ML25007A040

Licensee Post Exam Comments

ML25007A040
Person / Time
Site:	Saint Lucie
Issue date:	01/07/2025
From:	NRC/RGN-II
To:
References
Download: ML25007A040 (1)

Text

QUESTION10PSLL241POSTEXAMCOMMENTS TableofContents Applicantcomment Page1 Originalqueson,asgiven Page2 Originalquesonpedigree Page3to9 Staonrecommendaon Supporngdocuments Page10 Page11to15 Applicant(DocketNumber33534)commentsubmiedtotheStaon:

Recommendaccepngtwoanswerstoqueson10.AandBarebothcorrect.PerCEN152,BEE9.5, page35,the"rstparagraphstatetherearetwogoalsthatwillbeaccomplished.AnswerAiscorrect.

AnswerBisalsocorrect.MinimizethepotenalforborondiluonintheRCS.

ML25007A040

10

10.

Given:

Unit 2 experienced a Steam Generator Tube Rupture (SGTR) and Loss of Offsite Power (LOOP) 2-EOP-04, Steam Generator Tube Rupture, is in progress Safety Injection Actuation Signal (SIAS) has actuated 2-EOP-99, Appendix R, Steam Generator Isolation, has been COMPLETED for the affected SG Which ONE of the following completes the statement?

In accordance with 2-EOP-04, pressurizer pressure must be ESTABLISHED and MAINTAINED within 50 psia of the most affected SG pressure to ________.

A.

minimize RCS inventory loss to the isolated SG B.

minimize the potential for boron dilution in the RCS C.

maintain adequate subcooling in the lower regions of the isolated SG D.

maintain RCS leakage into the isolated SG as a method for cooling the upper portion of the SG tubes PSL L-24-1 NRC INITIAL WRITTEN EXAM - RO Original question, as given Page 2

10 EPE 038 EK3.01 Knowledge of the reasons for the following responses and/or actions as they apply to STEAM GENERATOR TUBE RUPTURE: Controlling RCS pressure for equalizing pressure on primary and secondary sides of ruptured S/G

10.

Given:

Unit 2 experienced a Steam Generator Tube Rupture (SGTR) and Loss of Offsite Power (LOOP) 2-EOP-04, Steam Generator Tube Rupture, is in progress Safety Injection Actuation Signal (SIAS) has actuated 2-EOP-99, Appendix R, Steam Generator Isolation, has been COMPLETED for the affected SG Which ONE of the following completes the statement?

In accordance with 2-EOP-04, pressurizer pressure must be ESTABLISHED and MAINTAINED within 50 psia of the most affected SG pressure to ________.

A.

minimize RCS inventory loss to the isolated SG B.

minimize the potential for boron dilution in the RCS C.

maintain adequate subcooling in the lower regions of the isolated SG D.

maintain RCS leakage into the isolated SG as a method for cooling the upper portion of the SG tubes PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 3

10 CORRECT ANSWER:

A DISTRACTOR ANALYSIS:

A.

Correct: The reason that the RCS depressurization is maintained within 50 psi of the ruptured generator is minimizing the pressure differential between the steam generator and the RCS in SGTRs to minimize the leakage, deliberately creating a primary to secondary differential pressure to establish backflow.

B.

Incorrect: Plausible because the 2-EOP-04 RCS depressurization step ALLOWS for backflow instead of preventing it.

C.

Incorrect: Because the S/G is at a high point in the natural circulation loop, steam may accumulate in the U-tubes, breaking the single-phase natural circulation cooldown. If voiding is suspected in the affected S/G, in accordance with 2-EOP-04, Step 43.F., feed and bleed of the S/G is directed, but to the secondary, not using the RCS.

D.

Incorrect: Plausible because 2-EOP-04, Step 44, provides guidance to drain the affected S/G via backflow then refilling the S/G as a method to cool down the S/G, but only if at least one RCP is running. Backflowing the S/G without forced circulation may lead to a dilution event.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 4

10 Question Number:

10 Tier:

1 Group:

1 K/A:

EPE 038 EK3.01 Knowledge of the reasons for the following responses and/or actions as they apply to STEAM GENERATOR TUBE RUPTURE:

Controlling RCS pressure for equalizing pressure on primary and secondary sides of ruptured S/G Importance Rating:

4.1 10CFR55 Section(s) 41.5 / 41.10 / 45.6 / 45.13 K/A Match:

K/A is matched because the applicant is required to have knowledge of the reason for controlling S/G and RCS pressures within a prescribed band during a SGTR.

Technical

Reference:

2-EOP-04, Steam Generator Tube Rupture CEN-152, BD-E-4, Steam Generator Tube Rupture Recovery Guideline Basis Document Proposed references to be provided:

None Learning Objective:

PSL OPS 0702825-12 Cognitive Level:

Higher Lower X

Question Source:

New Modified Bank Bank X

Question History:

LIOT Bank question from the 2023 Audit Exam Comments:

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 5

REVISION NO.:

PROCEDURE TITLE:

PAGE:

36 STEAM GENERATOR TUBE RUPTURE SGTR 13 of 63 PROCEDURE NO.:

2-EOP-04 ST. LUCIE UNIT 2 INSTRUCTIONS CONTINGENCY ACTIONS 4.0 OPERATOR ACTIONS (continued) 12.

(continued)

NOTE If the PORVs are unavailable, then only one RCGVS vent path is required. If venting is required, then the preferred venting flow path order is as follows:

PORVs when available Pzr vent to quench tank Pzr vent to containment atmosphere Pzr vent to RX Head Vent accumulator - last resort A.

ESTABLISH and MAINTAIN pressurizer pressure to meet all of the following criteria:

Within limits of Figure 1A, RCS Pressure Temperature Less than 930 psia Greater than minimum RCP NPSH Requirement per Figure 1A, RCS Pressure Temperature Within 50 psia of the most affected S/G pressure A.1 IF pressurizer pressure can NOT be lowered and maintained within the specified criteria, THEN OPERATE the PORVs or RCGVS to reduce pressure as follows:

1.

ENSURE pressurizer level is less than 68%.

2.

PERFORM the following to cycle the PORVs:

a.

OPEN PORV Block valves.

b.

ENSURE PORV control switches are in OVERRIDE:

V1474 V1475 c.

PULL at least TWO RPS Hi Pzr Press bistables.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 6

Background Information Revision 6.0 BD-E-4 36 of 113 EPG Step:

10 Depressurize the RCS in preparation for isolating the affected SG Intent The intent of this step is to establish control of RCS pressure.

The general goals associated with RCS pressure control are: providing subcooling to support the core heat removal process, avoiding overpressure situations for PTS and RT NDT considerations, minimizing the pressure differential between the steam generator and the RCS in SGTRs to minimize the leakage, deliberately creating a primary to secondary differential pressure to establish backflow to control SG level rise or reduce SG pressure/temperature, and controlling RCS pressure below the main steam safety valve (MSSV) lift pressure to prevent uncontrolled release of radioactivity to the environment.

Method Pressurizer pressure may be reduced by any of the following:

1. Operation of Pressurizer sprays and heaters.

2. Control of charging pumps, letdown and HPSI pumps (if HPSI Stop/Throttle criteria are met).

3. As a last resort, by operating the [PORV(s) or pressurizer vent(s)]. This method would likely be chosen in the event that main and auxiliary pressurizer spray are not available and it is necessary to lower pressurizer pressure. Pressure control by this method requires close operator attention, because the resultant pressure decrease when the PORV(s) is opened can be dramatic. In addition, the operator must closely monitor RCS inventory control and pressure/ temperature conditions in the RDT/containment while utilizing this method.

Maintaining the RCS pressure within the PT limits and approximately equal to the isolated steam generator pressure [+/-50 psi] and below (O02) will minimize the loss of primary fluid to the secondary side and the possibility of overfilling the isolated SG. This action will minimize the potential for release of radiation to the environment by minimizing RCS to steam generator leakage.

Maintaining RCS pressure approximately equal to or less than the affected SG pressure allows for the backflow of secondary water into the RCS which provides several operational benefits. These benefits include:

1. SG level can be maintained within the indicating range,

2. by controlling SG level, the probability of filling the main steam piping with water is greatly reduced,

3. use of the blowdown system for SG level control can be minimized, thus minimizing contamination of the secondary

4. depressurization of the isolated SG can be performed without steaming to the condenser or to the atmosphere,

5. less secondary makeup water is required for the RCS cooldown.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 7

REVISION NO.:

PROCEDURE TITLE:

PAGE:

36 STEAM GENERATOR TUBE RUPTURE SGTR 45 of 63 PROCEDURE NO.:

2-EOP-04 ST. LUCIE UNIT 2 INSTRUCTIONS CONTINGENCY ACTIONS 4.0 OPERATOR ACTIONS (continued) 43.

(continued)

F.

IF depressurization to SDC entry conditions is NOT possible and S/G U-tube voids are suspected, THEN PERFORM the following:

(1)

STEAM the suspect S/G using ONE of the following methods:

SBCS ADVs (2)

FEED and BLEED the suspect S/G using feedwater and SGBD to maintain level between 60 and 70% NR PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 8

REVISION NO.:

PROCEDURE TITLE:

PAGE:

36 STEAM GENERATOR TUBE RUPTURE SGTR 46 of 63 PROCEDURE NO.:

2-EOP-04 ST. LUCIE UNIT 2 INSTRUCTIONS CONTINGENCY ACTIONS 4.0 OPERATOR ACTIONS (continued)

NOTE The following listed methods for isolated S/G cooldown are arranged in a preferred order from the perspective of minimizing radiological impact.

The Main Feedwater Pump trip setpoint is 81% (2/4 narrow range).

44.

COOL and DEPRESSURIZE the isolated S/G during the cooldown as necessary using any of the following methods IN THE ORDER LISTED:

A.

IF at least one RCP is OPERATING, THEN PERFORM one of the following:

(1)

DRAIN the isolated S/G via backflow to 40% NR, and REFILL it to approximately 90%

NR in a repeating cycle.

(2)

PERFORM both of the following to expose the upper portion of the S/G tubes:

a.

DRAIN using backflow to 10% NR.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 9

QUESTION 10 PSL L-24-1 POST EXAM COMMENTS Staon Recommendaon:

The Cauon on page 38 of 2-EOP-04 discusses that the back"ow to the RCS from the isolated S/G can lower RCS boron concentraon in the faulted loop. The background informaon for this step, CEN-152, BD-E-4, Page 73, discusses the concern for a slug of reduced RCS boron concentraon accumulang in the aected RCS loop. As discussed in CEN-152, BD-E-9.5, Page 35, maintaining the RCS pressure approximately equal (+/- 50 psi) to the isolated steam generator pressure will accomplish two goals: 1) minimize the loss of primary "uid to the secondary side and the possibility of over"lling the isolated steam generator; 2) minimize the amount of unborated water "owing into the RCS from the steam generator which could reduce the RCS boron concentraon. Because of this, it is the Staon's recommendaon to accept both answers A and B for queson 10.

2-EOP-04, Page 38:

CEN-152, BD-E-4, Page 73:

CEN-152, BD-E-4, Page 37:

CEN-152, BD-E-9.5, Page 35:

Station recommendation Page 10

REVISION NO.:

PROCEDURE TITLE:

PAGE:

36 STEAM GENERATOR TUBE RUPTURE SGTR 38 of 63 PROCEDURE NO.:

2-EOP-04 ST. LUCIE UNIT 2 INSTRUCTIONS CONTINGENCY ACTIONS 4.0 OPERATOR ACTIONS (continued)

CAUTION Backflow to the RCS from the isolated S/G can lower RCS boron concentration in the faulted loop. Consideration should be given to starting the first RCP in the intact loop. After observing the impact on reactivity control safety function from the first RCP start, the RCP in the faulted loop should then be started.

38.

CHECK if RCP restart criteria are MET:

A.

VERIFY RCP restart DESIRABLE.

A.1 GO TO Section 4.0, Step 39.

B.

VERIFY isolated RCS loop has NOT been diluted by S/G leakage.

B.1 IF RCS dilution possibility exists, THEN SAMPLE RCS for boron concentration before starting RCP.

Supporting documents Page 11

Background Information Revision 6.0 BD-E-4 36 of 113 EPG Step:

10 Depressurize the RCS in preparation for isolating the affected SG Intent The intent of this step is to establish control of RCS pressure.

The general goals associated with RCS pressure control are: providing subcooling to support the core heat removal process, avoiding overpressure situations for PTS and RT NDT considerations, minimizing the pressure differential between the steam generator and the RCS in SGTRs to minimize the leakage, deliberately creating a primary to secondary differential pressure to establish backflow to control SG level rise or reduce SG pressure/temperature, and controlling RCS pressure below the main steam safety valve (MSSV) lift pressure to prevent uncontrolled release of radioactivity to the environment.

Method Pressurizer pressure may be reduced by any of the following:

1. Operation of Pressurizer sprays and heaters.

2. Control of charging pumps, letdown and HPSI pumps (if HPSI Stop/Throttle criteria are met).

3. As a last resort, by operating the [PORV(s) or pressurizer vent(s)]. This method would likely be chosen in the event that main and auxiliary pressurizer spray are not available and it is necessary to lower pressurizer pressure. Pressure control by this method requires close operator attention, because the resultant pressure decrease when the PORV(s) is opened can be dramatic. In addition, the operator must closely monitor RCS inventory control and pressure/ temperature conditions in the RDT/containment while utilizing this method.

Maintaining the RCS pressure within the PT limits and approximately equal to the isolated steam generator pressure [+/-50 psi] and below (O02) will minimize the loss of primary fluid to the secondary side and the possibility of overfilling the isolated SG. This action will minimize the potential for release of radiation to the environment by minimizing RCS to steam generator leakage.

Maintaining RCS pressure approximately equal to or less than the affected SG pressure allows for the backflow of secondary water into the RCS which provides several operational benefits. These benefits include:

1. SG level can be maintained within the indicating range,

2. by controlling SG level, the probability of filling the main steam piping with water is greatly reduced,

3. use of the blowdown system for SG level control can be minimized, thus minimizing contamination of the secondary

4. depressurization of the isolated SG can be performed without steaming to the condenser or to the atmosphere,

5. less secondary makeup water is required for the RCS cooldown.

Supporting documents Page 12

Background Information Revision 6.0 BD-E-4 37 of 113 Boron dilution of the RCS will occur due to unborated secondary water flowing through the tube rupture into the RCS. However, under most circumstances, this dilution will not threaten the maintenance of (P03).

An important point that the SGTR event strategy is to maintain or restore forced circulation as soon as possible to minimize the adverse affects of RCS dilution. On the other hand, maintaining adequate NPSH for RCP operation may cause the operator to hold RCS pressure above secondary pressure until the RCS is cooled down beyond the SG isolation temperature. If continued RCP operation is possible, the requirement to maintain NPSH takes precedence over the strategy of equalizing primary pressure and secondary pressure.

During the forced circulation cooldown process the isolated steam generator may cool faster in the lower regions (see Figure E4-8). The cooling of the isolated SG steam space will significantly lag in the cooldown and cause the fluid in the lower regions to be subcooled. If the tube rupture is located in this subcooled region, as it is most likely to be, then the primary fluid can be at the same pressure as the secondary fluid and still be subcooled. However the continued depressurization of the primary during the cooldown will now be limited by the ability to depressurize the isolated SG.

During natural circulation cooldown conditions the isolated steam generator will not cool unless there is a transfer of mass in the isolated SG. This complicates RCS pressure control during the cooldown. It is desirable to cool the RCS such that the tube bundle region of the affected SG remains subcooled. Voiding in the tube bundle region can be expected and may result in the region becoming a pressurizing source for the RCS.

Maintaining the presence of subcooled liquid in affected loop will be a complicated process under natural circulation conditions. Forced circulation conditions are much more desirable and if possible should be maintained or restored. During natural circulation conditions the cooldown and depressurization of the RCS will be limited to the operator's ability to control the conditions of the isolated steam generator.

Sequence This step is sequenced prior to the isolation of the most affected steam generator, after cooling down T-hot to less than (O02).

Plant Parameters Steam Generator pressure less than the lowest MSSV lift setpoint The bases for the engineering limit is the lift setpoint of the lowest MSSV, minus lift tolerance. The intent of the application is prevent lifting an MSSV which then sticks open, resulting in an uncontrolled release to the environment, because the operator can do nothing to stop it. The operational limit, [950 psia] was derived by taking the lowest MSSV setting [1000 psia.], subtracting the lift tolerance, typically +/-1%, [+ 10 psi], and additional operational margin [40 psi].

Supporting documents Page 13

Background Information Revision 6.0 BD-E-4 73 of 113 EPG Step:

34 Check if RCP restart criteria are met:

Intent The intent of this step is to ensure the RCS and the RCP(s) selected for restart are in the proper condition to support RCP restart and operation. This step is only applicable if RCP restart is desired.

In general, forced circulation is normally the preferred mode of core cooling because it provides more responsive control of RCS temperature and pressure. But the need for forced circulation operation must be weighed against any potential risks of RCP operation. Once RCPs have been stopped, restarting of RCPs should be considered a major plant evolution and given considerable thought before restart is attempted. In addition to the RCP restart criteria given in the instruction the following items should be considered when deciding whether RCP restart is desirable: 1) the effectiveness of RCS heat removal using natural circulation, and 2) the need for main pressurizer spray capability.

Method Verify isolated RCS loop has not been diluted. If there is doubt regarding the existence of a slug of water with reduced boron concentration in the RCS, then RCS boron concentration should be sampled prior to starting RCPs to ensure the boron concentration and distribution is sufficient to prevent loss of SDM after RCP restart.

During a SGTR there is a possibility that a slug of water with reduced boron concentration may accumulate in the affected loop. Therefore, RCS boron concentration in the affected loop should be considered (determined if possible) prior to starting the RCP in that loop if there is doubt about the existence and/or size of a slug of water with reduced boron concentration in that loop.

Ensure RCS pressure and temperature within (C01) limits for RCP operation, REFER TO Executive Volume Figure T2-9. This substep is a check to see that RCS pressure and temperature conditions will support RCP restart. (C01) should include RCP operation curves that ensure RCP NPSH and reactor core subcooling requirements are within their limits. Minimum RCS subcooling should be based on [representative CET]

temperature.

In addition, to checking the RCS (C01), the operators should be aware that the pressurizer water and steam space may not have reached equilibrium saturation conditions if there was a large level change (rise) a short time ago. This may result in the pressurizer being much less able to mitigate the pressure effects of a significant outsurge.

Ensure the duration of [CCW] interruption to RCPs is [within limits for RCP restart],

REFER TO [RCP technical manual]. This substep prompts the operator to check that any cooling interruption to RCPs is within limits that allow for RCP restart. Each plant should establish guidelines for RCP restart after cooling interruption. The guidelines should be based on RCP vendor information and other appropriate sources.

Supporting documents Page 14

Background Information Revision 6.0 BD-E-9.5 STEP DESCRIPTION TABLE 35 of 236 Pressurizer pressure approximately equal to SG pressure The engineering limits provide an operational band, nominally [+/- 50 psi] that minimizes backflow, while permitting some back flow to control SG level or to aid in cooldown of the affected SG. Maintaining the RCS pressure approximately equal (+/- 50 psi) to the isolated steam generator pressure will accomplish two goals: 1) minimize the loss of primary fluid to the secondary side and the possibility of overfilling the isolated steam generator; 2) minimize the amount of unborated water flowing into the RCS from the steam generator which could reduce the RCS boron concentration.

Pressurizer pressure within the post-accident PT curves Engineering limits are derived from plant specific PT curves. The plant Technical Specifications establish operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

10 CFR 50, Appendix G, requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. It also requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code,Section III, Appendix G.

Setpoints (A09)

Pressure band for maintaining PZR pressure approximately equal to isolated SG pressure, nominally +/-50 psid (A13)

Minimum RCP NPSH limit (C01)

Post-accident PT curves (O02)

Lowest MSSV setpoint or maximum expected post-trip S/G pressure, nominally 950 psia Critical Tasks HR-4 Prevent lifting affected SG safety valves Supporting documents Page 15

QUESTION 45 PSL L-24-1 POST EXAM COMMENTS Table of Contents Applicant comment Page 1 Original queson, as given Page 2 Original queson pedigree Page 3 to 11 Staon recommendaon Supporng documents Page 12 Page 13 to 14 Applicant (Docket Number 77430) comment submited to the Staon:

C should be correct answer. Ref: 1-AOP-09.02 "Actuaon" of AFAS occurs at 19.5% narrow range on 2/4 channels. The 235 second mer does not start unl AFAS has "actuated," thus one minute following the AFAS "actuaon" the Aux Feed Valves should sll be closed because components don't posion or start unl the 235 seconds mes out aer the 19.5% actuaon.

45

45.

Given:

Unit 1 TRIPPED from 100% power 1C Auxiliary Feedwater (AFW) pump is TAGGED OUT AFAS-1 and AFAS-2 have AUTOMATICALLY ACTUATED BOTH SG levels are 23% Narrow Range (NR) and slowly RISING 1-EOP-01, Standard Post Trip Actions, are in progress 5 minutes later:

Flows have been THROTTLED to 250 gpm to each SG A Loss of Offsite Power (LOOP) occurs BOTH SG levels are 27% and slowly RISING With NO operator action, which ONE of the following completes the statements?

One (1) minute following the Auxiliary Feedwater Actuation Signal (AFAS) actuation, the AFW valves to each SG ____(1)____ be FULL OPEN.

One (1) minute after the LOOP, the AFW valves to each SG will be ____(2)____.

A.

(1) will (2)

FULL OPEN B.

(1) will (2) throttled to 250 gpm C.

(1) will NOT (2)

FULL OPEN D.

(1) will NOT (2) throttled to 250 gpm PSL L-24-1 NRC INITIAL WRITTEN EXAM - RO Original question, as given Page 2

45 061 A4.02 Ability to manually operate and/or monitor the AUXILIARY/EMERGENCY FEEDWATER SYSTEM in the control room: AFW flow

45.

Given:

Unit 1 TRIPPED from 100% power 1C Auxiliary Feedwater (AFW) pump is TAGGED OUT AFAS-1 and AFAS-2 have AUTOMATICALLY ACTUATED BOTH SG levels are 23% Narrow Range (NR) and slowly RISING 1-EOP-01, Standard Post Trip Actions, are in progress 5 minutes later:

Flows have been THROTTLED to 250 gpm to each SG A Loss of Offsite Power (LOOP) occurs BOTH SG levels are 27% and slowly RISING With NO operator action, which ONE of the following completes the statements?

One (1) minute following the Auxiliary Feedwater Actuation Signal (AFAS) actuation, the AFW valves to each SG ____(1)____ be FULL OPEN.

One (1) minute after the LOOP, the AFW valves to each SG will be ____(2)____.

A.

(1) will (2)

FULL OPEN B.

(1) will (2) throttled to 250 gpm C.

(1) will NOT (2)

FULL OPEN D.

(1) will NOT (2) throttled to 250 gpm PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 3

45 CORRECT ANSWER:

A DISTRACTOR ANALYSIS:

A.

Correct: Part 1 Correct: On an AFAS actuation signal, the header valves, MV-09-09 and MV-09-10, go full open.

Part 2 Correct: The 1A and 1B AFW pump feed regulating valves, MV-09-09 and MV-09-10 are AC powered. On a LOOP, if the valves were throttled, the valves will go full open upon a restoration of AC power.

B.

Incorrect: Part 1 Correct: See analysis 'A' Part 2 Incorrect: See analysis 'D' C.

Incorrect: Part 1 Incorrect: See analysis 'D' Part 2 Correct: See analysis 'A' D.

Incorrect: Part 1 Incorrect: Plausible because per 1-NOP-99.07, the 150 gpm limitation is required if no AFAS signal is present and level is less than 31%. In this case, the AFAS signal did actuate and is still present.

Part 2 Incorrect: Plausible because upon a loss of power, the 1A and 1B AFW MOVs fail as is until power is restored. Upon restoration of power to the AC powered AFW valves, AND an AFAS signal still present, the motor driven AFW pump discharge valves will go full open. If the 1C AFW pump were the only source of feed, the 1C AFW pump discharge valves would not reposition due to no loss of power and flow would remain 250 gpm. The valves are DC powered from the 1AB DC bus, therefore they would not reposition on a loss of AC.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 4

45 Question Number:

45 Tier:

2 Group:

1 K/A:

061 A4.02 Ability to manually operate and/or monitor the AUXILIARY/EMERGENCY FEEDWATER SYSTEM in the control room:

AFW flow Importance Rating:

4.2 10CFR55 Section(s) 41.7 K/A Match:

K/A is matched because the applicant is required to monitor for expected AFW system response flowing an AFAS actuation, and flowing a LOOP.

Technical

Reference:

PSL OPS 0702412, Auxiliary Feedwater and Auxiliary Feedwater Actuation Systems 1-NOP-99.07, Operations Hard Cards 1-ADM-03.01C, Unit 1 Power Distribution Breaker List Motor Control Center Proposed references to be provided:

None Learning Objective:

PSL OPS 0702412-19 Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question for the L-24-1 NRC RO Exam Comments:

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 5

REVISION NO.:

PROCEDURE TITLE:

PAGE:

36 OPERATIONS HARD CARDS 27 of 48 PROCEDURE NO.:

1-NOP-99.07 ST. LUCIE UNIT 1 ATTACHMENT 6 Auxiliary Feedwater Operations (Page 1 of 6) 1.0 AUXILIARY FEEDWATER OPERATIONS 1.1 Restoring Steam Generator Levels using Auxiliary Feedwater WITHOUT AFAS present CAUTION

If Steam Generator level is below 31% NR, then initial feedwater flow should be controlled to less than 150 gpm until increasing S/G level is detected when using Aux Feed for water hammer and thermal shock concerns.

This flow limit can be exceeded to restore an EOP Safety Function prior to requiring transition to the Functional Recovery procedure or escalation of an Emergency Declaration.

Feedwater should NOT be restored to a dry S/G if another S/G still contains water. Feedwater shall only be restored to the S/G that is NOT dry. If both S/Gs become dry, then feedwater shall only be restored to one S/G to reinitiate core cooling.

1.

PERFORM one or more of the following to feed the desired Steam Generator(s):

A.

FEED the 1A Steam Generator using the 1A AFW Pump (1)

START 1A AUXILIARY FEEDWATER PUMP.

(2)

THROTTLE MV-09-9, 1A AFW PUMP DISCH TO 1A S/G, as necessary to establish and maintain desired AFW flow to 1A S/G.

AND/OR B.

FEED the 1B Steam Generator using the 1B AFW Pump (1)

START 1B AUXILIARY FEEDWATER PUMP.

(2)

THROTTLE MV-09-10, 1B AFW PUMP DISCH TO 1B S/G, as necessary to establish and maintain desired AFW flow to 1B S/G.

AND/OR PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 6

Valve Operation with AFAS

• On AFAS actuation the header valves will go full open.

• Once the valve is full open, it can be throttled closed [the solenoid valves cannot be throttled or closed].

• When level reaches 29% the header valve(s) will close.

• If the operator attempts to open the header valve, it will stroke open but will close as soon as the control switch is spring returned to AUTO.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 7

Valve Operation with AFAS (cont.)

• To be able to regain control of the valves, AFAS must be reset.

• AC-powered valves will go full open from throttled position if power is lost and then restored. This happens when a LOOP occurs after the valves were throttled or closed. DC valves remain as is (they do not lose power).

• This will be experienced and practiced many times in the simulator.

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 8

PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 9

REVISION NO.:

PROCEDURE TITLE:

PAGE:

39 UNIT 1 POWER DISTRIBUTION BREAKER LIST MOTOR CONTROL CENTER 15 of 42 PROCEDURE NO.:

1-ADM-03.01C ST. LUCIE UNIT 1 ATTACHMENT 6 480V MCC 1A5 Reactor (Page 2 of 3)

BKR. NO.

CUBICLE CWD EQUIPMENT NOTES 1-41229 HF2 SPACE 1-41230 HF3 1011 Non-essential Loads Breaker 1-41231 JF1 574 Reactor Cavity Sump Pump 1A 1-41232 JF2 Lighting Panel Transf LP-112 1-41233 JF3 Lighting Panel Transf LP-114 1-41234 JF3 SPACE 1-41235 JF5 SPARE 1-41235A JF6 Non-Essential Load Feeder no breaker 1-41284 KR1 SPACE 1-41285 KR2 SPACE 1-41286 KR3 SPACE 1-41287 KR4 SPACE 1-41288 KR5 SPACE 1-41289 KR6 SPACE REAR SECTION 1-41236 AR1 268 Aux. HPSI Flow Control Va. HCV-3647 1-41237 AR2 262 HCV-3617 Aux. HPSI Flow Control Va.

1-41238 AR3 SPARE 1-41239 AR4 991 Incoming Line no breaker 1-41240 BR1 468 Elect. Equip. Room Pwr. Roof Vent RV-3 1-41241 BR2 SPACE 1-41242 BR3 1001 Battery Charger 1A 1-41243 CR1 253 Shutdown Cooling Isol. Va V-3651 1-41244 CR2 SPACE 1-41245 CR3 584 Waste Management System Heat Tracing Transf 1A 1-41246 CR4 SPARE 1-41247 CR5 Lighting Panel Tr. LP-130 1-41248 DR1 SPACE 1-41249 DR2 168 Boric Acid Makeup Tank 1A Htr A 1-41250 DR3 279 HPSI Pump Discharge Va. V-3656 1-41251 DR4 597 Hydrogen Recombiner 1A 1-41252 ER1 SPACE 1-41253 ER2 1182 Kitchen Exhaust Fan Isol Valve FCV-25-24 1-41254 ER3 476 Aux. Bldg. SWGR Supply Fan HVS-5A 1-41256 FR1 1173 Control Room South Isol. Va. FCV-25-17 1-41257 FR2 SPACE 1-41258 FR3 SPARE 1-41259 GR1 608 AFW Pump 1A Disch Va. to S/G 1A MV-09-9 1-41260 GR2 610 AFW Pump 1A Disch Va. to S/G 1B MV-09-13 1-41261 GR3 616 FW Pump 1A Disch Valve MV-09-1 1-41262 GR4 LP-131 N/E Transf Blowdown Bldg PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 10

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39 UNIT 1 POWER DISTRIBUTION BREAKER LIST MOTOR CONTROL CENTER 29 of 42 PROCEDURE NO.:

1-ADM-03.01C ST. LUCIE UNIT 1 ATTACHMENT 15 480V MCC 1B5 Reactor (Page 2 of 3)

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CUBICLE CWD EQUIPMENT NOTES 1-42024 GF1 216 Air Recirc. Cool Outlet Hdr B Valve MV-14-7 1-42025 GF2 SPARE 1-42026 GF3 SPARE 1-42027 GF4 1013 Non-essential Loads Breaker 1-42028 HF1 574 Reactor Cavity Sump Pump 1B 1-42029 HF2 Lighting Panel Transf LP-113 1-42030 HF3 480V Power Rcpts. 58, 62, 66 (Cntmt) & Jib Boom Crane 1-42031 HF4 SPARE 1-42032 HF5 SPARE 1-42032A HF6 Non-essential Feed no breaker 1-42032B JF1 SPACE REAR SECTION 1-42033 AR1 270 Safety Inject Tank Disch Va. V-3614 1-42034 AR2 266 LPSI Flow Control Va. HCV-3645 1-42035 AR3 166 Boric Acid Gravity Feed Valve V-2509 1-42036 AR4 480V PP-136 1-42038 BR1 267 HPSI Flow Cont. Va. HCV-3646 1-42039 BR2 261 HPSI Flow Cont. Va. HCV-3616 1-42040 BR3 477 Aux. Bldg SWGR Room Supply Fan HVS-5B 1-42043 CR1 1177 Shield Bldg Vent Valve FCV-25-12 1-42044 CR2 SPARE 1-42045 CR3 SPARE 1-42046 DR1 609 AFW Pump 1B Disch. Va. to S/G 1B MV-09-10 1-42047 DR2 611 AFW Pump 1B Disch. Va to S/G 1A MV-09-14 1-42048 DR3 169 Boric Acid Makeup Tank 1A Htr B 1-42049 DR4 SPARE 1-42050 DR5 480V Power Rcpts. 47, 49, 52, 54, 72 1-42051 ER1 277 HPSI Pump Disch Va. V-3654 1-42052 ER2 1183 Kitchen Exhaust Fan Isol Vlv FCV-25-25 1-42053 ER3 525 HVE-3B Reactor Support Cooling Fan 1-42054 FR1 1170 Inlet Exh. Fan Isol. Valve FCV-25-14 1-42055 FR2 90 RCP Seal Injection Valve MV-02-1 Bkr NORMALLY OFF 1-42056 FR3 147 Chem. and Vol. Control Heat Tracing Transf. 1B PSL L-24-1 RO WRITTEN EXAM PKG FINAL Original question pedigree Page 11

QUESTION 45 PSL L-24-1 POST EXAM COMMENTS Staon Recommendaon:

When discussing AFAS, the word Iniate would convenonally be used to designate when Steam Generator level reaches the process setpoint of 19.5% lowering, which begins the 235 second countdown mer and the word Actuate would convenonally designate the point at which components begin to reposion following expiraon of the 235 second countdown mer. However, Operaons procedures and system design are not consistent in the use of this convenon. For example, procedure 1-AOP-09.02 uses the words Actuate and Iniate interchangeably when discussing AFAS as highlighted in the included examples. As another example, the AFAS Iniate switches on RTGB-202 bypass the mer and cause components to immediately posion when ulized by procedure.

Accordingly, the phrase 1 minute following AFAS Signal Actuaon is not speci"c and can mean either 1 minute following Steam Generator level reaching 19.5% or 1 minute following component reposioning.

Because of this, the staon recommends accepng both A and C as correct answers.

1-AOP-09.02:

Station recommendation Page 12

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16 AUXILIARY FEEDWATER 7 of 56 PROCEDURE NO.:

1-AOP-09.02 ST. LUCIE UNIT 1 INSTRUCTIONS CONTINGENCY ACTIONS 4.2 Subsequent Operator Actions (continued) 3.

IF all of the following conditions exist:

In MODE 3 through MODE 6

SIAS is blocked

NOT already implementing an optimal recovery procedure.

THEN VERIFY at 15 minute intervals that SFSC criteria are met per Low Mode ONP for the current plant conditions.

3.1 IMPLEMENT Low Mode ONP for current plant conditions.

NOTE

Manual initiation of AFAS is allowable under the following circumstances:

Automatic actuation of the system did NOT occur after the appropriate time delay has elapsed.

When cooling down the RCS using only one Steam Generator, if the operable Steam Generator is affected by the AFAS rupture identification circuit.

During the loss of off-site power conditions, after AFAS actuation, if one feedwater header pressurizes before the other. This assumes neither feed header is ruptured.

SIAS signal trips the Main Feedwater Pumps and allows the feedwater headers to depressurize at slightly different rates.

Automatic actuation of AFAS occurs at 19.5% narrow range steam generator level on 2 out of 4 logic.

, General Information, provides AFW information.

4.

VERIFY AFAS, if required.

4.1 Manually ACTUATE AFAS using actuation switches on AFAS cabinet.

Supporting documents Page 13

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16 AUXILIARY FEEDWATER 56 of 56 PROCEDURE NO.:

1-AOP-09.02 ST. LUCIE UNIT 1 ATTACHMENT 8 General Information (Page 1 of 1) 1.

Automatic actuation of AFAS occurs at 19.5% narrow range steam generator level on 2 out of 4 logic. Reset is at 29% 3 out of 4 logic.

2.

AFAS time delay is 235 seconds.

3.

The AFAS logic identifies a steam generator as faulted/ruptured if:

The S/G level is below the AFAS setpoint, and

The other S/G has NOT been identified as ruptured, and

The S/G Pressure is 275 psi less than the other S/G, or the associated FW Header Pressure is 150 psi less than the other FW Header.

4.

Positioning the manual actuation switch to the "ACT" position will cause the associated Actuation Relays and the Lockout Relays to de-energize, thus the AFAS-1(2) components controlled by that channel will actuate. To get a full AFAS-1(2) actuation all four channels' MANUAL ACTUATION switches must be positioned to "ACT". It should be noted that both the Rupture ID and Automatic level control features are defeated by Manual Actuation.

5.

The control circuits of the electric AFW Pumps have a 15 second auto-start delay when the EDG breaker is closed.

6.

Under certain plant conditions with low feedwater flow following a reactor trip, steam flashing in the secondary system may cause the main feedwater check valves (V09280 or V09248) to lose differential pressure required to fully seat. Under these conditions, back leakage of auxiliary feedwater through these main feedwater check valves can be mitigated by closing the associated main feedwater isolation valve (HCV-09-8 or HCV-09-7). Monitoring of the affected Steam Generator water level provides indication of the isolation of potential AFW back leakage.

Supporting documents Page 14