI/CVI crosstie is required, the following During refueling operation when ABIICVI conditions apply: (Refer to Tech Spec Bases 3.3.6, 3.3.8, 3.7.12, & 3.9.8.) 3.3.6,3.3.8,3.7.12,
- 1. 1-HS-90-410-A & 415-B are placed in the REFUEL position prior to the beginning of refuel operations (Mode 5) when the Contaiment and/or Annulus is open to the Auxiliary Bldg ABSCE spaces and returned to NORMAL postion prior to entering Mode 4 from Mode 5 unless repositioned by another approved configuration document (eg. 18 Mon-th 18Month GO Test, DG Blackout Test, etc.). When moving irradiated fuel inside Containment, both trains of the Containment Purge System must be operable and at least one train must be operating, or Containment must be isolated.
- 2. 1-HS-90-41 O-A & 415-B are required to be placed in the NORMAL If 1-HS-90-410-A position during movement of irradiated fuel in containment or the auxiliary bldg, then fuel movement must be stopped or containment must be isolated.
- 3. When moving irradiated fuel in the Auxiliary Bldg with the Containment open to the Auxiliary Bldg ABSCE spaces, Containment Purge System may be operated, but all Containment ventilation isolation valves and associated instrumentation must remain operable.
- 4. Both trains of ABGTS must remain operable if Contaiment and/or Annulus is open to the Auxiliary Bldg ABSCE spaces during movement of irradiated fuel inside Containment.
B. Fuel Handling Supervisor (FHS) will ensure persons involved in fuel handling are CERTIFIED / QUALIFIED per SPP-10.8. C. Caution must be used when working around open pits. D. Preventing loose articles from falling into refueling areas is accomplished per SPP-6.5. E. If communication is lost between the Auxiliary Building, the Control Room, or Containment, fuel loading should stop UNTIL communication is reestablished. If communication is lost between the Control Room and Containment then fuel loading MUST STOP. F. The following guidelines apply for movement of fuel assemblies.
- 1. Slow speed (lowest practical speed) is required in the following zones:
- a. Guiding and seating of handling tool/gripper mast onto fuel assembly.
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 Page 9 of 41 3.0 PRECAUTIONS AND LIMITATIONS (continued)
- b. Fuel assembly bottom nozzle approximately 10 inches above to 10 inches below top of Upender, RCCA change fixture, storage cell funnel, or within 10 inches of full down.
- c. Inserting fuel-bottom nozzle approximately 10 inches above to 10 inches below the top of seated fuel assemblies and within 10 inches of full down.
- d. Fuel movements when an assembly is in close contact with another assembly.
- e. Fuel movements outside of zones above, when NOT in contact with fuel assemblies may adjacent fuel or clearly above the top of seated flJel be 40 ftlmin maximum speed.
- 2. Operations must be stopped immediately IF :
- a. Unanticipated count rate increase by a factor of 2 on ALL responding nuclear monitoring channels during any single loading step after the initial nucleus of 8 fuel assemblies are loaded except an anticipated change.
- b. Count-rate on ANY nuclear channel increases by a factor of 5 during any single loading step after initial nucleus of 8 assemblies are loaded except an anticipated change due to movement of a source bearing assembly or detector to fuel assembly neutronic coupling.
- c. Water in SFP or Core is not clear enough to view Fuel top Nozzles with supplemental lighting.
- d. During Core loading, the Lower Core Support Plate and Guide Pins are not visible enough with supplemental lighting to check proper engagement of the Bottom Nozzle with the Lower Core Plate pins.
- e. C Bs decreases by more than 20 ppm, on 2 successive coolant samples, until the decrease is explained.
- f. Reactor Cavity leakage is such that the FHS is concerned with maintaining adequate inventory?
G. Current plant configuration does NOT permit irradiated fuel to be moved in the cask loading area. (Reference PER 96939 and DCN 52211). H. Maximum Number of Fuel Assemblies allowed out of approved storage is: 11
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 Page 10 of of4141 3.0 PRECAUTIONS AND LIMITATIONS (continued)
- 1. One un-irradiated fuel assembly shall be allowed within the fuel-handling area. The fuel handling area includes all areas of the refueling floor where un-irradiated fuel assemblies are handled outside of metal shipping containers. The fuel-handling area also includes the new fuel storage vault and the truck bay where metal shipping containers are unloaded.
- 2. One fuel assembly shaJI be allowed within the spent fuel storage pool boundary (excluding the inspection, reconstitution, or cleaning locations with appropriate evaluation for each configuration that must be performed prior to implementation). The spent fuel storage pool boundary includes the
. cask loading area, fuel transfer canal (excluding the transfer cart), and spent fuel pool.
- 3. Three fuel assemblies shall be allowed within the refueling canal. The refueling canal includes the fuel transfer tube boundary (including the transfer cart) and the rod cluster control changing fixture. This allows for two fuel assemblies to be in the rod cluster control changing fixture while the third fuel assembly is being transferred through the fuel transfer tube, is in the Upender, or is in transit to or from the reactor cavity.
) 4. One Assembly is allowed within the Rx Vessel. Assemblies not over the Vessel area are covered in 3.0H.3 above. I. Interlocks will only be bypassed with Fuel Handling Supervisor approval. J. Manual movement of Refueling Machine may be authorized by FHS. Manual movement is when brake is released released on trolley or bridge motor and the drive shaft is manually rotated to position mast. K. Refueling Machine Operation (See FHI-4, Refueling Machine, for description and interlocks):
- 1. Machine gripper device must be lowered onto an Assembly only when the .
Amber GRIPPER UNLATCH light is LIT.
- 2. Machine Gripper tube initial movement, up or down, when handling fuel, should use slow speed.
- 3. Gripper must be left in LATCHED POSITION while NOT in use.
- 4. Hoist Encoder or Z-Z Axis Tape (if available) must be checked for gripper location before moving bridge or trolley. Disregard all lights until this is done. 33
)
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 Page 11 of 41 3.0 PRECAUTIONS AND LIMITATIONS (continued)
- 5. Hoist Encoder or Z-Z Axis Tape (if available) is required to be within 1/4 in.
of the full down before unlatch can occur. In those cases where burnup induced fuel assembly growth results in Hoist Encoder or Z-Z Axis Tape (if available) reading outside this limit, the FHS and Westinghouse Refueling Supervisor must agree assembly is full down prior to unlatch. L. Fuel Assembly Handling:
- 1. When fuel is being moved or prepared for movement, a visual watch will be maintained to ensure adequate clearances for safe fuel and Refueling Machine movement.
- 2. Use of visual aids such as Binoculars, TV Camera, or Viewing Box should be used to ensure accurate visual sightings.
- 3. Assembly and full length FRCs (Fuel Related Components) must be handled in Vertical position only, unless supported by a suitable frame such as the shipping container.
- 4. Appropriate steps must be taken to ensure all handling Tools are properly seated on Assembly before engaging, and the Assembly is fully seated
) before releasing handling Tool.
- 5. When placing an Assembly in ANY storage location, its Reference Hole pOints points SOUTHWEST. -
- 6. Refueling Machine Operator should monitor load meter at all times during lifting or lowering fuel as described below 6. Any time the following changes are observed, the MAST will be stopped and conditions or actions evaluated according to the following:
- a. If a sudden change of 100 Ibs lbs or greater is observed while .
inserting/removing Assembly, (except when Assembly is properly seated on its bottom nozzle) Operator should perform the following followings: 5: (1) Movement should be reversed 2 in. (2) Lateral crane position should be adjusted to center suspended Assembly relative to surrounding Assemblies, Core position, Transfer system Upender, or RCCA Change Fixture, as applicable. )
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 Page 12 of 41 3.0 PRECAUTIONS AND LIMITATIONS (continued)
- b. If load changes +/- 250 Ibs or greater, while inserting/withdrawing Assembly (Assembly not seated), the FHS will be notified and the Assembly must be removed from the core or fixture and examined for damage before proceeding. In the case where overload occurs on lifting of an Assembly in the core, this may require setting Assembly back in position and removing adjacent Assemblies. Adjacent Assemblies should also be examined for damage.
- 7. Fuel Handling issues should be noted on the Fuel Assembly Transfer Form (TI-7.006) in the REMARKS Column. Adverse conditions shall be identified and controlled in accordance the SPP-3.1, "Corrective Action Program". Examples of Fuel Handling issues are:
- Mast overloading; overloading,
- Bowed fuel,
- Difficulties during unloading or loading during refueling operations, fuel receipt, or fuel manipulations in the spent fuel pit.
) 8. All unusual or abnormal events such as load changes must be reported to the Fuel Reliability Assessment Team (FRAT) team leader per TI-7.004. A description of the event, mitigative actions taken, and identification of all affected Fuel Assemblies, should be recqrded in the Unit 1 Narrative Log.
- 9. Westinghouse onsite representative or Fuel Projects must be notified of observed damage to or abnormal mechanical interference between fuel and any other material.
- 10. Fuel Handling Supervisor (FHS) should minimize the number of personnel in fuel handling area when fuel is being moved.
- 11. If an Assembly is dropped or damaged, action will be be taken per AOI-29, Dropped or Damaged Fuel or Reactor Cavity Seal Failure.
- 12. If fuel movement is interrupted for an extended period, return all Assemblies to approved storage.
M. If an Assembly cannot be placed in its specified core location, it may be stored temporarily in the RCCA Change Fixture, or in an alternate core location. If an Assembly is positioned in an alternate core location, it must be adjacent to the core baffle wall and be separated from the nearest single Assembly by at least one Assembly width, and from the nearest group of 2 or more Assemblies by at least 2 Assembly widths.4 ') N. Free-standing Assemblies are NOT allowed during core load and refueling.
WBN Fuel Handling And Movement FHI-7 Unit 1 Rev. 0031 ) Page 13 of 41 3.0 PRECAUTIONS AND LIMITATIONS (continued) o. O. To achieve engagement between an Assembly and the lower core pins, a "Shoe horn" is recommended to provide lower core pin alignment. P. If the top of an Assembly to be removed from the core is out of position such that the Refueling Machine gripper fingers cannot latch the top nozzle, the
. crane position should be adjusted laterally as necessary to permit the guide pins in the gripper to enter the top nozzle guide pin holes. After engagement, the crane should be moved back to index before latching and lifting.
Q. Manual manipulation of hoist cables is permissible to properly align gripper or . Assembly; however, sudden release of cables is NOT permitted. RADCON will be notified when this is needed since these cables will be contaminated. R. If it becomes necessary to secure RHR or reduce flow to seat an Assembly in the core near the nozzles, the FHS shall remind the MCR Operator to monitor RHR temp and assure Tech Spec compliance. S. Work in a Radiological Control Area (RCA) requires the use of existing RWPs, and may require additional ALARA Preplans. Failure to follow posted Rad control requirements can cause unnecessary radiation exposure. Rad Con should be notified of work having the potential to to change radiological conditions. 1. T. Appropriate sections of the FHI series may be performed concurrently, as necessary, to support continued fuel movement or prerequisites. U. Fuel Handling Supv. may authorize use of temporary markings as reference guides. Example: Marking Encoder racks to show Upender 50/50 position. V. If the gripper is stuck in the mid-position, then the gripper should be lowered to the Emergency Disengage Plate until the slack cable light is received. Then the gripper should be cycled going gOing to the Latch position first before going to the Unlatched position. W. The gripper should be latched and the assembly seated before Board transfers that effect the 480V Reactor Vent Board 1A-A,1A-A, otherwise the gripper may need to be cycled (See Precaution 3.0V). X. When BOTH trains of ABGTS are inoperable for any reason (including loss of ABSCE boundary), irradiated fuel movement is prohibited. Y. When moving fuel inside containment, the containment penetration penetration XI (equipment hatch) configuration shall be closed in accordance with Technical Specification 3.9.4.
Containment Penetrations B B 3.9.4 BASES APPLICABLE Containment penetrations satisfy Criterion 3 of the NRC Policy Statement. SAFETY ANALYSES (continued) LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE Reactor Building Purge and Ventilation System penetrations, and the containment personnel airlocks. For the OPERABLE Reactor Building Purge and Ventilation System penetrations, this LCO ensures that these penetrations are isolable by the Containment Ventilation Isolation System. The OPERABILITY requirements for this LCO ensure that the automatic purge and exhaust valve closure times specified in the FSAR can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit. The containment personnel airlock doors may be open during movement of irradiated fuel in the containment provided that one door is capable of being closed in the event of a fuel handling accident and provided that ABGTS is OPERABLE in accordance with TS 3.7.12. Should a fuel handling accident occur inside containment, one personnel airlock door will be closed following an evacuation of containment. The LCO is modified by a Note allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls. Administrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during movement of irradiated fuel assemblies within containment, 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident, 3) penetration flow paths, penetrating the Auxiliary Building Secondary Containment Enclosure (ABSCE) boundary, are limited to less than the ABSCE breach allowance, and 4) the ABGTS is OPERABLE in accordance with TS 3.7.12. Operability of ABGTS is required to alleviate the consequences of a FHA inside containment resulting in leakage of airborne radioactive material past the open airlock or penetration flow paths prior to their closure. (continued) Watts Bar-Unit 1 B 3.9-14 Revision 37, 45,73,74 Amendment 26,35 26, 35
WBN Refueling Operations GO-7 Unit 1 Rev. 0029 ') Page 9 of62 of 62 3.1 Precautions (continued) H. The fuel transfer canal has unidentified leakage that has not been found by el 692 was noted at the U1 and U2 penetration extensive NDE. Leakage at el692 rooms and AB el 692 wall just north of the refueling water purification pumps and at the U2 Fuel Transfer Tube (FTT) penetration into the U2 annulus. During The RF03 outage, it was determined that the leak is above el 735 and that leaving a leakage path through the U2 FITFTT would reduce the leakage into el 692 to a manageable level. I. The Spent Fuel Skimmer Filter delta P may rise following removal of the Transfer Canal gate, if the Transfer Canal had been recently filled. Filtration of Transfer Canal water with the "Trinuc" filter may be required before the Transfer Canal Gate is removed.
-removed. The use of the "Trinuc" filter after fill of the Transfer Canal, will require coordination between Operations, Radcon, and Maintenance.
J. N3~68-4001 requires the minimum RCS temperature of The system description N3-68-4001 60°F for reactor vessel and reactor coolant piping during Modes 5 and 6. To ensure that all of the RCS remains between 60 and 120°F, the cold leg and hot leg wide range and RHR heat exchanger inlet temperature indications should be monitored. Although the temperature limits of 60-120°F apply during Modes 5 and 6, the RCS should be maintained >60°Fwhen de-fueled, or no mode, as well, to ensure readiness to re-enter Mode 6. With no fuel in the core, maintaining containment air temperature 60°F may be necessary to keep the RCS temperature >60°F. 3.2 Limitations A. In Mode 6, with fuel in the ReactorVessel, Reactor Vessel, 2 Source Range (SR) channels must be in operation and OPERABLE, both channels should be recording on 1-NR-92-145, or trending on the ICS computer. B. SOURCE RANGE HI FLUX AT SHUTDOWN alarm must be in operation when the reactor is shut down with fuel in the Reactor Vessel. C. A method shall be available for monitoring the Cntmt Pit Sump (Keyway) for leakage. Potential methods include:
- 1. Using personnel to monitor for leak detection, or
- 2. Having a TV camera installed for remote leak detection. (Once the incore probe guide tubes are pulled, this area may be inaccessible due to High Radiation area).
D. Reactor Vessel head is not to come into contact with the refueling water.
)
WBN Refueling Operations GO-7 Unit 1 Rev. 0029 Page 10 of 62 3.2 Limitations (continued) E. In the worst case design basis single active failure scenario, spent fuel pool temperature will not exceed 1590F .11 F. Security must be notified prior to opening SFP to Transfer Canal Weir Gate. Radiation Protection: G. Radiation in Cntmt and fuel storage areas must be monitored continuously. H. Radiation near the surface of the water in the refueling cavity and SFP will be monitored. I. Exposure to the underside of Rx Vessel Head must be avoided. J. Rad Con Manager approval required for entering the following areas UNTIL radiation Dose Rates are determined by RADCON and the area cleared for entry:
- 1. Area under Rx Vessel.
- 2. Areas adjacent to Fuel Transfer Tube, particularly the North side, and underneath the cavity in lower Cntmt.
- 3. Annulus.
K. Before Tools and equipment are withdrawn from the refueling water, Radcon will be informed and Radcon requirements be complied with.
WBN Refueling Operations GO-7 Unit 1 Rev. 0029 Page 11 of 62 Date _ _ _ __ INITIALS 4.0 PREREQUISITES NOTE Prerequisites may be complied with in any order. If a prerequisite cannot be complied with, the SM shall initial, date, and write a brief explanation. Prerequisites that contain a must, shall, or will, cannot be N/A'd. [1] ENSURE all Fuel and Fuel Related Components (Fuel /I FRC) handling personnel and FHS are certified and trained per the requirements of SPP-10.8 9 [2] ENSURE RHR cooling IN SERVICE per SOI-74.01. NOTES
- 1) To ensure all of the RCS remains between 60 and 120°F, the cold leg and hot leg wide range and, RHR heat exchanger inlet temperature indications should be monitored.
- 2) Although the temperature limits of 60-120°F apply during Modes 5 and 6, the RCS should be maintained >60°F when de-fueled, or no mode, as well, to ensure readiness to re-enter Mode 6. With no fuel in the core, maintaining containment air temperature
>60°F may be necessary necessary to keep the RCS temperature >60°F.
[3] ENSURE RCS temperature is being maintained between 70 and 120°F. [4] ENSURE RHR letdown to CVCS and one charging pump IN SERVICE to maintain RCS chemistry. [5] ENSURE Section 6.4 of TI-5.007, Containment Purge Damper Adjustments For Pressure Control, has been completed during Mode 5. [6] INITIATE Appendix A, Mode 5-To-Mode 6 Checklist. [7] OBTAIN chemistry sample results from Chemistry to verify RCS and PZR DEGASSED and Specific Activity acceptable to open the RCS per CM-5.09. [8] ENSURE Incore Detectors in STORAGE in the polar crane wall. [9] ENSURE Cntmt Pit Sump (Keyway) checked for Operability, Cleanliness, Lighting, and Power BEFORE pulling Incore Probe Guide tubes.
11/2009 Watts 8ar Bar SRO NRC Exam - As submitted 10/2/2009
- 95. G 2.2.22095 2.2.22 095 Given the following conditions;
- Unit 1 is at 100% power.
- Monday at 0700 - RHR pump 18-8 1B-B is tagged for a scheduled 36 hour component outage.
- Tuesdayat 0300 - CCP 1A-A tripped and is damaged.
B-B is started and the plant stabilized. CCP 18-8 Which of the following identifies the latest time the plant could be placed in Mode 3 and be in compliance with Technical Specifications requirements? A. Tuesday at 0900 if neither pump has been restored to operable status. B. Tuesday at 1000 if neither pump has been restored to operable status. 8. C!' Friday at 0900 if RHR pump 18-8 1B-B is restored to service on Tuesday at 1900. D. Friday at 1300 if RHR pump 18-81B-B is restored to service on Tuesday at 1900. )
) /
Page 59
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, although ECCS equipment of both Trains is inoperable, the LCO has allowance for the condition provided there is the equivalent of 100% flow of a single Operable ECCS Train available. Therefore it is not an LCO 3.5.2 Action B entry. Plausible because Action B does allow 6 hours to get to Mode 3. B. Incorrect, although ECCS equipment of both Trains is inoperable, the LCO has allowance for the condition provided there is the equivalent of 100% flow of a single Operable ECCS Train available. Therefore it is not an LCO 3.0.3 entry. Plausible because LCO 3.0.3 requires Mode 2 entry within 7 hours. C. Correct, LCO 3.5.2 Action A allows a 72 completion time when EGCS ECCS components on both Trains inoperable provided there is the equivalent of 100% flow of a single Operable ECCS Train available. With the Train A A RHR pump, the Train B CCP and both SIPs operable the 100% flow equivalent exists. The LCO would have been entered at 0700 on Tuesday and not exited. Using the allowance for an LCO completion time extension as provided in in TIS Section 1.3, Completion Times, the completion time for the CCP would be the more restrictive of either the 72 hours after it was discovered to be inoperable (expire Friday at 0300) or an additional 24 hours after the initial entry into the condition (Friday at 0700). Friday at 0300 is the more restrictive, 0300 plus 6 hours to get to Mode 3 would be Friday at 0900. D. Incorrect" LCO 3.5.2 Action A A allows a 72 completion time when ECCS components on both Trains inoperable provided there is the equivalent of 100% flow of a single Operable ECCS Train available and if the conditions identified in TIS Section 1.3 'Completion Times' exist, the time can be extended as described in
'c' above. Plausible because Friday at 1300 is the time of the initial entry plus an additional 24 hours (72 hrs + 24 hours) as described in section 1.3 plus the 6 hours to get to Mode 3. However, the additional 24 hours is not the required 'More restrictive' of the 2 choices.
j Page 60
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 95 Tier: _3_ Group n/a KIA: G 2.2.22 Knowledge of limiting conditions for operations and safety limits. Importance Rating: 4.0 14.7 I 4.7 10 CFR Part 55: 41.5 I 43.2 I 45.2 10CFR55.43.b: 2 KIA Match: Applicant must recall the limitations for operations that exist when ECCS equipment on both trains is inoperable simultaneously and the restriction on how the maximum out of service time is determined. SRO because the question requires knowledge of Tech Spec section 1.3, Completion Times, and requires the ability to determine when conditions exist that allow LCO completion times to be extended. Technical
Reference:
Technical Specifications 3.5.2, ECCS, Amendment 55 Technical Specifications 1.3, Completion Times Proposed references None to be provided: Learning Objective: 3-0T-SYS063A
- 28. Identify the technical specifications and bases associated with each of the following components or conditions:
- a. Cold Leg Injection Accumulators
- b. ECCS Subsystems - Tavg ~ 350 0 F
- c. ECCS Subsystems - Tavg < 350 of
- d. RWST Question Source:
New Modified Bank X X Bank Question History: WBN bank question T/SOOOO.05 modified Comments: Question stem and all four choices modified. ) Page 61
ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . 3.5.2 ECCS - Operating LCO 3.5.2 TwoECCS Two ECCS trains shall be OPERABLE.
-----------------c------------~-------,--------------NOTES------------~----------------------------
---------------.--------------------------------------N 0 TES--------------------- ----~-- -------------
- 1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours to perform pressure isolation valve testing per SR 3.4.14.1. 3.4.14.1 .
- 2. In MODE 3, the safety injection pumps and charging pumps may be made incapable of injecting to support transition into or from the Applicability of the LCO 3.4.12, Cold Overpressure Overpressure Mitigation System (COMS) for up to four hours hours or until the temperature of all the RCS cold legs exceeds 375°F, whichever occurs first.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to OPERABLE 72 hours inoperable. status. AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours Watts Bar-Unit 1 3.5-4 Amend ment 55
ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS REOUIREMENTS SURVEILLANCE FREQUENCY FREOUENCY SR 3.5.2.1 Verify the following valves are in the 12 hours listed position with power to the valve operator removed. Number Position Function FCV-63-1 Open RHR Supply FCV-63-22 Open SIS Discharge SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked,
,.
sealed, or otherwise secured in position, isin is in the correct position. SR 3.5.2.3 Verify ECCS piping is full of water. 31 days NOTE: Surveillance performance not required req ui red for fo r safety injection hot leg injection lines until start up from the Fall 2003 refueling outage. SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the IInservice nservice developed head. Testing Program I SR 3.5.2.5 Verify each ECCS automatic valve in the flow path 18 months that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. SR 3.5.2.6 Verify each ECCS pump starts automatically on an 18 months actual or simulated actuation signal.
---- - -
((continued) continued) Watts Bar-Unit 1 3.5-5 Amendment 43
ECCS - Operating 3.5.2 ) SURVEILLANCE FREQUENCY SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each I 18 months position stop is in the correct position. Valve Number CCP Discharge SI Cold Leg SI Hot Leg Throttle Throttle Throttle Valves Valves Valves 63-582 63-550 63-542 63-583 63-552 63-544 63-584 63-554 63-546 63-585 63-556 63-548 SR 3.5.2.8 Verify, by visual inspection, each ECCS train 18 months containment sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion. Watts Bar-Unit 1 3.5-6
Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND Limiting Conditions for Operation (LCOs) specify mlnlmum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO LCD state Conditions that typically describe the ways in which the requirements of the LCO LCD can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s). DESCRIPTION DESCRI PTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation s.ituation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. LCD. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO LCD Applicability. If situations are discovered that require entry into more than one Condition at a time within a single LCO LCD (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition. Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition; Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. (continued) Watts Bar-UnH Bar-Unit 1 3-1 1 .3-1 1.
Completion Times 1.3 1.3 Completion Times 1.3* iI . DESCR I PTI ON DESCRIPTION However, when a subsequent train, subsystem, component, or (continued) variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:
- a. Must exist concurrent with the first inoperability; and
- b. Must remain inoperable or not within limits after the first inoperability is resolved.
The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:
- a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours; or b*.
b~ The stated Completion Time as measured from discovery of the subsequent inoperability. The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications. The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . . " Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended. (continued) Watts Bar-Unit 1 1.3-2
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 96. G 2.2.37096 Given the following plant conditions;
- The Unit is at 100% power.
- Annunicator 131-A 'CL ACCUM 1 LEVEL HIILO'HI/LO' alarms.
- The crew determines the following changes have occurred since the last shift:
- Leakage has caused CLA #1 level to increase from 7880 gallons to 7975 gallons.
- Pressure has increased from 645 psig to 650 psig.
Which ONE of the following identifies ... (1) the Technical Specification requirement for verifying the boron concentration is within limits in the Cold Leg Accumulator and (2) the number of CLAs required to inject into the core to ensure the acceptance criteria of 10 CFR 50.46 is met during the blowdown phase of a LOCA? Tech Spec Boron Requirement CLAs Required to meet 10CFR 50.46 A. Verify within 6 hours. 2 CLAs 2CLAs B~ Verify within 6 hours. B:o" 3CLAs C. Verify within 72 hours. 2 CLAs 2CLAs D. Verify within 72 hours. 3CLAs Page 62
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, the level change of more than 75 gallons does required the verification of the CLA boron concentration with six hours but three (not two) CLAs are required to inject into the core during the blowdown phase of a large break LOCA. Plausible because the verification of boron concentration change being required within 6 hours is correct and with Tech spec allowance for one being inoperable for a time period, that 2 of the other 3 must be required to inject because the one with the leak would be spilled onto to the containment floor. B. Correct, the level has changed greater than 75 gallons, therefore a surveillance requirement to determine boron concentration (which normally has a 31 day frequency) is required to be performed once within 6 hours. As identified in LCO 3.5.2 Bases three of the four CLAs are required to inject into the core during the blowdown phase of a large break LOCA. C. Incorrect, the level change of greater than 75 gallons requires the determination of the CLA boron concentration be performed once within 6 hours and three (not two) CLAs are required to inject into the core during the blowdown phase of a large break LOCA. Plausible because 72 hours is the time a CLA can be inoperable due to boron concentration being out of limits and with Tech spec allowance for one being inoperable for a time period, that 2 of the other 3 must be required to inject because the one with the leak would be spilled onto to the containment floor. D. Incorrect, the level change of greater than 75 gallons requires the determination of the CLA boron concentration be performed once within 6 hours but three CLAs are required to inject into the core during the blowdown phase of a large break LOCA is correct. Plausible because 72 hours is the time a CLA can be inoperable due to boron concentration being out of limits and 3 CLAs being required to inject contents boron to the core is correct. Page 63
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 96 Tier: _3_ Group n/a KIA: G 2.2.37 Ability to determine operability andlor availability of safety related equipment. Importance Rating: 3.6 3.6/4.6 14.6 10 CFR Part 55: 10CFRPart55: 41.7/43.5/45.12 10CFR55.43.b: 2 KIA Match: Applicant must determine the operability status of a Cold Leg Accumulator using information contained in TS Surveillance Requirements and the TS BASES. SRO because the question requires knowledge of the conditions identified in the Tech Spec required actions section (below the line) and knowledge of the bases for the LCO Technical
Reference:
Technical Specification 3.5.1 Technical Specification 3.5.1 Bases Amendment 21, Revision 3 Proposed references None to be provided: Learning Objective: 3-0T-SYS063A
- 28. Identify the technical specifications and bases associated with each of the following components or conditions:
- a. Cold Leg Injection Accumulators Question Source:
New Modified Bank X X Bank Question History: Comments: Page 64
Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 ACCUMULATORS LCO 3.5.1 Four ECCS accumulators shall be OPERABLE. APPLICABILITY: MODES 1 and 2, MODE 3 with pressurizer pressure> 1000 psig. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One aCcumulator accumulator A.1 Restore boron concentration to 72 hours inoperable due to boron within limits. concentration not within limits. B. One accumulator B.1 Restore accumulator to 1 hour inoperable for reasons OPERABLE status. other than Condition A. C. Required Action and C.1 Be in MODE 3. 6 hours. associated Completion Time of Condition A or B AND not met. C.2 Reduce pressurizer 12 hours Pressure to ~ 1000 psig. D. Two or more accumulators D.1 0.1 Enter LCO 3.0.3. Immediately inoperable. Watts Bar-Unit 1 3.5-1
Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. I 12 hours SR 3.5.1.2 Verify borated water volume in each accumulator is I 12 hours
~ 7630 gallons and C: and::;~ 8000 gallons.
SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is I 12 hours c:~ 610 psig and ~ 660 psig and::; SR 3.5.1.4 -------------------------------- I 31 days
-------------------------------NOTE -------------------------------
The number of TPBARs in the reactor core is contained in the Core Operating Limits Report AND (COLR) for each operating cycle.
-------NOTE -----------
Only required to be Verify boron concentration in each accumulator is as performed for provided below depending on the number of tritium affected producing burnable absorber rods (TPBARs) installed accumulators. in the reactor core for this operating cycle: Once within 6 hours after each solution L Number of TPBARs I Boron Concentration Ranges I volume increase of I 0-400 I c: 3000 ppm and::; 3300 ppm I c: 75 gallons, that is
~
not the result of addition from the refueling water storage tank. (continued) ./ Watts Bar-Unit 1 3.5-2 Amendment 7,21,40,48,67
Accumulators Accu m ulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify, power is removed from each accumulator 31 days isolation valve operator when pressurizer pressure
~ 1000 psig.
is ;? Watts Bar-Unit 1 3.5-3
Accumulators B 3.5.1 B ~\ J B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA. The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere. In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water. safety The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure. Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series. The motor operated isolation valves are interlocked by P-11 with the pressurizer pressure measurement channels to ensure that the valves will automatically open as RCS pressure increases to above the permissive circuit P-11 setpoint. (continued) Watts Bar-Unit 1 B 3.5-1
Accumulators B 3.5.1 BASES BACKGROUND This interlock also prevents inadvertent closure of the (continued) valves during normal operation prior to an accident. Although not required for accident mitigation, the valves will automatically open as a result of an SI signal. These features ensure that the valves meet the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) for "operating bypasses" and that the accumulators will be available for injection without reliance on operator action. The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a L~~A. Lr~A. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. APPLICABLE The accumulators are assumed OPERABLE in both the large SAFETY ANALYSES and small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is also considered to determine if it yields limiting results. The loss of offsite power assumption imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break. The limiting large break LOCA is a double ended guillotine break in the cold leg. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure. (continued) Watts Bar-Unit 1 B 3.5-2 Revision 39 Amendment 21
Accumulators B 3.5.1 BASES APPLICABLE As a conservative estimate, no credit is taken for ECCS SAFETY ANALYSES pump flow until an effective delay has elapsed. This (continued) delay accounts for the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA. The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase. This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46, Paragraph b (Ref. 3) will be met following a LOCA:
- a. Maximum fuel element cladding temperature is ~
$ 2200°F;
- b. Maximum cladding oxidation is ~$ 0.17 times the total cladding thickness before oxidation;
- c. Maximum hydrogen generation from a zirconium water reaction is ~$ 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and
- d. Core is maintained in a coolable geometry.
since Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46. For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained (continued) Watts Bar-Unit 1 B 3.5-3 Revision 39 Amendment 21
Accumulators B 3.5.1 BASES APPLICABLE water volume is the same as the deliverable volume SAFETY ANALYSES for the accumulators, since the accumulators are emptied, (continued) once discharged. The safety analysis assumes values of 7518 gallons and 8191 gallons. To allow for instrument inaccuracy, values of 7630 gallons and 8000 gallons are specified. The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron toncentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. The small break LOCA analysis is performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure analysis limit of 690 psig prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The LOCA analyses support a range of 585 to 690 psig. To account for the accumulator tank design pressure rating, and to allow for instrument accuracy values of ~2 610 psig and ~ 660 psig are specified for the pressure indicator in the main control room. The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. (Re f s. 2 and 4) . (continued) Watts Bar-Unit 1 B B 3.5-4 Revision 3, 39 Amendment 21
Accumulators B 3.5.1 BASES APPLICABLE The accumulators satisfy Criterion 3 of the NRC Policy SAFETY ANALYSES Statement. (continued) LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 3) could be violated. For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure
> 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.
This LCO is only applicable at pressures> 1000 psig. At pressures ~ 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 3) limit of 2200 oo F. In MODE 3, with RCS pressure ~ 1000 psig, and in MODES 4, 5, are and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators. (continued) Watts Bar-Unit 1 B 3.5-5
Accumulators B 3.5.1 BASES (continued) ACTIONS A.l A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours. In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially reflooding phase of a large recovered core during the early ref100ding break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not stearn line break for the discharge following a large main steam majority of plants. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours is allowed to return the boron concentration to within limits. B.l B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 1 hour. In this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 1 hour Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. C.l and C.2 C.1 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and pressurizer pressure reduced to (continued) Watts Bar-Unit 1 B 3.5-6
Accumulators B 3.5.1 BASES ACTIONS C.l and C.2 (continued)
~ 1000 psig within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.l If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately. SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully open every 12 hours. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. This Frequency is considered reasonable in view of other administrative controls that ensure a mispositioned isolation valve is unlikely. SR 3.5.1.2 and SR 3.5.1.3 Every 12 hours, borated water volume and nitrogen cover pressure are verified for each accumulator (refer to the note below). This Frequency is sufficient to ensure adequate injection during a LOCA. Because of the static design of the accumulator, a 12 hour Frequency usually allows the operator to identify changes before limits are reached. Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends. Note: In the discussion contained in the Applicable Safety Analyses of this Bases section, the borated water vollli~e volume and nitrogen cover pressure specified for SR 3.5.1.2 and SR 3.5.1.3 account for instrument accuracy (Ref. 6). (continued) Watts Bar-Unit 1 B B 3.5-7 Revision 29
Accumulators B 3.5.1 BASES (continued) SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued) The boron concentration should be verified to be within required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage. Sampling the affected accumulator within 6 hours after a 75 gallons (1% volume) increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. This is consistent with the recommendation of NUREG-1366 (Ref. 5). SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator when the pressurizer pressure is ~1000 psig enSures ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed. This SR allows power to be supplied to the motor operated isolation valves when pressurizer pressure is < 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves. Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA. This design feature still exists, but is no longer required for accident mitigation. (continued) Watts Bar-Unit 1 B 3.5-8
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 97. G 2.3.6097 2.3.6 097 Given the following:
Unit 1 is at 100% power.
- The Monitor Tank is to be released.
Which ONE of the following identifies ... 0-RM-90-122 "Liquid Radwaste Effluent Line" being inoperable would affect (1) how O-RM-90-122 the planned release, and (2) the permit requirements to restart the release if it isolated due to low Cooling Tower
.Blowdown Slowdown (CTBD) flow?
0-RM-90-122 inoperable Low CTBD flow A. Must have 2 independent Release can not be restarted until a samples, analysis, and valve new release permit generated after lineups performed. it is isolated due to low CTBD flow. B~ Must have 2 independent B:o" Release can be continued if samples, analysis, and valve restarted within 2 hours with lineups performed. existing release permit. C. Must have 2 independent analysis, Release can not be restarted until a and valve lineup performed but new release permit generated after obtaining only 1 sample is required. it is isolated due to low CTBD flow. D. Must have 2 independent analysis, Release can be continued if and valve lineup performed but restarted within 2 hours with obtaining only 1 sample is required. existing release permit. Page 65
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, 2 independent samples are taken, independently analyzed, and the valve lineup independently verified are all required but does not need a new permit if it is restarted within 2 hours. Plausible because the action required due to the radiation monitor being inoperable are correct and there are conditions where the release being isolated would require a new release permit. B. Correct, If the radiation monitor is out of service, a release can be made provided that 2 independent samples are taken, independently analyzed, and the valve lineup independently verified and if while in progress the release is isolated due to low Cooling Tower Blowdown flow the release can be restarted using the same release permit provided the restart is within than 2 hours of the isolation. C. Incorrect, there must be 2 independent samples taken and well as 2 independent analysis of the samples and independent valve lineup verification and a new release permit is not required if the release is restarted within 2 hours. Plausible because the action required due to the radiation monitor being are correct except for the taking of the sample and normally only one sample is required and there are conditions where the release being isolated would require a new release permit. D. Incorrect, there must be 2 independent samples taken and well as 2 independent analysis of the samples and independent valve lineup verification and a new permit is not required if the release is restarted within 2 hours but the release can be restarted using the same release permit provided the restart is within than 2 hours of the isolation. Plausible because the action required due to the radiation monitor being are correct except for the taking of the sample and normally only one sample is required and the release being restarted without requiring a new release permit is correct. Page 66
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 97 Tier: _3__ Group n/a KIA: G 2.3.6 Ability to approve release permits. Importance Rating: 2.0 13.8 2.0/3.8 10 CFR Part 55: 41.13 /I 43.4 /I 45.10 10CFR55.43.b: 2, 4, 5 2,4,5 KIA Match: Applicant is required to understand the requirements that must be met before approving a release package and the requirements for restarting a release after it had been terminated prior to completion. SRO because the question involves the application of ODCM required actions when equipment status is abnormal. Technical
Reference:
SOI-77.01, Liquid Waste Disposal, Rev 0062 ODCM ,Offsite Dose Calculation Manual, Rev 22 0-001-90-1, Liquid Radwaste Tank Release, Rev 0031 0-ODI-90-1, Proposed references None to be provided: Learning Objective: 3-0T-SYS077 A
- 19. Discuss how processed water is released.
- 21. State the parameters of the Liquid Radwaste Processing System that are governed by the Offsite Dose Calculation Manual (ODCM).
Question Source: New Modified Bank xX Bank Question History: WBN question SYS077A.19 SYS077A.1 9 051 modified Comments: Question stem and distractors modified, correct answer relocated Page 67
QUESTIONS REPORT for ILT Sank Current wo pw SYS077A.l9051 SYS077A.19 051 Given the following plant conditions:
- Unit is operating at 100% power.
- A release of the Monitor Tank to Cooling Tower Slowdown is planned.
RM-90-122, Liquid Radwaste Effluent Monitor is inoperable. Which of the following identifies the requirements for making the release under these conditions?
- a. No release can be made untiI1-RM-90-122 is returned to service.
- b. Sample results must show that activity of the release liquid is <2x1 0-44 microcuries per milliliter.
- c. " Two separate samples must be analyzed, and two independent qualified C.oI members of facility staff verify release rate and discharge valve lineup.
- d. The tank must be recirculated for two volumes, and a licensed Senior Reactor Operator must confirm release rate and discharge valve lineup.
The correct answer is C
- a. incorrect - per ODCM a release to the environment may be made provided that the conditions in c. below are met, but examinee may confuse what these conditions consist of and if all conditions are not met then no release can be made.
- b. incorrect - per ODCM a release to the environment may be made provided that the conditions in c. below are met, but examinee may apply the wrong ODCM section and select this since 2E-4 is the limit per ODCM 1.2.1.1.
- c. correct - per aDCMODCM 1.1.1 action A with 1-RM-90-22 inoperable the release can be made if prior to the release, (1)- two independent samples are made, (2)-two qualified members of the staff verify release rate and (3)-discharge valve lineup.
- d. incorrect - per aDCMODCM a release to the environment may be made provided that the conditions in c. below are met, but examinee may select this since the tank must be recirculated for normal releases.
REFERENCES:
ODCM 1.1.1 Action A NEW 10CFR55. 43.4 /I 45.10 RO- NA SRO - 95 11 :12:36 Tuesday, July 14, 2009 11: 12:36 AM 9
UE Watts Bar Nuclear Plant Unit 0 Offsite Dose Instruction 0-001-90-1 Liquid Radwaste Tank Release Revision 0031 Quality Related Level of Use: Continuous Use Effective Date: 02-06-2009 Responsible Organization: CEM, Chemistry Prepared By: Eddie Woods Approved By: Todd E. Rose
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 2 of 32 Revision Log Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change Rev 16 3/20/99 All Intent change. Procedure simplification to move all MIG and Operations performance steps to Appendices. Remove references to PAI-4.01. Relocated requirements for inoperable liquid radwaste radiation monitor 0-RE-90-122 from 0-001-90-4. Added OOCM SR for inoperable rad monitor actions to Section 1.2.2. Rev 17 8/17/99 2,7,8,10,12 Non-Intent Change. Removed expiration of
,15,17,18 release permits. Deleted Appendix o. O. Added requirement that an ongoing LRW release and 001 package must be completed prior to starting a new new 001 package. Moved background limits from Appendix A to Appendix C step [7] [b]. (PER99-007870-000)
Clarified wording in Appendix Band C to match the title on the release permit. Rev 18 08/20/99 2,7 Non-Intent change. Corrected typo. Rev 19 03/09/00 2,4,6,14, Non-Intent change. Rearranged Appendix C as 15,16,17, part as the corrective action to 18 WBPER99-01648-000. Removed reference to CM-9.99. Corrected referenced step number in Section 6.1 step [5]. Reassigned appendix letters. Rev 20 2/9/01 2,4,6,7,9 Intent Change. Added steps to allow the use of the new computer point that averages the background for 0-RE-90-122. Removed references to TI-18 Appendix O. Rearranged steps so that setpoint and permit data need be recorded if a setpoint change is required. Rev 21 10/29/01 2,4,8,9,10, Intent Change. Added permit generation 19-23 instructions in Appendix F and G. Removed IV requirements and source note 3 since pre-release is generated using default values. Removed reference to CM-9.91.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 3 of 32
.Revision Log Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change Rev 22 12/12/02 2,6,7,8,9, Intent Change. Added IV requirements if 12,16,17, monitor is inoperable. Added use of CHEM 6 23,24 screen to update monitor setpoints. Included the ability to use ICS point RE-90-122 for operations. Minor editorial changes. Removed step to record data on chart recorder paper Rev 23 12/01/03 2,15 Intent Change. Corrected Flow Indicator UNID referenced in Appendix C.
Rev 24 10/15/04 1,2 Intent Change. Change the level of use to continuous. Rev 25 06/13/06 All This procedure has been converted from Word 95 to Word 2002 XP using rev 24, by Austin Norris. Removed background limits and added place keeping box to step Rev 26 08/08/2007 1,3, 12 1,3,12 Add independent verification to Section 7.0 to ensure CHEM 6 screen was updated following a setpoint change. Rev 27 11/21/2007 1,9,10,14 Revised the administrative limit for total gamma activity from 3E-05 to 8.65E-07 JlCi/ml.j.1Ci/ml. Corrective action for PER 130473. Rev 28 2/22/2008 1,9, 10, 1,9,10, Added sampling signoffs to Appendix A. 14,22,25 Corrected numbering on Appendixes. Added instructions to use CHEM8 to obtain LRW concentration limit. Deleted Appendix E comparison guidelines. Added instructions to ensure the higher concentration of all nuclides determined during duplicate sampling are used to open the prerelease permit. Rev 29 08/18/2008 1,3, 15 1,3,15 Revised Appendix B to include SRO approval as step [1] versus standalone signoff.sign off. Rev 30 11/24/2008 3,7 Added a reference to SPP-6.4
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WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 4 of 32 Revision Log '-- Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change Rev 31 02/06/2009 1,3,26 Clarified editing of the concentration screen when the radiation monitor is inoperable.
WBN Liquid Radwaste Tank Release O-ODI-90-1 Unit 0 Rev. 0031 Page 5 of of3232 Table of Contents
1.0 INTRODUCTION
.......................................................................................................... 7 1.1 Purpose ........................................................................................................................ 7 1.2 Scope ............................................................................................................................ 7 1.2.1 Analyses To Be Performed Performed ............................................................................. 7 1.2.2 Surveillance Requirements Fulfilled and Modes ............................................. 7 1.3 Frequency/Conditions ................................................................................................... 7
2.0 REFERENCES
...............................................................................................................
............................................................................................................. 8 2.1 Performance References .............................................................................................. 8 2.2 Developmental References ........................................................................................... 8 3.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 8 4.0 PREREQUISITE ACTIONS ..............................................*...........................................
.......................................................................................... 9 4.1 Preliminary Actions ....................................................................................................... 9 4.2 Approvals and Notifications .......................................................................................... 9
) 5.0 ACCEPTANCE CRITERIA ........................................................................................... 9 6.0 PERFORMANCE ........................................................................................................ 10 6.1 Pre-Release Instructions ............................................................................................. 10 6.2 Post-Release Instructions .............................................................
...........................................................................................
,............................. 12 7.0 ................... ~ .................................................... 13 POST PERFORMANCE ACTIVITIES ........................................................................
8.0 RECORDS .................................................................................................................. 14 8.1 QA Records ................................................................................................................ 14 8.2 Non-QA Records ........................................................................................................ 14 Appendix A: Release Data .................................................... ~ ........................................ 15
.............................................................................................
Appendix B: Pre-Release Setpoint Adjustment ........................................................... 16 Appendix C: Release Instructions ..................................................................................
................................................................................. 19 Appendix D: Inoperable LRW (0-RE-90-122) ................................................................. 23 Appendix E: Instructions For Opening A LRW Permit ................................................ 24 Appendix F: Instructions For Closing A LRW Permit Permit...................................................
................................................. 29
WBN Liquid Radwaste Tank Release 0-001-90-1 UnitO Unit 0 Rev. 0031 "
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Page 6 of 32 Table of Contents (continued)
............................................................................................. 32 Source Notes ....................*........................................................................ ) )
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 7 of 32
1.0 INTRODUCTION
1.1 Purpose This Offsite Dose Instruction (001) provides the steps required to perform sampling, analyses, calculations, monitor source checks, and setpoint changes required by the Offsite Dose Calculation Manual (ODCM) to perform radioactive effluent releases from a Liquid Radwaste Tank via the Cooling Tower Blowdown. 1.2 Scope 1.2.1 Analyses To Be Performed Principal Gamma Emitters 1-131 Dissolved/Entrained Noble Gases 1.2.2 Surveillance Requirements Fulfilled and Modes This instruction fulfills the following ODCM Surveillance Requirements (SRs): Surveillance Applicable Performance Requirements Modes Modes ODCM p.2.2.1.1.1 Table 2.2-1 Item A All All ODCM p2.2.1.3.2 All All ODCM p2.2.1.1.2 Table 2.2-1 Item A All All ODCM p2.1.1 Table 2.1-1 Item 1.a All All ODCM p2.2.1.2 All All ODCM pControl 1.1.1 Table 1.1-1 Item 1.a Action A All All 1.3 Frequency/Conditions This instruction is initiated prior to each batch release from the following liquid radwaste tanks: A. Cask Decontamination Collector Tank B. Monitor Tank
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 of32 Page 8 of 32
2.0 REFERENCES
2.1 Performance References A. CM-7.18, Preparation of Liquid Marinelli and Composite Samples. B. CM-9.09, Effluent Radiation Monitor Alarm Response Guidelines. C. CM-9.30, Operation of the Gamma Spectroscopy Counting System. o. D. CM-9.73, Liquid RadwasteTanks and Liquid Effluent Radiation Monitor Sampling Methods. E. CM-9.90, Administration of the Effluent Management Program. F. SOI-77.01, Liquid Waste Disposal. G. 1-SI-0-2-00, Shift and Daily Surveillance Log Master. H. TI-18, Calculation Methods for Effluent Radiation Monitors. I. SPP- 6.4, Measuring and Test Equipment. 2.2 Developmental References A. Offsite Dose Calculation Manual. B. WBNTSR-066, "Design Flowrate for the Offline Liquid Radiation Monitors." 3.0 PRECAUTIONS AND LIMITATIONS A. Failure to utilize the ALARA principles when working with radioactive samples can result in unnecessary exposure or personnel contamination. B. An ongoing release from any liquid radwaste tank must be completed and the 001 package completed prior to initiating a new 001 package. C. The detection limits for the analyses are listed in the OOCM. ODCM. O. D. This 001 is limited to radwaste tanks which have a known recirculation time determined by testing.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 ') Page 9 of32 of 32 4.0 PREREQUISITE ACTIONS NOTE Subsections of Section 4.0 may be performed independently and in any sequence necessary to accomplish the desired task. 4.1 Preliminary Actions [1] RECORD start date and time on Surveillance Task Sheet. 4.2 Approvals and Notifications [1] NOTIFY MIG that support may be required during performance of this instruction. 5.0 ACCEPTANCE CRITERIA
') ;
NOTE The acceptance criteria A, B, and C are also listed on the release permit. A. The total post dilution concentration of dissolved and entrained noble gases does NOT exceed 2.0 X 10-4 J..lCilml. J...lCi/ml. B. The total effluent concentration limit (Eel)(ECl) ratio for nongaseous non gaseous radionuclides does NOT exceed 10. . C. The dose or dose commitment to a member of the public from radioactive material in liquid effluents discharged from each unit to unrestricted areas is to be limited as follows:
- 1. During any calendar quarter: To less than or equal to 1.5 mrem to the total body and to less than or equal to 5.0 mrem to any organ.
- 2. During any calendar year: To less than or equal to 3.0 mrem to the total body and to less than or equal to 10.0 mrem to any organ.
D. An upscale deflection should be observed when performing a source check on 0-RE-90-122. O-RE-90-122.
WBN Liquid Radwaste Tank Release O-ODI-90-1 0-001-90-1 Unit 0 Rev. 0031 Page 10 of 32 6.0 PERFORMANCE 6.1 Pre-Release Instructions [1] OBTAIN the following information from Operations: [1.1] INDICATE liquid radwaste tank to be sampled on Appendix A. [1.2] VERIFY minimum recirculation time has been met on Appendix A. [1.3] RECORD initial tank level on Appendix A. RECORD'initial [1.4] RECORD estimated start date and time of release on Appendix A. [1.5] INDICATE operability status of O-RE-90-122 on Appendix A. [2] SAMPLE the appropriate liquid radwaste tank per CM-9.73 AND ) RECORD initials on Appendix A. [3] PREPARE and PRESERVE samples per CM-7.18. [4] PERFORM a gamma isotopic analysis per CM-9.30, AND REVIEW, SIGN, and ATTACH the report to Data Package. [5] IF O-RE-90-122 is operable, THEN perform the following: [5.1] OBTAIN the averaged background for O-RE-90-122 from ICS by entering "CHEM4" on the yellow bar AND RECORD on Appendix A. [5.2] OBTAIN the current setpoint for O-RE-90-122 from ICS by entering "CHEM6" on the yellow bar AND 1 RECORD on Appendix A. 1 [6] IF O-RE-90-122 is inoperable, THEN PERFORM Appendix D.
WBN Liquid Radwaste Tank Release O-ODI-90-1 Unit 0 Rev. 0031 Page 11 of 32 6.1 Pre-Release Instructions (continued) [7] OBTAIN the lRW concentration limit from ICS CHEM8 screen. [8] IF the total gamma activity listed on the gamma isotopic analysis report exceeds the value on CHEM8, THEN SUBTRACT the noble gas activity from the total gamma activity. [9] IF the corrected gamma activity still exceeds the value listed on CHEM8, THEN OBTAIN the Chemistry Duty Manager's approval on
. Appendix A to permit the release.
[10] GENERATE a release permit per Appendix E, AND [10.1] VERIFY all input data on Appendix A matches permit [10.2] VERIFY permit nuclide data matches the gamma analysis of the tank [10.3] SIGN and ATTACH release permit to Data Package [11] IF Appendix D was performed, THEN OBTAIN independent verification of the release permit data, AND ENSURE verification is indicated by signature of the verifier on Appendix A. [12] LABEL the composite sample with the permit number AND STORE in the deSignated designated storage location. [13] IF the post-dilution ECl fraction listed on the pre-release permit exceeds 2, THEN [13.1] SUBTRACT the post-dilution ECl fractions for any non-gamma emitters and noble gas radionuclides from the post-dilution ECl given on the permit. [13.2] IF the post-dilution ECl fraction listed still exceeds 2, THEN
WBN Liquid Radwaste Tank Release 0-001-90-1 O-ODI-90-1 Unit 0 UnifO Rev. 0031 Page 12 of 32 6.1 Pre-Release Instructions (continued) INDICATE on the release permit that the sample flowrate through radiation monitor 0-RE-90-122 must be maintained above the minimum flow rate sample flowrate which corresponds to the corrected post-dilution ECl fraction from the table below. s5 Post-dilution ECl Minimum Allowed Monitor Sample Flowrate
- 2
~2 2.9 gpm
~3 2<ECl ::;3 3.6 gpm
~5 3<ECl ::;5 4.1 gpm
~10 5<ECl ::;10 - -
6.4 gel'Tl gpm [14] IF Acceptance Criteria are NOT met, THEN NOTIFY the Unit SRO and Chemistry Management or designee. [15] IF 0-RE-90-122 is inoperable OR the setpoint in Section 1.0 of Appendix A and on the release permit are within 0.101 Vdc, THEN ) TRANSMIT the Data Package to Operations, AND REQUEST performance of Appendix C. [16] OBTAIN Unit SRO approval on Appendix B to adjust 0-RE-90-122 setpoint. [17] TRANSMIT Data Package to MIG to adjust setpoint to the value indicated on the release permit in accordance with Appendix B. 6.2 Post-Release Instructions [1] IF 0-RE-90-122 is indicated as inoperable in Section 1.0 of Appendix A AND operable in Appendix C, THEN DELETE the release permit per CM-9.90, and PROCEED TO Section 7.0. 22 [2] CLOSE release permit per Appendix F, AND ) REVIEW, SIGN, and ATTACH Post Release printout to Data Package.
WBN Liquid Radwaste Tank Tank Release 0-001-90;.1 0-001-90-1 Unit 0 Rev. 0031 Page 13 of 32 6.2 Post-Release Instructions (continued) [3] IF Acceptance Criteria are NOT met, THEN NOTIFY the Unit SRO and Chemistry Management or designee. 7.0 POST PERFORMANCE ACTIVITIES [1 ] IF the Setpoint was changed (Appendix B performed), THEN [1.1 ] TYPE "CHEM6" on the yellow bar of an ICS screen and PRESS enter. [1.2] TYPE "CSL" on the yellow bar of the ICS screen and PRESS enter. [1.3] SELECT "Change Security Level" and SELECT OK. [1.4] TYPE "CHEMSET" for both the user name and password and SELECT OK. [1.5] SELECT "CHEMISTRY ICS MAIN MENU". [1.6] SELECT "LIQUID RAD MON SETPOINTS (CHEM 6)". [1.7] SELECT "PRESS BUTTON TO CHANGE SETPOINT" for 0-RE-90-122. [1.8] ENTER the "as left setpoint (Vdc)" from Appendix "A" Section 2.0. [1.9] SELECT "F3 = =save data". [1.10] SELECT "F1 = =main menu". [1.11 ] VERIFY the new setpoint matches the "as left" setpoint from Appendix "A". [1 .12] [1.12] REQUEST an independent verifier to ensure that the setpoint value recorded on CHEM 6 screen for 0-RE 122 matches the as left setpoint recorded on Appendix A Section 2. IV Date [2] RECORD completion date and time on Surveillance Task Sheet.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 14 of 32 8.0 RECORDS 8.1 QA Records The Data Package is a QA record, is handled in accordance with the Document Control and Records Management Program, and contains the following: A. Surveillance Task Sheet. B. Completed Appendices A through F. C. Other sheets added during the performance. 8.2 Non-QA Records None ) )
WBN Liquid Radwaste Tank Release O-ODI-90-1 0-001-90-1 Unit 0 Rev. 0031 Page 15 of 32 Appendix A (Page 1 of 1) Release Data 1.0 SECTION 1 - PRE-RELEASE PERMIT DATA o Cask Decontamination Tank Initials o Monitor Tank Minimum recirculation time met Sampled by Duplicate sampled by (if monitor inoperable) Initial Tank Level_ _ _ _ _ _ % Estimated Start Date/Time _ _ _ _ _ _ _,I/_ _ __ ____ __ _ __ O-RE-90-122 o Operable D [J Inoperable D Background_ _ _ _ _cpm Setpoint._ _ _ _ _Vdc Setpoint Approval to permit a release with the gamma concentration greater than the Chemistry Duty Manager Date CHEM8 LRW concentration limit: Independent Verification of Release Permit Data: Signature Date 2.0 SECTION 2 - SETPOINT ADJUSTMENT DATA As-found Setpoint_ _ _ _ _ _ _Vdc As-Left Setpoint Setpoint._ _ _ _ _ _ _Vdc 2nd Party 3.0 SECTION 3 - RELEASE DATA Release Start Date/Time,...--_,...--_ DatelTime _ _ _-
----------
Release End DatelTime_ __ ____ ___ ______ __ Final Tank Level_ _ _ _ _ _% Average CTBD Dilution Flowrate (O-FI-27-98)_ _ _ _ _ _ _gpm g,pm )
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 16 of 32 Appendix B (Page 1 of 3) Pre-Release Setpoint Adjustment 1.0 PRE-RELEASE SETPOINT ADJUSTMENT [1] Unit SRO approval to adjust O-RE-90-122 setpoint Unit SRO Date SRO Signature Date NOTE During the performance of this section, the monitor may be declared inoperable due to a performance step NOT being completed. [2] ENSURE the following M&TE is available, and [2.1] MEETS the minimum range, accuracy, and calibration due date. [2.2] COMPLETE the following table: MINIMUM ACTUAL REQUIRED CAL DUE DESCRIPTION RANGE RANGE ACCURACY TVA ID 10 NO. DATE DMM o to 10 Vdc ----
+/- 0.01 Vdc
-- --_._- '--------
[3] NOTIFY UO that the following devices and annunciators are affected by the performance of this ODI: 001: A. Radiation Module [0-M-12]: 0-RM-90-122A, WDS LIQUID RELEASE LINE B. Recorder [0-M-12]: 0-RR-90-122, WDS LIQUID RELEASE LINE C. Annunciator [0-M-12]: Window No. Description 0-XA-55-12B-181A WDS RELEASE LINE 0-RM-122 LlQ RAD HI 0 0-XA-55-12B-181C WDS RELEASE LINE 0-RM-122 INSTR MALF 0 [4] SET the DMM to measure 10 Vdc. [5] CONNECT DMM to test pOints points TP-3 (+) and TP-1 (-) on the radiation module (RM) for 0-RM-90-122A.
WBN Liquid Radwaste Tank Release O-ODI-90-1 Unit 0 Rev. 0031 Page 17 of 32 Appendix B (Page 2 of 3) 1.0 PRE-RELEASE SETPOINT ADJUSTMENT (continued) NOTE Function Switch is spring-return and may be held in position by hand or retained with a clip. [6] PLACE and RETAIN RM Function Switch in ALARM ADJ. [7] ENSURE Red HIGH alarm is NOT LIT by adjusting the ALARM REF potentiometer as necessary. [8] ADJUST ALARM REF potentiometer until the Red HIGH alarm light is just LIT, AND RECORD the as-found setpoint on Appendix A. [9] IF as-found setpoint is NOT +/- 0.101 Vdc of the previous calculated voltage (Section 1.0 of Appendix A), THEN NOTIFY Unit SRO to determine corrective actions. [10] ADJUST ALARM REF potentiometer until DMM indicates new setpoint voltage listed on the Batch Liquid Effluent Permit. [11] IF the Red HIGH alarm is actuated, THEN ADJUST HIGH SET potentiometer until the Red HIGH alarm can be cleared, AND CLEAR Red HIGH alarm light. [12] ADJUST HIGH SET potentiometer until Red HIGH alarm lamp is just LIT. [13] ADJUST ALARM REF potentiometer below Red HIGH alarm setpoint, AND CLEAR Red HIGH alarm light. [14] ADJUST ALARM REF .REF potentiometer until Red HIGH alarm lamp is just LIT. [15] VERIFY the as-left setpoint is +/- 0.1 Vdc of calculated voltage listed on the Batch Liquid Effluent Permit.
WBN Liquid Radwaste Tank Release 0-001-90-1 O-ODI-90-1 Unit 0 Rev. 0031 Page 18 of 32 Appendix B (Page 3 of 3) 1.0 PRE-RELEASE SETPOINT ADJUSTMENT (continued) [16] IF as-left setpoint is NOT within the above tolerance, THEN 1.0[11 ] through 1.0[16], OR REPEAT steps .1.0[11] NOTIFY Unit SRO to determine corrective actions. [17] RECORD as-left setpoint trip voltage on Appendix A. [18] ADJUST ALARM REF potentiometer until the DMM indicates a voltage of approximately 0 Vdc. [19] RETURN Function Switch to OPERATE. 2nd Party [20] PRESS alarm lights to RESET, AND \ I VERIFY the following: J A. Red HIGH alarm light - NOT LIT. B. Yellow ALERT alarm light - NOT LIT. C. Green OPERATE light - LIT. [21] VERIFY alarms 181A and 181C are NOT LIT. [22] REMOVE DMM from TP-3 (+) and TP-1 (-) on the RM. [23] PERFORM a Source Check, AND [23.1] VERIFY an observable upscale deflection. [23.2] IF an upscale deflection is NOT observed, THEN NOTIFY Unit SRO to evaluate monitor operability. [24] RETURN Data Package to Unit SRO for performance of Appendix C. )
WBN Liquid Radwaste Tank Release O-ODI-90-1 0-001-90-1 Unit 0 Rev. 0031 Page 19 of 32 Appendix C (Page 1 of 4) Release Instructions 1.0 RELEASE INSTRUCTIONS [1] OBTAIN Unit SRO approval signature to perform release on the Batch Liquid Effluent Permit. [2] INDICATE O-FIT-77-5042 operability status. o Operable 0 Inoperable [3] IF 0-FIT-77-5042 is inoperable, THEN ESTIMATE flow during release in accordance with 1-SI-0-2-00. [4] INDICATE 0-RE-90-122 operability status. o Operable 0 Inoperable [5] IF 0-RE-90-122 operability status is different than that indicated in Section 1.0 of Appendix A THEN RETURN Data package to Chemistry, AND REQUEST new 0-001-90-1 package for this release. 2 [6] IF 0-RE-90-122 is inoperable, THEN [6.1] ENSURE the discharge valve lineup is independently verified. [6.2] INITIATE and PERFORM release per SOI-77.01 in conjunction with this 001. [6.3] GOTO step 1.0[14] of this Appendix.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 20 of 32 Appendix C (Page 2 ofof 4) Release Instructions 1.0 RELEASE INSTRUCTIONS (continued) NOTE Monitor alarm or isolation may occur if the source check switch is NOT released as soon as an upscale deflection is observed. [7] IF O-RE-90-122 is operable, THEN PERFORM a Source Check, AND [7.1] VERIFY an observable upscale deflection. [7.2] IF an upscale deflection is NOT observed, THEN NOTIFY the Unit SRO to evaluate monitor operability. [8] ENSURE the following:
- Red HIGH alarm light - NOT LIT.
- Yellow ALERT alarm light - NOT LIT.
- Green OPERATE light - LIT.
[9] ENSURE the following:
- O-RM-90-122A, WDS LIQUID RELEASE LINE, NOT IN ALARM [O-M-12].
- Annunciator 181-A, WDS RELEASE LINE O-RM-122 LlQ RAD HI, NOT LIT.
- Annunciator 181-C, WDS RELEASE LINE O-RM-122 INSTR MALF, NOT LIT.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 PagE! 21 of 32 Page Appendix C (Page 3 of 4) Release Instructions 1.0 RELEASE INSTRUCTIONS (continued) NOTE The following step should be performed just prior to the start of the release. [10] RECORD the current reading from 0-RM-90-122A or 122B or ICS point RE-90-122. 0-RM-90-122 reading prior to release: _ _ _ _ _cpm [11] INITIATE and PERFORM release per Sol-77.01in SOI-77.01in conjunction with this 001. [12] IF the release is terminated due to a high radiation alarm, THEN NOTIFY Chemistry to perform CM-9.09, AND RECORD the highest observed reading from 0-RM-90-122A or 0-RM-90-122B or ICS point RE-90-122. Reading_ _ _ _ _ _cpm High Rad Alarm Reading, NOTE The first reading from 0-RE-90-122 taken during the release should be obtained between ten and 15 minutes after initiation of the release. [13] RECORD 0-RM-90-122A or 0-RM-90-122B or ICS point RE-90-122 reading once per hour during the release. Date/Time Date/ Time Monitor Reading during release (cpm) Initials [14] RECORD all start and stop dates/times.
WBN Liquid Radwaste Tank Release 0-001-90-1 O-ODI-90-1 Unit 0 Rev. 0031 Page 22 of 32 Appendix C (Page 4 of 4) 4of4) Release .Instructions 1.0 RELEASE INSTRUCTIONS (continued) Start DatelTime Stop DatelTime Date/Time Duration (minutes) Initials [15] release start date and time of the release on RECORD the relea.se Appendix A. [16] IF the tank was released in one segment, THEN ENTER the release end date and time on Appendix A. NOTE The calculated release end date and time is used to ensure compliance with the ODCM. This time may not match the actual release end date and time. [17] IF the tank was released in several segments, THEN [17.1] DETERMINE the total duration of release. [17.2] ADD the duration to the initial start date and time. [17.3] ENTER the calculated release end date and time as the release end date and time on Appendix A. A [18] RECORD the final tank level upon termination of release and average cooling tower blowdown dilution flow rate during the release on Appendix A A. [19] TRANSMIT Data Package to Chemistry.
WBN Liquid Radwaste Tank Release O-ODI-90-1 Unit 0 Rev. 0031
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Page 23 of 32 Appendix D (Page 1 of 1) Inoperable LRW (0-RE-90-122) 1.0 INOPERABLE LRW (0-RE-90-122) NOTE The sampling should be performed by a separate analyst working independently of the analyst performing Section 6.1 Step6.1 [2]. [1 [1]] OBTAIN a sample from the appropriate liquid radwaste tank per CM-S.73 CM-9.73 AND RECORD initials on Appendix A. [2] RECORD the sample date/time: Date Time Initials [3] PREPARE and PRESERVE the sample. sample per CM-7.18.
\
) I [4] PERFORM a gamma isotopic analysis per CM-S.30, CM-9.30, AND REVIEW, SIGN, and ATTACH the report to the Data Package.
,**[5]
[5] COMPARE the total ECl value from the two gamma isotopic reports that have been generated. [5.1] SELECT the report With with the higher total ECl value to open the release permit. [5.2] SELECT the sample with the higher total ECl value to use for composite sample. [6] PROCEED TO Section 6.1 step 6.1 [7].
WBN Liquid Radwaste Tank Release O-ODI-90-1 0-001-90-1 Unit 0 Rev. 0031 Page 24 of 32 Appendix E (Page 1 015) of 5) Instructions For Opening A LRW Permit 1.0 INSTRUCTIONS FOR OPENING A LRW PERMIT [1] IF NOT previously performed, THEN ACCESS the CAS Main Menu by entering the user 10 and password at the EMS system node. [2] SELECT "Effluent Management" from the Main Menu. [3] SELECT "Process Liquid Permit" from the Effluent Management Menu. NOTE The release point pOint number for the CDCT COCT is 10 and Monitor Tank is 11. [4] ENTER Release Point Number, AND PRESS RETURN. [5] ENTER the sample number from the gamma isotopic report for the liquid sample, AND PRESS RETURN. [6] SELECT "Define and Open A New Liquid Permit" by placing the cursor next to the menu item, AND
~O key.
PRESS the "DO" [7] ENTER "Y" and PRESS RETURN to define and open a permit at the screen displaying the last permit information. [8] INPUT the estimated date and time the release is to start, AND PRESS RETURN. [9] PRESS RETURN at the estimated "Release End" date and time prompt. '\
WBN Liquid Radwaste Tank Release O-ODI-90-1 Unit 0 Rev. 0031 Page 25 of 32 Appendix E (Page 2 of 5) 1.0 INSTRUCTIONS FOR OPENING A LRW PERMIT (continued) [10] PRESS the "TAB" key at the "Release Flow Rate" field to use the default flow rate, OR IF a value other than the default has been authorized by the Chemistry Management or designee, THEN INPUT the estimated flow rate as directed, AND PRESS RETURN. [11] PRESS RETURN at the "Start %" prompt. [12] PRESS TAB at the "Release Volume" prompt to use the default volume. [13] PRESS TAB at the Dilution Flow Rate prompt, OR IF a value other than the default has been authorized, THEN INPUT the estimated dilution flow rate as directed, AND PRESS RETURN. [14] PRESS RETURN at the Dilution Volume prompt. [15] PRESS TAB until the cursor is at the "Collected By" data field, AND ENTER the initials of the analyst that collected the sample, AND PRESS RETURN. [16] PRESS the "FILL" command key (F14), AND VERIFY all input data is correct. [17] IF any changes are required, THEN USE Control P, TAB, or RETURN keys to position the cursor to the appropriate data field, AND nec~ssary: ENTER changes as necessary.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 26 of 32 Appendix E (Page 3 of 5) 1.0 INSTRUCTIONS FOR OPENING A LRW PERMIT (continued) [18] PRESS the "Save" command key (F10). (F1 0). [19] PRESS the "Process" command key (DO). NOTE When the monitor is inoperable the permit is to be opened with the gamma isotopic report with the higher Eel concentration. [20] PRESS the VMS-GSP command key (F12) at the sample entry concentrations screen. NOTE The gamma isotopic report imported may require editing per steps 22 through 24. [21] IF the monitor is inoperable, THEN EDIT the sample entry concentration screen to include all identified nuclides and the higher concentration from the two gamma isotopic reports, as necessary. [22] IF a nuclide and concentration is to be added, THEN [22.1] PLACE the cursor below the last nuclide, and ADD the nuclide to the list. [22.2] PRESS TAB, and ENTER the concentration, and PRESS RETURN [22.3] .PRESS "Save" command key (F1 0) [22.4] ENTER "Y" at the prompt "Has this been authorized? (YIN)" [22.5] ENTER "EMS" when prompted for the password.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 27 of32 of 32 Appendix E (Page 4 of 5) 1.0 INSTRUCTIONS FOR OPENING A LRW PERMIT (continued) [23] IF natural occurring nuclides are present (listed as "OTHER"), THEN [23.1] POSITION the cursor on the entry to be deleted, AND PRESS the DELETE command key (Remove). [23.2] IF the entry and Its its concentration is to be deleted, THEN ENTER "R" and press RETURN. [23.3] PRESS "Save" command key (F10) (F1 0) [23.4] ENTER "Y" at the prompt "Has this been authorized? (YIN)" [23.5] ENTER "EMS" when prompted for the password ) [24] IF a concentration for a nuclide is to be changed, THEN [24.1] POSITION the cursor on the nuclide name. [24.2] PRESS TAB to move to the concentration column. [24.3] ENTER the corresponding concentration in the concentration column. [24.4] PRESS RETURN. [24.5] PRESS "Save" command key (F10) (F1 0) [24.6] ENTER "Y" at the prompt "Has this been authorized? (YIN)" [24.7] ENTER "EMS" when prompted for the password [25] PRESS the "Process" command key (DO). [26] INPUT the monitor background, AND PRESS RETURN. [27] PRESS the "Process" command key (DO).
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031
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Page 28 of 32 Paae Appendix E (Page 5 of 5)
.1.0 1.0 INSTRUCTIONS FOR OPENING A LRW PERMIT (continued)
[28] REVIEW the Liquid Effluent Permit Screen for any limits exceeded. [29] IF limits are exceeded, THEN CONTACT Chemistry Management or designee. [30] PRESS the "Process" command key (DO).(~O). [31] ENTER "Y" to confirm opening the release permit, AND PRESS RETURN. [32] PRESS the REPORT command key (F20), AND PRESS RETURN twice to print the release permit. P*RESS \j [33] PRESS the "Quit" command key (PF4).
WBN Liquid Radwaste Tank Release 0-001-90-1 O-ODI-90-1 Unit 0 Rev. 0031 Page 29 of 32 Appendix F (Page 1 of 3) Instructions For Closing A LRW Permit 1.0 INSTRUCTIONS FOR CLOSING A LRW PERMIT [1] IF NOT previously performed, THEN ACCESS the CAS Main Menu by entering the user 10 and password at the EMS System node. [2] SELECT "Effluent Management" from the main menu. [3] SELECT "Process Liquid Permit" from the Effluent Management Menu. [4] ENTER Release Point Number, AND PRESS RETURN. [5] PRESS RETURN when when prompted for the sample number from the gamma isotopic report for the liquid sample. [6] SELECT "Close A Liquid Permit" by placing the cursor next to the menu item, AND PRESS the "Process" command key (DO). (~O). [7] PRESS TABTAS to retrieve the open permit data. [8] VERIFY the Release Point Number and Permit Number are correct, AND PRESS the "Process" command key (DO). (~O). [9] ENTER the actual start date and time of the release, AND PRESS RETURN. [10] ENTER the stop date and time of the release .from from the 001, AND PRESS RETURN. [11] PRESS RETURN at the Release Flow Rate prompt.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 30 of 32 Appendix F (Page 2 of 3) Instructions For Closing A LRW Permit 1.0 INSTRUCTIONS FOR CLOSING A LRW PERMIT (continued) [12] ENTER the percentage level of the tank at the start of the release AND PRESS RETU RN. RETURN. [13] ENTER the percentage level of the tank at the end of the release, AND PRESS RETURN. NOTE The following step will allow EMS to calculate the release volume using actual start and and stop tank levels. . [14] PRESS RETURN and CTRL-P twice to position the cursor at the % STOP prompt, THEN PRESS TAB. [15] ENTER average dilution flow rate during the release at the dilution flow rate prompt, AND PRESS RETURN. [16] ENTER RETURN at the dilution volume prompt. [17] PRESS the "Fill" command key (F14). [18] VERIFY displayed data is correct. [19] PRESS "Save" command key (F10).(F1 0). [20] PRESS the "Process" command key (~O) (DO) twice. [21] REVIEW the "Liquid Release Permit" screen for any limits exceeded.
WBN Liquid Radwaste Tank Release 0-001-90-1 Unit 0 Rev. 0031 Page 31 of 32 Appendix F (Page 3 of 3) Instructions For Closing A LRW Permit 1.0 INSTRUCTIONS FOR CLOSING A LRW PERMIT (continued) [22] IF limits are exceeded, THEN CONTACT the Chemistry Management or designee. [23] IF no limits are exceeded, THEN PRESS the "Process" command key (DO). (~O). [24] ENTER "Y," AND PRESS RETURN when prompted "Are you sure (YIN)" to close the permit. [25] PRESS the "Report" command key (F20), AND PRESS RETURN to obtain the printed release permit. [26] PRESS the "Quit" command key (PF4). [27] PRESS RETURN at the prompt "Are you sure you want to quit?" [28] PRESS the "Prev Scr" key (PF4) to return to the CAS main menu.
WBN Liquid Radwaste Tank Release 0-001-90-:1 0-ODl-90,:1 Unit 0 Rev. 0031 Page 32 of 32 Source Notes (Page 1 of 1) Implementing Requirements Statement Source Document Statement Reference TI-18 Appendices Band C WBPER960331 1 to ensure correct values are used for monitor setpoint determination. Added steps to provide actions to be WBPER960465 2 taken to ensure that a package prepared for an inoperable rad monitor is NOT used if the monitor is returned to service after package has been transmitted to Operations, but prior to the release. Deleted in Rev 21 3 '\ Added requirement to obtain Chemistry WBPER971360 4 Duty Manager approval to release tanks with gamma activity greater than 3.0E-05 mCi/ml. mCilm!. Revised requirements for ensuring that WBPER971183 5 the radiation monitor flow requirements established in WBNTSR-066 are met.
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 48 of 146 Page 4.8 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown CAUTION Complete drain down of the CVCS Monitor Tank should be avoided since it could result in unplanned gaseous release from the tritiated drain system NOTE The Unit SRO should confirm there are no planned interruptions in Hydro Plant operation which would result in isolation of normal CTBO CTBD diffuser flow, in order to minimize to the yard holding pond. transportation of contaminants to.the [1] ENSURE Monitor Tank pump(s) running by PLACING one or more of the following handswitches to START [PNL 0-L-2] or local PB's may be used (N/A any pump NOT selected).
- 0-HS-77-2904B1 (0-HS-77-2904B2), MONITOR TANK PUMPA
- 0-HS-77-2906B1 (0-HS-77-2906B2), MONITOR TANK PUMPB
[2] VERIFY the Monitor Tank has recirced for at least 17 15 minutes. 17 [3] REQUEST Chemistry Countroom PERFORM 0-ODI-90-1. 0-001-90-1. [4] IF 0-RE-90-122 is OOS, THEN
- PERFORM 0-RE-90-122 INOPERABLE steps, OR
- REQUEST REQUESTMIG MIG to ENSURE 0-RE-90-122 properly aligned for service. 88
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 49 of 146 Date Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [5] 0-RE-90-122 is INOPERABLE, THEN IF O-RE-90-122 PERFORM the following: PERFORM*the [5.1 ] [5.1] REMOVE fuses N-17 and N-18 in PaneI1-R-76. Panel 1-R-76. CV [5.2] PLACE temporary jumper (clip-type) at TB-248 across terminals 8 and 10 in Panel PaneI1-R-72. 1-R-72. CV [5.3] IDENTIFY jumper with an Information Tag noting procedure and step number. [6] IF 0-LPF-27-98 (0-FR-27-98) is INOPERABLE, THEN PERFORM the following: [6.1 ] [6.1] PLACE temporary jumper (clip type) across terminals 197 and 198 in Panel 0-R-144 CV [6.2] IDENTIFY jumper with an Information Tag noting procedure and step number. [6.3] ENSURE 1-SI-0-2-00, INOPERABLE DIFFUSER DISCHARGE EFFLUENT FLOW MONITOR MONITOR log, is being 12 performed in conjunction with 0-001-90-1. 12 [6.4] RECORD total dilution flowrate as estimated from 1-SI-0-2-00 FLOWRATE gpm
,gpm
[7] OBTAIN 0-001-90-1 Release Package, AND VERIFY Release Permit has been approved by SM/UNIT SRO.
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 50 of 146 - Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [8] IF 0-RE-90-122 is OPERABLE, THEN ENSURE SOURCE CHECK has been performed. SRO [9] OBTAIN SRO approval and verification that release is authorized,and procedures are correct for release of the Monitor Tank. SRO NOTES
- 1) The following step may be performed earlier at the discretion of the SRO; however, Chemistry degradation in the SGs may coincide with flow reduction.
- 2) -Back Back pressure on the release line due to S/G blowdown flow can extend the time required for release by several hours. Reducing S/G blowdown to 150 gpm (or less) can reduce the release duration to between 1.5 to 2 hours.
[10] IF S/G blowdown flow is greater than 150 gpm AND is aligned to cooling tower blowdown, THEN: [10.1] RECORD S/G blowdown flow rate, FLOWRA FLOWRATE TE ,gpm gpm [10.2] CONTACT the Chemistry Duty Manager for recommended S/G blowdown flow during the release. RECORD FLOWRAFLOWRATE TE _ _ _ _ gpm [10.3] REFER TO SOI-15.01 and ADJUST S/G blowdown flow to the flow rate recommended from Step 8.13[10.2]. [11] IF Total dilution flow (72"DIFFUSER SUP FLOW) is being used to support release, THEN RECORD FLOWRATE gpm (Check indication used)(N/A if Step 8.14[6.4] performed) 0-FR-27-98, 72"DIFFUSER SUP FLOW [1-M-15] 0 Computer Pt Y8000A 0
WBN WSN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 51 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Slowdown Blowdown (continued) [12] ENSURE 0-FIT-77-5042 is OPERABLE, OR PERFORM 1-SI-0-2-00 data sheet for INOPERABLE LIQUID EFFLUENT LINE FLOW MONITOR, 0-FIT-77-5042. [13] ENSURE both of the following handswitches are in STOP [PNL 0-L-2] or local PB's may may be used:
- 0-HS-77-2904B1 (0-HS-77-2904B2),
(O-HS-77-2904B2), MONITOR TANK PUMP A.
- 0-HS-77-2906B1 (0-HS-77-2906B2), MONITOR TANK PUMP B.
CAUTION 0-THV-77-579, MONITOR TANK RECIRC THROTTLE, must remain OPEN ONE (1) TURN 11 to prevent component damage ifflowpath if flowpath isolates due to 0-RCV-77 0-RCV-77-43
-43 closure. 11
[14] FULLY CLOSE 0-THV-77-579, MONITOR TANK RECIRC [A 14U/692], THEN ISOL [A14U/692], THROTTLE OPEN 0-THV-77-579 one (1) turn. NOTE If 0-RE-90-122 is INOPERABLE the following steps that have IV blanks provided must be 16 IV'd* 16 [15] OPEN 0-ISV-77-573, MONITOR TANK PUMP OUTBOARD [A 13U/692]. (N/A IV if 0-RE-90-122 is OPERABLE.) ISOL [A13U1692]. IV [16] 0-ISV-77-2853, CVCS MONITOR TANK PUMP OPEN O-ISV-77-2853, DISCH ISOL [A7W/697]. (N/A IV if 0-RE-90-122 is OPERABLE.) IV
WSN WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062
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Page 52 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown Slowdown (continued) NOTE N/A Step 8.13[17] If 0-RE-90-122 is INOPERABLE. [17] PLACE 0-HS-90-122, WDS LlQ L1Q EFFLUENT MON PMP (0-RE-90-122) [A6W/692] to ON, AND switch (0-RE-90-122)[A6W/692] VERIFY Red Alarm Light NOT lit. [18] OPEN 0-RCV-77-43, CT BLDN LN RAD RELEASE CNTL, with 0-HS-77-43 [PNL 0-L-2] (N/A IV if 0-RE-90-122 is OPERABLE). . IV CAUTION Release via the pump which was NOT used for tank recirculation may result in unplanned isolation due to unanticipated contents in the "dead leg. [19] PLACE one or more of the following handswitches to START [PNL 0-L-2] or local PB's may be used (N/A any pump NOT selected). A. 0-HS-77-2904B1 (0-HS-77-2904B2), MONITOR TANK PUMPA B. 0-HS-77-2906B1 (0-HS-77-2906B2), MONITOR TANK PUMPB [20] UNLOCK and OPEN 0-ISV-77-660, COOLING TOWER BLOWDOWN RELEASE HEADER ISOL. [A7W/692] (N/A IV if 0-RE-90-122 is OPERABLE). IV
WBN WSN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 53 of 146 - - - - Date _ _ __ . Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown Slowdown (continued) NOTES
- 1) Simultaneous release of water from Cond DI 01 and Monitor Tank to CTBD may drastically reduce the flow rate resulting in the need to use two Monitor Tank pumps.
Flow rate must be maintained below maximum discharge rate listed on the permit.
- 2) 0-THV-77-5042B, WDS LIQUID THROTTLE ISOLATION VALVE, is the valve normally used to adjust discharge flow to meet permit requirements. If sufficient flow cannot be established via this valve, 0-ISV-77-5042A, WDS LIQUID DISCH ISOLATION, may be used for additional flow ifrequired.
[21] ADJUST 0-THV-77-5042B, WDS LIQUID THROTTLE ISOLATION VALVE and/or 0-ISV-77-5042A, WDS LIQUID DISCH ISOLATION, [A7W/692], as required to adjust the flowrate specified on the Release Permit, AND 15 RECORD FLOWRAFLOWRATE TE 15 gpm
,gpm
[22] IF O-FIT 5042, Liquid Radwaste Effluent Line flow 0-FIT-77-5042, indication is being used to support release, THEN RECORD flow as a channel check: _ _ _ _ _ _gpm g,pm (Check indication used) A. 0-FIT-77-5042 [0-L-848, A5W/692] o B. Computer Pt. F0112A o NOTE N/A Step 8.13[23] If 0-RE-90-122 is is INOPERABLE. [23] ENSURE 0-RE-90-122 flow is greater than low flow alarm setpoint and less than 10 gpm as indicated on 0-FI-90-122, AND [23.1] IF flow requirements are NOT adequate, THEN ADJUST 0-ISIV-90-122E as necessary, AND 15 FLOWRATE.
. RECORD FLOWRA TE. 15 gpm
,gpm
WBN WSN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 54 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown Slowdown (coritinued) (continued) NOTE This release will require termination; the remaining steps of this section (starting at Step 8.13[26]) performed; and a new release permit generated if any of the following occur during the release:
- 1. Loss of CTSD flow for >2 hours
- 2. Closure of O-RCV-77-43 by operator action for >2 hours
- 3. Auto closure of O-RCV-77-43 due to O-RE-90-122 exceeding its setpoint.
[24] IF O-RCV-77-43 isolates due to loss of CTSD flow, OR Operator action, THEN [24.1] FULLY OPEN O-THV-77-579, MONITOR TANK RECIRC ISOL [A14U1692]. [A14U/692]. [24.2] CL05E CLOSE O-ISV-77-660, COOLING TOWER \) SLOWDOWN RELEASE HEADER ISOL. [A7W/692] (N/A IV if O-RE-90-122 is operable). IV [25] IF CTSD flow is restored, THEN PERFORM the following: [25.1] FULLY CL05EO-THV-77-579, CLOSE O-THV-77-579, MONITOR TANK RECIRC ISOL [A14U/692, THEN THROTTLE OPEN O-THV-77-579 one (1) turn. [25.2] OPEN O-RCV-77-43, CT SLDN LN RAD RELEASE CNTL, with O-HS-77-43 [PNL O-L-2]. IV [25.3] OPEN O-ISV-77-660, COOLING TOWER SLOWDOWN
. RELEASE HEADER ISOL. [A7W/692]
IV [25.4] REPERFORM Step 8.13[21] if necessary (Note on Permit).
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062
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Page 55 of 146 p~~550f146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) CAUTION IF release is terminated by Auto pump cutoff, Operator Action or Auto closure of O-RCV 43, O-HS-90-122 should be placed in OFF within one hour to prevent radiation monitor pump damage. NOTES
- 1) IF release is terminated by Auto pump cutoff or Operator Action, PERFORM Steps 8.13[26] through 8.13[43] and N/A Steps 8.13[44] through 8.13[64].
- 2) . IF release is terminated by Auto closure of O-RCV-77-43 O-RCV-77-43 due to O-RE-90-122 exceeding its setpoint, PERFORM Steps 8.13[44] through 8.13[64] and N/A Steps 8.13[26] through 8.13[43].
[26] IF release is terminated by Auto pump cutoff or Operator Action, THEN ENSURE the following pumps have been placed in STOP using either [PNL O-L-2] or local PB's. (N/A pump NOT selected): NOMENCLATURE LOCATION UNID PERF INITIALS MONITOR TANK PUMP A O-L-2 PNL 0-L-2 O-HS-77 -290481 0-HS-77-2904B1 (LOCAL) -290482) (O-HS-77 -2904B2) MONITOR TANK PUMP B 8 PNL 0-L-2 O-HS-77 -290681 0-HS-77-2906B1 (LOCAL) (O-HS-77 -290682)
-2906B2)
[27] RECORD reading from O-RI-90-122B, WDS LlQ L1Q EFF, [PNL O-L-2] or computer point R1022A R1 022A cpm. (N/A if O-RE-90-122 is INOPERABLE).
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 56 of 146 ) Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [28] .PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIALS MONITOR TANK RECIRC ISOL A14U/692 CLOSED O-THV-77-579 0-THV-77 -579 MONITOR TANK PUMP B A13V/692 CLOSED O-ISV-77-571B . DISCHARGE ISOL MONITOR TANK PUMP A A13V/692 CLOSED O-ISV-77-571A 0-ISV-77-571A DISCHARGE ISOL NOTES
- 1) Approximately 85 gallons of release effluent is contained in the piping to the hig her can be expected to O-RE-90-122 rad monitor. Water flush volumes of 10 or higher reduce the 0-RE-90-122 chamber to less than its background setpoint.
- 2) Steps 8.13[29] through 8.13[32] direct Primary Water flush of 0-RE-90-122 and
\ \ associated release piping. Flushing should be performed even with 0-RE-90-122 inoperable..
inoperable
. [29] OPEN 0-ISV-77-3007, ISOLATION VLV VLVWASTE WASTE DISPOSAL SYSTEM [A14U1692].
[A 14U1692]. [30] ENSURE the following: NOMENCLATURE LOCATION POSITION POSITiON UNID PERF INITIALS WDS RAD MON O-RE-90-122 A7W/692 OPEN O-ISV-77-689A 0-ISV-77-689A RET TO CLG TOWER BLDN O-RE-90-122 WDS RAD MON 0-RE-90-122 A7W/692 CLOSED O-ISV-77-689B RETURN TO FDCT [31] FLUSH Release Line and 0-RE-90-122 monitor until the following conditions are met: A. An estimated 1000 gals has been flushed by 0-RE-90-122, AND RECORD flush water volume____ gal (estimated) OR
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 57 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) 8. B. IF O-RE-90-122 is OPERABLE, FLUSH UNTIL O-RI-90-122B reading stabilizes, AND RECORD O-RI-90-122B reading cpm. [32] CLOSE O-ISV-77-3007, ISOLATION VLV WASTE DISPOSAL SYSTEM [A 14U/692] [33] PLACE O-HS-77-43, COOLING TOWER BLDN LN RADIATION RELEASE CNTL, to CLOSE, AND ENSURE O-RCV-77-43 CLOSED. [PNL O-L-2] NOTES
- 1) MCR should be notified that Window 181 A, WDS RELEASE LINE LI N E 0 RM 122 LlQ RAD HI, may alarm during performance of following due to loss of power to control relay.
- 2) N/A Step 8.13[34]. IF O-RE-90-122 is INOPERABLE or not started in Step 8.13[17].
[34] PLACE O-HS-90-122, WDS LlQ EFFLUENT MON PMP.SW, PMPSW, to OFF. [A6W/692]
. [35] CLOSE and LOCK O-ISV-77-660, COOLING TOWER BLOWDOWN RELEASE HEADER ISOL [A7W/692].
CV [36] ENSURE the following discharge valves CLOSED [A7W/692].
- O-ISV-77-5042A, WDS LIQUID DISCH ISOLATION VALVE.
- O-THV-77-5042B, WDS LIQUID THROTTLE ISOLATION VALVE.
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 58 of 146 Date Date _ __ __ __ _ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [37] ENSURE the following: NOMENCLATURE LOCATION POSITION UNIO UNID PERF INITIALS MONITOR TANK RECIRC ISOL A14U/692 OPEN O-THV-77-579 MONITOR TANK PUMP A A13V/692 OPEN O-ISV-77-571A DISCHARGE ISOL MONITOR TANK PUMP B A13V/692 OPEN O-ISV-77-571B DISCHARGE ISOL MONITOR TANK PUMP A13V/692 CLOSED O~ISV-77-573 O-ISV-77-573 OVERBOARD ISOL WDS RAD MON O-RE-90-122 A7W/692 OPEN O-ISV-77-689A O-ISV-77 -689A RET TO CLG TOWER BLDN WDS RAD MON O-RE-90-122 A7W/692 CLOSED O-ISV-77-689B RETURN TO FDCT CVCS MONITOR TANK PUMP A7W/692 CLOSED O-ISV-77-2853 I DISCH I [38] IF 0-RE-90-122 fuses were removed and jumper installed in Step 8.13[5], THEN PERFORM the following: [38.1] REPLACE fuses N-17 and N-18 previously removed from Panel 1-R-76. cv [38.2] REMOVE temporary jumper at TB-248 T8-248 terminals 8 and 10 previously installed in Panel 1-R-72. cv [38.3] ENSURE information tag is removed. [39] IF 0-LPF-27-98 (0-FR-27-98) jumper was installed in Step 8.13[6], THEN PERFORM the following: [39.1] REMOVE temporary jumper previously placed across terminals 197 and 198 in Panel 0-R-144. cv
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 59 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [39.2] ENSURE information tag is removed. [40] IF S/G blowdown flow was adjusted to support the release, THEN REFER TO SOI-15.01 and ADJUST S/G blowdown flow to the flow rate recorded prior to the release (see Step 8.13[10.1] of this section). NOTE Current plant conditions may NOT require starting a monitor tank pump (e.g. low tank level). If desired to NOT start either pump then both pumps may be N/A'd. [41] START the applicable Monitor Tank Pump using either the [PNL 0-L-2] or local HS's. (N/A pump(s) NOT selected): NOMENCLATURE LOCATION UNID PERF INITIALS MONITOR TANK PUMP A PNL O-L-2 O-HS-77-2904B1 (LOCAL) (O-HS-77-2904B2) MONITOR TANK PUMP B PNL O-L-2 O-HS-77-2906B1 (LOCAL) (O-HS-77-2906B2L (O-HS-77-2906B2) ------ ----
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 61 of 146 Date Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [44] IF release terminated by AUTO closure of 0-RCV-77-43 closing due to O-RE 90-122 exceeding its setpoint, THEN PLACE 0-HS-77-43, COOLING TOWER BLDN LN RADIATION RELEASE CNTL, to CLOSE AND ENSURE 0-RCV-77-43 CLOSED. [PNL 0-L-2] [45] PLACE the operating pump(s) handswitch momentarily to STOP [PNL 0-L-2] or local PB's may be used. (N/A the non operating pump):
- 0-HS-77-2904B1 (O-HS-77-2904B2),
(0-HS-77-2904B2), MONITOR TANK PUMP A
- 0-HS-77 -2906B 1 (0-HS-77-2906B2),
0-HS-77-2906B1 (O-HS-77 -2906B2), MONITOR MON ITOR TANK PUMP B [46] RECORD reading from 0-RI:'90-122B, 0-RI-90-122B, WDS L1Q EFF, [PNL 0-L-2] cpm. N/A if 0-RE-90-122 is INOPERABLE. [47] CLOSE and LOCK 0-ISV-77-660, COOLING TOWER BLOWDOWN RELEASE HEADER ISOL [A7W/692]. CV [48] ENSURE the following discharge valves CLOSED [A7W/692].
- 0-ISV-77-5042A, WDS LIQUID DISCH ISOLATION VALVE.
- 0-THV-77-5042B, WDS LIQUID THROTTLE ISOLATION VALVE.
WBN WSN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 62 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown Slowdown (continued) NOTES
- 1) MeR should be notified that Window 181 A, WDS RELEASE LINE 0-RM-90-122 O-RM-90-122 LlQ RAD HI, may alarm during performance of following due to loss of power to control relay.
- 2) N/A Step 8.13[49] If 0-RE-90-122 is INOPERABLE or not started in Step 8.13[17].
[49] PLACE 0-HS-90-122, WDS LlQ EFFLUENT MON PMP SW, to OFF [A6W/692]. [50] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIALS WDS RAD MON O-RE-90-122 0-RE-90-122 A7W/692 CLOSED O-ISV-77-689A 0-1 SV-77 -689A RET TO CLG TOWER BLDN WDS RAD MON O-RE-90-122 0-RE-90-122 A7W/692 OPEN O-ISV-77-689B 0-ISV-77-689B RETURN TO FDCT - - -
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 63 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) NOTE5 NOTES
- 1) Approximately 85 gallons of release effluent is contained in the piping to the O-RE-90-122 rad monitor. Water flush volumes of 10 or higher can be expected to reduce the O-RE-90-122 chamber to less than its background setpoint.
- 2) Steps 8.13[51]through 8.13[51 ]through Step 8.13[57] direct Primary Water flush of O-RE-90-122 and associated release piping. Flushing should be performed even with O-RE-90-122 O-RE-90-122 inoperable.
[51] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIALS MONITOR TANK RECIRC ISOL A14U/692 CLOSED O-THV-77-579 MONITOR TANK PUMP B A13V/692 CLOSED O-ISV-77-571B O-ISV-77-571 B DISCHARGE ISOL MONITOR TANK PUMP A A13V/692 CLOSED O-ISV-77-571A DISCHARGE ISOL ---- [52] OPEN O-ISV-77-3007, ISOLATION VLV WASTE DISPOSAL SYSTEM [A 14U/692]. [A14U/692].
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 P~~4of146 Page 64 of 146 Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [53] PERFORM the following: [53.1] OPEN 0-ISV-77-661, WASTE CNDS PUMP DISCH HEADER TO FDCT ISOL. [A7W/692] [53.2] .FLUSH line for approximately 10 minutes. [53.3] CLOSE 0-ISV-77-661, WASTE CNDS PUMP DISCH HEADER TO FDCT ISOL. [A7W/692] [54] PLACE 0-HS-90-122, WDS LlQ EFFLUENT MON PMP SW, to ON. [A6W/692] [55] FLUSH Release Line and 0-RE-90-122 monitor until the following conditions are met: A. An estimated 1000 gals has been flushed by 0-RE-90-122, AND RECORD flush water volume _ _ _ gal (estimated) AND B. IF.0-RE-90-122 is OPERABLE, FLUSH UNTIL 0-RI-90-122B reading stabilizes, AND RECORD 0-RI-90-122B reading cpm. NOTES
- 1) MCR should be notified that Window 181 A, WDS RELEASE LINE 0-RM-90-122 LlQ powerto RAD HI, may alarm during performance of following due to loss of power to control relay.
- 2) N/A Step 8.13[56] if 0-RE-90-122 is INOPERABLE or not started in Step 8.13[17].
[56] PLACE 0-HS-90-122, WDS LlQ EFFLUENT MON PMP SW, to OFF. [A6W/692] [57] CLOSE 0-ISV-77-3007, ISOLATION VLV WASTE DISPOSAL SYSTEM [A 14U/692]
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 6~-'~fJ Page 65 of 146 46 ) Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [58] ENSURE the following: LOCATION POSITION UNID - PERF INITIALS NOMENCLATURE MONITOR TANK RECIRC ISOL A14U/692 OPEN O-THV-77-579 MONITOR TANK PUMP A A13V/692 OPEN O-ISV-77-571A DISCHARGE ISOL MONITOR TANK PUMP B A13V/692 OPEN O-ISV-77-571 B O-ISV-77-571B DISCHARGE ISOL MONITOR TANK PUMP A13V/692 CLOSED O-ISV-77-573 OVERBOARD OVERBOARDISOL ISOL WDS RAD MON O-RE-90-122 A7W/692 OPEN O-ISV-77-689A RET TO CLG TOWER BLDN WDS RAD MON O-RE-90-122 A7W/692 CLOSED O-ISV-77-689B RETURN TO FDCT CVCS MONITOR TANK PUMP A7W/692 CLOSED O-ISV-77-2853 DISCH [59] IF 0-RE-90-122 fuses were removed and jumper installed in Step 8.14[5], THEN PERFORM the following: [59.1] [59.1] REPLACE fuses N-17 and N-18 previously removed from Panel 1-R-76. cv [59.2] REMOVE temporary jumper at TB-248 T8-248 terminals 8 and 10 previously installed in Panel 1-R-72. cv [59.3] ENSURE information tag is removed. [60] IF 0-LPF-27 -98 (0-FR-27 0-LPF-27-98 -98) jumper was installed in (0-FR-27-98) Step 8.14[6], THEN PERFORM the following:
WBN Liquid Waste Disposal 501-77.01 Unit 1 Rev. 0062 Page 66 of 146 ') Date _ _ __ Initials 8.13 Release of Monitor Tank to Cooling Tower Blowdown (continued) [60.1] REMOVE temporary jumper previously placed across terminals 197 and 198 in Panel O-R-144. CV
- [60.2] ENSURE information tag is removed.
[61] IF S/G blowdown flow was adjusted to support the release, THEN REFER TO SOI-15.01 and ADJUST S/G blowdown flow to the flow rate recorded prior to the release (see Step 8.14[10.1] of this Section). NOTE Current plant conditions may NOT require starting a monitor tank pump (e.g. low tank level). If desired to NOT start either pump then both pumps may be N/A'd. [62] START the applicable Monitor Tank Pump using either the [PNL 0-L-2] or local HS's. (N/A pump(s) NOT selected): NOMENCLATURE LOCATION UNID PERF INITIALS MONITOR TANK PUMP A PNL 0-L-2 O-L-2 O-HS-77 -2904B 1 O-HS-77-2904B1 (LOCAL) (0-HS-77-2904B2) (O-HS-77-2904B2) MONITOR TANK PUMP B PNL O-L-2 0-L-2 0-HS-77-2906B1 O-HS-77-2906B1 (LOCAL) (0-HS-77-2906B2) (O-HS-77-2906B2) NOTE N/A Step 8.13[63] If 0-RE-90-122 is INOPERABLE [63] IF count reading on 0-RI-90-122B approaches max allowable background as specified by plant SSD 0-LPR-90-122-S, THEN INITIATE Work Order for decontamination of 0-RE-90-122. [64] RETURN Release Permit and SOl to Chemistry for package closure.
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 98. G2.3.7 098 Given the following plant conditions:
- The plant is in a refueling outage.
Irradiated fuel and control rod movements are in progress in the Spent Fuel Pit. It is now 0800 on Tuesday.
- An inspection which requires a diving operation in the Spent Fuel Pit is planned to commence at 2000 on Tuesday.
Due to the potential for high radiation dose, the Radiation Work Permit (RWP) has been written as an entry into a Very High Radiation Area (VHRA). Which one of the following describes who, in addition to the Radiation Protection Manager, must also authorize the RWP for the dive, and what restriction applies on irradiated fuel and control rod movements in the Spent Fuel Pit, in accordance with RCI-100, "Control of Radiological Work," and RCI-153, "Radiation Work Permits?" RWP Authorization SFP Fuel and Related Component Movements A":I A':" Shift Manager AND Plant Manager Irradiated fuel and fuel inserts movements stopped no later than 1600. B. Shift Manager ONLY Irradiated fuel and fuel inserts movements stopped no later than 1600. C. Shift Manager AND Plant Manager ONLY movements of fuel inserts may continue, and only if further than 10 feet away from the diving area. D. Shift Manager ONLY ONLY movements of fuel inserts may continue, and only if further than 10 feet away from the diving area. Page 68
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. CORRECT. Per RCI-100, "Control of Radiological Work," and RCI-153, "Radiation Work Permits," the RWP for diving operations (entry) in a Very High Radiation Area must be authorized by the Radiation Protection Manager, the Shift Manager, and the Plant Manager. The restriction on movement of irradiated fuel and fuel related the*Plant components is that it should be stopped at a minimum of 4 hours prior to entering the pool. B. Incorrect. Plausible, since the SFP Fuel and Related Component Movement column is correct. The Shift Manager has plausibility if applicant does not recognize that the SFP is considered a Very High Radiation Area. Even if the applicant does recognize this, it is plausible to select Shift Manager, and not think that the Plant Manager needs to authorize a diving operation, especially when given the information in the stem that the Radiation Protection Manager DOES have to authorize this RWP. C. Incorrect. Plausible, since the RWP authorization is correct. Applicant correctly recognizes that irradiated fuel movements cannot continue, but incorrectly thinks that rod movements may, perhaps reasoning that since they are not actually fuel, these movements can continue, but with a precautionary restriction of more than 10 feet from the diving operation. This value of 10 feet is recognizable, since it is a value listed in the System Description for the Spent Fuel Pool as a depth of shielding water which would remain above the fuel if SFP cooling is lost for 30 hours with no makeup water supplied to the pool. The applicant misapplies that shielding value here. D. Incorrect. The Shift Manager has plausibility if applicant does not recognize that the SFP is considered a Very High Radiation Area (since a VHRA entry also requires the Plant Manager to authorize the RWP). Even if the applicant does recognize this, it is plausible to select only the Shift Manager, and not think that the Plant Manager needs to authorize a diving operation, especially when given the information in the stem that the Radiation Protection Manager DOES also have to authorize this RWP. Applicant correctly recognizes that fuel movements cannot continue, but incorrectly thinks that rod movements may, perhaps reasoning that since they are not actually fuel, these movements can continue, but with a precautionary restriction of more than 10 feet from the diving operation. This value of 10 feet is recognizable, since it is a value listed in the System Description for the Spent Fuel Pool as a depth of shielding water which would remain above the fuel if SFP cooling is lost for 30 hours with no makeup water supplied to the pool. The applicant misapplies that shielding value here. Page 69
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 Question Number: 98 Tier: _3__ Group ~n/a KIA: G2.3.7 Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions. Importance Rating: 3.5/3.6 10 CFR Part 55: 41.12/45.10 10CFR55.43.b: Related to item 7. KIA Match: The question tests the applicant's knowledge of RWP requirements for entry into a Very High Radiation Area, and also for a specific type of work (diving) related to fuel handling. This is tested at the SRO level by testing approvals required, and management of the refueling process (Le., what type of fuel handling activity can be authorized and what are the restrictions). Technical
Reference:
RCI-100, "Control of Radiological Work," Section 2.1 O.A, 2.10.A, and B.2, Rev. 0036. RCI-153, "Radiation Work Permits," Section 2.3.10.H, 2.4.3.H, Rev. 0002. Proposed references None to be provided: Learning Objective: 3-0T-SYS078A, Spent Fuel Pit Cooling System Obj. 00. Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with the Spent Fuel Pit Cooling System that are rated 2.5 during Initial License training and 3.0 during License Operator Requalification training for the appropriate license position as identified in Appendix A. Question Source: New xX Modified Bank Bank Question History: New question Comments: Page 70
WBN Control of Radiological Work RCI-100 UnitO Unit 0 Rev. 0035 Page 35 of 62 2.9 Refueling Operations (continued) Q. Closed circuit audio and video systems will be set up in upper containment, keyway, and spent fuel pit areas to support refueling operations. Personnel are expected to use these systems to the extent practical to minimize access to radiation hazard areas. R. Both trains of Containment Purge ventilation system should be in operation during refueling activities and for activities likely to cause the generation of airborne radioactivity (such as cavity flood up, etc.), unless waived by the duty RPM. 2.10 Diving Operations
. A. Diving operations into a radiologically controlled area requires a Special ALARA Preplan Report and RWP which must be authorized by the RPM and Shift B.
Manag~ Manager.
-
The following are the minimum requirements for conducting diving operations: 55 ,6
- 1. Prior to any diving operations into a radiologically controlled area, detailed radiation surveys SHALL be conducted in the work area and areas where the diver will be traversing, including the descent area.
- 2. Handling or movement of irradiated fuel or irradiated fuel related components is not permitted during diving operations and should be restricted a minimum of 4 hours prior to entering the pool.
- 3. Irradiated fuel or irradiated fuel related comROnents comp-onents shall not be handled or*
or moved while divers are in the Reactor Cavity, SFP, Cask Loading Area, or Transfer Canal.
- 4. A staging and protective clothing (PC) doffing area should be established in a manner which promotes personnel safety and adequately supports the diving operation.
- 5. The survey shall be conducted using two independent radiation monitoring devices suitable for conducting underwater surveys (Le.,(i.e., Eberline RO-7, etc.).
- 6. If any operation is conducted that could cause the movement of activated Significant changes in dose rates, a or irradiated particles or could cause significant radiation survey SHALL be conducted.-
new radiation*survey conducted. Any surveys conducted after the start of the dive SHALL be reviewed with the diver(s).
WBN Radiation Work Permits (RWPs) RCI-153 RCI*153 Unit 0 UnitO Rev. 0002 Page 10 of 18 2.3.10 RWPs written for access to, and activities in, HRAs, LHRAs, and VHRAs shall include as a minimum: (continued) E. Specifications for placement of dosimetry. F. Specifications for RP surveillance requirements. G. Specifications for pre-job briefing. H. VHRAs require authorization by the Plant Manager, Operations Shift Manager and the RPM prior to accessing the area. A RWP logbook entry shall be made noting that permission was given by the above individuals to access the VHRA. NOTE: Persons qualified in radiation protection procedures may be exempted from the requirements for a RWP while performing radiation surveys in a HRA or LHRA, provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in the ~area. area. This exemption does NOT apply to VHRA areas. 2.3.11 RWPs written for Hot Particle Areas shall contain the following requirements as appropriate: A. Protective Clothing 19
- 1. (outer pair disposable, if possible OR single coveralls with Double coveralls {outer approval).
disposable aprons and sleeves (requires RPSS approval):
- 2. Two pair of rubber gloves (outer pair disposable, if possible)
- 3. booties (outer pair disposable, if possible)
Two pair of bootie-s
- 4. One pair of shoe covers
- 5. Face shields, as required to prevent facial contamination
- 6. Respiratory protection, as required to prevent inhalation of hot particles 2o B. Special Instructions2o ,21
- 1. Continuous RP coverage lAW RCI-100RCI-1 00
- 2. Assistance to help workers remove outer protective clothing upon exiting
-the the affected area
- 3. Pre-job, in-process and post-job decon requirements to minimize the spread of hot particles
- 4. Wipe down of respiratory protection equipment to remove hot particles
WBN Radiation Work Permits (RWPs) RCI-153 UnitO Rev. 0002 Page 11 of 18 2.3.11 RWPs written for Hot Particle Areas shall contain the following requirements as appropriate: (continued)
- 5. Instructions for the set-up of the area for the containment of hot particles
- 6. Job hold points and frequency for hot particle surveys of personnel and the work area .
- 7. Cleanliness requirements for system breaches and closures s.
S. Containment devices when reconstituting leaking fuel bundles
- 9. Use of sub-micron filters and fuel pool skimmers when hot particles are in the fuel pool and refuel cavity (when flooded) 2.4 RWP Approvals 2.4.1 At a minimum, the "Prepared By" and "Final Approval" signature is required. RWPs should be independently verified. The "Prepared By" and "Final Approval" should not be the same person.
2.4.2 The "Final Approval" signature can be given by group managers or Shift Supervisors (or designees). NOTE Prior to approving a new or revised RWP, you must SELECT the ALERT ACTIVE an-d and LOCKOUT ACTIVE check boxes under HIS-20 RWP/Maintain RWP Budgets/RWP Budgeting Tab. The Alert and Lockout percentages are normally set between SO and 95%.25 2.4.3 RPM, or designe_e, deSignee, approval is required for RWPs that involve: NOTE The RPM shall designate, deSignate, in writing, any and all designees authorized to approve RWPs, if dee~ed necessa'ry, necessary, on. hi~ behalf.. A ~opy Of. this li~t Ii~t should be maintained in the re~uired reading book so that thiS Information information IS readily available to all RP Department staff. 4
.A.
-A. Access to inside the containment polar crane wall at at power.
WBN Radiation Work Permits (RWPs) RCI-153 Unit 0 Rev. 0002 Page 12 of 18 2.4.3 RPM, or designee, approval is required for RWPs that involve: (continued) B. Access to the containment keyway area. C. .Access Access to either the containment side or the auxiliary buildingsisde building sisde of the fuel transfer canal. D. Access to the spent resin storage tank room. E. Access to high radiation areas where the expected whole body dose rate is <::;::: 3,000 mrem/hr. F. Expected whole body dose is <:: ;::: 500 mrem/entry. G. Expected extremity or skin dose rate is <::;::: 50,000 mrem/hr. H. Diving activities into radioactive pools, tanks, or ponds. I. Expected airborne levels are <:: 100 DAC. are;::: DAC. J. Access to a posted Very High Radiation Area. K. Access to the Volume Control tank room at power. L. Access to the Regenerative Heat Exchanger room. M. Access/work in the reactor cavity and/or the equipment pit. N. Movement of irradiated fuel or components (containment and AFP sides). O. Access to the fuel transfer tube during fuel movement (unless all surveys have been completed lAW RCI-100). RCI-1 00). P. Access to and/or work involving the exposure to irradiated incore instrumentation, cables or thimble tubes 12,13,14. Q. Industrial radiography radiography15. 15. 2.5 RWP Revisions 2.5.1 RWP revisions may be documented on the hard copy RWP if the RP Computer System is not available or if time does not allow for a system revision. If the hard copy changes are made, the changes must be approved as described in 2.4 16 above 16 .
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009
- 99. G 2.4.18099 2.4.18 099 Given the following Unit 1 plant conditions:
ECA-O.O, Loss of All AC Power is in effect. Depressurization of all intact SGs at the 'maximum rate' is in progress. Which ONE of the following identifies ... (1) the parameter used to determine the maximum rate and its basis and (2) a parameter used to determine which recovery procedure will be implemented when ECA-O.O is completed? A. (1) Ability to maintain pressurizer level above 10% to prevent loss of pressure control resulting in reactor vessel head voiding. (2) RCS pressure trend B. (1) Ability to maintain pressurizer level above 10% to prevent loss of pressure control resulting in reactor vessel head voiding. (2) RCS Subcooling value C. (1) Ability to maintain at least one SG narrow range level above 29% to prevent loss of sufficient heat transfer capability from the RCS. (2) RCS pressure trend D~ (1) Ability to maintain at least one SG narrow range level above 29% to prevent loss of sufficient heat transfer capability from the RCS. (2) RCS Subcooling value Page 71
11/2009 Watts Bar SRO NRC Exam - As submitted 101212009 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. The ability to maintain 1 SG level above 29% narrow range is the parameter. Loss of pressurizer level and reactor vessel head voiding may occur and are acceptable during the depressurization may stated in a note preceding step to depressurize the SGs. RCS pressure trend is not a parameter used to determine which recovery procedure to use. The parameters are RCS sub cooling subcooling and pressurizer level. Plausible because maintaining pressurizer level 10% or greater is a parameter monitored in the EOPs that results in actions being taken and RCS pressure stable or rising is used to determine if SI can be terminated or is still required. B. Incorrect. The ability to maintain 1 SG level above 29% narrow range is the parameter. Loss of pressurizer level and reactor vessel head voiding may occur and are acceptable during the depressurization may stated in a note preceding step to depressurize the SGs. The RCS subcooling value being used to determine the correct recovery procedure is correct. Plausible because maintaining pressurizer level 1 0% or greater is a parameter monitored in the EOPs that result in actions 10% being taken and using the RCS sub cooling value is correct. subcooling C. Incorrect. The ability to maintain 1 SG level above 29% narrow range is the parameter and the basis is to prevent the loss of sufficient heat transfer capability from the RCS but RCS pressure trend is not a parameter used to determine which recovery procedure to use. The parameters are RCS subcooling and pressurizer level. Plausible because SG level above 29% narrow range is correct and RCS pressure stable or rising is used to determine if SI can be terminated or is still required. D. Correct. The ability to maintain 1 SG level above 29% narrow range is the parameter and the basis is to prevent the loss of sufficient heat transfer capability from the RCS. A parameter used to determine the correct recovery procedure along with pressurizer level is the RCS subcooling value. Page 72
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 \ Question Number: 99 Tier: _3__ Group n/a KIA: G2.4.18 Knowledge of the Basis for EOPs. Importance Rating: 2.7 I 3.6 2.7/ 10 CFR Part 55: 41.10/45.13 41.10 I 45.13 10CFR55.43.b: 5 KIA Match: Applicant is required to identify the parameters that would be limiting while the steam generators were being depressurized and why. Also, the parameters that would be assessed to determine the correct recovery procedure during the loss of a shutdown power event. SRO because the question requires knowledge of the specific action implemented in a procedure and the assessment that would allow the proper procedure to be implemented when a transition from the current procedure is required. Technical
Reference:
ECA-O.O, Loss of Shutdown Power, Rev 19 Proposed references None to be provided: Learning Objective: 3-0T -ECAOOOO 3-0T-ECAOOOO
- 04. Discuss why it is important to maintain at least Intact SG NR level greater than 29% during SG depressurization per ECA-O.O.
- 08. Given a set of plant conditions, use ECA-O.O, ECA-0.1, and ECA-0.2 to correctly diagnose and implement: Action Steps, RNOs, Notes, and Cautions Question Source:
New Modified Bank x X Bank Question History: SON question 055 G2.4.18 Modified from SQN Comments: SON NRC EXAM 1/2008, from Sequoyah bank question ECA-0.0-B.1.B 002 modified Page 73
WBN LOSS OF SHUTDOWN POWER IECA-O.O ECA-O.O Rev 19 IStep II Action/Expected Response I I[Response Response Not Obtained
- 16. MONITOR Intact S/G levels:
- a. At least one Intact S/G NR level a. MAINTAIN maximum AFW greater than 29% [39% ADV]. flow until NR level greater than 29% [39% ADV].
- b. S/G NR level less than 50% b. IF any S/G NR level continues to and controlled. rise uncontrolled, THEN ISOLATE Ruptured S/G:
- ISOLATE AFW flow.
- ENSURE steam supply to TO AFW pump from Intact SG.
- ENSURE S/G Slowdown isolated.
- ENSURE Ruptured S/G PORV aligned:
- 1) ENSURE controller in AUTO set at 90%.
- 2) ENSURE HS in P-AUTO.
3). WHEN Ruptured S/G pressure less than 1130 psig, THEN a) ENSURE Ruptured S/G PORV CLOSED, OR b) OBTAIN RADPROT support and Locally CLOSE Ruptured S/G isolation valve:
- Loop 1, 1-ISV-1-619
[South Valve Room]
- Loop 2, 1-ISV-1-620
[North Valve Room]
- Loop 3, 1-ISV-1-621
[North Valve Room]
- Loop4, Loop 4, 1-ISV-1-622
[South Valve Room] 10 of 20
WBN LOSS OF SHUTDOWN POWER IECA-O.O ECA-O.O Rev 19 IStep III IAction/Expected Response II Response Not Obtained
- 17. CONTROL Intact S/G NR levels between 29% and 50%
[39% and 50% ADV].
- 18. EVALUATE DC bus loads:
- a. INITATE AOI-40, Station Blackout, to shed non-essential DC loads and restore AC power.
- b. MONITOR DC power supply:
- 125 Vdc vital batteries.
- 250 Vdc batteries.
- c. NOTIFY TSC to evaluate other non-essential DC loads.
- 19. MONITOR CST volume greater NOTIFY TSC to evaluated alternate than 200,000 gal. CST makeup source.
INITIATE CST refill USING SOI-59.01, Demineralized Water System. IF CST volume drops to less than 5000 gal, THEN DISPATCH operators to AFW pumps to manually perform suction transfer to ERCW. 11 of 20
WBN LOSS OF SHUTDOWN POWER IECA-O.O ECA-O.O Rev 19 IStep IIIIAction/Expected Response III[Response Response Not Obtained CAUTION RCS press should not be reduced to less than 250 psig to prevent injection of accumulator nitrogen into the RCS. NOTE
- ReS cooldown and depressurization should be performed as quickly as possible to minimize ReS inventory loss from the Rep seals.
- Pzr level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/Gs. Depressurization should NOT be stopped to prevent these occurrences.
- 20. DEPRESSURIZE Intact S/Gs to 300 psig:
- a. MONITOR reactor subcritical: a. CONTROL S/G PORVs to stop ReS depressurization, and Intermediate range startup rate ALLOW ReS heatup (negative zero or negative.
moderator coefficient). Source range startup rate zero or negative.
- b. MAINTAIN T-cold greater than OF to avoid PTS concerns.
270 of
- c. MAINTAIN ReS press greater than 250 psig.
- d. MAINTAIN at least one d. STOP S/G depressurization Intact S/G NR level greater AND than 29% [39% ADV]. MAINTAIN maximum AFW flow UNTIL NR level is restored in at least one S/G.
WHEN S/G NR level greater than 29% [39% ADV] , THEN PERFORM steps 20e thru g.
** GO TO Step 21.
Step continued on next page 12 of 20
WBN LOSS OF SHUTDOWN POWER I ECA-O.O ECA-D.D Rev 19 IStep I IAction/Expected Response II Response Not Obtained
- 20. (continued)
- e. ENSURE S/G PORV control e. DISPATCH operators to locally available. control S/G PORVs.
- f. DUMP steam at maximum rate with S/G PORVs:
1OO°F/hr cooldown limit not applicable.
- g. CHECK intact S/G pressures less g. WHEN intact S/G pressures less than 300 psig. than 300 psig, THEN CONTROL intact S/G PORVs to maintain intact S/G pressures at 300 psig.
** GO TO Step 21.
NOTE Due to possible instrument errors, accumulator injection may begin at an indicated RCS press of 950 psig.
- 21. MONITOR accumulator injection during RCS cooldown:
- a. MAINTAIN RCS press a. OPEN any unisolated greater than 250 psig to prevent accumulator nitrogen make up nitrogen injection. valve:
- 1-FCV-63-127 accumulator 1.
- 1-FCV-63-107 accumulator 2.
1-FCV-63-87 accumulator 3. 1-FCV-63-63 accumulator 4. OPEN 1-FCV-63-65 vent header. 13 of 20
WBN LOSS OF SHUTDOWN POWER IECA-O.O ECA-O.O 1Rev 19 IStep II Action/Expected Response III[Response Response Not Obtained NOTE If RCP seal cooling was previously isolated, further cooling of the RCP seals will be established by natural circulation cooldown as directed in subsequent guidelines.
- 31. DETERMINE recovery Instruction:
- a. CHECK RCS subcooling greater a. ** GO TO ECA-0.2, ECA-O.2, Recovery than 65°F [85°F ADV]. From Loss of Shutdown Power With SI Required.
- b. CHECK pzr level greater b. ** GO TO ECA-O.2, Recovery than 15% [33% ADV]. From Loss of Shutdown Power With SI Required.
- c. CHECK ECCS equipment: c. ** GO TO ECA-0.2, Recovery From Loss of Shutdown Power
- SI pumps in PULL TO LOCK.
With SI Required.
- RHR PUMPS in PULL TO LOCK.
- CCPs in PULL TO LOCK OR BIT isolated.
- d. ** GO TO ECA-0.1, Recovery From Loss of Shutdown Power Without SI Required.
--End-End-18 of 20
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 100. G2.4.3 100 Which one of the following describes the listed instrumentation as follows:
- If it is controlled by LCO 3.3.3, "PAM Instrumentation," and
- If it is included in EPIP-1, "Emergency Plan Classification Flowchart," as an indication for determining EALs?
Instrumentation LeO LCO 3.3.3 EPIP-1 A. Aux. Bldg. Passive Sump Level YES YES (0-LS-40-21 B) B. Aux. Bldg. Passive Sump Level NO NO (0-LS-40-21 B) C. Condenser Vacuum Pp. YES NO Exhaust Vent (Noble Gas) (1-RE-90-404A1B) D~ Condenser Vacuum Pp. D'r' NO YES Exhaust Vent (Noble Gas) (1-RE-90-404A1B) Page 74
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. It is plausible that if an applicant is not familiar with the categories of PAM instrumentation, they could think that a particular instrument is important enough to ensure it is available during the diagnostic phase of a potential accident; i.e., having the ability to diagnose a steam generator tube leak or rupture, and therefore, need to be controlled by Tech. Specs. They could be confused by the "post-accident" aspect of PAM instrumentation. It is also plausible that an applicant
post-accident
could believe that this instrument is used to determine which EAL applies, since it is reasonable to think that a sump level is in indication of leakage. B. Incorrect. The Auxiliary Building Passive Sump Level instrument is a Category 1 PAM instrument and as such, is subject to LCO LCD requirements for operability. However, it is highly plausible for an applicant to think that an instrument for sump level in the Auxiliary Building would NOT be governed by Tech. Specs., especially most references in Tech. Specs. for "sump" are related to the containment sump. This misconception is given even more plausibility by the fact that there are many PAM instruments that are not controlled by Tech. Specs. Of the over 100 PAM instruments at Watts Bar, approximately 26 of them are controlled by an LCO. LCD. The E-plan column is correct, adding further overall plausibility to this distractor. C. Incorrect. It is plausible that if an applicant is not familiar with the categories of PAM instrumentation, they could think that a particular instrument is important enough to ensure it is available during the diagnostic phase of a potential accident; i.e., having the ability to diagnose a steam generator tube leak or rupture, and therefore, need to be controlled by Tech. Specs. They could be confused by the "post-accident" aspect of PAM instrumentation and select this distractor. D. CORRECT. Per the given references, the Condenser Vacuum Pump Exhaust Vent Monitor (Noble Gas) is a Post Accident Monitoring instrument, and is listed in EPIP-1, Table 7-1 as an indicator to be used when determining EALs, however, the instrument is NOT controlled by any LCO.LCD. Page 75
11/2009 Watts Bar SRO NRC Exam - As submitted 10/2/2009 "\ Question Number: 100 Tier: _3__ Group n/a KIA: G2.4.3 Emergency ProcedureslPlan Procedures/Plan Ability to identify post-accident instrumentation Importance Rating: 3.7 1/ 3.9 10 CFR Part 55: 41.6 / 45.4 41.6/45.4 10CFR55.43.b: 1 KIA Match: Applicant must identify which instrument is controlled by the Post-Accident Monitoring Tech. Spec., and from the SRO perspective, is also used during Emergency Plan implementation. Technical
Reference:
EPIP-1, Table 7-1, "Effluent Radiation Monitor EALs" LCO 3.3.3, "PAM Instrumentation" Table 3.3.3-1, "Post Accident Monitoring Instrumentation" FSAR Table 7.5-2, "Post Accident Monitoring Variables
\ List" System Description N3-77C-4001, "Liquid Radwaste Processing System," Section 3.3.5 Proposed references None to be provided:
Learning Objective: 3-0T-T/S0303 Obj. 1 Demonstrate the ability to extract specific information from the Technical Specification, and Technical Requirements, as they pertain to Instrumentation Systems. Question Source: New x X Modified Bank Bank Question History: New question Comments: Page 76
TVA
Title:
LIQUID RADWASTE PROCESSING SYSTEM N3-77C-4001 3.3.5 Passive Sump Instrumentation The passive sump located in the auxiliary building is designed to receive shield building annulus floor drains and in the event of a pipe break, receive water from the containment spray pump room and/or pipe chase on elevation 676.0. The passive sump instrumentation is designed to detect eleva'tion two types of leaks. to detect a leakage rate First, the instrumentation is to. of 50 gal/min into the passive sump from areas throughout the auxiliary building. Secondly, the level indicators are part of the postaccident monitoring system which provide sump level indication to the operator in the control room. Water in the sump is drained to the Auxiliary Building floor and equipment drain sump (ABF&EDS) via 6" drain valve DRV-77-916 bottom of the sump at elevation 665'-9" with which is located at the 'bottom mechanical linkage provided to a hand wheel on the floor above on elevation 676'-2". The sump level indication is provided by a qualified differential pressure transmitter with capillary tubes mounted to the sump wall. The upper capillary tap is mounted 13 inches below the top of the passive sump at elevation 672'-6.5", and the lower tap is mounted near the floor of the sump at elevation 667' -6'"
-6" to accommodate the capillary length length...- The span.
span on the transmitters is 0-60 inches, and the span on the level indicators is from 12.5-72.5 inches. A signal of zero transmitted by the level transmitter means the level indicator is reading 12.5, and a level transmitter signal o£of 60 inches of water results in a level. indicator reading of 72.5. When the setpoint is reached the operator takes manual action to empty the sump. . A level alarm from level \ switch (0-LS-40-21B) will annunciate in the main control room when the setpoint (El(EI 668'-1") is reached. This alarm will alert the operator that a pipe break may have occurred. The operator in the control room shall monitor the level of water in the passive sump and take manual action to prevent flooding of the sump by opening the drain valve. The accuracy of the level indicators required at worst conditions to give indication of water level in the sump has been evaluated and documented in 7.1.5.23.. Refs 7.1.5.13 and 7.1,5.23 . 3.3.6 Deleted 3.4 System Precautions and Limitations (Ref 7.1.2.35)
- 1. Provisions shall be made to sample and analyze liquids prior to discharge to verify radioactive limits of 10 CFR 20 and 50 are not exceeded. A liquid release shall be made only after the issuance of a release permit.
- 2. Liquid, whose tritium concentration is 10 percent or more of the primary coolant tritium concentration, should be drained to the TDCT, while liquid having a tritium concentration of less than 10 percent of the primary coolant should be routed to the FDCT as nontritiated water.
- 3. The loading of radionuclides on the mobile waste demineralizer is limited 3
to 1 microCi/cm3 of isotopes with half-lives greater than 5 years. 24
WBNP-1 fS~~ TABLE 7.5-2 (Sheet 4 of 42) REGULATORY GUIDE 1.97 POST ACCIDENT MONITORING VARIABLES LIST VAR Redundant Minimum Minimum Range NUM Variable Name Type/Category Channels Range From Range To Units Notes 16 Sub cooling Margin Al B2 Cl C1 D2 PI P2 200 35 DegF 200 Deg. Subcooling Monitor to 35 Deg. Superheat Notes 9 && 10 17 Auxiliary Building B1 C1 PI P2 - 12.5 c 72.5 Inches Note 9 Passive Sump Level 18 Containment Isolation B1 D2 1 Per Valve Closed Not N/A* N/A Deviation #20 Valve Position Indication Closed 19 Containment Hydrogen B1 C1 D2 PI P2 0 10 % Deviation #2 Concentration 20 Control Rod Rod Position D3 1 Channel 0 235 Steps Deviation #35 Per Bank 21 Nuclear Instrumentation B1 D2 PI P2 1.0E-8 200 % Power Note 9 (Intermediate Range) 22 REACTOR VESSEL LEVEL B1 C1 D2 PI P2 See below (See (See Notes 5, 9, && 10) , 22a Static Mode 0 100 % O%represents reactor (Pumps Not Not Running) vessel empty. 100% represents reactor vessel full.
WBNP-1 TABLE 7.5-2 fsA~ fSA~ (Sheet 13 of 42) REGULATORY GUIDE 1.97 POST ACCIDENT MONITORING VARIABLES LIST VAR Redundant Minimum Minimum Range NUM Variable Name Type/Category Channels Range From Range To Units Notes 89 Plant and Environs E3 Portable 1.0E-3 1.0E4 Rad/hr. Radlhr. Radiation 90 Plant and Environs E3 Portable NA NA NA Multi Channel Radioactivity Gamma Ray Spectrometer 91 Auxiliary Building Vent E2 1 Channel 1.0E-6 1.0E-2 )lCi/cc Il Ci/ cc Deviation #13 (Noble Gas) 92 Auxiliary Building Vent E2.. E2 1 Channel 0 250,800 CFM (Flow Rate) 93 Auxiliary Building Vent E3 1 Channel ----See Note 11---- )lCi/cc Il Ci/ cc Sampling With (Particulates and Onsite Halogens) Analysis Capability Deviation
#14 94 Condenser Vacuum Pump E2 1 Channel 0 45 SCFM Exhaust Vent (Flow Rate) 95 Condenser Vacuum Pump C3 E2 1 Channel 4.0E-7 2.4E+3 )lCi/cc Il Ci/cc Deviation #33 Exhaust Vent (Noble Gas)
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2) Post Accident Monitoring Instrumentation APPLICABLE MODES CONDITION OR OTHER REFERENCED FROM SPECIFIED REQUIRED REQUIRED FUNCTION CONDITIONS CHANNELSfTRAINS ACTION E.1
- 1. Intermediate Range Neutron 1(a), 2(b), 3 2 F Flux(g)
- 2. Source Range Neutron Flux 2(C),3 2(C) , 3 2 F
- 3. Reactor Coolant System (RCS) Hot 1,2,3 1 per loop F Leg Temperature (T-Hot)
- 4. RCS Cold Leg Temperature (T- 1,2,3 1 per loop F Cold)
- 5. RCS Pressure (Wide Range) 1,2,3 3 F
- 6. Reactor Vessel Water Level (g)
(I) (g) 1,2,3 2 G
- 7. Containment Sump Water Level 1,2,3 2 F (Wide Range)
- 8. Containment Lower Compo 1,2,3 2 F Atm. Temperature
- 9. Containment Pressure (Wide 1,2,3 2 F Range) (g)
- 10. Containment Pressure (Narrow 1,2,3 4 F Range)
- 11. Containment Isolation Valve 1,2,3 2 per penetration flow F Position (g) path (d)(i)
- 12. Containment Radiation (High 1,2,3 2 upper containment G Range) 2 lower containment
- 13. Containment Hydrogen 1,2,3 2 F Concentration (g)
- 14. RCS Pressurizer Level 1,2,3 3 F
- 15. Steam Generator (SG) Water Level 1,2,3 1/SG F (Wide Range)(g)
- 16. Steam Generator Water Level 1,2,3 3/SG F (Narrow Range)
- 17. AFW Valve Status Ul 1,2,3 1 per valve F
- 18. Core Exit Temperature- 1,2,3 2(8) 2 (e) F Quadrant 1(I)
(continued) Watts Bar-Unit 1 3.3-44
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2) Post Accident Monitoring Instrumentation APPLICABLE MODES CONDITION OR OTHER REFERENCED FROM SPECIFIED REQUIRED REQUIRED FUNCTION CONDITIONS CHANNELSfTRAINS CHANNELSITRAINS ACTION E.1
- 19. Core Exit Temperature- 1,2,3 2 (e) F Quadrant 2(1)
- 20. Core Exit Temperature- 1,2,3 2 (e)
F Quadrant 3(1)
- 21. Core Exit Temperature- 1,2,3 .2 (e) F Quadrant 4(1)
- 22. Auxiliary Feedwater Flow 1,2,3 2/SG F
- 23. Reactor Coolant System 1,2,3 2 F Subcooling Margin Monitor (h)
- 24. Refueling Water Storage Tank 1,2,3 2 F Water Level
- 25. Steam Generator Pressure 1,2,3 2/SG F
- 26. Auxiliary Building Passive Sump 1,2,3 2 F Level UJ U)
(a) Below the P-10 P-1 0 (Power Range Neutron Flux) interlocks. (b) Above the P-6 (Intermediate Range Neutron Flux) interlocks. (c) Below the P-6 (Intermediate Range Neutron Flux) interlocks. (d) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, valve, closed manual valve, blind flange, pressure relief valve, or
. check valve with flow through the valve secured.
(e) A channel consists of two core exit thermocouples (CETs). (f) The ICCM provides these functions on a plasma display. (g) Regulatory Guide 1.97, non-Type A, Category 1 Variables. (h) This function is displayed on the ICCM plasma display and digital panel meters. (i) Only one position indication channel isrequired is required for penetration flow paths with only one installed control room indication channel. U)
- 0) Watts Bar specific (not required by Regulatory Guide 1.97) non-Type A Category 1 variable.
Watts Bar-Unit 1 3.3-45
EPIP-1 EPIP-l Revision 31 Page 43 of47 TABLE 7-1 EFFLUENT RADIATION MONITOR EALS NOTE: The values below, if exceeded, indicate the need to perform the specified assessment. If the assessment R can not be completed within 15 minutes (60 minutes for DE), UE), the declaration shall be made based on the A VALID reading. As used here, the radiation monitor indications as displayed on ICS are the primary indicators. InCS If ICS is unavailable, utilize the radiation monitor readings in the control room or local D indication as necessary. I o Monitor ICS Screen Units UE Alert Site General L o Total Site (GAS) EFFI /lCi/s j.,LCi/s (2) 2.59E+05 2.59E+07 2.71E+07 2.7IE+07 2.7lE+08 2.71E+08 G Ul U1 Shield Building I . l-RE-90-400 I-RE-90-400 EFFI j.,LCi/s
/lCi/s 2.59E+05 2.59E+07 2.71E+07 2.71E+08 2.7IE+08 C
U2 Shield Building 2-RE-90-400 EFFI j.,LCi/s
/lCi/s 2.59E+05 2.59E+07 2.71E+07 2.7lE+08 2.71E+08 A Auxiliary Building L 0-RE-90-101B O-RE-90-I01B 4RMI cpm 6.22E+04 6.22E+06 6.52E+06 ,'*****(1)
!' ~*t~bi;(1f),*i~ /
Service Building F 0-RE-90-132B O-RE-90-I32B 4RMl 4RMI cpm 1.42E+06 ~'¥*1:r~nf~
*****(1) ~tf;~£:(t~tf~
*****(1) *'!':f~li<**~aWi;
*****(1) U Ul UI Condenser Vacuum E Exhaust l-RE-90-404A 3PAM j.,LCi/cc(3)
/lCi/CC(3) 22E+01 1.22E+Ol
- 1. 1.22E+03 1.28E+03 1.28E+04 L
I-RE-90-404A l-RE-90-404B I-RE-90-404B 3PAM j.,LCi/cc
/lCi/cc 1.22E+Ol 1.22E+03 1.28E+03 1.28E+04 S/G Discharge H Monitors A l-RE-90-421 I-RE-90-42I thmthru 4RM2 mRlhr(4) NA 7.21E+02 7.55E+02 7.55E+03 N 424 D Total Site (LIQUID) N/A j.,LCi/ml(2)
/lCi/ml(2) 1.0lE-02 1.01E-02 1.0lE+00 1.01E+00 N/A N/A L
~~¥i~~~5 0-RE-90-122 O-RE-90-I22 4RM2 cpm 1.00E+06 *****(1) N/A N/A I
~~~~i.H~:
1-RE-90-120,121 I-RE-90-I20,12I 4RM2 cpm 5.68E+05 *****(1) N/A N/A 0-RE-90-225 O-RE-90-22S 4RM2 cpm 9.92E+05 *****(1) N/A N/A N 0-RE-90-212 O-RE-90-212 4RM2 cpm 7.64E+03 7.64E+05 N/A N/A G RELEASE DURATION minutes 60 15 15 15 U ASSESSMENT METHOD: ICS or radiation monitor (RM) readings in the MCR or local indication as necessary 1 Notes: (1) Table values are calculated values. The ***** mdlcates the momtor IS off scale, and other confirmatory confinnatory data is required for event classification. (2) In all cases, the total site EAL is the limiting value. Therefore, in the case where there are multiple release paths from the plant, it is the total release EAL (obtained from ICS or other analysis) that will determine whether an emergency classification is warranted. (3) This eberline channel reads out in cpm in the MCR. Indications of a radioactivity release via this pathway would be S/G blowdown monitors or other indications of primary-to-secondary leakage such as S/G level increase or pressurizer level decrease. ICS calculates /lCi/cc j.,LCi/cc and has a visual indication of an alarm condition when the indications exceeds 12.2 /lCi/cc.j.,LCi/cc. This channel was included in the table to provide a means to further assess a release detected by other indications and to provide a path for possible escalation. (4) These unit values are based on flow rates through one [1] PORV of 970,000 lb/hr Ib/hr at 1,185 psig, 600°F. Before using these values, ensure a release to the environment is ongoing (e.g. PORV).}}