Text

Draft - RO & SRO Written
	ML031710512
Person / Time
Site:	Three Mile Island
Issue date:	05/14/2003
From:	Gumbert R; AmerGen Energy Co
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-289/03-301 50-289/03-301
	Download: ML031710512 (323)
	v • d • e

UA# AA2.04 Page # 4.2-3 Tier # -1 W ROISRO Importance Rating 4.2 -4.3 Group # -1 Ability to determine and interpret the following as they apply to Continuous Rod Withdrawal:

reactor power and its trend.

Sequence of events:

- Reactor power is initially 50% with ICs in automatic.

- ICs T-Ave SETPOINT signal fails slowly, ramping HIGHER at 5"Flminute.

due to setpoint potentiometer malfunction.

- ICs control mode does NOT transfer to Tracking during this sequence.

Based on these conditions, identify the ONE statement below that describes initial reactor power trend and control system response to this event.

A. Reactor power RISES due to rod withdrawal.

6. Reactor power is REDUCED due to rod insertion.

C. Reactor power DOES NOT CHANGE due to CRD logic circuit operation.

D. Reactor power DOES NOT CHANGE due to T-Ave control transfer to Feedwater.

OPM Section F-03, Integrated Control System, pages 32 and 48, Rev. 11.

None.

V. D.03.01 d New 0 TMIBank TMI Question #

u 1 Modified TMl Bank Parent Question #

J Memory or Fundamental Knowledge E

Z Comprehension or Analysis 3 55.41 51.6 3 55.43 c 55.45 A Correct answer. Error produced by this failure makes T-Ave appear to be lower than setpoint. ICs will attempt to raise T-Ave by withdrawing rods - causes reactor power to rise.

B Incorrect answer. Plausible since this would be the effect of a T-Ave signal (rather than setpoint) failure in the same direction.

C Incorrect answer. T-Ave control the ICs Reactor Subsystem are enabled under conditions presented in the question.

D Incorrect answer. Tracking does not occur. Therefore Rod Control remains in automatic, and T-Ave control does not transfer to the FW Subsystem.

T-Ave setpoint potentiometer actually failed in the TMI plant in 2001.

TMI SRO Exam- May 2003 Friday, March 28, 2003

SECTION F-03 RErnSION 9 52.3 Adaptive Control

a. Large errors may develop in the subsystems as the various control hnctions are being coordinated to produce the desired load and may cause a unit Ioad offset or error, which is undesirable. Therefore the contrd system has adaptive control features which are designed to maintain the control in continuous proper calibration. The.se adaptive controls are known as calibrating integrals.

In the circuitry a calibrating integra1 operates off of an error signa1. The integral responds or changes its outpuf as long as that input error exists. Once the error is reduced to zero, the integral will hoId its output to a constant value. To describe the operation of an integral, lets assume an error signal is developed equal to one increment for a period of time and look at the output of the integral over that same period of time.

FIGURE 11- - INTEGRAL OPERATION

. . a .

+I Integral Output 0 I--% - J----- Integral Input (error sigaa At time zero the error deveIops; the integral then takes that error signal and repeats it.

The integrals output increases such that by the end of .rhe fmt time interval, its output is equal to the error signal. By the end of the second time imerval, the integrals output bas risen to two times the error signal. This action continues until the input error s i g d is reduced to zero. At time period four, the error signal is then reduced to zero. The integrals output has increased to four time the input or error signal. With the error signal at zero, the integral holds its output constant at the new Ievel. If an error of the opposite polarity developed, the integrals output would then decrease from its existing level until the error or input signal was again reduced to zero.

Integrals me used to calibrate or fine-tune the system. It can be seen that ifan error existed, the integral could then modify the demand signal in such a mariner as to have the system canceI that error. Since; it is a characteristic of an integral to hold its output constant with a zero input, once the error is canceled, the integral wit1 hold its output constant having provided the necessary modification to the demand signal to hold or maintain the system in a stable, error fie,condition, 32

SECTION F-03 REVISION 9 The heat transfer coefficient, p, is a constant as.fir as this discussion is concerned. Tsat is held constant by the action of the turbine controls or the turbine bypass valves. Thus we have three variables to deal with; Q - the heat produced in the primary which is dependent upon reactor power; A - the heat transfer area in the OTSG which is varied by changing the level on the secondary side of the OTSG; and, -

T which will vary depending on the combination of the above mentioned variables.

As reactor power is increased from 0% to 150/, Q increases. Since the steam generators are on low Ievel limits, the level in the OTSG is constant, thus A is constant. Ta is constant because of the action of the turbine bypass d v e s . So in looking at the equation, Q = pA (Twg- TSa),we can see that Tavgmust rise as the primary side heat is increased by raising reactor power for the equation to remain constant.

TQ = - -

p A (?T~% - TZ) ? indiktes increase or

-+ indicates constan t or no change From 15% to 100% load, OTSG Ievel will increase, thus increasing the Area. For this condition, increasing the Q while increasing the Area will result in a higher load with a constant Tsvg.Again, Tu is held constant, but tbis time by the turbine.

T Q = FA? (x - T;) ? indicates increase or

+ indicates cons tan t or no change 7.4 Average Reactor Temperature Control (Tw$

Tmgis the average temperature between reactor outlet temperature (Tho,) and reactor inlet temperature (Twla). The reactor demand signal is calibrated to produce a constant average reactor coolant temperature of 579OF; however, during startup this average temperature should be ramped from about 532°F to the normal average temperature value, while the load is ramped fkom 0 to 15%. Below 15%

loadythe reactor is controlled manually at the Reactor Demand station or on the Diamond Rod Control station.

From 15% to 100%power, with the reactor demand station in automatic, the reactor demand may need modifying in order to maintain Tavgat 579°F. &temperature error signal is developed by comparing

. measured versus setpoint. This error s i g n a h applied to the reactor d e d 5 r s t in the summing circuit and also by changing the gain of the multiplier. A steady state Tavg error indicates that the power level does not correspond with the necessary feedwater flow indicating a need to modi@ the reactor demand. For this purpose, integral action is applied to the, T error, which wiIl change the relationship between reactor demand and neutron power demand. The integral modifies the demand, or calibrates the demand signal, by applyjng the error to the summer and by changing the gain of the muIt iplier.

48

SYSlEP# 003 KA# AK2.05 Page t# 4.2-4 Tier # 1 2.8 Group # 1 i_/ ROlSRO Importance Rating 2.5 -

Knowledge of the interrelations between Dropped Control Rod and the following: control rod drive power supplies and logic circuits.

Initial plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- No surveillance testing or maintenance in progress.

- All CRDs energized from NORMAL power supplies.

Event:

- CRD Group 7 Rod #dropped I into the core.

Analyze the conditions above to diagnose the ONE operational event below that caused this SINGLE control rod to drop.

A. Two DC hold breakers failed open.

B. Two CRD motor fuses failed on energized phases.

C. Two Auxiliary Power Supply Programmer lamp fuses failed.

D. Two Group 7 Programmer phase indication lamp bulbs failed.

OPM section F-01, Control Rod Drive System, pages 32 and 33, Rev. 7.

u 13; New 3 TMlBank TMI Question #

3 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge

'3 Comprehension or Analysis E 55.41 .7 i 55.43 a 55.45 .7 A Incorrect answer. Entire groups 1-4 will drop, not just one rod.

B Correct answer. Two fuses blown on the same motor will drop that rod.

C Incorrect answer. Phase indication lamps will not cause any malfunction, for indication only.

D Incorrect answer. Distracter is plausible since if (only) one rod is on Aux power supply, it would drop.

However, NORMAL conditions are that all rods on normal power supplies.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

SECTION F-0 1 REVISION 7 7.2 System Power Supplies Reductant 24,215, and 5V DC power supplies (Figures 23,24 and 25 respectively) are utilized in the Control Rod Drive System. These DC supplies normally get their power from two vital busses.

Again, these sources are identical but separate, so that loss of one source would affect only one of the two power supplies.

Each redundant pair has a set of blocking diodes on its output. When these outputs are equal, both provide outputs in parallel. If an output becomes unbalanced due to a fault or drift its output is blocked. This prevents a fkult in one supply from affecting the other supply or the downstream circuitry.

7.3 DC Hold Supply The DC hold supply maintains the position of the four safety groups. Because the safety groups are normally left in one position (OUT LIMIT) during normal operation, it is not necessary to provide a able six-phase output; ormally energized through only two phases in the hold' ctually, one phase will 1 rod position ) There is no need for a capability to run the mechanism rotor, but only to maintain the rotor segment (latch) arms together. Diodes are used in place of SCRS to provide a DC output, and A and CC were chosen as the output phases. The DC hold supply, like the regulating supply, is designed with two redundant halves; one half supplies the A phase, the other the CC phase As mentioned previously, no one of the various groups may have more than 12 drives connected to it.

The DC hold supply could conceivably carry 48 CRDMs, which accounts for a good portion of the total CRD system power requirements. The output from the DC hold supply is routed through four DC breakers, two for each phase (see Figure 25).

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Form ES-401-6 Q # -003

- Page # 4.2-5 Tier # 1 v ROlSRO Importance Rating 4.0 4.1

- Group # 1 Ability to operate and / or monitor the following as they apply to Dropped Control Rod : RCS pressure and temperature 3 c; -B

- Reactor power is loo%, with ICs in full automatic.

- Group 7 Rod #Mdrops into the core, Quadrant YZ.

Based on these conditions, identify the ONE statement below that describes initial plant response when the rod reaches full insertion.

A. Pressurizer level rises.

B. RCS pressure and temperature lower.

C. Diamond rod control transfers to manual.

D. Quadrant power tilt becomes more positive in Quadrant YZ.

EP 1202-8 CRD Equipment Failure Rev 53, pg 8 section 2b-1 E New 2 TMlBank TMI Question #

2 Modified TMI Bank Parent Question #

u Memory or Fundamental Knowledge 2 Comprehension or Analysis 55.41 .7 55.43 E l 55.45 51.6 A Incorrect answer - rod drop causes power drop which causes temp/press drop which causes outsurge initially.

B Correct answer - rapid power reduction causes T-ave, and thus RCS pressure, to drop.

C Incorrect answer - a sequence fault will occur but not cause transfer to manual.

D Incorrect answer - opposite effect, tilt in that quadrant goes negative.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number W G PNUCLEAR W TMI Emergency Procsdure -l202-8 Title Revision No.

CRD quipment Failure 53

1. Unexpected In-limit indication.

2. CRD pattern asymmetrical alarm (G-2-1).

3. Absolute position shows one or more rods at 0% an both individual position indication inefers and the PPC.

4. Flux tilt as indicated by incore and out-of-core detectors and NAS display #I.

5. Reduction in reactor power level with accompanying fluctuations in reactor coolant temperature and pressure, and pressurizer level.

6. I C s RUNBACK alarm (H-1-1) if Rx power is greater than 60% and t h e Diamond Panel is in AUTO.

7. Power Distributian Limits Exceeded alarm (G-2-6).

8. PPC alarm L3039 "7 INCH ASYMMETRIC ROD".

9. Asymmetric Rod Alarm on the Diamond Rod Control Panel.

7 0. Out-lnhibit Alarm on the Ciamond Rod Control Panel if the Asymmetric Rod Alarm is received and reactor power is greater than 60%.

26. Immediate Action Automatic Action

a. If ICs is in full auto, the plant will run back to 482 MWe.

b. Possible reactor trip on Low RCS pr C. w pressurizer level may resuit from reduced reactor power a n d , T Manual Actions NOTE if the Diamond Rod Control Panel is in manual and a runback is required due to a dropped rod, the in-limit bypass piishbutton musi b e pressed to insert the group which contains the dropped rod.

8

Form ES-401-6 Q # 004 Page # 4.2-6 Tier # -1 W ROlSRO Importance Rating 3.1 -3.8 Group # -1 Knowledge of the operational implications of the following concepts as they apply to Inoperable/Stuck Control Rod: Axial power imbalance.

Initial conditions:

- Reactor power is loo%, with ICs in full automatic.

- Reactor power imbalance is 0%.

Sequence of events:

- Power reduction to 50%.

- Rod index during power reduction changed from 292 to 238.

- Stuck rod in CRD Group 7 was diagnosed at the end of the power reduction.

- CRD Group 8 position was NOT changed.

Operational plan for the next 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:

- Maintain Group 7 aligned with the stuck rod at index 238.

- Maintain reactor power at 50% by adjusting RCS boron concentration.

- CRD Group 8 position will NOT be changed.

Based on these conditions, identify the ONE set of conditions below that describes the operational implications regarding reactor power IMBALANCE for the power reduction and the stuck rod operational plan.

During the power reduction, power IMBALANCE became (1) I and during the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months power IMBALANCE will become (2) u A. (1) negative (2) more negative

6. ( I ) positive (2) more positive C. (1) negative (2) less negative D. ( I ) positive (2) less positive 11024 Power Operation, pages 16 and 17, Rev. 101.

Z New 'E TMIBank TMI Question #

'1 Modified TMI Bank Parent Question #

C Memory or Fundamental Knowledge E Comprehension or Analysis 31 55.43 .6 & 55.45 .I3 A Correct answer -1) rods inserting into top half of core lowers top half power and 2) Xenon builds up in upper u half of core, reducing power in that half even further since rods are not allowed to withdraw back out.

B Incorrect answer - both parts wrong, opposite of actual effects.

C Incorrect answer - first part right, second part wrong.

TMI SRO Exam May 2003 Friday, March 28,2003

D Incorrect answer - first part wrong, second part right (less positive is in negative direction).

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 u 0perating Procedure 1102-4 Title Revision No.

Power Operation 101 B. Stable Power Operation with Transient Xenon

1. STABILIZE reactor power at the power level specified by SM/CRS by performing the following:

IF ICs is in full automatic, THEN perform the following:

a) SET ULD Target Load Demand to desired setpoint b) SET ULD LOAD RATE OF CHANGE for 5 0.25 %/minute NOTE The TMI-1 Core response to Axial Power Changes is naturally damped.

For most power changes, the core imbalance should remain acceptable without operator intervention. &$gflmg Rods to the Steady-State band should reduce Imbalance to r ~ q u i k ~ i e v e l s .

2. MAINTAIN reactor power and establish desired control rod position by performing the following:

Perform Enclosure 2C.

'-J MONITOR reactor power using both Nl's and Calculated Heat Balance Power, (PPC ( 3 7 0 8 ) .

IF ICs is in AUTOMATIC, THEN ADJUST ULD Target Load to maintain reactor power at desired level.

IF ICs Feedwater is on Low Level Limit Control, THEN adjust reactor power using Reactor Demand Station.

IF ICs is in MANUAL, then INSERT or WITHDRAW control rods in sequence to ADJUST reactor power.

IF IMBALANCE is approaching the boundary of the normal operating envelope of the COLR Figures 4, 5 or 6, THEN refer to Section 3.4.

VERIFY Control Rods Position meets Rod Insertion Limits in COLR.

16

Nurnber TMI - Unit 1

~ W .

Operating Procedure 1102-4 Tirle Revision No.

Power Operation 101 NOTE Group 7 rods will be withdrawn to maintain stable power while Xenon concentration is increasing+:Jhis will continue until "Peak Xenon" is reached.

The time to Peak Xenon can approximated as follows:

TIME (hr.) = square root of (initial condition power OO/ - current power Yo) h) IF Reactor Power < I O % , skip the remaining steps in this section.

i) Establish and maintain desired control rod position, by performing the following:

1. Allow rod withdrawal until peak Xenon is reached by performing the following:

a) MONITOR Rod Position and NAS and RECORD Group 7 Rod position when Xenon peak is reached.

Rod Index at Peak Xenon:

b) DETERMINE the addition volume of borated water required for a 5% rod withdrawal of Group 7 from either Figure 3 or Figure 6.

Volume of 2500 ppm boron solution for 5% Rod Withdrawal:

gals/5Yo rods Volume of 15,000 ppm boron solution for 5% Rod Withdrawal:

gals/5% rods 17

Form ES-401-6 SYSIEP# QOJ KA# EK2.03 Page # 4.1-2 Tier # 1

-.J RO/SRO Importance Rating 3.5 3.6

- Group # -2 Knowledge of the interrelations between Reactor Trip and the following: Reactor trip status panel The Control Room Operator manually trips the reactor in accordance with EOP guidance.

Based on this action, identify the ONE condition below that confirms ALL CRD trip breakers have opened.

A. DSS Actuation (G-2-7) alarm actuates.

B. Breaker trip lights illuminate on 4 RPS Cabinets.

C. Protective subsystem trip lights illuminate on 4 RPS Cabinets.

D. CRD DC Trip Bkr Open (G-1-3) AND CRD AC Trip Bkr Open (G-1-4) alarms actuate.

PM Section F-02, Reactor Protection System, pages 116 and 117, Rev. 8 (10).

G New 2 TMlBank TMI Question # #I7 CRO 2000 0 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge 27 Comprehension or Analysis k-/

c 55.43 E 55.45 . I 3 A Incorrect answer - alarm indicates EITHER IG-2A or 1L-2A opened (power TO CRD system) OR that DSS relay actuates, but not that all CRD breakers are open.

B Correct answer.- power to these lights is dependent on CRD DC/AC breaker position switches.

C Incorrect answer - RPS Protective Trip lights energize when the RPS Subsystem Reactor Trip module actuates. They are not associated with actual CRD breaker positions.

D Incorrect answer - this alarm combination confirms that either breaker associated with each alarm has opened. Alarm actuation logic does not require ALL breakers to be open.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION F-02 REVISION 8 W 7.4 RPS Test Circuits and Trip Relay Operation (refer to Figures 61,62)

Each RPS main trip relay controls a single contact in the supply and return power lines for the Control Rod Drive (CRD) power supply breakers undervoltage coils (Refer to CRD section of training manual). The trip scheme for the CRD undervoltage coils is set up in such a way (refer to Figure 62) that when 2 out of 4 RPS master trip reiays are de-energized the contacts associated with those W S cabinets open in each CRD undervoltage coil supply and return power lines. The CRD undervokage coils then deenergize tripping all CRD power supply breakers, the segment arms in the Control Rod Drives release, the roller nuts disengage from the lead screw and control rods fall into the core.

Test circuitry is set up in the RPS cabinets so a single CRD breaker or breaker set (in the case of the CRD DC hold breakers) can be test tripped from its associated RPS cabinet through the use of simulate trip toggle switches on the Rx trip module in each RFS cabinet (see Figure 62). The simulate trip toggles switches will only open contacts for the CRD breaker or breaker sets W coil associated with that RPS cabinet and will not cause a reactor trip. The manual trip scheme (use of 2 or more out of 4 simulate trip toggle switches) is set up so the CRD breakers can be tested via following scheme.

B CB.11 C CB1, CB2 D I CB3. CB4 When any of the test signals are applied, (i.e., a module taken to the test position) the master trip relay is de-energized and that cabinets contact opens in the power lines of the CRD undervoltage coils. The test lamp on the test module in the RPS cabinet becomes bright and the protective subsystem lights for that cabinet will also become bright.

This will also occur if any modules shown in Figure 62 are withdrawn from the cabinet. This will place the RPS in a 1 out of 3 trip scheme. That is to say if one more of the remaining three untripped channels were to trip, the contact scheme for de-energizing the CRD W coils would be satisfied and a reactor trip would be initiated.

116

Qf,& Z P SECTION F-02 REVISION 8 Each RPS cabinet has indicating lights on the outside top to tell the operator that status of each cabinet (see Figure 61). The turbine trip and FW trip bypass lights are bright when the bypass bistables are energized (trip function bypassed). Fan failure lights will be bright when there is a loss of power to the cabinet fans or low flow as sensed by a AP switch. The Protective Subsystem lights come on bright when a channel trips due to a setpoint being exceeded, a test trip signal is inserted or a module I

withdrawn or taken to tesr, or when a simuiate test toggle switch in that RPS cabinet is taken to trip.

The Protective Subsystems lights are set up such that when a channel trips light # 1 signifies A channel tripped, light #2 signifies B channel tripped, light #3 signifies channel C and light #4 signifies channel D. When a test signal is inserted the respective light for that cabinet will become bright in the numbering sequence described above on both outside top of cabinet and inside on the test module.

When a channel is tripped or in test the respective Protective Subsystem light for that channel will be bright on all RPS cabinets @e., if A channel trips then light #I on all RPS cabinets becomes bright).

However, when the simulate trip toggle switches are used during CRD breaker testing the onIy lights that become bright are the lights above the toggle switch and the lights on the outside top of that RPS cabinet. The Manual Bypass light becomes bright when the channel is placed in manual or shutdown bypass and when the channel is placed in manual bypass a light on the test module (Figure 6 1) also becomes bright. When the shutdown bypass keyswitch is actuated not only does the outside top light become bright but a module in the right side of the RPS cabinet will have bright lights indicating the relay has picked up, the operators overhead alarm has actuated and the light above the cabinet has become bright. Thebxyaker trip light will become bright when the CRD breaker associated with that cabinet is tripped.

When an RPS channel trips due to exceeding a setpoint or putting a module in test the channel is reset by first clearing the trip condition (or taking the test module selector switch to operate), then resetting the initiating module by using a reset spring return toggle switch on the relay status module

-./

. shown in Figure 6 1. When the master trip relay is de-energized due to exceeding a limiting safery system setpoint the subsystem trip light on the relay module becomes bright.

The transfer circuitry for the non-nuclear instrument inputs to the ICs has been removed from the A RPS cabinet. Transfer of neutron power, RCS flow, and RCS pressure from Ato B RPS cabinzi input is accomplished by the SASS pushbuttons on console center. SASS operation is explained in section F-3 of the Operation Plant Manual.

117

W 4.1 Group # -

2 ROlSRO Importance Rating 4.2 -

Ability to operate and/or monitor the following as they apply to Reactor Trip: RCS Pressure and temperature.

Sequence of events:

- Reactor power was loo%, with ICs in full automatic.

- Loss of all Circulating Water pumps.

- Turbine and Reactor trip.

Based on these conditions, identify the ONE selection below that describes how to control (1) RCS temperature and (2) RCS pressure in the PT Plot post-trip window.

A. (1) Turbine bypass valves (TBVs) control OTSG pressures.

(2) Electric heaters cycle to control Pressurizer temperature.

B. (1) Atmospheric dump valves (ADVs) control OTSG pressures.

(2) Electric heaters cycle to control Pressurizer temperature.

C. (1) Turbine bypass valves (TBVs) control OTSG pressures.

(2) Spray valve (RC-V-1) cycles open to control Pressurizer temperature.

D. (I) Atmospheric dump valves (ADVs) control OTSG pressures.

(2) Letdown flow isolates to control Pressurizer level.

OP-TM-EOP-001, Reactor Trip, Steps 3.10 and 3.12, page 5, Rev. 3.

OP-TM-EOP-010 Guide 8 , RCS Pressure Control, page 20, Rev. 1.

EZ New J TMIBank TMI Question #

3 Modified TMI Bank Parent Question #

- Memory or Fundamental Knowledge Z Comprehension or Analysis B 55.41 .7 c 55.43 3l 55.45 3 . 6 A Incorrect answer. OTSG pressures control RCS temperature, but the Main Condenser is not available due to loss of circulating water pumps. Second part of the answer is correct.

B Correct answer. The Main Condenser is not available, therefore ADVs control OTSG pressures to control RCS temperature. RCS pressure will be returned to 2155 psig through operation of the Pressurizer heaters.

C Incorrect answer. First part is incorrect since the Main Condenser is not available due to loss of all circulating pumps. Second part is incorrect (RCS pressure will be low).

D Incorrect answer. First part is correct (Main Condenser is not available). Incorrect second part was an automatic action in the past.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

OP-TM-EOP -00I Revision 3 Page 5 of 11

.___--_ -_ ~ _ _ _ _ . _ _ _

I - A C T I O N / E X P E C T E D RESPONSE RESPONSE NOT OBTAINED I If neither 1D or 1E is energized 3.7 VERIFY 1D and 1E 4160V from Aux Xfmr, then INITIATE buses are energized from OP-TM-AOP-020, Loss of Station auxiliary transformers. Power .

If 1D or 1E is energized from an Aux Xfmr, then INITIATE 1107-4 and OP-TM-861-901 or 902.

3.8 VERIFY PZR Level > 20. INITIATE HPI IAW Guide 2.

3.9 VERIFY PZR Level and MU INITIATE Guide 9 RCS Inventory Tank Level are being controlled. Control.

3.10 ENSURE OTSG pressure is INITIATE Guide 6 OTSG Pressure being controlled at desired Con tro I.

J values using TBVs/ADVs.

3.1 1 DISPATCH an operator to check MSSVs status.

3.12 . VERIFY RCS pressure i INITIATE Guide 8, RCS Pressure ing toward desired Control.

condition.

3.13 VERIFY both Generator IAAT generator MW 5 zero Breakers are OPEN. or turbine speed < 1770 RPM, then OPEN G B I - 1 2 PLACE Emergency Rev PWR Bypass switch in BYPASS and OPEN GBI-02.

3.14 VERIFY the Generator field IAAT GBI-12 and GBI-02 are OPEN, then breaker is OPEN. OPEN the Generator Field Breaker 3.15 VERIFY primary and secondary INITIATE 1202-36. LOSS of Instrument Air pressure > 80 Instrument Air.

u psig.

OP-TM-EOP-010 Revision 1 Page 20 of 49 Guide 8 RCS Pressure Control (Page 1 of 2)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY the reactor is shutdown and SCM

> 25°F.

If it is required to MINIMIZE SCM and SCM

> 70°F, then lower RCS pressure to between 70' and 30" SCM.

RAISE or LOWER RCS pressure per the following direction, as needed, to maintain RCS pressure within the limits of Figure 1 and 1A.

--/

To RAISE RCS Pressure: '-

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY RCS temperature is stable or CONTROL RCS temperature.

cooldown rate is within desired band.

VERIFY Pressurizer level is stable or rising. RAISE RCS makeup or HPI.

VERIFY pressurizer level > 80" ENERGIZE Pressurizer Heater ban IF Pzr Htr power is not available, then ADJUST heater demand for the SCR INITIATE OP-TM-220-901 to transfer Grp 8 controlled heaters.

. . . . . . . . or 9 to ES power.

OP-TM-EOP-010 Revision 1 Page 21 of 49 Guide 8 RCS Pressure Control (Page 2 of 2)

To LOWER RCS Pressure:

ENSURE HPI is throttled per (Rule 2) and control RCS inventory.

VERIFY RCS temperature is stable or NOTIFY US.

decreasing.

I If pressurizer steam bubble is controlling RCS pressure, then ENSURE Pressurizer Heaters are OFF.

NOTE 7I Pressurizer cooldown rate is to be limited to less than I O O O F in any one hour VERIFY an -RCP is operating. ENSURE WDG-V-3 and WDG-V-4 are OPEN (Guide 19).

OPEN RC-V-44.

OPEN RC-V-28.

When RC Drain Tank pressure > 40 psig or RCS pressure is in desired range, then, CLOSE RC-V-28.

CLOSE RC -V-44.

G O TO END.

Form ES-401-6 KA# AK2.01

-c/

ROlSRO Importance Rating 2.7

- Group # -2 Knowledge of the interrelations between Pressurizer Vapor Space Accident (Relief Valve Stuck Open) and the following: Valves.

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- T-Ave is stable at 579°F.

- Pressurizer level is stable at 220 inches.

- OTSG pressures are stable at 910 psig.

- RCS pressure is 2110 psig and stable.

- ALL Pressurizer Heaters are energized.

- ALL Pressurizer relief tailpipe differential temperatures are rising - the highest has risen 25°F.

Based on these conditions, identify the ONE statement below that describes required manual action.

A. Close PORV Block valve RC-V-2.

B. Isolate RCS letdown using MU-V-3.

C. Secure ALL Pressurizer Heater Banks.

D. Vent the RC Drain Tank to the Vent Header.

EP 1202-29, Pressurizer System Failure, page 9, Rev. 59.

None.

u V.D. 11.02 i? New i? TMI Bank TMI Question #

CI Modified TMI Bank Parent Question #

a Memory or Fundamental Knowledge 2 Comprehension or Analysis a 55.41 .7 'a55.43 E 55.45 .7 A Correct answer. Refer to EP 1202-29.

B Incorrect answer, since this action does not comply with EP 1202-29. Distracter is plausible since this action is related to management of coolant inventory during leak conditions.

C Incorrect answer, since this action does not comply with EP 1202-29. Distracter is plausible since RCS pressure reduction could reduce leakage rate.

D Incorrect answer, since this action does not comply with EP 1202-29. Distracter is plausible since RC Drain Tank pressure will rise due to the leaking PORV.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

1 I Number

-u'-

TMI Unit I Emergency Procedure 1202-29 Title Revision No.

Pressurizer System Failure 59 1 .I Decreasing RCS pressure on Control Room indicators, recorders and computer.

1.2 Decreasing PZR level.

1.3 Decreasing RCS temperature.

1.4 Decreasing PZR temperature.

1.5 Open indications far PORV, Code Safeties2Tailpipe AT A517, A518, A519/Elbow Tap AP/Acoustics Monitor Ann G-1-7, PORV Open [Acoustic])"

1.6 PZR heaters fail to energize at desired setpoint.

BKI -

BK2 - BK3 -

BK4 -

BK5 FULL ON 2135 2135 2135 2120 2105

\//

OFF . 2155 2155 2147 2140 2125 2.0 IMMEDIATE ACTION 2.1 Automatic Action 2.1.1 RC-V-1 closed at 5 2155 psig.

2.1.2 All PZR heaters are "ON" at c 2105.

2.1.3 RPS trips reactor at -e 1900 wig.

2.1.4 ESAS actuation at c 1600 psig.

2.2 Manual Action 2.2.1 VERIFY PORV and Code Safeties are closed.

2.2.2 .' - :& PORV failed open, THEN BLOCK PORV (close RC-V-2).

2.2.3 E press c 2155 psig, THEN VERIFY PZR spray valve (RC-V-1) is shut.

9

SYSIEP# @t~ KA# 2.4.44 Page # 2-16 Tier # -3

'U 4.0 Group #

ROlSRO importance Rating 2.1 -

Knowledge of emergency plan protective action recommendations.(Emergency Procedures/Plan).

identify the ONE statement below that describes the purpose of the Protective Action Recommendations (PARS) associated with the Exelon Standard Emergency Plan.

A. Provides guidance to prevent plant workers from receiving radiation exposures in excess of 10 CFR 20 limits.

B. Provides sheltering and evacuation recommendations for protection of the general public.

C. Determines if potassium iodide tablets should be administered to reduce thyroid dose.

D. Determines if radioactive releases from the site will exceed 10 CFR 20 limits.

EP-AA-111, Emergency Classification and Protective Action Recommendations, Attachment 9, page 19, Rev. 5a.

None.

S-5A.5 Determine the Protective Action Recommendation (PAR) per EP-AA-111 E New 7 TMI Bank TMI Question # #8 1998 SRO exam I

g Modified TMI Bank Parent Question #

Z Memory or Fundamental Knowledge u G Comprehension or Analysis

__ 55.41 c; 55.43 M 55.45 .I1 A Incorrect answer - plausible misconception since this is a valid function of the Radiation Protection Pian rather than the Emergency Plan.

B Correct answer - PAR are recommendations to the state on General Public evackheltering.

C Incorrect answer - plausible distracter since decision to administer KI is a potential action during severe accident conditions.

D incorrect answer - Emergency Plan doesn't assess releases but provides recommended actions based on RAC programs which assess releases.

Rewrote several distractors and the entire stem.

TMI SRO Exam May 2003 Friday, March 28,2003

EP-AA-11 I Revision 5a Page 19 of 19 ATTACHMENT 9 THREE-MILE ISLAND PLANT-BASED PAR FLOWCHART Page Iof I

SYSIEP# KA# 2.2.25 Page # 2-7 Tier # 1 d

ROlSRO Importance Rating 2.5 3.7

- Group # -

2 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits: Small Break LOCA.

Identify the ONE statement below that describes the basis for the Technical Specification requirements for MINIMUM BORATED water inventory in the BWST (Borated Water Storage Tank).

A. Reflood the uncovered core to ensure the reactor coolant remains subcooled.

B. Recirculate the coolant in the core and RB sump, and ensure the core remains 1% delta WK subcritical.

C. Recirculate the coolant in the core and RB sump to limit peak fuel clad temperature to less than 1800°F.

D. Reflood the uncovered core to limit peak fuel clad temperature to less than 18OO0F,and borate the core to at least 10% deltaWK subcritical.

Technical Specification 3.3 bases, page 3-23, Amendment 229.

3 New C TMI Bank TMI Question #

I I Modified TMI Bank Parent Question #

E Memory or Fundamental Knowledge C Comprehension or Analysis 55.41 3 55.45 I

- pj 55.43 .2 A Incorrect answer. First part is incorrect (reflooding applies to Core Flood tanks rather than BWST). Second part is also incorrect.

B Correct answer. Borated volume is needed to ensure enough in mass in the RB Sump to recirculate through the reactor vessel and to ensure the core remains adequately shutdown.

C Incorrect answer. Second part is incorrect (peak clad temperature limit is 22OO0F),and this limit applies to overall acceptance criteria for ECCS operation, rather than to only the BWST.

D Incorrect answer. Both parts are incorrect. Reflood is incorrect (this is a Core Flood Tank function). Peak cladding temperature limit (2200°F) is final ECCS Acceptance Criteria, rather than the basis for minimum BWST volume.

TMI SRO Exam - May 2003 Friday, March 28,2003

3.3 EMERGENCY CORE COOLNG, REACTOR BUILDIXG EMERGENCY COOLING AND REACTOR BULDCNG SPRAY SYSTEMS (Contd.1 3.3.2 Maintenance or testing shall be allowed during reactor operation on any component(s) in the makeup and purification, decay heat, RB emergency cooling water, RB spray, B WST level instrumentation, or cooling water systems which will not remove more than one tmin of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more than 72 consecutive hours. Lf the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the reactor shall be placed in a HOT SHUTWWN condition within six hours." I 3.3.2.1 K the CFT boron concentration is outside of limits, or NaOH tank is outside the Iimits of 3.3.1.3.b or any manual valve in the NaOH tank discharge lines are not locked open, restore the system to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the reactor shall be placed in a HOT SHUTDOWN condition within six hours.

3.3.3 Exceptions to 3.3.2 shdl be as follows:

a. Both CFTs shdl be OPERABLE at aII rimes.

b. Both the motor operated valves associated with the CFTs shalt be fully open at all times.

c. One reactor buiIding cooling fan and associated coohg unit shall be permitted to be out-of-service for seven days.

3.3.4 Prior to initiating maintenance on any of the components, she duplicate (redundant) component shall be verified to be OPERABLE.

In accordance With AmerGen License Change Application dated February 14,2001, and any u requirements in the associated NRC Safety Evaluation, a portion of the Nuclear Service Water System piping between valves NR-V-3 and NR-V-5 may be removed from senice and Nuclear Services River Water flow realigned through a portion of the Secondary Services River Water System piping for up to 14 days. This note is applicable for one time use during TMX Unit 1 Operating Cycle 13.

Bases me requirements of Specification3.3.1 assure that, before the reactor can be made critical, adequate engineered safety features are operable. Two engineered safepards makeup pumps, two decay heat removal pumps and two decay heat removal coolers (along with their respective cooling water systems components) are specified. However. only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both CFTs are requked because a single CFT has insufficient inventory to reflood the core for hot and cofd line breaks (Reference 1).

The operability of the borated water storage tank (BWST)as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA (Reference 2).

The limits on BWST minimum volume and boron concentdon ensure that 1) sufficient water is available within containment to permit recirculation coding flow to the core, and 2) the reactor will remain at least one percent subcritical following a Loss-of-Coolant Accident (LOCA).

The contained water volume limit of 350,000 gallons inclucks an allowance for water no&usable because of tank discharge Iocation and sump recirculation switchover setpoint. The limits on contained water volume, NaOH concentration and baron concentration ensure a pH value of 3-23 Amendment No.+@+4+% )45,U8,22?, 229

Form ES-401-6 Q# Ti0 Page ## 4.1-4 Tier # -1 u ROlSRO Importance Rating 4.1 4.4 Group # 2 Knowledge of the operational implications of the following concepts as they apply to Small Break LOCA: Natural circulation and cooling, including reflux boiling.

z B:

Plant conditions:

Reactor tripped from 100% power due to LOCA.

Automatic ESAS/EFW actuations occurred.

All RCPs are tripped.

One HPI pump is operating.

Core Flood Tanks are now emptying, "floating on the RCS."

RCS is at 200 psig, with core exit thermocouple temperature at 382°F.

Pressurizer level indication = 20 inches.

OTSG pressures are both 150 psig, and steady.

OTSGs are at required levels, with 0 gpm EFW flow to each OTSG.

TBVs are closed.

RCS cooldown has STOPPED, and heatup rate is now +20"F/hr.

Based on these conditions, identify the ONE phrase below that describes current status of boiler-condenser core cooling.

A. Stopped due to low EFW flows.

B. Interrupted by introduction of the cold Core Flood Tank water C. Occurring because OTSG levels have been raised sufficiently.

W D. Occurring because the OTSG temperatures are lower than core exit temperature.

Emergency Operating Procedures Technical Bases Document Volume 3, pages 1ll.B-I3 and 111.B.14, Rev. 9.

None.

V.E.10.12

@ New E TMI Bank TMI Question #

Z Modified TMI Bank Parent Question #

C Memory or Fundamental Knowledge E Comprehension or Analysis E 55.41 .8/.10 E 55.43 2l 55.45 .3 A Incorrect answer. Plausible since there is a minimum EFW flow requirement to induce BCC until levels are raised to 75-85%.

B Correct answer. Introductionof cold Core Flood water into the core area can temporarily lower local coolant temperatures enough to stop (pool boiling) phase change at the core outlet. This would interrupt BCC, and the RCS would begin to heatup.

C Incorrect answer. Plausible since this a requirement to produce BCC.

D Incorrect answer. Plausible since this a requirement to produce BCC.

L-,

TMI SRO Exam - May 2003 Friday, March 28,2003

T E C H N O L O G I E S NUMBER TECHNICAL DOCUMENT 73-11524 14-09 Boiler condenser cooling can be lost due to a loss of adequate condensing surface in the SG tubes. This can occur due to a decrease in effective thermal center on the secondary side due to a low liquid level and insufficient EFW flow or if HPI is being injected, it may be due to an increase in RC level inside the tubes. Boiler condenser cooling may be cyclic in nature because, as the RC steam is condensed, RCS pressure will decrease toward SG pressure causing leak flow to decrease and HPI flow to increase. This may allow the RCS to refill which will reduce the available condensing surface inside the SG tubes. If the heat transfer to the SGs is lost, RC temperature and pressure will begin to increase causing the leak flow to increase and HPI flow to decrease. This will, in turn, cause a net loss in RC inventory which will restore the condensing surface. This self-regulating process may occur with only fluctuations in the amount of heat transfer or it may result in periodic losses of heat transfer. In either case, boiler condenser cooling should continue without hrther operator action required to restore heat transfer.

However, the operator may take the actions discussed in Chapter II1.C to aid restoration of heat transfer. Boiler condenser cooling should not be cyclic if PI is not available.

The primary system will not refill and the condensing surface inside the SG tubes will be "4

maintained as long as there is no net addition to RC inventory.

Boiler condenser cooling may also be partially lost due to a continued net loss of RC inventory while boiler condenser cooling exists. Boiler condenser cooling provides or aids core cooling in two ways. First, the condensation of the RC steam in the tube region removes heat from the RCS and, by reducing RC pressure, allows higher HPI flow which also provides cooling. Second, by maintaining a high enough RC liquid level in the SG tubes, coolant is forced over the RCP internal lip where it can flow into the core region.

If the RC inventory continues to decrease even with boiler condenser cooling, the level required to force coolant over the RCP internal lip will be lost. How-ever] the condensing surface inside the SG tubes will continue to grow and thus the heat removal obtained by condensing RC steam will continue. This case is essentially a combination of saturated cooldown with the SGs (3.6) and cooldown on break/HPI flow (3.7).

During a saturated cooldown, using either the SGs or break/" flow or a combination of both, it is possible for the RC pressure to appear to stabilize when the CFTs begin to discharge. If the net loss of RC inventory is relatively small, the CFTs will RCS should continue to cool and depressurize but possibly at a egins to reheat, the operator should attempt to increase heat removal by the If SG heat removal cannot be increased sufficiently, and HPI flow exists, the PORV can be used to decrease RCS pressure, increase HPI flow, and thus increase cooling due to HPI.

7 3;; 112000 r-- V01.3, 1IT.B -13

1 NUMBER I

'ECHNICAL DOCUMENT 74-1 152414-09

.6 Saturated Cooldown Using the SG(s)

Cooldown of the RCS, during saturated conditions, is accomplished in essentially the same way as a normal subcooled natural circulation cooldown; however, there are some major differences:

HPI Flow Required Full PI flow from two "1 pumps, if available, must be maintained while the RCS is saturated. This PI flow can also provide substantial core cooling, thus the operator does not have total control of the RCS cooldown rate with the SGs. If P I is not available, then the cooldown will be a hnction of heat transfer to the SGs and the energy that flows out of the break.

5 Unstable Cooldown A saturated cooldown using the SGs is inherently unstable. In order to maintain SG heat removal with saturated natural circulation, a mass and energy balance must exist. If the RCS is initially in saturated natural circulation and core decay heat exceeds the heat removal rate, a mass and energy balance does not exist. In this case steam formation, due to core boiling, can block natural circulation flow with a resultant loss of primary to

'4 secondary heat transfer. This may eventually result in boiler condenser cooling and restoration of primary to secondary heat transfer. RCS refill could result if core boiling is suppressed by PI cooling through the break. &fRCS refill does occur, and prevents boiler condenser cooling, then primary to secor%ary heat transfer will not be restored until single-phase natural circulation develops following RCS refill.

If the RCS is in saturated natural circulation or boiler condenser cooling with a mass and energy balance, a subsequent decrease in core decay heat or increase in SG heat removal will offset the balance. If the RCS is in natural circulation, this will result in RCS cooling and depressurization, which will decrease the break flow while increasing the HPI flow, and should allow eventual restoration of SCM while maintaining primary to secondary heat transfer. in boiler condenser cooling, this will result in refill and a possible interrupti sfer. This could then result in either cyclic boiler condenser cooling or continued refill and restoration of natural circulation. st Normal Cooldown Rates Do Not Apply When the RCS is saturated normal Technical Specification cooldown rates and the special case RCS cooldown rate for RV head void concerns do not apply.

DATE 3 i 3 1/2000 V01.3, 1II.B -14 PAGE

Form ES401-6 SYS/EP# Q l J KA# 2.1.33 Page # Tier #

\/

ROfSRO Importance Rating 3.4 -

4.0 Group # -1 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications: Large Break LOCA.

izl -

a Plant conditions:

- Reactor trip due to loss of offsite power (LOOP).

- ESAS actuation due to LOCA.

- EG-Y-1A failed to start.

- B Train HPI and LPI systems actuated, and continue to operate.

- Both Core Flood tank levels are at 2 feet.

- RCS pressure = 260 psig.

- Pressurizer level is offscale low.

- RCS temperature is 180°F.

Based on this event, identify the ONE condition below that requires NRC to be notified regarding VIOLATION of a Tech Spec LCO (Limiting Condition for Operation).

A. EG-Y-1A failure.

B. Pressurizer empty.

C. Low Core Flood tank levels.

D. RCS temperature at 180°F with HPI operating.

OP-TM-EOP-006, LOCA Cooldown, Step 3.23, Page 7, Rev. 2.

Technical Specification 3.1.12, pages 3-18d, 3-18e, 3.18f, Amendment 234.

TS 3.1.12, pages 3-18d, 3-18e, 3.18f, Amendment 234.

@ New TMlBank TMI Question ##

3 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge Comprehension or Analysis g 55.43 .I 55.45 .3 A incorrect answer - EDGs are Tech Spec, however conditions are not met to require them B incorrect answer - PZR is Tech Spec, however conditions are not met to require it C incorrect answer - Core Flood Tanks are Tech Spec, however conditions are not met to require them, and they have already performed intended function D correct answer - this is an LCO violation, however EOPs take priority. IAW EOP and TS, notify NRC.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

t 3.1.12 Pressurizer Power Operated Relief Valve (PORV),Block Valve, and Low Temperature Overpressure Protection &TOP)

Applies to the settings, and conditions for isolation of the foRV.

Objective To prevent the possibility of inadvertently overpressurizingor depressurizing the Reactor Coolant System.

SDecification 3.1.12.1 LTOP Protection Ifthe reactor vessel head is installed and indicated RCS temperature is 5 329*F, High Pressure Injection Pump breakers shall not be racked in unless:

a. MU-V16A/B/C/D are closed with their breakers open, and Mu-V217 is closed, and

b. Pressurizer level is maintained ,< 100 inches. If pressurizer level is > 100 inches,restore level to 5 100 inches within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3.1.12.2 The FORV settings shall be as follows:

a. Low TemperatureOverpressureProtection Setpoht

1. W h e n indicated RCS temperature is I329°F.the LTOP system shall be operable as kfmed in Specification3.1.12.1 and

2. The PORV will have a maximum lift setpoint of 552 psig.

With the PORV setpoint above the maximum value, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1. restore the setpoint below the maximum value, or

2. verify pressurizer level is S 100 inches indicated and satisfy the requirements of Technical Specification 3.1.12.3 allowing the PORV to be taken out of service.

b. Unless the Low Temperature Overpressure Protection Setpoint is in effect, the PORV lift setpoint will be a minimum of 2425 psig.

With the PORV setpoint below the minimum value, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1. restore the setpoint above the minimum value, or

2. close the associated block valve, or

3. close the PORV, and remove power fromPORV

4. otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

- Amendment No.5 G W - 4 9 , 157, %,2%

3-18d

3.1.12.3 When the indicated RCS temperature is below 329°F the POW shall not be taken out of service, nor shall it be isolated from the system unless one of the following i s in effect:

a. High Pressure hjection Pump breakers are racked out.

b- MU-V16A/B/C/D are closed with their breakers open, and MU-V217is closed.

C. Head of the Reactor Vessel is removed.

3.1.12.4 The PORV Block Valve shall be OPERABLE during HOT STANDBY, STARTUP, and POWER OPERATION:

a. With the PORV Block Valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

1. restore the PORV Block Valve to OPERABLE status or

2. close the PORV (verify closed) and remove power from the PORV

3. otherwise,be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the PORV block valve inoperable, restore the inoperable valve to OPERABLE status prior to stamp from the next COLD SHUTDOWN unless the COLD SHUTDOWN occurs W

within 90 EffectiveFull Power Days (EFPD)of the end of the fuel cycle. If a COLD SHUTDOWN occurs within this 90 day period, restore the hoperable valve to OPERAl3LE status prior to startup for the next fuel cycle.

Bases If the PORV is removed from service while the RCS is below 329"F, sufficient measures are incorporated to prevent severe overpressurization by either eliminating the high pressure sources or flowpaths or assuring that the RCS is open to atmosphere.

The PORV setpoints are specified with tolerances assumed in the bases for Technical Specification 3.1.2. Above 329'F. the PORV setpoint has been chosen to limit the potential for inadvertent discharge or cycling of the PORV.

Other action such as removing the power to &e PORV has the same effect as raising the setpoint which also satisfies this requirement. There is no upper limit on this setpoint as the Pressurizer Safety Valves (T.S. 3.1.1.3) provide the required overpressure relief.

Below 329'F, the PORV setpoint is reduced to provide the required low temperature overpressure reIief when high pressure sources and flowpaths me in service. There is no lower limit on the pressure actuation specified as lower setpoints also provide this same protection.

3-18e

, 1% ,234

-VI Amendment No. -6?

In both cases, the setting is specified to reflect the nominal value which allows for normal variations in the temperature setpoint while maintaining the tolerances assumed in the bases for T.S. 3.1.2. Either pressure actuation setpoint is acceptable within the temperature range between 313°F and 329°F.

With RCS temperaturesless than 329OF and the makeup pumps running, the high pressure injection valves are dosed and pressurizer level is maintained less than 100 inches to allow time for action to prevent severe overpressurizationin the event of any single failure.

The PORV block valve is required to be OPERABLE during tbe HOT STANDBY, STARTUP, and POWER OPERATION in order to provide isolation of the PORV discharge line t o positively control potential RCS depressurization.

For protection from severe overpressurization during HPI testing, refer to Section 4.5.2.1.c.

3- 18f Amendment No. 4-86 ,234

\d

0P-TM-EOP-006 Revision 2 Page 7 of 13 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 3.22 IAAT RCS 700 psig and SCM 2 25°F and RCS pressure is being controlled, then CLOSE the following breakers:

-CF-V-1A (1C-ESV-MCC, Unit 3C

-CF-V-I 8 (1C-ESV-MCC, Unit 4C)

CLOSE CF-V-1 and CF-V-1B. ....-

3.23 NOTIFY SM. IF HPI is required for RCS inventory control when RCS temperature approaches 329F, then Continue use of HPI for inventory control.

T I 3.24 Invoke 10CFR50.54X.

IAAT RCS 329F, then MONITOR RC,

. . - - I pressure and SWITCH the PORV contro to AUTO (to enable 552 psig PORV W

setpoint).

3.25 INITIATE OP-TM-EOP-001, VSSV and Follow-up Actions.

3.26 ENSURE the following valves are CLOSED:

- MS-V-1A

- MS-V-1B

- MS-V-1C

- MS-V-1D

Page # 4.1-7 Tier # -1 W

ROlSRO Importance Rating 4.1 4.4

- Group # -1 Knowledge of the operational implications of the following concepts as they apply to Large Break LOCA: Natural circulation and cooling, including reflux boiling.

From the list below, identify the ONE statement that describes operational implications of Primary-to-Secondary Heat Transfer during a large break LOCA, characterized by establishment of LPI flow into the reactor core.

A. Boiler-condenser cooling can be enhanced by raising OTSG levels to the upper tube sheet.

B. Single phase natural circulation cooling can be established after LPI flow refills the reactor vessel.

C. Boiler-condenser cooling is NOT required to augment ECCS cooling if 1250 gpm LPI flow is established to the reactor core.

D. Single phase natural circulation will NOT be possible until reactor vessel head temperature is lower than pressurizer temperature.

OP-TM-EOP-006, LOCA Cooldown, Steps 3.13 (page 3) and 3.26 (page 7), Rev. 2.

Emergency Operating Procedure Technical Bases Document 74-1152414-09, Volume 2, Chapter IV. Section A, LOCA Cooldown, pages 9 and 28, Rev. 9.

9 New TMlBank TMI Question ##

-- Modified TMI Bank Parent Question #

b Memory or Fundamental Knowledge 9

- Comprehension or Analysis 9 55.41 .8/.10 __- 55.43 g 55.45 .3 A Incorrect answer. This action would actually block B-C cooling, since it can eliminate all condensing action in the tubes.

B Incorrect answer. Even after LPI flow refills the vessel, coolant flow out the break will prevent establishment of natural circulation flow to the OTSGs.

C Correct answer. Specifyng LPI flow to the core, this is adequate flow to meet ECCS cooling requirements, and PSHT cooling augmentation is not required.

D Incorrect answer. Distracter is plausible since a head bubble can exist under these conditions.

TMI SRO Exam - May 2003 Friday, March 28,2003

T E C H N O L O G I E S NUMBER 74-1 153,414-09 23.0 IF AT ANY TIME LPI FLOW > [minimum flow rate], THEN THE FOLLOWING MAY BE PERFORMED:

23.1 SG cooling terminated. ~

23.2 RCPs secured.

Indicators and Controls Indicators: - LPI flow Controls: - TBV/ADV controls

- RCP motor controls Puruose of Step The purpose of this step is to allow termination of SG heat transfer and forced flow when no lonser required. .

Bases Once the RCS pressure uced suMiciently to allow LPI flo specific minimum flow oling may no longer be required jperation is also no longer required, and securing RCPs will reduce the heat load imposed on the

-- LPI system. Securing RCPs is allowed under these conditions even if they were required to be running due to failure to trip on loss of SCM.

If SG cooling was used to lower RCS pressure to establish LPI flow > [minimum flow rate], then it is not recoinmended to secure SG heat transfer if it is needed to maintain RCS pressure < LPI shutoff. If SG cooling is terminated and the RCS subsequently begins to reheat and repressurize, then SG cooling should be re-established per step 15.0.

The plant-specific value of [minimum flow rate] is a limiting value and should be error-corrected.

Sequence There is no specific sequence requirement.

TBD Volume 3 References III.B.3.8, IV.A.2.2, IV.A.2.4 and V.1.0 7 3 / 3 1/2000 PAGE V01.2, 1V.A-28

I NUMBER I TECHNICAL DOCUMENT 74-1 152414-09 7.0 IF AT ANY TIME RCS PRESSURE < LPI OPERATIONAL PRESSURE AND LPI FLOW EXISTS, THEN GO TO STEP 20.0.

Indicators and Controls Indicators: - RCS pressure

- LPI flow Controls: - NIA Purpose of Step The purpose of this step is to bypass unnecessary steps if LPI flow is already established.

Bases If RCS pressure is low enough for LPI operation, and LP'I flow is established, then steps to control cooldown rate with the SGs, PORV operation, etc., are no longer necessary, ,

Sequence There is no specific sequence requirement.

TBD Volume 3 References None. This is only a routing step that is needed because of the manner in which the GEOG is

-1 structured.

7 313 1/2000 I PAGE V01.2, 1V.A-9

0P-TM-EOP-006 Revision 2 Page 3 of 13 I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

__ 3.7 ENSURE performance of an alarn review.

- 3.8 REQUEST SM evaluate

- 3.9 Emergency Action Levels (EALs).

VERIFY Emergency Diesel

- If unloaded, then INITIATE

- _I Generators are in ES standby or OP-TM-86 1-901 and loaded on bus. OP-TM-861-902, and place Diesel Generators in ES standby, one at a time.

- 3.10 INITIATE Guide 20, PRIOR to Transfer to RB Sump.

- 3.1 1 INITIATE Rule 5, Emergency Boration. ~

3.12 IAAT BWST level 15 ft, or RB flood level > 54,then BRIEF, anc

.- INITIATE Guide 21, Transfer to RB Sump Re-circulation.

and LPI flow > 1250 G then GO TO Step 3.26.

n3.14 IAAT primary-to-secondary heat transfer does not exist and core cooldown rate < 40F/hr, then PERFORM Step 3.32. - -- - - - - .- . ....... .. .. .... .. .. . .... - .... .. ... .... .. .... . . . - .. .. ... . .. . ..... .. __ . . .. I

OP-TM-EOP-006 Revision 2 Page 7 of 13 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 3.22 IAAT RCS 700 psig and SCM 2 25°F and RCS pressure is being controlled, then CLOSE the following breakers:

__ CF-V-1A ( I C-ESV-MCC, Unit 3C

__ CF-V-1B (1C-ESV-MCC, Unit 4C)

-CLOSE CF-V-1 and CF-V-1B.

3.23 NOTIFY SM. IF HPI is required for RCS inventory control when RCS temperature approaches 329F, then Continue use of HPI for inventory contro I.

Invoke 10CFR50.54X.

. . . . . . . . . . . . . . . . . . . . . I 3.24 IAAT RCS < 329F, then MONITOR RCS pressure and SWITCH the PORV control to AUTO (to enable 552 psig PORV setpoint). ..... . . . . . . . .

3.25 INITIATE OP-TM-EOP-001, VSSV and Follow-up Actions.

3.26 ing v re MS-V-I A ~

- MS-V-1 B

- MS-V- 1 c

- . - .- . .- ..- .. D S-V-I -.. - . . . .

SYSIEP# 015 KA# 2.4.6 Page # 2-11 Tier # 1

-d ROlSRO Importance Rating 3.1 -4.0 Group # -1 Knowledge symptom based EOP mitigation strategies: Reactor Coolant Pump (RCP)

Malfunctions Sequence of events:

TIME EVENT 1800 - Reactor trip from 100% power.

1810 - RCS SCM (Subcooled Margin) reduces to less than 25°F.

1813 - Procedure directs operators to run all available RCPs.

1814 - All four RCPs are running.

1900 - RC-P-1A Seal # I failure is diagnosed.

- SCM is at 4°F.

Based on these conditions, identify the ONE statement below that describes required actions and procedure to be implemented.

A. Trip RC-P-1A in accordance with 1203-16, Reactor Coolant Pump and Motor Malfunction.

B. Trip RC-P-1A in accordance with OP-TM-EOP-010, Abnormal Transients Rules, Guides and Graphs.

C. Continue to operate RC-P-1A in accordance with 1203-16, Reactor Coolant Pump and Motor Malfunction.

D. Continue to operate RC-P-1A in accordance with OP-TM-EOP-010, Abnormal Transients Rules, Guides and Graphs.

OS-24, Conduct of Operations During Emergency and Abnormal Events, section 4.1.6.G, Page 8, Rev. 7.

New 13 TMIBank TMI Question #

El Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge 3 Comprehension or Analysis 3 55.41 . I O a 55.43 .5 I 55.45 . I 3 A Incorrect answer - plausible distractor as this is the EP action, however EOP takes priority in conflicting guidance.

B Incorrect answer - wrong action, correct reference.

C Incorrect answer - correct action, wrong reference.

D Correct answer.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 Operations Department

-u Administrative Procedure OS-24 Title Revision No.

Conduct of Operations During Abnormal and Emergency Events 7 4.1.6 Rules and Guides A. The Loss of Subcooling Margin Rule and the Excessive Heat Transfer Rule, when identified, are performed immediately, even if Reactor Trip Immediate Manual Actions are in progress. See section 4.3.5 Reactor Trip Actions.

6. The following sequence is used when Rule based action is required:

1. Announce the applicable Rule.

2. Pull the applicable Rule card.

3. The Unit Supervisor provides concurrence.

4. Perform the Rule based actions.

5. Report completion of Rule based action and r e x n card I holder.

C. Reactor Operators consider the priority of Rules in relationship to actions or steps in progress by the Unit Supervisor. If the Unit Supervisor is in the process of Immediate Manual Action verification, or other higher priority action, the Reactor Operator delays implementation of the lower priority action until the Unit Supervisor is able to provide concurrence for the action.

D. Rules are numbered according to priority. If mulfiple Rule based actions are required, the highest priority Rule is performed first. If resources allow, multiple Rules may be performed in parallel by separate personnel.

E. Rules and Guides are applicable whenever any EOP is entered. Rules and Guides may be implemented without specific direction from an EOP if the condition or intent is met.

F. After action has been initiated in accordance with a Rule or Guide, further US concurrence is required prior to any substantive change in the action being taken. Otherwise, t h e operator will continue to take action in accordance with the Rule or Guide.

G. When direction from Rules, Guides or procedures conflict, the following order of precedence should be applied: (1) Rules (including the order of priority within the Rules) (2) EOP steps (3) Guides and (4) other procedure requirements.

\.-

a

Page # 4.2-11 Tier # -1 Nu ROlSRO Importance Rating 3.7 -3.7 Group # -1 Ability to determine and interpret the following as they apply to Reactor Coolant Pump (RCP)

Malfunctions (Loss of RC Flow): When to secure RCPs on loss of cooling or seal injection Plant conditions:

- Reactor power loo%, with ICs in full automatic.

- Intermediate Closed Cooling Pump is IC-P-1B 00s for motor replacement.

- Total RCP seal injection flow is 18 gpm, controlled locally in Makeup Valve Alley.

Event:

- IC-P-1A trips.

- Plant remains steady at 100% power.

Based on these conditions, identify the ONE statement below that identifies the applicable procedure and required action@)to be implemented.

A. Initiate plant shutdown in accordance with OP 1102-4, Power Operations.

B. Increase RCP seal injection flow in accordance with OP 1104-2, Makeup and Purification System C. Attempt to restart IC-P-1A one time in accordance with OP-AA-103-103, Operation of Plant Equipment.

D. Trip the reactor AND then trip all 4 RCPs in accordance with OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel.

OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel, Rev. 0, Section 11 4.6.4.7, Page 6.

d New - TMl Bank TMI Question #

- Modified TMI Bank Parent Question #

- Memory or Fundamental Knowledge 7

- Comprehension or Analysis 3 G 5 . 4 3 .5 -9_55.45 . I 3 A Incorrect answer - Reactor/RCPs are required to be tripped due to failure of automatic RCP trip interlock.

B Incorrect answer - plausible, since seal injection flow is low, but reactor/RCPs are required to be tripped due to failure of the automatic RCP trip interlock.

C Incorrect answer - plausible, since this is historical guidance, but reactorlRCPs are required to be tripped due to failure of the automatic RCP trip interlock. OP-AA-103-103, Operation of Plant Equipment does not address this issue.

D Correct answer.

None.

u TMI SRO Exam - May 2003 Friday, March 28,2003

OP-AA-101-111 Revision 0 Page 6 of 8 4.6.1. REPORT to the Unit Supervisor.

4.6.2. OPERATE the plant in accordance with approved procedures, and within the Limiting Conditions for Operation of the Technical Specifications to ensure the reactor is operated in a safe, conservative, and efficient manner at all times.

NOTE: The ROs immediate actions to stabilize the plant during transient conditions take priority over verbalization to the Unit Supervisor. If possible, verbalization should be accomplished to inform the Unit Supervisor of actions being taken.

1. During transient conditions, the RO may perform immediate operator actions of abnormal procedures from memory, while verbalizing actions being taken to the Unit Supervisor.

2. Subsequent actions taken during transient conditions will be based on direction of the Unit Supervisor per the applicable procedure(s).

4.6.3. MAINTAIN an active Reactor Operators license.

4.6.4. One RO on each unit SHALL be designated the Unit RO and SHALL be at the L controls (as defined by each station).

1. ENSURE applicable Technical Specification time clocks are entered and exited and associated action requirement completed as appropriate based on the scope of the work.

2. MONITOR the reactor and ENSURE reactor operation remains within established bands.

3. MONITOR all assigned control room panels, and NOTIFY the Unit Supervisor regarding unusual or unexpected conditions.

4. MAINTAIN cognizance of the activities and work impacting the unit, and the work of the assist RO(s) assigned to the unit.

5. COORDINATE and/or PERFORM necessary reactivity changes on the unit during the shift.

6. SHUTDOWN the reactor when the RO determines the safety of the reactor is in jeopardy or when operating parameters exceed any of the reactor protection circuit setpoints and automatic shutdown does not occur.

7. Manually INITIATE safety systems automatic actions when operating parameters exceed the systems automatic initiation setpoints and automatic initiation does not occur.

Form ES-401-6 SYSIEP# 022 KA# AK1.02 Page # 4.2-12 Tier #

- I '

RO/SRO Importance Rating 2.7 -3.1 Group # -

2 Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship of charging flow to pressure differential between charging and RCS Sequence of events:

- Manual reactor trip due to excessive RCS leakage.

- ES actuation prevented loss of RCS subcooled margin (SCM).

- MU-V-18 (Make up isolation) is closed.

- HPI flow is now manually throttled to 100 gpm.

- Pressurizer level is steady at 120 inches.

- RCS temperature is steady at 535°F.

- RCS pressure is now 1600 psig.

- TBVs (turbine bypass valves) are 5% open.

Based on these conditions, identify the ONE statement below that describes operational impact on HPI flow, if the TBVs were to fail closed.

A. HPI flow REDUCES due to automatic closure of MU-V-17 (Pressurizer Level Control Valve).

B. HPI flow RISES to due reduction in coolant density.

C. HPI flow REDUCES due to rising RCS pressure.

D. HPI flow RISES due to rising RCS leakrate.

\-/'

OPM Section N-03, Mechanical Fundamentals, Figure E-26, Makeup Pump Curves, page 114, Rev. 6.

'Q New - TMIBank TMI Question #

- .-Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge v

- Comprehension or Analysis 2 55.41 .8/.10 __ 55.43 v 55.45 .3 A Incorrect answer. Although MU-V-17 closes, with HPI initiated, MU-V-18 is closed, blocking normal makeup flowpath B Incorrect answer. SCM reduction is result of heatup and pressurization, so HPI flow will be reduced.

C Correct answer. RCS pressure increase will reduce HPI flow as long as cavitating venturies are not controlling flow.

D Incorrect answer. Distracter is plausible, since RCS pressure rise will raise leakrate. HPI flow will be reduced as RCS pressure rises.

None.

--J TMI SRO Exam May 2003 Friday, March 28,2003

I s

E

\

Form ES-401-6 b

SYS/EP# 026 KA# AK3.03 Page # 4.2-18 Tier # -1 ROlSRO Importance Rating 4.0 -

4.2 Group # 1 Knowledge of the reasons for the following responses as they apply to Loss of Component Cooling Water (CCW): Guidance actions contained in EOP for Loss of CCW.

Identify the ONE statement below that describes the basis for procedural guidance to close NS-V-32 (cooling inlet isolation to evaporators, seal return coolers, and waste gas compressors) during implementation of 1203-20, Nuclear Services Closed Cooling System Failure.

A. Isolate a NSCC leak at the Spent Fuel Coolers.

B. Isolate a NSCC leak at the Seal Return Coolers.

C. Reduce NSCC system heat load when ESAS is actuated.

D. Reduce NSCC system heat load when only one NSCC pump is available.

AbP 1203-20, Nuclear Services Closed Cooling System failure, Rev. 20 OPM Section 8-1 1, Nuclear Services Closed Cooling System, page 8, Rev. 8.

v New - TMIBank TMI Question #

._ Modified TMI Bank Parent Question #

9 Memory or Fundamental Knowledge L

Comprehension or Analysis 9 55.41 .5/.10 - 55.43 v 55.45 .6/.13 A Incorrect answer. Valve does not supply Spent Fuel coolers, and closing this valve will not stop any leak.

B Incorrect answer. Although valve supplies seal return coolers, closing it will not stop a leak - the return side still feeds the leak.

C Incorrect answer. Distracter is plausible since this valve actually had an automatic (ES) closure interlock in the past.

D Correct answer. This valve blocks cooling supply to several non ESAS components to reduce non-essential heat load. The purpose of closing this valve is not to islate a leak.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

REVISION 8 u 7.6 Nuclear Services Surge Tank Pressure Lo Annunciator ahmi F-2-8. Alarm setpoint is 30 psig.

7.7 Too Many Nuclear Services Coolers Out-of-Service Annunciator alarm F-3-7. This alarm is actuated if 3 Nuclear Services Cooler river water d e t or outlet valves are closed. The valve positions are determined by limit switches on the valve.

7.8 Nuclear Services Surge Tank LowHigh Level Annunciator alann on F-1-8. The low level alarm setpoint in 1.6 ft. the high level alann setpoint is 7 ft.

8.0 CONTROLS The controls provided the operator for the NSCCW system are located on console center, console right and on Panel Center Bght.

. The controls on the console center and right include pump start-stop switches. On console center is the pump start-stop switch for NS-P-IA and the start-stop switch for NS-P-1B when it is powered from the P bus. On console right is the start-stop switch for NS-P-1C and NS-P-1B when it is powered from the S bus. Above each pump control switch are 3 lights green indicating the pump breaker is open, amber indicating a breaker overload or breaker position does not agree with the control switch flag and a red light indicating the breaker is closed.

L_r-Also a pump amp indicator is available above each control switch.

On console center are the controls for valves NS-V-4, NS-V-15, and NS-V-35. These are not jog control pushbuttons that is if close is pushed the valve will travel completely closed. Also on console center is a pushbutton control for NS-V-32. When NS-V-32 is closed the nuclear services closed cooling supply to various NON - ES components is secured.

The components that are serviced by the NON - ES isolation valve NS-V-32 are:

a. RCP Seal Return Coolers A, B

b. Waste Gas Compressor A, B C. RC Waste Evaporator Condenser

d. RC Waste evaporator Distillate Cooler

e. Misc. Waste Evaporator Condenser

f. Misc. Waste Evaporator Distillate Cooler Controls are also provided on PCR for NS-V-52A, B, C and NS-V-SA, B, C these valves are the cooling water inlet and outlet valves to the Reactor Building fan (AH-E-lA, B, C) motor coolers. These valves are normally open during Plant Shutdowi and Power Ops.

Also on PCR are special controls that allow testing of NS-V-4, 15, 3 5 . Since these are return and supply valves for cooling water to the RCPs it is necessary to allow testing without fully closing the valves. When the switch is in the normal position the valves will operate normally. When the switch is in the TEST position the valve will only close 10%. This allows for functional testing to be performed without interrupting the cooling water flow

-d for the RCPs.

8

TMI Unit 1 u

Abnormal Procedure 1203-20 Title Revision No.

Nuclear Services Closed Coolina Svstem Failure 2Q 1.D SYMPTOMS 1.1 "N.S. Surge Tnk. Press La" alarm, 30 psig decreasing (PS-188) (F-2-8).

1.2 "N.S.Clg Pmp. Disch. Press Lo" alarm, 60 psig decreasing (PS-169), (F-1-7).

3.3 N.S. surge tank: Comp. Pt. A0453 Lo Level alarm 30" decreasing Lo-2 Level alarm 18" decreasing

. Comp.Pt. A0454 Lo Level alarm IS" decreasing NOTE If low level alarm at 1.6 feet is received coincident with HPI the following valves will dose: NS-V-4. NS-V-15. NS-V-35.

"NS Surge Tank Level HilLo"alarm at 1.6 ft. decreasing (F-1-8). (LS-800or LS-801) 1.4 "N.S. Heat Exchangers Outlet Temp. Hi" alarm at ?OO"F increasing p - 2 4 1 ) (F-2-7).

1.5 "480V ES Motor Trip" alarm, N S pump breaker tripped (6-1-5).

IMMEDIATE ACTION 2.1. Automatic Actions 2.1 .I Standby NSCC Pump starts on breaker trip of the operating NSCC pump.

2.2 Manual Actions 3.0 NOTE Reactor cannot be made criticat unless two NSCCW pumps are operable (T.S.3.3.1.4). A 3.1 VERIFY surge tank level greater than Lo Level alarm (1.6 ft).

3.2 -IF surge tank is less than 1$6ft., THEN REFILL surge tank (comp. pt- A0453/A0454).

3.3 -

IF N.S. Surge Tank Low Level is coincident with HPI actuation (Line Break Isolation) M E N DO NOT reset without Shift Manager or Control Room Supervisor permission.

2

Number TMI - Unit 1 Abnormal Procedure 1203-20

.L/" Title Revision No.

Nuclear Services Closed Cooling System Failure 20 ATTEMPT to isolate system leakage by isolating portions of the system while maintaining flow to the I RCP motor coolers.

-IF only 1 NSCCW pump can be made available, THEN PERFORM the following:

3.5.1 REDUCE to 2 NS heat exchangers in service.

3.5.2 SHUTDOWMthe MWlRC evaporators.

/-----------+-

M yCLOSE NS-V-32.

L-.---- - ---

3.5.4 MONITOR system heat loads.

3.5.5 E NS heat exchanger outlet temperatures approach 95"F, THEN REDUCE non essential loads.

-IF non -ES selected pump cannot be started, THEN VERIFY/RESET 27/86 tockout relays.

IF either of the following conditions occur:

0 NO NSCCW Pumps can be started e NSCCW CANNOT maintain the RCP motor stator and bearing temperatures below their alarm point THEN PERFORM the following:

3.7.1 PROCEED with a normal plant shutdown in accordance with 1102-10, Plant Shutdown.

3.7.2 E conditions warrant, THEN TRIP the Reactor Coolant Pumps AND GO to 1210-1, Reactor Trip.

NOTE The maximum allowable amount of time without water flow to.motor air cooler is 10 minutes.

IF any of the following conditions occur:

a RCP motor stator temperature exceeds l5O0C c RCP motor radial bearing temperatures exceed 185°F OR I

3

. . ... ... _ 1

Number TMI Unit I Abnormal Procedure 1203-20

\.-' Title Revision No.

Nuclear Services Closed'Cooling System Failure 20 RCP motor thrust bearing exceeds 195OF THEN REFER TO 1203-16, Reactor Coolant Pump and Motor Malfunction.

3.9 Upon receipt of a low cooling water flow alarm to Make-up Pump "B" (L2834). PERFORM the following:

3.9.1 START the "A" Decay Heat River Pump (DR-P-1A) 3.9.2 START the "A" Decay Heat Closed Pump (DC-P-1A) 3.9.3 START the "A" Make-up Pump (MU-P-1A) 3.9.4 STOP the "B" Make-up Pump (MU-P-1 B)

- 3.10 START additional N.S. pumps AND VALVE-IN additional coolers as needed to maintain adequate cooling to all components in service.

3.11 -IF conditions warrant, THEN BACKWASH NUC Service coolers per 1104-30, Nuclear River Water.

4

Form ES-401-6 SYSiEP# 026 KA# AA1.07 Page # 4.2-18 Tier #

L.-/

ROiSRO Importance Rating 2.9 3.0

- Group # 1 Ability to operate and/or monitor the following as they apply to Loss of Component Cooling Water (CCW): Flow rates to the components and systems that are serviced by the CCWS; interactions among the components.

~

@ I C

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Intermediate Closed Cooling Pump IC-P-1A is operating.

- ESAS testing is in progress on IC-V-6 (CRDM cooling supply isolation).

- IC-V-6 is inadvertently CLOSED to 0% position.

Based on these conditions, identify the ONE statement below that describes operator action that will be required if IC-V-6 CANNOT be re-opened.

A. Open IC-V-74 (minimum flow recirculation valve).

B. Isolate RCS letdown flow.

C. Trip the reactor.

D. Start IC-P-1B.

EP 1202-8 CRD Equipment Failures, High Stator Temperature, page 28, Rev. 53.

OP 1105-9, Control Rod Drive System, Limit & Precaution 2.1 .a, page 5, Rev. 61.

u 7 New 1TMIBank TMI Question #

- Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge

- Comprehension or Analysis 7 55.41 .7 I 55.43 g 55.45 51.6 A Plausible distractor - for candidate who thinks recirc intlk is based on flowllC-V-6 position (need IC-V-4 closed also to make up intlk, or IC-V-2 or 3 alone)

B Plausible distractor - there is an interlock to isolate letdown, but lack of IC CRDM FLOW would NOT allow the high temp to be seen (must have some flow). Also, letdown isolation is not necessary in this case, as heat load is not a factor.

C Correct answer - CRD stator temperatures will rise rapidly, requiring the operators to trip the reactor in order to de-energize the motors.

D Incorrect answer - low system flow conditions does not exist, only crdm flow is blocked, system flow will remain >900gpm, well above intlk or any need for further pumps None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TM I NUCLfAR Emergency Procedure 1202-8 I I

.L/' Title Revision No.

CRD Equipment Failure 53 E. STATOR HIGH TEMPERATURE 1E. Symptoms

1. Computer alarms and printouts of the affected rnechanisrn(s) giving the stator temperature.

2. Possible low ICCW CRD cooling flow as indicated on ICIO-FI or MAP alarn~C-7-2.

3. Possible ICCW CRD filter dP High alarm on MAP C-1-4.

4. Possible ICCW cooler outlet temperature high on IC6-TI or MAP alarm C-2-3.

5. Possible CRD cooling water outlet temperature high on IC9-l-l.

6. Possible IC system CRD cooling water outlet temperature high alarm on MAP C-'1-3.

2E. Immediate Actions

1. Automatic Actions

a. MU-V-INB close when CRD cooling water outlet temperature reaches 160°F.

2. Manual Actions NOTE The steps with an asterisk (*) will be re-verified as the first step of the follaw-up actions.

a. Verify MU-V-INB close if CRD outlet temperature reaches 160°F.

b. Check the ICCW system for proper operation by verifying

1. The following parameters for the CRD system:

1. Flow to CRD stators 100 gpm.

ii. CRD return temperature'< 160°F iii. dP across IC-F-INB.

c. If required to provide continued cooling to the GRDMs, then start the standby IC-P-1 and/or valve in the standby IC-F-1, 3E. Fotlow-up Actions Objective: The objective of this procedure is to prevent stator failure due to a loss of cooling water flow or high temperature in the ICCW system.

1. Reverify the steps in the Immediate Manual Actions that are marked with an asterisk v).

Use redundant indications where available.

u-26

TMI NUCLEAR Emergency Procedure 1202-8 Title Revision N o .

CRD Equipment Failure 53

2. Check all stator temperatures for high temperature indication.

3. Check CRD cooling water outlet temperature on ICQ-TI. If temperature is high, then check river water flow path through the coolers. Start an additional river water pump as necessary.

4. Place the temperature input of the affected mechanism on analog trend on the PPC and reset the alarm setpoint for that point to 170°F.

5. If the temperature is between 160°F and 170"F, then continue normal operation.

6. If the mechanism is above 1TOOF, then de-energize the mechanism as follows:

a. Reduce reactor power to 60% of the thermal power allowed for the RC-P combination.

b. Obtain O&M Director's approval, or Plant Operations Director if O&M Director is not available and bypass the asymmetric rod alarm for the mechanism by placing toggle switch S2 on the individual PI amplifier in the down position.

c. Determine the mechanism number for the rod from EflClQSure 6 of 1105-9, Control Rod Drive System or from NAS display 5 .

-d- d. Transfer the affected rod to the Auxiliary Power Supply

e. Drive the rod fully into the core.

CAUTION Use extreme care because of the energized components in the vicinity of the fuses removed in the next step. -

9- Pull the six (6)motor output fuses located in the transfer cabinets for the mechanism.

h. Replace the programmer lamp fuses F1 and F3.

I. Press FAULT RESET on the Diamond Rod Control Panel.

L-'

27

Number TM I NUCLEAR Emergency Procedure 1202-8 L'

Title Revision No.

CRD Equipment Failure 53

7. If t h e stator temperature reaches 180°F,then de-energize the mechanjsm immediately See Steps 6.d and 6.f above.

8. If more than one stator reaches 180"F, then trip the reactor and go to ATP-1210-1.

28

Number TMI - Unit 1 Operating Procedure I105-9 Title Revision No.

Control Rod Drive System 61 1.4 Figures, Tables, Charts

a. Control Rod Drive Mechanism Control System Instruction Manual Vol. 1, Figure 1-2, Sheet 1 4 , "Motor Control Composite View", used to correlate mechanism number to motor output fuse location.

b. Figure 1, CRD Control Panel C. Figure 2, Position Indication Panel

d. Figure 3, Programmer Control Assembly

e. Table 1, Control Rod Group Assignment

f. Table 2, Position Indication AmplifierlMechanism Correlation Chart

g. Figure 4, Rod Map

h. Figure 5, 120V AC Power Distribution Assembly I. Rod insertion Limits from Fuel Densification Study (from 11024) 2.0 LIMITS AND PRECAUTIONS '

W 2.1 Equipment

a. A control rod drive must be de-energized if the stator temperature exceeds 180°F as indicated by the Plant Process Computer.

b. Maximum allowable temperature of cooling water entering the CRDs is 120°F. Inlet temperature may be read on IC6-TE indicator on console CR.

C. Prior to energizing CRDs, a Minimum of 100 GPM Intermediate Closed Cooling Water (ICCW) flow must be provided to the CRD stators as indicated on console CR Indicator ICIO-PI.

d. Maximum rod travel is 420"lhour.

e. Maximum CRD running time is 30 minutes per hour.

f. The maximum desired air temperature around the drives in the service structure is 160°F.

See computer points A0422, A0423 and A0424. When the CRDs are energized, assure that a reactor compartment fan should be running per 1104-14C.

5

SYSIEP# 027 KA# AK2.03 Page # 4.2-20 Tier # 1

%u ROISRO Importance Rating 2.6 -2.8 Group # -2 Knowledge of the interrelations between Pressurizer Pressure Control (PZR PCS) Malfunction and the following: Controllers and positioners Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Pressurizer pressure control systems are in automatic.

- Control console Pressurizer pressure SETPOINT signal fails to 2500 psig.

From the list below, identify the ONE statement that describes automatic response to this malfunction.

A. RC-V-I Pressurizer spray valve opens to 100%.

B. RC-V-1 Pressurizer spray valve opens to 40%.

C. Pressurizer heater banks 1,2and 3 energize.

D. RC-RV-2 PORV opens.

PM Section B-01, Reactor Coolant System, Rev 13, Page 28.

E New 9TMIBank TMI Question #

3 Modified TMI Bank Parent Question #

u 5 Memory or Fundamental Knowledge E Comprehension or Analysis g 55.41 .7 3 55.43 'a55.45 .7 A Incorrect answer. Spray valve does not operate off the control panel setpoint potentiometer.

B Incorrect answer - this is a demand signal, not feedback signal, no auto open positioning will occur.

C Correct answer. The elevated setpoint produces an error signal to fire all SCR controlled heaters.

D Incorrect answer. PORV is bistable controlled independent of the setpoint potentiometer on Console Center.

TMI SRO Exam May 2003 F J ~ I March

~ ~ v , 28,2003

i SECTION B-0 1 REVISION 13 10.0 CONTROLS

\ /

The operator is provided with controls to operate pressurizer heaters, PORV, spray valve and the PORV and spray valve block valves.

Pressurizer heaters are controlled fiom console right. The modulatedcontrol heaters work in nsole center. The setpoint for normal operating he heaters in banks 4 and 5 are on-off control and are strictly a function of RCS pressure. AI1 5 banks of pressurizer heaters are provided with red, green indicator lights.

The PORV is operated manually by using a autdopen switch on console center for an explanation of the electrical print see Figure 10.

The spray valve can also be operated manually from console center. The spray valve electrical diagram is described on Figure 11.

The spray and PORV block valves are operated using their respective pushbutton controls Iocated on console center.

28

Form ES-401-6 SYSIEP# 029 KA# EK2.06 Page # 4.1-9 Tier # -1

'u ROlSRO Importance Rating E 3.1* Group # 1 Knowledge of the interrelations between Anticipated Transient Without Scram (ATWS) and the following: Breakers, relays, and disconnects.

a C '-B The Diverse Scram System (DSS) provides backup protection for an ATWS (Abnormal Transient Without Scram) by (1) the shunt trip coils on the (2) 480V breakers.

A. (1) de-energizing (2) 1L-2A and 1G-2A (CRD System Feeders)

6. (1) energizing (2) 1L-2A and 1G-2A (CRD System Feeders)

C. (1) de-energizing (2) CRD #0Iand # I 1 D. (1) energizing (2) CRD #I 0 and # I 1 OPM Section F-02, Reactor Protection System, pages 123 and 124, Rev. 9 (IO).

E New 3 TMlBank TMI Question #

3 Modified TMI Bank Parent Question #

a Memory or Fundamental Knowledge

$5Comprehension or Analysis E 55.41 .7 E 55.43 a 55.45 .7 A Incorrect answer. First part is incorrect, second part is correct (see 6).

B Correct answer. First and second parts are correct, shunt trip is energize to actuate, and those are the correct breakers.

C Incorrect answer. Both parts of this answer are incorrect. Second part is incorrect, although these breakers do have shunt trips, DSS does not actuate them.

D Plausible distractor. First part is correct, second part wrong (although) it does have shunt trip as part of its source interruption device, this signal does not activate it.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION F-02 REVISION 9 During 13R two additional lights were added to each circuit breaker trip circuit. One across each breakers Under Voltage coil tu provide U/VDEF?CEENERGlZED status when the light is on and one across each breakers Shunt Trip coil to provide S T TRIP SIGNAL PRESENT status when the light is on. The U N light will normally be energized when a trip signal or under-voltage condition is not present. The Shunt Trip light will normally be de-energized except when the shunt trip coil is energized to trip the breaker by either the Source Interruption Device or by the U N trip device. The two indicating lights will be physically mounted on the front of the second cubicle door from the top of the 10 and 1 1 breaker cabinets on the lower left and right hand sides of the door for the UN DEVICE ENERGIZW status lights and the SHUNT TNP SIGNAL PRESENT status fights respectively.

A twelve point terminal block was also added to allow more convenient monitoring of voltage at various control circuit locations on the 10 and 1 1 breakers. The terminal blocks will be physically mounted on the back side of the second cubicle door from the top of the IO and 1 1breaker cabinets on the lower portion of the door.

The testholation switches and control relays are classified as NSR. The indicating lights are classified as ITS.

All test/isolation switches and indicating lights are mounted either on the door (see Figure 67), or in the cubicle with the associated undervohge relay. The control relays are mounted inside the .

associated AC or DC breaker cabinet enclosure.

8.5 Operation The four undervoltage relays are continuously energized during reactor operation via the respective RPS trip logic output voItage. Upon interruption ofthis voltage, due to an RPS or manual trip command, contacts ftom these undervoltage relays trip their associated CRDM power feeder circuit breaker by energizing its shunt trip coil. To prevent unnecessary tripping of any of these circuit breakers upon loss and restoration of DC control power to the undervoltage relays, the trip contacts logic is such that they remain open upon loss of control power.

Testing of the trip circuits is done monthly per 1303-4.1 by I&C.

9.0 ATWS MODIFICATIONS@SSb DINERSE SCRAM SYSTEM Purpose The Code of Federal Regulations was revised and a requirement intended to reduce the risk fiom Anticipated Transient Without Scram (ATWS)events is applicable to TMI. The modifications were made to meet the requirements as described in 10 CFR 50.62 and to resolve related equipment problems.

9.1 General Description Diverse Scram System @SS)

Equipment is installe d lL-2A a u ~ ~ c aiflRCS ~ ypressure than 2500 pig or m 123

SECTIUN F-02 REVISION 9 9.2 Components W

Diverse Scram System 9.2.1 A foxboro cabinet has been installed in the control tower on the third floor patio area across from the CRD AC breakers.

9.2.2 Shunt trip coils have been installed in the 1G-2A and 1L-2A breaker cubicles to permit remotely opening the breakers.

9.2.3 A manual actuation pushbutton has been installed into console center directly above the RPS Rx. trip pushbutton.

9.2.4 Additional modules have been installed in the Remote Shutdown Signal Conditioning Cabinets to interface the RCS pressure signals to the DSS.

9.2.5 Additional relays have been installed in the cubicle above CRD AC Breaker #I 1 to interface 1G-2A & 1L2A positions to the reactor trip confirm logic.

9.3 Operation Diverse Scram System (Figure 74) 9.3.1 The DSS will provide a backup to the RPS/CRD systems if the RPS fails to actuate or if CRD breaker malfunctions prevent the CRDM mechanism from being deenergized. If RCS pressure increases to >2500 psig or if the operator manually actuates the DSS, the shunt trip coils on the 1G-2A and 1G2A breakers will be energized to trip these breakers.

The DSS uses RC-PT-963 and RC-PT-949 (0-3000 psig) for indication of RCS pressure. If both indications are greater than setpoint (2500 pig) then both output relays are energized which-in turn energize the breaker shunt trip coils. The DSS requires both input channels to TRIPin order to energize either output relay. The output relays would energize together except for testing. (Testing will be performed on a refuehg intervaI.)

The DSS Manual Actuation Pushhaon directly energizes the breaker shunt trip c ushbutton bypasses all the actuation logic but is disabled by the DEFEA The DSS DEFEATENABLE switch located inside the DSS cabinet opens the circuit to prevent any type of accidental energizing of the breaker shunt trip coils.

The DSS uses non vital power (120 VAC from CT-3 Sw.#7A) for cabinet logic and to operate the breaker shunt trip coils.

124

orm ES401-6 SYSiE?# 029 KA# EA1.05 Page # Tier #

w ROISRO Importance Rating 3.7* 3.6' Group #

Ability to operate and/or monitor the following as they apply to Anticipated Transient Without Scram (ATWS): BIT outlet valve switches.

Sequence of events:

- Reactor power was initially 1OO%, with ICs in full automatic.

- No maintenance or surveillance tests in progress.

- Automatic reactor trip.

- CRD safety groups 1-4 fail to drop into the core.

Based on these conditions, select the ONE statement below that describes required response to this event.

A. Emergency borate from the BWST in acccordance with Rule 5, EB.

B. Manually insert CRD Groups 1-4 from the Diamond Control Panel.

C. Locally open the CRD DC Hold power supply breakers.

D. Trip all four RCPs.

P-TM-EOP-001, Reactor Trip, Step 3.3, page 3, Rev. 3.

E New 3 TMlBank TMI Question #

u 0 Modified TMI Bank Parent Question #

a Memory or Fundamental Knowledge d Comprehension or Analysis c 55.43 E 55.45 51.6 A Correct answer. BWST is the preferred source of emergency boration.

B Incorrect answer. Rods are already de-energized by the reactor trip. This action would actully require trip reset (blocked since Groups 1 4 are not at their inlimit.

C Incorrect answer. Although this would de-energize the rods, they are already de-energized by the upstream breakers.

D Incorrect answer. This is an action for loss of SCM, not stuck rods.

None.

TMI SRO Exam - May 2003 Friday, March 28, 2003

OP-TM-EOP-001

~-

Revision 3 Page 3 of I 1 3.0 VITAL SYSTEM STATUS VERIFICATION (VSSV)

T I 3.1 IAAT a symptom exists, then immediately treat the symptom using the following priority:

1. SCM 25°F GO TO OP-TM-EOP-002.

2. XHT GO TO OP-TM-EOP-003.

3. LOHT GO TO OP-TM-EOP-004.

4. OTSG tube leakage > 1 gpm GO TO OP-TM-EOP-005.

Time 3.2 ANNOUNCE Reactor Trip over plant page and radio.

u - 3.3 VERIFY control rod groups 1 through 7 INITIATE Emergency Boration are fully inserted. per RULE 5 - EB.

3.4 VERIFY MAIN FW Flow is ENSURE FW-V-5A AND excessive. FW-V-5B are stroking closed or are closed.

3.5 VERIFY OTSG level > setpoint. INITIATE RULE 4 - FWC.

3.6 VERIFY ICS/NNI HAND or AUTO Power 1. TRIP both MFW pumps.

are available. 2. ENSURE EFW actuation and INITIATE Guide 15.

3. CONTROL OTSG pressure using the ADV B/U loaders

4. INITIATE 1202-40, Loss of ICs Hand and Auto Power.

Form ES401-6 SYS/EP# 033 KA# AK3.01 Page # 4.2-26 Tier # 1 xu ROlSRO Importance Rating 3.2 -

3.6 Group # -

2 Knowledge of the reasons for the following responses as they apply to Loss of Intermediate Range Nuclear Instrumentation: Termination of startup following loss of intermediate-range instrumentation.

Plant conditions:

- Plant startup in progress.

- Intermediate Range nuclear instrument NI-3 is inoperable due to log amplifier failure. Replacement amplifier will arrive in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

- Reactor is critical at 1E-8 Amps.

Event:

- NI-4 Intermediate Range nuclear instrument log amplifier fails in same manner as NI-3, and is required to be replaced.

Based on these conditions, identify the ONE statement below that describes why startup is required to be terminated.

A. Loss of reactor protection automatic trip function.

B. Inability to detect an ejected rod accident.

C. Lack of protection against a continuous rod withdrawal accident.

D. Non-compliance with Remote Shutdown System instrumentation requirements.

W Technical Specification 3.5 including Table 3.5-1 (Amendment 189), FSAR sections 14.1.2.2, 7.3.1.3, 7.2.2.1.c.2, Update 154/00.

Technical Specification 3.5 including table 3.5-1, Amendment 189.

V.G.OI.01 New TMI Bank TMI Question #

c Modified TMI Bank Parent Question ##

Memory or Fundamental Knowledge 3 Comprehension or Analysis 55.41 .5/.10 g 55.43 .2 E 55.45 .6/.13 A incorrect answer - no trip functions from Inter. Range instruments cause rx trip.

B Incorrect answer. Ejected rod withdrawal accident is not specifically tied to IR instruments; other means of detection are available.

C Correct answer. The 3 DPM SUR withdrawal inhibit circuitry provides this protection at low power levels as documented in FSAR and TS bases.

D Plausible distractor. However since the minimum degree of redundancy for IR instruments is ZERO, minimum redundancy requirements are available.

Answers are all Tech Spec and FSAR bases. Allowing the availability of Tech Specs will assist L...,, in excluding only one distractor.

TMI SRO Exam - May 2003 Friday, March 28,2003

3.5 INSTRUMENTATION SYSTEMS 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION A p ~icatii l 1it v A p p l i e s to unit instrumentatfon and cantrol systems.

0bject i ve To delineate t h e conditions o f the unit instrumentation and safety circuits necessary to assure reactor safety.

Specifications 3.5.1.1 The reactor shall not be i n a startup mode or i n a critical state unless the requirements of Table 3.5-1, Column "A" and "W' a r e met, except as provided in Table 3.5-1, Column "C". Specification 3.0.1 appt ies.

3.5.1.2 The key operated channel bypass switch associated w i t h each reactor protection channel may be used tQ lock the reactor trip module in the untripped state as indicated by a light. Only one channel shall be lacked in t h i s untrfpped state at any one time. Unit operation at rated power shall be permitted to continue with Table 3.5-1, Column "A". Only one channel bypass key shall be kept in, the control room.

3.5-1.3 In t h e event the number o f protection channels operable falls below the limit given under Table 3.5-1, Column "A", operation shall be u limited as specified in Column 'IC". Specification 3.0.1 applies.

3.5.1.4 The key operated shutdown bypass switch associated with each reactor protection channel shall not be used during reactor power operation (except for required maintenance or testing).

3.5.1.5 During START-UP when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade.

3.5.1.6 During START-UP, HOT STAFSOBY or POWER OPERATION, in the event that a control rod drive trip breaker t s inoperable; within one hour .

place the breaker in trip. Specification 3.0.1 applies.

3.5.1.7 During START-UP, HOT STANDBY or POWER OPERATION, i n the event that one o f the control rod drive t r i p breaker diverse trip features (shunt trip or undewol tage trip attachment) i s inoperable:

a. Restore to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OF

b. Within one addttional hour place the breaker i n trip.

Specification 3 .O. 1 appt i es.

3- 27

TABLE 3.5-1 INSTRUMENTS OPERATING CONOIT IONS 2

0 Functional Unit (A) (B) (C) v

I Minimum Operable Minimum Degree Operator Action i f Conditions D

I.

Channels of Redundancy o f Column A and B Cannot be Met

-4

.89 , A. Reactor Protection System

+a.

'Q m 1. Manual push butt on

-5 02 2, Power range instrument channel

3. Intermediate range instrument channels

4. Source range instrument channels 0

5. Reactor coolant temperature 1 w instrument channels .

L Tu 0 . 6. (Del eted) ----

7. Flux/imbalance/flow (a) 8 .

.8. Reactor cool ant pressure

a. High reactor coolant pressure 1 instrument channels

b. Low reactor coolant' pressure 1 instrument channels

(6 TABLE 3.5-1 (Cont'd)

INSTRUMENTS OPERAT 1NG COND 1T IONS c

Functional U n i t (A) (B) (C)

Minimum Operable Minimum Oegree Operator Act ion i f Conditions Channels o f Redundancy o f Column A and 3 Cannot Be Met Reactor Protection System (cont'd).

9. Power/number of pumps 2 1 (a) instrument channels

10. high reactor building 2 1 (a) pressure channels (a) Restore the conditions o f Column (A) and Column (B) w i t h i n one hour or place the u n i t i n HOT StiUTDOWN withi an additlonal 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(b) When 2 o f 4 power range instrument channels are greater than 10 percent f u l l power, intermediate range

.. ' instrumentation i s not required.

(c) When 1 of 2 4ntermediate range instrument channels i s greater than IO-'* amps, or 2 o f 4 power range instrument channels are greater than 10 percent f u l l power, source range instrumentation i s not required.

TMI-1/FSAR Table 14.1-2 summarizes the reactivity changes and the corresponding change in the average moderator temperature. These reactivity changes are extremely u slow and allow the operator to detect and compensate for the change. Operator actions to counter the reactivity changes would include boration and control rod motion.

14.1.2.2 Startup Accident

a. Identification of Cause The objective of a normal startup is to bring a subcritical reactor to the critical or slightly supercritical condition, and then to increase power in a controlled manner until the desired power level and system operating temperature are obtained. During a startup, an uncontrolled reactivity addition could cause a nuclear excursion. The uncontrolled reactivity addition, through rod withdrawal from zero power, is a startup accident. This excursion is terminated by the strong negative Doppler effect if no other protective action operates.

I The following design provisions minimize the possibility of inadvertent continuous rod withdrawal and limit the potential for power excursions:

1) The control system is designed so that only one control rod group can be withdrawn at a time, except that there is a 25 percent overlap in travel between two regulating rod groups successively withdrawn. This overlap occurs at the minimum worth positions for each group because one group is at the end of travel and the other is at the beginning of travel.

2) Control rod withdrawal rate is limited.

3) A startup rate withdrawal stop and alarm are provided in the source range.

4) A startup withdrawal stop and alarm are provided in the intermediate range.

5) A high flux level and a high pressure trip are provided.

The criteria for the analysis of this accident is that the Reactor Protection System shall be designed to limit (1) the reactor thermal power to the design overpower condition (I 12 percent rated power) and (2) the Reactor Coolant System pressure so as not to exceed the ASME Code allowable pressure limit of 2750 psig (1I O percent of Design Pressure).

14.1-3 UPDATE-I5 4/00

TMI-l/FSAR calibration facilities provide a means for reading the output of the individual sections of the detector. Each detector has a combined sensitive volume extending approximately from the bottom to the top of the reactor core.

Dual element fission chamber detectors are used in the source range and wide range channels. Each detector, approximately 58 inches long, is connected via special Class 1E qualified cable to an amplifier assembly outside the Reactor Building. Connected to each amplifier is a signal processing unit that provides the output to the Control Room indicator.

The physical locations of the neutron detectors are shown on Figure 7.3-3. A power range detector is located external to each quadrant of the core.

The two intermediate range detectors are located on opposite sides of the core but rotated approximately 55 degrees from the source range detectors.

b. Test and Calibration Test and calibration facilities are built into the system. The facilities permit an accurate calibration of the system and the detection of system failures in accordance with the requirements of the reactor protective system design and IEEE Standard 279 (see 7.5, Reference 2).

7.3.1.3 System Evaluation The nuclear instrumentation will monitor the reactor over a minimum 10 decade range from source to 125 percent of rated power. The full-power neutron flux level at the power range detectors will be approximately 3.2 x I O 9 nv. The detectors employed will provide a linear response up to 2.5 x I O nv.

The intermediate range channels overlap the source range and the power range channels as shown on Figure 7.3-2, providing the continuity of information needed during startup.

7.3-3 UPDATE-I3 4/96

TMI-1/FSAR

-9 b. Reactivity Rate Limits The speed of the mechanism and group rod worth provide the reactivity change rates required. For design purposes, the maximum rate of change of reactivity that can be inserted by any group of rods has been set in Item b. of Section 14.1.2.2. The drive controls, Le., the drive mechanism and rods combination, have an inherent speed limiting feature.

C. Startup Considerations The CRDS design bases for startup are as follows:

I) Reactor regulation during startup is a manual operation.

2) Control rod out motion is inhibited when a high startup rate (short period) in the source range or intermediate range is detected.

d. Operational Considerations For operation of the reactor, functional criteria related to the rod drive control system are:

1) CRA Positioning P The CRDS provides for controlled withdrawal, controlled insertion and L U holding of the CRAs, to establish and maintain the power level required for a given reactor coolant boron concentration.

2) Position Indication Continuous rod position indication, as well as full-in and full-out position indication, is provided for each control rod drive.

3) System Monitoring The CRDS design includes provisions for routinely monitoring conditions that are important to safety and reliability.

7.2.2.2 Svstem Desicln The CRDS provides for withdrawal and insertion of the CRAs to maintain the desired reactor output. This is achieved either through automatic control by the ICs, discussed in Section 7.2.3, or through manual control by the operator. As noted previously, this control compensates for short term reactivity changes. It is achieved through the positioning in the core of 61 CRAs and eight axial power shaping rod assemblies.

7.2-2 UPDATE-8 I

7/89

F o ES401-6

~

SYSiEP# 038 KA# EA1.08 Page # 4.1-11 -

u RO/SRO importance Rating 3.7* -

3.8* Group # -2 Ability to operate and/or monitor the following as they apply to Steam Generator Tube Rupture: Core Cooling monitor (for TMI, this is SCM).

Plant conditions:

- Reactor is shutdown due to Steam Generator tube leak in OTSG IA.

- All RCPs are operating.

- Both OTSG levels are steady at 25 inches.

- T-Ave is 510°F.

- RCS subcooled margin is 93°F.

The Unit Supervisor directs the team to "minimize RCS subcooled margin." Based on the conditions above, identify the ONE set of actions below that complies with this order.

A. De-energize Pressurizer heaters and open RC-V-1 (spray valve) to reduce RCS pressure until SCM lowers to 45°F.

B. Close Turbine Bypass Valves to raise RCS temperature until SCM lowers to 45°F.

C. De-energize Pressurizer heaters and open RC-V-1 (spray valve) to reduce RCS pressure until SCM lowers to 75°F.

D. Open the PORV to reduce RCS pressure until SCM lowers to 75°F.

OP-TM-EOP-010 Guide 8, RCS Pressure Control, pages 20 and 21, Rev. 1.

OS-24, Conduct of Operations During Abnormal and Emergency Events, sections u 3.8 and 4.7.2 (pages 4 and 22), Rev. 7.

New Li TMlBank TMI Question #

5 Modified TMI Bank Parent Question #

c1 Memory or Fundamental Knowledge id Comprehension or Analysis I@ 55.41 .7 Ll 55.43 E 55.45 51.6 A Correct answer. In accordance with EOP-010 Guide 8, RCS Pressure Control, reduce pressure by turning heaters off and opening the spray valve. Proper SCM band is 30-70 degrees, as defined in OS-24, section 3.8.

B Incorrect answer. Method is not in accordance with EOP-010 Guide 8, RCS Pressure Control, although it reduces SCM. Lowering SCM to 45°F complies with OS-24 acceptable band.

C Incorrect answer. Method is correct, lowering SCM to 75°F is outside OS-24 acceptable band.

D Incorrect answer. Method is not correct (although it will reduce SCM). Lowering SCM to 75°F is outside OS-24 acceptable band.

None.

L-TMI SRO Exam - May 2003 Friday, March 28,2003

Number

- Tirle TMI - Unit 1 Operations Department Administrative Procedure OS-24 Revision No.

Conduct

- of Operations During Abnormal and Emergency Events 7 3.7 LACK OF PRIMARY-TO-SECONDARY HEAT TRANSFER (LOHT) is the inability of either OTSG lo remove sensible heat from the RCS. LOHT can be confirmed if :

Neither OTSG has water level control and pressure control.

AND 0 Core exit temperatures are rising 3.8 MINIMIZE SCM: An intentional reduction of the reactor coolant pressure temperature relationship as close as practical to the 25°F subcooling margin or emergency RCP NPSH limit. (Recommended band 30-70°F) 3.9 OTSG AVAILABLE AS A HEAT SINK:

A physical condition where the OTSG demonstrates level and pressure control, used to determine if primary to secondary heat transfer is possible. (i.e. heat sink) Primary to secondary heat transfer need not be demonstrated to determine this availability. Primary to secondary leakage should not be considered a means of OTSG level control. A dry OTSG is not considered available as a heat sink.

3.10 OVERS I G HT:

The independent monitoring of plant and crew performance and any subsequent intervention, as needed, to ensure the appropriate mitigation strategy is being pursued for the current plant conditions. Refer to Attachment B, SM Oversight Management Guidelines.

3.11 PLANNED REACTOR TRIP A scheduled shutdown, where a reactor trip, is directed by an approved procedure.

3.12 PRIMARY-TO-SECONDARY HEAT TRANSFER (PSHT) is the removal of sensible heat from the RCS to one or both OTSG(s). PSHT can be confirmed if:

0 Either OTSG has water level control and pressure control.

AND RCS T, is approximately the same as secondary T,, and responds to changes in OTSG pressure.

AND RCS forced or verified natural circulation is present.

3.13 RCP AVAILABLE -An available RC Pump is one which can be operated without extraordinary efforts. Pump service functions (motor cooling, seal cooling, etc. j are operable (redundancy not required} and all interlocks can be satisfied. Strict compliance with administrative shutdown criteria (vibration, seal leakoff flow, etc) is not expected when the operation of the pump is more important to safe plant operation.

4

Number TMI - Unit 1 Operations Department

~u Administrative Procedure OS-24 Tide Revision No.

Conduct of Operations During Abnormal and Emergency Events 7 4.6.4 Safety systems may be bypassed with US concurrence under the following conditions:

1. Any action required by an EVENT PROCEDURE that would cause a safety system actuation and actuation would complicate plant control when there is positive control of the actuation parameter 0 To BYPASS ESAS (HPI or LPI), SCM must be > 25°F and RCS pressure is being controlled (7 To BYPASS HSPS (Lo-Lo Pressure), the operations team must have control of OTSG pressure

2. When directed by the EVENT PROCEDURE or Operating Procedure for plant cooldown.

4.6.5 Safety system actuation signals may be defeated with US concurrence at any time after complete actuation of all components has been verified.

4.6.6 When safety systems are throttled, the Reactor Operator announces the action to the Control Room team. The Unit Supervisor acknowledges the team announcement.

4.7 Instrument Usage

'- 4.7.1 Safety Grade Instrumentation During degraded plant conditions, Safety Grade instruments are used when available to

, monitor and control plant parameters. When available, redundant or alternate indication is used to verify critical parameters.

4.7.2 Determination of Subcooling Margin (SCM)

1. Primary Method:

A. If RCS circulation exists (forced or natural), then use the most conservative safety grade Subcooling Margin monitor.

B. If RCS circulation does NOT exist. then use the value of computer point C4008, (Calculated lncore SCM).

2. Backup Method:

Use any combination of lncore temperature and RCS wide range pressure:

I Temperature Indications 1 Pressure Indications Computer point C4006 (average of five PI-949A (PCL) highest available incore thermocouples)

Highest reading available Backup Computer point A0404 (PT-963) lncore Read Out (BIRO) 22

OP-TM-EOP-01 0 Revision 1 Page 20 of 49 Guide 8 RCS Pressure Control (Page 1 of 2)

ACTIONlEXPECTED RESPONSE RESPONSE NOT OBTAINED

. ~ . ................................................. ....... ..........

VERIFY the reactor is shutdown and SCM

> 25°F.

If it is required to MINIMIZE SCM and SCM

> 7OoF, then lower RCS pressure to between 70" and 30" SCM.

RAISE or LOWER RCS pressure per the following direction, as needed, to maintain RCS pressure within the limits of Figure 1 and 1A.

u To RAISE RCS Pressure:

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY Pressurizer level is stable or rising. 1 RAISE RCS makeup or HPI.

VERIFY pressurizer level 80".

ENERGIZE Pressurizer Heater banks or IF Pzr Htr power is not available, then ADJUST heater demand for the SCR INITIATE OP-TM-220-901 to transfer Grp 8 controlled heaters. .. 9 to ES power.

or

OP-TM-EOP-010 Revision 1 Page21 o f 4 9 Guide 8 RCS Pressure Control (Page 2 of 2)

To LOWER RCS Pressure:

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ENSURE HPI is throttled per (Rule 2) and control RCS inventory.

VERIFY RCS temperature is stable or NOTIFY US.

decreasing.

If pressurizer steam bubble is controlling RCS pressure, then ENSURE Pressurizer Heaters are OFF.

.- -- _________ ~- -- - __.

\-.- NOTE Pressurizer cooldown rate is to be limited to less than IOOOF in any one hour VERIFY an RCP is operating. ENSURE WDG-V-3 and WDG-V-4 are OPEN (Guide 19).

OPEN RC-V-44.

OPEN RC-V-28.

When RC Drain Tank pressure > 40 psig or RCS pressure is in desired range, then, CLOSE RC-V-28.

CLOSE RC -V-44.

GO TO END.

- ENSURE RC-V-3 is open,

- OPEN RC-V-1.

- When RCS pressure is in the desired range, then CLOSE RC-V-1.

Form ES-QO1-6 KA# EA2.09 Page t# 4.1-12 Tier #

W ROlSRO Importance Rating 4.2 -

4.2 Group # -2 Ability to determine and interpret the following as they apply to Steam Generator Tube Rupture: Existence of natural circulation, using plant parameters Sequence of events:

E Ll m

- Reactor tripped due to loss of offsite power (LOOP).

- Tube rupture occurred in OTSG 1B.

- OTSG 1B was isolated and filled solid when T-Ave was 510°F.

- EF-P-1 was secured.

Current plant conditions:

- T-Ave is now 480°F.

- Subcooling Margin is 55°F.

- RCS pressure is 900 psig.

- EFW flowrates:

- OTSG 1A = 500 gpm.

- OTSG 1B = 0 gpm.

- OTSG 1A level is 30%, rising at ?%/minute.

- RCITS suggests the existence of a 8 inch reactor vessel HEAD BUBBLE.

Based on these conditions, identify the ONE statement below that describes CURRENT natural circulation core cooling capability.

Natural circulation core cooling is.._

W A. POSSIBLE in RCS Loop A, due to EFW spray effectiveness.

B. NOT POSSIBLE in RCS Loop A, due to low OTSG level.

C. POSSIBLE in RCS Loop B, due to large heat sink area.

D. NOT POSSIBLE in RCS Loop B, due to the head bubble.

Emergency Operating ProceduresTechnical Bases Document 74-1 152414-09, Volume 3:

Chapter 1II.G (Cooldown Methods) pages 12, 13, 20 and 21, Rev. 9.

Chapter 1V.C (MFWIEFW System Operation) pages 2 and 3, Rev. 9.

g New -__ TMI Bank TMI Question #

__ Modified TMI Bank I- Parent Question #

- Memory or Fundamental Knowledge

@ Comprehension or Analysis 5 55.41 3 55.43 d 55.45 .I3 A Correct answer. Subcooled natural circulation flow is possible if adequate EFW flow exists, even with low u OTSG level and a small head bubble.

B Incorrect answer - EFW spray flow provides elevated heat sink to support natural circulation cooling when reactor coolant is subcooled.

C Incorrect answer. Although plausible that it could be a large heat sink, this OTSG has been filled with water TMI SRO Exam - May 2003 Friday, March 28,2003

Form ES-401-6 Q # 023__

that is hotter than current RCS water conditions. OTSG 1B is now a heat source with respect to the RCS.

D Incorrect answer. Bubble is not the limiting factor for OTSG 1B. This OTSG has been filled with water that is t/ hotter than current RCS water conditions. OTSG 1B is now a heat source with respect to the RCS.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

NUMBER TECHNICAL DOCUMENT 74- 1152414-09 Control of natural circulation with low decay heat RV head cooling Idle RC loop and SG shell cooling Primary pressure control Longer cooldown time (FW requirements)

Possible thermal shock (if HPI on)

Typically, the EFW initiation and control systems are designed to provide sufficient EFW flow to remove maximum decay heat assuming a single failure of the system. When less than full decay heat exists, and especially when little or no decay heat exists, the EFW system can initially provide excess EFW flow to the SGs and may rapidly depressurize them. Proper initiation and control of the EFW system during a natural circulation event should be verified, and manual control taken, as necessary, to properly throttle and control EFW flow to prevent excessive overcooling and depressurization of the SGs.

- When little or no decay heat exists, little or no heat transfer by the SGs will be required.

Thus, while it is desirable to provide a continuous EFW flow to the SGs to reduce the thermal cycles on the EFW nozzles, it may not be possible to continuously feed the SGs without overcooling the SGs. Intermittent RCS flow stagnation may occur in one or both of the RCS loops when these conditions exist. However, if EFW flow has been stopped or throttled, the incore T/Cs will provide indication that heat transfer is necessary (temperature increasing), and that EFW flow needs to be increased. The resumption or increase in EFW flow wiII result in reinitiating natural circulation flow in the RCS. The requirements for establishing natural circulation levels in the SGs are discussed in Section IV.C.3.2, and the requirements for initiating and throttling EFW are discussed in Sections IV.C.4.2 and IV.C.4.3.

During natural circulation, the RV head fluid remains stagnant and COOIS very slowly (unless the plant has a passive vent system). To prevent head void formation, head cooling must be enhanced or the cooldown rate must be limited (see Section 3.7 for a more detaifed discussion of RV head cooling). A similar situation occurs in the hot leg during single loop natural circulation cooldowns. The idle loop of the RCS stagnates and cools slowly, unless action is taken to enhance idle loop coolingj(see Section 3.8 for more detailed discussion of idle loop cooling). Also, SG shell cooling difficulties may occur during single loop natural circulation cooldowns (see Section 3.6 for a more detailed discussion of SG shell cooling). Normal pressurizer spray is unavailable following a loss of RCPs.

Therefore, alternate primary depressurization methods are needed during natural circulation (see section 3.2).

r--

L-DATE 3/3 1/2000 VoI.3,III.G -12

E A MA T 0M E T E C H N O L O G I E S TECHNICAL DOCUMENT 74- 1152414-09 I Because of lower achievable cooldown rates and the previously mentioned difficulties, longer cooldown times may result before reaching DHRS initiation for natural circulation cooldowns. Longer cooldown times result in larger condensate requirements if steam is being vented to the atmosphere.

These cooldown times will vary depending on number of ADVs availabie, DHR cut in temperature, and decay heat level. Figure 1II.G-14 shows cooidown times and condensate requirements assuming maximum decay heat load.

In general, the more steaming capacity and the higher the DHR cut in point, the faster the cooldown can be accomplished. There are, however, times when head bubble considerations may force longer cooldown times regardless of other conditions.

3.3.2 Natural Circulation Cooldown Initiation and Verification To initiate a natural circulation cooldown when SCM exists, the water level in each SG should be raised to the natural circulation level setpoint. If the RCS is saturated, raise the level to the loss of SCM setpoint. The TBVs (condenser dump valves or ADVs) should be opened and controlled to limit the cooldown rate. Enough boron should be in the reactor coolant system to ensure preservation of shutdown margin should fluid in regions such as the pressurizer, idle loop or RV head mix with the remainder of the RCS and subsequently enter the core. Adequate shutdown margin can be complied with by RCS boron concentration or by boron injection as allowed by Technical Specifications.

Once natural circulation has been initiated and verified, RV head cooling considerations should be addressed (see Section 3.7). For the case of single loop natural circulation cooldowns, idle loop cooling and shell cooling difficulties should be addressed (see Section 3.8 and 3.6). Verification of natural circulation and recognition of a loss of natural circulation are discussed in Chapter 1I.B.

3.3.3 Impact of Steam Generator Tube Rupture' Natural circulation cooldowns with SGTRs may result in an idle loop because of SG isolation'or a loop with intermittent flow because of periodic steaming. In either case, one loop may be cooling much more slowly than the other and voids will result if saturated conditions occur in the loop that is cooling much more slowly. Discussions of the prevention and elimination of loop voiding are contained in Section 3.8.

3.4 Solid Pressurizer Operation Plant operation with a pressurizer steam bubble is preferable to solid pressurizer operation.

With a pressurizer steam bubble, primary pressure can be increased using the pressurizer heaters and decreased with pressurizer spray (if RCPs are operating) or with one of the 7 313 1/2000 I PAGE V01.3, 1II.G -13

I NUMBER I

'-1 TECHNICAL DOCUMENT 74-1 152414-09 I Ambient Heat Loss Induced Condensation.

RCPRestart Increasing the void size due to depressurization should be prevented after the vent is opened. To accomplish this, MU flow should be increased to compensate for venting. The RC pressure is to be maintained or slightly increased. Successful venting has occurred when primary pressure and pressurizer level increase suddenly. This level increase will occur even if a void exists in a hot leg.

RV head level measurements should only be used to indicate trends and may not suffice to determine if the void has been completely eliminated. RV head level measurements (if available) may not be accurate while venting is in progress.

RV head void elimination due to ambient heat loss induced condensation is a slow process and may require extremely long cooldown times. RCP restart with a RV head void is addressed in Chapter 1V.A.

3.7.4 Cooldown With Voided Head The RCS can be cooled and depressurized with a RV head bubble.* The depressurization rates achievable using the PORV will depend on the volume of steam in the system.

Depressurization will only be possible with the PORV while a steam bubble exists in the pressurizer. The exact reduction of the depressurization rate will depend on the sizes of the RV head bubble and the pressurizer bubble. As the cooldown progresses the RV head bubble will slowly expand. When the PORV is opened the bubble will expand rapidly and if large enough will expand into the hot leg. For natural circulation flowrates as slow as 3 percent, and temperatures in the hot leg within 10°F of saturation, a steam void escaping into the hot leg will be condensed long before it reaches the highest portion of the hot teg.

A 5O"F/hr cooldown with a 600°F steam void in the RV upper head is acceptable from an operations as well as a stress analysis standpoint. The flow regime expected at the RV outlet is one of bubbly flow as opposed to slug flow. This indicates that the condensation will not be violent. During this mode of expansion of the RV head void (steam escaping into the hot leg), the RV head void will not hinder depressurization.

The existence of a head void would lly require that the cooldown rate be limited to

< 5O"FMR. However, if the cooldown rate cannot be limited to this value, then continued SG cooling is preferable to HPI coolins? HPI cooling would induce greater thermal stress on the RV than continued SG cooling at > SO"F/HR with a head void.

3.7.5 Impact of Steam Generator Tube Ruptures L' If a fast depressurization and natural circulation cooldown is performed for SGTRs, RV head void formation is likely; Actions consistent with SGTR management may be taken to 7 313 1/2000 1 PAGE V01.3,III.G -20

TECHNICAL DOCUMENT 74-1 152414-09 eliminate the RV head void and prevent subsequent RV head void formation whenever possible. (See Chapter 1D.E).

3.8 .Idle Loop Cooling; '

During single loop natural circulation cooldowns, fluid in the idle loop remains stagnant and does not thermally communicate with the rest of thefRCS. It thus cools very slowly due to ambient losses. ,This has the following potential impact on the plant cooldown:

To avoid idle loop void formation the plant cooldown rates may have to be reduced The formation of loop voids will slow the depressurization and cooldown of the plant.

Unacceptable compressive tube-to-shell temperature differences in the idle SG could result if the idle loop is not cooled while the SG shell is cooling to ambient.

3.8.1 Idle RC Loop Void Prevention The temperature of the fluid in the idle RC loop may be kept below saturation temperature through either periodic injection of EFW to the SG in the idle RC loop or the bumping of a RCP in the active RC loop, provided the RCP restart criteria of 1V.A are met. The hot fluid in the idle RC loop is replaced with fluid closer to the active RC loop temperature by the bumping of a RCP. Injection of EFW should not be used if an unisolable steam leak exists in a location that could be hazardous to personnel or key equipment.

Using EFW to induce idle RC loop cooling involves periodically injecting EFW to the SG in the idle RC -loop and then observing the primary temperature response. Since natural circulation will start slowly, several minutes may elapse before the hot leg RTD begins to change.

The SG in the idle loop may be bottled up or dry. EFW (or MFW diverted through the EFW header) should be fed at less than 700 GPM to an idle SG (or within the limits of Rule 4.0 if the SG is dry) whenever the following conditions exist:

Idle RC loop tube-to-shell temperature differential (shell hotter than the S G tubes) is increasing to within 10°F of the tensile tube-to-shell temperature differential limit. It is expected that this action will momentarily increase the temperature differential before the proper thermal coupling is established between the shell and the tubes. A more complete discussion of shellhube temperature differences is contained in Section 3 . 6 .

The saturation temperature during the cooldown is reduced to within 10°F of the idle loop hot leg temperature.

7 313 112000 I PAGE V O I . ~ , I I I . -21 G

A M A T 0M E NUMBER T E C H N O L O G I E S 74-1 152414-09 B. Natural Circulation Setpoint - This setpoint is used when no RCPs are on and the RCS has adequate SCM.

C. Loss of Subcoolins Margin Setpoint - This setpoint is used when SCM does not exist.

D. ICC Setpoint - This setpoint applies only to Davis-Besse and is used when the RC temperature and pressure is in region 3 ofFigure 1II.F-1.

E. Shutdown - SG Overfill Setpoint - The SG level must be kept below this value to prevent water from entering the steam lines after a reactor trip.

F. EFW Start SG Level - This is the SG water level at which the EFW system starts following a loss of MFW.

3.0 STEAM GENERATOR WATER LEVEL This section discusses the bases for SG level setpoints. Setpoint values are plant specific. In general, the SG water level is maintained only high enough for adequate primary to secondary heat transfer to limit unnecessary filling of the steam generator.

The required SG water levels when two SGs are operating are discussed in the following sections. When only one SG is operating in natural circulation, raising the water level slightly higher may be beneficial. This action increases the heat removal ability of the one operating SG.

The rate at which the SG level setpoints are achieved is determined by the FW flow rate guidelines discussed in Section 4.0 3.1 SG Level With Operating RCPCs)

When using MFW or EFW and at least one RCP is on and SCM exists. the SG level SHOULD be controlled at or above the [low level setpoint].

This level is sufficient for removing core and RCP heat when the RC is subcooled and forced circulation of RC exists between the core and the SG.

3.2 SG Natural Circulation Setpoint If all the RCPs are deenergized and the RC SCM exists. the SG level SHOULD be controlled at or above the [natural circulation setpointl.

The RC will have to flow between the core and the SGs by natural circulati requires a higher SG level than does forced-RC flow. The SG lev enough to create a SG heat sink thermal center sufficiently above the r

L ,

DATE 313 1/2000 Vo1.3,IV.C -2

K *T E TC

M E H H O L O G I E S I NUMBER I TECHNICAL DOCUMENT 74- 1152414-09 1-center to induce adequate natural circulation of the RC. AIthough obtaining the necessary SG levels is the preferred method for establishing natural circulation, adequate natural circuIation can occur due to EFW flow onIy. For certain in-plant conditions (e.g., low decay heat level), it may not be necessary to increase SG levels as long as EFW flow is providing adequate heat transfer.

- 1 3.3 SG Loss of Subcooling Margin If the RC SCM is lost. the SG level MUST be controlled at or above the [loss of subcoolins margin setpointl.

This SG level is even higher than the natural circulation setpoint. When the RC SCM is lost, the RC has a potential of being saturated. Therefore, the SG water level is raised to the loss of SCM setpoint. This setpoint has been determined for a saturated RCS with steam in the hot leg pipes, for the case when core cooling is assisted by boiler condenser cooling. The core heats the surrounding water creating steam which flows through the hot leg pipes to the SGs where it is condensed. The resulting pool of water in the SG tubes must be higher than the elevation of the RCP internal spill-over so that the cold leg water will flow to the RV. For this to happen, the SG condensing surface has to be higher than the RCP internal spill-over. Also the condensing surface of the SG tubes must be adequate, combined with HPI cooling, to remove all the heat being generated by the core. The elevation of the EFW nozzles is high enough to provide the required condensing surface. The level setpoint is set high enough such that one SG can provide the required condensing surface during periods of no EFW flow.

3.4 If ICC conditions exist with an indicated fuel clad temperature greater than 1400°F the SG levels SHOULD be raised to the rICC level setpointl. (Onlv applicable to Davis-Besse)

The SG water level should be raised to the maximum level possible without causing water to enter the steam lines or losing SG level measurement or causing SG overfill protection system actuation; i.e., SFRCS trip on high SG level. This will provide the greatest SG condensing surface.

4.0 FEEDWATER CONTROL TO STEAM GENERATORS TJX4T CAN HOLD PRESSURE This section applies only to SGs that can hold pressure. Special considerations for SG(s) that cannot hold pressure are given in Section 5.0.

The FW flow rate should be controlled to increase and decrease the SG level to obtain the required setpoint.

I i

DATE 3/3 1/2000 V01.3, 1V.C -3 PAGE

Form ES-401-6

- Page # 4.2-33 Tier # 1

-4 ROlSRO Importance Rating 3.1 Group # -1 Knowledge of the reasons for the following responses as they apply to Loss of Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum.

Transferring OTSG pressure control from the turbine bypass valves to the atmospheric dump valves during loss of vacuum conditions in the Main Condenser prevents A. condenser tube leaks.

B. over-fill of the condenser hotwell.

C. damage to turbine bypass valve operators.

D. rupture of low pressure turbine exhaust hood diaphragms.

OPM Section H-4, Main and Auxiliary Condensers, page 3, Rev. 3.

OPM Section F-03, Integrated Control System, page 31, Rev. 9 (11).

None.

IV.C.06.07 a New B TMl Bank TMI Question # #99 SR021 Audit 3 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge

\

4 ,

a Comprehension or Analysis 55.41 .5/.10 CJ 55.43 55.45 .6/.13 A Incorrect answer, although plausible since tube leaks can occur with or without the vacuum.

B Incorrect answer, although plausible since mass flow into the condenser could cause level to rise.

C Incorrect answer, although plausible due to overdemand on the TBV (reduced valve dp) to maintain same OTSG pressure.

D Correct answer. Diaphragms rupture to prevent pressurizing the hotwell to >5#.

Incorrect original answer, changed answer to correct one, rearranged distractors, replaced one distractor (#I2 on SRO 7/2001)

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION F-03 REVISION 9 5.2.2 Turbine Bypass The unit is furnished with turbine bypass valves which permit the unit to operate at low steam flow rates. The bypass valves serve four (4) functions:

a. Provide pressure control at low loads before the turbine is capable of accepting pressure CantroI. (Start-up condition) (10 psi)

b. Provide a normal control range at operating conditions- Bypass valve setpoint is raised 75 psi.

c. Provide a hdependent high pressure relief that will operate proportionally to limit steam generator pressure (above 1040 psi), if in auto.

d. Provide pressure control after a reactor trip, at 125 psi above normal setpoint, to prevent excessive cooling of the reactor coolant fluid.

During the normal plant startup, the reactor and steam generator will be producing 10 to 15%

steam flow through the bypass valves before the turbine is rolled and synchronized. When synchronization is accomplished, an initial load of 5% is picked up by the turbine-generator.

The turbine is then manually increased until it is carrying 15% load and then placed in automatic. As soon as the turbine control station is placed on automatic, it will be controlling the pressure. At 15% load, the bypass valves pressure setpoint control will be biased up to setpoint plus 75 psi. Lfthe bypass valves were not biased to a higher setpoint, there would be two systemstrying to control pressure which could lead to undesirable interaction between the two systems and thus give unstable pressure controI. The bypass valves then serve as overpressure relief valves operating at the higher setpoint. If the reactor is tripped, the bypass valves pressure setpoint will be firther biased to a new level of setpoint ptus 125 psi.

An overpressurerelief setpoint is developed and also applied to the bypass valves. This control operates fiom comparing individual steam-generatorpressure to a setpoint which is higher than the normal bypass valve setpoint. The pressure error signal developed is applied to the bypass valves via high selecting auctioneers located downstream of a11 control components. (See plant specific section for details). This overpressure relief will cause the bypass valves to open at 1040psig if in auto.

Additional overpressure relief is provided by the atmospheric exhaust valves.

During a reactor tr;p, turbine bypass pressure control is supplemented by modufating the ahnospheric exhaust valves when the OTSG pressure ezeeds 1026 psig.

For the turbine bypass valves to be functional, the main turbine condenser must be available.

of cooting water (less than 2 CW-PS arc4 7 Hg absolute), the turbine bypass valves td valves would be utilized for ovqressure relief untii the condenser was again available.

Turbine bypass valve capacity - 6 = 22%, 4 = 15% steani flow Atmospheric dump valves - 2 = 6.4% stm flow 31

SECTION H-4 REVISION 3 If a leak develops i n one o f the tubes containing circ. water a change i n hotwell conductivl t y wi 11 result. The condenser circ. water sides (tube sides) are designed t o allow shutting down half of the tubes a t a time t o allow repair o f such leaks.

During this time condenser vacuum may decrease due t o excess heat load f o r the available circ. water capacity.

4.1.3 Emergency Operation The condensers are not needed during any plant emergency t o affect safe shutdown o f the plant. If they a r e available, t h e i r sole function is as a heat s i n k for waste steam energy from the OTSGs v i a the steam bypass system.

Z f during power operation the entire circ. water system i s l o s t the steam f l o w into the condensers may increase pressure in the 11s t o as high as 5 pstg.

phragms on t o p of the LP T densers w i l l blow open t o protect the co

. overpressure damage.

4.2 Components 4.2.1 Main Condensers (CO-C-1)

The U n i t 1 main condenser i s an I-R model 195-RET-45 classified as tandem t w i n shell, three pressure level, 678,000 square foot, single pass, vertically divided surface condenser. See Figure 1 attached f o r general physical layout o f internals, See Table 1 for specifications.

The CW inlet water boxes are located on the north wall o f the condenser. Two banks o f tubes run from the i n l e t water boxes t o the condenser inboard water box north wall. Two more banks

. - of tubes then run from the inboard water boxes south wall t o the condenser outlet water boxes located on the condenser south wall. Circulating water enters the tubes (see Figure 2) through the inlet water boxes and flows through the i n s i d e of the tubes t o the inboard water box. The water then flows into the inboard water box and from there into the tubes between the inboard water box and the outlet water boxes. From there the water flows through the tubes and discharges from the condenser through the outlet water boxes. This arrangement provide for two independent loops on the tube (CW) side so one can be shutdown f o r leak repair while the other loop remains i n operation.

All tube sheets are epoxy coated on t h e CW side to minimize corrosion o f the carbon steel sheets. Magnesium anodes are installed i n a l l water boxes t o further reduce corrosive attack on the carbon steel tube sheets by counteracting the galvanic cell created by the SST tubes and CS sheet.

Page t# 4.1-14

'W ROISRO importance Rating 2.7 -

3.4 Group # -1 Knowledge of the reasons for the following responses as they apply to Loss of Offsite and Onsite Power (Station Blackout): Length of time for which battery capacity is designed Identify the ONE statement below that describes the FSAR basis for the length of time for which the Station Batteries are designed during a station blackout (loss of ALL offsite power, with NO operable emergency diesel generators).

A. Maintain RCP lift oil pressure during pump coastdown.

B. Provide for turbine-generator bearing oil flow during shaft coastdown.

C. Provide emergency lighting in vital areas to support performance of time critical tasks.

D. Support remote control of emergency feedwater flow to the OTSGs during plant cooldown.

FSAR section 8.2.2.6 pg 8.7-7 bottom/ section 7.1.4.1 pg 7.1-26 bottom/ section 14.1.2.8.4 pg. 14.1-26a top June 1969 1.E.E.E report no. NSG/TCS/SC4-1 OPM section A-03 pg 5.

&l?j New E TMlBank TMI Question #

0 Modified TMI Bank Parent Question #

u L3 Memory or Fundamental Knowledge 0 Comprehension or Analysis 55.41 .5/.10 c 55.43 .1 3l 55.45 .6/.13 A Incorrect answer. DC does supply RCP lift oil pump power, but this is not a basis for battery time capacity.

Distracter is plausible because of the DC Oil Lift Pumps associated with the RCPs.

B Incorrect answer. Turbine coast down is not basis for battery time capacity. Distracter is plausible because of the DC Lube Oilt Pump associated with the Main Turbine.

C Incorrect answer. Distracter is plausible since there are emergency DC lights required in vital areas, and there are time critical actions.

D Correct answer. FSAR has restart commitment of minimum 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for EFW control using DC power.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

TMI-1/FSAR

7) The turbine-driven emergency feed pump takes suction from the condensate storage tanks and is driven by steam from either or both steam generators. The emergency feedwater system provides feedwater for decay heat removal and is discussed in Section 10.6. The controls and auxiliary systems for the emergency feed pum power from the battery backed DC bus for a minimum of two ttle was added to the air supply for the EFW turbine throttle valve (MS-V-6) to assure a heat sink for a minimum of two hours, under loss of AC power operation.

In view of the above sequence, the loss of all unit AC power (Station Blackout) does not result in any fuel damage or excessive pressures on the RCS. There is no resultant radiological hazard to operating personnel or to the public from this accident as only secondary system steam is discharged to the atmosphere.

14.1.2.8.5 Radiological Consequences of SBO Event The radiological consequences associated with the Station Blackout Event were evaluated at design RCS flow conditions using the design inputs specified in Reference

93. For this event, it is assumed that the plant has been operating with both 1 percent failed fuel and steam generator tube leakage of 1 gpm. The operation has been followed until the steam generator that does not have tube leakage can remove decay heat and the atmospheric dump valve associated with the leaking generator is closed.

This results in the following sequence:

1) The atmospheric dump valves are assumed available for this case. It is assumed that the operator further opens the atmospheric dump valves 10 minutes after the loss of power to cooldown the plant at the maximum allowable rate.

2) Cooling down at the maximum available rate requires an additional 45 minutes to reach a temperature below the saturation temperature corresponding to the set point pressure for the steam safety valve having the lowest setting.

3) The steam generator with tube leakage is then completely isolated by closing its atmospheric dump valve, and the other steam generator is used to remove decay heat.

As in the loss of load transient evaluation, the whole body dose does not change due to steam relief. The total integrated thyroid dose is shown in Table 14.1 -1 5 (Sheet 2).

14.1-26a UPDATE-15 4/00

I TMI-1/FSAR A manual bypass reset switch is also provided for increased flexibility during testing and/or operation.

The trip functions of the Reactor Building isolation and cooling can only be bypassed after actuation (see Item c.3 of Section 7.1.3.2).

The overriding of any one type of isolation signal (radiation, reactor trips, high Reactor Building pressure) to a containment isolation valve will not block any of the other signals from performing their isolation function.

g- Reset The ESAS trip bistables are adjusted for a minimum deadband, thereby allowing automatic reset when the monitored variable goes below or above its predetermined setpoint. After a trip has occurred and the bistable has reset, the operator can reset the logic remotely by pressing a Control Room console mounted reset switch. Thus, the operator is not required to leave the Control Room to reset the bistable. Resetting the trip bistable and the logic does not reset ESAS-actuated equipment.

Remote reset capability is provided for all of the channels in high pressure injection, low pressure injection, and Reactor Building isolation actuation "A",

and for all channels in high pressure injection, low pressure injection, and Reactor Building isolation actuation "B".

ESAS logic cannot be reset remotely if the trip signal is present.

7.1.4 EMERGENCY FEEDWATER SYSTEM 7.1.4.1 Desian Basis The TMI-1 Emergency Feedwater System (EFW) is required to remove heat from the primary system when the main feedwater system is not available. It is capable of holding the plant at hot standby and also capable of cooling down the plant to the point where the normal decay heat removal system can operate.

The EFW system is configured to insure the addition of EFW to the OTSGs assuming a single active failure concurrent with loss of offsite power.?$n addition, the modified +

system is capable of providing controlled emergency feedwater flow to the OTSGs for at least two hours without relying on alternating current (AC) power (Station Blackout).

The two hour analysis is based on a TMI-1 restart commitment: The TMI-1 Station Blackout (SBO) specified duration, however, is four hours. See Section 8.5 for the Station Blackout evaluation .

7.1-26 UPDATE-I 5 4/00

SECTION A-03 REVISION 12 FIGURE 1 - DC ELECTRICAL DISTRIBUTION SYSTEM 5.0 COMPONENTS 5.1 Storage Battery Banks 5.1.1 Purpose and General Description The heart of the plant DC EIectricaI Distribution System is the storage battery. It provides power on a loss of battery chargers and acts as a surge power supply when the chargers become overloaded.

The capacity of the two banks of batteries installed in Unit 1 is fixed to ensure that on a loss of all AC power there will be enough DC power for about eight hours, assuming that unnecessary loads are secured when no longer needed. The capacity of each badc is sufficient

\- to power its connected essential l a d for two hours continuously and P: perform three complete cycles of safe_rmardsbreaker cIosures and subsequent tripping.

5

1. 1000-PLN-7200.01, TMI-1 Operational Quality Assurance Plan.

2. Deleted

3. G/C Report: Appendix R Safe Shutdown Equipment and Circuit Evaluation Summary Report, August 15, 1985.

4. IEEE Report No. NSG/TCS/SC4-?,entitled: "Proposed IEEE Criteria for Class ?E ?

Electrical Systems for Nuclear Power Generating Stations," dated June, 1969.

5. GPU Nuclear Report 990-1879: TMI-I Station Blackout Evaluation.

6. C-I 10-l-700-E510-010, "TMI-1 AC Voltage Regulation Study."

7. Proposed AEC Criteria 11, 24, and 39, July 11, 1967

8. Reg Guide 1.89, Qualification of Class 1E Equipment for Nuclear Power Plants.

9. Reg Guide 1.100, Seismic Qualification of Electric Equipment for Nuclear Power Plants.

IO. Reg Guide 1.108, Periodic Testing of Diesel Generator Units Used as Onsite W

Electric Power Systems at Nuclear Power Plants.

11. IPCEA(ICEA) Standard P-46-426, 1962, Power Cable Ampacities Vol 1 Copper Conductors, reprinted as IEEE S-135, 1984.

12. NUMARC 87-00, Guidelines and Technical Basis for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors.

13. NSAC 108 The Reliability of Emergency Diesel Generators at US Nuclear Power Plants.

14. IEEE 279, Proposed Criteria for Nuclear Power Plant Protection Systems, August 1968.

15. IEEE 383-1974, Standard for Type Testing of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations.

16. NUREG 0737, Post TMI Requirements.

17. C-I 101-741-E510-005, Loading Summary of Emergency Diesel Generator and Engineered Safeguard Buses. I L- 8.6-1 UPDATE-15 4/00

Drawings B-201043, 8-201052, 8-201069, 8-201 044, B-201053, B-201062 and B-201063 list the electrical loads connected to the Engineered Safeguards 480 V Control Centers.

8.2.2.6 250f125 Vdc Svstem The 250/125 Vdc system provides a source of reliable continuous power for DC pump motors, control, and instrumentation. In general, DC motors are rated 240 Vdc and control circuits 125 Vdc.

The 250/125 Vdc system consists of two isolated bus sections, each supplied by a battery and battery chargers. A spare 125 Vdc battery charger is provided for each battery for backup.

The arrangement and number of batteries, chargers, and dc distribution panel boards are as shown on Drawing E-206-051. The output of spare battery chargers may be fed to either half of the corresponding 250/125 Vdc system, and provision has been made for manual cross-connection of the two systems during battery discharge tests. By this means, all battery chargers would be available for feeding the essential loads. The manually operated bus tie is protected on both ends by normally locked open fused switches. The fuses are removed and the switches are locked open on the DC tie in the 230 kV substation.

Each battery charger has its own input and output protective circuit breakers. Each battery charger is connected to its associated distribution bus through fused disconnect switches. The battery chargers utilize silicon controlled rectifiers (SCRs) and, thus, are inherently protected against becoming a load on the DC bus during AC power outages.

As shown on Drawing E-206-051, under plant operating conditions there are no DC ties between redundant engineered safeguards equipment, switchgear, motors, and so forth, and, therefore, no single failure of any DC component can adversely affect the operation of the 100 percent redundant diesel generators. The entire system satisfies the IEEE criteria given in Reference 4, Section 8.6.

The capacity of each of the two redundant batteries is sufficient to feed its connected essential load for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months continuously and perform three complete cycles of safeguard breaker closures and subsequent tripping. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rating is based on the time -

required to ensure that all nuclear and BOP emergency equipment can performjts intended function and on the criteria contained in Draft 3 of Reference 4, the IEEE criteria for Class 1 E electrical systems.

8.2-7 UPDATE-12 3/94

TM 1-1/FSAR Each battery has been sized to have no less than the ratings given below, based on the use of 116 cell batteries and discharge to 1.81 Vdc per cell, at 77F 210/105 Vdc across the battery:

Discharge AMP-Hour Reserve time Rate AMPS Capacity 8 Hours 177.5 1420 3 Hours 370 1110 2 Hours 475 950 1 Hour 665 665 IMinutes 1160 -

Each battery charger has been sized at 150 amperes continuous rating. This would allow a battery to be fully recharged in less than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , with the normal load of the battery system being carried simultaneously. For design basis loadings on each of the Station Battery systems A and B refer to Calculation C-I 101-734-5350-003 (Reference 25).

Batteries are sized in accordance with IEEE-485-1983 method, and load profile is determined for each battery. The sizing includes an ambient temperature correction and an aging factor. Because of space limitations in the battery rooms, the A battery is slightly undersized. The size limits the usable life of the battery. The acceptance criteria in the Station Battery Load surveillance is specified so that the battery will have the required capacity and temperature margin at the end of the surveillance interval.

The following alarms are provided in the Main Control Room for the DC power supply system:

Alarm Actuated By Battery 1A Ground Station battery 1A ground detector Battery I 6 Ground Station battery 1B ground detector Alarm Activated By 1A Battery Charger I A battery chargers power failure System Trouble I A battery chargers DC volts low/high IB Battery Charger 1B battery chargers power failure System Trouble 18 battery chargers DC volts lowlhigh Battery Discharging Either battery current high or battery voltage low on either battery section Substation Panels DC-A and DC-B have been provided with a backup DC power supply. The batteries are rated 200 ampere hours and are located in the 230 kV switchyard. In the unlikely event of loss of station DC supply, power to the substation panels DC-A and DC-B can be turned on manually from the Substation Battery.

L-A single line diagram of the DC system showing essential loads is given on Drawing E-206-051 8.2-8 UPDATE-15 4/00

Q#-026

~- SYSIEP# Q6J RO/SRO Importance Rating KA#

3.5 AA2.01

-3.7 Page # 4.2-48 Tier #

Group #

Ability to determine and interpret the following as they apply to Area Radiation Monitoring (ARM) System Alarms: ARM panel displays Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Pump down of Reactor Coolant Drain Tank (RCDT) in progress.

- Radiation monitor RM-G-20 (RCDT) high alarm actuates unexpectedly.

Based on this occurrence, from the list below identify the ONE set of procedures that is required to be implemented.

A. EP 1202-29, Pressurizer System Failure AbP 1203-15, Loss of RC MakeuplSeal Injection B. EP 1202-11, High RCS Activity EP 1202-12, Excessive Radiation Levels AbP 1203-15, Loss of RC MakeuplSeal Injection C. EP 1202-12, Excessive Radiation Levels EP 1202-29, Pressurizer System Failure AbP 1203-15, Loss of RC MakeuplSeal Injection D. EP 1202-11, High RCS Activity EP 1202-12, Excessive Radiation Levels EP 1202-29, Pressurizer System Failure u MAP C-1-1, Radiation Level Hi (RM-G-20), page 36, Rev. 27.

None.

EP Entry Conditions I? New a TMIBank TMI Question # #34 7l2001 SRO 3 Modified TMI Bank Parent Question #

El Memory or Fundamental Knowledge

!GI Comprehension or Analysis 3 55.41 3 55.43 .5 2 55.45 . I 3 A Incorrect answer. Although Pressurizer system failure can apply, there is no indication of loss of makeup or RCP seal injection listed.

B Incorrect answer. Although 1202-12 and 1202-11 apply (see D), no loss of makeup has occurred.

C Incorrect Answer. Two correct EPs, and one incorrect, (See A and D).

D Correct answer. All listed EPs apply: since the source of water is RCS, High activity EP applies, since there is an RMS alarm, Excessive Rad Levels EP applies, and since RCS water can only get to RCDT inadvertently

- thru PZR leak, and no other answer distractor applies, PZR system failure EP does apply.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1

\-1 Alarm Response Procedure MAP C Title Revision No.

Main Annunciator Panel C (See Cover Page)

ALARM:

RM-G-20 RC DRAIN TANK SET POINTS:

Refer to Operating Procedure 1101-2.1 RMS setpoints.

CAUSES:

Fuel damage (Hi Activity in RCS) coupled with RCS leakage to drain tank.

AUTOMATIC ACTION:

Closes WDL-V-303, WDL-V-304, WDG-V-3, WDG-V-4.

b' OBSERVATION (CONTROL ROOM):

1. RM-G-20 "Alert" Alarm on PRF.

2. RM-G-20 "Hi" Alarm on PRF.

3. RM-G-20 Indication on PRF > setpoints.

MANUAL ACTION REQUIRED:

1. Verify WDL-V-303, WDL-V-304, WDG-V-3, WDG-V-4 close.

2. Refer to Emergency Procedure 1202-11 (Hi RCS Activity); Emergency Procedure 1202-29 (Pressurizer System Failure).

3. Refer to EP 1202-12, Excessive Radiation Levels.

Page 36 of 50

Form ES-401-6 SYS/EP# Q6J KA# 2.1.32 Page # 2-4 Tier # -1 u 3.8 Group # -2 ROISRO Importance Rating 3.4 -

Ability to explain and apply all system limits and precautions: Area Radiation Monitoring (ARM)

System Alarms A warning is posted locally at RM-A-2 (Reactor Building Atmospheric Monitor) particulate housing to Contact the Control Room prior to opening. The reason for this is to ensure...

A. RM-A-2 pump is shutdown and isolated to prevent damage to the pump diaphragm.

B. RM-A-2 pump is shutdown and isolated to prevent inadvertent breach of containment integrity.

C. shift management notifies Work Management before initiating work.

D. shift personnel mark the control room charts due to potential radioactive particulate release in the Intermediate Building.

OP 1105-8, Radiation Monitoring System, page 89, Rev. 67.

New Ci TMlBank TMI Question #

Ll Modified TMI Bank Parent Question #

&? I Memory or Fundamental Knowledge 0 Comprehension or Analysis d

d 55.41 .10 @ 55.43 .2 M 55.45 . I 2 A Incorrect answer, although running the pump with no suction for more than a few minutes does damage the diaphragm, this listed action will not.

B Correct answer. Containment integrity will be breached if CM-V-1/4 are not first closed, which is the reason for the warning.

C Incorrect answer. Although Work Management schedules tasks, on shift personnel have overall responsibility for Technical Specifications compliance.

D Incorrect answer, although plausible under severe accident conditions, and marking charts is common practice.

None.

TMI SRO Exam - May 2003 Friduy, March 28,2003

Number

- Title I

TMI - Unit 1 Operating Procedure I105-8 RevisionNo.

Radiation Monitoring System 67 DO NOT DISTURB Containment Isolation Boundary CONTACT CONTROL ROOM PRIOR TO OPENING 89

u I ROlSRO importance Rating 3.4 -

3.8 Group #

Ability to explain and apply all system limits and precautions: Plant Fire on Site a 'c:

A fire in the relay room has occurred. Under this condition, identify the ONE reason below for an operator to depress and hold the C02 Discharge Delay Switch.

A. To prevent undesirable component actuation by allowing fire brigade personnel to combat fire quickly without needing to don breathing apparatus.

B. To prevent undesirable component actuation caused by C02-induced cold room temperatures before the plant is shutdown.

C. To await personnel working in the relay room to evacuate before an oxygen deficient atmosphere is created.

D. To allow the fire brigade leader to assess the fire condition before an oxygen deficient atmosphere is created.

1104-45F, C02 Fire Extinguishing System for 338' Elev. Relay Room, section 3.1.2, pages 5 and 6, Rev. 18.

New 'L TMI Bank TMI Question #

- Modified TMI Bank Parent Question #

2 Memory or Fundamental Knowledge L - Comprehension or Analysis

< 55.41 .IO

- 55.43 v 55.45 .I2 A Incorrect answer - although plausible, this is not a protocol for fire fighting.

B Incorrect answer - although plausible since the rapid cooling can cause component control to be erratic.

C Correct answer - this is a personnel safety design feature as described in fire system OP limits.

D Incorrect answer - although plausible, this is not a protocol for fire fighting.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Q -43 y Number TMI -Unit 1 0perati ng Procedure 110445F Title Revision No.

COn Fire Extinguishing System for 338' Elev. Relay Room 18 The vapor pilot valve (FS-V-341) is locked in the open position.

(Located above the 3-way switching valve inside the east cover).

The C 0 2 storage tank level indicated 2 8.75 ft.

The C 0 2 storage tank pressure indicates 2 285 psig.

5. FD-21 and 22 are open. (North wall of Relay Room)

-IF not open, THEN refer to 1303-12.5 for directions to set dampers.

3.1.2 Procedure 3.1.2.1 An Operator shall periodically check the following during normal operation:

CO2 Storage Tank pressure is greater than or equal to 285 psig C 0 2 Storage Tank level is greater than or equal to 8500 Ibm (85%)

Proper operation of the compressor System leakage as indicated by abnormal tank level changes The three-way switching valve is fully back-seated in the counter-clock-wise direction.

I WARN1NG I

I Remaining in the Relay Room after a discharge of C 0 2 has occurred will result in personal injury or death due to the displacement of oxygen by I 3.1.2.2 IF personnel are working in the Relay Room and are unable to exit the room within the 30 second delay associated with the discharge alarm, THEN Ensure an individual is present to perform the following with No other concurrent duties:

1. Standby the C02 Discharge Delay Switch located on the Relay Room north wall, west end at the entrance to ESAS Room.

5

Number TMI -Unit 1

'L/ Operating Procedure I 104-45F Title Revision No.

C02 Fire Extinguishing System for 338' Elev. Relay Room 18 THEN Depress AND Hold the delay switch button until all personnel have exited the Relay Room and ESAS Room

3. WHEN all personnel have exited the Relay Room and ESAS Room,

-THEN Perform the following:

(1) Release the delay switch button to restart the 30 timer.

(2)' Immediately Exit the Relay Room AND Close the door.

(3 ) Exit the Relay Room and ESAS Room AND Wait for the Fire Brigade personnel at the Control Building Stairwell.

3.2 Automatic (Automatic Actuation) Level 3 NOTE Automatic operation requires NO operator action, however, as a result of 1

the system actuation, the Control Room will receive alarm PRF 3-5. A C 0 2 system trouble will be received (PRF 3-4) when COz pressure exceeds high or low setpoints.

3.2.1 Prerequisites

1. System is in normal operation per Section 3.1.

3.2.2 Procedure 3.2.2.1 Actuation of the C 0 2 system occurs, THEN Perform the following:

1. Follow alarm response procedures for the specific alarm and fire pre-plans.

2. After the fire has been extinguished, Proceed to Section 3.4 for actions required after C 0 2 actuation.

3.3 Manual - Level 2 3.3.1 Prerequisites L -

1. System is in normal operation per Section 3.1.

6

Form ES-401-6 KA# AA2.01 Page # 4.2-57 Tier # -

1 i/ ROlSRO Importance Rating 3.7 -

4.3 Group # -1 Ability to determine and interpret the following as they apply to Loss of containment integrity:

loss of containment integrity.

I? a Plant is in refueling shutdown mode. Identify the ONE situation below that represents a condition that does NOT meet requirements for containment closure control during handling of irradiated fuel.

A. Service Air is in use inside the RB.

B. TWO RB purge exhaust valves are stuck at 30% open.

C. Containment Building nitrogen supply spool piece is installed.

D. BOTH air lock doors of the RB personnel hatch are open. An Operator is responsible for closing ONE door.

Technical Specifications, Section 3.8.7, Amendment 236. Pg 3-44.

IV.D.38.08 New fl TMIBank TMI Question #

E Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge Comprehensionor Analysis u w 55.43 .7 @ 55.45 . I 3 A Incorrect answer - although plausible misconception that this violates containment integrity requirements since hoses are run through penetrations with temporary seals.

B Correct answer - need to be closed or flanged to comply with TS.

C Incorrect answer - N2 is spooled and a closed system as related to low pressure containment integrity.

D Incorrect answer - both doors are allowed to be open, with closure capability per TS 3.8.7 Rev 236.

TMI SRO Exam - May 2003 Friday, March 28,2003

3.8 FUEL LOADING AND REFUELING Applicability: Applies to fuel loading and refueling operations.

.- Objective: To assure that fuel loading and refueling operations are performed in a responsibIe manner.

S w ification 3.8. I Radiation levels in the Reactor Building refueling area shall be monitored by RM-G6 and RM-G7.Radiation levels in the spent fuel storage area shall be monitored by RM-G9. If any of these instruments become inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refueling operation, shall be used until the permanent instrumentation is returned to service.

3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with continuous indication available, whenever core geometry is being changed. W h e n core geometry is not being changed, at least one neutron flux monitor shall be in service.

3.8.3 At least one decay heat removal pump and cooler shall be operable.

3.8.4 During reactor vessel head removaI and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at not less than that required for refueling shutdown.

3.8.5 Direct communications between the control room and the refueling personnel in the Reactor Building shall exist whenever changes in core geometry are taking place.

3.8.6 During the handling of irradiated fuel in the Reactor Building at least one door in each of the u personnel and emergency air locks shall be capable of being closed.* The equipment hatch cover shall be in place with ;1 minimum of four bolts securing the cover to the sealing surfaces.

3.8.7 During the handling of irrddiated fuel in the Reactor Building, each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

I. Closed by an isolation valve. blind flange, manual valve, or equivalent, or capable of being closed,* or

2. Be capable of being closed by an operable automatic containment purge and exhaust isolation valve.

Administrative controls shall ensure that appropriate personnel are aware that air lock doors and/or other penetrations are open, a specific individual(s) is designated and available to close the air lock doors and other penetrations as part of a required evacuation of containment. Any obstruction(s) (e.g.,'

cable and hoses) that could prevent closure of an air lock door or other penetration will be capabte of being quickly removed.

3-44 Amendment No, M , 236

Page t# 4.2-57 Tier # 1 u ROISRO Importance Rating 2.6 -3.1 Group # 1 Knowledge of the operational implications of the following concepts as they apply to Loss of containment integrity: Effect of pressure on leak rate.

Plant conditions:

- Reactor tripped due to loss of off-site power (LOOP).

- MS-V-4NB ADVs are controlling OTSG pressures at 500 psig.

- OTSG 1A tube leak rate = 5 gpm.

- In accordance with procedural guidance, RCS pressure is being reduced from 900 psig to 800 psig.

Based on these conditions, identify the ONE selection below that completes the following phrase:

The operational implication of this action to lower RCS pressure is to reduce the.. .

A. OTSG level rise.

B. off-site radioactive release rate.

C. time required for cooldown of the RCS.

D. potential of lifting Main Steam safety valves.

Emergency Operating Procedures Technical Bases Document, Volume 2, Chapter Ill,page IIIE-10, Rev. 9.

b PI New 3 TMIBank TMI Question #

0 Modified TMI Bank Parent Question #

- Memory or Fundamental Knowledge PI Comprehension or Analysis 37 55.41 .7 1 55.43 3 55.45 .7 A Incorrect answer. Plausible since reduction in delta-P will Primary-to-Secondary leak rate. This is a good answer for a larger OTSG tube leak.

B Correct answer. Lowering differential pressure across a leaking OTSG tube will lower leakage (mass flow rate) and curie transport (radioactive release to the environment) out the ADVs.

C Incorrect answer. This effect is achieved by steaming both OTSGs.

D Incorrect answer. As described in the stem, RCS pressure is already below the lowest MSSV setpoint.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

I NUMBER I TECHNICAL DOCUMENT 74-1 152414-09 7.0 'REDUCE RCS PRESSURE TO MAINTAIN MINIMUM SCW, AND IF APPLICABLE RCP NPSH, WITHOUT EXCEEDING [pressurizer emergency cooldown rate] (Rule 3.0).

Indicators and Controls Indicators - RCS pressure

- RCS temperature (incore thermocouples)

- SCh4 monitor

- P-T display

- RCP status

- SPDS

- Spray valve position

- Auxiliary high pressure spray valve indication

- Pressurizer vent valve position

- PORV position Controls: - Spray valve controls

- Auxiliary high pressure spray valve controls

- Pressurizer vent valve controls

- PORV controls

-2urpose of Step The purpose of this step is to reduce RCS pressure as much as possible while still maintaining SCM and RCP NPSH if applicable.

Bases During the cooldown it is desirable to maintain RCS pressure and temperature close to, but above, the minimum SCM. This minimizes the differential pressure between the RCS and the affected SG(s), thus minimizing the tube leak flow ra@!'If normal spray is not available (e.g., RCPs off) to reduce RC pressure, then high pressure auxiliary spray, if available, is used. In the absence of high pressure auxiliary spray, then the PORV or pressurizer vent is employed to decrease RCS pressure.

The RCS depressurization will require a pressurizer cooldown rate greater than the normal limit, but the cooldown rate should be kept within the [pressurizer emergency cooldown rate] for rapid depressurization, if possible.

If PTS guidance has been invoked, then the RCS pressure and cooldown rate must be maintained per PTS guidance.

The plant-specific control value for RCP NPSH does not require error correction.

le plant-specific value for [pressurizer emergency cooldown rate] is intended as a control

.,,arameter and therefore does not require error correction.

7 313 1/2000 r V01.2,III.E-lO

SYSIEP# 974 KA# EA2.06 Page # 4.1-17 Tier # -1 u

RO/SRO Importance Rating 4.0 -

4.6 Group # 1 Ability to determine and interpret the following as they apply to Inadequate Core Cooling:

Changes in PZR level due to PZR steam bubble transfer to the RCS during inadequate core cooling.

Initial plant conditions:

- Reactor tripped from 100% power due to loss of off-site power (LOOP).

- Emergency Feedwater Pump EF-P-2B tripped.

- One Makeup Pump operating.

- Pressurizer level = I00 inches, controlled in automatic.

- Core exit thermocouple temperature is steady at 570°F.

- RCS pressure steady at 2100 psig.

- OTSG pressures = 1000 psig.

- OTSG levels at 12% Operating Range, slowly rising.

- OP-TM-EOP-001, Reactor Trip, ImmediateActions complete.

- Initial post trip Symptom Check has been completed.

Sequence of events:

- PORV opened unexpectedly, and failed to reclose.

- RC-V-2 control power fuse failed during attempt to close PORV block.

- RCS pressure rapidly reduced to 1680 psig, and now is slowly lowering to 1660 psig.

- At the end of the pressure reduction, Pressurizer level rose rapidly to 300 inches, and is now rising slowly.

'u HPI has NOT been actuated at this time. Based on these conditions, identify the ONE set of statements below that describes ( I ) the reason for the Pressurizer insurge and (2) required actions.

A. (1) Displacement of water from under the RV head due to steam bubble formation.

(2) Continue with EOP-001 VSSVs.

6. (1) Expansion of RCS loop water due to depressurization.

(2) Exit EOP-001 and GO TO OP-TM-EOP-009 HPI Cooling - Recovery From Solid Operations.

C. (1) EFW capacity is insufficient to remove decay heat.

(2) Feed the OTSGs with Reactor River Water, using EOP-010 Guide 17, Alternate Inventory for Emergency Feedwater.

D. (1) RCS water in the Hot Legs is saturated.

(2) Implement EOP-010 Rule 1, SCM.

Analysis of Three Mile island - Unit 2 Accident OP-TM-EOP-001, Reactor Trip, page 3, Rev. 3.

None.

I II .c.07.04 W New G TMl Bank TMI Question #

0 Modified TMI Bank Parent Question #

E Memory or Fundamental Knowledge L

E Comprehension or Analysis r-' 55.41 k?! 55.43 .5 E 55.45 . I 3 TMI SRO Exam - May 2003 Friduy, March 28,2003

A Correct answer. Depressurization has resulted in formation of a steam bubble due to the hot metal W temperature and absence of forced flow under then head. The water displacement is manifested in a Pressurizer insurge. Since the RCS is still subcooled and adequate heat transfer exists, no symptom based criteria exist to exit EOP-001. As a side note, EOP-001 follow-up actions will direct the operators to GO TO EOP-006, LOCA Cooldown.

B Incorrect answer. At these conditions, coolant flashing to steam will occur only under the head due to low flow and high metal tempersture. Exiting EOP-009 would be incorrect at this time.

C Incorrect answer. Capacity of EF-P-1 with EF-P-2A is sufficient to remove deacay heat. Feeding OTSGs with river water at these conditions would be incorrect.

D Incorrect answer. Based on core exit thermocouple temperature, loops are subcooled >25"F, and therefore implementation of Rule 1 is not required.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Analysis of Three Mile Island - Unit 2 Accident NSAC-80-1 NSAC-1 Revised March 1980 Prepared by the Nuclear Safety Analysis Center Operated by the Electric Power Research Institute 3412 Hiflview Avenue .

P.O. Box 10412 Palo Alto, California 94303

Without t h e steam g e n e r a t o r s as an e f f e c t i v e h e a t s i n k , t h e primary system t e m p e r a t u r e continued t o i n c r e a s e . T h e thermal expansion of t h e system i n v e n t o r y d u e t o t h i s h e a t u p , t o g e t h e r with o p e r a t o r a c t i o n s t o provide i n c r e a s e d a m o u n t s of makeup w a t e r t o t h e primary system from t h e makeup t a n k s and t o reduce letdown flow from t h e system, r e s u l t e d i n r e v e r s i n g t h e p r e s -

s u r i z e r w a t e r l e v e l decrease. The stuck-open r e l i e f v a l v e , how-e v e r , c o n t i n u e d t o d i s c h a r g e c o o l a n t from t h e primary system, t h u s causing i t t o depressurize. This continued d e p r e s s u r i z a -

t i o n , c o i n c i d e n t with i n c r e a s i n g p r e s s u r i z e r c o o l a n t l e v e l , i s a s t r o n g i n d i c a t o r of t h e abnormal n a t u r e of t h e e v e n t .

Assuming t h a t t h e c o o l a n t volume i n t h e r e a c t o r v e s s e l dome i s a t t h e h o t l e g t e m p e r a t u r e ( 6 0 5 ° F ) b e f o r e the t r a n s i e n t , i t s s a t u r a -

t i o n p r e s s u r e is approximately 1600 p s i a . This p r e s s u r e was reached a b o u t t w o m i n u t e s a f t e r t h e i n i t i a t i n g e v e n t . If this volume s t a g n a t e s due t o flow r e s t r i c t i o n by the plenum c o v e r which s e p a r a t e s t h e plenum from t h e head, subsequent f l a s h i n g might occur d u r i n g t h e d e p r e s s u r i z a t i o n . However, when f o u r

-d r e a c t o r c o o l a n t pumps a r e o p e r a t i n g , normal flow a c r o s s t h e c o n t r o l rod g u i d e t u b e s , through t h e upper head and i n t o t h e o u t l e t annulus i s s u f f i c i e n t t o r e p l a c e one upper head volume i n a p p r o x i m a t e l y seven seconds. Thus, f l u i d i n t h e upper head is n o t s t a g n a n t , and f l a s h i n g would n o t n e c e s s a r i l y o c c u r . In fact, d u r i n g t h e f i r s t t w o minutes of the t r a n s i e n t , which i n c l u d e s t h e d e p r e s s u r i z a t i o n t o 1600 p s i a , no evidence of f l a s h i n g i s r e f l e c t e d i n t h e r a t e of p r e s s u r i z e r water l e v e l change. That i s , l e v e l change is e x p l i c a b l e i n terms of primary system t e m -

p e r a t u r e changes and makeup flow a l o n e .

A f t e r a b o u t t w o m i n u t e s , t h e primary system p r e s s u r e had d e c r e a s e d t o t h e Emergency Safeguards Actuation s e t p o i n t , and t h e emergency h i g h p r e s s u r e i n j e c t i o n system w a s a u t o m a t i c a l l y started. F u l l i n j e c t i o n flow w a s maintained f o r approximately t w o and a h a l f minutes. A c a l c u l a t i o n of p r e s s u r i z e r c o o l a n t

&eve1 was performed f o r t h i s p e r i o d . The r e s u l t s confirmed t h e indicated p r e s s u r i z e r coolant l e v e l .

A P P E N D I X TH 29

Page # 4.3-27 Tier # -1 L./,

ROISRO importance Rating 3.7 3.7

- Group # -2 Ability to operate and/or monitor the following as they apply to Plant Runback: Components, and functions of control and safety systems, including instrumentation, signals, failure modes, and automatic and manual features.

Initial conditions:

- Reactor power is loo%, with ICs in full automatic.

Sequence of events:

- FW-P-1A trip initiated an automatic ICs runback.

- During the runback, Group 7 Rod 6 dropped (fully inserted).

Based on these two concurrent conditions, identify the ONE selection below that identifies ( I ) the rate of change for the ICs runback, and (2) the final power level when the plant stabilizes.

A. (1) 50%/minute.

(2) 68% power.

B. (1) 30%/minute.

(2) 68% power.

C. (1) 50%/minute.

(2) 55% power.

D. (1) 30%/minute.

(2) 55% power.

L OPM Section F-03, Integrated Control System, Pages 24 and 25, Revision I1.

MAP H-1-1, ICs Runback, Rev. 21 None.

IV.E.27.09 E New la TMlBank TMI Question # #36 June 2001 SRO Audit 0 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge M Comprehension or Analysis E 55.43 d 55.45 51.6 A Incorrect answer. This is the original target and rate prior to the rod drop.

B Incorrect answer. This is the original target with the correct rate.

C Incorrect answer. This is the correct target with the wrong rate.

D Correct answer. In accordance with OPM section F-03, runback rate selection is based on lowest of the high load limits.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 i_/

Alarm Response Procedure MAP H Title Revision No.

H-f -1 Revision 21 ALARM:

ICs RUNBACK SET POINTS:

Loss ofone RCP and > 75 percent power Loss of two RCP and > 50 percent power Loss of one main feed DumD and > 60 percent power IOS~RC I .

RC flow limit percent core power 2 flow) 140.00x IO6 Ibs/hr Asymmetric rod 9" out of alignment with its group average rod position and > 60 percent power.

CAUSES:

RCP Trip MFP Trip Dropped Rod RCS Flow Degradation L/

AUTOMATIC ACTION:

Unit runs back to:

75 percent power on a loss of (1) RCP at a rate of 50 percenvmin. = 665 MWe 50 percent power on a loss of (2) RCP at a rate of 50 percent/min. = 405 MWe 6'8 percent power on a loss of (1) MFP at a rate of 50 percenUmin. = 560 MWez The RC flow limit at a rate of 20 percent min.

55 percent power on a Assy. rod fault at a rate of 30 percent/min. w 455 MWe NOTE Runbacks are based on a MWe. Input power may vary due to plant 1

efficiencies.

Page 1 of 2

TMI Unit 1 I Number u Alarm Response Procedure MAP H Title Revision No.

Main Annunciator Panel H (See Cover Page)

H-1-1 Revision 21 OBSERVATION [CONTROL ROOM):

RC flow low as indicated by RC 14 A/B on cc MW generated and neutron power decrease to above limits ICs in track Ann. H-2-1 RCP motor trip alarm on main annunciator 61-1 Feed pump Turb 1A or 16 trip alarm on Main Annunciator M-1-1 and M-1-7 CRD Pattern asymmetrical alarm on Main Annunciator G-2-1 L3126 indicates ULD HAND/AUTO control station is in HAND when in track MANUAL ACTION REQUIRED:

1. Verify the unit runs back to above limits CAUTION If operating in hand decrease at a rate so as not to cause a Rx trip.

a. If any ICs HAND/AUTO control stations are in HAND, decrease that station to within the limits of the runback using the raise-lower switch.

b. I f the Diamond rod control station is in MANUAL, reduce reactor power to within the capabilities of the plant.

C. If the Main Turbine Generator is under LOCAL-OWS control, reduce electrical load to within the capabilities of Main Feedwater and Reactor Power.

d. Monitor imbalance during runback.

2. Review transient cycle logbook and make all required entries.

Page 2 of 2

SECTION F-03 REVISION 10 u FIGURE 8 -UNIT LOAD D E l W VS,TJMX OPERATION UNIT LOAD STATION W I B N INMANUAL t

55 Wanual Set' Targer Laad Rate limited u n l t load atmand (rate sat a t 2% per mln.)

This rate of change can be manually adjusted fiom 0.25% up 1 10% per minute, but : automatically limited to be compatible with the speed ofthe unit. A 5% per minute rate limit is imposed at each end of the load range because the deviations caused by normal over and under pumping and BTU input which would be caused by faster rate changes are not acceptable in these ranges, so the response of the unit is purposely set slower. Overpump and overpower are discussed in Section 4.0.

The rate of change can be adjusted for rates of:

a. 0.25 to 1 0% per minute for all loads between 20 arid 90%.

b. Above 90% load, if an increase is required, the maximum rate is 5% per minute. If a deGTmSe is required above 90% load, 10% per minute is the maximum limit.

C. Below 20% load, 5% per minute is the maximum rate.

d. When a limiting cgndition is received by the Unit Load Demand Subsystem (other than an operator set limit) which requires a load reduction, the rate of change control is taken away fiom the operator. The rate o f change under these conditions is a function of the condition that exists.

I% = 9.25 MW I

The rates for each condition are:

Loss ofreactor coolant pump Loss of reactor coolant flow I CRD Asymmetric Rod Fault Unit in tracking mode 24

SECTION F-03 REVISION 10 The maximum load limit and rate o f change circuits are designed such that if the unit were responding to a load limiting condition and a more severe situation were to occur, the most limiting condition .

would take precedence in establishing the maximum load limit and the rate at which the unit will respond to reduce the load to less than that limit.

4.5 Frequency Regardless of the unit load, the generator output is maintained at a constant frequency. During normal operation, some adjustment may be necessary in maintaining fi-equency. This adjustment is accomplished by governor action. In attempting to maintain constant turbine speed, the governor opens or closes the turbine valves independent of the megawatt or load demand signal. If significant frequency error were to develop, the governor action would Iead to a change in the measured megawatts and unit header pressure. These changes would be reflected to the ICs control system and would result in errors. If iiothing were done, the system would tty to correct these errors, thus tending to counteract or fight the governor action. The unit bad demand is conected to accommodate this change in unit output power, otherwise the ULD would not permit the generated load to be different from the demand. To provide the correction to the unit load demand, the unit fiequency is compared to its desired value to produce a frequency error. This error signal, properly gained, is added to the unit load demand to provide a change in demand equal to the expected change in generation due to the governor action. For example, if a low frequency caused a 2% increase in generation, then the frequency error would cause a 2% increase in unit load demand.

The effect of frequency error on the unit load demand and to the turbine is not permitted if the unit is either in tracking (see TRACKING below), or on a high load limit. I f a high limit occurs (which removes frequency correction) and then dears, frequency control will only be restored if the correction from frequency will not put the unit back into a limiting condition.

NOTE TMI frequency is detuned so that frequency correction will not affect unit load under normal conditions.

4.6 Tracking The term tracking is used to define the mode of operation, when due to some abnormal condition, the unit is being limited in its production ofmegawatts. If one or more control stations (other than the ULD) are in manual control, or if large errors develop between the demand and the variable, unit load deniand must be made to follow the manual or limited control function to maintain coordination of all the control variables. The actual megawatts generated is used as the unit load demand.

The ULD will switch from hand or auto to the track mode if any track or limit conditions are satisfied.

Following the track or limit condition, the WLD STAR module will revert to band.

SYSIEP# A03 KA# AK2.2 Page # 4.3-30 Tier # -1 W

ROBRO importance Rating 3.3 3.3

- Group # 1 Knowledge of the interrelations between Loss of NNI-Y and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Initial plant conditions:

- Reactor is in Cold Shutdown condition.

- Decay Heat Removal Train A is operating.

Clearance and tagging activity:

- Vital Bus A is de-energized.

- Vital Bus A is re-energized 1 minute later.

Based on these conditions, identify the ONE set of statements below that describes DC-V-2A (DH Removal Cooler inlet valve) and DC-V-65A (DH Removal Cooler bypass valve) automatic responses to:

(1) Loss of power to Foxboro controllers.

(2) Restoration of power to Foxboro controllers.

A. (I) DC-V-2A fails fully open, DC-V-65A fails fully closed.

(2) DC-V-2A and DC-V-65A both return to their original positions.

B. (1) DC-V-2A and DC-V-65A both fail to random unpredictable positions.

(2) DC-V-2A repositions to fully closed, DC-V-65A reposition to fully open.

C. (1) DC-V-2A fails fully open, DC-V-65A fails fully closed.

u (2) DC-V-2A and DC-V65A both move to random, unpredictable positions.

D. (1) DC-V-2A fails fully closed, DC-V-65A fails fully open.

(2) DC-V-2A and DC-V-65A both move to random, unpredictable positions.

OP 1107-2B, 120 Volt Vital Electrical System, Limit 81Precaution G, page 3, Rev. 5.

EZ New a TMlBank TMI Question #

0 Modified TMI Bank Parent Question #

1Memory or Fundamental Knowledge Comprehension or Analysis il 55.43 i3 55.45 .7 A Incorrect answer. First part is correct, second part is not correct.

B Incorrect answer. Both parts are incorrect.

C Correct answer. Both responses are correct.

D Incorrect answer. First part wrong, second part right.

None.

TMI SRO Exam - May 2003 Friday, March 28, 2003

Number TMI - Unit 1 Operating Procedure 1107-25 Title Revision No.

120 Volt Vital Electrical System 5 2.0 LIMITS AND PRECAUTIONS A. Adhere to safety rules and regulations for operation and maintenance of electrical equipment.

B. Maintenance must be scheduled so as not to conflict with the requirements of Technical Specifications.

C. Do not parallel any inverter with the regulated bus.

D. Do not rely on a static switch to prevent paralleling an inverter with the regulated bus.

E. Following loss of any inverter verify loads are operable after inverter has been restored. If the "E"inverter is lost, de-energize the loose parts monitor at ATB (Breaker 1 1 ). To prevent blowing fuses on the voltage transient, re-energize the loose parts monitor only after restarting the inverter.

F. Prior to de-energizing ATA or ATB for any reason, refer to EP 1202-40,41,42 and OP 1105-6to evaluate consequences.

G. Foxboro Spec 200 controllers are used in control loops for EF-V-3OAfB/C/D, MS-V-4A/B, DC-V-X/B and DC-V-65A/B. If control power is lost, DC,-V-2A/Bfail open while the remaining valves listed fail closed. On power restoration the position of the valves listed will be indeterminate. Following control power restoration VERIFY POSITIONS of the respective components for plant conditions.

H. DO NOT close the 1A inverter static switch output breaker and ATA Breaker 18 at the same time.

Both breakers closed at the same time causes 1A inverter failure.

Temperature/Flow Control System Vital BLIS DC Distribution Bus 1E (1F)

Positioner

-A- DCCS 1 Positioner

-w-!

Cooler Inlet Valve DCCS Cooler Bypass Valve D C-V-65A(B) 27 DC-V-2A( B)

@-Q-23 3 62 SECTION B-07 REVISION X 6.4.3 Design

a. DH system flow (tube side) - 3,000 gpm

b. DC system flow (shell side) - 3,000 gpm C. DC system temperature, inlevoutlet - 95/115F

d. RCS temperature, inlet/outlet - 140/120°F

e. Shell design temperature - 200°F

f. Tube design temperature - 300°F g* Shell design pressure - 100 psig

11. Tube design pressure - 350 psig

1. Capacity at 140°F - 30 x IO6 BTU/hr at 250F - 125 x IO6 BTU/hr 6.5 Radiation Monitors (RM-L-2/3) 6.5.1 Description

a. NaI crystal detector

h. Refer to Enclosure 6.5.2 Location

a. 28 1 elevation Aux Building above DH vault A.

6.5.3 Power Supply

a. RM-L-2 powered from W A , Bkr 2.

b. RM-L-3 powered from VBB, Bkr 2.

6.6 Valve Control Station (DC-V-2A/B9DC-V-65A/B) 6.6.1 Description

a. The control stations for DC-V-2A aici DC-V-65A are locaceci on console center near the extension control for DH-P- 1A. The control staticins for DC-V-2B and DC-V-65B are located on console right near the extension control for DC-P- 1 B.

1). Located on panel PCR are two enable-defeat key switches. These switches enable the remote control of Train A or Train B valves or, in the Defeat position, will maintain DC-V-2s atid DC-V-65s in their normal ES alignment. These key switches are normally left in the Defeat position.

c. D C - V - 2 m will fail open and DC-V-65A/B will fail close on a loss of instrument air and/or loss of power. However. when power is restored, the clemancl 011 the control loop will change and the component position could change. Therefore. the component position must be verified correct for existing plant conditions.

7.0 INSTRUMENTATION 7.1 Refer to Enclosure 2, U

9

Form ES401-6 SYSIEP# A03 KA# AA1.l Page ## 4.3-31 Tier # 1 u ROlSRO Importance Rating 4.0 4.0 Group # 1 Ability to operate and/or monitor the following as they apply to Loss of NNI-Y: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Initial Conditions:

- Reactor power is IOO%, with ICs in full automatic.

- ICs Power Supplies are in their normal line-up.

Event:

- LOSS Of BUSATA.

Identify the ONE statement below that describes automatic equipment response to this event, and the reason for the response.

A. MU-V-5 controller fails to the mid position due to loss of ICs HAND Power.

B. MU-V-5 controller fails to the mid position due to loss of ICs AUTO Power.

C. MS-V-4NB control transfers to the Back-up Loaders due to loss of ICs HAND Power.

D. MS-V4NB control transfers to the Back-up Loaders due to loss of ICs AUTO Power.

EP 1202-42, Total or Partial Loss of lCS/NNl Auto Power, Pages 2 and 3, Rev. 38.

iL/

C New &!ITMlBank TMI Question # #207/2001 SRO .

7 Modified TMI Bank Parent Question #

9 Memory or Fundamental Knowledge Comprehension or Analysis a 55.41 .7 c 55.43 d 55.45 51.6 A Incorrect - MU-V-5 fails to mid position on loss of HAND power. AUTO power was lost.

B Incorrect - MU-V-5 fails to mid position on loss of HAND power. AUTO power was lost.

C Incorrect - Plausible, controllers swap to BU loader, but only on loss of AUTO power.

D Correct answer - Requires manual control via BU loader.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Total or Partial Loss of ICWNNI Auto Power 38 NOTE If a partial loss of auto power occurs, not all of the symptoms and automatic actions will occur.

1.0 SYMPTOMS

1. "ICS/NNI POWER LOST" alarm, H-1-8.

2. -

"ICs AUTO" light andlor "SUBFEEDS AUTO/HAND" subfeed light on ICS/NNI power monitor (Panel PCL) are off. If only a partial loss of auto power occurs, only one of the lights will go out.

3. Indications listed in Table 1 will fail. For a partial loss of auto power only the indications powered by the applicable subfeed will fail.

4. All ICSlNNl H/A station indicators (Measured Variable position) will fail. POS position remains operable (except for Turbine Control HIA Station).

5. All ICSINNI H/A station lamps will go off.

6. All the backlighted instrument selector pushbuttons will go off.

7. Invalid and inoperable alarms are listed in Table 2A and 2B.

2.0 IMMEDIATE ACTION 2.1 Automatic Action

1. Control transferred to "Final Control Elements" for Reactor, Feedwater, and Main Turbine Generator control.

a. Control transferred to HAND on all ICs "Final Control Elements" HANDIAUTO control stations retaining their last demand value (loss of ICs Auto power).

b. Main Turbine Generator transfers to "LOCAL-OWS" control on loss of ICs Auto power.

C. Reactor "Diamond" control rod panel transfers to MANUAL control on loss of ICs Auto power.

d. Control transferred to HAND on "MU-V-17" HANDlAUTO control station retaining its last demand value (loss of ICs Auto power).

e. Control transferred to HAND on "MU-V-32" HAND/AUTO control station retaining its last demand value (loss of MU Auto power).

NOTE Feedwater must be reduced manually if the reactor trips after ICs Auto power is lost.

2

Number TMI - Unit 1

.- Emergency Procedure 1202-42 Title Revision No.

Total or Partial Loss of ICSlNNl Auto Power 38

2. Signals transfer to provide valid main feedwater flow indication (recorder) and valid feedwater valve A P indication.

3. MU-V-1NB and MU-V-3 go closed due to auto powered temperature interlocks.

4. All pressurizer heater controls are "Off " due to pressurizer lo-lo level interlock.

5. MU-V-8 travels to "thru" position due to MU tank level interlock.

6. PORV (RC-RV-2) will not respond to automatic setpoints but is operable with the manual control switch.

7. Pressurizer Spray Valve (RC-V-1) will fail closed in Auto, but is operable in Manual mode.

a. MS-V4A/B will transfer to back-up manual loader ("BACKUP CTRL" Bailey Stations) 2.2 Manual Action NOTE Control Room indications affected and alternate indications are listed on Table 1. Additionally, Table 3 provides indicators unaffected by the loss of auto power.

1. VERIFY(ADJUST plant control to obtain a stable plant configuration.

CAUTION If 1. or 2. below cannot be performed as written (i.e., ATWS or failure of main turbine stop valves to close) go directly to ATP 1210-1 at that point for direction on performance of remedial actions. Refer to this procedure for additional guidance.

2. IF feedwater control cannot be established, THEN: (NA this step if not required.)

1. TRIP the reactor and VERIFY power less than 10%

2. TRIP the main turbine and VERIFY TIG stop valves CLOSED.

3. TRIP both main feedwater pumps.

4. GO TO ATP 1210-1 and refer to this procedure for additional guidance.

3

Form ES401-6 Q # 53 ..5

- Page # 4.3-36 Tier # -I u -1 ROlSRO Importance Rating 3.4 -

3.4 Group #

Knowledge of the operational implications of the following concepts as they apply to Shutdown Outside Control Room: Annunciators and conditions, indication signals, and remedial actions associated with the (Shutdown outside the Control Room).

Plant conditions:

- Cooldown outside the control room in progress.

- RCS temperatures are steady.

- MU-P-IC is operating.

- All MU system cross-tie valves are open.

- Pressurizer is observed to be rising in level, at a rate higher than attributable to RCP seal injection flow.

Based on these conditions, the action to be taken at the Remote Shutdown panels to limit this increase is to close A. MU-V-I7 (Normal Makeup Control Valve).

B. MU-V-217 (High Capacity Makeup Valve).

C. MU-V-18 (Normal Makeup Isolation Valve).

D. MU-V-16A & B (High Pressure Injection Valves).

OP 1105-20, Remote Shutdown Systems, pages I 1 and 12, Rev. 13.

L New @ TMlBank TMI Question # QR5D18-I 0-Q01 Modified TMI Bank Parent Question #

G Memory or Fundamental Knowledge EZ Comprehension or Analysis E27 55.41 .8/.10 G 55.43 E 55.45 .3 A Incorrect answer. There is no RSD Panel control for this valve.

B Incorrect answer. There is no RSD Panel control for this valve.

C Correct answer. MU-V-18 is cycled from the RSD Panel to control normal makeup to RCS.

D Incorrect answer. Based on conditions, there is no flow through these valves.

Modified answers. Reformatted question.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 W

Title I Operating Procedure 1105-20 Revision No.

Remote Shutdown Systems 13 ENCLOSURE 1 Page 2 of 4 REMOTE SHUTDOWN PANEL "B" Reactor Neutron Power Source Range Flux NI-YI- 12-1 10-1 to 105 CPS Reactor Coolant System B Coolant Temp TI-960 120- 920°F Inlet Temp TI-961 5 0 - 650°F B Coolant Press Pl-949 0 - 3000 psig Pressurizer Level Ll-777 0 - 400 inches RC-V-2 Control and indication RC-V-3 Control and indication Emergency Feedwater Control EF A Flow FI-788A digital EF B Flow FI-782A digital OTSG A Level LI-789A digital OTSG B Level LI-776A digital EF-V-30D Hand controller and demand indication EF-V-30B Hand controller and demand indication OTSG "B" Press Contr Outlet Press PI-951 0 - 1200 psig MS-V-4B Hand controller and demand indication MS-V-8A Control and indication MS-V-8B Control and indication Decay Heat Removal Temp CLR I B Out Tl-982 0 - 300°F Temp PP I6 Inlet TI-980 0 - 300°F Flow Loop B FI-803 0 - 5000 gpm MU Control and Status MU-V-20 Control and indication MU-V-18 Control and indication MU-V-16C Jog control and indication MU-V-I 6D Jog control and indication MU-V-14B Control and indication MU-V-37 Control and indication MU-V-2A Control and indication I1

Remote Shutdown Systems 13 MU Control and Status (Cont'd)

MU-V-2B Control and indication MU-V-8 Indication only MU-V-32 Flow status adequatelinadequate MU-P3B OnlOff status MU-P3C On/Off status MU-P-1B Control switch and status indication MU-P-1C Control switch and status indication MU-T-1 11-778 0 - I 0 0 inches DH-T-1 BWST Level LI-809 0 - 60feet Switch Gear 4160V Bus 1E White energized status light 480V Bus IS White energized status light 480V Bus I T White energized status light 480V 1B ES MCC White energized status light 480V 1B ES Screenhouse White energized status light 480V 1B ESV MCC White energized status light 480V I C ESV MCC White energized status light Instrument Air IA-P-1B On/Off Status Diesel Generator EG-Y-1B Green - Ready to load; Red - Running AUXILIARY REMOTE SHUTDOWN PANEL "B" Control Room Comm. Svs Isolation Gray page transferred - white light M&l Sys transferred white light Nuclear Services River Water NR-P-IC Control switch and indication NR-V-1C Control and indication NR-V-15B Control and indication NR-V-I8 Control and indication 12

Form ES-401-6 Q # 036

- Page # 4.3-41 Tier # -1

'w ROlSRO Importance Rating 3.8 4.0 Group # -

3 Ability to determine and interpret the following as they apply to Refuel Canal Level Decrease:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Plant conditions:

- Refueling outage in progress.

- Fuel Handling Bridge is over the reactor core with new fuel assembly.

- Annunciator PLB-4-9, Fuel Transfer Canal Lo Level, actuates.

- Fuel Transfer Canal water level is lowering at 7 inches per minute.

- RB Sump level is rising.

Based on these conditions, complete the following statement from the list below:

The fuel assembly engaged and in the mast of the main fuel handling bridge shall be . ..

A. inserted between the plenum and the upender in the deep end of the Fuel Transfer Canal, IAW EP 1203-43, Transfer Canal Level Loss.

B. inserted between the plenum and the upender in the deep end of the Fuel Transfer Canal, IAW annunciator PLB-4-9 alarm response procedure.

C. inserted into the nearest available core location, IAW EP 120343 Transfer Canal Level Loss.

D. transferred to the spent fuel pool, IAW annunciator PLB-4-9 alarm response procedure.

L AbP 120343, Transfer Canal Level Loss, page 2, Rev. I O .

3 New E! TMI Bank TMI Question # #43 7/2001 SRO Z Modified TMI Bank Parent Question #

7 Memory or Fundamental Knowledge i

Z l Comprehension or Analysis 55.43 .5 i3 55.45 .13 A Incorrect answer, although plausible since this is one of the storage locations for loss of pool level based on rod location.

B Incorrect answer, although plausible since this is one of the storage locations for loss of pool level based on rod location.

C Correct answer. Procedure requires rod to be lowered into the nearest core location.

D Incorrect -although plausible since this is one of the storage locations for loss of pool level based on rod location.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 b Abnormal Procedure 1203-43 Title Revision No.

Transfer Canal Level Loss 10 1.0 SYMPTOMS 1.I Rapid decrease in Fuel Transfer Canal Water Level.

1.2 R.B.Sump Level increasing faster than normal.

1.3 Fuel Transfer Canal Lo Level Alarm on PLB-4-9.

1.4 Possible higher than normal radiation levels as indic ted n RM-G6, RM-G7. RM-G22, RM-G23, andlor RM-G9.

1.5 Leakage from primary shield penetrations.

2.0 IMMEDIATE ACTION 2.1 Automatic Action None 2.2 Manual Action I NOTE In some instances, time required to respond could be critical depending on size of leak.

2.2.1 EVACUATE all unnecessary personnel from Reactor Building by sounding the Reactor Building Evacuation Alarm and making an announcement on the page system.

2.2.2 Efuel transfer canal level decrease rate is >6" per minute, THEN PERFORM the following steps as quickly as possible:

2.2.2.1 Eany fuel bridge has a component engaged, THEN PROCEED to the nearest available storage space & N J LOWER the fuel component, DO NOT DISENGAGE THE COMPONENT.

2.2

the bridge is cl to the core, THEN INSERT the component to the full down q&iition in the nearest available core locatio 2.2.2.3 E the bridge is nearer the deep end of the transfer canal, THEN LOWER the component between the plenum and up ender until the component just touches the fuel pool floor, as indicated by a rapid decrease in load cell reading. DO NOT DISENGAGE THE COMPONENT.

2.2.2.4 MOVE the fuel transfer carriages to the spent fuel pool.

2.2.2.5 CLOSE FH-V-1A and FH-V-15.

2

Form ES-QOI-6 Q # 037 .-

SYSIEP# KA# EK2.1 Page # 4.3-9 Tier # -I 3.8 -1 I_

4.0 Y

ROBRO importance Rating Group #

Knowledge of the interrelations between Excessive Heat Transfer and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Loss of ICS/NNI HAND and AUTO power occur.

Power should not be restored to ICs HAND power before the emergency procedure immediate manual actions are completed because this action could result in.. .

A. overloading the alternate power supply.

B. turbine bypass valves failing open.

C. loss of Emergency Feedwater.

D. an overfeed event.

EP 120240, Loss of ICs Hand and Auto Power, Page 3, Rev. 39.

0 New @ TMIBank TMI Question # QR5D20-03-QO6

'-1 1 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge

&?IComprehension or Analysis PI 55.41 .7 3 55.43 iz 55.45 .7 A Incorrect answer. Although plausible, it is not the reason.

B Incorrect answer. TBVs do not fail open on restoration of power.

C Incorrect answer. No loss of EFW will occur.

D Correct answer. Feedwater valves will fully open and to cause an overfeed condition.

shortened stem significantly, reordered distractors TMI SRO Exam - May 2003 Friday, March 28,2003

Number

- Title Loss of ICs Hand and Auto Power TMI Unit 1 Emergency Procedure 120240 Revision No.

39 CAUTION Do not select alternate ICS/NNI Power or otherwise attempt to restore power at this point. Upon restoration of HAND power, main and startup feedwater valves will stroke fully open.

If a. or b. below cannot be performed as written (Le., ATWS or failure of main turbine stop valves to close) go directly to ATP 1210-1 at that point for direction on performance of remedial actions. Refer to this procedurefor On a confirmed loss of ICs Hand and ICs Auto power.

1. TRIP the reactor and VERIFY power less than 10%.

2. TRIP the main turbine and VERIFY T/G stop valves closed.

3. TRIP both main feedwater pumps.

L'

4. GO TO ATP 1210-1 and refer to this procedure for additional guidance.

NOTE Control room indications listed in Table 1, are available for controlling plant parameters.

3.0 FOLLOW UP ACTION Objective: The objective of this procedure is to stabilize the plant in a hot shutdown condition and to restore ICSlNNl power. If unable to restore power, proceed with a controlled plant cooldown.

A. VERIFY EFW Controls OTSG Level at 2 25" startup range.

B. OPEN MS-V-4PJB with B/U loaders, ("BACK UP CTRL" Bailey Stations) to reseat main steam safety valves and control OTSG pressure.

C. IF MU-V-17 cannot be controlled in Hand or Auto, THEN (NA if MU-V-17 can be controlled)

a. USE MU-V-217 to control pressurizer level.

b. DISPATCH an operator to isolate MU-V-17 locally BY CLOSING MU-V-916.

3

- Page # 4.3-14 Tier # -1 W

ROlSRO importance Rating 3.3 4.0 Group # 2 Ability to determine and interpret the following as they apply to LOCA Cooldown: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Sequence of events:

- Reactor trip due to low RCS pressure.

- Automatic 1600 psig ESAS actuation.

- Loss of RCS Subcooled Margin.

- All RCPs were tripped.

Current plant conditions:

- RCS pressure is 780 psig.

- Core exit thermocouple temperature is 485°F.

Identify the ONE statement below that describes the MAXIMUM ALLOWABLE RCS cooldown rate limit for these conditions.

A. 40°F per hour in accordance with OP-TM-EOP-006, LOCA Cooldown.

B. 50°F per hour, since RCPs are not running, in accordance with OP-TM-EOP-010 Guide 11, Cooldown Rate (CDR) Limits.

C. 100°F per hour in accordance with TMI Technical Specifications.

D. NO maximum cooldown rate limit applies, since the RCS had been saturated, in accordance with OP-TM-u EOP-010 Guide 11, Cooldown Rate (CDR) Limits.

OP-TM-EOP-010 Guide 11, Cooldown Rate (CDR) Limits, page 24, Rev. I.

IZ New J TMlBank TMI Question ##

0 Modified TMI Bank Parent Question #

c! Memory or Fundamental Knowledge h? Comprehension or Analysis kz 55.43 .5 E 55.45 . I 3 A Incorrect answer. This is a plausible distracter since it actaully is a minimum cooldown rate in EOP-006.

B Correct answer - this is the maximum rate allowed under these conditions, as specified in Guide 1I.

C Incorrect answer. Plausible since this is the Tech Spec cooldown rate limit for normal conditions.

D Incorrect answer. The RCS is not saturated in present conditions.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

OP-TM-EOP-010 Revision 1 Page 24 of 49 Guide I 1 Cooldown Rate (CDR) Limits IAAT SCM > 25OF, then observe the following RCS cooldown rate limits:

or 30°F/HR if RCS Temperature < 255OF.

VERIFY RCS temperature > 255OF. RCS Cooldown rate limit is 30°F/HR.

RCS Cooldown rate limit is 100°F/HR.

Q # ..

039. -

- Page # 4.3-12 Tier # -1 u Group # -2 ROlSRO Importance Rating 3.3 3.5 Knowledge of the operational implications of the following concepts as they apply to LOCA Cooldown: Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Cooldown).

c a %1-7 Sequence of events:

- Reactor trip due to low RCS pressure (LOCA).

- Automatic ES actuations:

- 1600 psig.

- 500 psig.

- 4 psig RB pressure.

- 30 psig RB pressure.

Current plant conditions:

- RCS pressure is STABLE at 270 psig.

- RB flood level is 32 inches and rising.

- RB pressure is 7 psig and lowering.

- BWST level is 14.5 feet and lowering.

Based on the conditions above, identify the ONE statement below that describes (1) operational implications related to continuing to lower BWST and raising RB flood levels, and (2) required remedial action.

A. ( I ) Adequate pump NPSH (Net Positive Suction Head) will be lost.

(2) Transfer HPI Pump suction to LPI pump discharge per OP-TM-EOP-010 Guide 5, Transfer MU Pumps to "Piggyback" Mode.

u B. (1) Vital instrumentation could be damaged by flooding in the RB.

(2) Transfer LPI pump suction to RB sump per OP-TM-EOP-010 Guide 21, Transfer to RB Sump Recirculation.

C. (1) Adequate pump NPSH (Net Positive Suction Head) will be lost.

(2) Secure LPI pumps per OP-TM-EOP-010 Guide 22, RB Sump Recirculation.

D. (1) Vital instrumentation could be damaged by flooding in the RB.

(2) Secure RB Spray pumps per OP-TM-EOP-010 Guide 18, Containment.

EOP-010 Guide 5, Transfer MU Pumps to "Piggyback" Mode, page 16, Rev. 1.

Pj New fl TU1 Bank TMI Question #

0 Modified TMI Bank Parent Question #

LI Memory or Fundamental Knowledge Comprehension or Analysis 3 55.41 .8/.10 0 55.43 .5 @ 55.45 .3 A Correct answer. The concern listed is correct as well as the procedural actions to be performed.

b B Incorrect answer. Instrumentation flooding becomes a concern at higher flood levels. Conditions do not meet applicability requirements for Guide 21.

C Incorrect answer. Conditions do not meet applicability requirements for Guide 22.

TMI SRO Exam - May 2003 Friday, March 28,2003

D Incorrect answer. Conditions do not meet applicability requirements for Guide 18.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

OP-TM-EOP-010 Revision 1 Page 16 of 49 If required per Guide 20, then,

1. ENSURE all flow to the MU tank is isolated, as follows:

1. ENSURE MU-V-3 or [MU-V-2A and MU-V-2Bl are CLOSED.

2. ENSURE MU-V-25 or MU-V-26 are CLOSED.

3. ENSURE MU-V-36 or MU-V-37 is CLOSED.

2. OPEN DH-V-7A or DH-V-78.

t CAUTION 1 Opening MU-V-36 and 37 while in piggyback mode may cause MU tank design pressure 1 ( I 0 0 psiq) to b e exceeded.

1 1 3. IAAT HPI is throttled to MU pump flow c 115 GPM/Pump and two or more MU

- pumps are operating, then REDUCE the number of operating MU pumps.

n4. IAAT HPI must be throttled IAW Rule 2 to MU pump flow < 115 GPM/pump and only one M U pump is operating, then OPEN the PORV to ensure minimum MU pump flow.

OP-TM-EOP-010 Revision 1 Page 38 of 49 Guide 20 PRIOR to Transfer to RB Sump ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. OBSERVE the rate of BWST level reduction and ENSURE that the following actions are completed prior to BWST level < 15 ft.

A. UNLOCK and CLOSE the following breakers on I C ES Valves MCC:

DH-V-2 (Unit 3B).

CF-V-1A (Unit 3C)

CF-V-18 .(Unit 4C) . .

B. At Seal Injection Area (Aux. Bldg. 305' elev.):

OPEN DH-V-64.

UNLOCK and CLOSE u CA-V-371 (between CA-V-2 and RB wail).

OPEN MU-V-198 (SI Filter Bypass Valve).

C. If the Aux Bldg sump pumps are lined up in recirculation mode, then PERFORM the following at Aux. Bldg. 281' above AB Sump:

CLOSE W DL-V-714.

OPEN WDL-V-713.

2. F e n BWST Level -= 15 feet, then CONTINUE?

I

3. VERIFY SC GO TO Guide 5 "Transfer MU pumps to "piggyback" mode.

I I

4.1 OBTAIN US concurrence.

4.2 PLACE MU-P-IA, MU-P-1B and MU-P-1 C in Pull-To-Lock. . . -

L

OP-TM-EOP-010 Revision 1 Page 39 of 49 Guide 21 Transfer to RB Sump Recirculation I

ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED NOTE Maximum RB Flood Level limit (64) is to protect PZR Level and OTSG Level instruments.

Minimum RB Flood Level Limit (32) is to ensure adequate DH & BS pump NPSH.

1. ANNOUNCE initiation of RB Sump Recirculation over the plant page and radio.

2. VERIFY RB FLOOD LEVEL > 32.

~ ........ ........ ~ . .......... . ... . ..... .... ...... ........

3. When BWST LEVEL reaches 9.5 FT SHUTDOWN associated DH pump.

or RB Flood Level > 56 inches, then OPEN DH-V-6A and DH-V-6B. INITIATE Guide 3.

~

- - -. . - _.- -. __ -- - - .- __ - _F - __ -

CAUTION RB Pressure and Temperature will increase after establishing RB recirculation. A 4# ESAS

4. When BWST LEVEL reaches 6.33 FT or RB Flood Level

> 56 inches, then

4. ENSURE open DH-V-6A and CLOSE DH-V-SA.

4.2 ENSURE open DH-V-6B and CLOSE DH-V-5B.

4.3 CLOSE BS-V-2A and BS-V-2B.

4.4 ENSURE LPI Throttling criteria is met (RULE 2). .. - . .

5. INITIATE Guide 22 RB Sump

. u RecircuIati o n .

OP-TM-EOP-010 Revision 1 Page 40 of 49 Guide 22 RB Sump Recirculation (Page 1 of 2)

I ACTION/EXPECTED RESPONSE I

1. NOTIFY Rad Con to assess Aux Bldg radiation levels.. _

- - I

2. REQUEST DH sample (RB sump).

3. INITIATE OP-TM-212-911 Post LOCA Reactor Vessel boron concentration control.

4. VERIFY RB SUMP pH > 8.0 and ADD sodium hydroxide through the DHR

< 11.0. system. . .

5. VERIFY RCS / RB SUMP / DH 4 boron concentration > COLD SHUTDOWN boron concentration per 1103-4, So I ubI e Pois0n Conce nt rat io n

- t ro

. Con I. . .. . . _ - .......-.- .....-.-... - -..-.... . ...._ . . . ,

6. MONITOR RB FLOOD LEVEL REVIEW the guidance on page 2 if leakage is and VERIFY level is not suspected.

changing and is between 40 and 64.

OP-TM-EOP-010 Revision 1 Page 41 of 49 Guide 22 RB Sump Recirculation (Page 2 of 2)

A. IAAT leakage in the DH vault is detected, then OBTAIN ED concurrence and :

LPI Train A LPI Train B

1. VERIFY LPI train B operating 1. VERIFY LPI train A operating

2. SHUTDOWN DH-P-1A 2. SHUTDOWN DH-P-I B

3. CLOSE DH-V-6A 3. CLOSE DH-V-6B

4. CLOSE DH-V-4A 4. CLOSE DH-V-4B

5. ENSURE closed DH-V-7A 5. ENSURE closed DH-V-7B

6. ENSURE closed RC-V-4 6. ENSURE closed DH-V-1, 2 and DH-V-3 L d 7. ENSURE closed DH-V-5A 7 ENSURE closed DH-V-SB B. IAAT BWST in-leakage is indicated and DH-V-7A or B is OPEN, then CLOSE MU-V-14A and MU-V-14B.

OP-TM-EOP-010 Revision 1 Page 36 of 49 Guide 18 Containment ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY RB Pressure > 2 psig.

VERIFY 4 psig ESAS has actuated. INITIATE OP-TM-534-901,

..................................................................................... - ............................. ....- ................. RB Emergency DETERMINE energy source.

OBSERVE Radiation monitors and OTSG pressures and feed rates.

PRIMARY OTSGA OTSGB If source is a secondary side leak (either FW or MS), then PERFORM Phase I and Phase 2 Isolation of Rule 3 - XHT.

Shutdown of Building Spray Prerequisites:

RB pressure < 2 psig.

ED concurrence that atmospheric iodine removal has been accomplished.

Procedure:

1.I Shutdown one train of RB Spray

1. Stop BS-P-1A(B).

2. Close BS-V-3A(B).

3. Close BS-V-lA(B).

1.2 Monitor RB Temperature and Pressure.

1.3 Shutdown the remaining train of RB Spray

1. Stop BS-P-1 B(A).

2. Close BS-V-3B(A).

3. Close BS-V-lB(A).

Form ES-401-6 SYSIEW KA# 2.4.30 Page # 2-14 Tier # -1 W

ROlSRO Importance Rating 2.2 -

3.6 Group # -I Knowledge of which events related to system operations/status should be reported to outside agencies: Natural Circulation Cooldown.

Initial plant conditions:

- Reactor power is loo%, with ICs in full automatic.

Based on these conditions, identify the ONE operational event below that will require immediate NRC notification.

A. ICs runback due to trip of one feedwater pump.

B. Trip of all 4 RCPs due to loss of offsite power (LOOP).

C. Reduction to 50% power due to Main Condenser tube leak.

D. EF-P-1 start caused by invalid signal due to personnel error. No EFW flow was initiated into the OTSGs.

AP 1044, Event Review and Reporting Requirements, Section 4.1.8, Emergency Plan, page 5, Rev. 49.

2l New E TMI Bank TMI Question #

G Modified TMI Bank Parent Question #

u E Memory or Fundamental Knowledge E Comprehension or Analysis E 55.43 .5 E 55.45 . I 1 A Incorrect answer. Although a significant plant transient that affects heat removal capabilities, this is not reportable to the NRC.

B Correct answer. This condition is reportable under MUI, even without complications due to failed EDGs.

C Incorrect answer. Although a major transient, and may involve secondary chemistry limits, this is not reportable to the NRC.

D Incorrect answer. System actuations due to invalid signals, personnel error do not require immediate NRC notification.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

I TMI - Unit 1 L l Administrative Procedure 1044 Title Revision No.

Event Review and Reporting Requirements 49

a. With direction from Security personnel, evaluate the event under consideration utilizing the Exelon Reportability Reference Manual, Volume 1,

b. If the event is considered reportable under 73.71, then make notifications as described in 4.3.1. Senior Security personnel may make these notifications.

4.1.6 10 CFR 21, Defects and Noncompliance

a. 10 CFR 21 requires notification of the NRC upon discovery of substantial safety defects. Individuals generating reports in accordance with 10 CFR 21 shall provide a copy to the Plant Manager. Refer to the Exelon Reportability Reference Manual, Volume 1, 4.1.7 10 CFR 26, Fitness for Duty

a. Evaluate the event under consideration against the criteria specified in the Exelon Reportability Reference Manual, Volume 1.

b. If the event is considered reportable under 10 CFR 26, then make notification as described in 4.3.1.

4.1.8 Emergency Plan NOTE Initiation of the Emergency Plan is in itself reportable within one hour tinder 10 CFR 50.72. *.

a. The Emergency Plan requires special NRC notification for specific events.

The Emergency Plan shall be used for guidance in making those notifications.

b. Events reportable under the Emergency Plan may also be reportable under 10 CFR 20 or other requirements.

4.1.9 Events of Potential Public Interest NOTE The declaration of an Event of Potential Public Interest shall not be made in lieu of the declaration of a formal emergency classification (i,e.?Unusual Event).

a. These are events that may or may not be considered reportable under other sections of this procedure. Refer to Enclosure 1 for a list of these events.

5

Form ES-401-6 Q # 041 SYS/EP# KA# EK3.2 Page # 4.3-16 Tier # -1 W 3.8 Group #

ROlSRO Importance Rating 3.0 - -1 Knowledge of the reasons for the following responses as they apply to Natural Circulation Cooldown: Normal, abnormal and emergency operating procedures associated with (Natural Circulation Cooldown)

Plant conditions:

- Reactor is tripped from full power due to loss of offsite power (LOOP).

- Subcooled natural circulation RCS cooldown in progress.

- Pressurizer heaters are de-energized.

Based on these conditions, identify the ONE statement below that describes the required method to reduce RCS pressure, IAW OP-TM-EOP-010 Guide #8, RCS Pressure Control.

A. Vent the Pressurizer to the RC Drain Tank.

B. Open RC-V-1 (Spray Valve).

C. Initiate RCS letdown flow.

D. Raise EFW flow.

OP-TM-EOP-010 Guide #8, RCS Pressure Control, Rev. 1.

None.

V. E.10.02 3 New E TMlBank TMI Question #

L-0 Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge

@ Comprehension or Analysis Q 55.41 .5/.10 g 55.43 .5 55.45 .6/.13 A Correct answer. This method is required for these conditions since there is no forced RCS flow.

B Incorrect answer. Procedure specifies use of Pressurizer vent since there are no RCPs in operation.

Distracter is plausible since this is the prescibed method of pressure reduction for other conditions with RCP(s) operating.

C Incorrect answer. Procedure specifies use of Pressurizer vent since there are no RCPs in operation.

Distracter is plausible, since this action would lower RCS pressure due to reduction in coolant inventory (lower Pressurizer level).

D Incorrect answer. Procedure specifies use of Pressurizer vent since there are no RCPs in operation.

Distracter is plausible, since this action would result in lower RCS pressure due to cooldown and contraction of RCS water.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

OP-TM-EOP-010 Revision 1 Page 20 of 49 Guide 8 RCS Pressure Control (Page 1 of 2)

~

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY the reactor is shutdown and SCM

> 25°F.

If it is required to MINIMIZE SCM and SCM

> 70°F, then lower RCS pressure to between 70' and 30" SCM.

RAISE or LOWER RCS pressure per the following direction, as needed, to maintain RCS pressure within the limits of Figure 1 and 1A.

v To RAISE RCS Pressure:

1"-

~

VERIFY RCS temperature is stable or CONTROL RCS temperature.

cooldown rate is within desired band.

-1 .

VERIFY Pressurizer level is stable or rising. RAISE RCS makeup or HPI.

I- . .

ENERGIZE Pressurizer Heater banks or IF Pzr Htr power is available, then ADJUST heater demand for the SCR INITIATE OP-TM-220-901 to transfer Grp 8 controlled heaters. . . . . . . or 9 to ES power..

OP-TM-EOP-010 Revision 1 Page 21 of 49 Guide 8 RCS Pressure Control (Page 2 of 2)

To LOWER RCS Pressure:

ACTlONlEXPECTED RESPONSE RESPONSE NOT OBTAINED 1 ENSURE HPI is throttled per (Rule 2) and control RCS inventory.

VERIFY RCS temperature is stable or NOTiFY US.

de creasing.

If pressurizer steam bubble is controlling RCS pressure, then ENSURE Pressurizer Heaters are OFF.

NOTE Pressurizer cooldown rate is to be limited to less than 1OOOF in any one hour VERIFY an k C P is operati // ENSURE WDG-V-3 and WDG-V-4 are OPEN (Guide 19).

OPEN RC-V-44.

OPEN RC-V-28.

When RC Drain Tank pressure > 40 psig or RCS pressure is in desired range, then, CLOSE RC-V-28.

- . GLOSE RC -V-44. '

G O TO END.

- ENSURE RC-V-3 is open.

- OPEN RC-V-1.

- When RCS pressure is in the desired range, then CLOSE RC-V-1.

Form ES401-6 W

RO/SRO Importance Rating 3.4 4.1

- Group # -3 Knowledge of limiting conditions for operations and safety limits: EOP Rules.

rn m From the list below, identify the ONE condition that requires PLANT SHUTDOWN to be initiated due to unavailability of instrumentation needed to diagnose and respond to (implement guides and rules) situations that could result in inadequate core cooling or accidents outside the design basis for the plant.

A. Loop A SCM Monitor is operable; Loop B SCM Monitor has been inoperable for the past 8 days.

B. Pressurizer level instrumentation channels LT-1 and LT-3 inoperable for 3 days; Using alternate process computer indication for manual Pressurizer level control.

C. Both EFW Flow indication channels for OTSG IB are inoperable for 2 days; OTSG 1A level and EFP discharge pressure instrumentation is operable.

D. BlRO Display Channel Quadrant X has only 1 detector operable for 5 days; Both Loop A and Loop B SCM Monitors are operable.

Technical Specification 3.5.5 Accident Monitoring Instrumentation, pages 34Oa (Amendment IOO), 340b (242), and 340c (105).

Technical Specification 3.5.5 Accident Monitoring Instrumentation, pages 3-40a (Amendment IOO), 340b (242), and 340c (105).

V.F.01.08

\/

a New 1? TMlBank TMI Question ##

0 Modified TMI Bank Parent Question ##

FZ Memory or Fundamental Knowledge Ci Comprehension or Analysis

.- 55.41 L?? 55.43 .2 E 55.45 .2 A Incorrect answer. Condition meets minimum number of channels required by Technical Specifications.

Plausible distracter since the condition stated exceeds time limit for shutdown with one less than minimum operable channels.

B Correct answer. Condition is less than the minimum number of channels required by Technical Specifications. Operation is limited to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in HSD in next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , CSD in additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C Incorrect answer. Condition meets minimum number of channels required by Technical Specifications.

Plausible distracter since the condition stated exceeds time limit for shutdown with one less than minimum operable channels.

D Incorrect answer. Condition meets minimum number of channels required by Technical Specifications.

Plausible distracter since the condition stated exceeds time limit for shutdown with one less than minimum operable channels.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

To assure operabfl5.y oP key 5nStmenbtion useful fn diagnosing situations which could represent ar lead to inadequate core cooling or evaluate and predict .Me .course of accidents-beyondthe desfgr .

basis,

. specificatian 3.5.5-1 -The m i n i m u n nurrbe~aP channels identifled for the instruments In Table 3.5-2, shall be OPERABLE. Uith the number o f instrunentauun channels less than the minmurn required, restore the inoperable channel(s) to OPERABLE status within seven (7) days (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for pressucizez level) or be in a t least HOT ss.fLITDOlbJ withfn the next s i x

( 6 ) hours and in COLD SWTOOWN within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Prior i o start* f~llowitga COD s ", the rrdnhum number of channels shown in Table 3.5-2 shall be operable.

-I 3.5.5.2 The channels i d e d i f i e d for the instruments specified in Table 3-53 shall be OpERABLE. With the number of instrufmdation channels less than requhed, restore-theIncperable &annel(s) t o D F W in accurdame with the attion specified in Taule 3.5-3.

-Bases She saturation Margin fbnitm provides a qdck and relfatde means for determioation o f saturation temperature margins, Hand calculation or saturation pressure and saturation temperature margins can be easily and quickly performed as an alternate incficath far the Saturation Margin Monitors.

Discharge flarfqm ttre t w o (21 pressurizer code safety valves and the PORV is measured by differential pressure tmnsmittess connected across elbow taps downstream nf' ea& valye. A delta-pressme indicaffun fmm each pressure transmitter is availaole i n the contra1 room t o indicatecode safety or r e l i e f valve line flow. Rn alarm is a l s o provided i n the control mom . t indicate

~ W t discharge from a pressurizer code safety or relief valve is occurring. In addition, an acoustic monitor is provided t o detect flow i n t h e POW discharge line. An a l m is provided in the cuntrol rom for t k acoustic monitor, Amendment No. 78, - 3OD

L' 3.5.5 ACCIDENT MONITORING INSTRUMENTATION (Continued)

The Emergency Feedwater System (EFW) is provided with two channels of flow instrumentationon each of the two discharge lines, Local flow indication is also available for the ERN System.

Although the pressurizer has multiple level indications, the separate indications are selectable via a switch for dispfay on a single display. Pressurizer level, however, can also be determined via the patch panel and the computer log. In addition, a second channel of pressurizer level indication is available independent of the "1.

Although the instruments identified in Table 3.5-2are significant in diagnosing situations which could lead to inadequate core cooling, loss of any one of the instruments in Table 3.5-2would not prevent continued, safe, reactor operation. Therefore, operation is justified for up to 7 days (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for pressurizer level). Alternate indications are available for Saturation Margin Monitors using hand calculations, the PORVlSafety Valve position monitors using discharge line thermocouple and Reactor Coolant Drain Tank indications, and for EFW flow using Steam Generator level and ERN Pump discharge pressure. Pressurizer level has two channels, one channel from NNI (2 DIP instrument strings through a single indicator) and one channel independent of the NNI. Operation I

with the above pressurizer level channels out of service is permitted for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Alternate indicationwould be available through the plant computer.

The operability of design basis accident monitoring instrumentation as identified in Table 3.5-3,ensures that sufficient information is available on selected plant parameters to monitor and assess the variables following an accident. (This capabitity is consistent L' with the recommendationsof Regulatory Guide 1.97, "Instrumentationfor Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," Rev. 3, May ?983.) These instrumentswill be maintained for that purpose.

3-40b Amendment No. , ,242

TABLE 3.5-2 ACCIDENT MONITORING INSTRUMENTS FUNCTION INSTRUMENTS NUMBER OF CHANNELS MINIMUM NUMBER OF CHANNELS 1 Saturation Margin Monitor 2 1 2 Safety Valve Differential 1 per discharge line 1 per discharge line NI P Pressure Monitor N

3 PORV Position Monitor 2 1' 4 Emergency Feedwater Flow 2 per OTSG 1 perOTSG x

0 5 Pressurizer Level 2 1 0

6 Backup lncore Thermocouple 4 thermocoupleslcore 2 thermocoupleslcorequadrant Display Channel quadrant

With the PORV Block Valve closed in accordance with Specification 3.1.12.4.a,the minimum number of channels is zero.

Q # 043.

Page ## 4.3-21 Tier # -1 U'

ROISRO importance Rating 3

- 3.6

- Group# 3 Knowledge of the operational implications of the following concepts as they apply to EOP Rules: Normal, abnormal and emergency operating procedures associated with (EOP Rules).

Plant conditions:

- PORV-HPI cooling is in progress.

- OTSG 1A and OTSG I B are both dry.

- RCS subcooled margin is 3°F.

- RCPs are tripped.

Both Main Feedwater (MFW) and Emergency Feedwater (EFW) are now available to re-establish FW to the OTSGs.

OP-TM-EOP-010 guidance for this plant condition:

- Rule4, FWC:

- Minimum EFW flow rate is 215 gpm per OTSG.

- Minimum MFW flow rate is 1E6 lbmlhr per OTSG.

- Use of EFW is preferred.

- Guide 13, Dry OTSG:

- Maximum EFW flow rate is 185 gpm.

- Maximum MFW flow rate is 0.1E6 Ibm/hr.

- Use of MFW is preferred.

Based on these conditions, identify the ONE statement below that d scribes operator actic that demonstrates

</ proper application of this procedure guidance.

A. Establish and maintain greater than or equal to 215 gpm EFW flow to each OTSG.

B. Establish and maintain less than or equal to 185 gpm EFW flow to each OTSG.

C. Establish greater than or equal to 1E6 Ibm/hr MFW flow to each OTSG.

D. Establish and maintain less than or equal to 0.1 E6 Ibm/hr MFW flow to each OTSG.

OP-TM-EOP-010 Rule 4, FWC, Guide 13 (Dry OTSG), pages 8 and 26, Rev. 1.

OS-24, Conduct of Operations During Emergency and Abnormal Events, section 4.16.G, Page 8, Rev.-7.

i New 2 TMI Bank TMI Question # NOT IN CEBS, used on LOR quiz 3 Modified TMI Bank Parent Question #

G Memory or Fundamental Knowledge

&?IComprehension or Analysis 55.41 .8, .10 0 55.43 B 55.45 .5 A Correct answer. Rule 4 overrides Guide13 if there is conflict between the two. This is minimum required FW b flow to each OTSG when SCM is lost, and OTSG levels are ~ 7 5 % .

B Incorrect answer. Prioritization of Rule 4 over Guide 13 renders this response incorrect.

C Incorrect answer. Rule 4 states use of EFW is preferred over use of MFW if both are available.

TMI SRO Exam - May 2003 Friday, March 28,2003

D Incorrect answer. Prioritizationof Rule 4 over Guide 13 renders this response incorrect.

TMI SRO Exam - May 2003 Friday, March 28,2003

//3 Number TMI - Unit 1 Operations Department u Administrative Procedure OS-24 Title Revision No.

Conduct of Operations During Abnormal and Emergency Events 7 4.1.6 Rules and Guides A. The Loss of Subcooling Margin Rule and the Excessive Heat Transfer Rule, when identified, are performed immediately, even if Reactor Trip Immediate Manual Actions are in progress. See section 4.3.5 Reactor Trip Actions.

B. The following sequence is used when Rule based action is required:

1. Announce the applicable Rule.

2. Pull the applicable Rule card.

3. The Unit Supervisor provides concurrence.

4. Perform the Rule based actions.

5. Report completion of Rule based action and return card to holder.

C. Reactor Operators consider the priority of Rules in relationship to actions or steps in progress by the Unit Supervisor. If the Unit Supervisor is in the process of Immediate Manual Action verification, or other higher priority action, the Reactor Operator delays implementation of the lower priority action until the Unit Supervisor is able to provide concurrence for the action.

D. Rules are numbered according to priority. If multiple Rule based actions are required, the highest priority Rule is performed first. If resources allow, multiple Rules may be performed in parallel by separate personnel.

E. Rules and Guides are applicable whenever any EOP is entered. Rules and Guides may be implemented without specific direction from an EOP if the condition or intent is met.

F. After action has been initiated in accordance with a Rule or Guide, further US concurrence is required prior to any substantive change in the action being taken. Otherwise, the operator will continue to take action in accordance with the Rule or Guide.

G. r procedures conflict, the following order Rule? (including the order of priority nd (4) other procedure requirements.

8

OP-TM-EOl?-OIO Revision 1 P a g e 8 of 49 I ACTlON/EXPECTED RESPONSE RESPONSE NOT OBTAINED I VERIFY SCM > 25°F. MAINTAIN OTSG level 75 - 85%

OPERATING Range Level.

VERIFY a t least 1 RCP operating. MAINTAIN OTSG level 2 50% OPERATING Range Level.

... I MAINTAIN OTSG level 2 25" STARTUP Range Level.

B. If Level < minimum, then MAINTAIN the following MINIMUM required flow:

If SCM < 25OF and OTSGs are available, If EFW is not available, then FEED then FEED > 215 gpm/OTSG using EFW. > 1.0 Mlbm/hr using MFW.

If SCM < 25OF and only one OTSG is If EFW is E t available, then FEED available, then FEED > 430 GPM to the good > 1.0 Mlbm/hr using MFW.

OTSG using E M .

If RCPs are OFF, then FEED OTSG at EFW is not available, then FEED maximum available using EFW, within RCS > 1 .O Mlbm/hr using MFW.

Cooldown rate limit.

There is no minimum required flow rate.

OP-TM-EOP-010 Revision 1 Page 26 of 49 Guide 13 Dry OTSG

1. VERIFY OTSG SU Level < 6 and OTSG pressure at least 200 psi below Psatfor T,

2. MONITOR Tube to Shell Differential Temperature (TSDT) and REVIEW Guide 14.

3. VERIFY the other OTSG is available. GO TO Step 5.

4. VERIFY all RCPs are OFF or TSDT Limits are being challenged NOTE u

Automatic EFW actuation is not restricted by this guidance. Limit feedwater flow to the Dry OTSG 1 until OTSG pressure has been restored. RCP operation is desired.

5. VERIFY the DRY OTSG pressure boundary VERIFY the OTSG pressure boundary failure is is INTACT. -not in the Intermediate or Reactor Building.

6. If TSDT tensile limit is being challenged, If RCPs are OFF, then FEED the DRY OTSG at then, a maximum of 185 GPM using EFW.

I ) If OTSG pressure boundary is not intact, then VERIFY an RCP is operating.

2) FEED the DRY OTSG at a maximum flow of 0.1 MIb/HR using Main Feedwater.

7. If TSDT compressive limit is being If RCPs are OFF, then FEED the DRY OTSG challenged, then, at a maximum of 185 GPM using EFW.

I ) If at least one RCP is ON, then FEED the DRY OTSG at a maximum of u 435 GPM using EFW.

Q # 044 SYS/EP# 001 KA# K6.03 Page # 3.1-8 Tier # -2 u

ROlSRO Importance Rating 3.7 4.2 Group # - 1 Knowledge of the effect of a loss or malfunction of the following will have on the Control Rod Drive System: Reactor trip breakers, including controls Plant conditions:

- Reactor power is 100%.

- RPS surveillance testing in progress.

- ICs stations in manual:

- FW Loop masters A and B.

- Delta TC.

- Reactor Master control station.

- Steam Generator Reactor Master.

- Diamond Rod Control panel.

- CRD power supply breaker associated with 'B' RPS cabinet is open, and will not reclose.

- Repair parts will take two days to arrive.

Based on these conditions, AUTOMATIC CRD Diamond Panel control can A. be established, with normal CRD IN/OUT motion control.

B. be established, however CRD OUT motion will be inhibited.

C. NOT be established, due to MOTOR FAULT condition existing.

D. NOT be established, due to SYSTEM POWER FAULT condition existing.

'd OP 1105-9, Control Rod Drive System, Section 4.4.3 Auto Inhibit, page 74, Rev. 61.

M New 'JTMIBank TMI Question #

_i Modified TMI Bank Parent Question #

3! Memory or Fundamental Knowledge J Comprehension or Analysis 3 55.41 .7 0 55.43 z 55.45 .7 A Correct answer - although only one side of CRD is powered, normal ops is possible.

B Incorrect answer - no inhibit condition exists.

C Incorrect answer - no MOTOR FAULT condition exists.

D Incorrect answer - although power fault exists, it does not prevent normal auto motion.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1

---., Operating Procedure 1105-9 Title Revision No.

Control Rod Drive System 61 ENCLOSURE 2 Page 1 of 8 4.4 Definitions

1. Asymmetric Rod Alarm The asymmetric rod alarm is indicated by the amber FAULT lamp on the PI Panel and by the main annunciator, CRD PATTERN ASYMMETRICAL. The alarm is caused by a rod being misaligned 7 inches or more from the group average position.

2A. Asymmetric Rod Fault (ASYMM RODS)

The asymmetric rod fault is indicated by the ASYMM RODS lamp on the CRD Control Panel. This indication results from a rod being misaligned from the group average by 9 inches.

28. Asymmetric Rod Runback, With the ICs in AUTO, a runback to 55 Percent power occurs on any of the following conditions:

a. Any Safety Group group in-limit signal (Le., dropped safety rod)

b. Loss of any Safety Group, QrouD out-limit signal and any asymmetric rod signal (Le. a rod off the group average by more than 9")

C. Group In-limit signal for Group 5, greater than 60 Percent power and any asymmetric rod signal (i,e., dropped rod in group 5)

d. Group 5 is >80 Percent withdrawn, Group 6 has a group inlimit signal, and any asymmetric rod signal (Le., dropped rod in Group 6 )

e. Group 6 is >80 Percent withdrawn, Group 7 has a group inlimit signal and any asymmetric rod signal (Le. dropped rod in Group 7)

3. AUTO INHIBIT Indicated control cannot be switched to AUTO until the safety groups are at the OUTLIMIT and a - large neutron error signal is not present from the ICs and ICs auto power exists.

I NOTE CRD System will trip to manual if you have an auto inhibit due to loss of ICs auto power.

74

Form ES401-6 u ROlSRO Importance Rating 4.2 4.5 Group # 2 Knowledge of the effect that a loss or malfunction of the Reactor Coolant System (RCS) will have on the following: Fuel kz From the list below, identify the ONE process that constitutes the largest (volumetric) source of non-condensable gases in the RCS during severe accident conditions.

A. Release of pre-accident dissolved hydrogen in the RCS water.

B. Release of volatile fission product gases from failed fuel.

C. Zirconium-water reaction of the fuel cladding.

D. Radiolytic decomposition of water.

LP 11.2.01.177, Gas Generation and Hazards, Section 1I.A Sources of Hydrogen in Light Water Reactors, page 4, Rev. dated 3/11/02.

None.

V. E.08.05 0 New TMI Bank TMI Question #

3 Modified TMI Bank Parent Question ##

23 Memory or Fundamental Knowledge 0 Comprehension or Analysis u

0 55.43 0 55.45 A Incorrect answer, although plausible since H2 overpressure is maintained in the Makeup Tank.

B Incorrect answer, although trapped gasses are released from failed fuel pins.

C Correct answer. Zr -H20 reaction is the major contributor.

D Incorrect answer. Plausible distracter since this process does occur in the core region.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Conte nt/S kiI Is ActivitiedNotes

~~ ~ ~

Easy way to remember: Under the right conditions, this u reaction could create a chemically-induced explosion.

The violence of the chemical reaction is a function of at least three variables: the rate at which the cladding melts; the size of the melted cladding particles; and the ratio of the quantity of cladding metal to the quality of water vapor present. If the coolant were lost from the reactor vessel due to an accident, the temperature of the fuel would increase dramatically. When temperatures reached near 22OO0F, the zirconium-water reaction would be significant in the presence of water vapor. The reaction would then proceed autocatalytically, accelerating rapidly to a temperature of approximately 3O0O0F,where actual melting occurs. Once the heat source is removed, the reaction, in the presence of liquid water, stops by itself.

However, a complete and violent reaction with water is conceivable if the entire reactor core vaporizes, an event that is considered highly improbable. In the course of the zirconium-water reaction, hydrogen gas is produced in proportion to the amount of the cladding material that has reacted. A large concentration of hydrogen would therefore indicate a large amount of damaged cladding-:? During a LOCA, the hydrogen gas may escape through a break in the reactor vessel.

Since hydrogen is lighter than the surrounding air, it will tend to rise and collect in a "bubble" at the top of the containment dome. The concentration of hydrogen in the top of the dome would be high enough to prevent oxygen from entering the bubble and creating an explosion; however, an explosion could occur while the L'

W\Word9~LAOA1120117dcc PRESENTA TlON Page 4 of 25

F o ES-401-6

~ Q # 046-Page # 3.2-4 Tier # -

2 u RO/SRO Importance Rating 3.3 -

3.6 Group # -2 Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant System (RCS): Brittle fracture.

Plant conditions:

- Reactor is tripped from 100% power due to Loss of Offsite Power (LOOP).

- RCS pressure is 1750 psig.

- RCS temperature is steady at 570°F.

- MU-P-1A is operating.

- MU-V-17 (RCS Inventory Control Valve) is failed closed.

- MU-V-217 (High Capacity Normal MU Valve) is being used to control pressurizer level.

- EG-Y-1B failed to start.

Based on these conditions, identify the EOP procedure below that is required to be implemented.

A. OP-TM-EOP-002, Loss of Subcooled Margin.

B. OP-TM-EOP-010 Guide 4, HPI Failure.

C. OP-TM-EOP-006, LOCA Cooldown.

D. OP-TM-EOP-010 Rule 6, PTS.

OP-TM-EOP-010 Rule 6, PTS, Rev. 1.

u

@I New E TMlBank TMI Question#

0 Modified TMl Bank Parent Question #

1 Memory or Fundamental Knowledge F

d Comprehensionor Analysis 55.41 .5 a 55.43 .5 55.45 .7 A Incorrect. Conditions in stem are subcooled.

B Incorrect answer. Guide 4 does not apply to the conditions presented in the stem. Plausible distracter since EG-Y-1B has failed to start.

C Incorrect answer. Plausible because if candidate misinterprets intent of EOP-001 step 4.3.

D Correct answer. Use of MU-V-217 with RCPs off is defined as a pressurized thermal shock event, IAW Rule

6. This requires the operator to minimize RCS SCM (reduce pressure) to reduce possibility of brittle fracture.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

OP-TM-EOP-010 Revision I Page 10 of 49 PTS Rule 6 Pressurized Thermal Shock (PTS]

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED If RCS -= 525OF, then VERIFY RCS cooldown - REDUCE the cooldown rate to within Tech rate has been maintained within Tech Spec Spec 3.1 2. limits.

3.1 2 . limits.

- MINIMIZE SCM IAW Guide 8.

VERIFY MU-V-1GA,B,C,D and MU-V-217 are closed.

-.-.. - . . . . - .......................... ...... .-..-............... - ......-............................................. I - If no RCPs are operating, then MINIMIZE SCM IAW Guide 8. ^ ....

Form ES-4016 u -1 RO/SRO importance Rating 2.8 -

3.1 Group #

Knowledge of the effect that a loss or malfunction of the Reactor Coolant Pump System will have on the following: Feedwater and emergency feedwater Sequence of events:

- Reactor power is initially loo%, with ICs in full automatic.

- Turbine trip causes reactor trip.

- Low grid voltage causes RC-P-1A and RC-P-1B to trip.

Based on these conditions, identify the ONE statement below that describes automatic response of the (1) Main Feedwater system and (2) EFW system.

A. (1) Main Feedwater controls both OTSG levels at 25 inches on the Startup Range.

(2) EFW remains secured.

B. (1) Main Feedwater controls OTSG 1A level at 50% on the Operating Range, and OTSG IB level at 25 inches on the Startup Range.

(2) EFW remains secured.

C. (1) Main Feedwater controls OTSG 1B at 25 inches on the Startup Range.

(2) EFW actuates and controls OTSG 1A at 50% on the Operating Range.

D. (1) Main Feedwater controls both OTSG levels at 25 inches on the Startup Range.

(2) EFW actuates, and controls both OTSG levels at 25 inches on the Startup W Range.

OPM Section F-10, Heat Sink Protection System, page 13, Rev. 9.

None.

IV.E.05.04 14 New Ci TMI Bank TMI Question #

C!Modified TMI Bank Parent Question #

31 Memory or Fundamental Knowledge 3 Comprehensionor Analysis 55.41 .7 0 55.43 55.45 .6 A Correct answer. Main Feedwater controls at normal post trip levels, 4 RCPs must trip to actuate EFW control.

B Incorrect answer. Incorrect first part, right second part.

C Incorrect answer. Both parts of this answer are incorrect.

D Incorrect answer. First part is correct, but second part is incorrect.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION F- I 0 REVISION 9

d. -

Test Panel Components Behind alarmed door panels are located 1) test switches,

2) test input jacks, 3) bypass switches, and 4) instrument selector switches. Test switches, bypass switches, and corresponding external indicators are listed on Table 1. Refer to Figure 13 to explain operation of test and bypass switches.

1. A test switch if operated places an individual channel hnction in the actuated state. To prevent switches from being left in this condition, skvitch covers ensure that the switch is properly placed prior to closing the test panel.

2. Analog input test jacks are provided where needed for calibrations

_ - 3. Bypass switches prevent channel actuation from inputting to the 2 out of 4 logic. The bypass switch does not bypass an actuated signal caused by a test switch.

4. Instrument selector switches located in the train cabinets are available to swap input instruments for EF-V-30 control (See Figure 7 & 8 and Table 1). An additional set of toggle switches is used to select which level signals will be used for console indication and potential ICs inputs.

e. Power Supplies - Power supplies were discussed in previous section.

f. ExTemaI Indicators (See Table 1) - Indicating lights on the outside of Section A2 RK4 and Section AI RKj provide HSPS status locally for use during testing.

These lights can also be used to confirm control room HSPS status indicators.

There are ( 2 ) fuses for these lights located above the train test panel.

5.1.4 Logic and Actuation The HSPS performs two types of actuation. It initiates EFW or isolates MFW.

EFW Initiation There are two trains of EFW initiation. Both trains are designed to actuate following:

a. Lo OTSG Levei (<10 S/U Level)

b. High lU3 pressure (>4 psig)

C. Lost of all 4 RCPS (pump power monitors)

d. Loss of both FWPS (FWPturb hyd oil pressure switches)

High RB pressure, loss of all 4 RCPS and loss of both EWPS will initiate EFW to both OTSGs. Low level actuation will only initiate E W to the OTSG with low level. Train A when actuated starts EF-P-2A, opens MS-V-13A & B and transfers EF-V-30A & C setpoints fiom 0 to 25. Train B starts EF-P-2B, opens MS-V-13A & B and transfers EF-V30B & D setpoint from 0 to 35. On loss of all RCPS, setpoint selection is to 50%

operating range level instead of 25 startup range.

13

Q # .048 SYSIEW 003 KA# A2.02 Page # 3.4-8 Tier # -

2 v ROISRO Importance Rating 3.7 -

3.9 Group # -1 Ability to (a) predict the impacts of the following malfunctions or operations on the Reactor Coolant Pump System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctionsor operations: Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP Plant conditions:

- Reactor power is 70%, with ICs in automatic.

- All RCPs are operating.

- No maintenance or testing in progress.

Event:

- RC-P-1A number one seal leak-off flow indication lowers from 2.2 to 0.5 gpm.

- MAP F-1-3, RCP Seal # I Leak-off Flow HilLo actuates.

- RC-P-1A shaft and motor stand vibration alarms actuate.

- Alarms reset without re-actuation, but vibrations remain excessive.

- RC-P-1A parameters:

- Labyrinth seal delta pressure is steady at 46 inches H20.

- Radial bearing temperature indication is 127°F.

- Seal # I inlet temperature is 130°F, steady.

- L2755 RC-P-1A Standpipe Level High alarm actuates.

Based on these conditions, identify the ( I ) impact on the RC-P-1A operation and the (2) required procedure section of 1203-16 Reactor Coolant Pump and Motor Malfunction to be implemented.

L/ A. (1) Secure pump within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(2) NUMBER 2 SEAL FAILURE section.

B. (1) Trip RC-P-1A.

(2) PUMP AND MOTOR VIBRATION section.

C. (1) Commence plant shutdown and secure RC-P-1A.

(2) PUMP MOTOR SEPARATION - DROPPED IMPELLER section.

D. (1) Secure pump within 30 minutes, THEN isolate #I seal leakoff within 5 minutes.

(2) NUMBER 1 SEAL FAILURE section.

AbP 1203-16, Reactor Coolant Pump and Motor Malfunction, Rev. 42.

None.

V.C.06.01 0 New 0 TMlBank TMI Question # #24 1998 SRO Ei Modified TMI Bank Parent Question # SR5C06-01-QOI 0 Memory or Fundamental Knowledge id Comprehension or Analysis e

r-55.41 .5 'a55.43 .5 3 55.45 .3/.13 A Correct answer. Symptoms support Seal #2 failure. Actions listed support this condition.

\.-

B Incorrect answer. With the limited information given, it is plausible that this can be mis-diagnosed. Actions support misdiagnosis - correct for high vibration alarm that re-actuates after the operator attempts to reset the Bentley-Nevada.

TMI SRO Exam May 2003 Friday, March 28,2003

Form ES401-6 Q # 048 C Incorrect answer. With the limited information given, it is plausible that this can be mis-diagnosed. Actions support mis-diagnosis- correct for dropped impeller.

.u D Incorrect answer. With the limited information given, it is plausible that this can be mis-diagnosed. Actions support mis-diagnosis - correct for #Seal I Failure.

Augmented distractors, stem and raised to SRO level TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 L A Abnormal Procedure 1203-16 Title Revision No.

Reactor Coolant Pump and Motor Malfunction 42 If 1210-8, RCS Superheated requires operation of a RCP, the instruction in 1210-8 supercedes the requirements in this procedure.

1.0 SEAL FAILURE 1.1 Symptoms 1.1.1 RCP Labyrinth Seal D/P Lo Alarm c 10"for RCP-IA, I B , I D only (F-1-6).

1.1.2 RCP Seal No. 1 Leakoff Flow Hi/Lo Alarm Hi > 5 gpm*; Lo c 0.86 gpm (F-1-3).

or as set IAW 1101-2, Plant Setpoints NOTE If No. 1 seal leakoff has gradually increased over a period of time, a No. 1 seal leakoff of greater than 8 gpm may NOT be indicative of seal damage.

In these circumstances, Plant Engineering should be consulted.

Continued pump operation with No. 1 seal leakoff greater than 8 gpm (but within measurable range) may be authorized provided the follow-up actions are referred to for additional auidance.

1.1.3 RCP Seal No. 1 Inlet Water Temperature Hi Computer Alarm Pts. A0525, A0526, A0527, I A0528.

1.I.4 High Stand Pipe Level Computer Alarm Pts. L2755, L2757, L2759, L2127.

1.I.5 Excessive vibration Computer Alarm Pts. L3122, L3123, L3124, L3125.

1.1.6 Excessive pump shaft or motor stand vibration indicated on PLF.

1.I.7 RCDT level rate of increase abnormal.

1.2 Immediate Action 1.2.1 Automatic Action None 1.2.2 Manual Action 1.2.2.1 No. 1 Seal failure 2

Number TMI - Unit 1

,L, Abnormal Procedure 1203-16 Title Revision No.

Reactor Coolant Purnt, and Motor Malfunction 42 NOTE As No. 1 seal failure is indicated by uncontrolled No. 1 seal leakoff and sustained Lab seal AP for RCP-IA/IB/ID <IO" H20 (seal injection still in operation) with normal radial Brg temps.

(1) REDUCE power to between 50 - 75 percent and SECURE the affected pump within 30 minutes.

(2) CLOSE No. Iseal leakoff on the affected pump within 5 minutes after pump is secured (MU-V-33 A-D).

1.2.2.2 No. 2 Seal failure NOTE A No. 2 seal failure is indicated by high standpipe level: excessive RC Drain Tank level increase, and abnormal vibration (< limits of Section 3).

(11 REDUCE power to between 50 - 75%.

(2) SECURE the affected pump within 24 hrs.

1.3 Follow-up Action 0bjective:To stabilize RCP seal performance in a manner to either permit continued operation or permit a shutdown and cooldown with minimal damage to the seal package.

NOTE If a No. 1 or No. 2 seal failure has been mitigated by securing No. 1 seal leakoff and RCP shutdown, continued Power Operation may be authorized by the Plant Operations Director at power levels up to 75 I percent.

1.3.1 DO NOT attempt to restart the idle RCP without permission of Plant Operations with concurrence from Plant Engineering Directors.

NOTE If No. 1 or No. 2 Seal Degradation has occurred as determined by abnormal seal performance developing over a period of days, the Director operations with concurrence from Plant Engineering may authorize continued operation under these circumstances, additional monitoring requirements and extended limits may be provided by the Director operations.

1.3.2 continued operation of RCP is authorized, THEN PERFORM the following:

3

Number TMI - Unit 1 b Abnormal Procedure 1203-16 Title Revision No.

Reactor Coolant Pump and Motor Malfunction 42 1.3.2.2 REVIEW the following information to determine when an RCP should be secured under these degraded conditions:

a. Without loss of seal injection, if Lab seal AP remains < 20" H 2 0 for a sustained period of time, reverse flow thru the radial bearing should be suspected. This indication is available for RCP-I N l B / I D only. Continued operation in this mode is NOT recommended. If lab AP is unstable, its value is negative and with increasing radial Brg temps. then closure of No. 1 Seal leakoff isolation valve (MU-V-33A-D) is required.

NOTE The conditions in (a) may also be an indication of turning vane diffuser bolt failure (Ref. Westinahouse Advisow 88-508).

b. If # I seal leakoff (SLO) exceeds 8 gpm continuously, degradation of # I seal is to the extent that a subsequent loss of seal injection (SI)has the potential to cause overheating of the radial bearing and seal package in a relatively short period of time. Continued operation of up to two reactor coolant pumps with sustained SLO > 8 gpm (up to a maximum of 14 gpm) is allowable as long as:

1. SLO has degraded gradually over a period of time and the response and associated parameters of the seal continues to indicate an otherwise normal performance.

2. Increased monitoring of SLO flow is performed as specified by the Plant Operations Director.

3. Additional precautions are taken to ensure that a I standby makeup pump is available to provide seal injection in the case where the operating makeup pump would trip.

C. RC Drain Tank level control considerations may restrict continued operation with No. 2 Seal degradation. Limiting aspects of RC Drain Tank control include level, temperature and maintaining additional tank capacity to receive what must be rejected from the RC Drain Tank.

4

Number TMI - Unit 1 1203-16 Revision No.

Reactor Coolant PumD and Motor Malfunction 42

d. If No. 1 Seal leakoff is low and below the recommendations of OP 1103-6, Reactor Coolant Pump Operation, it may be indicative of I either an abnormally tight No. 1 seal or most likely an abnormally open No. 2 Seal. A bucket check of No. 1 seal leakoff per OP 1103-6 is recommended under these circumstances.

Continued operation of the RCP within the guidance provided above can be expected.

2.0 LOSS OF RCP MOTOR COOLING WATER 2.1 Symptoms 2.1 .I Loss of N.S.

2.1 . I .I Motor air temp. Hi computer alarm Pts. A0699, A0705, A071 1, A0717.

2.1..1.2 RCP stator temp. Hi computer alarm Pts. A0704, A0710, A0716, A0722.

2.1.I .3 Motor upper guide bearing temp. Hi computer alarm Pts. A0702, A0708, A0714, A0720. Motor lower guide bearing temp. Hi computer alarm Pts.

A0703, A0709, A0715, A0721.

2.1.I .4 Motor up thrust bearing temp. Hi computer alarm Pts. A0701, A0707, A0713, I A0719. Motor down thrust bearing temp. Hi computer alarm Pts. A0700, A0706, A0712, A0718.

2.2 Immediate Action 2.2.1 Automatic Action None 2.2.2 Manual Action 2.2.2.1 -

IF N.S. is lost AND any of the following temperatures are exceeded: I 0 Motor Stator > 150°C. (302°F) I e Motor Radial Bearings > 185°F I e Motor Thrust Bearings > 195°F THEN PERFORM the following:

(11 REDUCE power to 50 - 75 percent (2) TRIP the affected pump 5

Number TMI - Unit 1 L' Abnormal Procedure 1203-1 6 Title Revision No.

Reactor Coolant Pump and Motor Malfunction 42 3.0 PUMP AND MOTOR VIBRATION 3.1 Symptoms Pump shaft vibration alarms are Alert: 15 mils, Danger: 20 mils.

Motor stand vibration alarms are Alert: 3 mils, Danger: 7 mils. Setting the toggle switch to "Trip Multiply" on PLF doubles the alarm limits and 3.1. I Excessive pump shaft or motor stand vibration indicated on PLF.

3.1.2 High vibration computer alarm points L3122, L3123, L3124, L3125.

3.2 Immediate Action

'V 3.2.1 Automatic Action None 3.2.2 Manual Action 3.2.2.1 E a motor stand vibration alarm is received, THEN RESET motor stand vibration alarm on Bently Nevada System on I PLF.

3.2.2.2 E a shaft vibration alarm is received, I THEN RESET shaft vibration alarm on Bently Nevada System on PLF I 3.2.2.3 vibration alarms reoccur immediately, I THEN PERFORM the following:

(1) REDUCE power to 50 - 75 percent.

(2) TRIP affected RCP.

3.3 Follow-up Action Objective: To prevent damage to RCP due to excessive vibration.

3.3.1 DETERMINE cause and repair.

6

Number TMI -Unit 1 W Abnormal Procedure 1203-16 Title Revision No.

Reactor Coolant Pump and Motor Malfunction 42 4.1 .I Excessive vibration due to backstop failure.

4.1.2 Computer points L2875, L2876, L2877 or L2878 for applicable RC pump in alarm.

4.2 Immediate Action 4.2.1 Automatic Action None 4.2.2 Manual Action 4.2.2.1 START the oil lift systems on the affected pump.

4.3 Follow-up Actions Objective: To place the plant in a safe condition and prevent further damage to the RCP.

.. 4.3.1 Upon orders from the Shift ManagerKontrol Room Supervisor, COMMENCE a Normal Plant Shutdown and Cooldown of the RCS per 1102-10, Plant Shutdown, and 1102-1I, Plant Cooldown.

4.3.2 STOP all the Reactor Coolant Pumps, as soon as possible, in accordance with 1102-11 Plant Cooldown.

4.3.3 DETERMINE the cause and repair 5.0 LUBE OIL EMERGENCY 5.1 Symptoms 5.1.1 Low lift pump discharge oil pressure alarm 1750 psig on computer and affected RCP trouble ann. alarm (F-2-1, 2, 3,4).

5.1.2 High bearing temp. alarms on computer RCP A B C D Upper Guide A0702 A0708 A0714 A0720 Lower Guide A0703 A0709 A0715 A0721 Up Thrust A0701 A0707 A071 3 A0719 Down Thrust A0700 A0706 A0712 A0718 7

I I Number Reactor Coolant Pump and Motor Malfunction 42 5.2.2.1 VERIFY auto start of D.C. oil lift pump.

5.2.2.2 -IF D.C. oil lift pump does not start, THEN START the pump manually.

5.2.2.3 -

IF high motor bearing temps (Radial >185", thrust > 195°F) are being approached, THEN PERFORM the following:

(1) REDUCE reactor power to 50 - 75 percent.

(2) TRIP the affected pump.

5.3 Follow-up Action Objective: To prevent damage to RCP due to loss of oil to bearings.

5.3.1 low flow exists on lift and/or backstop system, THEN START lift andlor backstop pumps on affected RCP.

5.3.2 Hi/Lo levels exist in oil pots, THEN PERFORM the following:

5.3.2.1 REDUCE reactor power to 75 percent.

5.3.2.2 SECURE affected pump.

8

TMI - Unit 1 L' Abnormal Procedure 1203-16 Title Revision No.

Reactor Coolant Pump and Motor Malfunction 42 i NOTE A possibility of an oil fire exists when the RCP oil system ruptures. The plant must be cooled down to <400 O F which is the flash point of the oil in this system.

5.3.3 E an oil system has ruptured and a possibility of oil fire exists, THEN SHUTDOWN AND COOLDOWN the plant to < 400 O F in accordance with 7 102-10, Plant Shutdown, and 1102-11< Plant Cooldown.

- 5.3.4 DETERMINE cause and repair. I 6.0 -

PUMP MOTOR SEPARATION DROPPED IMPELLER 6.1 Symptoms 6.1 .I Low motor current ind. above RCP U S . I 6.1.2 Low R.C.S. flow ind. on Console "CC".

6.1.3 Hi vibration:

6.1.3.1 Motor stand on computer, Pts L3122, L3123, 13124, L3125.

6.1.3.2 -

Shaft Bently Nevada Sys. -Alarm lights.

6.2 Immediate Action 6.2.1 Automatic Action 6.2.1.1 Possible reactor trip on power to flow.

I 6.2.2 Manual Action 6.2.2.1 COMMENCE Plant Shutdown.

6.2.2.2 SECURE affected RCP.

6.3 Follow Up Action I Objective: To place the plant in a safe condition.

6.3.1 PROCEED to cold shutdown per OP 1102-1, Plant Cooldown 6.3.2 DETERMINE cause and repair.

9

SYSIEP# 004 KA# 2.4.4 Page # 2-11 Tier # -2

-u ROlSRO importance Rating 4.0 -

4.3 Group # -1 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures: Chemical and Volume Control System (CVCS)

Plant conditions:

- Reactor power is 1OO%, with ICs in full automatic.

- RCS T-ave is 579°F.

- Make up tank level is LOWERING at l"/minute.

- MU-V-17 is in manual control.

- Letdown flow is constant at 45 gpm.

- Indicated RCP total seal injection flow is 38 gpm.

- MU-V-32 is in automatic.

- RCP labyrinth seal D/P indicators are low off-scale (negative).

- Auxiliary Building airborne activity is rising.

Based on these conditions, diagnose the cause for the abnormal event and the identify proper response.

A. RCP seal # I leak-off flow is aligned to the Auxiliary Building sump; notify supervision for investigation and perform line-up per 1104-2 Makeup and Purification System B. RCP seal #1 leak-off flow has been isolated by closure of MU-V-26; notify supervision for investigation and open MU-V-26.

C. RCP total seal injection flow transmitter has failed; notify I AND C supervision for assessment and repair.

L, D. RCP seal injection flow is not reaching the RCPs, initiate 1203-15 Loss of Make-up/Seal injection.

P 1203-15, Loss of R.C. MakeuplSeal Injection, page 2, Rev 26.

Ci New 0 TMlBank TMI Question # SR4A05-04-QOI E l Modified TMI Bank Parent Question #

L l Memory or Fundamental Knowledge Comprehension or Analysis 2

l 55.41 .10 55.43 .2,.5 27 55.45 .6 A Incorrect answer. Answer does not support all symptoms provided, but it is plausible as a distracter due to rising Auxiliary Building activity concurrent with the reduction in Makeup Tank level.

B Incorrect answer. Answer is not consistent with rising Auxiliary Building activity concurrent with the reduction in Makeup Tank level. Plausible distracter since this would result in low labyrinth seal differential pressure.

C Incorrect answer. This answer is not consistent with low labyrinth seal differential pressure condition in the stem.

wer. Symptoms presented are evidence of a leak downstream of the seal injection flow sensor.

L, None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMl NUCLEAR Abnormai Procedure 1203-15 Revhion No.

L o s s of R.C. RllakeudSd Inlaction 26

.I

.Q SYMPTOMS 1.1 Makeup flow indication low as indicated on MU24FI on console "CC:".

1.2 MU Pump discharge header pressure high (3100 PSIG) or low (2400 PSIG) as indicated on P.I.

MU2 on console "CC".

1.3 Seal injection flow to RCP seals less than 22 gpm as indicated by MU42 FI on console 'CC".

1.4 RCP seal totat injection flow HilLo Alarm F-1 Law < 22 gpm.

4.5 Increasing No. 1 Seal Inlet (RC20-TE) and Radial Bearing Temperature. (RCl STE) computer aierns on PTS A0521-AO528.

NOTE The conditions in (1S) may also be an indication of turning vane diffuser bok failure (Ref. WestinghouseAdvisory 88508).

1.6 -

RC pump lab seal DP Lo Alarm F-1-6 (5 W )for RCP-IA, I&

seal AF will accelerate deposition of crud on #seal ID. Less than about 20 Inches lab I (electrophoresis). t 2.0 IRRNlEDIATE ACTION A. Automitic Action 1.. Decreasfng flow caused MU-V32 RCP seal InJection flow control valve to open and decreasing pressurizer level will open MU-V17.

2. . Low seal injection fIow (e22 gpm) with low ICCW flow (< 550 gpm) will CRUS@ trip of all RCP's after short Ume delay.

B. Manual Action I. Determine cause for loss of RC. Makeup/Seal Injection

a. Running make up pumpfAps (Green and amber control switch light)

b. MU-VI 7 closed. (Zero flow on MU24FI with MU-PI B operating.)

C. MU-V32 dosed. (Zem flow on MU42FI with MU91 B operating).

J u 2

Page ## 3.2-15 Tier # -

2

'W ROISRO Importance Rating 4.3 4.1 Group # -

I Ability to manually operate andlor monitor in the control room: Emergency borate valve 7

/BI Plant conditions:

- LWDS Panel Bleed Tank, Feed Tank, and Feed Pump selector switches in OFF position.

- LWDS Panel Deborating Demin selector switch in OFF position.

Based on these conditions, identify the ONE selection below that completes the following statement:

Backup (RBAT) emergency boration valve WDL-V-61 will open when A. MU-V-10, makeup batch isolation valve, is opened.

B. the Control Room Boric Acid Injection switch is positioned to INJECT position.

C. the batch is set and "Enter" is pushed on the Makeup System totalizer batch controller.

D. boric acid injection pump CA-P-I A or CA-P-1B is started with pump selector switch in "LWDS" position.

209-239, WDL-V-61, Rev. 1.

209-185 Deborating Demin. Sel. Sw. and Boraic Acid Inject Sel. Sw, Rev. 1.

None.

.09.25

'W a New E TMI Bank TMI Question #

E Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge 3 Comprehension or Analysis il 55.43 55.45 .5/.61.71.8 A Incorrect answer. There is no control interface between the valve control circuit and the batch controller.

Plausible distrcater because of the process flow path used to borate the Makeup Tank.

B Correct answer.

C Incorrect answer. There is no control interface between the valve control circuit and the batch controller.

Plausible distrcater because of the process flow path used to borate the Makeup Tank.

D Incorrect answer. There is no control interface between the valve control circuit and the batch controller.

Plausible distrcater because of the process flow path used to borate the Makeup Tank.

None.

TMI SRO Exam - May 2003 Friday, March 28, 2003

...- . -..La-

- MOTOR, DEWCES MOUNTED LOCALLY (NEAR OR AT VALVE, ETC.)

AT, 37 c3 BORIC ACID MIX TK. TO PRIMARY SYSTEM WDL-V61 BORIC ACID MIX 3A CLOSE P. B.

I SS-209-182

'I N1 IT 33Iw

- I 331bo

+20:;Rl; 4c ARC 0 A

SS-209-185 4

33x61 20x62 i

f 0

m m

0 I

E Q

0 f

r VI F

3 r;

rn U

r m

0 w

A 5

E m

s I

Form ES401-6 SYSIEP# 005 KA# K6.03 Page # 3.4-12 Tier # -

2 u ROlSRO Importance Rating 2.5 2.6 Group # -3 Knowledge of the effect of a loss or malfunction of the following will have on the Residual Heat Removal System (RHRS): HRH heat exchanger Plant conditions:

- Reactor is in Cold Shutdown condition.

- Decay Heat Removal Train A is operating.

- Decay Heat Closed Cooling flow through the Decay Heat Removal cooler is throttled to maintain the RCS at 130°F.

- Total loss of InstrumentAir (0 psig) occurred.

Identify the ONE statement below that describes the response of the cooling system and subsequent effect on RCS temperature for this situation.

A. Closure of DC-V-65A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in RCS heatup.

B. Opening of DC-V-65A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in RCS cooldown.

C. Closure of DC-V-2A (Cooler inlet) results in RCS heatup.

D. Opening of DC-V-24 (Cooler inlet) results in RCS cooldown.

EP 1202-36, Loss of InstrumentAir, page 6, Rev. 31.

- c3 New a TMIBank Ci Modified TMI Bank TMI Question #

Parent Question #

81 7/2001 SRO 0 Memory or Fundamental Knowledge

J Comprehension or Analysis M 55.45 .7 55.41 .7 [7 55.43 A Plausible distracter since closure of both valves (incorrect response for loss of Instrument Air) will result in loss of cooling and therefore RCS heatup.

I3 Plausible distracter since opening both valves (incorrect response for loss of Instrument Air) could possibly result in RCS cooldown.

C Plausible distracter since closure of DC-V-2A (incorrect response for loss of InstrumentAir) would result in RCS heatup.

D Correct answer.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

TMI Unit I I Number

\u Emergency Procedure 1202-36 Title Revislon No.

Loss of Instrument Air 31 Open MU-V-'I 10 13.2 Dispatch an auxiliary operator t o MU-V-3 and establish communication with the Control Room. I 13.3 Monitor letdown flow and prefilter A P and slowly re-establish letdown flow at 2.5 gpm/min.

(unless waived by the SWCRS) by opening MU-V-3.

I

14. If a normal seal return flowpath is desired, perform the following:

I 14.1 Dispatch an auxiliary operator to MU-V-26 and establish communication with the Control Room. t 14.2 14.3 Align the local handwheel and insert pin to engage the operator with the handwheel.

Open MU-V-26.

I

15. Upon verified loss of both Seal Injection and Intermediate Cooling Water to the RC pump seals:

I A. Verify tripped or trip the Reactor.

I Verify tripped or trip the Reactor Coolant Pump.

B C Go to ATP 1210-1 and refer to 1202-36for Loss of Instrument Air.

16. If Reactor is shutdown, verify 2 I% 4tVK shutdown margin. Use BWST and MU-V-14A to borate as necessary. I If on DHR, Monitor DHR suction and return temperatures throttle flows or secure pumps as 17.

necessary to minimize RCS temperature transient. DC-V-2's fail open and DC-V-65's fail closed providing full DC flow through the cooler.

1 NOTE T h e turbine jacking gear must be manually engaged locally when 1A-V-26 is shut.

18. Verify IA-V-26 closed if pressure is below 60 psig.

I If CO-V-8 fails open and condenser hotwell level is high, close CO-V-13.

19.

20. If Decay Closed Surge Tanks indicate high level verify DC-V-'lSA and B failed open due to loss of air I

and dose DC-V-20A and 5 to isolate surge tank makeup.

6

Form ES-401-6 Q# 052

- SYSIEP# 005 ROlSRO Importance Rating KA#

3.5 A1.O1 3.6 Page # 3.4-12 Tier #

Group #

2 3

Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the Residual Heat Removal System (RHRS) controls including:

Heatup/cooldown rates Plant conditions:

- Reactor shutdown, cooling down for refueling outage.

- RC-P-1C is the only RCP operating.

- 'A' Decay Heat Removal loop in operation.

- Decay Closed flow is fully bypassing 'A' Decay Heat Removal cooler.

Based on these conditions, identify the ONE parameter below that is required to be monitored and controlled in order to prevent exceeding RCS design limits after securing RC-P-1C.

A. Core exit thermocouples.

B. RCS Loop A wide range hot leg temperature.

C. ' A Decay heat cooler outlet temperature DH2-TE-1.

D. 'A' Decay heat pump suction temperature DH6-TE-1.

Technical Specifications Figure 3.1-1, page 3-5a, Amendment 234.

Technical Specifications Figure 3.1-1, page 3-5a, Amendment 234

- 8 New 2 TMIBank TMI Question #

0 Modified TMI Bank Parent Question #

&!I Memory or Fundamental Knowledge C Comprehension or Analysis 0 55.43 Ez 55.45 .5 A Incorrect answer. Plausible misconception that we would use Core outlet temperature.

B Incorrect answer. Plausible misconception that we would use Loop A hot leg temperature, since the DH drop line connects to Loop A hot leg.

C Correct answer.

D Incorrect answer. Plausible misconception that we would use Reactor Vessel outlet temperature.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

i ire 3.1-1 Reactor Coolant System Heatup/Cooldown Limitations 3000 2800 P

?!, ,2600 cn

13) '6

--I, a2400 IW on.

7 2200 2 3 moo Q ) L Cnb ne 9 1800 a d a 1600 fig u r m 1400 T > 255F t WFM or 15M MI. Steps a m TsmF 3 OW or 15F130 hfn, Steps 5 g 1200 -

4 RC Rnp CantIinatbns for Hatups:

--I No Rc Fuwp Ope,ratiig 100<Ts212 Any t ~2RarpCbddnat!m(llOIQll,~,012,1M) a 3 1000 212+Ts 329 Any RwpCombinaiion except212 si Any Rnp Conbination 3 800 -

5 RC Wrrp Corrbinatims for Cooklow ns:

600 400 50 100 150 200 250 300 350 400 450 500 550 600 Indicated Reactor Coolant Inlet Temperature (F) i

Page # 3.5-2 Tier # -

2

\/ ROlSRO Importance Rating 2.6 2.9 Group # -3 Knowledge of Pressurizer Relief TanWQuench Tank System design feature(s) and/or interlock@)which provide for the following: Quench Tank Cooling From the list below, identify the ONE signal that will initiate an automatic open signal for IC-V-20, RC Drain Tank Cooler cooling outlet valve.

A. RC Drain Tank pressure greater than the high pressure alarm setpoint.

B. RC Drain Tank level greater than the high level alarm setpoint.

C. Both RC Drain Tank containment isolation valves open (WDL-V-304 and WDL-V-305).

D. RC Drain Tank pump running.

SS-208-517, Reactor Coolant Drain Tank Outlet, IC-V-20, Rev. 2 El New @ITMI Bank TMI Question # AL4B03-05-QO5 17 Modified TMI Bank Parent Question # AL4B03-05-QO5

@ Memory or Fundamental Knowledge 0 Comprehension or Analysis c 55.45

-1 PI 55.41 .7 ci 55.43 A Incorrect answer. Plausible misconception of automatic pump down on high tank level to avoid bursting the rupture disk.

B Incorrect answer. Plausible misconception of automatic pump down on high tank pressure to avoid bursting the rupture disk.

C Incorrect answer. Plausible distracter since opening both of these valves aligns the pump down path to the MWST.

D Correct answer.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

T'HREE M I L E ISLAND N U C l E A R STATJON UNIT # 1 - CHK'D. PF B pII#NEoIwAHD~ANrs MOTOR OPERATED VALVE TITLE NOTE 4

'SPACE HTR.

I 1

LS16

I .: !i .

ERENCE.:.DWGS.:

t b

. . + 0..-/z-053-WEX:SS-208-403 1 .

NOTES:

_i_

. 1.. THE SET P O W S OF LS9 THRU IZ!:&t S 1 3 e

MET LECWUX: sk-m-ooi

. , ' . . THRU tS16 ARE ACJ33TABX AT M Y PECCENTAGE THR.EE #I L E

, I.. _- .- ., 2 O F VALVE WSITIOH.

. ._ . . i ~ o ~ i i ON t ~ LWDS o PANEL ID +tiit; 1 '-

1 1

' 3. ROTOR-+I -- L I M I T Si;. I TO 4. ' . i ROTOR $2 L.IMIT,'SWS. 5 TO 8 , : ~ i ELECT i .

, i ROTOR.#3 LlMIT!Sii 5 TO 12

1;'i ROTOR $4 LIMIT-SWS. 13.m 16. i ,

.. _I '

y. CONNECT WV. SPACE HEATER TO-,12OV L I N E ~ A S ~ S H O W N

- v Ec@WDLP8 -L 4.21WDLP8

'. SS-209- 172 I I !

a NOTE 2 LS 3 331 bo R'C G LS17

-TQC LS7 LS 4 33/ bo

- 7 **

2- 42-C

Page # 2-7 Tier # -2

.d ROlSRO Importance Rating 3.4 4.1

- Group # -

3 Knowledge of limiting conditions for operations and safety limits: Component Cooling Water System (CCWS)

Identify the ONE selection below that completes the description of the basis for the limiting conditions for operation (LCOs) for Nuclear Services Closed Cooling (NSCC).

(a) NSCC pump(s) idare required for normal operation heat loads; (b) NSCC pump(s) idare required for ECCS support during a LOCA.

A. (a)Two (b) Two B. (a) Two (b) One C. (a)One (b) Two D. (a)One (b) One Technical Specifications page 3-24, Amendment 227.

.d a New 3 TMIBank TMI Question #

3 Modified TMI Bank Parent Question #

3l Memory or Fundamental Knowledge Comprehension or Analysis 0 55.41 a 55.43 .2 Fzl 55.45 .2 A Incorrect answer. First part is correct that two are required for normal operations, but only one is needed for ECCS support (incorrect second part).

B Correct answer, in accordance with Tech Spec bases..

C Incorrect answer. First part is incorrect (two are required for normal operations), but only one is needed for ECCS support (correct second part).

D Incorrect answer. First part is incorrect (two are required for normal operations), but only one is needed for ECCS support (correct second part).

None.

TMI SRO Exam May 2003 Friday, March 28,2003

AND REACTOR BUflDlNG SPRAY SYSTEMS (Contd.)

Bases (Contd.)

Maintaining MUT pressure and level within the limits of Fig 3.3-1 ensures that MUT gas wifl not be drawn Into the pumps for any design basis accident. Preventing gas entrainment of the pumps is not dependent upon operator actions after the event occurs.

The plant operating limits (alarms and procedures) will include margins to account for instrument error.

The post-accident reactor building emergency cooling may be accomplished by three emergency cooling units, by two spray systems, or by a combination of one emergency cooling unit and one spray system. The specified requirementsassure that the required post-accident components are available.

The iodine removal function of the reactor building spray system requires one spray pump and sodium hydroxide tank contents.

The spray system utilities common suction lines with the decay heat removal system. Ifa single train of equipment is removed from either system, the other train must be assured to be operable in each system.

When the reactor is critical, maintenance is allowed per Specification 3.3.2and 3.3.3 provided

'v requirements in Specification 3.3.4are met which assure operability of the duplicate Components. The specified maintenancetimes are a maximum. Operability of the specified components shall be based on the satisfactory completion of surveillance and inservice testing and inspection required by Technical Specification 4.2 and 4.5.

The allowable maintenance period of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months may be utilized if the operability of equipment redundant to that removed from senrice is verified based on the results of surveillance and inservice testing and inspection required by Technical Specification 4.2 and 4.5.

Inthe event that the need for emergency core coofing should occur, operation of one makeup pump, one decay heat removal pump, and both core flood tanks wiil protect the core. In the event of a reactor coolant system rupture their operationwill limit the peak clad temperature to less than 2,200 O F and the metal-water reaction to that representingless than 1 percent of the clad.

Two nuclear service river water pumps and two nuclear service closed cycle cooling pumps are required for normal operation. The normal Operating requirements are greater than the emergency requirements following a loss-of-coolant.

REFERENCES (1) UFSAR, Section 6.1 - "Emergency Core Cooling System" (2) -

UFSAR, Section 14.2.2.3 "Large Break LOCA" 3-24 u

Amendment No. $42, '4-49, 4!Sl 227

FOWII ES-401-6 SYSIEP# Q lJ K A # A2.08 Page # 3.2-23 Tier # -2 u ROlSRO Importance Rating 2.6 2.8 Group # -

2 Ability to (a) predict the impacts of the following malfunctions or operations on the Pressurizer Level Control System (PZR LCS) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of level compensation.

Initial conditions:

- Reactor power loo%, with ICs in full automatic.

- Pressurizer level and temperature are normal.

Sequence of events:

- Temperature compensation to the selected Pressurizer level instrument fails LOW.

Based on these conditions, identify the ONE statement below that describes:

( I ) Automatic response of the Pressurizer level control system; (2) Immediate manual actions required.

A. (1) Makeup valve MU-V-17 OPENS to loo%, Pressurizer heaters DO NOT trip.

(2) Transfer MU-V-17 to manual and adjust to maintain Makeup Tank level constant.

B. (1) Makeup valve MU-V-17 OPENS to loo%, Pressurizer heaters trip.

(2) Raise RCS Letdown flow to maximum (140 gpm).

C. (1) Makeup valve MU-V-I7 CLOSES to 0%, Pressurizer heaters energize in L response to actual Pressurizer level reduction.

(2) Isolate RCS Letdown flow.

D. (1) Makeup valve MU-V-17 CLOSES to 0%, Pressurizer heaters energize in response to actual Pressurizer level reduction.

(2) Raise RCP Seal injection flow to compensate for reduced Makeup flow.

EP 1202-29, Pressurizer System Failure, pages 14 and 19, Rev. 59.

None.

V.D.11.01 L l New TMI Bank TMI Question # QR5D11-03-QOI 0 Modified TMI Bank Parent Question #

Ci Memory or Fundamental Knowledge Comprehension or Analysis 2 55.41 .5 lid 55.43 .5 M 55.45 .3/.13 A Correct answer, IAW EP 1202-29, Pressurizer System Failure B Incorrect answer. Impact of temperature compensation at these conditions will not reduce indication below 80-inch low level cut-off interlock.

C Incorrect answer. Failed value will be less than normal 220-inch level, therefore valve will open. Action u described is not in accordance with EP 1202-29.

D Incorrect answer. Failed value will be less than normal 220-inch level, therefore valve will open. Action described is not in accordance with EP 1202-29.

TMI SRO Exam - May 2003 Friday, March 28,2003

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI Unit 1 Emergency Procedure 1202-29 w Title Revision No.

Pressurizer System Failure 59 1.2 Rapid change in indicated/recorded level due to loss of compensation or loss of power or d/p cell failure or other malfunction, of the pressurizer.

1.3 Possible high or low pressurizer level alarms.

0 G-1-5, Pzr Level Hi-Hi 0 G-2-5, Pzr Level HVLo 0 G-3-5, Pzr Level Lo-Lo I.4 Pressurizer level indicator@)NOT responding to changes in pressurizer level.

1.5 Hi makeup flow alarm (D-3-1, MU Flow Hi).

1.6 Pressurizer temperature fails to agree with saturation temperature for RCS pressure.

1.7 RCS pressure changes does NOT agree with PZR level changes.

2.0 IMMEDIATE ACTION 2.1 Automatic Action 2.1.I If indication fails low

a. Pressurizer heaters trip at 80 inches.

b. Makeup valve MU-V-17 opens.

2.1.2 If indication fails high

a. Makeup valve MU-V-17 closes.

2.2 Manual Action 2.2.1 TAKE MU-V-17 under hand control& A ADJUST makeup flow to equal letdown flow minus seal injection to maintain makeup tank as constant as possible.

2.2.2 SELECT alternate pressurizer level transmitter.

2.2.3 SELECT alternate pressurizer temperature transmitter.

14

61 13A37 3 0 S3H3Nl 031VSN3dW03 OOP OS OOE: os2 002 os 1 00 1 0s 0 0

0s 001 os1 002 os2 rn OSC Z I

I I

1 1 I I 1 009

SYSIEP# 012 KA# K4.05 Page # 3.7-2 Tier # -2 u 2.9 Group # -

2 ROlSRO Importance Rating 2.7 -

Knowledge of Reactor Protection System design feature@) andlor interlock(s) which provide for the following: Spurious trip protection From the list below, identify the ONE selection that describes the Reactor Protection System (RPS) design features that provide:

(I) Spurious trip protection; (2) 100% reactor power on-line testing capability.

A. (1) 2 out of 4 trip logic (2) manual bypass B. (1) 2 out of 4 trip logic (2) shutdown bypass C. (1) electrical independence (2) module removal D. (1) electrical independence (2) shutdown bypass OPM Section F-02, Reactor Protection System, pages 6 and 7, Rev. 8 (IO).

B New TMIBank TMI Question #

u El Modified TMI Bank Parent Question #

LEI Memory or Fundamental Knowledge c3 Comprehension or Analysis 3 55.43 3 55.45 A Correct answer. 2 of 4 logic prevents spurious trips if one channel actuates, also allows testing while still meeting TS degree of redundancy. Manual bypass allows channel to be removed from trip logic to enable testing.

B Incorrect answer - first part correct, second part wrong, used for startup/shutdown, not testing at full power C Incorrect answer - plausible first part, as it does aid in preventing spurious trips; second part causes string to trip.

D Incorrect answer - plausible first part, as it does aid in preventing spurious trips; wrong second part - used for startup/hutdown, not testing at full power.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION F-02 REVISION s

3.0 INTRODUCTION

u The main intent of the Reactor Protection System is to protect the Reactor Core against fuel dadding damage and Coolant System from overpressure damage. This is performed by monitoring both nuclear and non-nuclear instrumentation signals. These signals are continuously compared with selected pre-set and calculated values which represent the safety system settings. There are four independent and redundant channels which compose the total Reactor Protection System(see Figures 8 and 9). Whenever two of the four channels agree that the safety system setting has been reached or exceeded a Reactor Trip signal is initiated.

The RPS trip string is shown in Figure 9.

The Reactor Protection System also provides a secondary finction for the nuclear and non-nuclear instrumentation via Control Room indication and control system input.

Nuclear Instrumentation is defined as an instrument which monitors the nuclear reaction. AIthough a11 three nuclear instrumentation strings (Le., Source Range, IntermediateRange, and Power Range) are contained within the RPS Cabinets only the Power Range Channel is considered as an input or the RPS. Nuclear instrumentation string module description will be presented here only as an aid for the student when going through WS cabinets.

Non-Nuclear Instrumentation is defined as an instrument which is not designed to monitor a nuclear reaction

[i-e., Pressure,Temperature and Flow Instruments). The following non-nuclear instruments are considered part of the RPS:

a. Reactor Coolant Loop Outlet Temperatures (THoT)

b. Reactor Coolant Loop Pressures (Narmw Range)

c. Reactor Coolant Loop Flows

--d d. Reactor Building Pressure 4.0 PRESENTATION

4. I Design Criteria Since the Reactor Protection System performs such an important protective finction, the design criteria applied are among the most extensive and restrictive of any applied to plant systems to insure reliable system operation. The design criteria for the Reactor Protection System was established from ,

IEEE-279 standard, August 1968 revision, Criteria for Nuclear Power Plant Protection Systems.

4.2.1 Single Failure No single failure of a component shall prohibit the protective system fiom performing its function. No single failure of a component shall initiate unnecessary protective action.

4.1.2 Redundancy There are (4) four separate and independent Reactor Protection System channels. The design criteria applied requires that the Reactor Protection System channels and their associated inputs and outputs shall be nt, to prevent any fmlt in one m affecting it eIectrical or rnec 6

SECTION F-02 REVISION 8 4.1.3 Electrical Independence and Separation Each Reactor Protection System channel i s powered from a different Vital Bus, so that a power supply fault can only affect one channel at a time. In order to prevent faults on output signal lines to the Non-Nuclear Instrumentation System, Integrated Control System and plant computer from the Reactor Protection System, isolation amplifiers andlor auxiliary relays separate the interconnecting points.

NOTE See Table 1 for listing of RPS power supplies.

4.1.4 Physical Independence and Separation Reactor Protection System channel cabinets, sensors, power cables and signal wiring are separated and physicaIIy protected to maintain redundancy and separation, thereby reducing the likelihood of a single event affecting more than one channel.

I 4.1.5 Manual Trip Manual trip can be accomplished .from the control console by a trip button which is independent of the automatic trip system and not subject to most of the failures which the automatic system is subject to.

4.1.6

'- The Reactor Protection System can be tested at all levels while shutdown or during noma1 operation without interfering with normal plant operation or trip functions. The RPS channel testing is performed with the channel in the chatme[ bypass mode. Ifany channel k in bypass, the trip logic is 2 out of 3 instead of the normal 2 out of 4. '

I NOTE See Table 2 for a list of Reactor Protection System related surveiIlance procedures.

4.1.7 Loss of Power Failure of a Vital Bus or a Channel Power Supply will Muse the affected channel to trip.

Whenever a Reactor Protection System is not in a trip condition, the trip string relays are energized providing a current path; therefore, the loss of power to the energized string from either the Ioss of the incoming vital bus power, loss of the internal DC power supply or a trip signal from one ofthe bistables will result in a channel trip.

4.1.8 Equipment Removal Removal of any module which is an integral part of the trip string initiates a channel trip.

Removal of auxiliary output modules or modules not definedeto be part of the Reactor Protection System (ie., Source and Intermediate Range Modules) will not initiate a Channel Trip. The removal of a Reactor Trip module trips the associated Control Rod Drive Breaker.

7

orm ES-401-6 SYSIEP# 012 KA# K6.11 Page ## 3.7-3 Tier # -

2 W 2.9 ROlSRO Importance Rating - Group # -

2 Knowledge of the effect of a loss or malfunction of the following will have on the Reactor Protection System: Trip setpoint calculators Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- No surveillance testing in progress.

Based on these conditions, identify the ONE statement below that describes an RPS cabinet trip string input failure that will cause the RPS CHANNEL to trip.

A. Contact Monitor switches to '3 RCP' setpoint.

B. Total RCS flow fails to 160 x E6 Ibmlhr.

C. Loop ' A RCS flow fails to ZERO Ibmlhr.

D. Power imbalance fails to ZERO %.

OPM Section F-02, Reactor Protection System, page 11, Rev. 8 (IO).

None.

IV.E.14.06 F! New TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

u 0 Memory or Fundamental Knowledge Comprehension or Analysis 2l 55.41 .7 0 55.43 3 55.45 .7 A Incorrect answer. This failiure mode will not result in a change to the high flux trip based on number of RCPs running, and therefore a channel trip will not occur. This high flux trip setpoint for 4 RCPs operating is the same as the setpoint for 3 RCPs operating.

B Incorrect answer. This failure mode will raise the overpower trip setpoint based on total RCS flow and power imbalance, and a channel trip will not occur.

C Correct answer. Zero (loop) flow with normal imbalance will generate an output (flux trip setpoint) signal that is less than 100% power, causing the RPS channel to trip on Flux/Flow/lmbalance.

D Incorrect answer. This failure mode will raise (rather than lower) the overpower trip setpoint based on total RCS flow and power imbalance, and a channel trip will not occur.

None.

TMI SRO Exam May 2003 Fridqy, March 28,2003

SECTION F-02 REVISION 8 4.4 Nuclear Overpower based on Flow and Imbalance 4.4.1 Tech Spec Setpoint: See Figure 4

4.4.2 Basis

The power level trip setpoint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power.

Analysis has demonstrated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any malfunction.

The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip setpoint produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations are as follows:

a. Trip would occur when four reactor coolant pumps are operating if power i s 108%

and reactor flow rate is 1OOV, or flow rate is 92.5% and power level is 1OO%$

b. Trip would occur when three reactor coolant pumps are operating if power is 80.6%

and reactor flow rate is 74.7% or flow rate is 69.4% and power level is 75%.

C. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 53.1 % and reactor flow rate is 49.2% or flow rate is 45.3% and the power level is 49%.

For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits '

from being exceeded. These thermal limits are either power peaking kW/ft Iimits or DNBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 4 are produced. The power-to-flow ratio reduces the power level trip and associated reactor powerheactor power-imbalance boundaries by 1.08% for a one percent flow reduction.

4.4.3 Description of circuit: Three specific parameters make up the Nuclear Overpower based on Flow and Imbalance trip; they are, as the title implies, reactor power, axial power imbalance and reactor coolant system flow rate. Imbalance and flow signals are inputs to the function generator. The function generator compares the two inputs and puts out an allowable power level signal based on imbalance and flow. This signal is then compared with the actual reactor power level, if the actual reactor power level exceeds the allowable power level the bistable associated with this protective parameter will trip initiating a channel trip (see Figure 6).

11

SECTION F-02 REVISION 8 FIGURE 6 - RCS =OW, RCP STATUS, & RX POWER LOGIC POWER A LOOP F U M - - SQUARE IMBALANCE D/P - ROUT - I EXTRACTOR i

FLOW FUNCTION WMUS 6,STABLE

-- GENERATOR- Rrm TRiP B LOOP FLOW : SQUARE b 8.

DIP ROOT

--.F,EXTRACTOR ROW TEST CIRCU IT RANGE

[CHANNEL I It I I I I m TRIPS MONITOR C 0N f ACTS CONTACT 27

Page ## 3.7-6 Tier # 2 u' ROlSRO Importance Rating 3.4? -

3.6? Group # -1 Knowledge of Nuclear Instrumentation System design feature(s) and/or interlock(s) which provide for the following: Slow response time of SPNDs.

E 0 1203-7 (Hand Calculations for Quadrant Power Tilt and Core Power Imbalance) limit and precaution states,

"'After a power change, wait ten minutes before taking data from the Minimum lncore System Backup recorders."

This limit ensures accurate recorder readings by allowing for decay effects of A. Palladium.

B. Rubidium.

C. Rhodium.

D. Cesium.

EP 1203-7, Hand Calculations for Quadrant Power Tilt and Core Power Imbalance,

~

page 4, Rev 40.

OPM Section (2-10, lncore InstrumentationSystem, page 4, Rev. 7.

None.

IV.E.16.02

@ New 2 TMlBank TMI Question #

L 0 Modified TMI Bank Parent Question #

4 Memory or Fundamental Knowledge 0 Comprehension or Analysis cl 55.43 c? 55.45 A Incorrect answer. Distracter is plausible, since Palladium is the decay product from Rhodium beta decay.

B Incorrect answer. Distracter will discriminate those who do not know the reaction involved in operation of the incore detectors.

C Correct answer.

D Incorrect answer. Distracter will discriminate those who do not know the reaction involved in operation of the incore detectors.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

TMI - Unit 1 I

Abnormal Procedure I 1203-7 Kevlslon NO.

Hand Calculations for Quadrant Power Tilt and Core Power imbalance 40 4.10 Power Operations Guidelines, Framatome Cogema Fuels 64-1234741-02 (Source Document) 4.11 EP-O07T, Numerical Analysis Computer Program Control 5.0 LIMITS AND PRECAUTIONS 5.1 A waiting period of at least twelve (12) minutes (2 complete NAS updates) should be observed prior to recording data from the incore detectors following any control rod group motion or power change to allow the detector output signals to reach steady-state. icient if taking data from the Minimum lncore System Backup recorders (IM1-NR and IM2-NR).

5.2 The acceptable difference between incore and out-of-core imbalance, as determined by 1302-1.1, Power Range Calibration, shall be adhered to.

5.3 Control room logs should be checked to determine that the 36 incore detector outputs on the two Minimum lncore System Backup recorders (IMI-NR and IM2-NR) and the.Out-Of-Core detectors were satisfactorily calibrated during the last periodic calibration period.

5.4 Quadrant tilt shall be monitored on a minimum frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the QPT alarm is inoperable and once every 7 days when the alarm is operable during power operation above 15 percent of rated power (T.S.3.5.2.4.g).

u 5.5 Power Imbalance shall be monitored on a minimum frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the axial power imbalance alarm is operable and every hour when the imbalance alarm is inoperable during power operation above 40 percent of rated power (T.S.3.5.2.7.f).

5.6 The computer - calculated heat balance, if valid, may be used whenever core power is required unless otherwise specified (Reference 4.5).

6.0 CALCULATIONS NOTE Enclosure 1 contains the definitions of Full lncore System, Out-of-Core Detector System, Minimum lncore System, Quadrant Power Tilt and Power Imbalance.

6.1 Full lncore System Tile and Imbalance Calculations 6.1 .I DETERMINE if the Nuclear Applications Software (NAS) results calculated by the Plant Process Computer (PPC) values are valid.

6.1.2 E the NAS calculated values are valid, THEN USE the full incore computer values to determine tilt and imbalance.

I NOTE I Figure 1 contains a flowchart outlining the guidance set forth in Steo 6.1.3.

4

REVISION 6

.W 5.0 SYSTEM DESCRIPTION 5.1 Modes of Operation - Theory The merial used in the incore flux detectors is the element RHODIUM ( R h d 3 ) . This material was

- chosen because of its high neutron absorption cross-section, its suitable mechanical and chemical properties and rhodium has only one mode of decay which is b*. This occurs with 2 separate half lifes, 43 seconds,and 4.4 minutes. The reaction that takes place is:

The decay of Rh-104to Pd-104by b decay with a half life of 4.4 minutes makes it necessary to use this system to monitor steadv-state neutron flux only.

The bdecay of Rh-104leaves a positive potential on the detector. If we have the detector connected to ground the resulting current flow from ground to the detector can be measured. The size of the current (signal) depends on the amount of Rhodium exposed to the neutron flux and the magnitude of the neutron flux in the core. The current flow is in the range of 600 nanoamps at 100% power.

u Since the flux itself produces the current flow these detectors require no external power source, Because ofthis they are commonly referred to as Self-Powered-Neutron Detectors or SPNDs.

Because of the high gamma flux in the core an additional current is produced in the wires coming from the detectors (Lead wires). This current is referred to as the background current. To obtain a correct reading we must compensate for this background current. To do this we use a background detector, this consists of the same material as the lead wires but no rhodium is present. Since the lead wires and the background wire are of the same material a current is produced in the background wire approximately equal to the background current produced on the SPND lead wires. This allows us to electrically compensate for the high gamma flux affects on the SPND lead wires.

Also included in the incore instrumentation string is a CHROMEL-ALUMEL incore thermocouple.

Because the two materials are different, when they are heated the electron motion of each material will be different. Ifthese two materials are connected at what is called the bot junction a difference in electrical potential wili exist. This difference of potential (voltage) can then be measured and corresponded to temperalure.

During the 12R Refbeling Outage, 47 of the 52 incore detector assemblies were replaced with Framatoms long emitter SPNDs.The long emitter refers to the physical length of the detector. The long emitter (15.75) SPNDs produce a much larger signal than the previousIy installed short emitter (4.75) SPNDs. The long emitter SPNDs produce a more accurate signal and have a longer operational life. Reer to Figure 2 for core pSacement of both type SPNDs.

c 4

SYS/EP# 015 KA# K6.04 Page # 3.7-7 Tier # -2 d RO/SRO importance Rating 3.1 -

3.2 Group # - 1 Knowledge of the effect of a loss or malfunction of the following will have on the Nuclear Instrumentation System: Bistables and logic circuits.

Plant conditions:

- Reactor power is 75%.

- 3 RCPs are operating.

- NI-5 power range instrument is selected for ICs input

- ICs stations in manual: SG/RX Master, Rx Master, Diamond CRD Control Panel, FW Loop Masters, Delta TC.

Sequence of events:

- NI-5 output signal fails slowly to ZERO % power.

- SASS actuation does NOT occur.

- Plant remains stable under manual control.

From the list below, identify the ONE statement that describes the effect of these conditions on plant control/logic circuits.

A. ICs Neutron Cross Limit circuit actuates.

B. CRD high startup rate Out Inhibit circuit is enabled.

C. RCP NI starting interlock designed to prevent a cold water accident is NOT functional.

D. RPS degree of redundancy for NI overpower trip defaults to ZERO.

b Drawing D553732, ICs Reactor Demand Subsystem, Rev M.

OPM Section F-03, Integrated Control System, page 72, Rev. 11.

FSAR Update-I5,page 14.1-13a, dated 4/00.

@ New TMlBank TMI Question #

@ Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge Comprehension or Analysis 0 55.43 55.45 .7 A Incorrect answer. Because both the Reactor Master and Diamond Control Panel are in Manual, neutron error is locked at 0. Therefore, actuation of Neutron Cross Limits (set at 5% neutron error) is not possible.

B Incorrect answer. Takes two power range channels, (NI-5 or NI-6) plus (NI-7 or NI-8), to enable thius circuit.

C Correct answer. 30% power starting interlock is not functional since it uses selected Power Range NI-5 or NI-6, and only uses a single instrument.

D Incorrect answer. Degree of redundancy drops to one, rather than the stated zero.

None.

b-TMI SRO Exam - May 2003 Friday, March 28,2003

b. Protective Basis Several different functions in the protective system are available ta terminate' any transient that might result f r b j the starting of an idle pump;.-For the case where the power rise is slow, as is the normal situation for this accident, the high pressure trip will terminate the accident if the high pressure set point is reached. However, if the power rise is rapid, the power/flow comparator is set so that when the reactor power is greater than the flow by a given amount (the set point), the reactor will trip. The overpower trip will limit the power to the maximum design overpower in all cases.

In addition to these primary protective devices, the power/pump comparator acts to fix the initial conditions from which this transient can be started if fewer than four reactor coolant pumps are operating.

rotective function is provided in the form of an interlock pump electrkal system wherein the operator is prevente pump if the reactor power is greater than a fixed power Protection exists at four distinct levels for this accident:

1) The pump control interlock acts to prevent the accident from occurring.

14.1-13a UPDATE-I5 4/00

TMI-l/FSAR

2) The power/pump comparator fixes the initial conditions in such a way that a severe accident (several pumps starting while at high power) cannot occur.

3) The Reactor Protection System operates so that the results of this accident stay within design limits.

4) Pumps cannot be started above 30 percent power:

C. Method of Analysis A detailed digital simulation of the plant (Reference 1) was used to evaluate the transient response to this accident. The model includes point kinetics, a multiregion fuel pin model, a pressurizer model, and a steam generator model.

To maximize the power response to the core temperature decrease, end-of-life core conditions were assumed in the analysis. At end-of-life the moderator temperature coefficient is most negative. Coupled with the reduction in core average temperature, the result was the largest power rise possible. To be consistent with end-of-life conditions, the end-of-life Doppler coefficient was used as well.

It was assumed that the plant was operating with two pumps (1 pump in each

'W loop) at 50 percent of rated power when the remaining two pumps were started.

Of the allowed operating conditions, the 111 pump status resulted in the greatest moderator temperature decrease for the cold water event. The mixed moderator average temperature as a function of time was independently calculated for the pump startup and imposed as input to the reactor system model. It was found that the maximum temperature decrease for this case was proportional to the core temperature rise and occurred at the end of one loop time. The assumed value of the moderator coefficient corresponds to end-of-life conditions. Using the most negative moderator coefficient results in the greatest positive reactivity feedback due to the core average temperature decrease which occurs, thereby maximizing the peak thermal power.

Conservative values for the moderator and Doppler feedback coefficients assumed were, - 3 . 0 ~0-41 (delta-k/k)/F and -12 x 1O'5 (delta-k/k)/F, respectively.

The tripped rod worth used corresponds to the minimum worth available with the maximum-worth rod stuck out.

It should be noted that while the analyses presented in this section assumed an MTC of -3.0 x delta-klWOF, it has been determined that adequate margin is available with an MTC of -4.0 x I0-4delta-k/W°F (Reference 65).

14.1-14 UPDATE-I5 4/00

SECTION F-03 REVISION 10 12.0 REACTOR CONTROL SUBSYSTEM 12.I Introduction The last subsystem which we need to discuss is the Reactor control subsystem. As was explained in the Basic Overview section the Reactor Control subsystem is used for controI of reactor power and to maintain a constant TaVg.

Subsystem Description The output of the 15% demand limiter IC 8.16 (refer to Inteegated Master analog print) is sent to RC 12.5 which is a function generator that calculates a neutron demand signa).for the demanded load. The output of RC 12.5 is then sent to summer amplifier (module) RC 12.9 where it is modified by feedwater cross limits.

Feedwater cross limits are developed by taking the feedwater error signal output of F W 8.9 {refer to Feedwater Subsystem analog print) and sending it to RC 153. RC 15.8 is a proportional controller which inverts the feedwater error signal. The output of RC 15.8 is sent to function generator RC 15.9 so that a useful signal is developed for the Reactor control subsystem. The output of the function getierator is sent to RC 14.9 which adds a 5% positive signal and then to RC 13.9 which is a limiter which will only pass a negative signal so that feedwater will only modify neutron demand when actual feedwater ffow is 5% less than demanded flow. The output of the limiter is then sent to the summer amplifier RC 12.9 to modify neutron demand.

The output of summer amplifier RC 12.9 is then sent to summer amplifier RC 12.10 where it is modified by TaVgThe selected Tmgsignal is sent to RC 7.10 where it is compared to an adjustable setpoint controller (normally set at 579OF) RC 7.12. The output of RC 7.10 is sent to RC 9.9 for use by the feedwater subsystem when the reactor cannot accept control of TaVg.The output of RC 7. IO is aIso sent to transfer switch RC 8.12 and to summer amplifier RC 12.10 for modification of the neutron demand signal. Transfer switch RC 8.12 will select the output of RC 7.10 when both the Reactor Master control station RC 15.13 and the Diamond Control panel are in automatic. If one or both of these control stations are in manual transfer switch selects RC 9.1 1 as its input (neutron error). The output of RC 8.12 is sent through integral RC 9.12 then through a limiter RC 10.12 and then to summer RC 12.10 and multiplier RC 12.12 to modify neutron demand for Tavg.

Integral correction to neutron demand is prevented when:

e Unit load demand changing at greater than 2% per minute as sensed by UL 1 1.13 Neutron demand limit circuit activated per RC 13.16 (MD) 6 Both OTSG's on low level limits and Tavgless than setpoint of RC 10.8 If the Reactor Master control station RC 15.13 is in manual than any difference between input and output is sent to RC 9.12 to make input of RC 15.13 match the output.

I 71

SECTION F-03 REVISIOhT10 If the reactor is unable to accept Tavgcontrol because the Reactor Master or Diamond contro! station are in manual then transfer switch RC 9.5 will select summer amplifier 9.6 output (RC 7.10 output signal that has been inverted and gone through integral and proportional controilers) onIy when the following conditions are met:

At least one OTSG off low level limits Neither OTSG on a BTU limit a At least one loop feedwater demand station in automatic If the above conditions are not met then transfer switch RC 9.5 selects a 0% setpoint RC 10.5.

The output of RC 12-12is sent through limiter RC 13.13 which will not pass neutron demand signals of less than 10% and greater than 103%. If either of these setpoints is reached an output is developed by RC 13.15 which causes RC 13.16 to change states and block the Tavsintegral RC 9.12.

The output of RC 13-13 is sent to HandAuto station RC 15.13 (Reactor Master control station) and if in auto the signal is sent to RC 16.13 where it is compared to actual neutron power (RC 16.14 output).

Any difference is sent to RC 19.12 (ND) and 19.14 (#AI)to cause rod motion to increase or decrease reactor power. The output of RC 18.13 is aIso sent to the neutron cross limit circuitry and RC 17.14 (A/D) which wilI prohibit transferring the Diamond control panel from manual to automatic when neutron error is greater than 1%.

The output of the functio 15.12 (AID) for use in the rod runback circuit and t which is a signal used for starting t Coolant pump.

The last circuit which needs to be developed is the "mini track circuit". This circuit is energized when both the Reactor Master and Diamond control stations are in manual and causes the output of RC 15.13 to be equal to the signal generator by neutron power RC 16.14 output.

13.0 INTEGRATED CONTROL SYSTEM 13.1 Power' Supplies During the 7R outage there were major changes in the area of reliable power to ICs, the ability to service a loss of ICs hand or auto power and limit the possibility of a loss of both hand and auto power. To reduce the possibility of a loss of both hand and auto power, they powered ICs hand power from ATB. ATA still powers ICs Auto Power. ICs Fan Power was left on ATA with a no load knife switch to TRB. Three A3T's are provided for the equipment that need power and are powered from ICs auto power and on loss of ICs auto power wilI transfer to hand power.

w FW Flow Recorder and Pzr HTR's Groups 1,2,3 0 FW Valve AP indication, and MS-V-3A-F control a FW Pump Control 72

orm ES-401-6 SYSIEP# 022 KA# K2.01 Tier #

W RO/SRO Importance Rating 3.o* 3.1

- Group #

Knowledge of bus power supplies to the following: Containment Cooling Fans.

Plant conditions:

- LOCA cooldown in progress.

- ES Actuations in effect:

- Automatic: 1600 A/B, 500 NB, 4 PSlG A/B

- Manual: 1600 NB, 4 PSlG A/B

- RCS pressure = 100 psig, slowly decreasing.

- RB pressure is 32 psig.

Sequence of events:

- Bus fault occurs on 1E 4KV Switchgear.

- Unit Supervisor directs CRO to re-energize I C ES Valve MCC to enable restart Of AH-E-I C.

Identify the ONE statement below that describes MINIMUM required actions to re-energize I C ES Valve MCC from its alternate power supply.

A. BypaWDefeat all AUTOMATIC ES signals, and then manually select the alternate power supply.

B. Bypass/Defeat/Resetall AUTOMATIC and MANUAL ES signals, and then select the alternate power supply.

C. BypasslDefeatlResetall AUTOMATIC and MANUAL ES signals, depress I C ES Valve MCC Power Reset pushbutton, to enable automatic ABT transfer to the alternate power supply.

\-

D. BypasslDefeatlResetall AUTOMATIC and MANUAL ES signals, depress 1C ES Valve MCC Power Reset pushbutton, and then MANUALLY select the alternate power supply.

OPM Section A-01, BOP and Class IE Electrical Distribution (Including Substation),

age 27, Rev. 17 (19).

&j! New c]TMlBank TMI Question #

0 Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge

@ Comprehension or Analysis 0 55.43 0 55.45 A Incorrect answer. Stated actions do not clear manual actuations. ABT Reset button is also required to be operated.

B Incorrect answer. ABT Reset button is also required to be operated.

C Correct answer.

D Incorrect answer. Once all ES signals are cleared, when the ABT Reset button is depressed the breaker will AUTOMATICALLY switch to the alternate supply -- even without manually operating the bus selector switch.

\--

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION A-0 1 REVISION 17

d. During an ES Condition all automatic transfers are locked out, that is, if an undervoltage condition existed on the 1C ESV MCC feeder Bus, during ESAS actuation the transfer switch woufd not transfer. The ES signal is locked in and therefore the ABT cannot be transferred, either automatically or manually by the ABT controI. BT reset button on panel PCR after the ES systems hay secured, the ABT is returned to the automatic mode. A transfer can only take after the ES signal has been reset.

6.0 OPERATION OF 6900V. 4160V AND 480V SWITCHGEAR UNDER ABNORMAL COPIDITIONS 6.1 Loss of one auxiliary transformer will cause auto transfer to the remaining transformer by causing the alternate feeder breaker to close within 100 milliseconds.

6-2 Upon a loss of both auxiliary transformers all feeder breakers which were closed to power the 6900V and 4160V busses will trip open. Due to the undervoltage present from the loss of the transformers undervoltage relays will cause the switchgear load breakers to trip. This will allow sequential re-energization of equipment when power is restored and, more importantly, prevents the undesired automatic starting of major pumps in the plant when power is restored.

6.3 Upon a loss of power to the 1A Circ Water Motor Control Center the circ water diesel fire pump (FS-P-1) will start.

6.4 It is possible to power C, G,J or L 480V bus from bus IN in an emergency. The tie bus current is limited to 600 amps as read by a 70 amp increase on feeder N1-02 ammeter.

To place C, G, J or L 480V bus on bus 1N perform the following steps:

6.4.1 Total load to be added must be less than 600 amps.

6.4.2 IN maintenance feed breaker 1N-12is closed before the respective tie breaker. is closed. 1N bus main breaker 1N-02 must be closed before IN-12 is closed.

6.4.3 T 2 - Q T1-02, T142 and T1-L2 can then be closed if the respective main breaker 1C-02, IG-02, 15-02 or 1L-02is open. (Dead-bus transfer) 7.0 BLOCK LOADING Block loading is the process by which selected loads required for core c o o h g and reactor building ventilation are loaded on the emergency diesels. If all loads were to be placed on the diesel at one time, the starting current from those loads would overload the diesel and possibly damage it. Therefore, selected loads are placed on the emergency bus at five second intervals to limit the starting current seen by the diesel generator.

M e r a start and bad signal is received by the diesel it takes ten seconds for the diesel to come to speed and voltage. When that occurs the diesel output breaker will close. Block 1 loads wili be energized.

Five seconds later Block 2 will Ioad on the bus with blocks 3 and 4 loading at five second intervals. EF-P-2A andlor B will load five seconds after that. This means that a total of thirty seconds will elapse from the receipt of a start and load signal until all required loads are operating.

27

Form ES-401-6 Q # 061 Page ## 3.5-10 Tier # -2 u ROISRO Importance Rating 4.2 4.2 Group # -1 Knowledge of the physical connections and/or cause-effect relationships between CSS and the following systems: ECCS Sequence of events:

Time RCS Pressure Event 0950 Reactor trip from full power due to LOCA.

1000 1200 psig HPI systems injecting BWST water into the RCS.

1030 800 psig RB Pressure peaked at 6 psig.

1130 700 psig LPllHPl Pumps in "Piggyback" Mode.

1200 550 psig Core Flood Tanks dumping.

1230 150 psig LPI systems injecting BWST water into RCS.

- RB Spray Pumps are NOT operating.

- RB Sump recirculation is NOT initiated.

Based on this event, identify the EARLIEST TIME below when sodium hydroxide (NaOH) will actually be injected into the reactor core.

A. Time = 1000.

B. Time = 1130.

C. Time = 1200.

D. Time = 1230.

W 302-640, Decay Heat Removal Flow Diagram, Rev. 78.

&!I New TMIBank TMI Question #

Ll Modified TMI Bank Parent Question #

El Memory or Fundamental Knowledge FZ Comprehension or Analysis E71 55.41 .2 thru .9 0 55.43 E71 55.45 .7/.8 A Incorrect answer. NaOH tank ties into LPI suction header, but not HPI suction.

B Correct answer. NaOH tank is being pumped to MUP suction by the DH Pumps, even though there is no direct LPI flow into the RCS.

C Incorrect answer. NaOH is already being added in a previous answer. Core Flood Tank water does not contain NaOH.

D Incorrect answer. NaOH is already being added in a previous answer. This would provide additional NaOH to the ECCS.

None.

TMI SRO Exam May 2003 Friduy, March 28,2003

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Form ES-401-6 KA# A3.01 Page # 3.5-12 Tier #

b ROlSRO Importance Rating 4.3 4.5 Group # -1 Ability to monitor automatic operation of the Containment Spray System (CSS) including:

Pump starts and correct MOV positioning w o Plant conditions:

- Reactor trip from full power due to LOCA.

- RB pressure reached 5 psig.

- Manual ESAS actuation signals were NOT initiated.

- Automatic ES Actuation status:

Train A Train B 160M Bypassed Bypassed 4## Defeated Actuated (NOT defeated)

Based on these initial conditions, identify the ONE statement below that describes automatic component response if RB pressure rises rapidly (spikes) to 40 psig.

A. Only BS-P-1B starts.

6. Both BS-P-1 A and BS-P-1B start.

C. Only BS-P-1A starts; Train 'A' RB Spray (BS) valves open.

D. Both BS-P-1A and BS-P-16 start; Both Train 'A' and Train 'B' RB Spray (BS) valves open.

OPM Section B-09, Reactor Building Spray, page 5, Rev. 6.

-u d New !. . TMI Bank TMI Question ##

C Modified TMI Bank Parent Question ##

CI Memory or Fundamental Knowledge Comprehension or Analysis 0 55.43 E 55.45 .5 A Correct answer.

B Incorrect answer. BS-P-1A does not start, since Train 'A' actuations are bypassedldefeated (Block 4 start permit does not exist for BS-P-IA).

C Incorrect answer. No new signals actuate to open Train A valves, and Train A Pump can not start without a Block 4 start permit.

D Incorrect answer. No new signals actuate to open Train A valves, and Train A Pump can not start without a Block 4 start permit. Train B will actuate.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION B-09 REVISION 6 5.1.2 Emergency There are two 50% capacity Reactor Building Spray Pumps (BS-P-IA and BS-P-1B) which take suction, through independent lines, fiom the decay heat removal pump suction headers.

The Decay Heat Removal Pump suction headers are connected to the BWST and to the Reactor Building Sump.

The discharge lines fiom each pump pass through separate penetrations into the Reactor Building. Each discharge line is connected to a separate spray header in the Reactor Building. A normally closed, motor-operated stop valve is installed in each spray supply line before the Reactor Building penetration. The spray headers are arranged and provided with sufficient spray nozzles to ensure complete coverage of the Reactor Building. Flow indication and alarms are displayed in the Control Room. Check valves (BS-V-30A and BS-V-3OB)are located in each spray line inside of the Reactor Building. These check valves insure the Reactor Building is isolated by preventing backflow through the building spray lines into the Auxiliary Building. Drain connections are installed on each spray header near the check valves to drain any leakage through the valves. This eliminates leakage through the nozzles and prevents moisture from building up in the spray headers.

The Reactor Building Spray System operates only during post-accident conditions and system testing. Post-accident building cooling and adsorption of iodine are provided initially by spraying water from the BWST and the NaOH Tank.

The Reactor Building Spray System is normally maintained in a standby condition during power operations. The operation of the system is accomplished in two segments. The system Walves are opened b ure at 4 psi. The Building Spray Pumps also receive a *START e. The Reactor Building Spray Pumps are started by HI-HIc psi;; When the 4 psi signal is received, the power operated valves, for the Sodium Hydroxide Tank outlet lines and the Reactor Building spray headers, open. When Reactor Building pressure increases to 30 psi, the two Reactor Building Spray Pumps start and take a suction from the BWST and the NaOH Tank to limit building pressure. When the Lo water level (9' 6")in the BWST is reached, the electric motor-operated stop valve (DH-V-6A and DH-V-6B)in each of the suction lines from the Reactor Building Emergency Sump are manually opened to permit recirculation of the Reactor Building Sump water. The sump would continuously collect water from the Reactor Coolant System leak. The outlet valves from the BWST are then closed (DH-V-SA and DH-V-SB), when BWST Lo-Lo level alarm is reached (6' 4"),

5.2 Components 5.2.1 Reactor Building Spray Pumps The Reactor Building Spray Pumps are single-stage, horizontal-shaft, single-suction, centrifugal pumps. They are rated at 1500 gpm with a total dynamic head of 450 feet. The Reactor Building Spray Pumps are located in their own separate vautts in the Auxiliary Building. The pump motors are powered as follows:

BS-P-1A 1D 4160 V ES BUS BS-P-1B - 1E 4160 V ES US 5

Page # 3.8-6 Tier #

u ROISRO Importance Rating 3.4 -3.7 Group # -2 Knowledge of the physical connections and/or cause-effect relationships between Containment Purge System (CPS) and the following systems: Gaseous radiation release monitors From the list below, identify the ONE statement that describes the response of the Reactor Building Purge System to a High alarm condition on RM-A-9G, Reactor Building Purge Duct monitor.

A. ONLY purge supply fans AH-E-GNB trip.

B. ONLY purge supply and exhaust isolation valves AH-V-IA/B/C/D close.

C. Only purge exhaust fans AH-E-7NB trip AND purge exhaust isolation valves AH-V-IAIB close.

D. Purge exhaust fans AH-E-7AIB trip, AND purge supply and exhaust isolation valves AH-V-INBICID close.

AP C-1-1, RM-A-9 Rx Bldg. Purge, page 16, Rev. 27.

0 New 0 TMI Bank TMI Question # SR4F02-06-QO3 Ci Modified TMI Bank Parent Question #

PI Memory or Fundamental Knowledge l-- , 0 Comprehension or Analysis E

d 55.41 .2 to .9 55.43 2 55.45 .7/.8 A Incorrect answer. Plausible misconception since tripping only supply fans and allowing exhaust fans to continue to operate could reduce RB pressure, release rate.

B Correct answer.

C Incorrect answer. Plausible misconception since this action would terminate the release.

D Incorrect answer. Plausible misconception since this action would terminate the release.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 u Alarm Response Procedure MAP C Title Revision No.

Main Annunciator Panel C (See Cover Page) c-I-1 Revision 27 ALARM:

RADIATION LEVEL HI SET POINTS:

Refer to appropriate sheet for individual action.

CAUSES:

RM-G-1 thru RM-G-7, RM-G-9 thru RM-G-21 or RM-G-24 thru RM-G-27 RM-L-1 thru RM-L-7 and RM-L-9 RM-A-1, RM-A-2, RM-A-4, RM-A-5, RM-A-6, RM-A-7, RM-A-8, RM-A-9, RM-A-15 AUTOMATIC ACTION:

Refer to appropriate sheet.

i, OBSERVATION (CONTROL ROOM):

Indication of the alarming monitor.

MANUAL ACTION REQUIRED:

1. Acknowledge the Alarm.

2. Refer to appropriate sheet.

3. Refer to 1202-12, Excessive Radiation Levels Page Iof 50

t TMI - Unit 1

\.J Alarm Response Procedure MAP C Tlile Revision No.

Main Annunciator Panel C I (See Cover Page) c-1-1 Revision 27 ALARM:

RM-A8 RX. BLDG. PURGE SET POINTS:

Refer to Operating Procedure 1101-2.1, RMS Setpoints CAUSES:

1. Filter Units AH-F-1 contamination

2. Improper Purge Flow

3. RCSLeak during purge AUTOMATIC ACTION:

1. R.B. Purge Valves AH-V-IA, B, C, and D will close.

2. R.B. Sump Isol. Valves WDL-V-534 and 535 will close on a Gaseous "Hi Alarm."

- 3. Remote sampler starts (MAP-5)

OBSERVATION (CONTROL ROOM):

1. RM-A-9 on PRF "Alert" Alarm (yellow light)

2. RM-A-9 on PRF "Hi Alarm" (red light)

3. RM-A-9 Indication > one or more of the above set points MANUAL ACTION REQUIRED:

1. Check RM-A-9 indication on PRF to insure discharge is within limits specified in the gas release permit.

If the RM-A-9 indication is above the limits of the gas release permit, notify GRCS.

2. If there is an obvious and unexpected trend toward the hi alarm setpoint, stop the purge and re-evaluate the release calculations.

3. If no release is in progress, verify closed AH-V-IA, B, C, and D.

Hi Alarm:

1. Verify AH-V-IA, B, C and D closed, WDL-V-534 and 535 closed.

2. Have chemistry take RM-A2 sample. notify GRCS of the problem.

3. Refer to Emergency Procedure 1202-12 Excessive Rad Levels.

Page 16 of 50

SYSIEP# 033 KA# A3.01 Page # 3.8-10 Tier # -2 u RO/SRO Importance Rating 2.5* 2.7*

- Group # -

2 Ability to monitor automatic operation of the Spent Fuel Pool Cooling System (SFPCS) including: temperature control valves I

B '

Initial conditions:

- Reactor power is loo%, with ICs in full automatic.

Based on these conditions, identify the ONE selection below that describes the automatic interface between Spent Fuel Cooling Pump, SF-P-IA, and SF Cooler inlet valve, NS-V-16A.

NS-V-I6A, cooling isolation to Spent Fuel Cooler A, automatically (1) when SF-P-1A extension control is switched from the (2)

A. (1) opens (2) Pull to Lock position to the Stop position.

B. (1) opens (2) Normal After Stop position to the Start position.

C. (1)closes (2) Stop position to the Pull to Lock position.

D. (1) closes (2) Normal After Start position to the Stop position.

SS-208-574, Spent Fuel Cooling Pump A SF-P-IA, Rev. 8.

SS-208-513, Spent Fuel Cooler Inlet Valve NS-V-IGNB, Rev. 5.

'LJ None.

@ New a TMIBank TMI Question #

E l Modified TMI Bank Parent Question #

31 Memory or Fundamental Knowledge 0 Comprehension or Analysis lid 55.41 .7 0 55.43 3 55.45 .5 A Incorrect answer. Plausible misconception that Pull-to-Lock position is interlocked with this action since it does intyerface with SF valve position indication on Panel PL.

B Correct answer.

C Incorrect answer. Plausible misconception, since there are other systems that have valve interlocks like this.

D Incorrect answer. Plausible misconceptionthat Pull-to-Lock position is interlocked with this action since it does intyerface with SF valve position indication on Panel PL.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

REFERENCE DWGS:

INDEX: SS-208-401 THRU 407 NOTES:

1. %E SET WlMS OF LS9 THRU 12 & LS13 I ' Mt LEGEND SS-208401 THRU LSl6 ARE ADJUSTABLE AT ANY PERCENTAGE THREE M I LOGIC MAG. 5-203-HI3 OF VALVE POSITION.

2. V-VERTICAL PNL SEC P L

3. ROTOR # 1-LIMIT SWS. 1 TO 4 VALVE SHOWN IN FULL CLOSED minau KOTOR # 2 -LIMIT SWS. 5 TO 6 ROTOR rY3-LIMIT SWS. 9 TO 12 I LlMlTORQUE VALVE OPERATOR ROTOR #4-LlMlT SWS. 13 TO 16

4. PREFIX WIRE MARKS WITH VALVE NO. i e V16A-X3.

1A H V165X3, ETC.

25w SPACE HTR.

I METROPOL I TAN EO I SON COt&ANY INAIM LGI I ULeRTAssocIATW,INc.

THREE MILE i w o NUCLEAR STATION ' CnK'D.

laa.cF.

JK c5c i -ANDQowwKT-0 MOTOR OWRATED VALVE TITLE 1 V 25w 4oW SPACE HT R.

MOTOR HtR.

a-I ,

15 1 LS16

+

REFERENCE DWGS. -

INDEX: SS-208-401 THRU W7 NOTES: M I. V PL-VERTICAL PANEL IN CONTROL ROOM.

LEGEND: SS -208-001 LOCK: DIAG.: 5-203-102 2. CLOSE IN FILL POSITION.

THREE M C . S. DEVELOPMENTS: SS-208-414 3. CLOSE IN MPTT POSITION. EL 43/559 DEVEL. : . Ss-209-009

4. OPEN OH HIGH LEVEL.

5. OPEN Ow LOU LfVYEL.

6 . 0 -i A ENGIMEEAED SAFEGUARD MCC UNlT 13A, 1.-

SEE O K . B-ZiO-SZIC.

- L A F 5 b T\M M E f R I

NE'jtHLER TYPE 4H33Q x3 T

4 1 cs4 RES RES 4cT v I 7 t 0 43lSS9 0

c1 R G 13 271866 SH. 311 1;

3 42 t 2T -

CSFl cs4 csF 42 I

1- LS212A 1 LS212B 7T

CSFL

<- SS-2Y)8-513

.- SPARES WL) WL) --

42 ts I FUEL TRANSFER FUEL TRANSFER CANAL HIGH CANAL LOW SS-209-91U SS-209-9 I U ASF 1 U

METROPOLITAN EDISON COMPANY MADE LGI GILBERT ASSOCIATES, IN.

CHK'D. gpg THREE MILE ISLAND NUCLEAR STATION ELECTRl CAL ELEMENTARY 0I.AGRAM UNIT 1 1

m. CF. G C ,

ENGINEERS AND CONSULTANTS IKIWWB, 480V. CONTROL CENTERS FCN-C085931 I U

ATS i

I I t I REVI SION I I TITLE

13E IA-ES SPENT FUEL COOLING PUMP A

F-PIA E CSlSFPlA 51 L I I

i-SF - 2 t 1 1 1 1 AI SFI ENI--.-&L J

--B 42 COOL.

Y SS-209-087 2

SS-209-088

Form ES-4016 Page # 3.8-13 Tier # 2

-u RO/SRO Importance Rating 3.4 -

3.9 Group # -

2 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling Equipment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped cask.

Plant conditions:

- Spent Fuel shipment activities are being conducted in compliance with applicable procedure and Tech Spec requirements.

- A cask containing spent fuel is being transferred from the Cask Loading Pit to the transport vehicle in the truck bay.

- During transfer, the fuel handling crane hoist brake fails, and the cask is dropped.

Based on these conditions, identify the ONE selection below that predicts (1) the impact of the load drop, and (2) the mitigating action required to be taken.

A. ( I ) New Fuel Elevator may be damaged.

(2) Implement 1503-1, Receipt of New Fuel and Control Components.

B. (1) The transport vehicle may be damaged.

(2) Follow Rad Con and Safety Engineering guidance.

C. (I) Fuel Transfer Carriage may be damaged.

(2) Implement 1503-2, Damaged Fuel and Control Components

- D. (1) One or more fuel assemblies in the Spent Fuel Pool may be damaged.

(2) Implement 1503-2, Damaged Fuel and Control Components.

Technical Specification 3.1 1, pages 3-55 (log), 3-56 (157), 3-56a (157), 3-56b (109).

Technical Specification 3.1 1, Pages 3-55 to 3-56b, Amendmends 109 and 157.

2l New 0 TMI Bank TMI Question #

Ll Modified TMI Bank Parent Question #

Cl Memory or Fundamental Knowledge Ld Comprehension or Analysis 55.41 .5 23 55.43 .6 @ 55.45 .3/.13 A Incorrect answer. New Fuel elevator is outside travel limits imposed by Tech Spec Figure 3.11-1.

B Correct answer.

C Incorrect answer. Technical Specification Figure 3.1 1-1 Transfer Path does not allow travel over the New Fuel Elevator.

D Incorrect answer. Technical Specification Figure 3.1 1-1 Transfer Path does not allow travel over the Spent Fuel Pool.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

i h

3.11 Handt fnq o f Irradiated Fuel Appl icabi 1 i t y Applies to the operation of the fuel handling building crane when withfn the

-confines o f Urlit 1 and there is any spent fuel in storage in the U n i t 1 f u e l handling building..

Object4 ve To define the lift conditions and allowable areas o f travel when loads to be lifted and transported with the fuel handling buildfng crane are i n excess o f 15 tons or between 1.5 tons and 25 tons or consist o f frradiated fuel elements.

I Specif iceition_

3.11.1 Spent fuel elements havfng less than 120 days far decay o f their frradiated fuel shall not be loaded into a spent fuel transfer cask fn the shipping cask area.

3.13.2 The key operated travel interlock system for automatically 1 irniting the travel area of the fuel handling building crane shall be imposed whenever_ loads in excess of 15 tons are tu be lifted and.transported with the exception o f fuel hand1 ing brfdge mafntenance.

3.11.3 The lowest surface of a l l loads f n excess of 15 tons shall be adMfniStratfVt?ly limited to an elevatfon one foot or less above the concrete surface a t the nomSnal 348 ft-0 in. elevation in the fuel handling building, 3.11.4 loads in excess o f hook capacity shall not be lifted, except for load testirig.

3.11.5 Following modifications or repairs to any o f the load bearing members, the crane shall be subjected to a t e s t l i f t o f 125 percent of its rated load.

3.11.6

Administrative controls shall require the use o f an approved procedure with an identified safe load path for loads i n excess o f 3,000 lbs. handled above the Spent fuel Pool Operating Floor (348' el evation 1.

3.11.7 During transfer o f the cask t o and from the cask loading pit, t h e cask will be restricted t o the transfer path shown i n Figure 3.11-1.

Administrative controls w f ? ? be used to ensure that all lateral movements of the cask are perfonned a t slow bridge and trolley speeds. During thfs transfer the cask llftfng yoke shall be oriented in the East-West dfrection, 3-55

Bases T h i s s p e c i f i c a t i o n will l i m i t acttvity releases t o unrestricted areas

-1) r e s u l t i n g from damage to spent fuel stored tn the spent fuel storage pools in the postulated event o f t h e dropping o f a heavy load from t h e fuel handling buflding crane. A Fuel Handling accident analysis was performed assuming that the cask and 9 t s entire contents o f t e n I fuel assemblies are sufflciently damaged as a result o f dropping the cask, to .allow the escape o f all noble gases and iodtne tn the gap

[Reference 1). This release was assumed to be directly t o the atmosphere and to occur instantaneously. The s i t e boundary doses 1 resulting from t h i s accident are 5.25 R whole body and 1.02 R t o thyroid, and are withSn t h e limits spec'lfied in 10 CFR 100. .

Specification 3.11.1 requires that spent fuel, having less than 120 days decay post-irradiation, ncrt be loaded i n a spent f u e l transfer cask fn order to ensure that the doses resulting from a highly improbable spent fuel transfer cask drop would be within those cal cul ated above.

Specification 3.11.2 requires the key operated interlock system, which automatically limits t h e travel area o f the fuel handling crane while it is liftjng and transporting the spent fuel shipping cask, t o be imposed whenever loads In excess o f 15 tons are to be lifted and.

transported while there i s any spent f u e l in storage in the spent fuel storage pools in Unit 1. This automatically ensures that these heavy loads travel in areas where, in the unl ikely event of a load drop accident, there would be no possfbility o f t h i s event resulting in any damage to the spent fuel stored in the pools, any unacceptable structural-damage to the spent fuel pool structure, or damage t o redundant trafns of safety related components. The shipping cask area i s designed to wjthstand the drop o f the spen't fuel shipptng cask f r o m the 349 ft-0 in. elevation without'unacceptable damage to the spent fuel pool structure (Reference 2 ) .

S p e c i f S c a t i o n 3.11.3 ensures that the lowest surface o f any heavy load never gets higher than one f o o t above the concrete surface o f the 348 ft-0 in. elevatfon In the fuel handljng buildSng (nominal elevation 349 f t - O in.) thereby keeping any Smpact force from an unlikely load drop accident within acceptable limits.

Specification 3.11.4 ensures that the proper capacity crane hook is used for lifting and transporting loads thus reducfng the probabilfty of a load drop accident.

Following modiffcation or repairs, specification 3.11.5 confirms the load rating o f the crane.

References (1) UFSAR, Sectjon 14.2.2.1 - "Fuel Handling Accident" (2) UFSAR, Section 14.2.2.8 - "Fuel Cask Drop Accident" 3-56

4-25 S p e c i f i c a t i o n 3.11.6 imposes a d m i n i s t r a t i v e l i m i t s on handling loads weighing in excess o f 3000 lbs. t o minimize the potential far heavy 3 loads, i f dropped, to impact irradiated fuel i n the spent fuel p o o l ,

o r to impact redundant safe shutdown equipment. The safe load path shall follow, to the e x t e n t p r a c t i c a l , structural floor members, beams, e t c . , such t h a t f f the load I s dropped, the structure i s more likely to withstand the impact. Handling loads o f less than 3000 1bs. without these restrictions i s acceptable because the consequences of dropping loads in this weight range are comparable to those produced by the fuel hand?ing accident considered In the FSAR and found acceptabl e.

Specification 3.11.7 in combination w i t h 3.11.3 ensures the spent fuel cask i s handled i n a manner consistent with the load drop analysis (Reference 3 ) .

1 Reference

( 3 ) GPU Evaluation of Heavy Load Handling Operations at TMI-1 February 21, 1984, as transmitted to the NRC i n GPUN tetter No. 5211-84-2013.

. 47565-3

.I I 4 L

1

Page # 2-11 Tier #

'- RO/SRO importance Rating 3.1 -

4.0 Group # 2 Knowledge symptom based EOP mitigation strategies: Main and Reheat Steam System 17 q Plant conditions:

- Reactor trip from full power due to low RCS pressure.

- All 4 RCPs are operating.

- OTSG 1A pressure is 800 psig and lowering.

- OTSG 18 pressure is 950 psig and lowering.

- Both OTSG levels are 33 inches in Startup Range and lowering.

- RCS T-Ave is 549°F and lowering.

- RCS pressure is 1700 psig and lowering.

- RB pressure is 0.1 psig and stable.

- Pressurizer level is 73 inches and lowering.

Based on these conditions, identify the OP-TM-EOP-010 procedural guidance required to be implemented to mitigate the transient.

A. Guide 15, EFW Actuation Response.

B. Rule 3, Excessive Heat Transfer.

C. Rule 2, HPI/LPI Throttling.

D. Rule 1, Loss of SCM.

OS-24, Conduct of Operations During Emergency and Abnormal Events, section 3.5, page 3, Rev, 7.

' d None.

V.E.10.08 New C TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 . I O PI 55.43 .5 E 55.45 .I3 A Incorrect answer. Based on coditions given, there is no reason for automatic EFW actuation.

B Correct answer.

C Incorrect answer. Based on coditions given, HPI is not operating.

D Incorrect answer. Based on coditions given, loss of subcooled margin has not occurred.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 Operations Department

'L/ Administrative Procedure OS-24 Conduct of Operations During Abnormal and Emergency Events 7 1.o PURPOSE Establish the structure, responsibilities, and guidance to promote consistent and reliable operator performance when implementing usage level "EP" Abnormal Procedures (1203), Emergency Procedures (1202), Emergency Operating Procedures (OP-TM-EOP) and support procedure sections identified with usage level "EP".

2.0 APPLICABILITY/SCOPE Operations Department personnel performing or responding to direction provided in "EP" Usage level procedures.

3.0 DEFINITIONS 3.1 ADEQUATE CORE COOLING:

A demonstrated cooldown rate accomplished by removing coce decay heat either through primary to secondary heat transfer, RCS feed and bleed or a combination of both methods. A core cooldown rate of 4O"F/hr demonstrates core heat removal. At less than 250°F, adequate core cooling may not result in a cooldown rate > 40°F/hr.

3.2 CARRYOVER STEP:

A conditional step that uses the logic: IF AT ANY TIME, THEN. These steps are denoted by a rectangular box and remain open until the condition is satisfied, at which point the procedure user refers to the step for implementation. A Carryover step only applies to the EVENT PROCEDURE in which the step is contained and does not carryover to another EVENT PROCEDURE. As an aid to the procedure user, CARRYOVER STEP(s) may appear on the facing page following the appearance of the original step and remain in the individual procedure until no longer applicable.

3.3 ENSURE

Take necessary or appropriate action to guarantee component, or parameter, etc. is as specified.

Exception: ENSURE and VERIFY in Emergency (1202 series) and Abnormal (1203 series) procedures are used interchangeably until they are revised.

3.4 EVENT PROCEDURE:

Abnormal Transient Procedures, Emergency Procedures and Abnormal Procedures designated Usage Level "EP", used to mitigate ABNORMAL or EMERGENCY EVENT(s).

3.5 EXCESSIVE PRIMARY-TO-SECONDARY HEAT TRANSFER (XHT) is undesired heat removal by one or both OTSGs. XHT can be confirmed if b . RCS temperature have lowered below the desired range

'.AND OTSG pressure is low or lowering in one or OTSGs AND b Secondary Tsat5 T, for the affected OTSG(s).

3.6 FEEDWATER

A water source from either Main or Emergency Feedwater supplied to the OTSG(s).

3

Page # 3.4-22 u -2 ROlSRO Importance Rating a 2.9

- Group #

Ability to manually operate and/or monitor in the control room: Steam dump valves.

LZ Plant conditions:

- Reactor tripped from 100% power.

- ICs Turbine Bypass Valve/Atmospheric Dump valve control stations are in automatic.

- Condenser vacuum is low at 20 inches due to condenser air leak.

- OTSG pressures are stable at 1010 psig.

Assuming NO operator action, based on these conditions, identify the ONE statement below that describes the response of OTSG pressure control systems when the Main Condenser vacuum interlock resets to NORMAL.

A. ADVs will automatically control OTSG pressures between 1026-1052 psig; TBVs remain closed.

B. TBVs open to automatically control OTSG pressures at 1010 psig; ADVs close, but will open if OTSG pressure exceeds 1040 psig.

C. ADV control is transferred to the (control room) Backup Manual Loaders; TBVs remain closed.

D. TBV control is automatically transferred to ICs Manual; ADVs close, but will open if OTSG pressure exceeds 1040 psig.

OP 1105-4, Integrated Control System, pages 40, 56 and 57, Rev. 66.

Training Lesson Plan 11.02.01.055, book 2, page 32. Rev. 12.

u

@JNew 0 TMI Bank TMI Question #

C Modified TMI Bank Parent Question #

Z Memory or Fundamental Knowledge G Comprehension or Analysis 55.43 E 55.45 .5/.6/.7/.8 A Correct answer. TBVs will not operate until RESET button is depressed. ADVs will resort to normal operation (high pressure relief at 1026 - full open at 1052 psig).

B Incorrect answer. TBVs will not open until the RESET button is depressed on Console Center.

C Incorrect answer. ADVs will not transfer to the backup Loaders under these conditions.

nswer. TBVs are locked out until RESET is depressed.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Revision 12 05ii 012000 11.201.055 INSTRUCTOR NOTES W

Cap - 3.2 percent each cj Atmospheric Dump Valves 27.19 (1) For the turbine bypass valves to be functional, the main turbine condenser must be available.

When the condenser is lost. due either to loss of coolinq water (less than 2 CW-P-s are runnind or hiah condenser wessure jareater than 7 Ha absolutei, the turbine bypass valves to the condenser fail shut. (Latched closed)

The atmospheric exhaust valves, if in auto would be utilized for pressure control at setpoint +Bias or open at 1040 until the

- 27.18 condenser was again available..

(2) Normal High Pressure Relief - 1026 PSI .

Full Open at 1052 PSI u

Number TMI - Unit 1 u Operating Procedure 1105-4 Tide Revision No.

Integrated Control System 66 3.7 Restoration of Turbine Bypass and Atmospheric Dump Valves to ICs Control following a LOSSof Condenser Vacuum - LEVEL 2

3.7.1 Prerequisites

CAUTION Prior to resting vacuum, transfer MS-V-4A(B) control from the ICs MS-V-3D, E, F or 4A (MS-V3A, B, C or 4B) HANDlAUTO control stations to the MS-V-4A(B) BACKUP CTRL stations by matching demands between the ICs and B/U LOADER stations and depressing the B/U LOADER Dushbutton.

A. Condenser Vacuum Restored 3.7.2 Procedure

1. Place the ICs MS-V-3D, E,F or 4A (MS-V3A, B. C or 48) HANDlAUTO control stations in HAND and reduce demand to zero percent.

2. Press MS-V-3 A-F Power RESET pushbutton on CC to return to ICs control and verify White Lamp is lit on pushbutton.

3. Check Steam Press Error on the MS-V-3D, E, F or 4A (MS-V3A, B, C or 4B) HAND/AUTO control stations and place in AUTO, or control MS-V-3 A-F in HAND at desired OTSG pressure. N/A option not used.

AUTO HAND

4. Gradually close MS-V-4A(B) using the MS-V-4A(B) BACKUP CTRL stations while transferring steam load to MS-V-3A-F. If MS-V-A-F in HAND the operator will have to pickup steam load.

5. When MS-V-4A(B) are closed, transfer MS-V-4A(B) control to ICs by pressing the ICs pushbutton on CC.

6. Check, Steam Press Error on the ICs MS-V3D, E,F or 4A (MS-V-3A, B,C or 4B) HAND/AUTO control stations and place MS-V-3A-F in AUTO, if performed previously N/A this step.

40

Number 3 f.4 TMI - Unit 1 Operating Procedure 1105-4 Tirle Revision No.

Integrated Control System 66 ENCLOSURE 3 Page 1 of 2 0perato r Inforrnatio n Part 1: Definitions:

"Full Automatic" -AI1 ICs stations in AUTO except:

(1) ULD station may be in HAND or AUTO (2) One FW Pump may be in HAND Part 2 : Turbine Bypass and Atmosphere Dump Valve Control Options NOTE Auto control input is always SASS/Operator selected OTSG pressure SPGA(BLPT112).

A. Normal Operations

a. Turbine Bypass Valves (MS-V-3A-F) operated in AUTO control at Turbine Hdr Pressure setpoint plus appropriate bias (+IO, +75, +I25 psig) open at 1040 whichever is the greater open signal.

b. Turbine Bypass Valves (MS-V-3A-F) operated in HAND respond only to demand changes from the

'-, ICs MS-V-3D, E, F or 4A (MS-V-3A, B, C or 48)" HANDlAUTO control stations.

c. Atmospheric Dump Valves (MS-V-4NB) will open at 1026 - 1052 psig.

d. Operator may select Atmospheric Dump Valves control from "MS-V-4A(B) BACKUP CTRL" stations by depressing the "BIU LOADER" pushbutton and position valves as desired.

B1. Loss of Condenser Vacuum (< 2 CW pumps or < 23" Hg Vac)

a. Turbine Bypass Valves (MS-V-3A-F) are closed and latched. The ICs "MS-V-3D, E, F or 4A (MS-V-3A, B, C or 4B)" HANDlAUTO control stations have no effect on their position.

b. Atmospheric Dump Valves operated in auto control at Turbine Hdr Pressure setpoint plus appropriate bias (+IO, +75, +I25 psig) or open at 1040 whichever demand is greater.

C. Atmospheric Dump Valves (MS-V4A/B) operated in HAND respond only to demand changes from the ICs "MS-V-3D, E,F or 4A (MS-V-3A, B,C or 4B)" HAND/AUTO control stations.

d. Operator may select Atmospheric Dump Valves control from "MS-V-4A(B) BACKUP CTRL" stations by depressing the "B/U LOADER" pushbutton and position valves as desired.

B2. Loss of ICs Auto Power

a. Turbine Bypass Valves (MS-V-3A-F) revert to HAND and respond only to demand changes from the ICs "MS-V3D, E, F or 4A (MS-V-3A, B, C or 4B)" HAND/AUTO control stations.

b. Atmospheric Dump Valves (MS-V-4NB) automatically transfer to BU Loader" and may be positioned as the operator desires from the "MS-V-4A(B) BACKUP CTRL" stations.

55

4-L p -

I Number TMI - Unit 1 Operating Procedure 1105-4 Ti:le Revision No.

_-integrated Control System 66 B3. Loss of ICs Hand Power

a. Turbine Bypass Valves (MS-V-3A-F) operated in AUTO control at Turbine Hdr Pressure setpoint plus appropriate bias (+IO, +75, +I25 psig) open at 1040 whichever is the greater open signal.

b. I C s "MS-V-3D, E, F or 4A (MS-V-3A, B, C or 4B)" HANDlAUTO control stations fail to 50 percent demand in HAND. Turbine Bypass Valves (MS-V-SA-F) operated in HAND fail to mid-position.

Operator action: Transfer HAND/AUTO station to AUTO or close locally.

C. Atmospheric Dump Valves (MS-V-4NB) will open at 1026 - 1052 psig.

d. Operator may select Atmospheric Dump Valves control from "MS-V4A(B) BACKUP CTRL" stations by depressing the "BIU LOADER" pushbutton and position valves as desired.

B4. Total Loss of ICS/NNI Power

a. Turbine Bypass Valves (MS-V-SA-F) fail closed and latched.

b. Atmospheric Dump Valves (MS-V-4NB) automatically transfer to BU Loader" and may be positioned as operator desires from "MS-V-4A(B) BACKUP CTRL" stations.

56

SYSfEP# 055 KA# A3.03 Page # 3.4-35 Tier # -2 yu ROlSRO Importance Rating 2.5* 2.7*

- Group # -2 Ability to monitor automatic operation of the Condenser Air Removal System (CARS) including: Automatic diversion of CARS exhaust.

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- VA-P-1 C is the only operating Main Vacuum Pump.

- Loss of 'A' DC Distribution System power occurs.

Based on these conditions, identify the ONE statement below that describes related response to the loss of 'A' DC power.

A. RM-Ad Condenser Offgas Monitor indication rises to the high alarm level.

B. Immediate LOW Vacuum trip of the main turbine.

C. Immediate start of the standby Vacuum Pump, D. Vacuum exhaust flow indication lowers.

1202-9A, Loss of "A" DC Distibution System, page 3, Rev. 43.

New 3 TMIBank TMI Question ##

'u E l Modified TMI Bank Parent Question #

El Memory or Fundamental Knowledge LE Comprehension or Analysis

& 55.41 .7 0 55.43 M 55.45 .5 A Incorrect answer. Plausible misconception that reduced offgas flow will cause off-gas radiation monitor indication to rise.

B Incorrect answer. Plausible misconception.

C Incorrect answer. Plausible misconception that running vacuum pump trips, initiating automatic start of standby vacuum pump.

D Correct answer. Vacuum Pump Suction valve, VA-V-9C, closes on loss of DC power.

TMI SRO Exam - May 2003 Fridw, March 28,2003

Number TMI Unit 1 Emergency Procedure 12028A

, Title Revision No.

Loss of "A" DC Distribution System 43 Loss of Control Power to the following:

a. 1A and 1B 6900V Reactor Plant Swgr. Feeder Breakers

b. lD-4160V ES Swgr.

C. lA-4160V Turbine Plant Swgr.

d. lC-4160V Turbine Plant Swgr.

e. 1C-48OV Turbine Plant Swgr.

f. 1E-480V Reactor Bldg. Ht. and Vent Swgr

g. 1G-480V Reactor Plant Swgr.

h. 1H-480V Aux. and Fuel Bldg. Ht. and Vent Swgr.

i. 1J480V Turbine Plant Swgr.

j. 1N-480V Turbine, Reactor and Control Bldgs. Htg. Swgr.

k. 1P-48OV ES Swgr.

I. 1R480V Scrn. Hse. ES Swgr.

2.0 IMMEDIATE ACTION A. Automatic Action

1. Suction Valves to VA-P-1A and VA-P-1C close.

2. Emergency Feed Pump EF-P-1 Starts.

3. Emergency Diesel Generator EG-Y-1A starts.

4. Panel 1M automatically transfers to "B" DC Distribution System.

5. RC-RV-2 fails closed and is inoperable.

6. MU-V-6A fails open and MU-V-6B fails closed placing " A makeup and purification demineralizer in service.

7. MU-V-I 16 fails open and MU-V-11A fails closed placing MU-F-1 B in service.

8. CM-V-1 and CM-V-3 fail closed making RM-A-2 inoperable.

3

SYSIEW 061 KA# K5.02 Page # 3 . 4 4 6 Tier # -

2 u ROISRO Importance Rating 3.2 -

3.6 Group ## 1 Knowledge of the operational implications of the following concepts as they apply to the AuxiliarylEmergency Feedwater (AFW) System: Decay heat sources and magnitude Plant conditions:

- Reactor tripped from 100% power due to loss of offsite power (LOOP).

- Both OTSG pressures are 1010 psig and stable.

- Both OTSG levels are at proper setpoint and stable.

- EFW flowrates are stable.

- RCS cooldown rate is 0" per hour.

- RCS natural circulation is verified.

- Time is 30 minutes post trip.

Based on these equilibrium conditions, from the list below, complete the following phrase:

This trip from full power results in a , compared to a trip from 15% power.

A. lower core delta T B. higher EFW flowrate C. longer (RCS) loop transport time D. lower Atmospheric Dump Valve position (% open)

OPM Section N-07, Reactivity Control and Fuel Behavior, page 160, Rev. 5.

None.

LL*,

lll.C.09.10 kZj New c]TMIBank TMI Question #

2 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge PI Comprehension or Analysis Ll 55.43 'a55.45 .7 A Incorrect answer. Exact opposite of true effect of higher decay heat due to trip from high power.

B Correct answer. Higher decay heat level will require more EFW flow to maintain OTSG levels at setpoint.

C Incorrect answer. Exact opposite of true effect if higher decay heat due to trip from high power.

D lncorect answer. Exact opposite of true effect of higher decay heat due to trip from high power.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION N-07 REVISION 5 L' 5.9 Decay Heat 1 5.9.1 Source of Decay Heat Decay heat is heat generated by the deca? of fission products after the actual fission event.

Decay heat generation is included as part of the approximate 200 MeV fission energy release.

5.9.2 Thermal Magnitudes "

Approximately 15 MeV of the total 200 MeV energy release from fission comes from the decay of fission products. This means that while at full power (2568 MW) about 7.5% 1

( 1 5/200) of our heat production comes from decay heat generation (0.075 x 2568 =

192.6 MW). Th'k percentage of heat production is good for any appreciable power level.,

Following a reactor trip almost all core heat production comes from decay of the fission products. This decay heat decreases with time as shown in Figure 55. Notice the curve does not look like a natural decay curve, since many, many fission products are decaying, each with its own half-life, as well as many daughter products.

The production of heat by decay of fission products is of major concern to designers of ECCS Systems and Emergency Feedwater Systems, since these systems must be available for emdrgency conditions to remove core decay heat and prevent significant fuel cladding damage due to extended core uncovery and fuel overheating, even though the reactor is shut down.

160

SYSIEP# Q6J KA# K6.01 Page # 3.4-46 Tier # -

2 L

ROISRO Importance Rating 2.8* 2.5

- Group # 1 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners Plant conditions:

- Reactor trip from 100% power due to loss of both Main Feedwater Pumps.

- EFW flows:

- OTSG 1A = 0 gpm.

- OTSG 1B = 400 gpm

- Status indications above the EF-V-30 controllers for OTSG IA:

- AUTO light is illuminated.

- OP light is illuminated.

- SU light is NOT illuminated.

Based on these conditions, identify the ONE selection below that completes the following statement to describe the only method to control the required level in OTSG IA.

Transfer EF-V-30A andlor EF-V30D controller(s) to A. Local Auto and dial in a 25% setpoint.

B. Local Auto and dial in a 50% setpoint.

C. Manual and control level manually at 50%.

11 D. Manual and control level manually at 25 inches.

OPM Section F-10, Heat Sink Protection System, pages 13 and 16, Rev. 9.

None.

IV.E.05.04

- New , . TMI Bank TMI Question # 2001 SRO Audit

56 2l Modified TMI Bank Parent Question # 2001 SRO Audit

56 El Memory or Fundamental Knowledge Ei?l Comprehension or Analysis 131 55.41 .7 El 55.43 a 55.45 .7 A Incorrect answer. Plausible if the examinee does not diagnose the failure to select SU Range level (setpoint should be 25-inches on SU Range).

B Incorrect answer. Plausible if the examinee does not diagnose the failure to swap to the proper level range.

C Incorrect answer. Incorrect setpoint and level range for these conditions - correct for other conditions.

D Correct answer. This is the correct setpoint and range for loss of both FW Pumps. Manual control is required since the control system did not select the proper (SU) level instrument for use bu automatic control.

TMI SRO Exam - May 2003 Friduy, March 28,2003

SECTION F-10 REVISION 9 U

d. Test Panel Components - Behind alarmed door panels are located 1) test switches,

2) test input jacks, 3) bypass switches, and 4) instrument selector switches. Test switches, bypass switches, and corresponding external indicators are listed on Table 1. Refer to Figure 13 to explain operation of test and bypass switches.

1. A test switch if operated places an individual channel function in the actuated state. To prevent switches fiom being left in this condition, switch covers ensure that the switch is properly placed prior to closing the test panel.

2. Analog input test jacks are provided where needed for calibrations

- - 3. Bypass switches prevent channel actuation from inputting to the 2 out of 4 logic. The bypass switch does not bypass an actuated signal caused by a test switch.

4. Instrument selector switches located in the train cabinets are available to swap input instruments for EF-V-30 control (See Figure 7 & 8 and Table 1). An additional set of toggle switches is used to select which level signals will be used for console indication and potential ICs inputs.

e. Power Supplies - Power supplies were discussed in previous section.

f. Extern$ Indicators (See Table 1) - Indicating lights on the outside of Section A2 RK4 and Section A1 RKj provide HSPS status locally for use during testing.

These lights can also be used to confirm control room HSPS status indicators.

There are ( 2 ) fuses for these lights located above the train test panel.

5.1.4 Logic and Actuation The HSPS performs two types of actuation. It initiates EFW or isolates MFW.

EFW Initiation There are two trains of EFW initiation. Both trains are designed to actuate following:

a. Lo OTSG Level (<lo S/U Level)

b. High RB pressure (>4 psig)

C. Lost of all 4 RCPS (pump power monitors)

d. , Lossofboth FWP turb hyd oil pressure switches) pressure, loss of all 4 RCPS and lossof both FWPS will initiate EFW to both Low level actuation will only initiate EFW to the OTSG with low level. Train A ated starts EF-P-2A, opens MS-V-13A & B and transfers EF-V-30A & C setpoints from 0 to 25. Train B starts EF-P-2B, opens MS-V-13A & B and transfers EF-V30B & D setpoint from 0 25. On loss of all RCPS, setpoint selection is to 50%

operating range level instead o 13

SECTION F- 10 REVISION 9 Cont. Press Hi Hi Hi Hi Lit if channel senses RB press >4 psig (see Figure 6 )

OTSG A Level Lit if channel senses OTSG level <lo (Pump Logic) (see Figure 1)

OTSG B Level Press Lo Lo Lo Lo OTSG A Lo light lit if channel senses press <600 (see Figure 5 )

Level HiHi Hi& Hi& &Hi Press Lo Lo Lo HI Hi light lit if channel senses OTSG Level >97.5% (see OTSG B Figure 4)

Level HiHi HiHi &Hi HiHi I I Feed Lit if both feed pump turbine switches sense low hyd. oil press. (<75 psig) (see Figure 3).

pumps 1 1 E p s Lit ifall four RC pump power monitors sense low power (see Figure 2) n signal is present or locked into related EF-V-3 0 controller.

OTSG level Iayed on the 30 5.2.2 Defeamnable (Refer to Figure 19)

These controls are normally operated in accordance with OP 1 102-1, 1102-2, 1 102-10 &

I 1102-11 .

a. EFW Actuation Defeafinable Controls There are Tmin A and B defeatenable switches for each condition which initiates EFW. There are four associated with Train A on console left and four associated with Train B on console center. To completely defeat actuation due to any one of the following conditions both switches (Train A and B) must be placed in defeat.

0 .Loss of Fw pumps 8 Loss of all RCPs 0 Lo Lo OTSG level (single DEFEN switch for OTSG A or OTSG B) 0 Hi RB pressure 16

Page # 2-2 Tier # -

2

'd ROlSRO Importance Rating 2.5 -

3.3 Group # -

1 Knowledge of system status criteria which require the notification of plant personnel: D.C.

Electrical Distribution System Plant conditions:

- Reactor power is 1OO%, with ICs in full automatic.

- MU-P-1A (Makeup Pump I A ) is supplying normal makeupkeal injection.

Sequence of events:

- The following alarms, actuate simultaneously:

- A-I -7 Battery 1A Discharging

- A-2-7 Batt Charger l N l C / 1E Trouble

- A-3-7 Inverter I A l l C / lE System Trouble

- PRFI-1-1 CRDM Breaker Test Trouble

- "-3-1 230 KV Substation Trouble

- AA-3-2 7KV Bus Trouble

- AA-3-3 4KV BOP Bus Trouble

- AA-3-5 480V BOP Bus Trouble Based on these conditions identify the ONE selection below that describes ( I ) the controlling procedure and (2) required actions.

A. (1) 1202-9A, Loss of "A" DC Distribution System (2) Notify Auxiliary Operator to close EG-V-l5A, air start isolation for Emergency Diesel Generator IA.

L' B. ( I ) 1202-9A, Loss of "A" DC Distribution System.

(2) Notify Auxiliary Operator to verify MU-P-1C is ready for start.

C. (1) Alarm response for AA-3-2, 7KV Bus Trouble.

(2) Notify Transmission System Operator (TSO) and trip the reactor.

D. (1) Alarm response for "-3-1, 230 KV Substation Trouble.

(2) Notify Transmission System Operator (TSO) and trip the reactor.

EP 1202-9A, Loss of "A" DC Distribution, page 4, Rev. 43.

None.

IV.G.08.21 E New a TMIBank TMI Question #

El Modified TMI Bank Parent Question #

Li Memory or Fundamental Knowledge

@ Comprehension or Analysis F 55.41 a 55.43 .5 Ed 55.45 .I2 A Correct answer.

B Incorrect answer. Part # I is correct procedure, but part #2 is not correct action u C lncorrect answer. Incorrect procedure, and incorrect actions.

D Incorrect answer. Incorrect procedure, and incorrect actions.

TMI SRO Exam May 2003 Friday, March 28, 2003

Q # 071 None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 Emergency Procedure 12028A

~ Title Revision No.

Loss of "A" DC Distribution Svstem 43 B. Manual Action CAUTION

1. Because of loss of control power to the switchgear listed in symptoms, affected breakers must be operated manually.

2. Closing the DC Tie Switches may cause a total loss of DC.

3. Loss of control power to MU-P-1A with the normal Make-up system alignment leaves only MU-P-I B for make-up and seal injection. Monitor seal injection flow if a plant transient requires augmentation of RCS make-up flow via MU-V-217.

4. FW-P-1A motor speed changer is inoperable.

NOTE The unit should continue to operate unless tripped by a loss of main 1

condenser vacuum or high RCS pressure.

I. DISPATCH operators to perform the following:

a. DETERMINE cause of failure at DC Switchgear

b. At EG-Y-1A

1. CLOSE EG-V-15A to allow lube oil booster to refill and starting air receiver to repressurize.

2. VERIFY Starting Air Pressure gauges EG-PI-535NB on the Engine Mounted Instrument Panel decrease to 0 psig.

3. TRIP EG-Y-1A fuel rack.

C. IF DC is lost to the D/G AND its 4160V ES bus while operating in parallel with the grid, THEN immediately TRIP the generator breaker G1-02 using the manual plunger at breaker 1D2 on 1D ES 4160V Bus.

d. TRANSFER "Excitation Cntrl Alt Power Supply" to "B" side DC power at Alterrex Cabinet on the 322' elevation south end of the turbine bldg in the 4KV switchgear room with controlled key #22.

4

L RO/SRO Importance Rating 2.6? -

2.9? Group # -

1 Knowledge of D.C. Electrical Distribution System design feature(s) and/or interlock(@which provide for the following: Trips.

Sequence of events:

- Reactor power is loo%, with ICs in full automatic.

- Loss of B 125/250V DC distribution system.

- Reactor is manually tripped.

- Shift management directs CRO to immediately secure all RCPs.

- CRO announces that all 4 RCP breaker closed lights are de-energized, but the ammeters are functional.

Based on the conditions above, identify the ONE statement below that describes how to accomplish this task.

A. Insert key and rotate local 69 Switch to the TRIP position.

B. Depress local TRIP pushbuttons on front of RCP breaker cubicle doors.

C. Open RCP motor breakers using normal control room extension controls.

D. De-energize both 7KV buses, using Panel Right feeder breaker extension controls.

OS-24, Conduct of Operations During Emergency and Abnormal Events, section 4.2.2.8, page 16, Rev. 7.

.-/

New 17 TMI Bank TMI Question #

c1 Modified TMI Bank Parent Question #

EZj Memory or Fundamental Knowledge C Comprehension or Analysis E 55.41 .7 13 55.43 5 55.45 A Incorrect answer. This operation is not possible.

B Incorrect answer. Therse buttons are not functional in the fully connected position.

C Incorrect answer. Control switches are not functional during loss of DC control power.

D Correct answer. Control power for these breakers is from 1A DC.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

TMI - Unit I Operations Department I Number u Administrative Procedure OS-24 Title Revision No.

Conduct of Operations During Abnormal and Emergency Events 7 B. It is not practical to provide a contingency action for every step in the EVENT PROCEDURE. e completion of a step and no continge rvisor may direct contingency actions that meet the intent of the EVENT PROCEDURE step.

Examples:

When SCM is lost, RCPs must be immediately tripped (within 1 minute). If:= RCP breaker fails to open, the 6900V bus can be<

de-energized to stop the RCP.

a If a valve identified in an EVENT PROCEDURE cannot be closed then the corresponding block valve is closed to perform the isolation function. If a block valve is not included in the system then a valve immediately upstream or downstream is closed to perform the isolation function.

C. if an EVENT PROCEDURE item cannot be completed due to plant conditions, continuing with the procedure may be permitted with the concurrence of the Unit Supervisor.

0. Efforts to recover failed equipment that impacts mitigation strategy are continued unless otherwise directed by the Emergency Director.

4.3 Command and Control 4.3.1 The Unit Supervisor directs &IEVENT PROCEDURE actions by using the following methods:

A. Direct the EVENT PROCEDURE with the highest priority at the time by providing specific commands or direction to Reactor Operators for each step.

B. Direct a Reactor Operator to perform an AP, EP, EOP Guide, Alarm Response procedure or other referenced procedure while the Unit Supervisor continues to direct a higher priority EVENT PROCEDURE.

C. Direct a Reactor Operator to perform an AP, EP, EOP Guide, Alarm Response procedure or other referenced procedure while the Unit Supervisor monitors the progress of the procedure.

4.3.2 The Unit Supervisor is expected to proceed through an EVENT PROCEDURE without unnecessary delay. The Unit Supervisor avoids the following activities when performing an EVENT PROCEDURE:

In-depth diagnosis Communication with personnel outside the Control Room 0 Any activities that impede use of the EVENT PROCEDURE.

16

u 3.6 Group # -

2 ROISRO Importance Rating 3.2* -

Knowledge of bus power supplies to the following: Control Power (Emergency Diesel Generator System)

E El Plant conditions:

- Reactor power is 1OO%, with ICs in full automatic.

- Normal electrical line-up.

- No testinglmaintenance in progress

- No emergency system actuation signals

- Auxiliary Operator reports discovering EG-Y-1A is,operating.

Based on these conditions, identify the ONE selection below that describes where to send maintenance to investigate this loss of control power event.

A. 1P DC bus.

B. 1Q DC bus.

C. 1M DC bus.

D. 1H DC bus.

EP 1202-9A, Loss of "A"DC Distribution, pages 2 and 3, Rev. 43.

None.

.10.02 w 4 New TMI Bank TMI Question #

0 Modified TMI Bank Parent Question ##

-v Memory or Fundamental Knowledge

-. . Comprehensionor Analysis d 55.41 .7 c 55.43 3 55.45 A Correct answer.

B Incorrect answer. Plausible since loss of this DC bus starts the other emergency diesel generator.

C Incorrect answer. Plausible answer since this bus has redundant DC power supplies, and is used for vitaVemergency equipment.

D Incorrect answer. Plausible since this bus is energized by the same power supply as the 1P DC Bus (correct answer).

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 Emergency Procedure 1202-9A

.-, Title Revision No.

Loss of "A" DC Distribution System 43 NOTE Partial loss of DC caused by a blown fuse to one of the DC Distribution Panels may give similar symptoms. This procedure does NOT provide specific guidance for a blown fuse.

1.0 SYMPTOMS A. Loss of Main Distribution Panel 1A as indicated by alarms:

0 AA-3-2, 7KV Bus Trouble 0 AA-3-3,4KV BOP Bus Trouble 0 AA-3-5, 480V BOP Bus Trouble 0 A-1-7, Battery 1A Discharging (Rate above 100 amps) 0 A-2-7, Battery Charger IN1CI1E Trouble 0 A-3-7, Inverter 1A/lC/I E Trouble 0 K-3-4, MN Turb. PC oil pmp stNtroub 0 L-1-3, Voltage Regulator DC Loss 0 B-3-1, 4KV ES Bus Trouble 0 "-3-1, 230KV Substation Trouble (loss of DCA) 0 PRF-1-1-1, CRDM Breaker Test Trouble (loss of shunt trip) 0 H & V, A-4-2 Cont. Bldg. Batt. Chargers A Damper Tbl, Fire-Smoke 0 Loss of breaker status lights at control switches B. Loss of Main Distribution Panel 1A will result in the following:

0 Loss of all power on the "A" Distribution System.

0 Inability to remotely trip or close breakers on A ESAS System.

0 Loss of Engineered Safeguards Distribution Panel 1E 0 Loss of ES Diesel Generator Dist. Pnl. 1P 0 Loss of 230KV Substation Dist. Pnl. DCA.

Loss of Distribution Panel I C .

2

I Number TMI - Unit 1 Emergency Procedure 1202-9A

\

- Title Revision No.

Loss of "A" DC Distribution System 43 Loss of Control Power to the following:

a. 1A and 1B 6900V Reactor Plant Swgr. Feeder Breakers

b. 1D-416OV ES Swgr C. lA-4160V Turbine Plant Swgr.

d. . lC-4160V Turbine Plant Swgr.

e. 1C-480V Turbine Plant Swgr.

f. 1E-480V Reactor Bldg. Ht. and Vent Swgr.

9. 1G-480V Reactor Plant Swgr.

h. 1H-480V Aux. and Fuel Bldg. Ht. and Vent Swgr.

I. 1J-Q80VTurbine Plant Swgr.

j. 1N-480V Turbine, Reactor and Control Bldgs. Htg. Swgr.

k. 1P-480V ES Swgr.

I. 1R-480V Scrn. Hse. ES Swgr 2.0 IMMEDIATE ACTION A. Automatic Action

1. Suction Valves to VA-P-1A and VA-P-IC close.

2. Emergency Feed Pump EF-P-1 Starts.

3. Emergency Diesel Generator EG-Y-1A starts.

4. Panel 1M automatically transfers to "B" DC Distribution System.

5. RC-RV-2 fails closed and is inoperable.

6. MU-V-6A fails open and MU-V-6B fails closed placing "A" makeup and purification demineralizer in service.

7. MU-V-11B fails open and MU-V-1 1A fails closed placing MU-F-1B in service.

8. CM-V-1 and CM-V-3 fail closed making RM-A-2 inoperable.

3

Page # 3.9-2 Tier # -

2 U

ROlSRO Importance Rating 2.7 -

2.9 Group # -1 Knowledge of the physical connections and/or cause-effect relationships between Liquid Radwaste System (LRS) and the following systems: Sources of liquid wastes for LRS.

a 5 E Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

Based on these conditions, identify the ONE selection below that describes a NORMAL source of water to the Liquid Waste Disposal System.

A. PORV pilot valve leakoff.

B. Leakoff from between the reactor vessel flange 0-Rings.

C. Intermittent drain flow from the Waste Gas Compressor Separator.

D. Valve packing leakage from Letdown isolation valves MU-V-1A and MU-V-1B.

302-696, Waste Gas Compressors Flow Diagram, Rev. 1.

New 0 TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

-\-/

IZj Memory or Fundamental Knowledge El Comprehension or Analysis 55.41 .2 to .9 L l 55.43 M 55.45 .7/.8 A Incorrect answer.

B Incorrect answer. This line is normally isolated.

C Correct answer. Level switch operates solenoid operated valve to drain excess water from the separator to the Auxiliary Building Sump.

D Incorrect answer. Packing leakoff lines are capped.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

I 4

I 10w.m

.lw-w.ID.al >

i I. VNVE POSITICNS Porn FCR REFEREKE W r .

A C T W PDSITICNS CDITRaLED 81 PROCEDSES.

2. ALL INSlGWNI TAG W E U S @? PREFIXED BI s w w DESIGMTCRVCESS OIWWISE WKD.

3. SEl POIN1 VALUS FCR RELIEF V*LVPS AI(E a I N A L Am FCR REFER- MI. KlW D h I A m 1 W BY QS2.

I 4 . ENTIRE C O P a i N T IS N - 3 . S-11.

1. NL mms AFTER usr VUVE PULL BE s-1x1 HDH-HXXEW.

-I

Q # .075 .

Page # 3.9-4 Tier # -2 W Group # -1 ROlSRO Importance Rating 3.6 36 Ability to monitor automatic operation of the liquid Radwaste System (LRS) including:

Automatic isolation.

Form the list below, identify the ONE operating condition that will automatically terminate process flow during a liquid transfer from the Reactor Building Sump to the Auxiliary Building sump.

A. High alarm on RM-A4 Gas (Aux Building Atmospheric Monitor).

B. High alarm on RM-G-5 (RB Area Monitor).

C. Low Reactor Building sump level.

D. High Auxiliary Building sump level.

SS-209-097, RB Sump Level Instrumentation, Rev. 8.

SS-209-306, WDL-V-534, 535,Rev. 10.

None.

IV.B.13.07

@ New 3 TMlBank TMI Question #

L! Modified TMI Bank Parent Question #

_I Memory or Fundamental Knowledge Comprehension or Analysis A

55.43 a 55.45 .5 A Incorrect answer. This parameter does not initiate an automatic interlock to terminate process flow from the RB Sump to the Auxiliary Building Sump.

B Incorrect answer. This parameter does not initiate an automatic interlock to terminate process flow from the RB Sump to the Auxiliary Building Sump.

C Correct answer. This parameter initiates an automatic interlock to terminate process flow from the RB Sump to the Auxiliary Building Sump.

D Incorrect answer. This parameter does not initiate an automatic interlock to terminate process flow from the RB Sump to the Auxiliary Building Sump. Plausible since this is the initial destination for water being transferred out of the RB Sump, as stated in the stem.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

tEfERENCE CWGS.: NOTS: ME N M X . S6-209#1 h AUX RELAY RACK %CR '

LEGEM) ss-maat RELAYS TO 8 E CLARK SU6-2-76 w THREE a$

3. DEV S I T U LPS TO 8EHkYLlH PIN LFFrS&# EL E 4 . SEE DWG LS-29659 (WM EM-10, " O W N FOR SUMP EMPTY 5, JT-LOC.I:N CONT. WTR AYL. CAB "C" Y

Q I

i 1

AS1 RB SUMP LEVEL LIGHTS 7

LEVEL ALARM ELEV. 283'0" LS-116C 2 W 3 112" 279'3

ON occ

"/).

OfF WTTOH

CAE] FILE NO.: 209097-Rt3 RB SUMP LEVEL INSTRUMENTATJ. THIS A COMPUTER GENERATED D R A W " ,

DO NOT REVISE IT MANUALLY.

~ x 756c

/4 FULL

'I'T t

7 1XSt(CI 3

T I "IT T SPARE cs1 Re T

(a)

SUMP LEVEL H w.819

2575 i

SPARE X? T TWS SH.

1T T 1T iED Plf 1 1 1 SH T

REFERENCE DWGS.:

INDEX & LEGEND: SS-209-151 THRU 155 NOTES:

I . ITEMS 58 A h B ON CWOS PNL.

LOGIC DIAGRAM: s-203-Fl94 2. DEVICES MOUNTED ON VERTICAL LWOS PANEL _THREE SS DEYELOPMEHT: SS-209- I59 DEVICES MDUNTED ON E. S. VERT1 CAL P 5 N a PCR .r x DEVICES MOUNTED ON Rw P EL- AUX. BLDG.

- DEV~CES mu rw LOC~LLYTHEAR OR AT MmR, VALVE, EX.!

ENERGIZE VALES TO OPEN NUCLEAR SAFETY R R . 6 . SUMP OUTLE ISOLATION VV. WDL-kV534 3A i

331 bc I

t 5PM 4

ENPINEERS AND CONSULTANTS A

RB SUMP OUTLET ISOLATION

-_I u) 3 ca c!

P z

2 7 c1 1 -t

--331 ao 331 bo A

A

SYSIEP# Q7J KA# A2.02 Page # 3.9-7 Tier # -2 u 3.6 Group # -1 RO/SRO Importance Rating 3.6 -

Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System (WDGS) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Use of waste gas release monitors, radiation, gas flow rate, and totalizer Plant conditions:

- Releasing Waste Gas Decay Tank WDG-T-1A.

- RM-A-7 High Alarm actuates.

Based on these conditions, identify the ONE statement below that describes (1) automatic interlocks actuated and (2) operator actions required.

A. (1) AH-E-I 1, Auxiliary Building Supply Fan, trips.

(2) Close WDG-V-25, WDG-T-1A outlet isolation valve.

B. (1) WDG-V-47, Waste Gas release control valve, closes.

(2) Stop AH-E-I 1, Auxiliary Building Supply fan.

C. (1) WDG-V-47 Waste Gas release control valve, closes.

(2) Implement EP 1202-12, Excessive Radiation Levels.

D. (1) WDG-V-25, WDG-T-1A outlet isolation valve, closes.

(2) Implement alarm response procedure for MAP C-1-1, High Radiation.

EP 1202-12, Excessive Radiation Levels, page 11, Rev. 50.

\/

New 0 TMIBank TMI Question #

0 Modified TMI Bank Parent Question #

L Memory or Fundamental Knowledge

@ Comprehension or Analysis 2 55.41 .5 I E 55.43 .5 E4 55.45 .3/.13 A Incorrect answer. This answer describes an incorrect interlock and incorrect actions.

B Incorrect answer. This answer describes the correct automatic interlock, but the manual actions described are incorrect.

C Correct answer.

D Incorrect answer. The interlock listed is incorrect. MAP C-1-1 will lead the operator to implement EP 1202-12.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 Emergency Procedure 1202-12

.-, Title Revision No.

Excessive Radiation Levels 50 ENCLOSURE 1 Page 2 of 3 E. Gas Waste Tank Discharge Monitor, RM-A7 closes Waste Gas Release Valve, WDG-V47.

F. Auxiliary and Fuel Handling Building Exhaust Duct (gas channel), RM-A8.

a. MAP-5 Iodine-Particulate sampling is initiated upon high alarm on RM-Ai8 gas channel.

b. The following fans shutdown:

A. . Fuel Handling Building Supply Fan (AH-E-I0).

B. Auxiliary Building Supply Fan (AH-E-I 1).

C. Waste Gas Release Valve, WDG-V47, closes.

G. Reactor Building Purge Duct (gas channel), RM-AS:

a. MAP-5 Iodine-Particulate sampling is initiated upon RM-A-9 high alarm on gas channel.

b. The following valves close:

A. Outside Reactor Building Purge Exhaust, AH-VIA.

B. Inside Reactor Building Purge Exhaust, AH-VI B.

C. Inside Reactor Building Purge Supply, AH-VI C.

D. Outside Reactor Building Purge Supply, AH-VI D.

E. Reactor Building Sump Outlet Isolation, WDL-V534.

F. Reactor Building Sump Outlet Isolation Valve, WDL-V535.

B. Interlock functions actuated by Hi Alarms from specific liquid radiation monitor are as follows:

A. RCS Letdown Line Monitor, RM-L-1 Hi closes RCS Letdown Isolation Valves MU-V-2A and MU-V-2B.

B. Radioactive Waste Water Discharge Monitor, RM-L6 closes Mechanical Draft Cooling Tower Stop Valve, WDL-V257.

11

Page # 3.7-13 u 3.5 Group # 1 ROlSRO Importance Rating 3.1 -

Knowledge of the effect that a loss or malfunction of the Area Radiation Monitoring (ARM)

System will have on the following: Fuel handling operations Refueling operations are scheduled to begin in one hour. From the list below, identify the ONE operational condition that would require the Unit Supervisor to delay handling of irradiated fuel inside the RB.

A. RM-A-2 (RB Atmospheric Monitor) sample pump trips.

B. RM-G-22 (RB High Range Monitor) is declared inoperable.

C. RM-A-9 (RB stack) automatic interlock surveillance test fails.

D. RM-G-5 (RB Personnel Access Hatch) monitor FaiVReset light de-energizes due to off-scale low indication.

Technical Specification 3.8.9, page 345, Amendment 236.

Technical Specification 3.8.9, page 3-45, Amendment 236.

New E TMI Bank TMI Question #

r? Modified TMI Bank Parent Question #

PI Memory or Fundamental Knowledge Comprehension or Analysis L-f lid 55.43 .7 1355.45 .6 A Incorrect answer. RM-A-2 is not required for Fuel handling Operations.

B Incorrect answer. RM-G-22 (High Range monitor) is not required for Fuel handling Operations.

C Correct answer.

D Incorrect answer. RM-G-5 is not required for Fuel handling Operations.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

- 3.8.8 If any of the above specified limiting conditions for fuel loading and refueling iue not met, movement of fuel into the reactor core shall cease; action shall be initiated to ciirrect the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.

3.8.9 The reactor building purge system, including the radiation monitors which initiate purge isolation, shall be tested and verified to be operable no more than one week prior to refueling 0peEltiOn.S.

3.8.10 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.8.11 During the handling of irradiated fuel in the Reactor Building at least 23 feet of water shall be maintained above the level of the reactor pressure vessel flange. If the water level is less than 23 feet above the reactor pressure vessel flange, place the fuel assernbly(s) being handled into a safe position, then cease fuel handling until the water level has been restored to 23 feet or greater above the reactor pressure vessel flange.

-Bases Detailed written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.7 of the UFSAR incorporating built-in interlocks and safery features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. If no change is being made in core geometry,one flux monitor is sufficient. This pennits maintenance on the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a unifcrm boron b

concentration. The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core (Reference 1). The boron concentration will be sufficient to maintain the core kc^< 0.99 if all the control rods were removed from the core, however only a few controt rods will be removed at any one time during fuel shuffling and replacement. The bit with all rods in the core and with refueling boron concentration is approximately0.9. Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.

Per Specification 3.8.6 and 3.8.7, the personnel and emergency air lock doors, and penetrations may be open during movement of irradiated fuel in the containment provided a minimum of one door in each of the air locks, and penetrations are capable of being closed in the event of a fuel handling accident, and the plant is in REFUELING SHUTDOWN or REFUELING OPERATION with at least 23 fket of water above the fuel seated within the reactor pressure vessel. The minimum water level specified is the basis for the accident analysis assumption of a decontamination factor of 200 for the release to the containment atmosphere from the postulated damaged fuel rods located on top of the fuel core seated in the reactor vessel. Should a fuel handling accident occur inside containment, a minimum of one door in each personnel and emergency air lock, and the open penetrations will be closed foltowing an evacuation of containment. Administrative controls will be in place to assure closure of at least one door in each air lock, as well as other open containment penetrations, following a containment evacuation.

Provisions for equivalent isolation methods in Technical Specification 3.8.7 include use of a material (e.g. temporary sealant) that can provide a temporary, atmospheric pressure ventilation barrier for other containment penetrations during fuel movements.

3-45 v Amendment No.W b W , 236

Form ES-401-6 SYSIEW 072 KA# A1.O1 Page # 3.7-14 Tier # -

2 W'

ROISRO Importance Rating 3.4 3.6

- Group # -1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Area Radiation Monitoring (ARM) System controls including:

Radiation levels.

Plant conditions:

- Refueling operations in progress.

- DH Train A in service.

- Refueling Canal water level is reducing.

Identify the ONE monitor below that, if rising, would indicate the leak described above is into the operating Decay Heat Closed Cooling System.

A. RM-L-2.

B. RM-L-3.

C. RM-L-4.

D. RM-L-5.

OPM Section F-07, Radiation Monitoring System, page 58, Rev. 17 None.

.01.06 0 New TMI Bank TMI Question # AL4E06-02-QO3

'.--/'

0 Modified TMI Bank Parent Question #

E Memory or Fundamental Knowledge C Comprehension or Analysis d 55.43 .4 @ 55.45 .5 A Correct answer. RM-L-2 is associated with Loop A DCCS System.

B Incorrect answer. RM-L-3 is associated with Loop B DCCS System.

C Incorrect answer. RM-L-4 is associated with the NSCC System.

D Incorrect answer. RM-L-5 is associated with the Spent Fuel Cooling System.

TMI SRO Exam - May 2003 Friday, March 28,2003

i SECTION F-07 REVISION 17 TABLE 1 Awr. Bldg. (Near Top of A DfIR Vault)

RM-L-3 B DHCCW NaI 10-10 cpm r

Elev. 28 1 RM-L-4 NSCCW Aux. Bldg. (Near NSCCW pumps) Elev. 305 NaI 10-lo6 cpm RM-L-5 Spent Fuel FH Bldg. (Near SF Coolers/Pumps) Elev. 305 GM TU~X io- 1o6 cpm RM-Ld Liquid Radioactive Waste Discharge Am. Bldg. (near WECST) Elev. 281 Nd 10-106cpm Closes WDL-V-257 J

. . RM-L-7 Plant Water Discharge Rh4-L-7 Bldg. (near hfJX3) NaI 10-106 cpm Closes WDL-V-257 i

RM-L-9 ICCW FN Bldg. (Near ICCW Surge Tank)Elev. 348 NaI 10- 1o6 cpm Trips IW-P- 1 6,17& 18 RM-L-12 IWTS/IWFS Effluent IWTS/IWFS Bldg. NaI 10-106cpm Trips IW-P-29 & 30 Closes IW-V-73 & 279 J I 58

SECTION k -h REVISION 17 TABLE 1 I

Aux. Sldg. (Near Top of A DHR Vault)

RM-L-3 RM-L-4 B DHCCW NSCCW Elev. 281 Aux. Bldg. (Near NSCCW pumps) Elev. 305 NaI NaI 10-lo6cpm IO-lo6cpm RM-L-5 Spent Fuel FHBldg. Wear SF Coolers/Pumps) Elev. 305 GM TU& IO-io6cpm RM-L-6 Liquid Radioactive Waste Discharge Aux. Bldg. {near WECST) EIev. 281 NaI IO-lo6cprn closes WDL-V.257 J RM-L-7 Plant Water Discharge RM-L-7 Bldg. (near MDCT) NaI 10-106 cpm Closes WDL-V-257 RM-L-9 ICCW FH Bldg. (Near ICCW Surge Tank)Elev. 348 NaI 10-lo6 cpm Tips I W-P- 1 6,17& 18 Rh4-L-I2 JWTSDWFS Effluent IWTWWFS Bldg. Nal 10-106cpm Trips [W-P-29 & 30 Closes IW-V-73 & 279 J I 58

- Page # 3.7-15 v ROISRO Importance Rating 4.0 4.3 Group # -2 Knowledge of Process Radiation Monitoring (PRM) System design feature@)and/or interlock(s) which provide for the following: release termination when radiation exceeds setpoint.

From the list below, identify the ONE set of conditions that can automatically terminate an inadvertent radioactive liquid release from the Waste Evaporator Condensate Storage Tank (WECST) in the event the wrong tank is released.

A. RM-L-7 in ALERT alarm OR loss of sample flow through RM-L-6.

6. RM-L-6 HIGH alarm OR high tank release rate.

C. High MDCT effluent flow OR low tank release rate.

D. RM-L-6 ALERT alarm OR RM-L-7 ALERT alarm.

OPM Section E-02, Waste Disposal Liquid System, Rev 13, Page 23.

E New d TMlBank TMI Question # #64 1998 SRO 0 Modified TMI Bank Parent Question #

&?I Memory or Fundamental Knowledge

, / G Comprehension or Analysis PI 55.41 .7 PI 55.43 .4 c 55.45 A Plausible distracter since RM-L-6 and RM-L-7 monitor liquid release flow paths. Misconception that loss of detector sample flow will terminate the radioactive liquid release or that an ALERT alarm will conservatively terminate a liquid release prior to reaching the HIGH alarm..

B Correct answer.

C Plausible distracter which reverses high and low flow interlocks between tank release rate and effluent (dilution) flow rate.

D Plausible misconception that an ALERT alarm will conservatively terminate a liquid release prior to reaching the HIGH alarm..

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION E-02 REVISION 12 FIGURE 3 - WL-V-257 CONTROL CIRCUIT 1

1 1,RM-L7 33x257 l- sv-267 20x257 1

I RESET

\ NORMAL (KEY SWITCH)

< 20x251 WDLV-257 CONTROL CIRCUIT (WECST EFFLUENT RELEASE ISOLATIONVALVE)

Valve shown CLOSED or SV-257deenergized. If the similar Unit 2 valve is closed, and RM-L-6 not at HIGH level, and the release rate not HIGH, and Mechanical Draft Cooling Tower effluent flow not LOE, and RM-L-7 to a HIGH level, all of the above associated contacts would be closed. Also, the 33 x 257 Relay would be energized indicating the Unit 2 valve closed at the Radwaste Panel. Then upon RESET of WDL-V-257 Key Switch, the 20 x 257 Relay is energized and its associated contacts wiiI cIose, seaIing in the WDL-V-257 Key Switch allowing the switch to be returned to the N O W position. When WDLV-257 is desired open, the CLOSE/AUTO Switch on the Radwme Panel is placed in the AUTO position. This energizes the SV-257Relay (Solenoid) and will allow operation of the valve through the WDL-V-57valve position controller.

Anytime a contact opens in series with the 20 x 257 relay, the relay is de-energized which opens its associated contacts and de-energizes the SV-257d a y (solenoid), causing closure of WDLV-257,'Closuxe may also be accomplished by placing the CLOSE/AUTO Switch for WDL-V-257to CLOSE position. If the similar valve on Unit2 is opened, WDGV-257will close and both Unit 2 Valve and WDL-V-257 positions indicated on Radwaste Panel..

23

Page # 3.7-16 Tier # -2 u ROISRO Importance Rating 3.2 3.5

- Group # -

2 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Process Radiation Monitoring (PRM) System controls including:

Radiation levels.

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- RM-L-1, RCS Letdown High Range Monitor, is out of service for calibration testing.

Event:

- Fuel pin failure occurs.

Based on these conditions, assuming no operator action, complete the statement below that predicts the effect of this event.

Auxiliary Building general area radiation levels.. .

A. RISE until letdown automatically isolates.

B. RISE with NO automatic letdown isolation.

C. DO NOT CHANGE since MU-V-IAIB are required to be closed before any RM-L-1 testing.

D. DO NOT CHANGE since MU-V-2AIB are required to be closed before any RM-L-1 testing.

EP 1202-12, High Radiation Levels, page 11, Rev. 50.

4 New T TMI Bank TMI Question #

4 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge H Comprehension or Analysis H 55.43 .4 ! I55.45 A Incorrect answer. Plauseible that the Examinee might think letdown isolation on high radiation is initiated by RM-L-1 Lo, which is still in service.

B Correct answer.

C Incorrect answer. Plausible distracter since the monitor that initiates automatic letdown isolation is out of service during this test.

D Incorrect answer. Plausible distracter since the monitor that initiates automatic letdown isolation is out of service during this test.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 Emergency Procedure 1202-12 Title Revision No.

Excessive Radiation Levels 50 E. Gas Waste Tank Discharge Monitor, RM-A7 closes Waste Gas Release Valve, WDG-V47.

F. Auxiliary and Fuel Handling Building Exhaust Duct (gas channel), RM-A8.

a. MAP-5 Iodine-Particulate sampling is initiated upon high alarm on RM-A-8 gas channel.

b. The following fans shutdown:

A. , Fuel Handling Building Supply Fan (AH-E-I 0).

B. Auxiliary Building Supply Fan (AH-E-I 1).

C. Waste Gas Release Valve, WDG-V47, closes.

G. Reactor Building Purge Duct (gas channel), RM-AS:

a. MAP-5 Iodine-Particulate sampling is initiated upon RM-A8 high alarm on gas channel.

b. The following valves close:

A. Outside Reactor Building Purge Exhaust, AH-VIA.

B. Inside Reactor Building Purge Exhaust, AH-VI B.

C. Inside Reactor Building Purge Supply, AH-VI C.

D. Outside Reactor Building Purge Supply, AH-VI D.

E. Reactor Building Sump Outlet Isolation, WDL-V534 F. Reactor Building Sump Outlet Isolation Valve, WDL-V535.

6. Interlock functions actuated by Hi Alarms from specific liquid radiation monitor are as follows:

A. RCS Letdown Line Monitor, RM-L-1 Hi closes RCS Letdown Isolation Valves MU-V-2A and MU-V-2B.

B. Radioactive Waste Water Discharge Monitor, RM-L6 closes Mechanical Draft Cooling Tower Stop Valve, WDL-V257.

11

Q # ,081

- Page # 2-16 Tier # 2

.u RO/SRO Importance Rating 4.0 -

4.0 Group # 2 Ability to perform without reference to procedures those actions that require immediate operation of (Steam Generator System) system components and controls.

ci 0 Plant conditions:

- RCS cooldown in progress due to OTSG tube leakage.

- RCS T-Ave = 510°F.

- OTSG 1B tube leakage = 5 gpm.

- OTSG 1A TBV status: 10% open in Manual.

- OTSG 1B TBV status: 0% open in Manual.

- Report from A 0 that "BI side Main Steam Safety (MSSV) valve is stuck open.

- OTSG pressures = 700 psig.

Based on these conditions, procedural guidance is to steam the MOST AFFECTED OTSG to the condenser while monitoring condenser vacuum.

Identify the ONE selection below that describes:

(1) How to implement these actions.

(2) The objective of the procedure guidance.

A. (1) Open OTSG 1B TBVs to reduce OTSG 1B pressure.

(2) Attempt to reseat the MSSV.

B. (1) Open OTSG 1B TBVs while closing OTSG 1A TBVs.

(2) Raise RCS cooldown rate.

xu C. (1) Open OTSG 1B TBVs while closing OTSG IATBVs.

(2) Reduce radioactive release rate out the MSSV.

D. (1) Open OTSG 1B TBVs.

(2) Increase OTSG 1B FW flow to dilute the radioactive release out the MSSV.

OP-TM-EOP-005, OTSG Tube Leakage, page 1I,Rev. 1.

@ New 0 TMI Bank TMI Question #

E l Modified TMI Bank Parent Question #

c1 Memory or Fundamental Knowledge

@Comprehension I or Analysis k!l 55.41 . I O 0 55.43 .4 &!I 55.45 .6 A Incorrect answer. Correct operation, but incorrect objective.

B Incorrect answer. Incorrect operation, and incorrect objective (plausible due to Guide 6 instructions for stuck open MSSV).

C Correct answer.

id- D Incorrect answer. Incorrect operation, and incorrect objective (plausible, since this operation would actually dilute the release out the MSSV).

TMI SRO Exam - May 2003 Friday, March 28,2003

0P-TM-EOP-005 Revision 1 Page 11 of 21 ACTlONlEXPECTED RESPONSE RESPONSE NOT OBTAINED 3.24 INITIATE Emergency Boration IAW Rule. 5.

3.25 When SCM has been minimized, then INITIATE plant cooldown IAW 1102-1I.

CAUTION I- .. ...

Steaming an OTSG that has a release path to the atmosphere can reduce the release rate.

However, the additional non-condensable gas flow may degrade condenser vacuum. At 23 -

HgVac TBV control will be lost and all steam flow will go to atmosphere through the ADVs.

3.26 VERIFY there is poJ a steam If condenser is available, then release path directly to the preferentially STEAM the MOST atmosphere. affected OTSG to the condenser and MONITOR condenser vacuum.

IAAT HPVPORV cooling was initiated and SCM 2 25°F and at least

-one OTSG is available as a heat sink, RCS pressure is being controlled with a steam bubble in the pressurizer and pressurizer level can be maintained stable above 80 without the use of HPI, then PERFORM the Recover from HPI Cooling

....... - ........ .....................Section....... -of .......................

OP-TM-EOP-009. - ..

SYSIEW 075 KA# K3.07 Page # 3.8-15 Tier # -

2

-v 3.5* Group # -2 RO/SRO Importance Rating 3.4* -

Knowledge of the effect that a loss or malfunction of the Circulating Water System will have on the following: ESFAS.

At full power, a loss of will cause all three EFW (Emergency Feedwater) pumps to auto-start A. Both Intermediate Closed Cooling Water Pumps.

B. Six Circulating Water Pumps.

C. 1A Auxiliary Transformer.

D. 1B Station Battery.

MAP N-1-6, Main Cond Vac Lo, Rev. 17.

MAP M-1-1, FWP 1A Trip Rev. I O .

New Z TMIBank TMI Question # # I 3 on SR021 Audit I- Modified TMI Bank Parent Question #

- Memory or Fundamental Knowledge

.'L

' --- E Comprehension or Analysis 9 55.41 .7 0 55.43 i? 55.45 .6 A Incorrect answer. Both Feedpumpswill not trip. All 4 RCPs will not trip. RB pressure will not rise to 4 psig.

OTSG levels will not reduce to < I O-inches.

B Correct answer.

C Incorrect answer. Loads auto transfer to 1B Auxiliary Transformer, and EG-Y-1A. Both Feedpumps will not trip. All 4 RCPs will not trip. RB pressure will not rise to 4 psig. OTSG levels will not reduce to 40-inches.

D Incorrect answer. Both Feedpumps will not trip. All 4 RCPs will not trip. RB pressure will not rise to 4 psig.

OTSG levels will not reduce to 40-inches.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

I Number TMI - Unit 1 1-Alarm Response Procedure MAP N Title Revision No.

Main Annunciator Panel N (See Cover Page)

N-I -6 Revision 17 ALARM:

MN COND VACUUM LO SET POINTS:

Variable Setpoint based on load. (See DTCS Screen #13)

CAUSES:

Loss of main condenser vac. pumps Loss of circulating water system 1 t o w circ. water system flow High circ. water temperature Loss of Auxiliary Boiler on startup Abnormal air in leakage to the main condenser due to improper valve lineup or system leaks Air binding in water boxes (low probability)

AUTOMATIC ACTION:

1. The standby vac. pump starts if the running vac. pump tripped

2. If less than 2 CW pumps are running or vac. 23" the atmosphere dump valves (MS-V-4NB) assume pressure control from the Steam Dump valves, MS-V-3A-F.

3. On an increase of backpressure to trip setpoint shown on DTCS Screen # I 3 (2 out of 3 switches) on any one hood, the main turbine trips. Reactor trips if turbine trips and > 45% power OBSERVATION (CONTROL ROOM):

1. Cond. vac. abnormal as indicated by CO-PI-340, CO-P1-73/74/75 on panel "CL".

2. Point No. TLOVAC "Mn. Cond. Vac. Low" alarm, and Pt. A0033, 34, 35 3rd, 2nd, 1st stage cond.

pressure on computer.

3. Increased air flow through the vac. pumps (PR Recorder) (RM-R-15).

4. Main Turbine exhaust hood temp. increasing as indicated by Points TA018, TA019, TA020.

5. "Aux. Boiler Trouble" alarm on the Mn. Annunciator Panel "-2-6.

6. If MS-V-4's assume pressure control from the MS-V-3's, the MS-V-3A-F Power Reset Light will be out.

7. Computer alarm S5331 will be actuated in an alarm state on the PPC alarm printer if the interlock actuated to transfer control to the MS-V-4's from the MS-V-3's.

8. Condenser vac. abnormal as indicated on DTCS Screen #13.

Page 1 of 3

Number TMI - Unit 1

'L-i Alarm Response Procedure MAP N Title Revision No.

Main Annunciator Panel N (See Cover Page)

MANUAL ACTION REQUIRED:

a. Verify Auto Action

b. Attempt to restart the tripped vac. pump if more than one tripped.

C. If an increase in air inleakage is suspected, start additional vacuum pump(s) as needed.

d. Start additional circulating water pumps or verify current C.W. lineup per OP 1104-9

e. If the Aux. Blr. trips during startup, restore same to service as soon as possible in accordance with OP 1106-04, Aux. Blrs.

f. If air binding in the water boxes is suspected, manually vent them by opening CW-V-108 A/B/C/D.

Although unlikely, air binding can occur if the water boxes auto vents do not function or are isolated AND if air is entrained in CW flow. Air entrainment should be suspected if the fume level is low and the CW

g. Check the area of the main condenser for air in leakage due to improper valve position, leaking flanges, etc. Fill the exhaust boots.

Use absolute pressure indicator CO-PI-73 on CL for determining LP-C backpressure. This indicator will normally agree closely with computer point A0033. Disagreementsof > 0.2" Hg should be investigated and repaired. If normal indication is suspected of being out of calibration, an independent check of absolute backpressureon LP-C can be obtained by using an absolute pressure test instrument on exhaust hood taps at EX-V- 134f 141I1 49f 145.

h. If vac. continues to degrade, or absolute pressure has been verified greater than allowable for plant load, notify dispatching and reduce load as necessary to limit back pressure. Load reduction for the purpose of improving exhaust pressure should not go below 30% rated output while above 5.0" Hga.

If load reduction does not reduce pressure below 5.0" Hga when 30% load is reached, the turbine should be tripped off line and shutdown.

I. If reactor trips, go to ATP 1210-1.

j. If turbine trips, refer to Alarm Response K-1-1, Page 2 of 3

Number TMI - Unit 1

.u Alarm Response Procedure MAP N Title Revision No.

Main Annunciator Panel N (See Cover Page)

N-I -6 Revision 17 MANUAL ACTION REQUIRED: (Cont'd)

I NOTE If vacuum is restored with ICs controlling MS-V4NB, OTSG pressure control will be lost (MS-V-3 & 4 closed) until 'reset' PB (CC) for MS-V-3A-F is depressed or OTSG pressure reaches 1026.

k. If OTSG pressure control transfers to MS-V4NB as indicated by the MS-V-3A-F Power Reset Light being out and alarm S5331 being actuated on the alarm printer, then send an operator to the ICSlNNl power monitor cabinet to obtain indications of MS-V-3's and 4's status and report it to the Control Room.

I. Restore MS-V-4NB to the backup loaders prior to restoring vacuum per 1105-4.

Page 3 of 3

Number TM I NUCLEAR Alarm Response Procedure MAP M I

'\, TLte Revision No.

Main Annunciator Panel M (See Cover Page)

ALARM:

F W 1A TRIP SET POINTS:

2 of 3 pressure switches indicate 75 psig decreasing oil pressure (FW-PS-?6A, FW-PS-I5A, RIV-PS-1484A)

CAUSES:

1. LOSSof condensate or condensate booster pumps (when needed in condensate circuit).

2. Low bearing oil pressure.

3. Loss of vacuum. 'I

4. Excessive thrust bearing wear.

5. Feed pump suction valve closed.

6. Local manual trip handle pulled,

7. Overspeed AUTOMATIC ACTION:

I. Auxiliary oil pump starts when oil pressure decreases to 43 psig.

2. If both feed pump turbines are operating and plant is above 60 percent load, and ICs in auto, Unit Load Demand will run back to 60 percent.

3. If only one feed pump turbine is operating, loss of this feed pump turbine will result in main turbine trip, EFW Actuation and Rx Trip if Pwr is > 7%.

4. FW-V-1A will close on pump trip except if caused by loss of vacuum, overspeed, or [ocal manual trip.

OBSERVATfON (CONTROL ROOM):

I. RCS pressure increasing

2. OTSG levels and feedwater flow decreasing

3. Change in feed water valve position L, Page 1 of 2

Form ES401-6

'-/

RO/SRO importance Rating 2.2 3.6

- Group ## -2 Knowledge of which events related to system operationdstatus should be reported to outside agencies: Containment System From the list below, identify the ONE selection that describes (1) an operating condition that exceeds a design limit, requiring NRC notification, and (2) the basis for the design limit.

A. (1) Reactor power I%, with average RB temperature above 320' elevation at 135°F.

(2) Maintain Core Flood Tank water within acceptable temperature limit.

B. (1) Reactor at Hot Shutdown, with average RB temperature above 320' elevation at 135°F.

(2) Prevent exceeding containment design temperature and pressure during a LOCA.

C. (1) Reactor power I%, with RB pressure at 2.0 psi vacuum.

(2) Prevent exceeding reactor building structure differential pressure limits.

D. (1) Reactor at Hot Shutdown with RB pressure at 2.0 psi vacuum.

(2) Maintain sample flow to RB atmospheric radiation monitor.

Technical Specification 3.6.4, page 3-41, Amendment 198.

Technical Specification 3.17, page 3-80, Amendment 157.

Technical Specification 3.6.4, page 3-41, Amendment 240.

Technical Specification 3.17, page 3-80, Amendment 157.

V.B.01.01 M New 0 TMlBank TMI Question #

0 Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge Comprehension or Analysis 55.43 .2 E 55.45 .I 1 A Incorrect answer. This would be reportable as violation of an LCO, but the basis stated is incorrect.

B Incorrect answer. This LCO is not applicable at Hot Shutdown, and the basis is incorrect.

C Correct answer.

D Incorrect answer. This LCO is not applicable at Hot Shutdown, and the basis for the limit is incorrect.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

3.6 REACTOR BUILDING d holicability Applies to the containment integrity of the reactor building as specified below.

Obiective To assure containment integrity.

SDecification 3.6.1 Except as provided in 3.6.6,3.6.8,and 3.6.12, CONTAINMENT INTEGRITY (Section 1.7) shall be maintainedwhenever all three of the following conditions exist:

a. Reactor coolant pressure is 300 p i g or greater.

b. Reactor coolant temperature is 200°F or greater.

c. Nuclear fuel is in the core.

3.6.2 Containment integrity shall be maintainedwhen both the reactor coolant system is open to the containment atmosphere and a shutdown margin exists that Is less than that for a refuelingshutdown.

... 3.6.3 Posifiii reactivity insertions which would result in a reduction in shutdown margin to less than 1% AWk shall not be made by control rod motion or boron dilution unless containment integrity is being maintained.

3.6.4 The reactor shall not be critical when the reactor building internal pressure exceeds 2.0 psig or I.O psi vacuum.

3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed and are conspicuously marked.

3.6.6 While the reactor is critical, if a reactor building isolation valve (other than a purge valve) is determined to be inoperable in a position other than the required position, the other reactor building isolation valve in the line shalt be verified to be OPERABLE. If the inoperable valve is not restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the OPERABLE vabe will be dosed or the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to the COLD SHUTDOWN condition within additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.7 DELETED I 3-4 1 Amendment No. 87,W,408,%,+!I8 240

I ,-

3 17 REACTOR BUSLXNG A I R TEMPERATURE Apipl icabil i t y

. / This specification applies t o the average a i r temperature o f t h e primary containment during power operations, Obj ect t ve TO assure that the temperatures assumed in the structural analysis o f the Reactor Building are not exceeded.

Specification 3.17.1 Primary containment average air temperature above Elev. 320 sha?l not exceed 13OoF and average a f r temperature below Elev. 320 shall not exceed 1 2 O O F .

3.17.2 I f , whi la the reactor is critical, t h e above stated temperature limtts are exceeded, the average temperature shall be reduced to the above limits wlthin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be i n a t l e a s t HOT STANDBY w i t h i n the next s i x (6) hours and in COLD SHUTDOWN wi thf n the fol 1owing thirty { 30) hours.

3.17 - 3 The primary containment average air temperature shall be caf cul ated as fo11 ows:

a) The average temperature above elevation 320 will be

.calculated by taking the arithmetic average o f the temperatures from at least 13 locations above

."-' elevation 320. A list o f locattons is given below, b) The, average temperatures below elevation 320 will be calculated by taking the arithmetic average of the temperatures from at least 4 locations below Elev. 320 A list o f locations I s given below.

Location Location SE Wall Elev. 352' NE Wall Elev 314'*

NW Sec Shield s v 352' s Wall l e v 3W*

NE Sec Shield Elev 352' NIJ Wall E l e v T 4 '

E Wall lev 382' N Set S h j e l d l e v 352' E See Shield mv 352' S Rx Wall Elev 321' .

NW Sec Shield Elev 352' NE Wall lev 287'"

NE Sac Shield Elev 352' S Wall lev 2W*

NW Sec Shield Elev 352' NW Wall Elev287'*

NW Wall lev 352' E Sec Shield mv 352' E Wall Elev 4007 NW Sec Shield Elev-7'*

S Sec S h i e l d m e v 352' NE Sec S h i e l d Elev 364' NW Sec Shield E1ev-Z' N Sec ShJeld Elev NOTE: (1)

Detectors located below elev 326'.

3-80 j -__

I Amendment No. 41, 78, 157

SYSIEW Gen KA# 2.1.34 Tier # -

3

.L' ROlSRO Importance Rating 2.3 2.9 Group #

Ability to maintain primary and secondary plant chemistry within allowable limits. (Conduct of OPS)

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Main Condenser tube leak in progress in 'B side of Main Condenser.

- Chemistry reports validated readings:

- 6.5 umholcm at CE-6A, corrected feedwater cation conductivity.

- 7.5 umholcm at CE-6, feedwater cation conductivity.

Based on these conditions, identify the ONE statement below that describes required operator actions.

A. Trip the reactor and go to OP-TM-EOP-001, Reactor Trip.

B. Secure all Moisture Separator Drain Pumps and continue power operation.

C. Reduce power to less that 50% and isolate the ' B side Circulating Water loop.

D. Perform a normal plant shutdown and cooldown to Decay Heat Removal operation.

AbP 1203-5, High Contaminants in the Condensate andlor Feedwater System, Immediate Manual Action #2, page 2, Rev. 25.

None.

.02.03

\- 7New 3 TMI Bank TMI Question # #77 Waple Re-exam 2000 Modified TMI Bank Parent Question #

a Memory or Fundamental Knowledge 0 Comprehension or Analysis Lq 55.41 . I O 0 55.43 3 55.45 .I2 A Correct answer.

B Plausible because this would stop recirculation of the contaminant.

C Plausible because this is the action required for lower contamination levels.

D Plausible if the examinee does not realize the immediacy of the problem.

rearranged distractors, updated procedure reference in distractor, augmented stem TMI SRO Exam - May 2003 Friday, March 28,2003

Number Title Revision No.

High Contaminants in the Condensate and/or Feedwater System 25 1.0 SYMPTOMS

1. Increasing conductivity on Control Room conductivity recorder.

2. Increasing conductivity on secondary sampling recorders.

3. Increasing sodium on the sodium monitor at the condensate pump discharge

3. Alarm PLB-8-6, Conductivity Recorder Abnormal

5. Alarm PRF 6-1, ICE-6A Conductivity Hi.

6. Alarm PRF 2-6, CO-C1-A Conductivity Trouble.

7. Alarm PRF 2-7, CO-C1-B Cndtvty Trouble.

8. Alarm PRF 2-8, CO-C2NB Cndtvty Trouble.

9. Alarm PLB-5-7, Turbine Sampling Room Trouble, caused by increasing sodium in the OTSG Feedwater and in the Condensate System. (OTSG Feedwater over 3 ppb and increasing.)

IO. Alarm on the following PPC (Plant Process Computer) points:

1. A1031 - output from instrument CE-773 (normally monitors CE-3, Condensate Pump Discharge)

2. A1032 - output from instrument CE-772 (normally monitors CE-2, Condensate Pump Outlet and also monitors Ecolochem Makeup Water)

3. A1033 - output from instrument CE-801 (normally monitors CE-6, OTSG Feedwater) 2.0 IMMEDIATE ACTION A. Automatic - None B. Manual

1. CONTACT Chemistry Dept. to confirm abnormal conductivity with the cation conductivity recorders in the secondary sample room.

2. IFCE-6A (corrected feedwater cation conductivity) is confirmed > 5.0 pmhokm or CE-6 (feedwater cation conductivity) is confirmed > 6.Opmho/cm, THEN immediately TRIP the reactor AND GO TO ATP 1210-1, Reactor Trip.

2

Q # 085 SYS/EP# @r~ KA# 2.1.5 Page # 2-1 Tier # -3 U' ROlSRO Importance Rating 2.3 -3.4 Group # -1 Ability to locate and use procedures and directives related to shift staffing and activities.

(Conduct of Ops)

Identify the ONE selection below that describes a condition when it is permissible for a Reactor Operator to be the ONLY NRC LICENSED person in the Control Room.

A. Reactor power is 100%;

There is one other person, a CRO trainee (NOT LICENSED) in the Control Room.

B. Plant is in Hot Shutdown condition; The Shift Technical Advisor (NOT LICENSED) is also in the Control Room.

C. RCS temperature is 210°F; The duty Shift Manager (LICENSED) is in the Shift Manager's Office.

D. RCS temperature is 189°F; The duty Shift Manager (LICENSED) is in the Operations Office Building.

Technical Specification 6.2.2.2.d, page 6-1, Amendment 219.

Technical Specification 6.2.2.2.a, page 6-1, Amendment 219.

Technical Specification Table 6.2-1, page 6-2, Amendment 219.

Technical Specification6.2.2.2.d, page 6-1, Amendment 219.

Technical Specification6.2.2.2.a, page 6-1, Amendment 219.

Technical SpecificationTable 6.2-1, page 6-2, Amendment 219.

V.A. 10.08 0 New &TMl !I Bank TMI Question # June2001 SRO Audit #84 0 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge 3 Comprehension or Analysis 55.41 .10 55.43 .I 3 55.45 . I 2 A Incorrect answer. Plausible misconception that requirement is for other licensed personnel to be ON SITE rather than in the Control Room.

B Incorrect answer. Plausible misconception since reactor is shutdown, and STA is in the Control Room.

C Incorrect answer. Plant is above 200"F, and licensed SRO is required to be in the Control Room.

D Correct answer. Plant is <200"F.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

6.0 ADMINISTRATIVE CONTROLS 4.1 RESPONSr3lLFrY 6.1.1 The V i e President-TMT Unit 1 SI@be responsible for TMI-I operations and m y , at any time, I delegate his responsibilitiesin writing to the Plant Manager. He shall delegate the succession of his xeqmnsibilities in wriiing duringhis absence.

6.1.2 The shift Manager (or during his absence from the Control Room,a designated individual), shall I be responsible for the Control Room command fhcxjon. A management directive IO this effect signed by the Chief Nuclear Officer shall be reissued to all unit personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 COrnRATE 6.2.1.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onshe and offsite organkition shall include the positions for activities afliecting the &e? of the nuclear power plant.

6.2.1.2 Lines of authority, responsibility and c a d c a t i o n shall be established and defined from the highest manageinent levels through intermediate levels lo and including operating organization positions. Tttese relationshipsshall be documented and updated as appropriate, in the form of organizational charis. These organizational charts will be documented in rhe Updated FSAR and updated in accordance with 10 CFR 50.71e.

6.2.1.3 The Chief Nuclear OEcer shall have corporateresponsibiIityfor overall plant nuclear saf'cty and shall take measqes to ensure acceptable performance of the staff in operating, maintaining, and providing technical support so lhat continued nuclear safety is assured.

6.2.2 UNIT STAFF 6.2.2.1 The Vice President-TMI Unit 1 be responsibIe for overall site safe operation and shall have control over those on site activities neas'sary for safe operation and maintenance of the site.

6.2.2.2 The unit staff organization shall meet the following:

a. Each onduty shift shall be composed of at least Ihe minimum ShiA crew composition shown in Table 6.2-1.

b. At least one licensed Reactor Operator 5balI be present in the control room when fie1 is in the reactor.

6-1 Amendment No. +a 219

c. At least two licensed Reactor Operators shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips

d. The Shft Manager or Control Room Supenisor f shall be in the control room at all times I I

other than cold shutdown conditions (T c 200OF) when he shall be onsite.

e. An individual i% qualified pursuant to 6.3.2 in radiation protection procedures shall be on site when fuel is in the reactor.

f. All REFUELING OPERATIONS shaa[lbe observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handing who has DO other concurrent responsibilities during this operatiori.

g. A Sire Fire Brigade ## of at least 5 members shall be maintained onsite at all times. The Site Fire 3rigade shall not include members of the minimumshift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

h. The Shift Technical Advisor shall serve in an advisory capachy to ?he Shifi Manager on matters pertaining to the ergineering aspects assuring safe operation ofthe unit. t 6.2.2.3 Individuals who train the operating staff and those who carry out the health physics and qualip assurance h c t i o n shall have sufficient organizational freedom to be independent from operating pressures, however they may report to ihe appropriate manager on site.

b

If not SRU.licensed, he shall have completed the SRO Training propam.

1. The individual of item 6.2.2.2e and the Fire Brigade composition may be less than the minimum requirements for a period o f h e not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodaie unexpected absence provided immedhte action is taken to fill the required positions.

6-la

TABLE 6.24 MTNIMWMSHIFT CREW COMPOSITION'^^^)

Non-Licensed Auxiliary 2 I Operator Shift Technical Advisor I(4 None Required (i) Does not include the Licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling,supervising (a) irradiated he1 handling and transfer activities onsite, and (b) all unimtdiatedhe1 handling and transfer activities to and from the Reactor Vessel.

(ii) May be on a different shifi rotation than licensed personnel.

(iii) Except for the Shift Manager, shifl crew composition may be one less than the minimurn requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accOmOdate unexpected absence of onduty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements ofTable 6.2-1. This provision does not permit any shift crew position to be unmanned upon shifi change due to an incoming shift crewman being late or absent.

(iv) Pursuant to the requirements of I O CFR 50.54(m).

6-2 hendment No. 1i, 32, ??, '49 219

L 4.4 Group #

ROISRO importance Rating 3.7 -

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (Conduct of Ops).

Plant conditions for the past 45 days:

- NORMAL OPERATIONS.

- Reactor power is loo%, with ICs in full automatic.

- STAR module has maintained Heat Balance Power at 2568 MW thermal.

- Control Rod Index has been maintained at 290-292.

- RCS boron concentration was reduced 75 ppm, ramped over the 45 days.

- Power Range NI readings have slowly risen from (initial) 100% to 101%.

Based on plant performance data above, identify the ONE selection below that'describes (1) impact and (2) credible reason for the NI-Heat Balance Calculation mismatch.

A. (1) Heat balance calculation is non-conservative.

ACTUAL reactor power = 101%.

FW temperature 'instrument error drift (indicating LOWER than actual) could be producing a heat balance calculation error to produce this mismatch.

Heat balance calculation is accurate.

ACTUAL reactor power is still 100%.

INDICATED reactor power is rising due to (outward) radial flux redistribution as fuel depletion continues.

Heat balance calculation is accurate.

ACTUAL reactor power is still 100%.

INDICATED reactor power is rising due to LOWER thermal neutron leakage resulting from the ramped reduction in RCS boron concentration.

Heat balance calculation is non-conservative.

ACTUAL reactor power = 101%.

Plant efficiency could be reduced due to rising FW heater tube leakage.

PM Section N-06, Reactor Theory, page 52, Rev. 6.

New 0 TMl Bank TMI Question #

0 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge

@IComprehension or Analysis 0 55.41 55.43 .6 id 55.45 .12/.13 A Incorrect answer. FW temperature indication drifting LOW would raise heat balance power, allowing the STAR module to lower ULD, FW flow, and reactor power. A MISMATCH would occur - ACTUAL and INDICATED power would both lower, and Heat Balance power would (erroneously) remain at 100%

u (conservative).

B Correct answer. Excore detectors normally drift high, requiring re-calibration due to outward radial flux shift as fuel in the center of the core is depleted.

TMI SRO Exam - May 2003 Friiiuy, March 28,2003

C Incorrect answer. First part is correct - actual reactor power would remain at loo%, and Heat Balance power would also be 100%. The second part incorrectly identifies the (75 ppm) boron concentration reduction as the reason for the Excores indicating (1%) higher than actual reactor power.

L D Incorrect answer. If FW heater leakage reduces plant efficiency, STAR will lower MW Demand to maintain 100% heat balance power, and the Nls will also lower by a corresponding amount. Result would produce NO MISMATCH between Heat Balance and Excores. Final MW would be lower, with Heat Balance maintained at 100% by ICs STAR module, and Excores showing reactor still at 100% power.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

SECTION N-06 REVISION 6

c. Core Age As the core ages, the fuel mix changes. The new fuel isotopes (Pu-239,240,24 1, and other minor contributors) have different fast fission characteristics than U-235 and 235. As the U-235 b u m out, the fast fission factor tends to decrease. But as the plutonium isotopes build in, the fast fission factor tends to increase. The net effect is small.

d. Control Rods and Lumped Burnable Poison Assemblies Inserting control rods or loading lumped burnable position (LBP) assemblies causes the average moderator density to decrease. The effect is the same as noted above for moderator density.

10.4.2 Factors Affectingthe Fast Non-Leakage Factor Mathematically the fast non-leakage probability can be expressed as:

1 L -

-f - 1+ T B ~

WHERE: -c is the Fermi Age and B2is the Geometric Buckling.

The geometric buckling is a patimeter dependent on the size of the core which indicates the curvature ofthe flux. lit an infinitely large core, B = O. A commercial power reactor core is nearly infinite, thus B2 is very small. A Navy propulsion core, being much smaller, has a neutron flux with a much tighter curvature, thus B2 is larger for a Navy core than for TMI- 1.

Besides core size, which is generally not contrdlable by the operator, the flux shape can change the buckling. For example, if the fuel loading is not uniform, the resulting flux will be Iumpy. A Iumpy flux has a larger buckling than a smooth flux.

a. Moderator Density A moderator density decrease will decrease the C, and C, of the moderator due to the drop in atom density. This will result in an increase in neutron slowing down length, Fermi Age, slowing down time, and absorption within the moderator - increasing fast neutron leakage, and decreasing &f -

b. Core Age r&.the core ages, theaxialflux shifts to the upper and lower regions of the core. ,The radial flux shifts outward (though not as dramatically as in the early cycles). This increases leakage, decreasing k .

c. Control Rods and Lumped Burnable Poison Assemblies Inserting control rods or loading LBPs has the same effect as decreasing the moderator density.

52

Group #

\'

ROISRO Importance Rating 2.7 39 Knowledge of conditions and limitations in the facility license. (Conduct of Ops)

From the list below, identify the ONE set of conditions that constitutes a VIOLATION of conditions and limitations in the facility license.

A. Operation at 50% power for 20 hour2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months s:

- Control rod positions inside the RESTRICTEDoperating region.

- Corrective actions initiated immediately when condition was discovered.

- Power peaking was verified within acceptable limits the entire time.

B. Operation at 44% power for 9 hour1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months s:

- 3 RCPs operating.

- 1 asymmetric rod cannot be aligned with its group, declared inoperable.

- Ejected rod worth and nuclear peaking factors were verified within limits.

- RPS nuclear overpower trip setpoints are set at 102%.

C. Operation at 100% power for 1 week:

- RCS net unidentified leakage = 0.02 gpm.

- OTSG tube leakage = 0.5 gpm.

- Baseline OTSG leakage = 0.01 gpm.

D. Operation at 100% power for Imonth:

- RCS net unidentified leakage = 0.05 gpm.

- OTSG tube leakage = 0.02 gpm.

- Baseline OTSG leakage = 0.01 gpm.

- MU Pump seal leakage = 1.O gpm.

Condition of License (8.2) page 6, Amendment 218.

TS 3.5.2.5.a.1.. Daae 3-35, Amendment 219.

TS 3.5.2.2.e, page 3-33, Amendment 21 1.

TS 3.1.6.1 ., page 3-12, Amendment 180.

TS 3.1.6.2 , page 3-12, Amendment 180.

Condition of License (8.2) page 6, Amendment 218.

TS 3.5.2.5.a.1, oaae 3-35, Amendment 219.

TS 3.5.2.2.e, page3-33, Amendment 21 1.

TS 3.1.6.1., page 3-12, Amendment 180.

TS 3.1.6.2 , page 3-12, Amendment 180.

@ New d TMlBank TMI Question # .

El Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge

&!I Comprehension or Analysis i3;l 55.43 .1 55.45 .I3 A Incorrect answer. This is not reportable since conditions satisfy TS LCO 3.5.2.5.a.1 (page 3-35), as long as

- acceptable rod positions are attained within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B Incorrect answer. This is not reportable since conditions satisfy TS LCO 3.5.2.2.e (page 3-33), as long as overpower trip setpoints are reduced to 70% of (75%) thermal power allowed (52.5%)

C Correct answer. This violates Condition of License (8.2) on Page 6 (Amendment 218) of Operating License TMI SRO Exam May 2003 Friday, March 28,2003

Docket No. 50-289, which requires the plant to be shut down, etc., if OTSG leakage exceed baseline leakage by >0.1 gpm.

D Incorrect answer. This is not reportable since conditions satisfy total leakage TS 3.1.6.1 limit ( I O gpm), TS u 3.1.6.2 limit (unidentified leakage is 4 gpm), and identified leakage has been evaluated as not unsafe (TS 3.1.6.2).

None.

TMI SRO Exam May 2003 Friday, March 28, 2003

Repaired Steam Generators In order to confirm the ieak-tight integrity of the Reactor Coolant System, including the steam generators, operation of the facility shall be in accordance with the following:

I.Prior to initial criticality, the licensee shall submit to NRC the results of the steam generator hot test program and a summary of its management review.

2. The licensee shall confirm baseline primary-to-secondary leakage rate established during the steam generator hot test program. If leakage exceeds the baseline leakage rate by more than 0.1 gprn', the facility shall be shut down and leak tested. If any increased leakage above baseline is due to defects in the tube free span, the leaking tube@)shall be removed from service. The baseline leakage shall be re-established, provided that the leakage limit of Technical Specification 3.1.6.3is not exceeded.

3. The licensee shall complete its post-criticaltest program at each power range (0-5%,5%-50%, 50°/6-100%)in conformance with the program described in Topical.Report 008, Rev. 3, and shall have available the results of that test program and a summary of its management review, prior to ascension from each power range and prior to normal power operation.

4. The licensee shall conduct eddy-current examinations, consistent with the extended inservice inspection plan defined in Table 3.3-1 of NUREG-1019, either 90 calendar days after reaching full power, or 120 calendar days after exceeding 50% power operation, whichever comes first. In the event of plant operation far an extended period at less than 50% power, the licensee shall provide an assessment at the end of 180 days of operation at power levels between 5% and 50%, such assessment to contain recommendations and supporting informationas to the necessity of a special eddy-current testing (ECT)shutdown before the end of the refueling cycle. (The NRC staff will evaluate that assessment and determine the time of the next eddy-current examination, consistent with the other provisions of the license conditions.) In the absence of such an assessment, a special ECT shutdown shall take place before an additional 30 days of operation at power above 5%.

'If jeakage exceeds the baseline leakage rate by more than 0.1 gpm during the remainder of the Cycle 8 operation, the faality shall be shutdown and leak tested. Operation at leakage rates of up to 0.2 gpm above the baseline leakage rate shall be acceptable during the remainder of Cycle 8 operation. After the 9R refueling outage,the leakage limit and accompanying shutdown requirements revert to 0.1 gpm above the baseline leakage rate.

I Amendment No. :C2, qG3 Amendment Nm-207 218

3.1.6 -,

ADDlicability Applies to reactor coolant leakage from the reactor coolant system and the makeup and purification system.

Objective, To assure that any reactor coolant leakage does not compromise the safe operation o f the facil ity.

SDeci f f cat i on 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be placed i n hot shutdown.within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

o\r?, 3.1.6.2 If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds one gpm o r if any reactor coolant leakage i s evaluated as unsafe, the reactor shall be placed I n hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection, 3.1.6.3 If primary-to-secondary 1 eakage through the steam generator tubes exceeds 1 gpm total for both steam generators, the reactor shall be placed in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months of detection.

3.1.6.4 If any reactor coolant leakage exists through a nonisolab3e fault in an RCS strength boundary (such as the reactor vessel, piping, valve body, etc., except the steam generator tubes), the reactor s h a l l be shutdown, and cool-dawn to the cold shutdown condition shall be

\-e initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

3.1.6-5 If reactor shutdown i s required by Specification 3.1.6.1, 3.1.6.2, 3.1.6.3, or 3.1.6.4, the rate of shutdown and the conditions o f shutdown shall be determined by the safety evaluation f o r each case.

3.1.6-6 Action to evaluate the safety implication o f reactor coolant leakage shall be initiated within four hours o f detection. The nature, as well as the magnitude, of the leak shall be considered in this evatuatfon. The safety evaluatlon shalt assure that the exposure o f 3.1.6.7 offsite personnel t o radiation is within the limits of Specification 3.22.2.1.

If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, I

3.1.6.3 or 3.1.6.4, the reactor shall not be restarted until the leak i s repaired or until the problem is otherwise corrected.

3.1.6.8 When the reactor is critical and above 2 percent power, two reactor coolant leak detection systems o f different operattng principles shall be in operation for the Reactor Building with one o f the two systems sensitive to radioactivity. The systems sensitive to radioactivtty may be aut-of-service for no more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided a sample is taken o f the Reactor Building atmosphere every eight hours and analyzed for radioactivity and two other means are available to detect leakage.

3-12 Amendment No. $7, f i d y 180 (12-22-78)

3.5.2 CONTROL ROD GROUP AND POWER DISTRIBUTION LIMITS mbility This specification applies to power distribution and operation of control rods during power operation.

Obiective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcritidity after a rcactor trip.

Ssxcification 3.5.2.1 The available shutdown margin shall not be less than o m percent aWK with the highest worth control rod filly withdrawn.

3.5.2.2 Operation with inoperable rods:

a. Opcration with more than one inopcrable rod as dcfincd in Specification 4.7.1 and 4.7.2.3 in the safety or.regulating rod banks shall not be permittcd. Vcrify SDM 2 1%

aklk or initiate booration to restore within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The reactor shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. If a control rod in the regulating andlor safety rod banks is declarcd.inopcrable in thc withdrawn position as ddincd in Specification Paxagaph 4.7.1.1 and 4.7.1.3, an evaluation shaiI be initiatd immediately ro veri@ the existencc of one percent AI& hot shutdown margin. Boration may be initiated to incrcase the available rod worth either to compensate for the worth ofthe inoperable rod or until thc regufating banks arc fully withdrawn, whichever occurs first. SimuItaneously a program of exercising the remaining regulating and safety rods shalI be initiated to verify operability.

c. If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, and Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, it is not determined that a one percent & hot shutdown w i n exists combiningthe worth ofthe inoperable rod with each of the other rods, the reactor shall be brought to the HOT SHUTDOWN condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months until this marsin is established.

d. F o h v i n g the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and exercised weekly until the rod problem is solved.

e. If a control rod in the regulatingor safety rod groups is declared inoperable per 4.7.1.2,

'and cannot be aligned per 3.5.2.2.c power shall be reduced to 3 60% ofthe thermal power aIlowablc for the rcactor coolant pump combination within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the overpower trip setpoint shall be reduced to 5 70% ofthe thermal power allowable within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Verify the potential ejected rod worth (ERW) is within the assumptions ofthe ERW analysis and Verify peaking factor (FP(2) and ):F limits per the COLR have not been excccded within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3-33 Amendment No. W, 21 1 (5-18-76)

- Q -Ids'7 3 3.5.2.5 Control Rod Positions

a. Operating rod group overlap shall not exceed 25 percent 55 percent, between two s q u e n r i d groups except for physics tests.

b. Position limits are specified for regulating control rods. Excepr for physics tests or exercising control rods, the regulating control rod insertiodwithdrawal limits are specified in the COW OPERATMG LIMITS REPORT.

1. If regulating rods are inserted in the restricted operating region, corrective pr\iS"'efL measures shall be taken immediately to achieve an acceptable control rod

,%fi IS

. position. Acceptable control rod positions shall be attained within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and FQ(Z) and FL shall be verified within iimits once every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or power shall be reduced to 3 power allowed by insertion limits.

2. If regulating rods are .inserted in the unacceptable operating region, initiate boration within 15 minutes to restore SDM to 21% AK/K, and restore regulating rods to within restricted region within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce power to 5 power allowed by rod insertion limits.

c. Safety rod limits are given in 3.1.3.5.

3.5.2.6 The control rod drive parch panels shall be locked at all times with limited access to be authorized by the Plant Manager.

3.5.2.7 AxiaI Power Imbalance:

a. Except for physics tests the axial power imbalance, as determined using the full incore system (FIS), shall not exceed the envelope defined in the CORE OPERATMG LIMITS FLEPOR'T.

The FIS is operable for monitoring axial power imbalancc provided the number of valid self powered neutron detector (SPND) signals in any.one quadrant is not less than the limit in the CORE OPERATING LIMITS REPORT.

b. When the full incorc detector system is not OPERABLE and except for physics tests axid power imbalance, as determined using the power range channels (out of core detector system)(OCD), shall not exceed the envelope defined in the CORE OPERATMG LIMITS REPORT.

c. When neither detector system above is OPERABLE and, except for physics tests axial power imbalance, as determined using the minimum incore system (MIS),shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.

d. Except for physics tests if axial power imbalance exceeds the envelope, corrective measures (reduction of imbalanceby APSR movements and/or reduction in reactor power) shall bc taken to maintain operation within the envelope. Verify SQ(2) and N

F~ are within limits of the COLR once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when not within imbalance limits.

3-35 I

Amendment No. 4 4 I + $ 9 + 3 & 3 9 , ~ M - , 2 I9

SYSIEP# Gen KA# 2.2.26 Page # 2-7 Tier # -

3 u 3.7 Group #

ROlSRO Importance Rating 2.5 -

Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (Equipment Control)

Identify the ONE statement below that describes the basis for maintaining Control Rods in the PERMISSIBLE Region of the COLR rod insertion limits.

A. Limit ejected rod worth.

B. Ensure shutdown margin capability.

C. Minimize fuel pellet swelling during power operations.

D. Limit peak fuel clad temperatures during the initial phase of a LOCA.

Technical Specification page 3-35a, Amendment 21 1.

Technical Specification page 3-35a, Amendment 21 I.

New c]TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

3 Memory or Fundamental Knowledge C Comprehensionor Analysis u 0 55.41 55.43 .2 @ 55.45 . I 3 A Incorrect answer. Plausible distracter since this is a function of the Not Allowed Region rod insertion limit.

B Incorrect answer. Plausible distracter since this is a function of C Incorrect answer. Plausible distracter since fuel pellet swelling does occur during power operations.

D Correct answer. Refer to TS page 3-35a, Amendment 211.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

If an acceptable axial power imbalance is not achieved wirhin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , reactor power e.

shall be reduced to 340% FP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . I

f. Axial power imbalance shall be monitored on a minimum Frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when axial power imbalance alarm is OPERABLE, and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when imbalancealarm is inoperable during power operation above 40 pcrcent Q rated power 3.5.2.8 A power map shall be taken at intervals not to exceed 3 1 efFectivc fuII power days using the incore instrumentation detection system to veri* the power distribution is within the limits shown in the CORE OPERATING LIMITS REPORT.

Bases Thc axial power imbalance, qua I rod position limits arc based on LOCk

, analyses which These limits are developed in a manner that cnsures rate wiii not cause the masimum clad

.tcmpcrature to exceed 10 CFR50 Appendix-E*Operation outside of any one limit alone does not neccssarily constitute a situation that would cause the Appendix K Criteria to be exceeded should a LOCA occur. Each limit represents the boundary of operation that will preserve the Acceptance Critcria even if all three limits are at their maximum allowable values simultancously. The effects of thc APSRs are included in the limit development. Additional conservatism included in the limit devclopment is introduced by application of:

a. Nuclear uncertainty factors

b. Thermal calibration uncertainty

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors

e. Postulated fuel rod bow effects

f. Peaking limits based on initial condition for Loss of Coolant Flow transients.

Thc incore instrumenfation system uncertainties used to develop thc axial power imbdancc and quadrant tilt limits accounted for various combinations of invalid Sclf Powcrcd Neutron Dctcctor (SPND) sipals. If the number of valid SPND signals falls below that used i.n fie uncertainty analysis, thcn anothcr system &a11 be used for monitoring axial powcr imbaiancc andor quadrant tilt.

For asjal power imbalance and quadrant power tilt measurements using thc incorc detector system, the minimum incon detector system consists of OPERA3LE detectors configured as foIIows:

Axial Power Imbalance

a. Three detectors in each of thee strings shall lie in the same axial plane with one p h c in each axial core half.

b. The axial planes in each core half shall be symmetrical about the cote mid-planes.

c. The detectors shall not have radial symmetry.

Quadrant Power Tilt

a. Two scts of four detectors shall lic in sach core half. ach set of fou'r shall lie in the same axial plane. The two sets in the same core half may lie in the Same axiaf plane.

b. Detectors in the m e plane shall have quarter mrc radial syrnmctry.

3-3Sa Amendment No. -I?,??, w 11'2, , W , 2 1I

I ,

ROlSRO Importance Rating 2.1 3.1

- Group #

Knowledge of maintenance work order requirements. (Equipment Control) 0 m Plant conditions:

- A maintenance work order requires a pneumatic operated valve to be closed as part of the clearance order safety boundary.

- The valve fails OPEN on loss of air or loss of power to the solenoids controlling the air.

Based on these conditions, identify the ONE selection below that completes the following phrase:

This valve may be used as part of the safety boundary if ...

A. the fluid controlled by the valve is less than 200°F and less than 500 psig.

B. the power supply to the solenoids is tagged to ensure a continuous power supply.

C. a temporary air bottle is installed to ensure a continuous air supply.

D. an appropriate gag is used on the valve operator.

OP-MA-109-101, Clearance and Tagging, page 22, Rev. 1.

None.

C New TMIBank TMI Question # XQRSAOI -0, Q02 El Modified TMI Bank Parent Question #

4 Memory or Fundamental Knowledge 0 Comprehension or Analysis

@I 55.43 .5 @I 55.45 .I3 A Incorrect answer. This is a mis-application of double valve isolation criteria described in OP-MA-109-101.

B Incorrect answer. This does not ensure a fail-safe mode for ensuring the integrity of the pressure boundary.

C Incorrect answer. This approach is not allowed by procedure OP-MA-109-101.

D Correct answer. In accordance with 1002.1 section 4.8.5, application of an appropriate gag is satisfactory if another valve can not be used for the pressure boundary to ensure personnel and plant safety.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

OP-MA-IO~~OI Revision I Page 22 of 126 I

2. Afail open POV may be used as a closed isolation point if the following conditions are met: A' A. The valve control station(s) shall be tagged closed.

B. The valve shall be tagged and forcibly held closed with an instalfed gag or blocking device.

C. The integrity of the gag or blocking device installation shall be verified by:

I. Close and vent the air supply to ensure POV remains as positioned.

2. Restore the air supply.

D. The air supply valve to a gagged POV shall be information tagged to aid in establishing the component in a safe condition prior to gag removal.

3. A fail close POV may be used as an isolation point if all the following conditions are met:

A. The valve control station(s) shall be tagged.

8. The valve should be visually checked closed.

C. The air supply shall be tagged closed or the power supply to the POV solenoid shall be tagged in the de-energized state.

D. The air supply should be vented.

1. If the POV solenoid valve does not vent the air path when isolated, then the air supply should be vented between the air isolation valve and the valve operator. It should be tagged as a clearance point (information tagged with vent path open or equivalent).

2. If a vent path is not available, then breaking a flange connection, fitting, gauge line, or opening a pressure regulator petcock valve, etc. is acceptable to accommodate a vent path.

All appropriate FME prevention steps shall be taken.

3. If a vent path cannot be established, then the clearance shall be designated as an Exceptional Clearance.

4. A fail-as-is POV may be used as an isolation point if all of the following conditions are met:

A. The valve control station(s) shall be tagged.

Form ES-401-6 Page # 2-6 Tier #

b 3.4* Group #

ROISRO Importance Rating 2.5 -

Knowledge of the process for controlling temporary changes. (Equipment Control) 3 E Identify the ONE operation below that is subject to the requirements of AP 1013, Temporary Modifications and Bypass of Safety Functions.

A. Using a power buggy to energize a portable sump pump in the Amertap pit.

B. Installationof a pipe cap on the outlet of a drain valve to stop leakage to the floor at 'A 12th stage heater.

C. Bolting a stainless steel collar around a valve stem to prevent vibration induced damage.

D. Installation of a calibrated test gauge to support performance of a Tech Spec surveillance on a temporary basis.

1013, Temporary Modifications and Bypass of Safety Functions, pages 1-4, Rev 51.

3 New E TMI Bank TMI Question # #5 SR021 AUDIT E Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge Comprehension or Analysis

-' 55.41 .IO 0 55.43 .3 P 55.45 .I3 A Incorrect answer. This is a temporary installationthat is excluded from 1013 control in section 2.2.3.4, page 4.

B Incorrect answer. This is a temporary installation that is excluded from 1013 control in section 2.2.3.6, page 4.

C Correct answer. This is a temporary installation that is NOT excluded from control of 1013.

D Incorrect answer. Excluded in section 2.2.3.2, page 3.

TMI SRO Exam May 2003 Friday, March 28,2003

Number TMI - Unit 1 W Ad minist rat ive Procedure 1013 Title Revision No Temporary Modifications and Bypass of Safety Functions 51 The purpose of this procedure is to establish a method for:

1.I Controlling and documenting temporary modifications on operating systems that have a safety or operational impact.

1.2 For documenting lifted leads and jumpers on systems which are tagged out-of-service.

1.3 Use of bypass features on safety related equipment.

2.0 APPLlCABlLlTYlSCOPE 2.1 Applies to the use of the design bypass features listed in Enclosure 1 NOTE Enclosure 3 is a Flowchart to help a user determine if AP 1013 requirements apply. Enclosure 3 is a guide. The requirements are all contained in the body of the procedure.

2.2 Applies to temporary modifications to TMI plant system. A "temporary modification" (Refer to 3.3)

L IS A minor change in the facility configuration.

e Only intended as an interim measure.

Required for testing, as an interim corrective action or to support plant operation.

IS NOT A means for permanent plant configuration change.

e Tools, fixtures, instruments or devices that are commonly used to facilitate or assist a component's operation without altering the functional performance or configuration of the system, structure or component 2.2.1 Applies to the installation of temporary ventilation equipment as defined in Section 3.5.

2.2.2 Applies to the installation of temporary test equipment as defined in Section 3.4.

2.2.3 -The requirements of this procedure DO NOT apply to:

Systems equipment outside of Configuration Management scope per AP 1029.8.1 Attachment 3.

L..,

N equipment used to monitor or record the performance of a piece of equipment on a temporary basis i.e. typically not left in place over shift turnover.

3

I I Number 'z 8" TMI - Unit 1

\/ Administrative Procedure 1013 Title Revision No.

Temporary Modifications and Bypass of Safety Functions 51 NOTE Temporary monitoring equipment such as test gauges can be substituted for the Tech. Spec. required gauges during the performance of a Tech.

Spec surveillance provided the temporary equipment is currently calibrated, meets the range requirements and accuracy of the gauge being replaced and documentation (i.e., calibration data sheets) is attached to the Tech. Spec. surveillance package. Temporary gauges must be removed upon completion of the Tech. Spec. surveillance.

2.2.3.3 Disabled annunciator alarms under the control of AP 1036, Instrument Out-Of-Service Control or Disabled alarm inputs under the control of OP 1105-21 Main Annunciator

~

Panel Beta Control System or Plant computer alarm changes under the control of OP 1105-10, Plant Computer Operations or DTCS alarm changes under the control of 1106-1, Turbine Generator lize existing e

'd power buggies with uiring special ns (Le., drop lights, recorders, hand held tools or sump pumps).

2.2.3.5 "Tygon" or "layflat" used to route fluid or gas. This exclusion is limited to uses where the tubing will not and cannot be pressurized. Also, the attachment to the discharge of a relief valve (pressure relief device) is prohibited.

2.2.3.6 Pipe caps or blank flanges installed to stop vent or drain valve leakage, provided that the cap/flange is visible from the vent or drain valve 224 -

Partial Exclusions Limited requirements of this procedure apply to:

2.2.4.1 Equipment configuration changes to systems in service when the change is performed in accordance with an approved procedure which includes the following provisions:

a. The procedure requires Shift Management notification for installation and removal.

b. The procedure requires and documents independent verification for installation and removal for NSR or RR systems and single person (as a minimum) documentation on other systems.

C. The procedure specifies that control of the "change" be transferred to AP I 0 13 if the procedure is complete and the system has not been returned to its normal configuration.

4

SYSIEP# Gen KA# 2.2.27 Page # 2-7 Tier #

b 3.5 Group #

ROlSRO Importance Rating 2.6 -

Knowledge of the refueling process. (Equipment Control)

During transfer of an irradiated fuel assembly with the Main Fuel Handling Bridge, the bridge stops travel, and will NOT move south to complete the operation. Investigation reveals that the south wall interlock has failed in the actuated position.

Based on these conditions, identify the ONE person below required to approve bypass of this interlock to continue the transfer.

A. Core Load Engineer.

B. Duty Shift Manager.

C. Fuel Handling Supervisor.

D. Refueling Outage Manager.

RP 1507-3, Main Fuel Handling Bridge Operating Instructions, Section 5.1.Ipage, 4, Rev. 22.

New d TMlBank TMI Question #

- Modified TMI Bank

.. Parent Question #

L

& Memory or Fundamental Knowledge

__ Comprehension or Analysis 7 55.41 r/J 55.43 .6, .7 $ 55.45 .I3 A Incorrect answer. Approval of (active) SRO licensed person in charge of Fuel Handling is required for this function.

B Incorrect answer. Approval of (active) SRO licensed person in charge of Fuel Handling is required for this function.

C Correct answer.

D Incorrect answer. Approval of (active) SRO licensed person in charge of Fuel Handling is required for this function.

TMI SRO Exam - May 2003 Friday, March 28,2003

I Number U'

I TMI - Unit 1 Refueling Procedure I 1507-3 Title Revision No.

Main Fuel Handling Bridge Operating Instructions 22 The purpose of this document is to provide detailed operating instructionsfor the main fuel handling bridge for all fuel handling operations and training evolutions.

2.0 DESCRIPTION

2.1 The main fuel handling bridge, located in the reactor containment building, lifts and transfers the following:

Fuel assemblies between reactor core locations and transfer system upenders Fuel assemblies between core locations (fuel shuffle)

Control components between fuel assemblies 2.2 The equipment consists of a bridge which runs on rails traversing the full length of the transfer canal.

The bridge has a trolley which runs transverse to the main axis of the bridge, thereby providing full x-y coverage of the transfer canal and core. Two types of hoist systems are provided on the trolley, each consisting of a gear/motor driven cable hoist lifting a grapple inside a fixed mast. One hoist system handles fuel assemblies and the other handles control components.

3.0 REFERENCES

,.d 3.1 1303-11.4, Refueling System Interlocks 3.2 1507-7, Fuel Transfer System Operation 3.3 VM-TM-0976, Main Fuel Handling Bridge 3.4 1013, Bypass of Safety Functions and Jumper Control 3.5 VM-TM-2395, Frequency Drive - FH Bridge, Toshiba Corp.

3.6 FH-A-1 Reactor Building Main Fuel Handling Bridge Weight Sensing System Calibration 3.7 M-83A, Main Fuel Bridge Hoist Brake Maintenance 3.8 1505-3, Fuel Handling Problems 4.0 TOOLS, EQUIPMENT AND SUPPLIES None 5.0 LIMITS AND PRECAUTIONS 5.1 Equipment 5.1. I Interlocks shall only be bypassed as specifically approved by the Fuel Handling Supervisor. (SRO in charge of refueling operations.)

4

L ROISRO Importance Rating 3.7 3.6

- Group #

Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (Equipment Control)

Plant conditions:

- Plant startup in progress.

- Reactor power is stable at 1E-8 Amps.

Critical data shows that actual criticality has occurred outside (above) the estimated critical position (ECP) band.

From the list below, identify the ONE statement that describes required action(s) for this condition.

A. Manually trip the reactor.

B. Go to 1203-10, Unanticipated Criticality.

C. Maintain criticality at current rod position, and initiate an assessment.

D. Insert the control rods to achieve at least 1% Delta WK subcritical condition, and initiate an assessment.

OP 1103-8, Approach to Criticality, Section 2.1.7, page 4, Rev. 47 E New 3 TMIBank TMI Question #

K l Modified TMI Bank Parent Question #

v

&!IMemory or Fundamental Knowledge 0 Comprehension or Analysis Ci 55.41 i? 55.43 @ 55.45 .1 A Incorrect answer. Not required by procedure, but plausible since it is very conservative action.

B Incorrect answer. Not required by procedure since the event occurred duyring a planned approach to criticality.

C Incorrect answer. This action would not address the requirements of OP 1103-8, Approach to Criticality, section 2. I.8.

D Correct answer. Refer to OP 1103-8, Approach to Criticality, section 2.1.8.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI Unit 1 L,' Operating Procedure I103-8 Title Revision No.

Approach to Criticality 47 (2.1.7 I' The estimated critical position (ECP) calculated in accordance with 11 03-156 I (Reference 1.5)specifies the rod position tolerance band. If criticality occurs outside the

, specified band:

0 An assessment shall be performed.

0 Regulating rods shall be inserted to maintain at least a 1% Ak/k subcritical condition during the assessment.

0. Entry into this procedure (1103-8)is a "planned evolution". Entry into Y- 1203-10,UnanticipatedCriticality (Ref. 1 .IO)is not required unless criticality outside the ECP tolerance band is uncontrollable.

2.1.8 The Nuclear Instrumentation shall be continuously monitored during any reactivity addition.

2.1.9 Reactor power shall not exceed 1% of rated power unless normal conditions are established for:

0 Operating temperature 0 Operating pressure 0 Control rod configurations 2.1.IO Safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin I by deboration or regulating rod withdrawal during the approach to criticality with the following exceptions (T.S. 3.1.3.5):

0 Inoperable rod 8 Physics testing 0 Shutdown margin may not be reduced below 1 Oh Ak/k 0 Exercising control rods 2.1.11 The available shutdown margin shall not be less than 1% Aklk with the highest worth control rod fully withdrawn (T.S. 3.5.2.1).

I 2.1.I2 This procedure shall not be used for the initial criticality following refueling (when 1550-01 I and 1550-02apply).

4

Form ES-401-6 SYSIEP# @t~ KA# 2.3.8 Page # 2-9 Tier # -3

\/ ROISRO Importance Rating 2.3 3.2

- Group #

Knowledge of the process for performing a planned gaseous radioactive release. (Radiation Control)

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Fuel pin failure has occurred.

- RCS degassification has resulted in higher than normal activity in the Waste Gas System vent header.

- Waste Gas Decay Tank WDG-T-1A has been isolated in support sampling in preparation for releasing its contents to the environment.

Based on these conditions, identify the ONE statement below that describes a procedural method used to limit the amount of radioactivity discharged when this tank is released to the environment.

A. Shorten the WDG-T-1A holdup time prior to release.

8. Introduce nitrogen dilution flow into the release effluent piping.

C. Filter the release through the normal roughing, HEPA and charcoal filters.

D. Verify both WDG-T-1B and WDG-T-I C are LESS than 65 psig prior to release.

1104-27, Waste Disposal - Gaseous, Prequisites for Waste Gas Decay Disposal, pages 21-24, Rev. 68.

w 0 New n TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

Memory or Fundamental Knowledge 0 Comprehension or Analysis 55.41 m 55.43 .4 2 55.45 .IO A Incorrect answer. This action would increase the curies released.

B Incorrect answer. Although N2 is used to purge tanks and vent header, no dilution of effluent with N2 is proceduraked.

C Correct answer. Normal release flowpath is through the AuxlFHB ventilation system, including filters. This is a prerequisite for the actual release by the Auxiliary Operator, per OP 1104-27, section 3.7.1 .J.

D Incorrect answer. 6610-ADM4250.11, Prerequisite 4.1.C requires another tank to be greater than 65 psig, based on ALARA considerations, prior to release of a tank. This maximizes hold-up time for additional radioactive decay prior to release.

None.

TMI SRO Exam May 2003 Friduy, March 28,2003

I Number

\/,-

TMI Unit 1 Operating Procedure I104-27 Title Revision No.

Waste Disposal - Gaseous 68

3. Close gas recycle valve WDG-V-25 (27 or 29), using the pushbuttons on the L.W.D. panel in the Auxiliary Building.

4. Vent the remaining 10 psig of gas in Waste Gas Decay Tank WDG-T-1A ( I B or I C ) if desired to the Station Ventilation Stack per Section 3.7 prior to refilling the Waste Gas Decay Tank. Note that this is only necessary if it is desired to fully depressurize the Waste Gas Tank.

Performed By Date Signature Reviewed By SRO or RO License Date Signature CAUTION All releases to the vent stack must be accomplished with an approved waste gas release permit.

3.7 Waste Gas Decay Disposal - LEVEL 2 3.7.1 Prerequisites A. Waste gas decay tank to be released is isolated IAW Section 3.12 B. VERIFY IA-V-1213 is OPEN and the instrument air system is available.

C. VERIFY the waste gas decay tank selected for disposal has had a hold time to ensure compliance with ODCM, Sec. 2.2.2. (SM/CRS)

D. ' NOTIFY the Radiological ControlslPlant Chemistry Dept. that the Operations Dept. wants to release a waste gas decay tank.

E. During release of gaseous waste from the waste gas decay tanks, the following conditions must be met:

a. Waste gas discharge monitor RM-A7 and Auxiliary and Fuel Handling Building exhaust gas, iodine and particulate monitor RM-A8 must be operable.

b. Waste gas decay tank discharge valve WDG-V-47 must be operable.

C. WDG-FR 123 must be operable.

21

Number TMI - Unit 1 u Operating Procedure I 104-27 Title Revision No.

Waste Disposal - Gaseous 68 F. SOURCE CHECK Waste gas discharge monitor RM-A-7 by Method 1 or 2: (N/A Method not used)

Method I:

1. Obtain key from Radiological Controls for RM-A-7 check source control box.

2. In Beckman analyzer room, unlock source control box. Open lead container and remove rod source.

3. Notify Control Room of intent to source check RM-A-7.

4. At RM-A-7 sampler, remove lead plug located on top of sampler and insert rod source.

5. Time and date charts for WDG-FR-123/RM-A-7 to indicate source check.

6. Confirm RM-A-7 ratemeter reading in Control Room qualitatively increases.

7. Remove rod source from sampler and replace lead plug in RM-A-7 sampler.

8. Place rod source in lead container and close. Close and lock source control box.

9. Return key to Radiological Controls.

22

Number TMI - Unit 1 I 11' Operating Procedure I 104-27 Title Revision No.

Waste Disposal Gaseous 68

1. Power-down RM-A-7 at ratemeter.

2. Remove lead cover from RM-A-7 (if necessary)

3. Remove detector from sampler.

4. Attach primary calibration source, #318A to the face of the detector, using the source cup holder.

5. Replace the detector in the sampler 6 Power-up RM-A-7 at ratemeter.

7. Confirm that RM-A-7 ratemeter reading increases significantly.

8. Power-down RM-A-7 at ratemeter.

9. Remove detector from sampler and remove calibration source from the face of the detector.

IO. Replace the detector in the sampler and replace all fasteners.

11. Replace lead cover.

12. Power-up RM-A-7 at ratemeter.

23

I I TMI - Unit 1

'U.

Operating Procedure 1104-27 G. VERIFY RM-A-7, WDG-FT-123 and WDG-V-47 are in service and that there are no Equipment Deficiency Tags (EDT's) on the I equipment.

H. VERIFY RM-A-7 interlock defeat switch is in "ENABLE".

I. VERIFY RM-A-8 interlock defeat switch is in "NORMAL" J. @ERIFY AUX/FH BLDG Ventilation is in service. ,

3.7.2 Procedure CAUTION All releases to the vent stack must be accomplished with an approved

/I waste gas release permit.

The disposal of waste gas is done by controlled venting of the Waste Gas Decay Tanks to the Auxiliary Building Stack.

1. The Chemistry Dept. has taken a sample of the tank that is to be released.

2. The release rate is determined from the results of the gas sample and a waste gas release permit per 6610-ADM-4250.11, that has been signed by Rad Con and the Shift ManagerKontrol Room Supervisor, and has been received by the Control Room Operator.

3. VERIFY the TEST/NORMAL switch on WDG-FR-123 is in the NORMAL position.

COMPLETE Waste Gas Release Data Sheet as required and periodically during the release.

4. HANG the "Gas Release in Progress" sign on L.W.D. panel in Auxiliary Building.

5. NOTIFY the Control Room to HANG the "Gas Release in Progress" sign on RMS Panel.

6. SET WDG-FR-123 alarm setpoint to the value which corresponds to the flow rate specified in the Waste Gas Release Permit by using the dial behind the recorder inside the Radwaste Panel.

EXAMPLE If the alarm setpoint per WG Permit is 8.5 cfm., a total of 8 112 complete revolutions will be required. An eight (8) will be showing in the window above the dial and the dial itself will be set at fifty (50).

7. VERIFY the proper WDG-FR-123 setpoint to the value corresponding to the Waste Gas Release Permit Form 1622-2.

24

Form ES401-6 SYSIEP# Gen KA# 2.3.3 Page t# 2-9 Tier # -3

'- 2.9 Group #

ROlSRO Importance Rating 1.8 -

Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems). (Radiation Control)

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

- Fuel pin failure occurred 3 days ago.

- RC Bleed Tank WDL-T-I B cleanup is in progress to reduce activity.

- Paperwork (P&ID and OPI 104-29A) to process the water specifies use of:

- Waste Transfer Pump WDL-P-GB.

- ' A Precoat Filter.

- 'A Cation Demineralizer.

- Leak develops at 'A Cation Demineralizer.

Based on conditions above, identify the ONE statement below that describes Unit Supervisor responsibility regarding operation of the Liquid Waste Disposal System.

A. Temporarily suspend the process until the leak is repaired, and then direct the operator to continue the process using the current paperwork.

B. Direct the operator to continue the process using 'B' Cation Demineralizer, and replace the paperwork after the process is completed.

C. Modify current paperwork to use 'B' Cation Demineralizer, and then direct the operator to continue cleanup of WDL-T-1B.

u D. Terminate the process, and annotate the paperwork, describing cause for pre-empting the process.

OP 1104-29, Liquid Waste Disposal System, section 4.3.1 , page 28, Rev. 81.

M New i7 TMIBank TMI Question #

0 Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge id Comprehensionor Analysis 55.43 .4 55.45 .IO A Incorrect answer. Does not conform to required actions in OP 1104-29, Liquid Waste Disposal System, section 4.3.1, page 28, Rev. 81.

B Incorrect answer. Does not conform to required actions in OP 1104-29, Liquid Waste Disposal System, section 4.3.1, page 28, Rev. 81.

C Incorrect answer. Does not conform to required actions in OP 1104-29, Liquid Waste Disposal System, section 4.3.1 page 28, Rev. 81.

D Correct answer None.

TMI SRO Exam May 2003 Friduy, March 28,2003

I -

TMI Unit 1 Number

'L/ Operating Procedure 1104-29 Title Revision No.

Liquid Waste Disposal System 81

2. Request sampling by Chem. Dept. of tank contents as required to establish the proper process to be used if the water chemistry might have a bearing on that evaluation.

Request Chem. Dept. recommendations on processing if their technical input is needed.

Also, indicate intended destination of tank contents so the appropriate analysis may be performed.

3. If a release of the Waste Evaporator Condensate Storage Tank is intended a release application must be initiated per 6610-ADM-4250.01.

4. The PROCESS INSTRUCTION AND DATA SHEET (See ENCLOSURE 2) must be completed by the Shift Manager/Control Room Supervisor or his designee in order to authorize and direct the operators to perform the process indicated on the form. The Control Room Supervisor indicates which LWDS procedure should be used, assigns a batch number, and enters the required information in computerized Radwaste Log.

5. The operator completes the process in accordance with the process instructions, appropriate release forms, and the applicable procedure.

6. The completed procedure with the PROCESS INSTRUCTION and DATA SHEET is returned to the Control Room Supervisor for his review.

7. Send any Data back to the Operations Department Office or the Manager, Radwaste &

Chemistry as required.

8. Request samples be taken by the Chemistry Department as appropriate.

4.3 Pre-emption of Processes 4.3.1 If a process must b ' ted due to equipment malfunction, interruption by other hould be shut down in accordance with the procedure being nd appropriate notations should be made to the PROCESS INSTRUCTION AND DATA SHEET and Procedure Checkoff Sheet as to the cause of the process termination.

4.3.2 A new PROCESS INSTRUCTION AND DATA SHEET should be initiated by the CRS if the conditions have changed which governed the original process. (e.g.. the chemistry of the source tank has significantly changed, a different process route is selected, etc.).

5.0 LWDS PLANNING AIDS AND ADMINISTRATIVE GUIDES 5.1 Estimation of RC Evaporator Feed Tank Boron Conc.

The desired concentration of boron in the RC Evaporator Feed Tank (bottoms) can be calculated (rather than chemical analysis) by using a mass balance equation given below. Evaporation is continued while periodically inserting the feed tank volume and volume processed into the equation until it balances. At that point the desired boron concentration will exist in the feed tank and the concentrates can then be transferred to the Reclaimed Boric Acid Storage Tanks.

28

SYS/EP# Gen KA# 2.3.1 Page # 2-9 Tier # -

3

'- ROlSRO Importance Rating 2.6 3.0

- Group #

Knowledge of I O CFR 20 and related facility radiation control requirements. (Radiation Control)

A male radiation worker at Three Mile Island (TMI) has just returned from 3 weeks of outage support at Peach Bottom.

- His Total Effective Dose Equivalent (TEDE) received at Peach Bottom was 150 mrem.

- After a fall at home, the worker had an ankle x-ray estimated at 10 mrem exposure to the ankle.

- This worker's current TEDE from TMI for this year is 75 mrem.

Based on these figures, choose the calculated MAXIMUM annual non-emergency TEDE that he can receive at TMI for the remainder of this year without exceeding the Federal Exposure Limits.

A. 4765mrem.

B. 4775mrem.

C. 4850mrem.

D. 4925mrem.

10 CFR 20.1201 None.

.02.01

- G New kd TMI Bank TMI Question # Peach Bottom 2001 SRO Modified TMI Bank Parent Question #

D Memory or Fundamental Knowledge Comprehensionor Analysis 55.41 .I2 ci 55.43 0 55.45 .9/.10 A Incorrect answer. Federal exposure limit is 5000 mrem. This answer erroneously includes the ankle x-ray in the TEDE exposure.

B Correct answer. Federal exposure limit is 5000 mrem - does not include the ankle x-ray exposure.

C Incorrect answer. Federal exposure limit is 5000 mrem. This answer erroneously does not include the Peach Bottom TEDE exposure.

D Incorrect answer. Federal exposure limit is 5000 mrem. This answer (erroneously) does not include the Peach Bottom TEDE exposure, and (erroneously) does include the ankle x-ray exposure.

None.

TMI SRO Exam May 2003 Friday, March 28,2003

10 CFR Subpart C -- Occupational Dose Limits Page 1 of 1

,**lrd.,

$6 ->

Index I Site Map I FAQ I Help I Glossary I Contact Us L ,

U.S.Nuclear Regulatory Commission Nuclear Nuclear Radioactive Public Home Reactors Materials Waste Involvement Home > Electronic Readina Room z Document Collections > NRC Requlations (10 CFR) > Part Index > 10 CFR Subpart C -

Dose Limits

§20.1201 Occupational dose limits for adults;?

(a) The licensee shall control the occupational dose to individual adults, except for planned special exposure 520,1206, t o the following dose limits.

(1)An annual limit, which is the more limiting of --

(i) The total effective dose equivalent being equal t o 5 rems (0.05 Sv); or (ii) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissu the lens of the eye being equal to 50 rems (0.5 Sv).

(2) The annual limits to the lens o f the eye, to the skin of the whole body, and t o the skin of the extremities (i) A lens dose equivalent of 15 rems (0.15 Sv), and (ii) A shallow-dose equivalent o f 50 rem (0.5 Sv) to the skin of the whole body or to the skin o f any extremit

' u (b) Doses received in excess of the annual limits, including doses received during accidents, emergencies, a special exposures, must be subtracted from the limits for planned special exposures that the individual may the current year (see §20.1206(e)(l)) and during the individual's lifetime (see §20.1206(e)(2)).

(c) The assigned deep-dose equivalent must be for the part o f the body receiving the highest exposure. The shallow-dose equivalent must be the dose averaged over the contiguous 10 square centimeters of skin rece highest exposure. The deep-dose equivalent, lens-dose equivalent, and shallow-dose equivalent may be ass surveys or other radiation measurements for the purpose of demonstrating compliance with the occupationa the individual monitoring device was not in the region of highest potential exposure, or the results of individ are unavailable.

(d) Derived air concentration (DAC) and annual limit on intake (AU) values are presented in table 1of appe 20 and may be used to determine the individual's dose (see tj20.2106) and to demonstrate compliance with occupational dose limits.

(e) I n addition to the annual dose limits, the licensee shall limit the soluble uranium intake by an individual t milligrams in a week in consideration of chemical toxicity (see footnote 3 of appendix B to part 20).

(f) The licensee shall reduce the dose that an individual may be allowed t o receive in the current year by the occupational dose received while employed by any other person (see §20.2104(e)).

[56 FR 23396, May 21,1991, as amended a t 60 FR 20185, Apr. 25, 19951 Privacv Statement I Site Disclaimer L

Last revised Tuesday, December 03, 2002 http ://~~~.nrc.gov/reading-rm/doc-collections/cfr/pa~O2O/pa~O20- 1201.litid 2/14/03

Form ES-4014

\ KA# 2.3.10 Page # 2-10 Tier #

d ROlSRO Importance Rating 2.9 3.3

- Group #

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. (Radiation Control)

Identify the ONE action below that is performed to reduce excessive levels of radiation and guard against personnel exposure during accident conditions in accordance with OP-TM-EOP-010, Abnormal Transients, Rules, Guides and Graphs.

A. After Subcooled Margin is restored, maintain natural circulation for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to starting a Reactor Coolant Pump per Guide 7, RCP Restart.

B. Continue RB Spray pump operation with RB pressure below 2 psig per Guide 18, Containment.

C. Maintain RB flood level below 64 inches per Guide 22, RB Sump Recirculation.

D. Verify RB flood level >32 inches prior to transferring to RB Sump reciculation per Guide 21, Transfer to RB Sump Recirculation.

OP-TM-EOP-010 Guide 18, Containment, Rev. 1.

None.

V.E.05.09

@lNew 0 TMI Bank TMI Question #

0 Modified TMI Bank Parent Question #

u 0 Memory or Fundamental Knowledge PI Comprehension or Analysis

@ 55.43 .4 a 55.45 .10 A Incorrect answer. The purpose of this action is to allow for adequate mixing of RCS water to prevent reactor restart when an RCP is restarted under specific conditions.

B Correct answer. RB pressure reduction reduces containment leakage, and allows time for iodine absorption by the spray fluid.

C Incorrect answer. The purpose of this action is to prevent flooding vital instrumentation that will be needed in the Accident Recovery Mode.

D Incorrect answer. The purpose of this action is to ensure adequate DH and BS pump NPSH when suction is transferred to the RB from the BWST..

None.

TMI SRO Exam May 2003 Friday, March 28, 2003

OP-TM-EOP-010 Revision 1 Page 36 of 49 Guide 18 Containment ACTlONlEXPECTED RESPONSE RESPONSE NOT OBTAINED VERIFY 4 psig ESAS has actuated. INITIATE OP-TM-534-901,

. . . _ ._ - .. . . . .. .. . . -. - RB Emergency Cooling.

DETERMINE energy source.

OBSERVE Radiation monitors and OTSG pressures and feed rates.

PRIMARY OTSGA OTSGB If source is a secondary side leak (either FW or MS), then PERFORM Phase 1 and Phase 2 Isolation of Rule 3 - XHT.

Shutdown of Building Spray Prerequisites:

RB pressure 2 psig.

ED concurrence that atmospheric iodine removal has been accomplished.

Procedure:

1.I Shutdown one train of RB Spray

1. Stop BS-P-lA(B).

2. Close BS-V-3A(B).

3. Close BS-V-lA(B).

1.2 Monitor RB Temperature and Pressure.

1.3 Shutdown the remaining train of RB Spray

1. Stop BS-P-1B(A).

2. Close BS-V-3B(A).

3. Close BS-V-1B(A).

Form ES-401-6 b' ROlSRO Importance Rating 2.4 2.8 Group #

Knowledge of the process used track inoperable alarms. (Emergency ProcedureslPlan) 0 rn Plant conditions:

- Reactor power is 1OO%, with ICs in full automatic.

- SR-P-1 B is tripped, SR-P-IC started automatically.

- Applicable procedure ImmediateActions are complete, and Follow-up Actions are in progress in response to SR-P-1 B trip.

- Control Room Alarm PLB 6-8, River,Rakeor Screen Trouble, is repeatedly actuatinglresetting, interrupting control room communications.

- There is no valid reason for the alarms, according to the Auxiliary Operator's field report.

- You, the Unit Supervisor, want to authorize the alarm to be removed from service, in order to eliminate the distraction.

- No alarm repair is expected during this shift.

- The nuisance alarm has been disabled.

Based on conditions above, identify the ONE statement below that describes how to track the inoperable alarm per OP-AA-103-102, Watchstanding Practices.

A. Initiate an Equipment Status Tag (EST), and direct the crew to denote the inoperable alarm in turnover documents.

B. Apply an Equipment Deficiency Tag (EDT), and document the condition in a Control Room Log entry.

- C. Apply an Instrument Out of Service (00s) documents.

tag, and direct the crew to denote the inoperable alarm in turnover D. Apply an Information Tag, and document the condition in a Control Room Log entry.

OP-AA-103-102, Watchstanding Practices, pages 5 and 6, Rev. 1.

PI New 0 TMIBank TMI Question #

0 Modified TMI Bank Parent Question #

@ Memory or Fundamental Knowledge 0 Comprehension or Analysis EZ 55.41 . I O 0 55.43 .3 &? 55.45 . I 3 A Correct answer. Conforms to OP-AA-103-102.

B Incorrect answer. Plausible distracter since this could be misconceived as an equipment deficiency.

Removal of the Panalarm horn fuse assembly would silence the annunciator for all alarms on that panel, but would not affect light operation.

C Incorrect answer. This answer describes the former AmerGen process used to track this type of problem.

D Incorrect answer. This is not associated with the BETA alarm system. EDTs are not used to track nuisance alarms.

None.

TMI SRO Exam - May 2003 Friday, March 28,2003

Form ES-401-6 Q # 097 TMI SRO Exam May 2003 Friday, March 28,2003

OP-AA-I 03-1 02 Revision 1 Page 5 of 6 4.3.5. Certain procedures and surveillances cause large numbers of Main Control Room annunciators/alarms to come in. Examples of this are annunciator cabinet surveillances or PWR RPS testing. In these cases it is not practical or expected to announce each alarm and the Unit Supervisor may suspend the normal annunciator/alarm response. Exceptions for expected alarms are as follows:

I. Many plant activities are known to cause alarms. When an activity is planned or in progress that will bring in an alarm/annunciator, the Reactor Operator or Radwaste Operator should be notified ahead of time.

2. The operator should review the Annunciator Response Procedure/Alarm Response Card associated with the planned alarm(s) either before or after the alarm is received the first time the alarm comes in on a shift for applicability.

3. The Reactor Operator should inform the Unit Supervisor of the planned alarms prior to their actuation.

- Alarms caused by rounds or local panel annunciator tests do not require annunciator procedure review.

- Flagging of expected alarms using colored dots or stickers is recommended to denote the fact that they have been discussed ahead of time, especially for alarms that are expected to be repetitive. (re:

HU-AA-101, Human Performance Tools and Verification Practices)

- Expected alarms need not be announced in detail provided they have been discussed with the Unit Supervisor. Expected Alarm is an acceptable announcement for these occurrences.

4. Nuisance Alarms.

A. Nuisance alarms are an operator distraction. Efforts should be made to eliminate the source of the alarm signalhnput by operational or maintenance actions (e.9. the T-mod Process).

1. Annotate alarm windows disabled in this manner.

B. If operational or maintenance corrective actions are ineffective, utilization of the annunciator mode the duration of the nuis input or g the nuisance input, under the following conditions:

1. Obtain Unit Supervisor permission.

2. If the annunciator mode switch will be re-positioned or the alarm disabled for more than one shift, ihitiate an Equipment Status Tracking (EST) tag to track the abnormal condition or annunciator mode switch position. (Note: use of EST is not u

required if procedure directed.)

OP-AA-103-102 Revision 1 Page 6 of 6

3. Disabled alarms shall be noted within the appropriate turnovers. ei-

5. Local Alarm Response.

A. Take action in accordance with the appropriate Annunciator Response Procedure/Alarm Response Card and direction as required from the Main Control Room.

B. The operator who acknowledged the local alarm shall notify the applicable Main Control Room Operator which local alarm actuated and the response action taken.

4.4. Control Room Personnel Administrative Duties.

4.4.1. Limit on-shift administrative tasks that could compromise the operators primary shift responsibilities.

1. Shift supervisory personnel remain in a supervisory role:

A. Directing plant operations.

B. Supervising control board operations.

\

L C. Providing crew oversight and coaching.

5. DOCUMENTATION - None

6. REFERENCES 6.1. INPO 01-002, Guidelines for the Conduct of Operations at Nuclear Power Stations 6.2. SOER 96-01.

7. ATTACHMENTS - None

Q # 098 Page # 4.2-8 Tier #

b RO/SRO Importance Rating 4.1 4.6 Group # -

2 Knowledge of the reasons for the following responses as they apply to Pressurizer Vapor Space Accident (Relief Valve Stuck Open): Actions contained in EOP for PZR vapor space accident/LOCA Plant conditions:

- Reactor power is 1OO%, with ICs in full automatic.

- A Pressurizer Safety Valve leak has been diagnosed.

In accordance with the applicable procedure, the Unit Supervisor requests chemistry samples for RCS and Pressurizer boron concentrations.

The basis for this chemistry sampling is to ensure that the Pressurizer boron concentration does NOT become significantly (1) than the RCS boron concentration in order to prevent a reactivity excursion if an unexpected plant (2) would occur.

A. (1) lower (2) heatup B. (1) lower (2) cooldown C. (1) higher (2) heatup D. ( I ) higher

.u (2) cooldown PM Section B-01, Reactor Coolant System, page 17, Rev. 13.

None.

3 New TMI Bank TMI Question ## #26 6/2001 SRO Audit Ci Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge Ed Comprehension or Analysis M 55.41 .5/.10 rl 55.43 3 55.45 .6/.13 A Incorrect - Plausible with misconceptionof the pressurizer response to a heatup, however pressurizer boron will be increasing.

B Incorrect - Reactivity would be positive on a pressurizer outsurge with a lower pressurizer boron concentration, however pressurizer boron will be increasing.

C Incorrect - right boron change, wrong effects.

D Correct answer.- right boron change and effects.

Modified plausibility statements and stem format, reordered answers, changed distractors, but u does not meet NRC modification screening, changed distractors TMI SRO Exam - May 2003 Friday, March 28,2003

SECTION B-0 1 REVISION 13 with a throttle valve (RC-V-24) is provided around the spray valy@. This spray ded to prevent thermal shock to the spray valve and also because of the small bypass spray flow it means flow of RCS liquid will always be coming out the pressurizer surge line. This rge flow will minimize the buildup of boron pressurizer. Normal bypass spray flow-is .5 to 1.5 gpm.

6.6 Reactor Coolant Piping The major piping components of the system are four 28 inch inside diameter cold legs connecting the OTSGs lower head to the RCP inlet and outlet to the reactor vessel, and the two 36 inch inside diameter hot legs connecting the reactor vessel to the upper OTSG head. These sections of piping are carbon steel clad with austenitic stainless steel.

Short sections of stainless steel piping are provided as transition pieces from the RCP casings to the cold leg piping.

The pressurizer surge line is a l0inch outside diameter pipe made completely from austenitic stainless steel as is the 2 1/2 inch outside diameter spray line.

Thermal sleeves are installed where needed to limit the thermal stresses due to rapid changes in fluid temperature. The thermal sleeves are located at the four high pressure injection nozzles on the reactor inlet piping, the two core floodlow pressure injection nozzles on the reactor vessel and on the pressurizer surge and spray lines.

6.7 Reactor Coolant Pumps L-Each reactor coolant loop contains two vertically mounted single stage centrifugal pumps, which employ a controlled leakage seal assembly. A more detailed description of the Reactor Coolant Pumps will be in Section B-2 Reactor Coolant Pumps/Motors of this manual.

6.8 Pressurizer Code Safety Valves Two pressurizer code safety valves (RC-RV-1A and RC-RV-1B) are mounted on individual nozzles on the top head of the pressurizer. The valves have a closed bonnet with bellows and supplementary balancing piston. The valve inlet and outlet is flanged to facilitate removal for maintenance. Each code safety valve is rated at 297,846 Ibm/hr at 2575 psig (setpoint +3%

accumulation).

6.9 Pilot Operated Electromatic Relief Valve (PORV)

The PORV (RC-RV-2) is mounted on a separate nozzle on the top of the pressurizer. The main valve operation is controlled by the opening of a solenoid operated pilot valve. When the pilot valve is opened an unbalanced force acts on the main valve disc causing it to open. The solenoid for the pilot valve is powered from DC Distribution Panel 1C. The PORV is rated for 113,350 l b d h r at 2450 psig.

A two position Auto/Open maintained contact switch allows the operator to manually open the valve to reduce RCS pressure. A protective cover is installed over the switch to prevent inadvertent opening such as when spraying the pressurizer in manual since this similar switch is in close proximity.

17

Page # 2-12 Tier # -

3 u RO/SRO Importance Rating 3.4 3.6

- Group #

Knowledge of abnormal condition procedures.(Emergency Procedures/Plan)

Initial conditions:

- Reactor power is 1OO%, with ICs in full automatic.

Sequence of events:

- Reactor power is rising at O.l%/minute.

- RCS pressure has lowered to 2130 psig, and is now steady.

- Average RCS temperature has lowered to 578.5"F, and is now steady.

- MU-V-17 is opening to maintain pressurizer level at 220 inches.

- Generator load is 900 MW and lowering at 1 MW every 2 minutes.

- Alarm B-2-8, RB Coolers Excess Condensate has actuated.

- RB pressure is 0.1 psig and rising at a rate of 0.1 psig every 5 minutes.

Identify the ONE statement below that describes required response to these conditions.

A. Trip the reactor and go to OP-TM-EOP-001.

B. Reduce power to ~ 4 5 % and trip the turbine.

C. Commence a one hour RCS leak rate calculation.

D. Commence plant shutdown at the rate specified by the Unit Supervisor.

AbP 1203-24, Steam Leak, pages 2 and 3, Rev 29.

-' None.

V.C.10.03 New E TMI Bank TMI Question # #98June2000 RO Make-up CJ Modified TMI Bank Parent Question #

0 Memory or Fundamental Knowledge

@ I Comprehension or Analysis

@ 55.41 .IO 55.43 51.13 c7 55.45 A Incorrect answer. Plausible since this is an IMA # I , determined by leak location and equipment hazards.

B Plausible since this is IMA #2, determined by leak location.

C Plausible since this would address the RCS conditions.

D Correct answer. Refer to AbP 1203-24, Steam Leak, immediate manual action #3.

TMI SRO Exam - May 2003 Friday, March 28,2003

Number TMI - Unit 1 L-J Abnormal Procedure 1203-24 Title Revision No.

Steam Leak 29

1. Decreasing secondary steam pressure.

2. Electrical load reducing (mismatch between electrical load and Rx Power).

3. Decrease in pressurizer level, R.C. Pressure, and cold leg temperature.

4. For a leak inside the Reactor Building; Indication of increasing Reactor Building pressure and tern perature.

5. For a leak outside the Reactor Building; Noise may be heard in Control Room or a report made from personnel outside the Control Room.

2.0 IMMEDIATE ACTION A. Automatic Action

1. If HSPS MFW Isolation actuates ( ~ 6 0 0psig) on the affected OTSG (could be both), the following valves auto close.

a. Startup Feedwater Control Valve FW-V-IGA(B)

b. Main Feedwater Control Valve FW-V-I-/A(B)

C. Main Feedwater Block Valve FW-V-SA(B)

d. Startup Feedwater Block Valve FW-V-92A(B)

2. Possible Reactor trip on low pressure.

B. Immediate Manual Action

1. If the steam leak is upstream of the turbine stop valves (or the leak location is unknown) and either:

a. HSPS actuates on either SG

b. Continued operation presents a hazard to personnel or equipment required for safe shutdown.

Then manually trip the reactor and go to OP-TM-EOP-001.

2. If the steam leak is downstream of the turbine stop valves and time permits.

Then reduce power to c 45 percent and trip the turbine IAW 1102-4.

2

Number TMI - Unit 1 L-J Abnormal Procedure 1203-24 Title Revision No.

Steam Leak 29

3. If a. Continued operation is not posing a hazard to personnel or equipment required for safe shutdown but is severe enough to require shutdown

b. RB pressure exceeds 2 psig.

Then reduce load at rate specified by US and go to following steps.

3.0 FOLLOW-UP ACTION The objective of this procedure is to continue to shutdown and cooldown the unit while monitoring the steam leak.

1. Continue to shutdown and cooldown the unit per OP 1102-10 and 11 respectively.

2. Determine which OTSG has the steam leak and if possible the location of the leak. If possible, the leak should be isolated.

NOTE If at anytime during the unit shutdown the steam leak degrades to where personnel or plant equipment required for safe shutdown are threatened

- 3.

then trip the reactor and go to OP-TM-EOP-001.

Monitor RB Pressure and Temperature if leak is in the RB.

4. If the steam leak is in the RB, start containment cooling to keep RB Pressure < 2.0 PSlG and RB temperature c 130°F (if necessary). Initiate OP-TM-534-901.

5. If steam leak is in RB, drain RB sump as needed to prevent overflowing the sump. (Containment Isolation Valve opening criteria must be met).

-\-

3

SYS/EP# @ J KA# 2.4.10 Page # 2-12 Tier # -3 u 3.1 Group #

RO/SRO Importance Rating 3.0 -

Knowledge of annunciator response procedures.(Emergency Procedures/Plan).

Plant conditions:

- Reactor power is loo%, with ICs in full automatic.

Event:

- Value for Process Computer point TA168, Stator Bar Water Out TC16, unexpectedly exceeds the high alarm limit.

Based on this condition, identify the ONE statement below that describes where the primary alarm response guidance for this alarm is identified.

A. OP 1106-1, Turbine Generator, Section 3.7 (Emergency Trip).

B. OP-TM-MAP-LO307, Gen. Stator Liquid Clg Trouble.

C. OP 1105-10, Plant Computer Operations, Enclosure 5, Plant Computer Alarm.

D. OP 1105-1OA, Plant Computer Alarm Attributes, "Reference Document" column.

OP-TM-103-102-1002 Additional Alarm Response Expectations, Section 4.2.2, page 2, Rev. 0.

OP 1105-1OA, Plant Computer Alarm Attributes, page E2-94, Rev. 48.

OP 1106-1, Turbine Generator, pages 118 and 119, Rev. 109.

None.

W 2l New E TMlBank TMI Question #

G Modified TMI Bank Parent Question #

@I Memory or Fundamental Knowledge 0 Comprehension or Analysis (3) 55.41 . I O 0 55.43 id 55.45 .I3 A Incorrect answer. Although this alarm is associated with the turbine generator, the limited information given does not by itself require following the emergency turbine trip guidance.

B Incorrect answer. This is the general alarm response procedure for MAP L-3-7. This document merely directs the operator to look at the Stator Cooling local alarm panel alarm response procedure.

C Incorrect answer. OP 1105-10 Enclosure 5 is a specific alarm response for PPC system failure.

D Correct answer. OP-TM-103-102-1002, Additional Alarm Response Expectations, section 4.2.2.2 identifies this as the primary PPC alarm guidance to be used.

TMI SRO Exam - May 2003 Friday, March 28,2003

0P-TM-q 03-1 02-1002 Revision 0 Page 2 of 3 TMI Training and Reference Material 4.2. Response to Process Computer Alarms 4.2.1. Expected alarm response shall b e as described in OP-AA-I 03-1 02.

4.2.2. Unexpected alarm response;;shall b e as described in OP-AA-103-102 with the exception that:

1, The primary alarm response guidance to b e used is listed in the "Reference Document" column in 1105-1OA Plant Process Computer Attributes

2. T h e secondary alarm response guidance to b e used in found in t h e "Alarm Basis" section of I 105-1OA Plant Process Computer Attributes and the applicable system operating procedure.

4.3. Nuisance Alarms 4.3.1. In addition to the requirements of OP-AA-103-102 section 4.3.5.4.B, the directions for how to disable a specific alarm function is found in the following procedures:

1. Main Annunciator Panel alarmdisabled -1 105-21

'U 2. Plant Process Computer alarm disabled, setpoint change or point deleted -

I 105-1 0

3. DTCS alarm disabled - 1106-1 4.3.2. An EDT or EST (depending on whether it is a equipment problem o r administrative issue) sticker o r tag is applied to the affected window.

4.3.3. For the PPC or DTCS alarms, the log is used to identify t h e abnormal condition and no sticker or tag is applied.

1. The EDT or EST process is applied (except for the sticker or tag not being applied) and the EST log or AR entry (for EDT) is completed.

5. DOCUMENTATION 5.1. None

6. REFERENCES

1. 1036 Instrument Out of Service

<> I

2. 1 105-10 Plant Computer Operations 6.3. 1 105-1OA Plant Process Computer Attributes

OP11OS-lOA Rcvisioll43 r 01:ra POIN'T SC.4N ALM NLM JYtl DESCHIP1'OR IPi-"I SOTIIICE UTE COND 1'A 16C S'1'.4'TOR BAR WATER O7,T TC 8 DEW TG'lT-6008 I20 ON 2 HI72 HI282 000211 3.0 1,2387 TRBI, OP1106-1 TAlGI S1.4rOR BAR \'.4'l'LR OLT TC 9 1IECK TG-TE-0003 I20 ON 2 HI72 HI^ 82 0002u 3.0 L2987 Ttil3L OPllGGl TA162 STATOR Bfi3 \V.YKR OUT TC 10 DEW TG-TE-0010 I20 ON 2 HI72 HI282 OOK1J 3.0 1,2989 TF.BL OPllClG-1 TA 163 S1'ArOOR BAR W.4TER OUT TC 1 1 DECK TG-lTAOll 120 ON 2 H1 72 I112 82 0002U 3.0 L2987 TRBI, OPlI06-1

'l'AlG4 ST.4'TOR DAR W.4TER OUT 'IC 12 DECK TG-TG0012 I20 ON 2 I31 72 III? 82 0002U 3.0 L2969 TRBL OPllOGl TA165 ST.4TOR BAR W.4TEX OUT I C 13 DEGC TG-TS0013 I20 ON 2 HI72 HI282 0002U 3.0 L2989 TRBL OPlIOG-1 TAi 66 STATOR BjJ? W.4TER OUT TC 14 DEGC TO-lE001.1 I20 ON 2 H I 72 1112 82 0002.U 3.0 1,2983 TRBL OPll06-1 TA167 ST.4TOR BAR WATER OUT 're 15 DEGC TG-'KO015 I20 ON 2 H I 72 1112 82 0002U 3.0 L2989 TRBL OP1106-1

'rp16% 'v VATOR BAR W.4TER OUT TC 16

DEN TG-TE-0016 120 ON 2 H I 72 I112 82 0002U 3.0 I2989 TRUL Th 163 ST.4TOR BAR VI'.%TTI',K OUT I C 17 DEGC TG-TL0017 I20 ON 2 HI72HI282 0002u 3.0 L2389 TkBL OP11OG-1 TA 170 ST-4TOR BAR WATER OUT TC 18 DEGC TG-TZ-0018 120 ON 2 H I 72 HI2 82 0002U 3.0 1,2387 TRBL OP1lOG.l 1'A171 ST.4TTJK BPJ? W.4TEA OUT TC 19 DEGC TC3-TGOO 19 I20 ON 2 HI72 HI282 0002U 3.0 L2989 TRBL OPllO6-1 7'4.1 72 s-r.KroR BAR 148.4~13 OUT rc 20 DEGC TG-TG0020 120 ON 2 11172 HI2 82 0002u 3.0 L2989 TkBL OP1106-1 TA 1 ?3 STATOR BAR WATER OUT TC 2 I DEGC TG-'E-0021 I20 ON 2 1i172 Ill2 82 0002U 3.0 L29E9 TRBL, OPllO6-1 TA 174 STATOfI BAR W.4TI;R OUT TC 22 DEGC TG-TL0022 120 ON 2 HI72 HI282 0002U 3.0 L2969 1'WL OP1106-I TA175 STATOR BAX W.4'lW OUT rC 23 DEW 'I'G-'Il--0023 I20 ON 2 111 72 I112 82 000211 3.0 L2983 TRBL OPllOGl TA17B STArOR BAR W.4TLR OUT I C 24 DEW TG-'E0024 I20 ON 2 HI 72 1112 82 0002U 3.0 L29S9 TGBL OPll0Gl TA175 ST.4TOR BAR W.4TER OUT TC 25 DECK TG-'I%OO?S I20 ON 2 111 72 1112 82 0002IJ 3.0 1,2989 TRRL 01'1 106-1 TA 1 78 ST.4TOR BARVIf,4'~1iROUT'IC 26 DEW TC;-*E-00?6 I20 ON 2 HI72 HI282 0002U 3.0 L2983 TP.BL OP11061

'~~179 s*r.vrmBAR WATER OUT TC 27 DEW T G - T I ~ ~ ~ I20 ON 2 111 72 MI2 82 000211 3.0 L2989 nux OPll06-1 TA180 ST.4TOR BAX W.4TE.R OUT I C 28 DEE 1'G-TJ>0028 I20 ON 2 1 1 1 7 2 HI282 0002IJ 3.0 13783 TEDL OP1106-1 TA181 s-r.mxBAR W . ~ E , ROUT I C 29 DEW TG-'CF,6029 I20 ON 2 111 72 I112 82 0002U 3.0 L2989 TRUL OPllOG-1 TA 182 ST.4TOR BAR W.4'IXR O U l 'IC 30 DEOC TG-7'E-0030 120 ON 2 MI72 HI2 82 0002U 3.0 L2983 TmL OYlI0G.1 TA 183 ST.4TOR BAR N'ATEI! OUT TC 3 1 DEW TO-TE-0031 I20 ON 2 111 72 HI2 82 0002U 3.0 L2967 'I'R131, UP1 106-1 TA 1 84 s'r.4'r'oK BAR WATER OUT rc j 2 DECK TG-TL0032 I20 ON 2 131 72 MI2 82 0002u 3.0 L2987 TRBL or1106-1 "A185 s-r.<roK rm w.xri;x OUTW 33 DEK TG'L-Z-~JO~~ I20 ON 2 131 72 HI2 82 0002u 3.0 L29E9 'I'HBL 01'1 106-1 E2- 94

Number TMI - Unit 1

1106-1 %

Revision No.

109 Page 1 of 2 ALARM Main Generator Stator Temperature High SETPOINT L3058, L3059 STATOR BAR RTD LIMIT EXCEEDED (either LO-K or HI-K)

L3060, L3061 STATOR BAR TC LIMIT EXCEEDED (either LO-K or HI-K) o An individual stator bar temperature indication is hotter than the expected temperature for the present plant load. The alarm setpoints vary with cooling water temperature and generator load. The alarm setpoints are calculated by PPC based on GSC supply temperature (A01091) and Generator output current (A1046).

L3062 STATOR BAR TC DEVIATION a Alarm if any thermocouple reading deviates from the average by more than f 8°C L3063 STATOR BAR RTD DEVIATION a Alarm if any RTD reading deviates from the average by more than +13"C or -12°C.

TA081 -TA152 STATOR BAR (#1 through #72) RTD TEMPERATURE (OC) o Hi alarm 72OC Hi-2 alarm 82°C

~ T A 1 5 3- TA224 STATOR BAR (#1 through #72) WATER OUTLET TC ("6) a Hi alarm 72OC 4 Hi-2 alarm 82°C a) Flow restriction in an individual stator bar (i.e. inadequate flow through an individual stator bar) b) Invalid instrument indication c) Cooling water supply flow inadequate (pump failure, etc.)

d) High temperature of GS cooling supply AUTOMATIC ACTIONS If GSC return temperature (A0192) exceeds 80°C, then DTCS will initiate a runback (MAP L-1-7)

OBSERVATION Main Generator Output Current (A1046)

Generator Stator Cooling Supply and Return Temperatures (PPC A0191 and A0192)

Individual stator bar temperature and outlet water temperatures (TA081 to TA224). PPC Area 15 Groups 25 to 33 118

Number TMI - Unit 1

'L, Operating Procedure I 106-1 Tide Revision No Turbine Generator I09 APPENDIX F Page 2 of 2 MANUAL ACTION REQUIRED

1) If GSC supply temperature is high (Alarms GSC-2-2 or A01 91) or approaching alarm limits, then VERIFY actions are taken per GSC-2-2 alarm response

2) If L3058 through L3063 alarms or a individual Hi-1 alarm on a stator bar RTD or water outlet TC, then R REVIEW all of the RTD or TC indications (Area 15 Groups 25 to 33)

R DETERMINE the stator bar with highest reading associated with the alarm received o INITIATE an AR to measure the RTD or TC circuit resistance and check the individual temperature instrument localIy .

3) If two or more individual stator bar RTD or stator bar outlet TC temperatures > 82OC, then REDUCE generator load IAW 1102-4 until all stator bar RTD and stator bar outlet TC temperatures < 82 OC

..I 119

ML031710512

Contents

Text

3.0 INTRODUCTION

4.4.2 Basis

3.3 ENSURE

3.6 FEEDWATER

3.7.1 Prerequisites

2.0 DESCRIPTION

3.0 REFERENCES

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ML031710512

Text

3.0 INTRODUCTION

4.4.2 Basis

3.3 ENSURE

3.6 FEEDWATER

3.7.1 Prerequisites

2.0 DESCRIPTION

3.0 REFERENCES

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